ML021280606
| ML021280606 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 05/02/2002 |
| From: | Jernigan D Florida Power & Light Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-2002-082, TAC MB0554 WCAP-15446, Rev 1 | |
| Download: ML021280606 (123) | |
Text
0 FPL Florida Power & Light Company, 6501 South Ocean Drive, Jensen Beach, FL 34957 May 2, 2002 L-2002-082 10 CFR 50.4 10 CFR 50.60(a)
U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 RE:
St. Lucie Unit 1 Docket No. 50-335 Reactor Vessel Surveillance Capsule Report of Test Results - Revision 1 (TAC MB0554)
Florida Power & Light Company (FPL) is submitting revision I of Westinghouse report titled Analysis of Capsule 2840 From the Florida Power and Light Company St. Lucie Unit I Reactor Vessel Radiation Surveillance Program, WCAP-15446, dated January 2002. The enclosed is a revised summary technical report for the capsule removed during the Fall 1999 St. Lucie Unit 1 refueling outage (SL1-16).
Revision 1 of the Westinghouse report (WCAP-15446) was issued to correct a non-conservative error in the fluence analysis. Reevaluation with the new fluence data verified the conclusions reported in revision 0 as correct.
No changes in any conclusions resulted from this revision.
The original report was submitted pursuant to the requirements of 10 CFR 50.60(a) and 10 CFR 50 Appendix H, Reactor Vessel Material Surveillance Program Requirements, paragraph I.A, by FPL letter L-2000-193 on September 27, 2000.
The report includes the data required by ASTM E-185, as specified in paragraph 1II.B.1 of 10 CFR 50 Appendix H, and the results of all applicable fracture toughness tests conducted on the beltline materials in the irradiated and unirradiated conditions. FPL has confirmed no technical specification or operating procedure changes are required due to the revision.
This letter does not contain any regulatory commitments. If there are any questions ab t is submittal, please contact George Madden at 772-467-7155.
iiery uly yours, o 0nal E. Je ri Vice P r r i\\1" St. Lucie Plant Enclosure DEJ/GRM an FPL Group company
Westinghouse Non-Proprietary Class 3 ANALYSIS OF CAPSULE 2840 FROM THE FLORIDA POWER AND LIGHT COMPANY ST.
LUCIE UNIT 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM Westinghouse Energy Systems LLC
WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-15446, Revision 1 Analysis of Capsule 284' from the Florida Power & Light Company St. Lucie Unit 1 Reactor Vessel Radiation Surveillance Program T. J. Laubham D.M. Chapman J. Conermann January 2002 Prepared by the Westinghouse Electric Company for the Florida Power & Light Company Approved:
C. H. Boyd, Manager Equipment & Materials Technology WESTINGHOUSE ELECTRIC COMPANY LLC Nuclear Services Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355
© 2002 Westinghouse Electric Company All Rights Reserved
TABLE OF CONTENTS L IS T O F T A B L E S................................................................................................................................................
H L IS T O F F IG U R E S............................................................................................................................................
iv PREFACE & RECORD OF REVISION.......................................................................................................
vi EXECUTIVE
SUMMARY
(OR) ABSTRACT...........................................................................................
vii I
SU M M A R Y O F R E SU LT S................................................................................................................
1-1 2
IN T R O D U C T IO N...............................................................................................................................
2 -1 3
B A C K G R O U N D...............................................................................................................................
3-1 4
D ESC R IPTIO N O F PR O G R A M........................................................................................................
4-1 5
TESTING OF SPECIMENS FROM CAPSULE 2840......................................................................
5-1 5.1 O V E R V IE W...........................................................................................................................
5 -1 5.2 CHARPY V-NOTCH IMPACT TEST RESULTS................................................................
5-3 5.3 TEN SILE TE ST R ESU LTS...................................................................................................
5-4 6
RADIATION ANALYSIS AND NEUTRON DOSIMETRY.............................................................
6-1 6.1 IN T R O D U C T IO N 6-1 6.2 DISCRETE ORDINATES ANALYSIS.................................................................................
6-2 6.3 N EU TR O N D O SIM ETRY...................................................................................................
6-4 6.4 PROJECTIONS OF REACTOR VESSEL EXPOSURE...................................................
6-13 7
SURVEILLANCE CAPSULE REMOVAL SCHEDULE..................................................................
7-1 8
R E F E R E N C E S....................................................................................................................................
8-1 APPENDIX A LOAD-TIME RECORDS FOR CHARPY SPECIMEN TESTS APPENDIX B CHARPY V-NOTCH PLOTS FOR EACH CAPSULE USING HYPERBOLIC TAGENT CURVE-FITTING METHOD APPENDIX C CHARPY V-NOTCH SHIFT RESULTS FOR EACH CAPSULE HAND-FIT VS.
HYPERBOLIC TANGENT CURVE-FITTING METHOD (CVGRAPH, VERSION 4.1)
APPENDIX D ST. LUCIE UNIT 1 SURVEILLANCE PROGRAM CREDIBILITY ANALYSIS Analysis of St. Lucie Unit I Capsule 284'
LIST OF TABLES Table 5-1 Charpy V-Notch Data for the St. Lucie Unit 1 Lower Shell Plate C-8-2 Irradiated to a Fluence of 1.45 x 1019 n/cm 2 (E > 1.0 MeV) (Longitudinal Orientation)................. 5-6 Table 5-2 Charpy V-notch Data for the St. Lucie Unit 1 Lower Shell Plate C-8-2 Irradiated to a Fluence of 1.45 x 1019 n/cm2 (E > 1.0 MeV) (Transverse Orientation)...................
5-7 Table 5-3 Charpy V-notch Data for the St. Lucie Unit 1 Surveillance Weld Metal Irradiated to a Fluence of 1.45 x 10'9 n/cm 2 (E > 1.0 M eV).............................................................
5-8 Table 5-4 Charpy V-notch Data for the St. Lucie Unit 1 Heat Affected Zone (HAZ) Metal Irradiated to a Fluence of 1.45 x 1019 n/cm2 (E > 1.0 MeV)............................................
5-9 Table 5-5 Instrumented Charpy Impact Test Results for the St. Lucie Unit 1 Lower Shell Plate C-8-2 Irradiated to a Fluence of 1.45 x 1019 n/cm 2 (E > 1.0 MeV)
(Longitudinal O rientation)................................................................................................
5-10 Table 5-6 Instrumented Charpy Impact Test Results for the St. Lucie Unit 1 Lower Shell Plate C-8-2 Irradiated to a Fluence of 1.45 x 1019 n/cm 2 (E > 1.0 MeV)
(Transverse O rientation)...................................................................................................
5-11 Table 5-7 Instrumented Charpy Impact Test Results for the St. Lucie Unit 1 Surveillance Weld Metal Irradiated to a Fluence of 1.45 x 1019 n/cm 2 (E > 1.0 MeV)........................ 5-12 Table 5-8 Instrumented Charpy Impact Test Results for the St. Lucie Unit 1 Representative Heat Affected Zone (HAZ) Material Irradiated to a Fluence of 1.45 x 1019 n/cm2 (E > 1.0 M eV )..........................................................................................................
.. 5-13 Table 5-9 Effect of Irradiation to 1.45 x 1019 n/cm 2 (E > 1.0 MeV) on the Notch Toughness Properties of the St. Lucie Unit 1 Reactor Vessel Surveillance Materials....................... 5-14 Table 5-10 Comparison of the St. Lucie Unit 1 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, R evision 2, Predictions............................................................................................
5-15 Table 5-11 Tensile Specimens From Lower Shell Course Plate C-8-2, Weld, and Heat Affected Z on e M aterial....................................................................................................................
5 -16 Table 6-1 Calculated Fast Neutron Exposure Rates and Iron Atom Displacement Rates at the Surveillance C apsule C enter.............................................................................................
6-16 Table 6-2 Calculated Azimuthal Variation of Fast Neutron Exposure Rates and Iron Atom Placement Rates at the Reactor Vessel Clad/Base Metal Interface..................................
6-17 Table 6-3 Relative Radial Distribution of 4 (E > 1.0 MeV) Within the Reactor Vessel Wall......... 6-18 Table 6-4 Relative Radial Distribution of 0 (E > 0.1 MeV) Within the Reactor Vessel Wall......... 6-19 Analysis of St. Lucie Unit I Capsule 2840
iii LIST OF TABLES (CONTINUED)
Table 6-5 Relative Radial Distribution of dpa/sec Within the Reactor Vessel Wall........................ 6-20 Table 6-6 Nuclear Parameters Used in the Evaluation of Neutron Sensors.....................................
6-21 Table 6-7 Monthly Thermal Generation During The First Fifteen Fuel Cycles of the St. Lucie Unit 1 Reactor (Reactor Power of 2700 MWt).................................................
6-22 Table 6-8 Measured Sensor Activities and Reaction Rates Surveillance C apsule 2840...........................................................................
6-25 Surveillance C apsule 1040........................................................................... 6-26 Surveillance C apsule 970..............................................................................
6-27 Table 6-9 Summary of Neutron Dosimetry Results Surveillance Capsules 970, 104', and 2840.... 6-28 Table 6-10 Comparison of Measured, Calculated, and Best Estimate Reaction Rates at the Surveillance C apsule C enter...........................................................................................
6-29 Table 6-11 Best Estimate Neutron Energy Spectrum at the Center of Surveillance Capsule C ap su le 2 840................................................................................................ 6 -3 0 C ap sule 10 40................................................................................................ 6 -3 1 C ap su le 9 70..................................................................................................
6 -3 2 Table 6-12 Comparison of Calculated and Best Estimate Integrated Neutron Exposure of the 970, 1040, and 2840 Surveillance Capsules...........................................................
6-33 Table 6-13 Azimuthal Variations of the Neutron Exposure Projections on the Reactor Vessel Clad/Base M etal Interface at Core M idplane...................................................................
6-34 Table 6-14 Neutron Exposure Values Within The St. Lucie Unit 1 Reactor Vessel.......................... 6-36 Table 6-15 Updated Lead Factors for St. Lucie Unit 1 Surveillance Capsules..................................
6-40 Table 6-16 Cj Values for the St. Lucie Unit 1 Sensor Reaction Rate Evaluation.................................
6-41 Table 7-1 St. Lucie Unit 1 Reactor Vessel Surveillance Capsule Withdrawal Schedule................. 7-1 Analysis of St. Lucie Unit 1 Capsule 2840
iv LIST OF FIGURES Figure 4-1 Arrangement of Surveillance Capsules in the St. Lucie Unit 1 Reactor Vessel............... 4-3 Figure 4-2 Typical St. Lucie Unit I Surveillance Capsule Assembly................................................
4-4 Figure 4-3 Typical St. Lucie Unit 1 Surveillance Capsule Charpy Impact Compartment A ssem bly............................................................................................................
..... 4-5 Figure 4-4 Typical St. Lucie Unit 1 Surveillance Capsule Tensile and Flux Monitor C om partm ent A ssem bly...................................................................................................
4-6 Figure 5-1 Charpy V-Notch Impact Energy vs. Temperature for St. Lucie Unit 1 Reactor Vessel Lower Shell Plate C-8-2 (Longitudinal Orientation)...........................................
5-17 Figure 5-2 Charpy V-Notch Lateral Expansion vs. Temperature for St. Lucie Unit 1 Reactor Vessel Lower Shell Plate C-8-2 (Longitudinal Orientation)............................................
5-18 Figure 5-3 Charpy V-Notch Percent Shear vs. Temperature for St. Lucie Unit 1 Reactor Vessel Lower Shell Plate C-8-2 (Longitudinal Orientation)............................................
5-19 Figure 5-4 Charpy V-Notch Impact Energy vs. Temperature for St. Lucie Unit 1 Reactor Vessel Lower Shell Plate C-8-2 (Transverse Orientation)..............................................
5-20 Figure 5-5 Charpy V-Notch Lateral Expansion vs. Temperature for St. Lucie Unit 1 Reactor Vessel Lower Shell Plate C-8-2 (Transverse Orientation)...............................................
5-21 Figure 5-6 Charpy V-Notch Percent Shear vs. Temperature for St. Lucie Unit 1 Reactor Vessel Lower Shell Plate C-8-2 (Transverse Orientation)..............................................
5-22 Figure 5-7 Charpy V-Notch Impact Energy vs. Temperature for St. Lucie Unit 1 Reactor V essel Surveillance W eld M etal.......................................................................................
5-23 Figure 5-8 Charpy V-Notch Lateral Expansion vs. Temperature for St. Lucie Unit 1 Reactor V essel Surveillance W eld M etal.......................................................................................
5-24 Figure 5-9 Charpy V-Notch Percent Shear vs. Temperature for St. Lucie Unit 1 Reactor V essel Surveillance W eld M etal......................................................................................
5-25 Figure 5-10 Charpy V-Notch Impact Energy vs. Temperature for St. Lucie Unit 1 Reactor V essel H eat A ffected Zone M aterial.................................................................................
5-26 Analysis of St. Lucie Unit I Capsule 284'
v LIST OF FIGURES (CONTINUED)
Figure 5-11 Charpy V-Notch Lateral Expansion vs. Temperature for St. Lucie Unit I Reactor V essel H eat A ffected Zone M aterial.................................................................................
5-27 Figure 5-12 Charpy V-Notch Percent Shear vs. Temperature for St. Lucie Unit 1 Reactor V essel H eat A ffected Zone M aterial.................................................................................
5-28 Figure 5-13 Charpy Impact Specimen Fracture Surfaces for St. Lucie Unit 1 Reactor Vessel Lower Shell Plate C-8-2 (Transverse Orientation)...........................................................
5-29 Figure 5-14 Charpy Impact Specimen Fracture Surfaces for St. Lucie Unit I Reactor Vessel Lower Shell Plate C-8-2 (Longitudinal Orientation)........................................................
5-30 Figure 5-15 Charpy Impact Specimen Fracture Surfaces for St. Lucie Unit 1 Reactor Vessel W eld M etal Specim ens.....................................................................................................
5-3 1 Figure 5-16 Charpy Impact Specimen Fracture Surfaces for St. Lucie Unit 1 Reactor Vessel H eat A ffected Z one (H A Z )...............................................................................................
5-32 Figure 5-17 Tensile Properties for St. Lucie Unit 1 Reactor Vessel Lower Shell Plate C-8-2 (L ongitudin al O rientation)...............................................................................................
5-33 Figure 5-18 Tensile Properties for St. Lucie Unit 1 Reactor Vessel Weld Metal................................
5-34 Figure 5-19 Tensile Properties for St. Lucie Unit 1 Reactor Vessel Heat-Affected-Zone (HAZ)...... 5-35 Figure 5-20 Fractured Tensile Specimens from St. Lucie Unit 1 Reactor Vessel Lower Shell Plate C-8-2 (Longitudinal Orientation)..................................................................
5-36 Figure 5-21 Fractured Tensile Specimens from St. Lucie Unit 1 Reactor Vessel Weld Metal........... 5-37 Figure 5-22 Fractured Tensile Specimens from St. Lucie Unit 1 Reactor Vessel Heat Affected Z o n e (H A Z ).......................................................................................................................
5 -3 8 Figure 5-23 Engineering Stress-Strain Curves for Lower Shell Plate C-8-2 Tensile Specimens 1J4, 1JL and 1JM (Longitudinal Orientation)..................................................................
5-39 Figure 5-24 Engineering Stress-Strain Curve for Weld Metal Tensile Specimens 3J2, 3JJ, an d 3 JY.............................................................................................................................
5 -4 0 Figure 5-25 Engineering Stress-Strain Curves For Heat-Affected Zone (HAZ) Material Tensile Specim ens 4KJ, 4KK, and 4KY...........................................................................
5-41 Analysis of St. Lucie Unit I Capsule 2840
vi PREFACE This report has been technically reviewed and verified by:
Reviewer:
Sections 1 through 5, 7, 8, Appendices A, B, C and D J. H. Ledger Section 6 T.J. Hall RECORD OF REVISION Revision 1: This WCAP report is being revised to correct errors in the DORT models used in the original calculations. This issue was originally identified in Reference 35. The errors included a non conservative assumption in the thickness of the core support barrel and a conservative assumption in the moderator density of the down-comer region. The incorrect thickness was present in the models used for Cycles I through 15, as well as the future projections. The incorrect moderator density was present only in Cycles 1 through 4. Additionally, some changes have been made (relative to the original issue) in the selection of the sensor measurements used in the neutron dosimetry analysis as a result of the updated DORT calculations.
Revision I to this WCAP report supersedes the original issue in its entirety.
Analysis of St. Lucie Unit I Capsule 2840
vii EXECUTIVE
SUMMARY
The purpose of this report is to document the results of the testing of surveillance capsule 2840 from St.
Lucie Unit 1. Capsule 2840 was removed at 17.23 EFPY and post irradiation mechanical tests of the Charpy V-notch and tensile specimens was performed, along with a fluence evaluation based methodology and nuclear data including recently released neutron transport and dosimetry cross-section libraries derived from the ENDF/B-VI database. The calculated peak clad base/metal vessel fluence after 17.23 EFPY of plant operation was 1.45 x 109 n/cm2 and the surveillance Capsule 2840 calculated fluence was 1.45 x 10i9 n/cm 2.
A brief summary of the Charpy V-notch testing results can be found in Section 1 and the updated capsule removal schedule can be found in Section 7. A supplement to this report is a credibility evaluation, which can be found in Appendix D, that shows the St. Lucie Unit 1 surveillance plate and weld data is credible.
Analysis of St. Lucie Unit 1 Capsule 284°
1-1 1
SUMMARY
OF RESULTS The analysis of the reactor vessel materials contained in surveillance capsule 2840 the third capsule to be removed from the St. Lucie Unit 1 reactor pressure vessel, led to the following conclusions: (General Note:
Temperatures are reported to two significant digits only to match CVGraph output.)
The capsule received an average fast neutron calculated fluence (E > 1.0 MeV) of 1.45 x 10"9 n/cm2 after 17.23 effective full power years (EFPY) of plant operation.
Irradiation of the reactor vessel lower shell plate C-8-2 Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major working direction of the plate (longitudinal orientation), to 1.45 x 1019 n/cm2 (E> 1.0MeV) resulted in a 30 ft-lb transition temperature increase of 87.930 F and a 50 ft-lb transition temperature increase of 95.38°F. This results in an irradiated 30 ft-lb transition temperature of 95.18°F and an irradiated 50 ft-lb transition temperature of 130.62°F for the longitudinally oriented specimens Irradiation of the reactor vessel lower shell plate C-8-2 Charpy specimens, oriented with the longitudinal axis of the specimen normal to the major working direction of the plate (transverse orientation), to 1.45 x 10' 9 n/cm2 (E> 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 84.99°F and a 50 ft-lb transition temperature increase of 97.55°F. This results in an irradiated 30 ft-lb transition temperature of 100.37°F and an irradiated 50 ft-lb transition temperature of 143.57°F for transversely oriented specimens.
Irradiation of the weld metal Charpy specimens to 1.45 x 10'9 n/cm2 (E> 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 68.00°F and a 50 ft-lb transition temperature increase of 67.19'F. This results in an irradiated 30 ft-lb transition temperature of 10.01°F and an irradiated 50 ft-lb transition temperature of 31.58°F.
Irradiation of the weld Heat-Affected-Zone (HAZ) metal Charpy specimens to 1.45 x 1019 n/cm2 (E >
1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 74.79°F and a 50 ft-lb transition temperature increase of 82.27°F. This results in an irradiated 30 ft-lb transition temperature of
-4.87°F and an irradiated 50 ft-lb transition temperature of 44.91 OF.
The average upper shelf energy of the lower shell plate C-8-2 (longitudinal orientation) resulted in an average energy decrease of 29 ft-lb after irradiation to 1.45 x 1039 n/cm 2 (E> 1.0 MeV). This results in an irradiated average upper shelf energy of 110 ft-lb for the longitudinally oriented specimens.
The average upper shelf energy of the lower shell plate C-8-2 (transverse orientation) resulted in an average energy decrease of 15 ft-lb after irradiation to 1.45 x 1019 n/cm2 (E > 1.0 MeV). This results in an irradiated average upper shelf energy of 88 ft-lb for the transversely oriented specimens.
The average upper shelf energy of the weld metal Charpy specimens resulted an average energy decrease of 34 ft-lb after irradiation to 1.45 x 1019 n/cm 2 (E> 1.0 MeV). Hence, this results in an irradiated average upper shelf energy of 110 ft-lb for the weld metal specimens.
Analysis of St. Lucie Unit 1 Capsule 2840
1-2 The average upper shelf energy of the weld HAZ metal Charpy specimens resulted an average energy decrease of 40 ft-lb after irradiation to 1.45 x 1019 n/cm2 (E > 1.0 MeV). Hence, this results in an irradiated average upper shelf energy of 93 ft-lb for the weld HAZ metal.
A comparison of the St. Lucie Unit 1 reactor vessel beltline material test results with the Regulatory Guide 1.99, Revision 2[1], predictions led to the following conclusions:
The measured 30 ft-lb shift in transition temperature values for all the surveillance program materials (Weld and Plate) for capsule 284' is less than the Regulatory Guide 1.99, Revision 2, predictions.
The measured percent decrease in upper shelf energy of the Capsule 2840 surveillance material is less than the Regulatory Guide 1.99, Revision 2, predictions.
The peak calculated and best estimate end-of-license (32 EFPY) neutron fluence (E> 1.0 MeV) at the core midplane for the St. Lucie Unit 1 reactor vessel using the Regulatory Guide 1.99, Revision 2 attenuation formula (ie. Equation # 3 in the guide; fýdepUhx) = ffkce
- e (-024x)) is as follows:
Calculated:
Vessel inner radius*
=
2.55 x 1019 n/cm2 Vessel 1/4 thickness
=
1.52 x 1019n/cm 2 Vessel 3/4 thickness
=
5.40 x 1018 n/cm 2 Best Estimate:
Vessel inner radius*
=
2.46 x 1019 n/cm 2 Vessel 1/4 thickness
=
1.47 x 1019n/cm 2 Vessel 3/4 thickness
=
5.21 x 1018 n/cm 2
"* The credibility evaluation of the St. Lucie Unit 1 surveillance program presented in Appendix D of this report indicates that the surveillance results for lower shell plate C-8-2 and the weld metal are credible.
"* All beltline materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are expected to maintain an upper shelf energy greater than 50 ft-lb throughout the life of the vessel (32 EFPY) as required by 10CFR50, Appendix G Analysis of St. Lucie Unit I Capsule 284'
2-1 2
INTRODUCTION This report presents the results of the examination of the Capsule located at 2840, the third capsule to be removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation on the St. Lucie Unit 1 reactor pressure vessel materials under actual operating conditions.
The surveillance program for the Florida Power and Light Company St. Lucie Unit 1 reactor pressure vessel materials was designed and recommended by Combustion Engineering. A description of the surveillance program and the preirradiation mechanical properties of the reactor vessel materials is presented in Reference
- 3. The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM E185-70, "Standard Practice for conducting Surveillance for light-water cooled Nuclear Power Reactor Vessels". Capsule 284' was'removed from the reactor after 17.23 EFPY of exposure and shipped to the Westinghouse Science and Technology Center Hot Cell Facility, where the postirradiation mechanical testing of the Charpy V-notch impact and tensile surveillance specimens was performed.
This report summarizes the testing of and the post-irradiation data obtained from surveillance capsule located at 284', removed from the St. Lucie Unit 1 reactor vessel and discusses the analysis of the data.
Analysis of St. Lucie Unit I Capsule 2840
3-1 3
BACKGROUND The ability of the large steel pressure vessel containing the reactor core and its primary coolant to resist fracture constitutes an important factor in ensuring safety in the nuclear industry. The beltline region of the reactor pressure vessel is the most critical region of the vessel because it is subjected to significant fast neutron bombardment. The overall effects of fast neutron irradiation on the mechanical properties of low alloy, ferritic pressure vessel steels such as A533 Grade B Class 1 (base material of the Florida Power and Light Company St. Lucie Unit I reactor pressure vessel beltline) are well documented in the literature.
Generally, low alloy ferritic materials show an increase in hardness and tensile properties and a decrease in ductility and toughness during high-energy irradiation.
A method for ensuring the integrity of reactor pressure vessels has been presented in "Fracture Toughness Criteria for Protection Against Failure," Appendix G to Section XI of the ASME Boiler and Pressure Vessel Code[41. The method uses fracture mechanics concepts and is based on the reference nil-ductility transition temperature (RTNDT).
RTNDT is defined as the greater of either the drop weight nil-ductility transition temperature (NDTT per ASTM E-208151) or the temperature 60'F less than the 50 ft-lb (and 35-mil lateral expansion) temperature as determined from Charpy specimens oriented perpendicular (transverse) to the major working direction of the plate. The RTNDT of a given material is used to index that material to a reference stress intensity factor curve (KIa curve) which appears in Appendix G to the ASME Codel41. The KI, curve is a lower bound of dynamic, crack arrest, and static fracture toughness results obtained from several heats of pressure vessel steel. When a given material is indexed to the Kia curve, allowable stress intensity factors can be obtained for this material as a function of temperature. Allowable operating limits can then be determined utilizing these allowable stress intensity factors. Note that Code Case N-640 now allows the use of the KI, curve as an alternative to the Ki*a curve.
RTNDT and, in turn, the operating limits of nuclear power plants can be adjusted to account for the effects of radiation on the reactor vessel material properties. The changes in mechanical properties of a given reactor pressure vessel steel, due to irradiation, can be monitored by a reactor surveillance program, such as the St.
Lucie Unit 1 reactor vessel radiation surveillance programi 61, in which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens tested. The increase in the average Charpy V-notch 30 ft-lb temperature (ARTNT) due to irradiation is added to the initial RTNDT, along with a margin (M) to cover uncertainties, to adjust the RTNDT (ART) for radiation embrittlement. This ART (RTNDT initial + M + ARTNDT) is used to index the material to the KIa curve and, in turn, to set operating limits for the nuclear power plant that take into account the effects of irradiation on the reactor vessel materials.
Analysis of St. Lucie Unit 1 Capsule 2840
4-1 4
DESCRIPTION OF PROGRAM Six surveillance capsules for monitoring the effects of neutron exposure on the St. Lucie Unit 1 reactor pressure vessel core region (beltline) materials were inserted in the reactor vessel prior to initial plant start-up. The capsules were positioned in the reactor vessel between the thermal shield and the vessel wall at locations shown in Figure 4-1. The vertical center of the capsules is opposite the vertical center of the core.
Capsule 2840 was removed after 17.23 effective full power years (EFPY) of plant operation. This capsule contained Charpy V-notch impact and tensile specimens made from reactor vessel lower shell course Plate C-8-2, submerged arc weld metal identical to the beltline region girth weld seam and heat-affected-zone (HAZ) metal. All HAZ specimens are obtained within the heat-affected-zone of Plate C-8-2. Standard Reference Material from HSST-01MY Plate was included in the program in addition to the reactor vessel materials, but not within capsule 284'.
Test specimens obtained from lower shell plate C-8-2 (after the thermal heat treatment and forming of the plate) was taken at least one plate thickness from the quenched ends of the plate. All test specimens were machined from the 1/4 thickness location of the plate after performing a simulated post-weld stress-relieving treatment on the test material and also from weld and HAZ metal of a stress-relieved weldment joining plates C-8-3 and C-8-1. All heat-affected-zone specimens were obtained from the weld heat-affected-zone of Plate C-8-2.
Charpy V-notch impact specimens from Plate C-8-2 were with the longitudinal axis of the specimen parallel to the major working direction of the plate (longitudinal orientation). Charpy V-notch impact specimens from Plate C-8-2 were with the transverse axis of the specimen perpendicular to the major working direction of the plate (transverse orientation). The Charpy V-notch specimens from the weld metal were machined with the longitudinal axis of the specimen transverse to the weld direction with the notch oriented in the direction of the weld.
Tensile specimens from Plate C-8-2 were machined in with the longitudinal axis of the specimen parallel to the major working direction of the plate (longitudinal orientation). Tensile specimens from the weld metal were oriented with the longitudinal axis of the specimen transverse to the weld direction.
Capsule 2840 contained dosimeter wires of sulfur, iron, titanium, nickel (cadmium-shielded), aluminum cobalt (cadmium-shielded and unshielded), copper (cadmium shielded) and uranium (cadmium-shielded and unshielded).
The capsule contained thermal monitors made from four low-melting-point eutectic alloys and sealed in glass capsules. These thermal monitors were used to define the maximum temperature attained by the test specimens during irradiation. The composition of the four eutectic alloys and their melting points are:
80% Au, 20% Sn Melting Point 5361F (2801C) 90% Pb, 5% Sn, 5% Ag Melting Point 5581F (2920C) 2.5% Ag, 97.5% Pb Melting Point 5801F (3040C) 1.75% Ag, 0.75% Sn, 97.5% Ag Melting Point 590°F (310°C)
Analysis of St. Lucie Unit 1 Capsule 2840
4-2 The arrangement of the various mechanical test specimens, dosimeters and thermal monitors contained in capsule 2840 is shown in Figure 4-2.
A typical St. Lucie Unit 1 surveillance capsule Charpy impact compartment assembly is shown in Figure 4-3.
A typical St. Lucie Unit 1 surveillance capsule tensile and flux-monitor compartment assembly is shown in Figure 4-4.
The heat treatment for the plate material consisted of austenitization at 1600'F +/-25°F for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; water quenched and tempered at 1225 'F +25'F for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. After a 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> stress relief at 11507F +/-25'F the plates were furnace cooled to 600 'F. The weldment received a final 41 hour4.74537e-4 days <br />0.0114 hours <br />6.779101e-5 weeks <br />1.56005e-5 months <br /> and 45 minute stress relief at 1100 to 1150 0F.
Analysis of St. Lucie Unit I Capsule 284'
4-3 Enlarged Pla 1-,
\\,"Oulle Nozzie
'W Ou_ N\\Nozzle Core Shr Core Support Barrel
,VeVessel SReactor Vessel
!2630Cr Vse*
Midplane]
Vesselseml Vessel aCapsule 2770 Assembly "Vessel Core Reacto.r Support Vessel Barrel
\\
/
n Vie Elevation View Figure 4-1.
Arrangement of Surveillance Capsules in the St. Lucie Unit I Reactor Vessel Analysis of St. Lucie Unit 1 Capsule 284'
4-4 Lock Assembly Tensile -Monitor.
Compartment Tensile -Monitor Compartment Tensile -Monitor Compartment
}
Wedge Coupling Assembly Charpy Impact Compartments Charpy Impact Compartments Figure 4-2 Typical St. Lucie Unit 1 Surveillance Capsule Assembly Analysis of St. Lucie Unit 1 Capsule 284'
4-5 Coupling - End Cap Charpy Impact Specimens Rectangular Tubing Nedge Coupling - End Cap Figure 4-3 Typical St. Lucie Unit 1 Surveillance Capsule Charpy Impact Compartment Assembly Analysis of St. Lucie Unit I Capsule 2840
4-6 Wedge Coupling - End Flux Monitor Housing Stainless Steel Tubing, Threshold Detector -
Temperature Monitor-Temperature Monitor Housing Split Spacer Tensile Specimen Housing
.,-Flux Spectrum Monitor Cadmium Shielded Stainless Steel Tubing S
Cadmium Shield Threshold Detector
-Quartz Tubing S
Weight
_j Low.Melting Alloy
.ectangular Tubing Nedge Coupling - End Cap Figure 4-4 Typical St. Lucie Unit 1 Surveillance Capsule Tensile and Flux-Monitor Compartment Assembly Analysis of St. Lucie Unit 1 Capsule 284'
5-1 5
TESTING OF SPECIMENS FROM CAPSULE 2840 5.1 OVERVIEW The post-irradiation mechanical testing of the Charpy V-notch impact specimens and tensile specimens was performed in the Remote Metallographic Facility (RMF) at the Westinghouse Science and Technology Center. Testing was performed in accordance with 10CFR50, Appendices G and HW 23, ASTM Specification El 85-82t1', and Westinghouse Procedure RMF 8402, Revision 2 as modified by Westinghouse RMF Procedures 8102, Revision 1, and 8103, Revision 1.
Upon receipt of the capsule at the hot cell laboratory, the specimens and spacer blocks were carefully removed, inspected for identification number, and checked against the master lists in TR-F-MCM-00513 1 and CENPD-391 61. No discrepancies were found.
Examination of the four low-melting, eutectic alloy thermal monitors indicated that the two lowest melting point monitors melted. Based on this examination, the maximum temperature to which the test specimens were exposed to was between 5591F and 5791F.
The Charpy impact tests were performed per ASTM Specification E23-98N and RMF Procedure 8103, Revision 1, on a Tinius-Olsen Model 74, 358J machine. The tup (striker) of the Charpy impact test machine is instrumented with a GRC 930-I instrumentation system, feeding information into an IBM compatible computer. With this system, load-time and energy-time signals can be recorded in addition to the standard measurement of Charpy energy (ED). From the load-time curve (Appendix A), the load of general yielding (PGA), the time to general yielding (toy), the maximum load (PM), and the time to maximum load (tM) can be determined. Under some test conditions, a sharp drop in load indicative of fast fracture was observed. The load at which fast fracture was initiated is identified as the fast fracture load (PF), and the load at which fast fracture terminated is identified as the arrest load (PA). The energy at maximum load (Em) was determined by comparing the energy-time record and the load-time record. The energy at maximum load is approximately equivalent to the energy required to initiate a crack in the specimen. Therefore, the propagation energy for the crack (E,?) is the difference between the total energy to fracture (ED) and the energy at maximum load (EM).
The yield stress (cy) was calculated from the three-point bend formula having the following expression:
uy=(P0 y *L) / [B * (W-a) 2
- C]
(1) where:
L
=
distance between the specimen supports in the impact machine B
=
the width of the specimen measured parallel to the notch W
=
height of the specimen, measured perpendicularly to the notch a
=
notch depth Analysis of St. Lucie Unit I Capsule 284'
5-2 The constant C is dependent on the notch flank angle (4)), notch root radius (p) and the type of loading (i.e.,
pure bending or three-point bending). In three-point bending, for a Charpy specimen in which 4,= 450 and p =
0.010 inch, Equation 1 is valid with C = 1.21. Therefore, (for L = 4W),
cy= (PGy *L) / [B * (W - a) 2 *1.21] = (3.33 *PGy
- V) / [B * (W - a) 2]
(2)
For the Charpy specimen, B = 0.394 inch, W = 0.394 inch and a = 0.079 inch. Equation 2 then reduces to:
c= 33.3 *Per (3) where cy, is in units of psi and PGY is in units of lbs. The flow stress was calculated from the average of the yield and maximum loads, also using the three-point bend formula.
The symbol A in columns 4, 5, and 6 of Tables 5-5 through 5-8 is the cross-section area under the notch of the Charpy specimens:
A=B * (W - a) = 0.1241 sq.in.
(4)
Percent shear was determined from post-fracture photographs using the ratio-of-areas methods in compliance with ASTM Specification A370-9719 1. The lateral expansion was measured using a dial gage rig similar to that shown in the same specification.
Tensile tests were performed on a 20,000-pound Instron, split-console test machine (Model 1115) per ASTM Specification E8-99"101 and E21-92E1"1, and RMF Procedure 8102, Revision 1. All pull rods, grips, and pins were made of Inconel 718. The upper pull rod was connected through a universal joint to improve axiality of loading. The tests were conducted at a constant crosshead speed of 0.05 inches per minute throughout the test.
Extension measurements were made with a linear variable displacement transducer extensometer. The extensometer knife edges were spring-loaded to the specimen and operated through specimen failure. The extensometer gage length was 1.00 inch. The extensometer is rated as Class B-2 per ASTM E83-93P21.
Elevated test temperatures were obtained with a three-zone electric resistance split-tube furnace with a 9-inch hot zone. All tests were conducted in air. Because of the difficulty in remotely attaching a thermocouple directly to the specimen, the following procedure was used to monitor specimen temperatures.
Chromel-Alumel thermocouples were positioned at the center and at each end of the gage section of a dummy specimen and in each tensile machine griper. In the test configuration, with a slight load on the specimen, a plot of specimen temperature versus upper and lower tensile machine griper and controller temperatures was developed over the range from room temperature to 550'F. During the actual testing, the grip temperatures were used to obtain desired specimen temperatures. Experiments have indicated that this method is accurate to +20F.
The yield load, ultimate load, fracture load, total elongation, and uniform elongation were determined directly from the load-extension curve. The yield strength, ultimate strength, and fracture strength were calculated using the original cross-sectional area. The final diameter and final gage length were determined from post-fracture photographs. The fracture area used to calculate the fracture stress (true stress at fracture) and percent reduction in area was computed using the final diameter measurement.
Analysis of St. Lucie Unit 1 Capsule 2840
5-3 5.2 CHARPY V-NOTCH IMPACT TEST RESULTS The results of the Charpy V-notch impact tests performed on the. various materials contained in capsule 2840, which received a fluence of 1.45 x 1019 n/cm2 (E > 1.0 MeV) in 17.23 EFPY of operation, are presented in Tables 5-1 through 5-8 and are compared with unirradiated results as shown in Figures 5-1 through 5-12.
The transition temperature increases and upper shelf energy decreases for the capsule 2840 materials are summarized in Table 5-9. These results led to the following conclusions:
Irradiation of the reactor vessel lower shell plate C-8-2 Charpy specimens, oriented with the longitudinal axis of the specimen parallel to the major working direction of the plate (longitudinal orientation), to 1.45 x 1019 n/cm2 (E> 1.0MeV) resulted in a 30 ft-lb transition temperature increase of 87.93'F and a 50 ft-lb transition temperature increase of 95.38'F. This results in an irradiated 30 ft-lb transition temperature of 95.18'F and an irradiated 50 ft-lb transition temperature-of 130.62°F for the longitudinally oriented specimens Irradiation of the reactor vessel lower shell plate C-8-2 Charpy specimens, oriented with the longitudinal axis of the specimen normal to the major working direction of the plate (transverse orientation), to 1.45 x 1019 n/cm2 (E> 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 84.99°F and a 50 ft-lb transition temperature increase of 97.55°F. This results in an irradiated 30 ft-lb transition temperature of 100.371F and an irradiated 50 ft-lb transition temperature of 143.57'F for transversely oriented specimens.
Irradiation of the weld metal Charpy specimens to 1.45 x 1019 n/cm 2 (E> 1.0MeV) resulted in a 30 ft-lb transition temperature increase of 68.00'F and a 50 ft-lb transition temperature increase of 67.19'F.
This results in an irradiated 30 ft-lb transition temperature of 10.0 IF and an irradiated 50 ft-lb transition temperature of 31.580 F.
Irradiation of the weld Heat-Affected-Zone (HAZ) metal Charpy specimens to 1.45 x 1019 n/cm 2 (E > 1.0 MeV) resulted in a 30 ft-lb transition temperature increase of 74.79IF and a 50 ft-lb transition temperature increase of 82.271F. This results in an irradiated 30 ft-lb transition temperature of
-4.87IF and an irradiated 50 ft-lb transition temperature of 44.9 1°F.
The average upper shelf energy of the lower shell plate C-8-2 (longitudinal orientation) resulted in an average energy decrease of 29 ft-lb after irradiation to 1.45 x 1019 n/cm2 (E> 1.0 MeV). This results in an irradiated average upper shelf energy of 110 ft-lb for the longitudinally oriented specimens.
The average upper shelf energy of the lower shell plate C-8-2 (transverse orientation) resulted in an average energy decrease of 15 ft-lb after irradiation to 1.45 x 1019 n/cm2 (E > 1.0 MeV). This results in an irradiated average upper shelf energy of 88 ft-lb for the transversely oriented specimens.
The average upper shelf energy of the weld metal Charpy specimens resulted an average energy decrease of 34 ft-lb after irradiation to 1.45 x 1019 n/cm2 (E> 1.0 MeV). Hence, this results in an irradiated average upper shelf energy of 110 ft-lb for the weld metal specimens.
The average upper shelf energy of the weld HAZ metal Charpy specimens resulted an average energy decrease of 40 ft-lb after irradiation to 1.45 x 1019 n/cm2 (E > 1.0 MeV). Hence, this results in an irradiated average upper shelf energy of 93 ft-lb for the weld HAZ metal.
Analysis of St. Lucie Unit I Capsule 2840
5-4 A comparison, as presented in Table 5-10, of the St. Lucie Unit I reactor vessel beltline material test results with the Regulatory Guide 1.99, Revision 2ý'I, predictions led to the following conclusions:
The measured 30 ft-lb shift in transition temperature values for all the surveillance program materials (Weld and Plate) for capsule 2840 is less than the Regulatory Guide 1.99, Revision 2, predictions.
The measured percent decrease in upper shelf energy of the Capsule 2840 surveillance material is less than the Regulatory Guide 1.99, Revision 2, predictions.
Further comparisons are made in the credibility evaluation presented in Appendix D The fracture appearance of each irradiated Charpy specimen from the various surveillance capsule 2840 materials is shown in Figures 5-13 through 5-16 and show an increasingly ductile or tougher appearance with increasing test temperature.
All beltline materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are expected to maintain an upper shelf energy of no less than 50 ft-lb throughout the life of the vessel (32 EFPY) as required by 10CFR50, Appendix G.
The load-time records for individual instrumented Charpy specimen tests are shown in Appendix A.
The Charpy V-notch data presented in this report is based on a re-plot of all capsule data using CVGRAPH, Version 4. 1, which is a hyperbolic tangent curve-fitting program. Hence, Appendix C contains a comparison of the Charpy V-notch shift results for each surveillance material (hand-fitting versus hyperbolic tangent curve-fitting). Additionally, Appendix B presents the CVGRAPH, Version 4.1, Charpy V-notch plots and the program input data.
5.3 TENSILE TEST RESULTS The results of the tensile tests performed on the various materials contained in capsule 2840 irradiated to 1.45 x 1019 n/cm2 (E > 1.0 MeV) are presented in Table 5-11 and are compared with unirradiated results as shown in Figures 5-17 through 5-19.
The results of the tensile tests performed on the lower shell plate C-8-2 (longitudinal orientation) indicated that irradiation to 1.45 x 1019 n/cm2 (E> 1.0 MeV) caused an approximate increase of 9 to 11 ksi in the 0.2 percent offset yield strength and approximately a 9 to 12 ksi increase in the ultimate tensile strength when compared to unirradiated datat11 (Figure 5-17).
The results of the tensile tests performed on the surveillance weld metal indicated that irradiation to 1.45 x 1019 n/cm2 (E > 1.0 MeV) caused a 10 to 17 ksi increase in the 0.2 percent offset yield strength and a 7 to 9 ksi increase in the ultimate tensile strength when compared to unirradiated data (Figure 5-18).
The results of the tensile tests performed on the surveillance HAZ metal indicated that irradiation to 1.45 x 1019 n/cm2 (E > 1.0 MeV) caused a 12 to 16 ksi increase in the 0.2 percent offset yield strength and a 10 to 13 ksi increase in the ultimate tensile strength when compared to unirradiated data (Figure 5-19).
Analysis of St. Lucie Unit 1 Capsule 2840
5-5 The fractured tensile specimens for the lower shell plate C-8-2 material are shown in Figure 5-20, while the fractured tensile specimens for the surveillance weld metal and heat-affected-zone material are shown in Figures 5-21 and 5-22, respectively.
The engineering stress-strain curves for the tensile tests are shown in Figures 5-23 through 5-25.
Analysis of St. Lucie Unit I Capsule 284'
5-6 Table 5-1 Charpy V-notch Data for the St. Lucie Unit 1 Lower Shell Plate C-8-2 Irradiated to a Fluence of 1.45 x 1019 n/cm 2 (E> 1.0 MeV) (Longitudinal Orientation)
Sample Temperature Impact Energy Lateral Expansion Shear Number F
C ft-lbs Joules mils mm 155 5
-15 8
11 3
0.08 3
136 25
-4 12 16 12 0.30 10 14M 50 10 7
9 9
0.23 5
145 72 22 16 22 9
0.23 15 16D 80 27 19 26 14 0.36 20 146 100 38 40 54 28 0.71 30 164 120 49 55 75 40 1.02 35 157 135 57 49 66 32 0.81 40 143 150 66 56 76 40 1.02 40 117 200 93 88 119 59 1.50 70 14U 275 135 107 145 78 1.98 100 14E 325 163 112 152 75 1.91 100 Analysis of St. Lucie Unit I Capsule 2840
5-7 Analysis of St. Lucie Unit I Capsule 2840 Table 5-2 Charpy V-notch Data for the St. Lucie Unit 1 Lower Shell Plate C-8-2 Irradiated to a Fluence of 1.45 x 1019 n/cm2 (E> 1.0 MeV) (Transverse Orientation)
Sample Temperature Impact Energy Lateral Expansion Shear Number F
C ft-lbs Joules mils mm 252 10
-12 15 20 10 0.25 10 25M 50 10 14 19 11 0.28 10 25L 90 32 28 38 24 0.61 20 25C 100 38 27 37 19 0.48 20 26C 115 46 37 50 24 0.61 20 26A 125 52 42 57 28 0.71 25 25J 135 57 37 50 34 0.86 30 251 150 66 57 77 43 1.09 40 253 175 79 66 89 49 1.24 50 26B 220 104 77 104 60 1.52 95 25P 275 135 92 125 69 1.75 100 24Y 325 163 83 113 62 1.57 100
5-8 Table 5-3 Charpy V-notch Data for the St. Lucie Unit 1 Surveillance Weld Metal Irradiated to a Fluence of 1.45 x 1019 n/cm2 (E> 1.0 MeV)
Sample Temperature Impact Energy Lateral Expansion Shear Number F
C ft-lbs Joules mils mm 31C
-50
-46 5
7 4
0.10 10 36L
-10
-23 12 16 8
0.20 15 343 0
-18 26 35 20 0.51 20 325 10
-12 18 24 13 0.33 20 317 15
-9 36 49 25 0.64 30 33K 25
-4 34 46 26 0.66 40 35J 30
-1 63 85 46 1.17 50 367 40 4
65 88 47 1.19 50 31D 100 38 87 118 63 1.60 85 34M 150 66 107 145 75 1.91 95 32T 200 93 115 156 83 2.11 100 353 250 121 107 145 81 2.06 100 Analysis of St. Lucie Unit I Capsule 2840
5-9 Analysis of St. Lucie Unit I Capsule 284' Table 5-4 Charpy V-notch Data for the St. Lucie Unit 1 Heat Affected Zone Metal Irradiated to a Fluence of 1.45 x 1019 n/cm 2 (E> 1.0 MeV)
Sample Temperature Impact Energy Lateral Expansion Shear Number F
C ft-lbs Joules mils mm 45L
-40
-40 9
12 6
0.15 20 46C
-5
-21 21 28 20 0.51 40 46B 10
-12 31 42 20 0.51 30 421 20
-7 55 75 37 0.94 50 45M 30
-1 41 56 35 0.89 60 461 40 4
62 84 40 1.02 65 463 60 16 36 49 30 0.76 40 46D 72 22 85 115 57 1.45 90 45U 100 38 73 99 49 1.24 75 45T 135 57 57 77 41 1.04 60 462 200 93 79 107 59 1.50 95 464 250 121 107 145 74 1.88 100
5-10 Table 5-5 Instrumented Charpy Impact Test Results for the St. Lucie Unit 1 Lower Shell Plate C-8-2 Irradiated to a Fluence of 1.45 x 1019 n/cm2 (E>1.0 MeV)
(Longitudinal Orientation)
Normalized Energies (ft-lb/in 2)
Charpy Yield Time to Time to Fast Test Energy Load Yield tcy Max.
Max.
Fract.
Arrest Yield Flow Sample Temp.
ED Charpy Max.
Prop.
PGY (msec)
Load PM T,
Load PF Load PA Stress Sy Stress No.
('F)
(ft-lb)
ED/A EM/A Ep/A (ib)
(Ib)
(msec)
(lb)
(Ib)
(ksi)
(ksi) 155 5
8 64 33 32 3435 0,16 3452 0.16 3435 0
114 115 136 25 12 97 56 40 4133 0.17 4340 0.20 4340 0
138 141 14M 50 7
56 28 29 3138 0.15 3138 0.15 3138 0
104 104 145 72 16 129 69 59 4025 0.17 4395 0.22 4395 0
134 140 16D 80 19 153 75 78 3764 0.17 4298 0.24 4156 82 125 134 146 100 40 322 239 83 3974 0.17 4791 0.51 4738 166 132 146 164 120 55 443 337 106 3906 0.17 4854 0.68 4732 234 130 146 157 135 49 395 240 155 3950 0.17 4786 0.52 4641 682 132 145 143 150 56 451 333 118 3931 0.17 4848 0.67 4647 866 131 146 117 200 88 709 330 379 3891 0.17 4819 0.67 4141 2230 130 145 14U 275 107 862 311 552 3262 0.17 4471 0.69 n/a n/a 109 129 Analysis of St. Lucie Unit I Capsule 2840
5-1 1 Table 5-6 Instrumented Charpy Impact Test Results for the St. Lucie Unit 1 Lower Shell Plate C-8-2 Irradiated to a Fluence of 1.45 x 10'9 n/cm2 (E>1.0 MeV)
(Transverse Orientation)
Normalized Energies (ft-lb/in2)
Charpy Yield Time to Time to Fast Test Energy Load Yield tGY Max.
Max.
Fract.
Arrest Yield Flow Sample Temp.
ED Charpy Max.
Prop.
PGY (msec)
Load PM Tm Load PF Load PA Stress Sy Stress No.
(OF)
(ft-lb)
ED/A EM/A Ep/A (Ib)
(lb)
(msec)
(ib)
(lb)
(ksi)
(ksi) 252 10 15 121 68 53 4076 0.17 4580 0.22 4563 0
136 144 25M 50 14 113 64 49 4244 0.17 4571 0.21 4571 0
141 147 25L 90 28 226 146 80 3683 0.17 4412 0.37 4403 372 123 135 25C 100 27 218 70 148 4124 0.17 4427 0.22 4389 674 137 142 26C 115 37 298 152 146 3746 0.17 4389 0.38 4378 446 125 135 26A 125 42 338 201 138 3861 0.17 4592 0.46 4530 908 129 141 25J 135 37 298 222 76 3807 0.17 4633 0.5 4606 1022 127 141 251 150 57 459 234 225 3921 0.17 4711 0.51 4592 2560 131 144 253 175 66 532 248 284 3986 0.17 4910 0.52 4649 2294 133 148 26B 220 77 620 233 387 3471 0.17 4558 0.54 4407 3181 116 134 25P 275 92 741 307 435 3590 0,17 4595 0.65 n/a n/a 120 136 24Y 325 83 669 231 438 3537 0.17 4455 0.53 n/a n/a 118 133 Analysis of St. Lucie Unit 1 Capsule 2840
5-12 Table 5-7 Instrumented Charpy Impact Test Results for the St. Lucie Unit 1 Surveillance Weld Metal Irradiated to a Fluence of 1.45 x 1019 n/cm2 (E>1.0 MeV)
Normalized Energies (ft-lb/in2)
Charpy Yield Time to Time to Fast Test Energy Load Yield tcy Max.
Max.
Fract.
Arrest Yield Flow Sample Temp.
ED Charpy Max.
Prop.
Pcy (msec)
Load PM Tm Load PF Load PA Stress Sy Stress No.
(OF)
(ft-lb)
ED/A EM/A E,/A (lb)
(Ib)
(msec)
(Ib)
(Ib)
(ksi)
(ksi) 31C
-50 5
40 17 23 2056 0.15 2231 0.13 2056 0
68 71 36L
-10 12 97 38 59 3739 0.17 3757 0.17 3739 291.38 125 125 343 0
26 209 72 138 4130 0.17 4522 0.23 4393 128 138 144 325 10 18 145 69 76 4209 0.17 4655 0.22 4620 126 140 148 317 15 36 290 213 77 4350 0.17 4779 0.45 4773 321 145 152 33K 25 34 274 69 205 4193 0.17 4536 0.22
.4295 881 140 145 35J 30 63 508 242 266 4183 0.17 4642 0.52 4417 492 139 147 367 40 65 524 237 286 3904 0.17 4449 0.53 4054 1015 130 139 31D 100 87 701 321 380 3915 0.17 4589 0.67 3786 2385 130 142 34M 150 107 862 314 549 3687 0.17 4381 0.69 3200 2328 123 134 32T 200 115 927 315 611 3279 0.17 4371 0.71 n/a n/a 109 127 353 250 107 862 295 567 3342 0.17 4172 0.69 n/a n/a 111 125 Analysis of St. Lucie Unit I Capsule 2840
5-13 Analysis of St. Lucie Unit I Capsule 284*
Table 5-8 Instrumented Charpy Impact Test Results for the St. Lucie Unit 1 Heat Affected Zone Material Irradiated to a Fluence of 1.45 x 10i9 n/cm 2 (E>1.0 MeV)
Normalized Energies (ft-lb/in2)
Charpy Yield Time to Time to Fast Test Energy Load Yield tGY Max.
Max.
Fract.
Arrest Yield Flow Sample Temp.
ED Charpy Max.
Prop.
PGY (msec)
Load PM Tm Load PF Load PA Stress Sy Stress No.
( 0F)
(ft-lb)
ED/A EM/A E1/A (lb)
(lb)
(msec)
(lb)
(Ib)
(ksi)
(ksi) 45L
-40 9
73 31 41 3411 0.16 3420 0.16 3411 108 114 114 46C
-5 21 169 73 96 4445 0.17 4882 0.22 4677 1497 148 155 46B 10 31 250 180 70 4213 0.18 4648 0.41 4635 467 140 148 421 20 55 443 254 189 4268 0.17 4843 0.52 4671 677 142 152 45M 30 41 330 72 259 4296 0.17 4661 0.22 4356 2461 143 149 461 40 62 500 249 250 4123 0.17 4749 0.53 4579 1924 137 148 463 60 36 290 70 220 4198 0.17 4574 0.22 4367 1486 140 146 46D 72 85 685 353 332 4187 0.17 4920 0.69 4738 3500 139 152 45U 100 73 588 328 260 4030 0.17 4750 0.67 4260 1516 134 146 45T 135 57 459 243 217 4029 0.17 4686 0.52 4151 1826 134 145 462 200 79 637 315 322 3540 0.17 4491 0.69 4262 3189 118 134 464 250 107 862 324 539 3695 0.17 4653 0.68 n/a n/a 123 139
5-14 Table 5-9 Effect of Irradiation to 1.45 x 10' 9 n/cm 2 (E>1.0 MeV) on the Notch Toughness Properties of the St. Lucie Unit 1 Reactor Vessel Surveillance Materials Average 30 (ft-lb)(')
Average 35 mil Lateral(b)
Average 50 ft-ib(')
Average Energy Absorption(.)
Material Transition Temperature (°F)
Expansion Temperature ('F)
Transition Temperature ('F) at Full Shear (ft-lb)
Unirradiated Irradiated AT Unirradiated Irradiated AT Unirradiat&d Irradiated AT Unirradiated Irradiated AE Lower Shell 7.25 95.18 87.93 35,58 131.61 96.03 35.23 130.62 95.38 139 110
-29 Plate C-8-2 (Longitudinal)
Lower Shell 15.37 100.37 84.99 44.98 136.08 91.1 46.01 143.57 97.55 103 88
-15 Plate C-8-2 (Transverse)
WeldMetal
-57.99 10.01 68
-32.8 28.64 61.44
-35.6 31.58 67.19 144 110
-34 HAZ Metal
-79.66
-4.87 74.79
-22.74 42,01 64.76
-37.36 44.91 82.27 133 93
-40 "Average" is defined as the value read from the curve fit through the data points of the Charpy tests (see Figures 5-1, 5-4, 5-7 and 5-10).
"Average" is defined as the value read from the curve fit through the data points of the Charpy tests (see Figures 5-2, 5-5, 5-8 and 5-11)
Analysis of St. Lucie Unit 1 Capsule 2840
- a.
b.
5-15 Table 5-10 Comparison of the St. Lucie Unit 1 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decrease with Regulatory Guide 1.99, Revision 2, Predictions Material Capsule Fluence 30 ft-lb Transition Upper Shelf Energy (x 1019 n/cm 2) (a)
Temperature Shift Decrease Predicted Predicted Measured Predicted Measured (OF) (d)
(OF) (e)
(OF)
(%)(b, d)
(%)
Lower Shell Plate 970 0.591 88.6 65.3 68.7 21 23 C-8-2 1040 0.918 105.8 78.0 79.87 24 17 (Longitudinal) 2840 1.45 119.2 87.9 87.93 27 21 Lower Shell 970 0.591 88.6 65.3 63.83 21 24 Plate C-8-2 1040 0.918 105.8 78.0 N/A (c) 24 N/A (c)
(Transverse) 2840 1.45 1.19.2 87.9 84.99 27 15 Surveillance Program 970 0.591 87.1 57.7 72.34 32 31 Weld Metal 1040 0.918 104.0 68.9 67.4 37 25 2840 1.45 117.3 77.7 68.0 42 24 Heat Affected Zone 970 0.591 19.48 14 Material 1040 0.918 59.8 22 2840 1.45 74.79 30 Notes:
(a)
Calculated Fluences from capsule 2840 dosimetry analysis results (E > 1.0 MeV)
(b)
From Figure 2 of Regulatory Guide 1.99, Revision 2, using the Cu wt. Percent and capsule fluence values. See note (d).
(c)
No Transverse Material in Capsule 1040 (d)
The Lower Shell Plate weight percent copper/nickel was 0.15 and 0.57, while the surveillance weld weight percent copper/nickel was 0.23 and 0.07. The copper and nickel values were used to determine the chemistry factor, which in turn is used to calculate the predicted ARTNDT.
(e)
Based on Reg. Guide 1.99, Rev. 2, methodology using chemistry factor calculated from Surveillance data (Plate CF = 79.9, Weld CF = 70.6).
See credibility evaluation in Appendix D.
Analysis of St. Lucie Unit 1 Capsule 2840
5-16 Sample Test 0.2% Yield Ultimate Fracture Fracture Fracture Uniform Total Reduction Number Material Temperature Strength Strength Load Stress Strength Elongation Elongation in Area (F)
(ksi)
(ksi)
(kip)
(ksi)
(ksi)
(%)
(%)
(%)
1J4 PLATE 125 82.5 103.3 3.45 173.6 70.2 11.2 23.6 60 IJL PLATE 250 77.4 98.4 3.23 192.8 65.7 10.7 23.1 66 IJM PLATE 550 73.3 98.6 3.58 167.4 72.9 10.7 22.2 56 3J2 WELD 35 82.5 97.4 3.08 202.9 62.7 12.9 28.4 69 3JJ WELD 150 78.9 91.6 2.90 173.2 59.1 12.0 25.9 66 3JY WELD 550 79.5 93 3.54 176.2 72.2 11.8 22.7 59 4KJ HAZ 72 83 99.4 3.33 193.6 67.9 9.0 30.6 65 4KK HAZ 225 75.4 91.5 3.14 166.6 64 7.8 27.0 62 4KY HAZ 550 74.9 94.4 3.48 113.1 71 10.1 13.8 37 Analysis of St. Lucie Unit 1 Capsule 284' Table 5-11 Tensile Specimens From Lower Shell Course Plate C-8-2, Weld, and Heat Affected Zone Material
5-17 LOWER SHELL C-8-2 (LONGITUDINAL)
CVGRAPH 4.1 Hyperbolic Tangent Curve Printed at 10"2:59 on 05-12-2000 Results Curve Fluence
[SE d-LSE USE d-USE T o 30 d-T o 30 T o 50 d-T o 50 219 0
139 0
219 0
107
-32 219 0
116
-23 Z19 0
110
-29 725 75.95 97.12 95.18 0
3523 68.7 114.91 79.87 1=62 87.93 130.62 0
79.67 8728 9528 U) b.
-300
-200
-100 0
100 200 300 400 500 600 Temperature in Degrees F Curve Legend 3 e 1i-4-
Curve Plant Casule 1
2 3
4 Data Set(s) Plotted Material SLI UNIRR PLATE SA533MI SLI W-97 PLATE SA533BI SLI W-104 PLATE SA533BI SLI W-284 PLATE SA533BI Figure 5-1 Charpy V-Notch Impact Energy vs. Temperature for St. Lucie Unit 1 Reactor Vessel Lower Shell Plate C-8-2 (Longitudinal Orientation)
Analysis of St. Lucie Unit 1 Capsule 284' 1
2 3
4 0
o 0
0 Or.
Beat/#
LT C-8-2 LT C-8-2 LT C-8-2 LT C-8-2 5-17 Cuv Plant..
.Casle-
LOWER SHELL C-8-2 (LONGITUDINAL)
CVGRAPH 4.1 Hyperbolic Tangent Curve Printed at 10:1811 on 0-12--00 Results Curve Fluence USE d-USE T o LE35 d-T o IM35 1
2 3
4 0
0 0
0 87.47 80.41 96.66 79.13 0
-7.06 9-9
-.-33 3558 1119 10839 131.61 0
75.6 72B3 96.03
-300
-200
-100 0
100 200 300 400 50 60 Temperature in Degrees F Curve Legend 2
Data Set(s) Plott Curve Plant Causule UNIRR W-97 W-104 Y-284 Material PLATE SA533B1 PLATE SA533B1 PLATE SA533B1 PLATE SA533BI 3
Ha ted Oni.
Heatt#
LT LT LT LT C--8-2 C-8-2 C-8-2 C-8-2 Figure 5-2 Charpy V-Notch Lateral Expansion vs. Temperature for St. Lucie Unit 1 Reactor Vessel Lower Shell Plate C-8-2 (Longitudinal Orientation)
Analysis of St. Lucie Unit I Capsule 2840 10--
4,
Ski SLi SIA SIA 4
3 4
Curve Plant CaDsule
5-19 LOWER SHELL C-8-2 (LONGITUDINAL)
CVGRAPH 4.1 Hyperbolic Tangent Curve Printed at 10:46'22 on 05-12-2000 r1nrvo FlhlpnrI'p 2
3 4
0 0
0 0
Results T o 5frz Shenr Curve Fluece T 50xShea 69.37 153.58 125.97 154.85
-3W0
-200
-100 0
100 200 3W0 400 Temperature in Degrees F 10-Curve Legend 3 e-d-T
- 50,/ Shear u
842 56.6 85.47 5M0 600 4-Data Set(s) Plotted Material PLATE SA533BI PLATE SA533BI PLATE SAM533B PLATE SA533BI Figure 5-3 Charpy V-Notch Percent Shear vs. Temperature for St. Lucie Unit 1 Reactor Vessel Lower Shell Plate C-8-2 (Longitudinal Orientation)
Analysis of St. Lucie Unit 1 Capsule 2840 Q~)
4cn C) 1 2
3 4
SLI SLI SL1 SL1 UNIRR Yf-97 W-104 W-284 Ori.
Heat#
Curve~L Plnt-aliMaera LT LT LT LT C-8-2 C--8-2 C--8-2 f'nrvo Pl*nl l*'[*'1fl e
5-20 Figure 5-4 Charpy V-Notch Impact Energy vs. Temperature for St. Lucie Unit 1 Reactor Vessel Lower Shell Plate C-8-2 (Transverse Orientation)
Analysis of St. Lucie Unit I Capsule 2840
5-21 LOWER SHELL C-8-2 (TRANSVERSE)
CVGRAPH 41 Hyperbolic Tangent Curve Printed at 1055:16 on 06-12-2000 Results Curve Fluence USE d-USE T o LE35 d-T o LE35 0
0 0
72.66 0
4418 66.98
-5.68 125.42 67.81
-4.84 136.08 0
80.44 91-0.)
-300
-200 10-
-100 0
100 200 300 400 500 600 Temperature in Degrees F Curve Legend D
20-Sets Pl 30 Data Set(s) Plotted riirvp Planf
('*renile Material Ori.
Heats/
UNIRR PLATE SA533BI W-97 PLATE SA533BI W-284 PLATE SA533BI Figure 5-5 Charpy V-Notch Lateral Expansion vs. Temperature for St. Lucie Unit 1 Reactor Vessel Lower Shell Plate C-8-2 (Transverse Orientation)
Analysis of St. Lucie Unit I Capsule 2840 1
2 3
1 2
3 SI' SL1 SLI TL TL TL C-B--2 C-8-2 C-8-2 5-21 Curve Plant Carmile Ori.
Heat#
-J-k Figure 5-6 Charpy V-Notch Percent Shear vs. Temperature for St. Lucie Unit 1 Reactor Vessel Lower Shell Plate C-8-2 (Transverse Orientation)
Analysis of St. Lucie Unit 1 Capsule 284'
SURVEILLANCE WELD CVGRAPH 4.] Hyperbolic Tangent Curve Printed at 11:10:14 on 05-12-20M0 Results Curve Fluence ISE d-1.SE USE d-USE T o 30 d-T o 30 T o 50 d-T o 50 1
0 Z19 0
144 0
-57.99 0
-35.6 0
2 0
2.9 0
100
-44 1435 72.34 42,73 78.33 3
0 2.19 0
108
-36 9.41 67.4 39.05 75.45 4
0 2.19 0
110
-34 10.01 68 3128 67.19
-30o
-200
-100 0
100 200 300 400 5*0 6o0 Temperature in Degrees F 1 0-Curve Legend 3 e--
4 -
Data Set(s) Plotted Curve Plant Capsule Material Oni.
Heat#
1 SUI UNIRR WELD 90136,LNDE 0091LOT 3999 2
SLi W-97 WELD 901360DNUE 0091WM13999 3
SLI W-104 WELD 90136,JNDE 0091JUT 3999 4
SUi W-284 WELD 90136LINDE 0091lOT999 Figure 5-7 Charpy V-Notch Impact Energy vs. Temperature for St. Lucie Unit 1 Reactor Vessel Surveillance Weld Material Analysis of St. Lucie Unit I Capsule 284' 5-23 cn
- 0 z3
5-24 SURVEILLANCE WELD CVGRAPH 41 Hyperbolic Tangent Curve Printed at 12M33:42 on 05-12-2000 Curve Fluence 1
2 3
4 0
0 0
0 Reults USE d-USE T o LE5 d-T o LE35 9Z06 76.62 8713 7729 0
-15.44
-493
-14.6
-32.8 40M3 2319 28.54 Curve luen0 0
73.63 55.99 61.44
-300
-200
-100 0
100 20o 300 400 5o0 600 Temperature in Degrees F Curve Legend to-3 4
Data Set(s) Plotted Material WELD WELD WELD Ori.
Heat#
90136,LINDE O091UJ 3999 90136,LINDE O093I999 90136MNDE 0091LOT 3999 90136,LINDE 091GT999 Figure 5-8 Charpy V-Notch Lateral Expansion vs. Temperature for St. Lucie Unit I Reactor Vessel Surveillance Weld Metal Analysis of St. Lucie Unit I Capsule 284' C,)
-4
-4 2
3 4
Eui SuI Sid EuI UNIRR W-97 W-104 WI-294 Curve Plant Camile Prvo Plnn!
f.arau]e
5-25 SURVEILLANCE WELD CVGRAPH 41 Hyperblic Tangent Curve Printed at 1244:55 on 05-12-2000 Results Curve Fluence 0
0 0
0 T o 5Nz Shear
-1312 61.4 30.93 3953 d-T o 50Y. Shear 0
74.53 44.06 52.65
-300
-200
-100 0
100 200 300 400 500 600 Temperature in Degrees F Curve Legend 20"....
30 4-*
1 D-Curve Planit Cla'rnle 1
2 3
4 SLI SLI Sid UNIRR 1-97 W-104 W-234 Data Set(s) Plotted Mlaterial WELD WELD WELD If ELD Oi.
Heat#
90136JINDE M091NOT 3999 9013,L1NDE 0091JO999 90136,UNDE 0091JXT 3999 90136,UNDE 00910999 Figure 5-9 Charpy V-Notch Percent Shear vs. Temperature for St. Lucie Unit 1 Reactor Vessel Surveillance Weld Metal Analysis of St. Lucie Unit 1 Capsule 284' 1
2 3
4
- _)
4-'
Q)
C)
Curve Fluence Curve Plant CaTm)le Material
Figure 5-10 Charpy V-Notch Impact Energy vs. Temperature for St. Lucie Unit 1 Reactor Vessel Heat Affected Zone Material Analysis of St. Lucie Unit I Capsule 2840
ý)-Z()
Charpy V-Notch Lateral Expansion vs. Temperature for St. Lucie Unit 1 Reactor Vessel Heat Affected Zone Material Analysis of St. Lucie Unit 1 Capsule 284' 5-27 Figure 5-11
5-28 Figure 5-12 Charpy V-Notch Percent Shear vs. Temperature for St. Lucie Unit 1 Reactor Vessel Heat Affected Zone Material Analysis of St. Lucie Unit 1 Capsule 2840
25C, 100TF 26A, 125F 25P. 275°F 25J, 135°F 251, 150°F 253, 175°F 26B, 220°F 24Y, 325 0F "aigure 5-13 Charpy Impact Specimen Fracture Surfaces for St. Lucie Unit 1 Reactor Vessel Lower Shell Plate C-8-2 (Transverse Orientation)
Analysis of St. Lucie Unit 1 Capsule 284o 252, 1 0°F 25M, 50OF 25L, 90-F
'26C, 115°F
14M, 50-F 146, 100°F 164, 120F 157, 135 0F 143, 150°F 117, 200°F 14U, 275 0F Figure 5-14 14E, 325 0F Charpy Impact Specimen Fracture Surfaces for St. Lucie Unit 1 Reactor Vessel Lower Shell Plate C-8-2 (Longitudinal Orientation)
Analysis of St. Lucie Unit I Capsule 284 155, 5°F 136, 25°F 145, 72°F 16D, 80°F V
36L, -10°F 31 D, 100°F 34M, 1507F 32T, 200°F Figure 5-15 353, 250°F Charpy impact Specimen Fracture Surfaces for St. Lucie Unit 2 Reactor Vessel Weld Metal Specimens Analysis of St. Lucie Unit 1 Capsule 284' 33K, 25°F 35J, 300F 367, 40°F
)I C, -50'F 343, 0°F 325. 10"F 317, 15°F
45L, -40°F 45M. 30°F 45U, 100TF 45T, 135 0F 462, 200°F Figure 5-16 464, 250TF Charpy Impact Specimen Fracture Surfaces for St. Lucie Unit 1 Reactor Vessel Heat Affected Zone (HAZ)
Analysis of St. Lucie Unit 1 Capsule 2840 461, 40F 463, 60°F 46D, 72°F 46C, -5°F 46B, I0°F 421, 20-F
5-33 (0C) 0 50 100 150 200 250 300 LEGEND:
0oL UNIRRADIATED
- A IRRADIATED TO A FLUENCE OF 1.45 X 1019 n/cm2 (E>1.OMeV) AT 5500F REDUCTION IN AREA
~2/
-2AC) 100 200 TOTAL ELONGATION
/3 UNIFORM ELONGATION =A I
I 12 300 400 500 TEMPERATURE (*F)
Figure 5-17 Tensile Properties for St. Lucie Unit 1 Reactor Vessel Lower Shell Plate C-8-2 (Longitudinal Orientation)
Analysis of St. Lucie Unit I Capsule 284' 120 110 100 90 80 70 60 50 40 800 700 in
-Y.
C-C-,
I I
I I
I I
I AL ULTIMATE TENSILE STRENGTH A
S0.2%/
YIELD STRENGTH 600 d
0L-1500 400 300
-J IL) 80 70 60 50 40 30 20 10 0
0 600 I
I
5-34 (0 )
0 50 100 150 200 250 LEGEND:
o
- UNIRRADIATED 9A IRRADIATED TO A FLUENCE OF 1.45 X 1019 n/cm2 (E>1.0MeV) AT 550°F I
-4
-J
-4 I
L) 80 70 60 50 40 30 20 10 0
0 100 200 300 400 500 TEMPERATURE (OF)
Figure 5-18 Tensile Properties for St. Lucie Unit 1 Reactor Vessel Weld Metal Analysis of St. Lucie Unit 1 Capsule 2840 300 800 700 120 110 100 90 80 70 60 50 40 I
i I
I I
I A
ULTIMATE TENSILE STRENGTH A &
0.2% YIELD STRENGTH C-1 600 6
0l-500 1400 300 y2 ZREDUCTION IN AREA A
TOTAL ELONGATION N2
-AI_
A UNIFORM ELONGATION I
I I
I I
600
(°C) 0 50 100 150 200 250
,,k.
ULTIMATE TENSILE STRENGTH A
120 110 100 90 80 70 60 50 40 0
,v 2
YI 1 D c I.;,
,,i I..
in LEGEND:
o AUNIRRADIATED IRRADIATED TO A FLUENCE OF 1.45 X 1019 n/cm 2 (E>L.OMeV) AT 550°F 100 200 300 400 500 TEMPERATURE (OF)
Figure 5-19 Tensile Properties for St. Lucie Unit 1 Reactor Vessel Heat-Affected-Zone (HAZ)
Analysis of St. Lucie Unit 1 Capsule 284' 300
-2 iLn y,-
800 700 600 d
0L, 500
-1 400 300 80 70 S.-..
I._.
I-C-)
60 50 40 30 20 10 0
REDUCTION IN AREA
-A A
A TOTAL ELONGATION UNIFORM ELONGATION A
0 600 300 L
A L
.A
Specimen 1J4 Tested at 125°F Specimen 1JL Tested at 250'F Specimen IJMTested at 550'F Fractured Tensile Specimens from St. Lucie Unit 1 Reactor Vessel Lower Shell C-8 2 (Longitudinal Orientation)
Analysis of St. Lucie Unit I Capsule 284' Figure 5-20 5 6
Specimen 3J2 Tested at 35°F Specimen 3JJ Tested at 150OF Specimen 3JY Tested at 550'F Fractured Tensile Specimens from St. Lucie Unit I Reactor Vessel Weld Metal Analysis of St. Lucie Unit 1 Capsule 284' Figure 5-21 5-37,
5-38 Specimen 4KJ Tested at 72°F Specimen 4KK Tested at 225°F Specimen 4KY Tested at 550'F Fractured Tensile Specimens from St. Lucie Unit 1 Reactor Vessel Heat-Affected Zone (HAZ)
Analysis of St. Lucie Unit I Capsule 284' Figure 5-22
5-39 STRESS-STRAIN CURVE ST. LUCIE UNIT 1 284 CAPSULE 100 80 1JL 250 F 0
0.05 0.1 0.15 0.2 0.25 0.3 STRAIN, IN/IN STRESS-STRAIN CURVE ST. LUCIE UNIT 1 CAPSULE 284 1JM 550 F 0.05 0.1 0.15
- STRAIN, INIIN 0.2 0.25 0.3 Engineering Stress-Strain Curves for Lower Shell Plate C-8-2 Tensile Specimens 1J4, 1JL and 1JM (Longitudinal Orientation)
Analysis of St. Lucie Unit 1 Capsule 2840 1J4 125 F 0.05 0.1 0.15 0.2 0.25 0.3 STRAIN, INIIN STRESS-STRAIN CURVE ST. LUCIE UNIT 1 CAPSULE 284
,6 UJ I,
U,)
60 40 20 0
100 90 80 70 60 5o 40 30 20 10 0
100 90 80 70 60 50 40 30 20 10 0
0 Figure 5-23
5-40 STRESS-STRAIN CURVE ST LUCIE UNIT 1 284 CAPSULE 100 90 so 70 60 50 40 30 20 10 0
0.05 01 0.15 STRAIN, IN/IN 0.2 0.25 0.3 STRESS-STRAIN CURVE ST. LUCIE UNIT 1 CAPSULE 284 3JJ 150 F 0.05 0.1 0.15 STRAIN. IN/IN STRESS-STRAIN CURVE ST. LUCIE UNIT 1 CAPSULE 284 3JY 550 F 0.1 0.15 STRAIN, IN/IN Figure 5-24 Engineering Stress-Strain Curves Weld Metal Tensile Specimens 3J2, 3JJ, and 3JY Analysis of St. Lucie Unit 1 Capsule 284'
.o" C~o w
I Qo 3J2 35 F 0
Co CO w
100 90 80 70 60 50 40 30 20 10 0.2 0.25 0.3 100 90 80 70 60 50 40 30 20 10 0
c6 Co Co 0
0.05 0.2 0.25 0.3
5-41 STRESS-STRAIN CURVE ST. LUCIE UNIT 1 CAPSULE 284 100 90 80 70 60 50 40 30 20 10 0
0.05 0.1 0.15 0.2 0.25 STRAIN, INIIN 0.3 STRESS-STRAIN CURVE ST.LUCIE UNIT 1 284 CAPSULE 4KK 225 F 0
0.05 0.1 0.15 STRAIN, INIIN STRESS-STRAIN CURVE ST. LUCIE UNIT 1 CAPSULE 284 4KY 550 F 0.1 0.15
- STRAIN, INIIN Figure 5-25 Engineering Stress-Strain Curves for Heat-Affected-Zone (HAZ) Material Tensile Specimens 4KJ, 4KK, and 4KY Analysis of St. Lucie Unit 1 Capsule 2840 09 09 w
I0 09 4KJ 72 F 0
09 w
n,-
100 90 1 80 70 60 50 40 30 20 10 0
L 0.2 0.25 0.3
£3 09 09 100 9D 80 70 60 50 40 30 20 10 0
0.05 0.2 0.25 0.3
6-1 6
RADIATION ANALYSIS AND NEUTRON DOSIMETRY
6.1 INTRODUCTION
Knowledge of the neutron environment within the reactor vessel and surveillance capsule geometry is required as an integral part of LWR reactor vessel surveillance programs for two reasons. First, in order to interpret the neutron radiation induced material property changes observed in the test specimens, the neutron environment (energy 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 reactor 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 (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 for 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 neutron energy 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 reactor vessel wall.
Because of this potential shift away from a 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 fluence (E > 1.0 MeV) to provide a data base for future reference. The energy dependent dpa function to be used for this evaluation is specified in ASTM Standard Practice E693, "Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements per Atom." The application of the dpa parameter to the assessment of embrittlement gradients through the thickness of the reactor vessel wall has already been promulgated in Revision 2 to Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials."
This section provides the results of the neutron dosimetry evaluations performed in conjunction with the analysis of test specimens contained in the 97', 104', and 284' surveillance capsules which were withdrawn at the end of the fifth, ninth, and fifteenth fuel cycles, respectively. This evaluation is based on current state-of the-art methodology and nuclear data including neutron transport and dosimetry cross-section libraries derived from the ENDF/B-VI data base. This report provides a consistent up-to-date neutron exposure data base for use in evaluating the material properties of the St. Lucie Unit I reactor vessel.
In each capsule dosimetry evaluation, fast neutron exposure parameters in terms of neutron fluence (E > 1.0 MeV), neutron fluence (E > 0.1 MeV), and iron atom displacements (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 used to project the integrated exposure of the vessel wall. Also, uncertainties associated with the derived exposure parameters at the surveillance capsules and with the projected exposure of the reactor vessel are provided.
Analysis of St. Lucie Unit I Capsule 2840
6-2 All of the calculations and dosimetry evaluations presented in this section have been based on the latest available nuclear cross-section data derived from ENDF/B-VI and the latest available calculational tools and are consistent with the requirements of Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence." Additionally, the methods used to develop the best estimate pressure vessel fluence are consistent with the NRC approved methodology described in WCAP-14040-NP A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," January 1996.
6.2 DISCRETE ORDINATES ANALYSIS A plan view of the reactor geometry at the core midplane is shown in Figure 4-1. Six irradiation capsules attached to the reactor vessel wall are included in the reactor design to constitute the reactor vessel surveillance program. The capsules are located at azimuthal angles of 83', 970, 1040, 263', 277', and 2840 relative to the core cardinal axis as shown in Figure 4-1.
A view of a surveillance capsule assembly is shown in Figure 4-2. A total of seven stainless steel specimen containers hold the charpy and tensile monitors. The assembly is positioned axially centered on the core midplane, thus spanning the central portion of the active fuel zone.
From a neutronic standpoint, the surveillance capsules and associated support structures are significant. The presence of these materials has a marked effect on both the spatial distribution of neutron flux and the neutron energy spectrum in the water annulus and on the inside of the vessel wall near the capsule locations.
In order to determine the neutron environment at the test specimen location, the capsules themselves must be included in the analytical model.
In performing the fast neutron exposure evaluations for the surveillance capsules and reactor vessel, several transport calculations were needed to accommodate changes in geometry, loading patterns, and downcomer temperature. Each is described in the table below.
Cycles 1-4:
Out-In Loading Pattern, Thermal Shield in Place Cycle 5:
In-Out Loading Pattern at Uprated Power Cycles 6-9:
Thermal Shield Permanently Removed Cycles 11,13:
Fresh Fuel Assemblies Containing full length Hafniium rods in Peripheral Locations Cycles 12,14:
Once Burned Fuel Assemblies Containing full length Hafnium rods in Peripheral Locations Cycles 10,15:
No Hafnium rods in Peripheral Assemblies These calculations were combined to produce average relative neutron energy distributions throughout the reactor geometry as well as to establish relative radial distributions of exposure parameters {4i(E > 1.0 MeV),
4(E > 0.1 MeV), and dpa/sec} through the vessel wall integrated over time. The neutron spectral information was required for the interpretation of neutron dosimetry withdrawn from the surveillance capsules as well as for the determination of exposure parameter ratios, i.e., [dpa/sec]/[f(E > 1.0 MeV)], within the reactor vessel Analysis of St. Lucie Unit 1 Capsule 284'
6-3 geometry. The relative radial gradient information was required to permit the projection of measured exposure parameters to locations interior to the reactor vessel wall, i.e., the 1/4/T and %,T locations.
Two-dimensional rO forward transport calculations for the reactor models were carried out using the DORT two-dimensional discrete ordinates code Version 3. 1131 and the BUGLE-96 cross-section library [14]. The BUGLE-96 library is a 47 energy group ENDF/B-VI based data set produced specifically for light water reactor applications. In these analyses, anisotropic scattering was treated with a P3 expansion of the scattering cross-sections and the angular discretization was modeled with an S order of angular quadrature.
The core power distribution utilized in the reference forward transport calculations were generated from input relative pin-by-pin and assembly power data through the SORCERY program. Cycle specific axial power distributions provided axial peaking factors for the core midplane and peak vessel locations that were applied to the results of the forward calculations.
Selected results from the neutron transport analyses are provided in Tables 6-1 through 6-5. The data listed in these tables establish the means for absolute comparisons of analysis and measurement for the 970, 1040, and 2840 capsule irradiation periods and provide the means to correlate dosimetry results with the corresponding exposure of the reactor vessel wall.
In Table 6-1, the calculated exposure parameters [ý(E > 1.0 MeV), 4i(E > 0.1 MeV), and dpa/sec] are given at the geometric center of the two 140 surveillance capsule positions (1040 and 2840) and for the 70 capsule position (970) for the cycle specific core power distributions. Similar data are given in Table 6-2 for the reactor vessel inner radius. It is important to note that the data for the vessel inner radius were taken at the clad/base metal interface, thus representing the maximum predicted exposure levels of the vessel plates and welds.
Radial gradient information applicable to 4(E > 1.0 MeV), 4(E > 0.1 MeV), and dpa/sec is given in Tables 6 3, 6-4, and 6-5, respectively. The data obtained from the reference forward neutron transport calculations are presented on a relative basis for each exposure parameter at key 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 Tables 6-3 through 6-5.
For example, the neutron flux 4(E > 1.0 MeV) at the 1/4AT depth in the reactor vessel wall along the 00 azimuth is given by:
,114T (00) = 0(221.55, 00) F(227.03,00) where:
(t/.T(0O) =
Projected neutron flux at the 1/4T position on the 00 azimuth.
ý(221.55,00) = Projected or calculated neutron flux at the vessel inner radius on the 0' azimuth.
F(227.03,0-) = Ratio of the neutron flux at the 'AT position to the flux at the vessel inner radius for the 00 azimuth. This data is obtained from Table 6-3.
Similar expressions apply for exposure parameters expressed in terms of ý(E > 0. 1 MeV) and dpa/sec where the attenuation function F is obtained from Tables 6-4 and 6-5, respectively.
Analysis of St. Lucie Unit I Capsule 2840
6-4 6.3 NEUTRON DOSIMETRY The passive neutron sensors included in the St. Lucie Unit I surveillance program are listed in Table 6-6.
Also given in Table 6-6 are the primary nuclear reactions and associated nuclear constants that were used in the evaluation of the neutron energy spectrum within the surveillance capsules and in the subsequent determination of the various exposure parameters of interest [O(E > 1.0 MeV), 4(E > 0.1 MeV), dpa/sec]. The relative locations of the neutron sensors within the capsules are shown in Figure 4-4.
The use of passive monitors such as those listed in Table 6-6 does not yield a direct measure of the energy dependent neutron flux at the point of interest. Rather, the activation or fission 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, and The neutron energy spectrum at the monitor location.
The specific activity of each of the neutron monitors was determined using established ASTM procedures[( 6 hrou 281 Following sample preparation and weighing, the activity of each monitor was determined by means of a lithium-drifted germanium, Ge(Li), gamma spectrometer. The irradiation history of the St. Lucie Unit I reactor was obtained from plant personnel[2 9 -d 30] and data reported in NUREG-0020, "Licensed Operating Reactors Status Summary Report," for the Cycles 1 to 15 operating periods. The irradiation history applicable to the exposure of the 970, 1040, and 2840 capsules is given in Table 6-7.
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 were determined from the following equation:
A R=
Pj N0 F Y E Pr* Cj [l-e-tJ] [e"*td]
where:
R Reaction rate averaged over the irradiation period and referenced to operation at a core power level of Prf (rps/nucleus).
A
=
Measured specific activity (dps/gm).
No
=
Number of target element atoms per gram of sensor.
F Weight fraction of the target isotope in the sensor material.
Y
=
Number of product atoms produced per reaction.
Analysis of St. Lucie Unit I Capsule 284'
6-5 Pj
=
Average core power level during irradiation period j (MW).
Pref =
Maximum or reference power level of the reactor (MW).
Cj
=
970 Capsule:
Calculated ratio of 4(E > 1.0 MeV) during irradiation periodj to the time weighted average u(E > 1.0 MeV) over the entire irradiation period.
1040 and 2840 Capsules:
Determined as the ratio of the calculated reaction rate during periodj to the spectrum average reaction rate over the irradiation period for each reaction.
This was done to account for the spectrum changes due to removal of the thermal shield.
X
=
Decay constant of the product isotope (1/sec).
tj
=
Length of irradiation period j (sec).
td
=
Decay time following irradiation period j (sec).
and the summation is carried out over the total number of monthly intervals comprising the irradiation period.
In the equation describing the reaction rate calculation, the ratio [PjI/[P~ef] accounts for month-by-month variation of reactor core power level within any given fuel cycle as well as over multiple fuel cycles. The ratio Cj, which can be calculated for each fuel cycle, accounts for the change in sensor reaction rates caused by variations in flux level induced by changes in core spatial power distributions from fuel cycle to fuel cycle.
The actual Cj values that were used to assess the sensor reaction rates from the 97', 104', and 2840 St. Lucie Unit 1 surveillance capsules are presented in Table 6-16.
These Cj values are typically derived from the information provided in Table 6-1 and they are determined on a fuel cycle-specific basis in order to account for core spatial power distribution differences in each cycle. The rationale behind this approach is that the impact of changing flux levels for constant power operation can be quite significant for sensor sets that have been irradiated for many cycles in a reactor that has transitioned from non-low leakage to low leakage fuel management, for sensor sets contained in surveillance capsules that have been moved from one capsule location to another, or when major changes to the reactor internals occur, such as removal of the thermal shield. This latter effect may be observed from the Cj data that has been provided in Table 6-16. Furthermore, since the neutron transport calculations are steady-state approximations of each fuel cycle, the radial power distribution change is not modeled explicitly. Hence, an average assembly power is used for the calculation. This is an adequate approximation since the effect of the relative radial power shift in the peripheral assemblies over a fuel cycle is small compared to cycle-to-cycle loading pattern changes.
The cycle-average flux values are used in the calculation of Cj assuming that the neutron energy spectrum at the capsule is unchanged. The irradiation period of the 1040 and the 2840 capsules encompassed the removal of the thermal shield following Cycle 5 which would have a marked change in the neutron energy spectrum at the capsule. The spectrum change affects each of the dosimeters differently since the response functions of the dosimeters are different, as shown below:
Analysis of St. Lucie Unit 1 Capsule 2840
6-6 Monitor Reaction of Detector Material Interest
Response
Copper 63Cu (n,y)
E > 4.7 MeV Titanium 46Ti (n,p)
E > 4.4 MeV Iron 54Fe (n,p)
E > 1.0 MeV Nickel 58Ni (n,p)
E > 1.0 MeV Uranium-238 238U (n,f)
E > 0.4 MeV Neptunium-237 237Np (n,f)
E > 0.08 MeV Cobalt-Al 59Co (n,y)
E > 0.015 MeV To account for the spectrum change for the 1040 and the 2840 capsule irradiations, the C, terms were derived from the calculated reaction rates provided in the cycle-specific neutron transport calculations.
Measured and saturated reaction product specific activities as well as the derived full power reaction rates are listed in Table 6-8. The reaction rates of the 238U sensors provided i Table 6-8 include corrections for 235U impurities, plutonium build-in, and gamma ray induced fissions.
In the determination of the Best Estimate fast neutron exposures of the surveillance capsules, least squares analysis is used to combine the plant-specific calculation and available measurements within the constraints of the respective uncertainties to produce a Best Estimate of the radiation field at each measurement location.
These Best Estimate values have associated uncertainties less than those associated with the input parameters. The overall data base of [Best Estimate]/[Calculation] (BE/C) comparisons is then used to bias the plant-specific calculation to produce Best Estimate values at the pressure vessel wall with an associated uncertainty less than that applicable to the stand-alone calculation.
Least squares adjustment methods provide the capability of combining the measurement data with the neutron transport calculation resulting in a Best Estimate neutron energy spectrum with associated uncertainties. Best Estimates for key exposure parameters such as 4(E > 1.0 MeV) or dpa/s along with their uncertainties are then easily obtained from the adjusted spectrum. The use of measurements in combination with the analytical results reduces the uncertainty in the calculated spectrum and acts to remove biases that may be present in the analytical technique.
In general, the least squares methods, as applied to pressure vessel fluence evaluations, act to reconcile the measured sensor reaction rate data, dosimetry reaction cross-sections, and the calculated neutron energy spectrum within their respective uncertainties. For example,
'5' 0-,g+/- '
)(Og +/-+350 g
relates a set of measured reaction rates, Ri, to a single neutron spectrum, 4 g, through the multigroup dosimeter reaction cross-section, aog, each with an uncertainty 6.
Analysis of St. Lucie Unit I Capsule 2840
6-7 The primary objective of the least squares evaluation is to produce unbiased estimates of the neutron exposure parameters at the location of the measurement. The analytical method alone may be deficient because it inherently contains uncertainty due to the input assumptions to the calculation. Typically these assumptions include parameters such as the temperature of the water in the peripheral fuel assemblies, by pass region, and downcomer regions, component dimensions, and peripheral core source. Industry consensus indicates that the use of calculation alone results in overall uncertaintes in the neutron exposure parameters in the range of 15-20% (1c;).
By combining the calculated results with available measurements, the uncertainties associated with the key neutron exposure parameters can be reduced. Specifically ASTM Standard E 944 states:
"The algorithims of the adjustment codes tend to decrease the variances of the adjusted data compared to the corresponding input values. The least squares adjustment codes yield estimates for the output data with minimum variances, that is, the "best estimates ". This is the primary reason for using these adjustment procedures".
In the current analysis, the FERRET' 311 code was employed to combine the results of plant specific neutron transport calculations and multiple foil reaction rate measurements to determine best estimate values of exposure parameters (4(E > 1.0 MeV) and dpa) along with associated uncertainties at the surveillance capsule measurement locations.
The application of the least squares methodology requires the following input:
1 - The calculated neutron energy spectrum and associated uncertainties at the measurement location.
2 - The measured reaction rate and associated uncertainty for each sensor contained in the multiple foil set.
3 - The energy dependent dosimetry reaction cross-sections and associated uncertainties for each sensor contained in the multiple foil sensor set.
The calculated neutron spectrum is obtained from the results of plant specific neutron transport calculations applicable to the irradiation period experienced by the dosimetry sensor set. This calculation is based on the application of the benchmarked transport calculational methodology using the DORT discrete ordinates transport code. The sensor reaction rates are derived from the measured specific activities obtained from the counting laboratory using the specific irradiation history of the sensor set to perform the radioactive decay corrections. The dosimetry reaction cross-sections and uncertainties are obtained from the SNLRML dosimetry cross-section libraryt33]. The SNLRML library is an evaluated dosimetry reaction cross-section compilation recommended for use in LWR evaluations by ASTM Standard E10 18, "Application ofASTM Evaluated Cross-Section Data File, Matrix E 706 (1iB)". There are no additional data or data libraries built into the FERRET code system. All of the required input is supplied externally at the time of the analysis.
The uncertainties associated with the measured reaction rates, dosimetry cross-sections, and calculated neutron spectrum are input to the least squares procedure in the form of variances and covariance's. The assignment of the input uncertainties also follows the guidance provided in ASTM Standard E 944.
The following provides a summary of the uncertainties associated with the least squares evaluation of the surveillance capsule dosimetry sets.
Analysis of St. Lucie Unit 1 Capsule 2840
6-8 Reaction Rate Uncertainties The overall uncertainty associated with the measured reaction rates includes components due to the basic measurement process, the irradiation history corrections, and the corrections for competing reactions. A high level of accuracy in the reaction rate determinations is assured by utilizing laboratory procedures that conform to ASTM National Consensus Standards. In all cases, the latest available versions of the applicable standard are used in the dosimetry evaluations.
From these standards, it is noted that the achievable uncertainties in the measured of the sensors comprising typical multiple foil sensor sets are as follows:
Reaction 63Cu(n,a)6°Co 46Ti(n,p)46Sc 54Fe(n,p) 54aM 58Ni(n,p) 58Co 238U(nf)FP 59Co(n,y)Co Precision 1%
1%
1%
1%
1%
1%
These uncertainties include the impacts of counting statistics, sample source/detector geometry corrections, and product nuclide branching ratios.
specific activities of each Bias 3%
3%
3%
3%
5%
5%
weighing, detector calibration, In determining reaction rates from the measured specific activities, the following additional uncertainties are incurred:
Fission Reaction 63Cu(n,a)6°Co 46Ti(n,p)46Sc 54Fe(n,p) 54Mvn 58Ni(nOp)SCo 238U(n,f)FP "59Co(nY) 6°Co Yield 1%
Product Half-Life 0.02%
0.2%
0.2%
0.2%
0.1%
0.02%
Competing Reactions 4%
Analysis of St. Lucie Unit I Capsule 2840
6-9 After combining all of these uncertainty components, the sensor reaction rates derived from the counting and data evaluation procedures result in the following net uncertainties associated with the sensor reaction rates that are input to the least squares evaluation:
Reaction Rate Reaction 63Cu(n,)6°Co
¶Ti(n,p)VSc
'4Fe(np)' 4Mn 58Ni(n,p) 58Co "238U(n,f)FP 59Co(n Y)60CO Uncertainty 5%
5%
5%
5%
10%
5%
The listed uncertainty values are at the Iet level.
In addition to the use of ASTM National Consensus Standards in the evaluation of sensor reaction rates, these procedures have been periodically tested via round robin counting exercises included as a part of the NRC Sponsored Light Water Reactor Surveillance Dosimetry Improvement Program (LWR-SDIP) as well as by evaluation of fluence counting standards provided by the National Institute of Science and Technology (NIST).
Dosimetry Cross-Section Uncertainties The reaction rate cross-sections used in the neutron fluence evaluations are taken from the SNLRML library[33). This data library provides reaction cross-sections and associated uncertainties, including covariances, for 66 dosimetry sensors in common use. Both cross-sections and uncertainties are provided in a fine multigroup structure for use in least squares adjustment applications. These cross-sections were compiled from the most recent cross-section evaluations and they have been tested with respect to their accuracy and consistency for least squares evaluations. Further, the library has been empirically tested for use in fission spectra determination as well as in the fluence and energy characterization of 14 MeV neutron sources.
Detailed discussions of the contents of the SNLRML library along with the evaluation process for each of the sensors is provided in Reference 33.
For sensors of interest to LWR dosimetry applications, the following uncertainties in the fission spectrum averaged cross-sections are provided in the SNLRML documentation package.
Reaction 63Cu(n,a)60Co 46Ti(n,p)46Sc 54Fe(n,p)SnMn "58Ni(n,p)58Co 23'U(n,f)FP 59Co(n,y)6°Co Uncertainty 4.08-4.16%
4.514.87%
3.05-3.11%
4.49-4.56%
0.54-0.64%
0.79-3.59%
Analysis of St. Lucie Untit 1 Capsule 284'
6-10 These tabulated ranges provide an indication of the dosimetry cross-section uncertainties associated with typical sensor sets used in LWR irradiations.
Calculated Neutron Spectrum The neutron spectrum input to the least squares adjustment procedure is obtained directly from the results of plant specific transport calculations for each sensor location. The spectrum at each location is input in an absolute sense (rather than as simply a relative spectral shape). Therefore, within the constraints of the assigned uncertainties, the calculated data are treated equally with the measurements.
While the uncertainties associated with the reaction rates are obtained from the measurement procedures and counting benchmarks and the dosimetry cross-section uncertainties are supplied directly with the SNLRML library, the uncertainty matrix for the calculated spectrum is constructed from the following relationship:
M9,
= R+/- Rg Rg,'P where R. specifies an overall fractional normalization uncertainty and the fractional uncertainties, Rg and Rg,,
specify additional random groupwise uncertainties that are correlated with a correlation matrix given by:
P
[I -
[]
5gg. +
+/-
e-H where H= (g-g')
2 2y2 The first term in the correlation matrix equation specifies purely random uncertainties, while the second term describes the short range correlation's over a group range y (0 specifies the strength of the latter term). The value of 8 is 1.0 when g = g' and 0.0 otherwise. For the trial spectrum used in the current evaluations, a short range correlation ofy = 6 groups was used. This choice implies that neighboring groups are strongly correlated when 0 is close to 1. Strong long-range correlation's (or anti-correlation's) were justified based on information presented by R. E. Maerker 341. The uncertainties associated with the measured reaction rates included both statistical (counting) and systematic components. The systematic component of the overall uncertainty accounts for counter efficiency, counter calibrations, irradiation history corrections, and corrections for competing reactions in the individual sensors.
The set of parameters defining the input uncertainties for the calculated spectrum is as follows:
Flux Normalization Uncertainty (R&)
15%
Flux Group Uncertainties (Rg, Rg,)
(E > 0.0055 MeV) 15%
(0.68 eV < E < 0.0055 MeV) 29%
(E < 0.68 eV) 52%
Analysis of St. Lucie Unit 1 Capsule 2840
6-11 Short Range Correlation (0)
(E > 0.0055 MeV) 0.9 (0.68 eV < E < 0.0055 MeV) 0.5 (E < 0.68 eV) 0.5 Flux Group Correlation Range (y)
(E > 0.0055 MeV) 6 (0.68 eV < E < 0.0055 MeV) 3 (E < 0.68 eV) 2 Having the reaction rate, dosimetry cross-section, and calculated spectrum input with associated uncertainties, the overall least squares evaluation of the data set can be conveniently divided into the following two components:
1 -
A pre-adjustment procedure performed by the SAND [32] module that processes the calculated neutron spectrum and SNLRML dosimetry cross-sections into the 53 energy group structure required by FERRET 2 -
The subsequent application of the least squares algorithm in the FERRET module itself.
The pre-adjustment processing can be summarized as follows:
1 -
The calculated neutron energy spectrum in the BUGLE-96 group structure is input to the SAND module.
2 -
The input spectrum is expanded to 620 energy groups to provide compatibility with the SNLRML dosimetry cross-section library.
3 -
The 620 group spectrum is combined with the dosimetry cross-section library to compute spectrum weighted cross-sections in the 53 energy group structure used in the FERRET module.
4 -
The 620 group spectrum is likewise collapsed to the 53 energy group structure used in FERRET.
The application of this pre-processing procedure allows the fine group dosimetry cross-sections to be spectrally weighted by a calculated spectrum representative of the actual measurement location in the reactor.
This approach, if executed properly, is superior to the use of broad group cross-sections that have been collapsed with an arbitrary spectrum.
The second step in the least squares adjustment procedure may be summarized as follows:
1 -
The 53 group neutron energy spectrum and dosimetry reaction cross-sections output from the SAND module are input to the FERRET module along with the measured reaction rate for each sensor included in the multiple foil set. This input includes uncertainty estimates for the neutron spectrum, the dosimetry cross-sections, and the sensor reaction rates.
Analysis of St. Lucie Unit 1 Capsule 2840
6-12 2 -
The least squares evaluation of the input data is performed by the FERRET module.
3 -
Best Estimate values of neutron exposure parameters [L(E > 1.0 MeV) and dpa/s] along with associated uncertainties are output from the FERRET module.
Results of the least squares evaluation of the 970, 1040, and 2840 capsule dosimetry sets are given in Table 6
- 9. The data summarized in that table include fast neutron exposure evaluations in terms of 'D(E > 1.0 MeV),
ct(E > 0.1 MeV), and dpa. In general, good results were achieved in the fits of the best estimate spectra to the individual measured reaction rates. The measured, calculated, and best estimate reaction rates for each foil reaction are given in Table 6-10. The best estimate neutron spectra from the least squares evaluations of the capsule dosimetry sets are given in Table 6-11 in the FERRET 53 energy group structure.
The resultant uncertainties (I c) associated with the Best Estimate fast neutron exposure of the three St. Lucie surveillance capsules are summarized as follows:
Percent Standard Deviation Quantity 2840 Capsule 1040 Capsule 970 Capsule c1(E > 1.0 MeV) 6 6
6 dpa 6
6 6
c1(E > 0.1 MeV) 9 9
10 It is important to note that the least squares adjustment procedure performed by the FERRET code is limited to the data evaluation at the measurement location. The purpose of this stage of the overall fluence evaluation methodology is to obtain the best estimates of the neutron exposure at the measurement location in terms of 4(E > 1.0 MeV) and dpa as well as to estimate the uncertainty associated with each of these capsule exposures. The FERRET code does not perform an adjustment of the neutron spectrum at the pressure vessel wall.
In Table 6-12, absolute comparisons of the best estimate and calculated fluence at the center of the 970, 1040, and 284' capsules are presented. The results for the dosimetry evaluation (BE/C ratio of 0.96 for (D(E > 1.0 MeV)) are consistent with results obtained from similar evaluations of dosimetry from other reactors using methodologies based on ENDF/B-VI cross-sections.
Analysis of St. Lucie Unit 1 Capsule 2840
6-13 6.4 PROJECTIONS OF REACTOR VESSEL EXPOSURE The best estimate exposure of the St. Lucie Unit 1 reactor vessel was developed using a combination of absolute plant specific transport calculations and all available plant specific measurement data. In the case of St. Lucie Unit 1, the measurement data base contains measurements from the three surveillance capsules discussed in this report.
Combining this measurement data base with the plant-specific calculations, the best estimate vessel exposure is obtained from the following relationship:
(D~BestEst.
K (IDcair where:
0Best Est The best estimate fast neutron exposure at the location of interest.
K The plant specific best estimate/calculation (BE/C) bias factor derived from the surveillance capsule dosimetry data.
ca~c
=
The absolute calculated fast neutron exposure at the location of interest.
A distinction should be made between the Best Estimate/Calculation, or [BE]/[C], ratios and the Measurement/Calculation, or [M]/[C], ratios. In this case, Best Estimate values refer to the combination of calculation and measurement via a least squares adjustment procedure to arrive at the best estimate of the neutron flux (E > 1.0 MeV) with an associated uncertainty. The least squares procedure provides a weighting of calculated and measured input based on the energy response and uncertainty associated with each input parameter. The [BE]/[C] ratios, therefore, represent a comparison of the results of the least squares adjustment with the analytical prediction of the neutron flux (E > 1.0 MeV). The [M]/[C] ratios, on the other hand, provide a direct comparison of actual calculated and measured individual foil reaction rates. Using the
[M]/[C] data, a direct comparison of calculated and measured neutron flux (E > 1.0 MeV) can not be made without a suitable weighting of the individual foil results.
The approach defined in the above equation is based on the premise that the measurement data represent the most accurate plant-specific information available at the locations of the dosimetry; and, further that the use of the measurement data on a plant-specific basis essentially removes biases present in the analytical approach and mitigates the uncertainties that would result from the use of analysis alone.
That is, at the measurement points the uncertainty in the best estimate exposure is dominated by the uncertainties in the measurement process. At locations within the reactor vessel wall, additional uncertainty is incurred due to the analytically determined relative ratios among the various measurement points and locations within the reactor vessel wall.
The use of the bias factors derived from the measurement data base acts to remove plant-specific biases associated with the definition of the core source, actual versus assumed reactor dimensions, and operational variations in water density within the reactor. As a result, the overall uncertainty in the best estimate exposure projections within the vessel wall depends on the individual uncertainties in the measurement process, the uncertainty in the dosimetry location, and in the uncertainty in the calculated ratio of the neutron exposure at the point of interest to that at the measurement location.
Analysis of St. Lucie Unit I Capsule 2840
6-14 The uncertainty in the derived neutron flux for an individual measurement is obtained directly from the results of a least squares evaluation of dosimetry data. The least squares approach combines individual uncertainty in the calculated neutron energy spectrum, the uncertainties in dosimetry cross-sections, and the uncertainties in measured foil specific activities to produce a net uncertainty in the derived neutron flux at the measurement point. The associated uncertainty in the plant specific bias factor, K, derived from the BE/C data base, in turn, depends on the total number of available measurements as well as on the uncertainty of each measurement.
In developing the overall uncertainty associated with the reactor vessel exposure, the positioning uncertainties for dosimetry are taken from parametric studies of sensor position performed as part a series of analytical sensitivity studies included in the qualification of the methodology. The uncertainties in the exposure ratios relating dosimetry results to positions within the vessel wall are again based on the analytical sensitivity studies of the vessel thickness tolerance, downcomer water density variations, and vessel inner radius tolerance. Thus, this portion of the overall uncertainty is controlled entirely by dimensional tolerances associated with the reactor design and by the operational characteristics of the reactor.
For St. Lucie Unit 1, the bias factor for (c(E > 1.0 MeV) was developed as follows:
Best Estimate Calculated (D (E>I.0 MeV)
I)(E>1.0 MeV)
Capsule lnCM2]
% Std Dev
[n/CM21 BE/C 2840 1.53E+19 6
1.45E+19 1.05 1040 8.29E+18 6
9.18E+18 0.90 970 5.52E+18 6
5.91E+18 0.93
[BE]/[C] Bias Factor K 0.96
% standard deviation 8
In regard to the irradiation of these three surveillance capsules, it is noted that the 970 capsule was irradiated with the thermal shield in place for 5 fuel cycles. The 1040 and 2840 capsules were irradiated for 5 fuel cycles with the thermal shield in place and for 3 fuel cycles and 10 fuel cycles, respectively, with the thermal shield removed. A comparison of the BE/C ratios for the 1040 and 2840 capsules relative to the 970 capsule shows a lower BE/C ratio for the 1040 data set and a higher BE/C for the 2840 data set. There is no systematic trend in BE/C that could be attributable to either the removal of the thermal shield or to the earlier power uprate of the St. Lucie reactor. Therefore, the three capsule data set has been treated as a single population for the purpose of determining an overall BE/C. The variability of the BE/C ratios observed at St. Lucie are within the variability observed at other operating PWR's and are consistent within the standard deviation of the best estimate results.
Based on this set of [Best Estimate]/[Calculation] comparisons, the bias factor to be applied to the plant specific calculated neutron flux distributions is 0.96 with an associated standard deviation of 8%. Thus, the Best Estimate fluence at locations within the pressure vessel wall is given by:
(IBestEst 0.96(DciC In a similar manner, bias factors of 0.99 + 6%, and 0.97 +/- 7% were developed for c)(E > 0.1 MeV), and dpa, respectively. The overall uncertainty in the Best Estimate exposure projections for the pressure vessel wall stem primarily from two sources.
Analysis of St. Lucie Unit I Capsule 2840
6-15 1 -
The uncertainty in the average [BE]/[C] normalization factor that is applied to the plant specific transport calculations, and 2 -
The additional analytical uncertainty associated with relating the [BE]/[C] results based on data from the measurement locations to the desired results within the pressure vessel wall.
Uncertainty in the [BE]/[C] bias factor derives directly from the individual uncertainties in the measurement process, in the least squares adjustment procedure, and from the number of data points comprising the overall
[BE]/[C] data base for the reactor being evaluated. The additional positioning uncertainties are taken from analytical sensitivity studies of sensor positioning and pressure vessel wall tolerances performed as a part of the overall benchmarking of the fluence methodology.
Having the uncertainty in the [BE]/[C] bias factor (crK) and the uncertainty in the relative position of the reactor vessel wall relative to the measurement points (cp), the total uncertainty in the Best Estimate neutron exposure at locations within the pressure vessel wall (ar) is determined from the following relationship:
(GT) 2 =-
2 + (
P)2 For the St. Lucie application, the uncertainty in the BE/C bias factor was determined to be GK = 8% and the uncertainty associated with the tolerances in dosimetry positioning, vessel inner radius, and downcomer water temperature were determined to be ar = 6%. Thus, the total uncertainty in the Best Estimate projection of neutron fluence at the pressure vessel wall is CT = 10.2%. This level of uncertainty is well within the 20% la uncertainty required by 10CFR50.61.
Tables 6-13 and 6-14 provide best estimate neutron exposure projections (which includes the bias factor, K) and calculated neutron exposure projections (which do not include K) for the 00, 150, 30', and 450 azimuths at the vessel inner radius and within the vessel wall, respectively. It should be recognized that the Upper Shelf Energy (USE) projections are based on the peak fluence values at 01. In addition, the data that is provided in Table 6-14 is based on both a cI(E > 1.0 MeV) slope and a plant-specific dpa slope through the vessel wall.
Exposure projections for future operation are presented in Tables 6-13 and 6-14 for exposure periods of 25, 32, 35, 48, and 54 EFPY The basis for all future exposure projections, beyond the current operating exposure of 17.23 EFPY, is that the Cycle 15 fuel management (In-out loading pattern and no Hafnium rods in the peripheral fuel assemblies, is representative of the fuel management to be used in all future fuel cycles.
In order to assess RTNDT versus fluence curves, dpa equivalent fast neutron fluence levels for the '/T and 3/4T positions were defined by the relations:
dpa(1/4 T) dpa(3/4 T) 0 (3/4T) = 0(OT) dpa(OT) and 0 143/4T) 0 0(OT) dpa(OT) dp a
0 T)dpa(OT)
Using this approach results in the dpa equivalent fluence values listed in Table 6-14.
In Table 6-15, updated lead factors are listed for each of the St. Lucie Unit 1 surveillance capsules.
Analysis of St. Lucie Unit I Capsule 2840
6-16 Table 6-1 Calculated Fast Neutron Exposure Rates and Iron Atom Displacement Rates at the Surveillance Capsule Center 4(E > 1.0 MeV) (nrcm 2-sec)
Capsule Positions Cycle(s) 1-4 5
6-9 10 11 12 13 14 15 70 4.19E+10 3.64E+ 10 4.69E+10 2.69E+10 2.09E+10 2.44E+10 2.23E+10 2.53E+10 3.22E+10 140 2.96E+10 2.40E+ 10 3.29E+10 2.33E+10 1.93E+10 2.15E+10 2.OOE+10 2.25E+10 2.48E+10
ý(E > 0.1 MeV) (n/cm -sec)
Capsule Positions Cycle(s) 1-4 5
6-9 10 11 12 13 14 15 70 9.5 1E+10 8.32E+10 8.49E+10 4.85E+10 3.75E+10 4.38E+10 3.99E+10 4.54E+10 5.81E+10 140 6.70E+10 5.40E+10 5.91E+10 4.16E+10 3.42E+10 3.81E+10 3.54E+10 3.98E+10 4.43E+10 Iron Atom Displacement Rate (dpa/sec)
Capsule Positions Cycle(s) 1-4 5
6-9 10 11 12 13 14 15 70 6.3 6E-11 5.55E-11 6.8 1E-11 3.9 1E-11 3.05E-I1 3.55E-11 3.25E-11 3.68E-11 4.68E-1I 140 4.50E-11 3.65E-11 4.80E-11 3.40E-11 2.83E-11 3.15E-11 2.93E-11 3.29E-11 3.61 E-11 Analysis of St. Lucie Unit I Capsule 2840
6-17 Table 6-2 Calculated Azimuthal Variation of Fast Neutron Exposure Rates and Iron Atom Displacement Rates at the Reactor Vessel Clad/Base Metal Interface 4i(E > 1.0 MeV) (n/cm2-sec) 00 2.76E+10 2.69E+10 3.70E+10 1.95E+10 1.65E+10 1.92E+10 1.75E+10 1.94E+10 2.36E+10 150 1.70E+10 1.41E+10 2.12E+10 1.56E+10 1.37E+10 1.49E+10 1.40E+10 1.53E+10 1.59E+10 300 1.65E+10 1.21E+10 1.81E+10 1.83E+10 1.94E+10 1.99E+10 1.85E+10 1.86E+10 1.35E+10 450 1.22E+10 8.5 1E+09 1.33E+10 1.40E+10 1.45E+10 1.42E+10 1.41E+10 1.46E+10 1.09E+10
ý(E > 0.1 MeV) (n/cm2-sec) 00 7.03E+10 6.85E+10 8.03E+10 4.25E+10 3.57E+10 4.15E+10 3.80E+10 4.21E+10 5.13E+10 150 4.56E+10 3.78E+10 4.79E+10 3.48E+10 3.03E+10 3.31E+10 3.1OE+10 3.41E+10 3.56E+10 300 4.19E+10 3.06E+10 3.96E+10 3.98E+10 4.22E+10 4.33E+10 4.02E+10 4.04E+10 2.94E+10 450 3.04E+10 2.13E+10 2.88E+10 3.02E+10 3.11E+10 3.06E+10 3.04E+10 3.14E+10 2.35E+10 Iron Atom Displacement Rate (dpa/sec) 00 4.38E-11 4.26E-11 5.64E-11 3.OOE-11 2.53E-11 2.93E-11 2.68E-11 2.97E-11 3.62E-11 150 2.75E-11 2.29E-11 3.29E-11 2.42E-11 2.12E-11 2.32E-11 2.17E-11 2.38E-11 2.46E-11 Analysis of St. Lucie Unit I Capsule 2840 Cycle(s) 1-4 5
6-9 10 11 12 13 14 15 Cycle(s) 1-4 5
6-9 10 11 12 13 14 15 Cycle(s) 1-4 5
6-9 10 11 12 13 14 15 300 2.62E-1 1 1.92E-11 2.79E-11
- 2. 80E-11 2.98E-11 3.05E-11 2.84E-11 2.86E-11 2.07E-11 450 1.94E-11 1.36E-11 2.06E-11 2.17E-11 2.24E-1 1 2.20E-11 2.18E-11 2.25E-I 1 1.69E-11
6-18 Table 6-3 Relative Radial Distribution of ý(E > 1.0 MeV)
Within the Reactor Vessel Wall Azimuthal Angle Radius (cm)j 221.55 222.08 223.12 224.16 225.20 226.25 227.29 228.33 229.38 230.42 231.46 232.51 233.55 234.59 235.64 236.68 237.72 238.77 239.81 240.85 241.89 242.94 244.73 Note:
Base Metal Inner Radius Base Metal '/T Base Metal 1/22T Base Metal %/T Base Metal Outer Radius 00 1.000 0.966 0.879 0.785 0.695 0.612 0.536 0.469 0.409 0.356 0.310 0.269 0.233 0.202 0.174 0.150 0.129 0.111 0.094 0.080 0.067 0.054 0.047 150 1.000 0.966 0.872 0.779 0.693 0.611 0.537 0.470 0.412 0.359 0.314 0.273 0.238 0.205 0.179 0.153 0.133 0.114 0.098 0.084 0.071 0.059 0.051 300 1.000 0.966 0.879 0.785 0.696 0.614 0.539 0.472 0.412 0.359 0.313 0.272 0.236 0.204 0.177 0.153 0.132 0.114 0.097 0.083 0.070 0.058 0.052 221.55 cm 227.03 cm 232.51 cm 237.98 cm 243.46 cm Analysis of St. Lucie Unit I Capsule 2840 450 1.000 0.965 0.876 0.785 0.695 0.613 0.539 0.472 0.412 0.360 0.313 0.273 0.237 0.205 0.178 0.154 0.133 0.115 0.098 0.085 0.072 0.060 0.054
6-19 Table 6-4 Relative Radial Distribution of 4(E > 0.1 MeV)
Within the Reactor Vessel Wall Azimuthal Angle Radius (cm) 221.55 222.08 223.12 224.16 225.20 226.25 227.29 228.33 229.38 230.42 231.46 232.51 233.55 234.59 235.64 236.68 237.72 238.77 239.81 240.85 241.89 242.94 244.73 00 1.000 1.007 0.991 0.958 0.919 0.876 0.831 0.787 0.741 0.697 0.654 0.611 0.569 0.528 0.488 0.449 0.410 0.372 0.335 0.297 0.259 0.219 0.195 Note:
Base Metal Inner Radius Base Metal 'AT Base Metal '1/2T Base Metal 3/T Base Metal Outer Radius 150 1.000 1.008 0.990 0.960 0.925 0.884 0.842 0.799 0.757 0.714 0.673 0.631 0.591 0.550 0.511 0.472 0.434 0.397 0.360 0.324 0.286 0.247 0.223 300 1.000 1.008 0.993 0.962 0.925 0.884 0.841 0.798 0.754 0.711 0.669 0.628 0.587 0.547 0.509 0.471 0.434 0.397 0.361 0.326 0.290 0.254 0.232 221.55 cm 227.03 cm 232.51 cm 237.98 cm 243.46 cm Analysis of St. Lucie Unit 1 Capsule 2840 450 1.000 1.008 0.995 0.967 0.931 0.890 0.850 0.807 0.765 0.723 0.681 0.641 0.602 0.562 0.525 0.487 0.451 0.415 0.380 0.346 0.311 0.276 0.255
Table 6-5 Relative Radial Distribution of dpa/sec Within the Reactor Vessel Wall Azimuthal Angle Radius (cm) 221.55 222.08 223.12 224.16 225.20 226.25 227.29 228.33 229.38 230.42 231.46 232.51 233.55 234.59 235.64 236.68 237.72 238.77 239.81 240.85 241.89 242.94 244.73 00 1.000 0.971 0.899 0.821 0.747 0.678 0.614 0.556 0.503 0.455 0.412 0.373 0.337 0.304 0.274 0.246 0.221 0.197 0.174 0.153 0.132 0.110 0.098 150 1.000 0.972 0.895 0.820 0.750 0.683 0.621 0.564 0.514 0.467 0.425 0.386 0.351 0.318 0.289 0.260 0.235 0.211 0.188 0.167 0.146 0.126 0.113 Note:
Base Metal Inner Radius Base Metal 'AT Base Metal '/T Base Metal /T Base Metal Outer Radius 221.55 cm
=
227.03 cm
=
232.51 cm 237.98 cm
=
243.46 cm Analysis of St. Lucie Unit I Capsule 2840 6-20 300 1.000 0.971 0.899 0.822 0.750 0.682 0.619 0.562 0.510 0.463 0.420 0.381 0.346 0.313 0.283 0.257 0.231 0.208 0.186 0.165 0.146 0.126 0.115 450 1.000 0.970 0.896 0.822 0.748 0.680 0.618 0.561 0.509 0.464 0.420 0.383 0.348 0.315 0.287 0.260 0.236 0.213 0.191 0.171 0.152 0.134 0.124
6-21 Table 6-6 Nuclear Parameters Used in the Evaluation of Neutron Sensors Monitor Material Copper Titanium Iron Nickel Cobalt-Al Uranium-238 Reaction of Interest 63Cu (n,c) 46Ti (n, p)
- Fe (n,p)
"58Ni (n,p)
' 9Co (n,y) 238U (n,f)
Target Atom Fraction 0.6917 0.0825 0.0585 0.6808 0.0015 0.9996
Response
Range E>4.7 MeV E > 4.4 MeV E> 1.0 MeV E > 1.0 MeV non-threshold E> 0.4 MeV Fission Product Yield Half-life
(%)
5.271 y 83.79 d 312.3 d 70.82 d 5.271 y 30.07 y 6.02 Notes:
- 1. Various monitors are cadmium shielded.
- 2.
Target atom fraction for 238U assumed 350 ppm 235U.
Analysis of St. Lucie Unit I Capsule 2840
6-22 Table 6-7 Monthly Thermal Generation During the First Fifteen Fuel Cycles of the St. Lucie Unit 1 Reactor (Reactor Power of 2700 MWt)
Thermal Thermal Thermal Generat.
Generat.
Generat.
Year Month (MW-hr)
Year Month (MW-hr)
Year Month (MW-hr) 1976 5
440372 1979 7
1836859 1982 9
1872813 1976 6
97690 1979 8
1897496 1982 10 1897947 1976 7
308019 1979 9
1082095 1982 11 1816532 1976 8
0 1979 10 1439455 1982 12 1956855 1976 9
0 1979 11 1820321 1983 1
1872765 1976 10 0
1979.
12 1901210 1983 2
1659657 1976 11 0
1980 1
1818926 1983 3
0 1976 12 347603 1980 2
1775685 1983 4
0 1977 1
1317560 1980 3
892672 1983 5
0 1977 2
1389034 1980 4
0 1983 6
0 1977 3
1592703 1980 5
890501 1983 7
0 1977 4
407185 1980 6
625543 1983 8
0 1977 5
1519476 1980 7
1874604 1983 9
0 1977 6
1377938 1980 8
1697623 1983 10 0
1977 7
1705677 1980 9
1810963 1983 11 0
1977 8
1865335 1980 10 1885676 1983 12 0
1977 9
1450241 1980 11 1814059 1984 1
0 1977 10 1270278 1980 12 1902530 1984 2
0 1977 11 1764776 1981 1
1873725 1984 3
0 1977 12 1841408 1981 2
1717790 1984 4
0 1978 1
1569897 1981 3
1901230 1984 5
798009 1978 2
1678630 1981 4
1517646 1984 6
1884155 1978 3
1642702 1981 5
1881121 1984 7
1732947 1978 4
0 1981 6
1838626 1984 8
1882130 1978 5
0 1981 7
1885109 1984 9
1415580 1978 6
1488742 1981 8
1893652 1984 10 1990103 1978 7
1877051 1981 9
438971 1984 11 1930083 1978 8
1781953 1981 10 0
1984 12 1834566 1978 9
1742015 1981 11 0
1985 1
1950365 1978 10 1598034 1981 12 1127901 1985 2
1801216 1978 11 1284581 1982 1
1946002 1985 3
1964733 1978 12 1705329 1982 2
1791664 1985 4
1927520 1979 1
1505641 1982 3
1984448 1985 5
1990987 1979 2
1641249 1982 4
1922445 1985 6
1930233 1979 3
1891957 1982 5
768617 1985 7
1971995 1979 4
0 1982 6
1897597 1985 8
1975243 1979 5
0 1982 7
1990892 1985 9
1929908 1979 6
1025125 1982 8
1965785 1985 10 1214386 Analysis of St. Lucie Unit I Capsule 2840
6-23 Table 6-7 Cont'd Monthly Thermal Generation During the First Fifteen Fuel Cycles of the St. Lucie Unit 1 Reactor (Reactor Power of 2700 MWt)
Thermal Thermal Thermal Generat.
Generat.
Generat.
Year Month (MW-hr)
Year Month (MW-hr)
Year Month (MW-hr) 1985 11 0
1989 1
1977328 1992 3
2008800 1985 12 98138 1989 2
1800643 1992 4
1903500 1986 1
1963711 1989 3
1987159 1992 5
2008800 1986 2
1773239 1989 4
1934223 1992 6
1865700 1986 3
2004406 1989 5
1994532 1992 7
2008800 1986 4
1862951 1989' 6
1783978 1992 8
1919700 1986 5
2009078 1989 7
849129 1992 9
1085400 1986 6
1081761 1989 8
1989644 1992 10 2003400 1986 7
2008970 1989 9
1834971 1992 11 1930500 1986 8
1975834 1989 10 1982298 1992 12 1979100 1986 9
1722036 1989 11 1907837 1993 1
1962900 1986 10 1998087 1989 12 1894712 1993 2
1787400 1986 11 1933003 1990 1
1248855 1993 3
1806300 1986 12 1993847 1990 2
0 1993 4
0 1987 1
2002084 1990 3
0 1993 5
0 1987 2
386909 1990 4
229500 1993 6
734400 1987 3
0 1990 5
1560600 1993 7
2003400 1987 4
832420 1990 6
1852200 1993 8
1965600 1987 5
1891173 1990 7
99900 1993 9
1350000 1987 6
1877939 1990 8
1476900 1993 10 1971000 1987 7
2003731 1990 9
1944000 1993 11 1898100 1987 8
2004433 1990 10 1876500 1993 12 1944000 1987 9
1933003 1990 11 1941300 1994 1
1906200 1987 10 1286517 1990 12 1952100 1994 2
1814400 1987 11 1937190 1991 1
1914300 1994 3
1765800 1987 12 1973080 1991 2
1800900 1994 4
1741500 1988 1
2007107 1991 3
2008800 1994 5
1962900 1988 2
1877156 1991 4
1898100 1994 6
1536300 1988 3
1903649 1991 5
1825200 1994 7
1941300 1988 4
1941214 1991 6
1809000 1994 8
1871100 1988 5
2000153 1991 7
1806300 1994 9
1903500 1988 6
1878506 1991 8
1952100 1994 10 1657800 1988 7
612621 1991 9
1838700 1994 11 0
1988 8
2511 1991 10 1144800 1994 12 1863000 1988 9
1556411 1991 11 0
1995 1
2008800 1988 10 2011022 1991 12 480600 1995 2
1679400 1988 11 1944291 1992 1
2008800 1995 3
1482300 1988 12 1938835 1992 2
1879200 1995 4
1944000 Analysis of St. Lucie Unit I Capsule 284°
6-24 Table 6-7 Cont'd Monthly Thermal Generation During the First Fifteen Fuel Cycles of the St. Lucie Unit 1 Reactor (Reactor Power of 2700 MWt)
Thermal Thermal Generat.
Generat.
Year Month (MW-hr)
Year Month (MW-hr 1995 5
2008800 1997 8
2008800 1995 6
1930500 1997 9
1941300 1995 7
1760400 1997 10 1225800 1995 8
37800 1997 11 0
1995 9
0 1997 12 0
1995 10 1109700 1998 1
1228500 1995 11 1768500 1998 2
1336500 1995 12 2008800 1998 3
2008800 1996 1
1998000 1998 4
1938600 1996 2
1655100 1998 5
2008800 1996 3
1925100 1998 6
1868400 1996 4
1822500 1998 7
2008800 1996 5
0 1998 8
2008800 1996 6
0 1998 9
1930500 1996 7
288900 1998 10 2008800 1996 8
1952100 1998 11 1919700 1996 9
1331100 1998 12 2008800 1996 10 1981800 1999 1
2006100 1996 11 1917000 1999 2
1795500 1996 12 1976400 1999 3
2003400 1997 1
1995300 1999 4
1938600 1997 2
1806300 1999 5
2008800 1997 3
1806300 1999 6
1941300 1997 4
1649700 1999 7
2006100 1997 5
2006100 1999 8
1676700 1997 6
1944000 1999 9
764100 1997 7
1984500 Analysis of St. Lucie Unit 1 Capsule 2840
6-25 Table 6-8 Measured Sensor Activities and Reaction Rates 2840 Surveillance Capsule Reaction "63Cu (n,a) 6OCo (Cd) 46Ti (n, p) 46Sc 54Fe (n,p) 54Mn 58Ni (n,p) 58Co (Cd) 59Co (n,y) 6°Co 59Co (n,_Y) 60Co (Cd) 238U (n,f) 137 Cs Location CAP37614 CAP67641 CAP67673 CAP37614 CAP67641 CAP67673 CAP37614 CAP67641 CAP67673 CAP37614 CAP67641 CAP67673 CAP37614 CAP67641 CAP67673 CAP37614 CAP67641 CAP67673 CAP37614 CAP67641 CAP67673 Measured Activity (dps/g*n) 1.96E+05 1.82E+05 1.94E+05 1.46E+05 1.26E+05 1.38E+05 1.20E+06 1.07E+06
- 1. 14E+06 5.1OE+06 4.35E+06 4.76E+06 1.41E+07 1.50E+07 1.14E+07 1.89E+06 1.97E+06 1.93E+06 1.04E+06 1.02E+06 8.62E+05 Saturated Activity (dps/gm) 2.74E+05 2.55E+05 2.71E+05 8.98E+05 7.75E+05 8.49E+05 2.19E+06 1.96E+06 2.08E+06 4.47E+07 3.81E+07 4.17E+07 2.86E+07 3.04E+07 2.3 1E+07 3.84E+06 4.OOE+06 3.92E+06 3.41E+06 3.35E+06 2.83E+06 Average Including 235U, 239Pu, and y,fission corrections Analysis of St. Lucie Unit I Capsule 284° Reaction Rate (rps/atom) 4.18E-17 3.88E-17 4.14E-17 8.30E-16 7.17E-16 7.85E-16 3.48E-15 3.10E-15 3.3 GE-15 6.3 9E-15 5.45E-15 5.97E-15 1.87E-12 1.99E-12 1.51E-12 2.50E-13 2.61 E-13 2.56E-13 2.24E-14 2.20E-14 1.86E-14 1.48E-14
6-26 Table 6-8 cont'd Measured Sensor Activities and Reaction Rates 104' Surveillance Capsule Reaction 63Cu (n,ca) 60Co (Cd) 46Ti (n, p) 46Sc 54Fe (n,p) 14Mn 5'Ni (n,p) 5"Co (Cd) 59Co (n,7) 6°Co 59Co (n,y) 6°Co (Cd) 238U (n,f) 137Cs Location 1-7314 1-7341 1-7373 2-7314 2-7341 2-7373 2-7314 2-7341 2-7373 1-7314 1-7341 1-7373 2-7314 2-7341 2-7373 1-7314 1-7341 1-7373 2-7314 2-7341 2-7373 Measured Activity (dps/gm) 1.78E+05 1.66E+05 1.79E+05 1.84E+05 1.55E+05 1.61E+05 1.40E+06 1.28E+06 1.3 1E+06 6.18E+06 5.34E+06 5.19E+06 1.92E+07 1.96E+07 1.24E+07 2.43E+06 2.48E+06 2.43E+06 8.55E+05 8.96E+05 6.78E+05 Saturated Activity (dps/gm) 2.66E+05 2.48E+05 2.68E+05 6.13E+05 5.16E+05 5.36E+05 1.87E+06 1.71E+06 1.75E+06 2.75E+07 2.37E+07 2.3 1E+07 3.93E+07 4.01E+07 2.54E+07 4.97E+06 5.07E+06 4.97E+06 4.44E+06 4.65E+06 3.52E+06 Average Including 231U, 239Pu, and yfission corrections 23 8U (n,f) 137Cs (Cd) 1-7314 1-7341 1-7373 3.18E+05 3.03E+05 3.49E+05 1.65E+06 1.57E+06 1.81E+06 Average Including 235U, 239pu, and y,fission corrections Analysis of St. Lucie Unit I Capsule 2840 Reaction Rate (rps/atom) 4.06E-17 3.79E-17 4.09E-17 5.67E-16 4.77E-16 4.96E-16 2.97E-15 2.71 E-15 2.78E-15 3.93E-15 3.40E-15 3.30E-15 2.5 6E-12 2.61 E-12 1.65E-12 3.24E-13 3.3 1E-13 3.24E-13 2.9 1E-14 3.05E-14 2.3 1E-14 1.92E-14 1.08E-14 1.03E-14 1.19E-14 7.67E-15
6-27 Table 6-8 cont'd Measured Sensor Activities and Reaction Rates 97 0Surveillance Capsule Reaction 63Cu (n,'a) 60CO (Cd) 46Ti (n, p) 6Sc 54Fe (n,p) 54Mn "58Ni (n,p) 58Co (Cd) 59Co (n,y') 6°Co 59Co (n,Y) 60Co (Cd) 238U (n,f) 137Cs Location A
B C
A B
C A
B C
A B
C A
B C
A B
C A
B C
Measured Activity (dps/mn) 1.43E+05 1.38E+05 1.52E+05 5.87E+05 6.30E+05 6.60E+05 1.82E+06 1.97E+06 1.93E+06 3.25E+07 3.45E+07 2.90E+07 3.05E+07 2.20E+07 3.001E+07 4.58E+06 4.63E+06 4.60E+06 6.1OE+05 8.27E+05 7.17E+05 Saturated Activity (dps/gm) 3.42E+05 3.30E+05 3.63E+05 6.98E+05 7.49E+05 7.85E+05 2.36E+06 2.55E+06 2.50E+06 3.83E+07 4.07E+07 3.42E+07 7.29E+07 5.26E+07 7.17E+07 1.1OE+07 1.11E+07 1.1OE+07 6.13E+06 8.32E+06 7.21 E+06 Average Including 235U, 239Pu, and yfission corrections 238U (n,f) 137Cs (Cd)
A B
C 2.48E+05 3.35E+05 3.83E+05 2.49E+06 3.37E+06 3.85E+06 Average Including 235U, 239pu, and y,fission corrections Analysis of St. Lucie Unit 1 Capsule 2840 Reaction Rate (rps/atom) 5.22E-17 5.03E-17 5.54E-17 6.46E-16 6.93E-16 7.26E-16 3.73E-15 4.04E-15 3.96E-15 5.49E-15 5.82E-15 4.90E-15 4.76E-12 3.43E-12 4.68E-12 7.15E-13 7.22E-13 7.18E-13 4.03E-14 5.46E-14 4.74E-14 3.72E-14 1.64E-14 2.21E-14 2.53E-14 1.67E-14
6-28 Table 6-9 Summary of Neutron Dosimetry Results 970, 1040, and 2840 Surveillance Capsules Best Estimate Flux and Fluence for 284' Capsule Quantit 4 (E > 1.0 MeV) 4 (E > 0.1 MeV)
- 4) (E < 0.414 eV) dpa/sec Flux fn/cm 2-sec]
2.82E+10 5.40E+10 6.30E+10 4.11E-11 Quantit0 cD1(E > 1.0 MeV)
(D (E > 0. 1 MeV)
(D (E < 0.414 eV) dpa Best Estimate Flux and Fluence for 1040 Capsule Quantity S(E > 1.0 MeV)
S(E > 0.1 MeV)
S(E < 0.414 eV) dpa/sec Flux
[n/cm 2-sec]
2.76E+10 5.95E+10 8.01E+10 4.10E-Il Quantity cD (E > 1.0 MeV) cD (E > 0.1 MeV) cD (E < 0.414 eV) dpa Best Estimate Flux and Fluence for 970 Capsule Quantity 4 (E > 1.0 MeV) 4 (E > 0.1 MeV)
S(E < 0.414 eV) dpa/sec Flux
[n/cm _secl 3.78E+10 8.71E+10 1.48E+ 1I 5.77E-1 1 Quantity cD (E > 1.0 MeV)
D (E > 0. 1 MeV)
S(E < 0.414 eV) dpa Analysis of St. Lucie Unit I Capsule 2840 Fluence
[n/CM2]
1.53E+19 2.93E+19 3.43E+19 2.23E-02 Uncertaint 6%
9%
8%
6%
Fluence
[n/cm 2]
8.29E+18 1.78E+19 2.40E+19 1.23E-02 Uncertainty 6%
9%
8%
6%
Fluence
[n/cm2]
5.52E+18 1.27E+19 2.16E+19 8.43E-03 Uncertainty 6%
10%
9%
6%
6-29 Table 6-10 Comparison of Measured, Calculated, and Best Estimate Reaction Rates at the Surveillance Capsule Center Reaction 63Cu (n,a) 46Ti (n, p) 54Fe (n,p) 59Co (n,y)
"9Co (n,y) (Cd) 238U (n,f) (Cd)
Reaction 63CU (n,x) 46Ti (n, p) 54Fe (n,p)
-8Ni (n,p)
' 9Co (n,y)
"59Co (n,y) (Cd) 238U (n,f)
Reaction 63Cu (n,ot) 46Ti (n, p) 54Fe (n,p) 58Ni (n,p) 59Co (n,y)
"59Co (n,y) (Cd) 231U (n,f) (Cd)
Measured 4.07E-17 7.77E-16 3.29E-15 1.79E-12 2.55E-13 1.48E-14 Measured 3.98E-17 5.13E-16 2.82E-15 3.54E-15 2.28E-12 3.26E-13 1.92E-14 Measured 5.26E-17 6.88E-16 3.91 E-15 5.40E-15 4.29E-12 7.18E-13 1.67E-14 Surveillance 284' Capsule Best Calculated Estimate BE / Meas 4.20E-17 4.25E-17 1.05 6.45E-16 7.04E-16 0.91 3.57E-15 3.64E-15 1.11 1.51E-12 1.78E-12 1.00 3.07E-13 2.58E-13 1.01 1.20E-14 1.26E-14 0.86 Surveillance 1040 Capsule Best Calculated Estimate BE / Meas 4.43E-17 3.69E-17 0.93 6.85E-16 5.36E-16 1.04 3.87E-15 3.09E-15 1.10 5.07E-15 3.99E-15 1.13 1.99E-12 2.27E-12 1.00 4.01E-13 3.29E-13 1.01 1.35E-14 1.18E-14 0.61 Surveillance 970 Capsule Best Calculated Estimate BE / Meas 4.41E-17 4.86E-17 0.92 7.00E-16 7.07E-16 1.03 4.27E-15 4.12E-15 1.05 5.66E-15 5.49E-15 1.02 3.81E-12 4.28E-12 1.00 7.56E-13 7.20E-13 1.00 1.67E-14 1.59E-14 0.95 Analysis of St. Lucie Unit 1 Capsule 2840 BE/ Calc 1.01 1.09 1.02 1.18 0.84 1.05 BE/ Calc 0.83 0.78 0.80 0.79 1.14 0.82 0.87 BE/ Calc 1.10 1.01 0.97 0.97 1.12 0.95 0.95 Meas/Calc 0.97 1.20 0.92 1.18 0.83 1.23 Meas/Calc 0.90 0.75 0.73 0.70 1.14 0.81 1.43 Meas/Calc 1.19 0.98 0.92 0.95 1.13 0.95 1.00
6-30 Table 6-11 Best Estimate Neutron Energy Spectrum at the Center of Surveillance Capsules 2840 Capsule Group #
1 2
3 4
5 6
7 8
9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 Energy (MeV) 1.73E+01 1.49E+01 1.35E+01 1.16E+0 1 1.OOE+01 8.61E+00 7.41E+00 6.07E+00 4.97E+00 3.68E+00 2.87E+00 2.23E+00 1.74E+00 1.35E+00 1.11E+00 8.21E-01 6.39E-01 4.98E-01 3.88E-01 3.02E-0 1 1.83E-01 1.11E-01 6.74E-02 4.09E-02 2.55E-02 1.99E-02 1.50E-02 Flux (n/cm 2-sec) 6.59E+06 1.40E+07 4.89E+07 1.27E+08 2.82E+08 4.80E+08 1.24E+09 1.77E+09 3.07E+09 2.84E+09 4.54E+09 4.38E+09 4.46E+09 3.35E+09 4.66E+09 4.30E+09 3.83E+09 2.67E+09 3.02E+09 4.54E+09 3.74E+09 2.98E+09 2.59E+09 1.91E+09 1.37E+09 1.17E+09 2.01E+09 Group #
28 29 30 31 32 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 Energy (MeV) 9.12E-03 5.53E-03 3.36E-03 2.84E-03 2.40E-03 2.04E-03 1.23E-03 7.49E-04 4.54E-04 2.75E-04 1.67E-04 1.01E-04 6.14E-05 3.73E-05 2.26E-05 1.37E-05 8.32E-06 5.04E-06 3.06E-06 1.86E-06 1.13E-06 6.83E-07 4.14E-07 2.51E-07 1.52E-07 9.24E-08 Note: Tabulated energy levels represent the upper energy in each group.
Analysis of St. Lucie Unit 1 Capsule 284' Flux (n/cm2-sec) 2.06E+09 2.08E+09 6.63E+08 6.44E+08 6.38E+08 1.91E+09 1.90E+09 1.82E+09 1.72E+09 1.80E+09 1.68E+09 1.83E+09 1.85E+09 1.89E+09 1.92E+09 1.93E+09 1.96E+09 2.03E+09 2.05E+09 2.06E+09 2.02E+09 1.86E+09 2.42E+09 8.59E+09 1.53E+10 3.67E+10
6-31 Table 6-11 cont'd Best Estimate Neutron Energy Spectrum at the Center of Surveillance Capsules 1040 Capsule Group #
1 2
3 4
5 6
7 8
9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 Energy (MeV) 1.73E+01 1.49E+01 1.35E+01 1.16E+0 I 1.00E+01 8.61E+00 7.41E+00 6.07E+00 4.97E+00 3.68E+00 2.87E+00 2.23E+00 1.74E+00 1.35E+00 1.11E+00 8.21E-01 6.39E-01 4-98E-01 3.88E-01 3-02E-01 1.83E-01 11 IE-01 6.74E-02 4.09E-02 2.55E-02 1.99E-02 1.50E-02 Flux (n/cm2-sec) 5.74E+06 1.19E+07
- 4. 1OE+07 1.05E+08 2.3 1E+08 3.8§E+08 9.87E+08 1.42E+09 2.57E+09 2.55E+09 4.36E+09 4.49E+09 4.84E+09 3.77E+09 5.46E+09 5.21E+09 4.77E+09 3.33E+09 3.78E+09 5.69E+09 4.67E+09 3.71E+09 3.20E+09 2.34E+09 1.66E+09 1.42E+09 2.43E+09 Group #
28 29 30 31 32 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 Energy (MeV) 9.12E-03 5.53E-03 3.36E-03 2.84E-03 2.40E-03 2.04E-03 1.23E-03 7.49E-04 4.54E-04 2.75E-04 1.67E-04 1.01E-04 6.14E-05 3.73E-05 2.26E-05 1.37E-05 8.32E-06 5.04E-06 3.06E-06 1.86E-06 1.13E-06 6.83E-07 4.14E-07 2.5 1E-07 1.52E-07 9.24E-08 Note: Tabulated energy levels represent the upper energy in each group.
Analysis of St. Lucie Unit 1 Capsule 2840 Flux (n/cm2-sec) 2.48E+09 2.52E+09 8.05E+08 7.82E+08 7.72E+08 2.33E+09 2.33E+09 2.26E+09 2.15E+09 2.24E+09 2.16E+09 2.30E+09 2.33E+09 2.38E+09 2.4 1E+09 2.4 1E+09 2.45E+09 2.54E+09 2.56E+09 2.58E+09 2.53E+09 2.33E÷09 3.04E+09 1.08E+10 1.94E+10 4.68E+10
6-32 Table 6-11 cont'd Best Estimate Neutron Energy Spectrum at the Center of Surveillance Capsules 970 Capsule Group #
1 2
3 4
5 6
7 8
9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 Energy (MeV) 1.73E+01 1.49E+01 1.35E+01 1.16E+0 1 1.OOE+01 8.6 1E+00 7.41E+00 6.07E+00 4.97E+00 3.68E+00 2.87E+00 2.23E+00 1.74E+00 1.35E+00 1.11 E+00 8.21 E-0 1 6.39E-01 4.98E-01 3.88E-01 3.02E-01 1.83E-01 1.11E-01 6.74E-02 4.09E-02 2.55E-02 1.99E-02 1.50E-02 Flux (n/cm 2-sec) 6.83E+06 1.46E+07 5.26E+07 1.40E+08 3.11E+08 5.26E+08 1.28E+09
- 1. 84E+09 3.42E+09 3.46E+09 5.86E+09 6.14E+09 6.77E+09 5.34E+09 7.85E+09 7.67E+09 7.21E+09 5.08E+09 5.86E+09 9.18E+09 7.77E+09 6.42E+09 5.67E+09 4.26E+09 3.05E+09 2.65E+09 4.60E+09 Group t 28 29 30 31 32 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 Energy (MeV) 9.12E-03 5.53E-03 3.36E-03 2.84E-03 2.40E-03 2.04E-03 1.23E-03 7.49E-04 4.54E-04 2.75E-04 1.67E-04 1.01E-04 6.14E-05 3.73E-05 2.26E-05 1.37E-05 8.32E-06 5.04E-06 3.06E-06 1.86E-06 1.13E-06 6.83E-07 4.14E-07 2.5 1E-07 1.52E-07 9.24E-08 Note: Tabulated energy levels represent the upper energy in each group.
Analysis of St. Lucie Unit I Capsule 284' Flux (n/crn-sec) 4.75E+09 4.88E+09 1.58E+09 1.55E+09 1.54E+09 4.67E+09 4.76E+09 4.69E+09 4.52E+09 4.73E+09 4.79E+09 4.94E+09 5.OOE+09 5.08E+09 5.12E+09 5.13E+09 5.14E+09 5.35E+09 5.41E+09 5.44E+09 5.29E+09 4.88E+09 6.07E+09 2.08E+10 3.66E+10 8.46E+10
6-33 Table 6-12 Comparison of Calculated and Best Estimate Integrated Neutron Exposure of the 970, 1040, and 284' Surveillance Capsules 2840 CAPSULE CD(E > 1.0 MeV) [n/cm 2]
c(1(E > 0.1 MeV) [n/cm2]
dpa Calculated 1.45E+19 2.80E+19 2.12E-02 Best Estimate 1.53E+19 2.93E+19 2.23E-02 1040 CAPSULE (D(E > 1.0 MeV) [n/cm 2]
(D(E > 0.1 MeV) [n/cm 2]
dpa Calculated 9.18E+18 1.84E+19 1.35E-02 Best Estimate 8.29E+18 1.78E+19 1.23E-02 970 CAPSULE (D(E > 1.0 MeV) [n/cm 2]
qc(E > 0.1 MeV) [n/cm 2]
dpa Calculated 5.91E+18 1.35E+19 8.86E-03 Best Estimate 5.52E+18 1.27E+19 8.43E-03 AVERAGE BE/C RATIOS BE/C (D(E > 1.0 MeV) [n/cm 2]
0.96 c1(E > 0.1 MeV) [n/cm 2]
0.99 dpa 0.97 Analysis of St. Lucie Unit I Capsule 2840 BE/C 1.05 1.05 1.05 BE/C 0.90 0.97 0.91 BE/C 0.93 0.94 0.95
6-34 Table 6-13 Azimuthal Variations of The Neutron Exposure Projections On The Reactor Vessel Clad/Base Metal Interface At Core Midplane Best Estimate 01[a]
150 300 450 17.23 EFPY E>1.0 MeV 1.39E+19 8.96E+18 9-05E+18 6.76E+18 E>O.1 MeV 3.25E+19 2.16E+19 2.10E+19 1.55E+19 dpa 2.17E-02 1.42E-02 1.41E-02 1.06E-02 Projection Data:
E>1.0 MeV Fluence/EFPY 7.19E+17 4.82E+17 4.11E+17 3.32E+17 E>0.1 MeV Fluence/EFPY 1.60E+18 1.11E+18 9.18E+17 7.32E+17 dpa/EFPY 1.11E-03 7.53E-04 6.37E-04 5.18E-04 25 EFPY E>1.0 MeV 1.95E+19 1.27E+19 1.22E+19 9.34E+18 E>O.1 MeV 4.49E+19 3.02E+19 2.81E+19 2.12E+19 dpa 3.03E-02 2.OOE-02 1.91E-02 1.46E-02 32 EFPY E>1.0 MeV 2.46E+19 1.61E+19 1.51E+19 1.17E+19 E>0.1 MeV 5.61E+19 3.80E+19 3.45E+19 2.63E+19 dpa 3.81E-02 2.53E-02 2.35E-02 1.83E-02 35 EFPY E>1.0 MeV 2.67E+19 1.75E+19 1.64E+19 1.27E+19 E>0.1 MeV 6.09E+19 4.13E+19 3.73E+19 2.85E+19 dpa 4.14E-02 2.75E-02 2.54E-02 1.98E-02 48 EFPY E>1.0 MeV 3.61E+19 2.38E+19 2.17E+-19 1.70E+19 E>O.1 MeV 8.17E+19 5.57E+19 4.92E+19 3.80E+19 dpa 5.58E-02 3.73E-02 3.37E-02 2.65E-02 54 EFPY E>1.0 MeV 4.04E+19 2.67E+19 2.42E+19 1.90E+19 E>O.1 MeV 9.13E+19 6.23E+19 5.48E+19 4.24E+ 19 dpa 6.25E-02 4.18E-02 3.76E-02 2.97E-02 Note:
a) Maximum neutron exposure projection.
Analysis of St. Lucie Unit I Capsule 2840
6-35 Table 6-13, cont'd Azimuthal Variations of The Neutron Exposure Projections On The Reactor Vessel Clad/Base Metal Interface At Core Midplane Calculated O[a]
150 300 450 17.23 EFPY E>1.0 MeV 1.45E+19 9.29E+18 9.39E+18 7.01E+18 E>O.1 MeV 3.29E+19 2.19E+19 2.13E+19 1.57E+19 dpa 2.23E-02 1.46E-02 1.45E-02 1.09E-02 Projection Data:
E>1.0 MeV Fluence/EFPY 7.46E+17 5.OOE+17 4.26E+17 3.45E+17 E>O.1 MeV Fluence/EFPY 1.62E+18 1.12E+18 9.30E+17 7.42E+17 dpa/EFPY 1.14E-03 7.75E-04 6.56E-04 5.33E-04 25 EFPY E>1.0 MeV 2.03E+19 1.32E+19 1.27E+19 9.69E+18 E>O.1 MeV 4.55E+19 3.06E+19 2.85E+19 2.14E+19 dpa 3.12E-02 2.06E-02 1.96E-02 1.5 1E-02 32 EFPY E>I.0 MeV 2.55E+19 1.67E+19 1.57E+19 1.21E+19 E>O.1 MeV 5.68E+19 3.85E+19 3.50E+19 2.66E+19 dpa 3.92E-02 2.60E-02 2.42E-02 1.88E-02 35 EFPY E>I.0 MeV 2.77E+19 1.82E+19 1.70E+19 1.31E+19 E>O.1 MeV 6.17E+19 4.18E+19 3.78E+19 2.88E+19 dpa 4.26E-02 2.83E-02 2.62E-02 2.04E-02 48 EFPY E>I.0 MeV 3.74E+19 2.47E+19 2.25E+19 1.76E+19 E>O.1 MeV 8.27E+19 5.64E+19 4.99E+19 3.85E+19 dpa 5.75E-02 3.84E-02 3.47E-02 2.73E-02 54 EFPY E>1.0 MeV 4.19E+19 2.77E+19 2.51E+19 1.97E+19 E>O.1 MeV 9.24E+19 6.31E+19 5.55E+19 4.29E+19 dpa 6.43E-02 4.3 1E-02 3.86E-02 3.05E-02 Note:
a) Maximum neutron exposure projection.
Analysis of St. Lucie Unit I Capsule 284'
Table 6-14 Neutron Exposure Values Within The St. Lucie Unit 1 Reactor Vessel Best Estimate Fluence (n/cm2) Based on E > 1.0 MeV Slope 17.23 EFPY, E>1.0 MeV Projection Data Beyond Current 17.23 EFPY:
E> 1.0 MeV Fluence/EFPY Attenuation Functions From Table 6-3:
1/4T 25 EFPY Surface
/4T
%T 32 EFPY Surface
/4 T 3/4T 35 EFPY Surface V4T 33/4T 48 EFPY Surface
/4T 33/4T 54 EFPY Surface "AT M
150_° 300 450 1.39E+19 8.96E+18 9.05E+18 6.76E+18 7.19E+17 4.82E+1 7 4.11E+17 3.32E+17 0.536 0.129 1.95E+19 1.05E+19 2.52E+ 18 2.46E+19 1.32E+19 3.17E+ 18 2.67E+19 1.43E+19 3.45E+18 3.61E+19 1.94E+19 4.66E+18 4.04E+19 2.17E+19 5.21E+18 0.537 0.133 1.27E+19 6.82E+18 1.69E+18 1.61E+19 8.63E+18 2.14E+18 1.75E+19 9.41E+18 2.33E+18 2.38E+19 1.28E+19 3.16E+ 18 2.67E+19 1.43E+19 3.55E+18 0.539 0.132 1.22E+19 6.60E+18 1.61E+18 1.51E+19 8.15E+18 1.99E+18 1.64E+19 8.81E+18 2.15E+18 2.17E+19 1.17E+19 2.86E+18 2.42E+19 1.30E+19 3.18E+18 0.539 0.133 9.34E+1 8 5.03E+18 1.25E+18 1.17E+19 6.28E+18 1.56E+18 1.27E+19 6.82E+18 1.69E+18 1.70E+19 9.15E+18 2.26E+18 1.90E+19 1.02E+19 2.53E+18 Note:
a)
Maximum neutron exposure projection.
Analysis of St. Lucie Unit 1 Capsule 2840 6-36
Table 6-14, cont'd Neutron Exposure Values Within The St. Lucie Unit 1 Reactor Vessel Best Estimate Fluence (n/cm2) Based on dpa Slope 17.23 EFPY, E>1.0 MeV Projection Data Beyond Current 17.23 EFPY:
E>1.0 MeV Fluence/EFPY Attenuation Functions From Table 6-5:
1/T 3AT 25 EFPY Surface 1/4T 3/T 32 EFPY Surface 1/4T 3/4T 35 EFPY Surface 1/T
%T 48 EFPY Surface 1/T 3/T 54 EFPY Surface 1/T 3/T 00O1a 150 300 450 1.39E+19 8.96E+18 9.05E+18 6.76E+18 7.19E+17 4.82E+17 4.1 1E+17 3.32E+17 0.614 0.221 1.95E+19 1.20E+19 4.31E+18 2.46E+19 1.51E+19 5.42E+18 2.67E+19 1.64E+19 5.90E+ 18 3.61E+19 2.22E+19 7.96E+18 4.04E+19 2.48E+19 8.91E+18 0.621 0.235 1.27E+19 7.89E+18 2.98E+18 1.61E+19 9.99E+18 3.78E+18 1.75E+19 1.09E+19 4.11E+18 2.38E+19 1.48E+19 5.59E+18 2.67E+19 1.66E+19 6.27E+18 0.619 0.231 1.22E+19 7.58E+18 2.83E+18 1.51E+19 9.3 6E+ 18 3.49E+18 1.64E+19 1.01E+19 3.78E+18 2.17E+19 1.34E+19 5.01E+18 2.42E+19 1.49E+19 5.58E+18 0.618 0.236 9.34E+18 5.77E+18 2.20E+18 1.17E+19 7.21E+18 2.75E+18 1.27E+19 7.83E+18 2.98E+ 18 1.70E+19 1.05E+19 4.OOE+ 18 1.90E+19
- 1. 17E+ 19 4.47E+18 Note:
a) Maximum neutron exposure projection.
b) The 1/4T and /T values were determined using the calculational methods described in Section 6.2 and not by the empirical relation described in Regulatory Guide 1.99, Rev. 2.
Analysis of St. Lucie Unit I Capsule 2840 6-37
Table 6-14, cont'd Neutron Exposure Values Within The St. Lucie Unit I Reactor Vessel Calculated Fluence (n/cm 2) Based on E > 1.0 MeV Slope 17.23 EFPY, E>1.0 MeV Projection Data Beyond Current 17.23 EFPY:
E> 1.0 MeV Fluence/EFPY Attenuation Functions From Table 6-3:
/4T 33/4T 25 EFPY Surface 11/4T 3/4T 32 EFPY Surface 1/4T 3/4,T 35 EFPY Surface
/4T 3/T 48 EFPY Surface 1/4T
%T 54 EFPY Surface 1/4T 3/4T M
150 300 45___0 1.45E+19 9.29E+18 9.39E+18 7.01E+18 7.46E+17 5.OOE+17 4.26E+17 3.45E+17 0.536 0.129 2.03E+19 1.09E+19 2.61E+18 2.55E+19 1.37E+19 3.29E+1 8 2.77E+19 1.49E+19 3.58E+18 3.74E+19 2.01E+19 4.83E+18 4.19E+19 2.25E+19 5.41E+18 0.537 0.133 1.32E+19 7.08E+18 1.75E+18 1.67E+19 8.95E+18 2.22E+ 18 1.82E+19 9.76E+18 2.42E+ 18 2.47E+19 1.33E+19 3.28E+18 2.77E+19 1.49E+19 3.68E+18 0.539 0.132 1.27E+19 6.84E+ 18 1.67E+18 1.57E+19 8.45E+18 2.06E+ 18 1.70E+19 9.14E+18 2.23E+18 2.25E+19 1.21E+19 2.96E+18 2.51E+19 1.35E+19 3.30E+18 0.539 0.133 9.69E+18 5.22E+ 18 1.29E+18 1.21E+19 6.52E+18 1.61E+18 1.31E+19 7.07E+18 1.75E+18 1.76E+19 9.49E+18 2.35E+18 1.97E+19 1.06E+19 2.62E+18 Note:
a) Maximum neutron exposure projection.
b) The 1/4T and 3/T values were determined using the calculational methods described in Section 6.2 and not by the empirical relation described in Regulatory Guide 1.99, Rev. 2.
Analysis of St. Lucie Unit I Capsule 2840 6-38
6-39 Table 6-14, cont'd Neutron Exposure Values Within The St. Lucie Unit I Reactor Vessel Calculated Fluence (n/cm2) Based on dpa Slope 17.23 EFPY, E>1.0 MeV Projection Data Beyond Current 17.23 EFPY:
E> 1.0 MeV Fluence/EFPY Attenuation Functions From Table 6-5:
'1/4T 33/4T 25 EFPY Surface 1/44T 3/T 32 EFPY Surface 1/44T 3/T 35 EFPY Surface
'AT 33/4T 48 EFPY Surface 1/44T
% T 54 EFPY Surface 1/4T
%3/4T 00fal 150 300 450 1.45E+19 9.29E+18 9.39E+18 7.01E+18 7.46E+17 5.OOE+17 4.26E+17 3.45E+17 0.614 0.221 2.03E+19 1.24E+19 4.47E+ 18 2.55E+19 1.56E+19 5.62E+18 2.77E+19 1.70E+19 6.12E+18 3.74E+19 2.3 OE+ 19 8.25E+1 8 4.19E+19 2.57E+19 9.24E+18 0.621 0.235 1.32E+19 8.19E+18 3.09E+18 1.67E+19 1.04E+19 3.92E+1 8 1.82E+19 1.13E+19 4.27E+18 2.47E+19 1.53E+19 5.79E+18 2.77E+19 1.72E+19 6.50E+18 0.619 0.231 1.27E+19 7.86E+18 2.94E+ 18 1.57E+19 9.70E+ 18 3.62E+18 1.70E+19 1.05E+19 3.92E+18 2.25E+19 1.39E+19 5.20E+18 2.51E+19 1.55E+19 5.79E+18 0.618 0.236 9.69E+18 5.99E+18 2.28E+18 1.21E+19 7.48E+18 2.85E+18 1.31E+19 8.12E+18 3.10E+18 1.76E+19 1.09E+19 4.15E+18 1.97E+19 1.22E+19 4.64E+18 Note:
a) Maximum neutron exposure projection.
b) The 1/4AT and 3/4/T values were determined using the calculational methods described in Section 6.2 and not by the empirical relation described in Regulatory Guide 1.99, Rev. 2.
Analysis of St. Lucie Unit I Capsule 2840
6-40 Table 6-15 Updated Lead Factors for St. Lucie Unit 1 Surveillance Capsules Capsule Lead Factor 9701a]
1.53 10 4 ob]
0.98 2840[c]
1.00 2 63 otd' 1.36 277o[dl 1.36 8 3 o[d]
1.36
[a] - Withdrawn at the end of Cycle 5.
[b] - Withdrawn at the end of Cycle 9.
[c] - Withdrawn at the end of Cycle 15.
[d] - Not withdrawn; standby.
The surveillance capsule lead factor is defined by:
I)Surveillance Capsule Calculated (D Clad I Base Metal Interface Axial Peak Calculated where (D is the neutron fluence (E > 1.0 MeV) at the time of the capsule withdrawal. In the case of the standby capsules, the neutron fluence is at the time of the latest withdrawn capsule.
Analysis of St. Lucie Unit I Capsule 284'
6-41 Table 6-16 C, Values for the St. Lucie Unit 1 Sensor Reaction Rate Evaluation CU-63 TI-46 FE-54 NI-58 CO-59 U-238 970 Capsule EOC 5 Cycles 1-4 ALL MATERIALS = 1.032 Cycle 5 ALL MATERIALS = 0.899 104' Capsule EOC 9 Cycles 1-4 0.803 0.810 0.846 0.856 1.304 0.926 Cycle 5 0.672 0.673 0.695 0.702 1.038 0.752 Cycles 6-9 1.279 1.272 1.230 1.219 0.687 1.136 284' Capsule EOC 15 Cycles 1-4 0.834 0.847 0.903 0.917 1.682 1.023 Cycle 5 0.698 0.703 0.741 0.752 1.339 0.831 Cycles 6-9 1.328 1.330 1.313 1.306 0.885 1.255 Cycle 10 0.973 0.967 0.943 0.937 0.614 0.893 Cycle 11 0.859 0.843 0.806 0.799 0.487 0.748 Cycle 12 0.941 0.926 0.890 0.883 0.548 0.830 Cycle 13 0.886 0.870 0.832 0.826 0.506 0.773 Cycle 14 0.984 0.969 0.930 0.923 0.575 0.868 Cycle 15 1.015 1.013 0.994 0.989 0.663 0.947 Analysis of St. Lucie Unit 1 Capsule 284°
7-1 7
SURVEILLANCE CAPSULE REMOVAL SCHEDULE The following surveillance capsule removal schedule meets the intent of ASTM E185-82 and is recommended for future capsules to be removed from the St. Lucie Unit 1 reactor vessel. This recommended removal schedule is applicable to 32 EFPY of operation.
Notes:
(a) Updated in Capsule 284' dosimetry analysis.
(b) Effective Full Power Years (EFPY) from plant startup.
(c) Plant specific evaluation.
(d) Capsule 263' will reach a EOL (32 EFPY) fluence of 2.55 x 1019 n/cm 2 (E > 1.0 MeV) at approximately 23 EFPY.
(e) Capsules 2770 and 83' will reach an EOL license renewal (48 EFPY) fluence of 3.74 x 1019 n/cm 2 (E > 1.0 MeV) at 35 EFPY. Thus, at this time Capsule 830 should removed/tested, while Capsule 277' should be removed and placed in storage.
Analysis of St. Lucie Unit I Capsule 2840 TABLE 7-1 St. Lucie Unit 1 Reactor Vessel Surveillance Capsule Withdrawal Schedule Removal Time Fluence Capsule Location Lead Factor(a)
(EFPY)(bl (n/cm2,
E > 1. 0 MeV)()*
970 970 1.53 4.67 5.91 x 108 (c) 1040 1040 0.98 9.515 9.18 x 108 (C) 2840 2840 1.00 17.23 1.45 x 1019 (c) 2630 2630 1.36 23 2.55 x 1 0 19 (d) 830 830 1.36 35 3.74 x 10"9(e) 2770 2770 1.36 Standby (e)
8-1 8
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- 1.
Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials, U.S.
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- 2.
Code of Federal Regulations, 10CFR50, Appendix G, Fracture Toughness Requirements, and Appendix H, Reactor Vessel Material Surveillance Program Requirements, U.S. Nuclear Regulatory Commission, Washington, D.C.
- 3.
TR-F-MCM-005, "Florida Power and Light Company St. Lucie Unit 1 - Evaluation of Baseline Specimens Reactor Vessel Materials Irradiation Surveillance Program", S.D. Crossman, August 9, 1984.
- 4.
Section XI of the ASME Boiler and Pressure Vessel Code, Appendix G, Fracture Toughness Criteria for Protection Against Failure
- 5.
ASTM E208, Standard Test Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA
- 6.
CENPD-39, "Summary Report on Manufacture of Test Specimens and Assembly of Capsules For Irradiation Surveillance of Hutchinson Island Plant - Unit 1 Reactor Vessel Materials", A.D. Emery, April 1972.
- 7.
ASTM E185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, E706 (IF), in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 1993.
- 8.
ASTM E23-98, Standard Test Methods for Notched Bar Impact Testing ofMetallic Materials, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 1998.
- 9.
ASTM A370-97, Standard Test Methods and Definitions for Mechanical Testing of Steel Products, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 1997.
- 10.
ASTM E8-99, Standard Test Methods for Tension Testing ofMetallic Materials, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 1999.
- 11.
ASTM E21-92 (1998), Standard Test Methods for Elevated Temperature Tension Tests ofMetallic Materials, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 1998.
- 12.
ASTM E83-93, Standard Practice for Verification and Classification of Extensometers, in ASTM Standards, Section 3, American Society for Testing and Materials, Philadelphia, PA, 1993.
- 13.
RSIC Computer Code Collection CCC-650, "DOORS 3.1 One, Two-and Three Dimensional Discrete Ordinates Neutron/Photon Transport Code System, ", August 1996.
- 14.
RSIC DLC-185, "BUGLE-96 Coupled 47 Neutron, 20 Gamma-Ray Group Cross-Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications", March 1996 Analysis of St. Lucie Unit I Capsule 2840
8-2
- 15.
R. E. Maerker, et al., Accounting for Changing Source Distributions in Light Water Reactor Surveillance Dosimetry Analysis, Nuclear Science and Engineering, Volume 94, Pages 291 308, 1986.
- 16.
ASTM Designation E482-89 (Re-approved 1996), Standard Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1999.
- 17.
ASTM Designation E560-84 (Re-approved 1996), Standard Recommended Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1999.
- 18.
A STM Designation E693-94, Standard Practice for Characterizing Neutron Exposures in Iron and Low Alloy Steels in Terms of Displacements per Atom (dpa), in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1999.
- 19.
ASTM Designation E706-87 (Re-approved 1994), Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standard, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1999.
- 20.
ASTM Designation E853-87 (Re-approved 1995), Standard Practice for Analysis and Interpretation ofLight-Water Reactor Surveillance Results, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1999.
- 21.
ASTM Designation E261-96, Standard Practicefor Determining Neutron Fluence Rate, Fluence, and Spectra by Radioactivation Techniques, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1999.
- 22.
ASTM Designation E262-97, Standard Method for Determining Thermal Neutron Reaction and Fluence Rates by Radioactivation Techniques, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1999.
- 23.
ASTM Designation E263-93, Standard Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Iron, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1999.
- 24.
ASTM Designation E264-92 (Re-approved 1996), Standard Method for Measuring Fast-Neutron Reaction Rates by Radioactivation ofNickel, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1999.
- 25.
ASTM Designation E481-97, Standard Method for Measuring Neutron-Fluence Rate by Radioactivation of Cobalt and Silver, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1999.
Analysis of St. Lucie Unit I Capsule 2840
8-3
- 26.
ASTM Designation E523-92 (Re-approved 1996), Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Copper, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1999.
- 27.
ASTM Designation E704-96, Standard Test Method for Measuring Reaction Rates by Radioactivation of Uranium-238, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1999.
- 28.
ASTM Designation E1005-97, Standard Test Method for Application and Analysis ofRadiometric Monitors for Reactor Vessel Surveillance, in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1999.
- 29.
Letter from R.S. Boggs to Ed Terek supplying Tcold data and monthly operating history reports from 01/90 through 12/00 for St. Lucie Unit 1, March 15, 2000.
- 30.
Letter from R.S. Boggs to Ed Terek supplying axial power distributions and relative assembly powers for St. Lucie Unit No. 1 Cycles 110, 11 and 12, April 4, 2000.
- 31.
F. A. Schmittroth, FERRETData Analysis Core, HEDL-TME 79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979.
- 32.
W. N. McElroy, S. Berg and T. Crocket, A Computer-Automated Iterative Method ofNeutron Flux Spectra Determined by Foil Activation, AFWL-TR-7-4 1, Vol. I-IV, Air Force Weapons Laboratory, Kirkland AFB, NM, July 1967
- 33.
RSIC Data Library Collection DLC-178, "SNLRML Recommended Dosimetry Cross-Section Compendium", July 1994.
- 34.
EPRI-NP-2188, Development and Demonstration of an Advanced Methodology for LWR Dosimetry Applications, R. E. Maerker, et al., 1981.
- 35.
Letters FPL-01-145, from D.H. Warren to R.A. Symes, "Model Errors in Fluence Calculations for Surveillance Capsules", Dated 10/31/01.
[Sub-
Reference:
Letter FPL-01-146, Surveillance Capsule Analysis Applicability Time of PT Curves Given Corrected Fluence Values]
Analysis of St. Lucie Unit I Capsule 284°
APPENDIX A INSTRUMENTED CHARPY IMPACT TEST CURVES
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