ML20127N513
| ML20127N513 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 09/30/1992 |
| From: | Shaun Anderson, Lau F WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20127N507 | List: |
| References | |
| WCAP-13520, NUDOCS 9301290219 | |
| Download: ML20127N513 (161) | |
Text
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n Egeket No. 50-213 ~
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Attachment No. 1 Haddam Neck P1 ant.
WCAP-13520 Reactor Cavity Neutron Measurement Program for Connecticut Yankee Atomic Power Company 4
January 1993 9301290219 930115 PDR ADOCK 0500G213' P PDR c.- -]
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WCAP-13520 WESTINGHOUSE CLASS 3 REACTOR CAVITY NEUTRON MEASUREMENT PROGRAM FOR CONNECTICUT YANKEE ATOMIC POWER COMPANY HADDAM NECK PLAlli Stanwood L. Anderson September 1992 Work performed under Shop Order No. CIGP-450 APPROVED:
F. L. Lau, Manager Radiation and Systems Analysis Prepared by Westinghouse for the Connecticut Yankee Atomic Power Company-Purchase Order No. 874899 WESTINGHOUSE ELECTRIC-CORPORATION Energy Systems Business Unit P.O. Box 355 Pittsburgh, Pennsylvania 15230 C 1992 Westinghouse Electric Corporation All Rights Reserved
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EXECUTIVE
SUMMARY
At the conclusion of Fuel Cycle 15 and coincident with the removal of the thermal shield from the reactor internals, a reactor cavity measurement program was instituted at Connecticut Yankee to provide continuous monitoring of the beltline region of the reactor pressure vessel and reactor vessel support structure. .The use of the cavity measurement program coupled with available dosimetry data from five surveillance capsules withdrawn prior to removal of the thermal shield provides a plant.
specific data base that enables the evaluation of the_ vessel exposure and the uncertainty associated with that exposure over the service life of the unit.
At the conclusion of fuel Cycle 16, the first irradiated set of reactor
-cavity dosimetry was removed and analyzed. The results of the evaluation of this cavity dosimetry data set coupled with the analysis of the five previously withdrawn surveillance capsule dosimetry sets indicated that the materials comprising the beltline region of the Connecticut Yankee reactor had accumulated the following maximum fast neutron exposures through the first 6 cycles of operation (17.5 Effective Full Power Years):
BELTLINE MATERIAL EXPOSURE AT 17.5 EFPY FLUENCE (E > 1.0 MeV) IRON DISPLACEMENTS MATERIAL in/cm21 Idoal N0ZZLE SHELL COURSE All Plates 2.03E+19 0,0327 10 Deg. Long. Weld 1.51E+19 0.0243 130 Deg. Long. Weld 3.81E+18 0.00595 250 Deg. Long. Weld 7.20E+18 0.0114 UPPER CIRC. WELD 2.03E+19 0,0327 INTERMEDIATE SHELL COURSE All Plates 4.01E+19 0.0644 70 Deg. Long. Weld 1.42E+19 0.0225 190 Deg. Long. Weld 2.98E+19 0.0479
,, O Deg. Long. Weld 7.50E+18 0.0117 LOWER CIRC. WELD 1.41E+19 0.0227 LOWER SHELL COURSE All P1ates 1.41E+19 0.0227 10 Deg. Long. Weld 1.06E+19 0.0168 130 Deg. Long. Weld 2.64E+18 0.00413 250 Deg. Long. Weld 5.00E+18 0.00794 Based on the continued use of a non-low leakage fuel management approach using zirconium clad fuel assemblies, the projected maximum fast neutron exposure of the vessel beltline materials at 32 effective full power years of cperation is summarized as follows:
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BELTLINE MATERIAL EXPOSURE AT 32 EFPY FLUENCE (E > 1.0 MeV) IRON DISPLACEMEN15 MATERIAL In/cm21 idoal N0ZZLE SHELL COURSL All Plates 4.98E+19 0.0750 10 Deg. Long. Weld 3.71E+19 0.0557 130 Deg. Long. Weld 9.74E+18 0.0146 250 Deg. Long. Weld 1.84E+19 0.0277 UPPER CIP.C. WELD 4.98E+19 0.0750 INTERMEDIATE SHELL COURSE All Plates 9.82E+19 0.148 70 Deg. Long. Weld 3.63E+19 0.0546 190 Deg. Long. Weld 7.31E+19 0.110 310 Deg. Long. Weld 1.92E+19 0.0287 LOWER CIRC. WELD 3.47E+19 0.0522 LOWER SHELL COURSE All Plates 3.47E+19 0.0522 10 Deg. Long. Weld 2.57E+19 0.0387 130 Deg. Long. Weld 6.77E+18 0.0102 250 Deg. Long. Weld 1.28E+19 0.0193 These exposure projections for future operation could be significantly reduced by a transition to a low leakage fuel management strategy employing high burnup fuel assemblies on the periphery of the core. The degree of neutron flux reduction afforded by the use of low leakage loading patterns depends on the time of implementation as well as on the charistics of the specific fuel assemblies located on the core periphery.
Neutron flux reductions approaching a factor of 2.0 have been achieved using conventional low leakage fuel management at a number of operating plants.
As further data is accumulated from subsequent cavity dosimetry irradiations, the neutron environment in the vicinity of the Connecticut Yankee pressure vessel will become better characterized and the uncertainties in the vessel exposure projections will be reduced. Thus, the measurement program will permit the assessment of vessel condition to be based on realistic exposure levels with known uncertainties and will eliminate the need for any unnecessary conservatism in the determination of vessel operating parameters.
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TABLE OF CONTENTS' Page-1 TABLE OF CONTENTS ' i:
LIST OF FIGURES iii LIST OF TABLES y 1.0 OVERVIEW 0F THE PROGRAM l
2.0 DESCRIPTION
OF THE MEASUREMENT PROGRAM 2 2.1 Description of Reactor Cavity Dosimetry 2-1 2.2 Description of Surveillance Capsule Dosimetry- 2-7 3.0 NEUTRON TRANSPORT AND DOSIMETRY EVALUATION METHODOLOGY 3-1 3.1 Neutron Transport Analysis Methods 3-1 3.2 Neutron Dosimetry Evaluation Methodology 3-7 4.0 RESULTS OF NEUTRON TRANSPORT CALCULATIONS 4-1 5.0 EVALUATIONS OF SURVEILLANCE CAPSULE 00SIMETRY 5-1 5.1 Measured Reaction rates E1 5.2 Results of the Least Squares Adjustment Procedure 5-2 5.3 Consistency Check on Least Squares Adjustment Results. 5-3 '
6.0' EVALUATIONS OF' REACTOR LAVITY DOSIMETRY 6-1 6.1. Cycle 16 Results 16-1 7.0 COMPARISON OF CALCULATIONS WITH MEASUREMENTS 7-1.
7.1 Comparison of Least Squares Adjustment Results - 7-2 with Calculation 7.2 Ccmpariso.ns of Measured and Calculated Sensor 7-2 Reaction Rates 1
- . ~ . . .-. - - . - _ . - . . _ - . . ._ - . . . . . . . ..-..- ~.-
t TABLE OF CONTENTS Page 8.0 BEST ESTIMATE NEUTRON EXPOSURE OF PRESSURE VESSEL MATERIALS 8-1 8.1 Exposure Distributions Within the Beltline Region 8-1 8.2 Exposure of Specific Beltline Materials 8-20 ,
8.3 Uncertainties in Exposure Projections 8-35 ,
9.0 REFERENCES
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APPENDIX A MEASURED SPECIFIC ACTIVITY AND IRRADIATION HISTORY A-1 0F SURVEILLANCE CAPSULE SENSOR SETS APPENDIX B MEASURED SPECIFIC ACTIVITY AND 1RRADIATION HISTORY B-1
- 0F REACTOR CAVITY SENSOR SETS P
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a 5
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.. N LIST OF FIGURES Fioure Title Paae l
2.1-1 Azimuthal Location of Sensor Strings 2-4 j 2.1-2 Axial Location of Multiple Foil' Sensor Sets 2-5 2.1-3 Irradiation Capsule for Cavity Sensor Sets 2-6 2.2-1 Neutron Sensor Locations Within Internal 2-8 Surveillance Capsules 3.1-1 Reactor Geometry Showing a 45' Sector 3-5 3.1-2 Internal Surveillance Capsule Geometry 3-6 6.1-1 Fast Neutron Flux (E >-1.0 MeV) as a Function 6-17 of Axial Position Along the O Degree Traverse
- Cycle 16 Irradiation 6.1-2 Fast Neutron Flux (E > 1.0 MeV) as a function 6-18 of Axial Position Along the 7.5 Degree Traverse
- Cycle 16 Irradiation 6.1-3 Fast Neutron Flux (E > 1.0 MeV) as a Function 6-19 of Axial Position Along the 37.5 Degree Traverse
- Cycle 16 Irradiation 6.1-4 Fast Neutron Flux (E > 1.0 MeV) as a Function 6-20 of Axial Position Along the 45 Degree Traverse
- Cycle 16 Irradiation 6.1-5 Fast Neutron Flux (E > 1.0 MeV) as a function 6-21 of Axial Position Along the 90 Degree Traverse
- Cycle 16 Irradiation 6.1-6 Fast Neutron Flux (E > 1.0 MeV) as a Function 6-22 of Axial Position Along the 180 Degree Traverse
- Cycle 16 Irradiation 6.1-7 Fast Neutron Flux (E > 1.0 MeV) as-a Function 23 of Axial Position Along the 270 Degree Traverse
- Cycle 16 Irradiation 8.2-1 Maximum Fast Neutron Fluence (E > 1.0 MeV) as a 8-26
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Function of Azimuthal Angle at the Inner Radius of the Intermediate Shell Plates iii
' .i ilST OF FIGURES Fioure Title Pace 8.2-2 Maximum f ast Neutron Fluence (E > 0.1 MeV) as a- 8-27 Function of Azimuthal Angle at the Inner Radius of the Intermediate Shell Plates -
8.2-3 Maximum Iron Atom Displacements [dpa] as a 8-E8 Function of Azimuthal Angle at the Inner Radius of the Intermediate Shell Plates 8.2-4 Maximum fast Neutron Fluence (E > 1.0 MeV) as a 8-29' Function of Azimuthal Angle at the Inner Radius' of the Upper Circumferential Weld anithe Nozzle Shell Course 8.2-5 Maximum fast Neutron Fluence (E > 0.1 MeV) as a 8-30 Function of Azimuthal Angle at the Inner Radius of the Upper Circumferential Weld and the Nozzle Shell Course-8.2-6 Maximum Iron Atom Displacements (dpa] as a 8-31 Function of' Azimuthal Angle at the Inner Radius of the Upper Circumferential Wold and the Nozzle Shell Course 8.2-7 Maximum fast Neutron Fluence (E > 1.0 MeV) as a 8-32' <
Function of Azimuthal Angle at the Inner Radius of the Lower Circumferential Weld and the Lower Shell Course 8, b 'd 8-33 Maximum Fast Neutron Fluence (E > 0.1 MeV) as. a' Function of Azimuthal Angle at the Inner Radius of the Lower Circumferential Weld and the lower Shell Course 8.2-9 Maximum Iron Atom Displacements (dpa] as a 8-34
~ Function of Azimuthal Angle at the Inner Radius of the Lower Circumferential Weld and the Lower Shell' Course iv-j 1
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LIST OF= TABLES Table Title _. Pace 4.1-1 Calculated Reference Neutron Energy Spectra at :
4-4 Cavity Sensor Set Locations 4.1 Calculated Neutron Sensor Reaction Rates-and 4 Exposure Rates at the Cavity Sensor Set Locations 4.1-3 Calculated Reference Neutron Energy spectra at 4-6 Surveillance Capsule locations 4.1-4 Calculated Niutron Sensor Reaction Rates and 4-7 Exposure Rates at the Center of the Surveillance Capsules 4.1-5 Radial Gradient Corrections for Sensors Contained 4-8 in Connecticut Yankee Internal Surveillance Capsules 4.1-6 Summary of Calculated Exposure Rates at the 4-9 Location of Midplane Cavity Sensor Sets 4.1-7 Summary of Calculated-Exposure Rates at the 4-10 Pressure Vestol Clad / Base Metal Interface 4.1-8 Summary of Calculated Exposure Rates at the 4-11 Center of Surveillance Capsules 4.1-9 Relative Radial Distribution of Neutron Flux 4 12 (E > 1.0 MeV) Within the Pressure Vessel Wall 4.1-10 Relative Radial Distribution of Neutron Flux 4-13 (E > 0.1 MeV) Within the Pressure Vessel Wall 4.1-11 Relative Radial Distribution of Iron Displacement 4-14.
Rate (dpa)-Within the Pressure Vessel Wall-5.1-1 Summary of Reaction Rates Derived from Multiple 5-5 Foil Sensor Sets Withdrawn from Internal
. Surveillance Capsules 5.2-1 Derived Exposure Rates from Surveillance Capsule A 5-6 Withdrawn at the End of Fuei-Cycle 1
, 5.2-2 Derived Exposure Rates from Surveillance Capsule F 5-7 Withdrawn at the End of Fuel Cycle 4 5.2-3 Derived Exposure Rates from Surveillance Capsule H 5-8 withdrawn at the End of-Fuel Cycle 7 v
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r E 't LIST OF TABLES Table ' Title Paae 5.2-4 Derived Exposure Rates from Surveillance Capsule D--5 Withdrawn at the End of Fuel Cycle 10 5.2-5 Derived Exposure Rates from Surveillance Capsule E 5-10 Withdrawn at the End of fuel Cycle 15 5.3-1 Normalized Neutron Sensor Reaction Rates corrected- 5-11 to the Capsule Center for Capsules E, D, H, F, and A 5.3-2 Reference Reaction Rate Data Set for Surveillance 5 Capsules Positioned at 43.5 Degrees '
5.3-3 Derived Exposure Rates from the Dosimetry 5-12 Evaluation of the Reference Reaction Rate Data Set 5.3-4 Calculated Spectrum Averaged Reaction Cross- 5-13 Sections and Exposure Parameter Ratios at the 43.5 Degree Surveillance Capsule Location 5.3-5 Fast Neutron exposure Parameters Derived-Using the 14 Spectrum Averaged Reaction Cross-Section Approach 5.3.6 Comparison of Exposure Parameters Derived Using 5-15 FERRET Approach with Results Using the Spectrum Averaged Cross-Section Method 6.1-1 Summary of Reaction Rates Derived from Multiple 6-4 Sensor Sets Irradiated During Cycle 16 6.1-2 Fe-54 (n p) Reaction Rates Derived from the 6-5 Stainless Steel Gradient Chains -Irradiated During -
Cycle 16 6.1-3 Ni-58 (n,p) Reaction Rates Derived from the. 6-6
,_ Stainless Steel Gradient Chains Irradiated During Cycle 16 6.1-4 Co-59 (n,y)-Reaction Rates Derived-from the .6-7 Stainless Steel Gradient Chains Irradiated During Cycle 16 g vi L
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LIST OF TABLES Table _ Title -Pace 6.1-5 Derived Exposure Rates from the Capsule B 6 Dosimetry Evaluation 0 Degree Azimuth-
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- Core Midplane 6.1-6 Derived Exposure Rates from the Capsule D 6-9 Dosimetry Evaluation 7.5 Degree Azimuth
- Core Midplane ,
6.1-7 Derived Exposure Rates from the Capsule E 6-10.
Dosimetry Evaluation 37.5 Degree Azimuth
- Core Midplane 6.1-8 Derived Exposure Rates from the Capsule F 6-11 Dosimetry Evaluation 45 Degree Azimuth
- Core Hidplane 6.1-9 Derived Exposure Rates from the Capsule A 6-12 Dosimetry Evaluation 0 Degree Azimuth
- Core Top 6.1-10 Derived Exposure Rates from the Capsule C 6-13 ,
l Dvsimetry Evaluation 0 Degree Azimuth l - Core Bottom l 6.1-11 Fast Neutron Flux (E > 1.0 MeV) as a Function 6-14 of Axial Position Within the Reactor Cavity t - Cycle 16 Irradiation 6.1-12 Fast Neutron F. lux (E > 0.1 MeV) as a Function 6-15 of Axial Position Within the' Reactor Cavity
- Cycle 16 Irradiation 6.1-13 Iron Atom Displacement- Rate as a Function 6-16 of Axial Position Within the Reactor Cavity-
- Cycle 16 Irradiation 7.1-1 . Comparison of Measured and Calculated Exposure. 7-4 Rates from Surveillance Capsule and Cavity Dosimetry Irradiations vii
. ~ . . . - .. .
j,= ,
. LIST OF TABLES' Table _ Tit 1.e_ Pace 7.2-1 Comparison of Measured and Calculated Neutron 7 Sensor Reaction Rates From Surveillance Capsule ,
and Cavity Dosimetry irradiations 8.1-1 Summary of Best Estimate Fast Neutron (E > l.0 MeV) 8-5 Exposure Projections for the Beltline Region of_
the Connecticut Yankee Reactor Pressure Vessel 0 Degree Azimuthal Angle 8.1-2 Summary of Best Estimate Fast Neutton (E > 1.0 MeV) 8-6 Exposure Projections for the Beltline Region of the Connecticut Yankee Reactor Pressure Vessel
- 10 Degree Azimuthal Angle 8.1-3 Summary of Best Estimate Fast Neutron (E > 1.0 MeV) 8-7 Exposure Projections for the Beltline Region-of the Connecticut Yankee Reactor Pressure Vessel-
- 20 Degree Azimuthal Angle ,
8.1-4 Summary of Best Estimate fast Neutron (E > l 0 MeV)_8-8 Exposure Projections for the Beltline Region of the Connecticut Yankee Reactor Pressure Vessel
- 40 Degree Azimuthal Angle-8.1-5 Summary of Best Estimate Fast Neutron (E > l.0 MeV)l8-9 Exposure Projections for the Beltline Region _of the Connecticut Yankee _ Reactor Pressure Vessel
- 45 Degree Azimuthal Angle 8.1-6 Summary of Best Estimate Fast Neutron (E > 0.1 MeV) 8-10 Exposure Projections for the Beltline Region of the Connecticut Yankee Reactor-Pressure-Vessel
- 0 Degree Azimuthal Angle 8.1-7 Summary of Best Estimate fast Neutron (E > 0.1 MeV) _8 Exposure Projections for the Beltline Region:of the Connecticut Yankee Reactor Pressure Vessel
- 10 Degree Azimuthal ' Angle viii
LIST OF TABLES i
Table Title Paae 8.1-8 Summary of Best Estimate Fast Neutron (E > 0.1 MeV) 8-12 Exposure Projections for the Beltline Region of the Connecticut Yankee Reactor Pressure Vessel
- 20 Degrae Azimuthal Angle .
8.1-9 Summary of Best Estimate Fast Neutron (E > 0.1 MeV) 8-13 Exposure Projections for the Beltline Region of the Connecticut Yankee Reactor Pressure Vessel
- 40 Degree Azimuthal Angle 8.1-10 Summary of Best Estimate fast Neutron (E > 0.1 MeV) 8-14 Exposure Projections for the Beltline Region of the Connecticut Yankee Reactor Pressure Vessel
- 45 Degree Azimuthal Angle 8.1-11 Summary of Best Estimate Iron Atom Displacement 8-15 Exposure Projections for the Beltline Region of the Connecticut Yankee Reactor Pressure Vessel
- 0 Degree Azimuthal Angle 8.1-12 Summary of Best Estimate Iron Atom Displacement 8-16 Exposure Projections for the Beltline Region of the Connecticut Yankee Reactor Pressure Vessel
- 10 Degree Azimuthal Angle 8.1-13 Summary of Best Estimate Iron Atom Displacement 8-17 Exposure Projections for the Beltline Region of the Connecticut Yankee Reactor Pressure Vessel
- 20 Degree Azimuthal Angle ,
8.1-14 Summary of Best Estimate Iron Atom displacement 8-18 Exposure Projections for the Beltline Region of the Connecticut Yankee Reactor Pressure Vessel
- 40 Degree Azimuthal Angle 8.1-15 Summary of Best Estimate Iron Atom Displacement 8-19 Exposure Projections for the Beltline Region of the Connecticut Yankee Reactor Pressure Vessel
- 45 Degree Azimuthal Angle ix I
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LIST 0F. TABLES Table Title Paa e ._
8.2-1 Maximur f ast Neutron- Exposure of Connecticut. 8 Yankee Beltline Circumferential Welds 8.2-2 Maximum fast Neutron Exposure of Connecticut 8-23 Yankee Intermediate Shell Plates 8.2-3 Maximum fast Neutron Exposure of Connecticut 8-24 Yankee Nozzle Course and Lower Shell Plates 8.2-4 Maximum Fast Neutron Exposure of Connecticut. 8 q Yankee Celtline Longitudinal Welds 1
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- '; g SECTION.I.0-OVERVIEW 0F THE PROGRAM The Reactor Cavity Neutron Measurement Program [1] initiated at Connecticut Yankee at the start of fuel Cycle 16 was designed to provide a-mechanism for.the long term. continuous monitoring of.the neutron exposure of those portions of the reactor vessel and shield tank which may experience radiation induced increases in reference nil ductility transition temperature-(RTNDT) over the nuclear power plant lifetime.
When used in conjunction with dosimetry from previously withdrawn internal surveillance capsules and with the results of neutron transport calculations, the reactor cavity neutron dosimetry provides the means for determination of the neutron exposure of the pressure vessel and the projection of embrittlement gradients through the vessel wall with a '
minimum uncertainty. Minimizing the uncertainty.in-the neutron exposure projections will, in turn, help to assure that the reactor can be operated in the least restrictive mode possible with respect to 1 - 10CFR50 Appendix G pressure / temperature limit curves for normal heatup and cooldown of-the reactor coolant system.
2 - Emergency Response Guidline (ERG) pressure / temperature limit Curves.
3 - Pressurized Thermal Shock (PTS) RTPTS_ screening criteria.
In addition, an accurate measure of the neutron exposure of the reactor vessel and shield tank can provide a sound basis for requalification should. operation of the plant beyond the current design and/or licensed lifetime prove to be desirable.
The Connecticut Yankee reactor vessel materials surveillance program [2]
consisting of several surveillance capsules attached to the-thermal shield in the downcomer region near the pressure vessel wall was in place at the -
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initial startup of the' reactor and remained in service through the completion _of Fuel Cycle 15. During the Cycle 15/16 refueling outage the thermal shield and the remaining surveillance capsules'were removed from the reactor due to mechanical problems with the shield supports, Consequently.the internal surveillance capsule irradiations within the Connecticut Yankee reactor were terminated at that time.
In order to provide a continuing surveillance of the Connecticut Yankee reactor vessel materials, an integrated surveillance program was initiated ]
subsequent to the removal of the thermal shield. The materials portion of' the integrated program consisted of the re-encapsulation of a set of test j specimens irradiated at Connecticut Yankee for the first fifteen fuel cycles and the_ subsequent insertion of those specimens into a vacant capsule position at the Millstone Unit 3 reactor. Continued fluence monitoring of the Connecticut Yankee reactor vessel itself was accomplished by the initiation of the Reactor Cavity Measurement Program. ;
Within the nuclear industry it has been common practice to base estimates of the fast neutron exposure of pressure vessels either directly on the results of neutron transport calculations or on the analytical results.
normalized to measurements obtained from internal surveillance capsules.
However, there are potential drawbacks associated with both of these approaches to exposure assessment.
In performing neutron transport calculations for pressurized water reactors, several design and operational variables have an impact on the
- magnitude of the analytical prediction of exposure rates within the-pressure vessel wall as well as on the uncertainties associated with that
] prediction. Of particular note in this regard are cycle to cycle
- variations in core power distributions (particularly with implementation
[ of low leakage loading patterns), variations of water temperature in the downcomer regions 'of the reactor internals, and deviations-in as-built
. versus design-dimensions for the reactor internals and pressure vessel.
- _ The. manner in which thase important variables are treated in -the analysis j may lead to an increased uncertainty in the exposure evaluations for the-i pressure vessel; and, these increased uncertainties may well result in the a
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.use of overly conservative estimates of vessel embrittlement in thu-assessment of' pressure temperature limitations as well as of the expected:
service life of the component.
The neutron dosimetry contained in the internal surveillance capsules ~.
provide measurement capability to determine the fast neutron exposure of >
the materials test specimens also located within the capsules, but at the same time produced measured data only at a single locatiun within the reactor geometry. Therefore, the surveillance capsule dosimetry, by itself, cannot provide information cegarding the azimuthal, radial, and.
axial gradients of neutron exposure within the pressure vessel.
Furthermore, data from internal surveillance capsules are, by design, obtained at rather infrequent intervals; and surveillance measurement locations may not be in proximity to critical areas on the pressure vessel. These limitations place a heavy reliance on analytical results tu project exposure levels to the vessel wall as well as to provide predictions of vessel exposure for time periods beyond the last capsule ,
withdrawal.
With the addition of supplementary passive neutron sensors in the reactor cavity annulus between the reactor vessel wall and the biological shield, the deficiencies in both surveillance dosimetry and analytical prediction !
can be mitigated and the uncertainties associated with exposure estimates for the pressure vessel can be minimized. With state of the art neutron sensors deployed to establish the absolute magnitude.of the azimuthal and axial exposure rate distributions in the reactor cavity, the burden placed on the_ neutron transport calculation is reduced to the determination of relative neutron energy spectra for sensor set interpretation and relative.
. spatial distributions for extrapolation of the measurement results to positions at the inner radius and through the thickness of the pressure vessel- wall . Studies have shown that the operational and design variables cited above that have a strong impact on the calculated magnitude of exposure rates have only a minor effect on both the interpretatian of cavity dosimetry and on the extrapolation of measurement results to key vessel locations. It is possible, therefore, to employ cavity measurements and a set of reference neutron transport calculations to 1-3
produce vessel exposure projections with a reduced uncertainty over that inherent in an approach based on analysis alone, furthermore, since the cavity neutron measurements are not directly tied to the materials surveillance program, measurement intervals can be chosen to easily provide integral vessel exposure over plant lifetime.
The use of fast neutron fluense (E > 1.0 MeV) to correlate measured materials properties changes to the neutron exposure of the material for light water reactor applications 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 enargy spectra between surveillance capsule locations ano positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves as well as to a more accurate evaluation of damage gradients through the pressure vessel wall.
Because of this potential shift away from threshold fluence toward an energy dependent damt.ge 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 ([ > 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 E093, " Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom".
The application of the dpa parameter to the assessment of embrittlement gradients has already been promulgated in Revision 2 to Regulatory Guide 1.99, " Radiation Damage to Reactor Vessel Materials".
With the aforementioned views in mind, the Reactor Cavity Measurement Program was established to meet the following objectives:
1 - Det W ine azimuthal and axial gradients of fast neutron exposure over the beltline region of the reactor pressure vessel.
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l 2 - Provide measurement capability sufficient to allow the i determination of pressure vessel exposure in terms of both '
fluence (E > 1.0 MeV) and iron displacements per atom. .
3 - Establish a methodology for the projection of exposure :
gradients through the thickness of the pressure vessel wall. 1 4 - Provide a long term continuous monitoring capability for the beltline region of the pressure vessel.
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This report provides the results of neutron dosimetry evaluations i performed subsequent to the completion of fuel Cycle 16. Fast neutron I exoosure in terms of fast neutron fluence (E > 1.0 MeV) and dpa is f established for all measurement locations in the reactor cavity. The analytical formalism describing the relationship among the measurement points and-locations within the pressure vessel wall is described and used to project the Cycle 16 exposure of the vessel itself.
P Results of exposure evaluations from surveillance capsule dosimetry -
withdrawn at the end of fuel Cycles 1, 4, 7, 10, and 15 are incorporated to provide the integrated exposure of the pressere vessel from plant startup through the end of Cycle 16. Also, uncertainties associated with !
the derived exposure parameters at the measurement locations and with the ,
1 projected exposure of the pressure vessel are provided. '
19 addition to the evaluation of the current exposure of the reactor -
vessel beltline materials, projections of the future exposure of the vessel are provided. Curren' evaluations and future projections are L provided for the beltline circumferential welds as well as for the nozzle shell course, intermediate and lower shell plates, and the associated t longitudinal welds that comprise the highly irradiated portions of the reactor vessel.
'When applying the results presented in this report, it should be noted that during the course of the operating history of the Connecticut Yankee '
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reactor a power uprating from 1473 HWt to 1825 MWt took place. in the evaluations presented in this report, all transport calculations and dosimetry an& lyses were completed based on operation at the 1825 MWt power level. To compensate for these higher flux levels during the lower power operating periods the effective full power irradiation time for the entire plant life was also indexed to a reference power icvel of 1825 HWt. Thus, the resultant fluence levels (product of flux and equivalent full power operating time) are accurate for the entire operating period.
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SECTION 2.0 a DESCRIPTION Of THE MEASUREMENT PROGRAM 2.1 - Description of Reactor Cavity Dosimetry To achieve the goals of the Reactor Cavity Neutron Measurement program, comprehensive multiple foil sensor sets including radiometric monitors '
(RM) and solid state track recorders (SSTR) were installed at several locations in the reactor cavity to characterize the neutron energy spectra i within the beltline region of the reactor vessel. In addition, gradient chains were used in conjunction with the encapsulated sensors to complete the azimuthal and axial mapping of the neutron environment over the regions of interest.
Placement of the multiple foil sensor sets was such that spectra evaluations could be made at four azimuthal locations within a single octant at an axial elevation representative of the midplane of the reactor core. The intent here was to determine changes in spectra caused by varying amounts of water located between the core and the pressure vessel. Due to the irregular shape of the reactor core, water thickness varies significantly as a function of azimuthal angle, in addition to the four midplane sensor sets, two multiple foil packages were positioned opposite the top and bottom of the active core at the azimuthal angle corresponding to the maximum neutron flux within the octant. Here the !
intent was to measure variations in neutron spectra over the core height; particularly near the top of the fuel where backscattering of neutrons
-from primary loop nozzles and vessel support structures could produce significant perturbations. At each of the four azimuthal locations selected for core midplane spectra measurements, gradient chains extended over a fourteen foot height centered on the core midplane. As a check of symetry in the neutron ficid, additional gradient chains were also located at the maximum flux location within each of the remaining three reactor quadrants.
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i 2.1.1 Sensor Placement in the Reactor Cavity A detailed description of the cavity dosimetry hardware and plant specific installation can be found in Reference 1. However, the following i information is provided in this report to orient the reader to the plant '
geometry and the specifics of the sensor sets.
The placement of the individual multiple foil sensor sets and gradient chains within the reactor cavity is illustrated in figures 2.1-1 and 2.1-2. In figure 2.1-1 a plan view of the azimuthal locations of the seven strings of sensor sets is depicted. The strings were located at azimuthal positions of 0, 7.5, 37.5, 45, 90,180, and 270 degrees relative to the plant 0.0 azimuth (called West). The sensor strings were hung in the annular gap between the pressure vessel insulation and the shield tank at a nominal radius of 93 inches relative to the core centerline. !
In figure 2.1-2 the axial extent of each of the sensor set strings is illustrated along with the locations of the multiple foil holders. At the O degree azimuth, multiple foil sets were positioned at the core midplane and opposite the top and bottom of the active fuel. At the 7.5, 37.5, and !
45 degree azimuthal locations, multiple foil sets were positioned only opposite the core midplane. At each of these four locations as well as at the 90, 180, and 270 degree azimuths, stainless steel gradient chains extended i 7 feet relative to the midplane of the active core.
The sensor sets and gradient chains were suspended from local stainless-steel dosimetry attachment plates that were in turn affixed to the stainless steel reactor vessel reflective insulation using four No.14 x 3/4-inch long self-tapping stainless steel screws. The seven individual attachment plates (one for each dosimetry traverse) were installed in the vicinity of the reactor vessel nozzles at plant elevation 13'-3".
Due to a complete lack of air flow in the annular region external to the reactor vessel, the bottom ends of the dosimetry chains were not secured to stationary plant features. The weight of the freely hanging chains ,
were sufficient to maintain azimuthal and radial positioni19 of the sensor sets.
2-2
_ _ _ _ _ _ . . ~ _ ___ _ _,__ _
~
- i. 3
- u * ; ... of Irradiation Capsules v
, ti osor sets used to characterize the neutron spectra within the
[ + %or cavity were retained in 3.87 inch x 1.00 inch x 0.50 inch rectangular aluminum 6061 capsules such as that shown in Figure 2.1-3.
Each capsule included three compartments to hold the neutron sensort. The top compartment (position 1) was intended to accomodate bare radiometric monitors and SSTR packages, whereas, the two remaining compartments (positions 2 and 3) were meant to house cadmium shielded packages. The separation between positions 1 and 2 was such that cadmium shields inserted into position 2 did not introduce perturbations in the thermal flux in position 1. Aluminum 6061 was selected for the dosimeter capsules, in order to minimize neutron flux perturbations at the sensor set locations as well as to limit the radiation levels associated with post-irradiation shipping and handling of the capsules. A summary of the contents of the multiple foil capsules used during each cycle of irradiation is provided in the appendices to this report.
2.1.3 - Description of Gradient Chains Along with the multiple foil sensor sets placed at discrete locations within the reactor cavity, gradient chains were cmployed to obtain axial variations of fast neutron exposure along each of the seven traverses.
Subsequent to irradiation these gradient chains were removed from the cavity and segmented to provide neutron reaction rate measurements at one foot intervals over the height of the axial traverses. These gradient chains consisted of Type 304 stainless steel bead chain of 0.188 inch diameter. When coupled with a chemical analysis, the stainless stac1 yielded activation results for the fe-54 (n p), Ni-58 (n p), and Co-59 (n,y) reactions. The high purity iron, nickel, and cobalt-aluminum foils contained in the multiple foil sensor sets established a direct correlation with the measured reaction rates from the stainless steel chain; and provided an overcheck on the chemical analysis of the Type 304 steel.
2-3
FIGURE 2.1-1 AZlHUTHAL LOCATION Of SENSOR STRINGS NOZZLE
. __ 1 . SHELL 180 90 270 0
INTERMEDIATE
. SHELL
\
. . LOWER SHELL
. Indicates Dosimetry Location 2-4
'4 4 FICVRE 2.1-2 AXIAL LOCATION Of MULTIPLE f0ll SENSOR SETS AZIMUTHAL ANGLE (Degrees) p$$
0.0 7.5 37.G 45.0 90.0 180.0 270.0 o
d
- 2 g x a
w d 8
u di w
5 -
g x x x x D B!
q x E
h)
@ X 6 Muhl-foil Set gradient chain 2-5 1
i
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+ .!
f FIGURE'2.1-3 I
i IRRADIATION CAPSVLE FOR CAVITY SENSOR SETS i
, t 1.L M A.1 4'!h 1 J S 1 1 'Th4 8 4 ,
..o t.at 4 % .It4.Stuut..th Twau
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NI . 4 ' II ' l 2.2 - Description of Surveillance Capsule Dosimetry Dyer the course of the first 15 fuel cycles at Connecticut Yankee, five materials surveillance capsules were withdrawn from their positions between the thermal' shield and the reactor vessel. The neutron dostmetry contained within these capsules provided a measure of the integral exposure received by each of the capsules during its respective irradiation period; and established a measurement continuity between the startup of the reactor and the initiation of the Reactor Cavity Heasurement Program. The specific withdrawal dates of these five capsules were as follows:
Capsule A End of Cycle 1 04/70 Capsule f End of Cycle 4 07/73 Capsule H End of Cycle 7 10/77 Capsule D End of Cycle 10 09/81 Capsule E End of Cycle 15 09/89 The type and location of the neutron sensors included in the materials surveillance program are described in some detail in Reference 2; and, are illustrated schematically in figure 2.2-1 of this report.
Relative to Figure 2.2-1, copper, nickel, and cobalt-aluminum monitors, in wire form, were placed in holes drilled in spacers at several axial levels within each capsule. The cadmium-shielded uranium and neptunium fission monitors were accomodated within a dosimeter block located near the center of the capsule, in addition to these high purity sensors, iron dosimeters were also obtained by removing samples from several charpy test specimens from various locetions within the capsule. Specific information pertinent to the individual sensor sets included in Capsules A F, H D, and E are provided in the appendices to this report.
2-7
FIGURE 2.2-1 NEUTRON SENSOR LOCATIONS WITHIN INTERNAL SUP.VEILLANCE CAPSVLES 7
SPACER RADIUS (Cm)
WIRES - charpy O charpy - 183.77 CENTER - X 184.14 WIRES - charpy O charpy - 184.77 2-8
4 a SECTION 3.0 NEU1RON TRANSp0RT AND 00SIMETRY EVALUATION NETH000LOGIES 3.1 - Neutron Transport Analysis Nethods fast neutron exposure calculations for the reactor and cavity geometry were carried out using a series of forward discrete ordinates transport calculations representative of a reactor geometry both with and without a thermal shield in place. These calcu17.tions provided the relative energy distribution of neutrons and gamma rays for use as input to neutron
~
dosirnetry evaluations as well as for use in relating measurement results to the actt.a1 exposure at key locations in the pressure vessel wall; and established the means to compute absolute exposure rate values using fuel cycle specif,c core power distributions; thus, providing a direct comparison with all dosimetry results obtained over the operating history
-of the reactor.
In combination, this series of computations provided the means to: .;
1 - Evaluate neutron dosimetry from reactor cavity and surveillance capsule locations.
2 - Enabic a direct comparison of analytical prediction with measurement.
3 - Extrapolate dosimetry results to key locations at the inner radius and through the thickness of the pressure vessel.
4 - Establish a mechanism for projection of pressure vessel exposure as the design of each new fuel cycle evolves.
3-1
3.1.1 - Forward Calculation for Cycles 1 through 15 4 A plan view of the reactor geometry at the core midplane elevation is shown in Figure 3.1-1 with the thermal shield in place. Since the reactor exhibits 1/8 core symmetry only a 0-45 degree sector is depicted. In addition to the core, reactor internals, pressure vessel, and the shield tank, the model also included explicit representations of the surveillance capsules, the pressure vessei cladding, and the mirror insulation located external to the vessel.
A description of a single surveillance capsule attached to the thermal ~
shield is shown in Figure 3.1-2. From a neutronic standpoint, the inclusion of the surveillance capsules and essociated support structures in the analytical model is significant. Since the presence of the capsules and structure has a marked impact on the magnitud of the neutron flux as well as on the relative neutron and gamma ray energy spectra at dosimetry locations within the capsules, a meaningful comparison of-measurement and calculation can be made only if these perturbation effects are properly accounted for in the analysis, in contrast to the relatively massive stainless steel and carbon steel structures associated with the internal surveillance capsules, the small aluminum capsules used in the reactor cavity measurement program were designed to minimize perturbations in the neutron flux and, thus, to provide free field data at the measurement locations. Therefore, explicit modeling of these small capsules in the forward transport model was not required.
The forward transport calculation for the reactor model depicted in Figures 3-1 and 3-2 was carried out in R,0 geometry using the DOT two-dimensional discrete ordinates code (3) and tk SAILOR cross-section library (4). The SAILOR library is a 67 group coupled notitron-gamma ray ENDFB-IV based data set produced specifically for light water reactor applications. In these analyses, anisotropic scattering was treated with a P3 expansion of the cross-sections and the angular discretization was 3-2 1
.. . = . . - _ . . _ ___-_ - -
modeled with an 58 order of angular quadrature. The reference forward calculation was normalized to a core power density characteristic of operation at a thermal power level of 1825 MWt.
The core power distribution used in the analysis of Fuel Cycles 1 through 15 represented plant specific data provided by Northeast Utilities. This power distribution information, obtained for each operating cycle of the reactor, was burnup weighted to derive a composite distribution representative of the average operating conditions for the plant over the first 15 fuel cycles. This approach permitted the use of the single transport calculation to compute the integrated exposure of the reactor vessel during the period of operation with the thermal shield in place..
In i similar fashion, supplied axial power distributions were also iveraged to provide a composite distribution that could be used as a relative weighting function to derive axial distributions of fast neutron exposure over the beltline region of the reactor.
3.1.2 - Forward Calculations for Cycles 16 and Beyond.
In addition to the forward calculation applicable to Fuel Cycles 1 through 15, two additional computations were carried out with a geometric model characteristic of reactor operation without a thermal shield. In the first instance, a calculation using the Cycles I through 15 average core power distribution was performed with the thermal shield removed. The results of this evaluation were reresentative of continued operation of the reactor using a fuel management approach that does not-incorporate the l
low leakage approach. Since no commitment to future operation with low leakage has been made at this time, these computational results were used ,;
in conjunction with measurement data provided in Sections 5.0 and 6.0 of this report to provide projections of vessel exposure for time periods L subsequent to the end of Cycle 16.
l The second forward calculation using a geometry with the thermal shield y removed was carried out using the plant specific cycle 16 core power L
! 3-3
distribution supplied by Northeast Utilities. The results of this computation provide information necessary for direct comparison of analytical results with cavity dosimetry measurements obtained during Cycle 16. The power distribution utilized in this case incorporated some high burnup fuel assemblies on the core periphery as a one time implementation. Thus, while appropriate for the Cycle 16 comparisons, the results of this calculation were not representative of long term future reactor operation.
Both forward calculations with the thermal shield removed were also carried out using an S8 order of angular quadrature and the P3 _
cross-section approximation from the SAILOR library. Thus, in all respects, these calculations were consistent with the baseline analysis discus.ed in Section 3.1.1.
3-4
- / --- _ - . - - - - - _ - - - - - - - - - - ~
pinuRE 3.1-1 g g(OMETRY SHO W A o'
aurit!!
45' jjtio>i C7 0p V ,
{l SSEL T
Hyp /
7777777 S N/p(O
,,,. i r
/
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- r<>,,,,1 C "i ~~ SURVEILLANCE
, / capsule l
/
/,-"
- REr.CTOR /$
CORE
/
/
~
/
/
/
p 3-5
flGURE 3.1-2 INTERNAL SURVEILLANCE CAPSULE GE0 METRY SPACER RADIUS (Cm)
WIRES - charpy O charpy 183.77 CENTER - X 184.14 WIRES . charpy O charpy 184.77 3-6
3.2 - Neutron Dosimetry Evaluation Methodology The use of passive neutron sensors such as those included in the internal surveillance capsule and reactor cavity dosimetry sets does not yield a direct treasure of the energy dependent neutron flux level at the measurement location. Rather, the activation or fission process is a measure of the integrated offect 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 flux level and, hence, time integrated exposure (fluence) experienced by the sensors may be developed from the neasurements only if the sensor characteristics and the _
parameters of the irradiation are well known. In particular, the following variables are of interest:
1 - The measured specific activity of each sensor 2 - The physical characteristics of each sensor 3 - The operating history of the reactor 4 - The energy response of each sensor 5 - The neutron energy spectrum at the sensor location In this section the procedures used by Westinghouse to determine sensor s .ific activities, to develop reaction rates for individual sensors from the measured specific activities and the operating history of the reactor, and to derive key fast neutron exposure parameters from the measured reaction rates are described, for the most part these procedures apply to all of the evaluations provided in this report. However, in some cases, the specific activities pertaining to individual internal serveillance capsules were determined from prior analysis by a radiochemical laboratory other than Westinghouse., in those cases, the source of the measured specific activity data was referenced and the remainder of the data evaluation proceeded using the methodology described in this section.
3-7
3.2.1 - Determination of Sensor Reaction Rates following irradiation, thn multiple foil sensor sets from surveillance capsule and cavity irradiations along with reactor cavity gradient chains were recovered and transported to Pittsburgh for evaluation. Analysis of all radiometric foils and gradient chains was performed at the Westinghouse Analytical Services Laboratory; while the evaluation of the SSTR sensors from the cavity irradiations was carried out at the Westinghouse Science and Technology Center Track Recorder Laboratory. -
i 3.2.1.1 - Radiometric Sensors The specific activity of each of the radiometric sensors and gradient chain segments was determined using established AS1H procedures [5 through-15), following sample preparation and weighing, the specific activity of each sensor was determined by means of a lithium drifted germanium, Ge(Li), gamma spectrometer. In the case of the surveillance capsule and cavity multiple fall sensor sets, these analyses were performed by direct counting of each of the individual foils or wires; or, as in the case of U-238 and Np-237 fission monitors from internal surveillance capsules, by direct counting preceded by dissolution and chemical separation of cesium from the sensor. For the stainless steel gradicat chains used in the cavity irradiations, individual sensors were obtained by cutting the-chains into a series of segments to provide data points at one foot intervals over an axial span encompassing 17 feet relative to the reactor core midplane.
The irradiation history of the reactor over its operating lifetime was obtained from NUREG-0020, " Licensed Operating Reactors Status Summary Report". In particular, operating. data were extracted from that report on a monthly bases from' reactor startup to the end of the current evaluation period. For the sensor sets utilized in surveillance capsule and reactor cavity irradiat lons, the half-lives of the product isotopes are long ~
enough that a Conthly histogram describing reactor operation has proven to be an adequate representation for use in radioactive decay corrections for l
3-8 y we r p . pi w u..d---T >
n
'. i t
the reactior.s of interest in the exposure evaluations.
I hav ug the measured specific activities, the operating history of the reactor, ;
an6 the physical characteristics of the sensors, reaction rates referenced to ;
full power operation at 1825 MWt were determined from the following equation!
= A R
il 0 fY [(Pj/ Pref) Cj (1-e-Atj)[g-Atq) j t-where: A = nsasswd succMic activity (dps/gm)
R = reaction iate overaged over the irradiation neriod and referenceu to operation at a core power lev.I of Pref (rps/ nucleus).
fio number of target element atoms per gram of sensor. -
F = weight fraction of the target isotope in the sensor material.
Y o number of product atoms produced per reaction.
Pj = average core power level during irradiation period j (MW). ,
Pref = maximum or reference core power level of. the reactor (MW).
Cj = calculated ratio of ( (E > 1.0 MeV) during irradiation period j to the time weighted average 4 (E > 1.0 MeV) over the entire irradiation period.
A = decay constant of the product isotope (sec'I).
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 total irradiation period.
In the above equation, the ratio Pj/ Pref accounts for month by month variation of power level within a given fuel cycle. The ratio Cj is calculated for each fuel cycle and accounts for the change in sensor reaction rates caused by variations in flux level due to changes in core power spatial distributions from fuel cycle to fuel-cycle. For a single cycle ,
3-9
- .- - , , - - - --- a. - - -- - - - . . - - --
[
irradiation Cj = 1.0. However, for multiple cycle irradiations, (
particularly those employing low leakage fuel management the additim al Cj correction must be utilized. ,
3.2.1.2 - Solid State Track Recorders ;
following preparation of the mica discs, all of the solid state track recorders were scanned either manually or with the Westinghouse STC Automated Track Scanner to determine the number of fissions that occured during the course of the irradiation of the sensor sets.- Since the SSTR sensors are integrating devices not sJsceptible to radioactive decay of a product isotope, the measurements of total fissions per atom, A, were ,
converted directly to reaction rates using the following relationship:
A R-E[Pj/Prefl tj j
where the denominator in the above equation represents the total effective full power seconds of reactor operation during the irradiation of the solid state track recorders.
The SSTR fissionable deposits were designed for reuse in the long term monitoring program. Therefore, following processing each sensor was carefully examined to assure that the deposits were neither damaged nor contaminated during irradiation, handling, and post-irradiation processing.
In particular, these examinations _were designed to assure that, in all cases, the fission tracks were confined to an-area corresponding to the active portion of the fissionable deposit and that-the edges of-the active area were sharply-defined with a sufficient drop-off-in track density to 3-10
indicate acceptable signal to background ratios for the measurements. '
Each mica SSTR and fissionable deposit was also closely inspected under a microscope to verify that no physical damage had occured during exposure or shipment. Selected deposits were also subjected to mass recalibration to verify that no deposit mass had been lost during shipping or exposure.
9 3.2.1.3 - Corrections to Reaction Rate Data Prior to using the measured reaction rates in the least squares adjustment procedure discussed in Section 3.2.2 of this report, additional corrections were made to the U-238 foil and SSTR measurements to account for the presence of U-235 impurities in the sensors as well as to adjust for the build-in of plutonium isotopes over the course of the irradiation. Likewise, corrections were made to both U-238 and Np-237 sensors to account for gamma ray induced fission reactions occuring over the course of the irradiation. These corrections were location and fluence dependent and were derived from a combination of data from the forward transport calculations and the measurements made with the U-235 solid state track recorders, in performing the dosimetry evaluations for the internal surveillance capsules, the sensor reaction rates measured at the locations shown in -
Figure 2.2-1 were indexed to the geometric center of the capsules prior to use in the spectrum adjustment procedure. This procedure required correcting the measured reaction rates by the application of analytically determined spatial gradients. For the Connecticut Yankee surveillance capsules, the gradient correction factors for each sensor reaction were obtained from the forward transport calculation with the thermal shield in place and were used in a multiplicative fashion to relate individual measured reaction rates to the corresponding value at the geometric center of the surveillance capsule. -In the case of the reactor cavity sensors, all of the monitors were located at the same radial location.- Thus, gradient corrections were not required in the evaluation of these dosimetry sets.
3-11
i 7
3.2.2 - Least Squares Adjustment Procedure j
'i Values of key fast neutron exposure parameters were derived from the ;
measured reaction rates using the FERRET least squares adjustment code ;
[16). The FERRET approach used the measured reaction rate data and the i calculated neutron energy spectrum at the sensor set locations as input j and proceeded to adjust a priori (calculated) group fluxes to produce a #
best fit (in a least squares sense) to the reaction rate data. The ;
exposure parameters along with associated uncertainties were then obtained from the adjusted spectra.
In the FERRET evaluations, a log-normal least-squares algorithm weights j both the a priori values and the measured data in accordance with the i assigned uncertainties and correlations. In general, the measured values, f, are linearly related to the flux, p, by some response matrix, A:-
I t
I (s,o) (s) (a) fj -
[ Ajg p g 9 ,
where i indexes the measured values belonging to a single data set s, g ,
designates the energy group and a delineates spectra that may be -i simultaneously adjusted. For example, Rj - E og g p g
9 relates ~a set of measured reaction rates Rg to a single spectrum pgby the multigroup cross section agg. (In this case,--FERRET.
also adjusts- the cross-sections.)- The log-normal- approach automatically. !
~
3-12 s
,--,,,.~,nn, , e-. 4 w- 4 , ~ . . ,,a-w-v,- - ,,n., n,-,.,,-nA,, . _ , - . . , , , , , , . . - , . , . , ,-,,n., ., ,,c,,,,w, -n, -w, w, --v r w, < w ,
accounts for the physical constraint of positive fluxes, even with large assigned uncertainties. '
In the FERRET analysis of the dosimetry data, the continuous quantities r (i.e., fluxes and cross-sections) were approximated in 53 groups. The calculated fluxos from the reference forward calculation were expanded intotheFERRETgroupstructureusingtheSAND-Ilcode(17]. This procedure was carried out by first expanding the a priori spectrum into '
the SAND-Il 620 group structure using a SPLINE interpolation procedure for interpolation in regions where group boundaries do not coincide. The 620-point spectrum was then easily collapsed to the group scheme used in FERRET.
The cross-sections were also collapsed into the 53 energy-group structure using SAND 11 with calculated spectra (as expanded to 620 groups) as weighting functions. The cross sections were taken from the ENDF/8-V dosimetry file. Uncertainty estimates and 53 x 53 covariance matrices ,
were constructed for each cross section. Correlations between cross sections were neglected due to data and code limitations, but this omission does not significantly impact the results of the adjustment.
For each set of data or a priori values, the inverse of the corresponding relative covariance matrix M is used as a statistical weight, in some
-cases, as for the cross sections, a multigroup covariance matrix is used.
More often, a simple parameterized form is employed:
2 M gg. - Rn+Rg R,P g gg, wh-re Rn specifies an overall fractional normalization uncertainty (i.e... complete correlation) for the corresponding set of values. The fractional uncertainties Rgspecify additioral random uncertainties for group g that are correlated with a correlation matrix: .
3-13
e P gg. - (1-0) 6g9. + 0 el h3 The first term specifies purely random uncertainties while the second term describes short-range correlations over a range 1 (0 specifies the strength of the latter term),
for the a priori calculated fluxes, a short-range correlation of y 6 groups was used. This choice implies that neighboring groups are strongly correlated when 0 is close to 1. Strong long-range correlations (or anticorrelations) were justified based on information presented by R.E.
Haerker(18). Haerker's results are closely duplicated when 7 6.
For the integral reaction rate covariances, sirnple normalization and random uncertainties were combined as deduced from experimental uncertainties.
In performing the least squares adjustment with the FERRET code, the input spectra from the reference forward transport calculation were normalized to the measured re-54 (n.p) Hn-54 reaction rates to remove any constant calculation to measurement bias and, thus, to permit the adjustment to take place on a relative basis. The specific normalization factors for individual evaluations depended on the location of the sensor set as well as on the neutron flux level at that location.
The specific assignment of uncertainties in the measured reaction rates and the input (a priori) spectra used in the FERRET evaluations was as follows:
REACTION RATE UNCERTAINTY 5%
FLUX NORMAll2AT10N UNCERTAINTY 30%
3-14
FLUX GROUP UNCERTAINTIES
_(E > 0.0055 HeV) 30%
(0.68 ev < E < 0.0055 HeV) 58% :
(E < 0.68 ev) 104%
SHORT RANGE CORRELATION (E > 0.0055 HeV) 0.9 (0.68 ev < E < 0.0055 HeV) 0.5 (E < 0.68 ev) 0.5 FLUX GROUP CORRELATION RANGE (E > 0.0055 HeV) 6 (0.68 ev < E < 0.0055 HeV) 3 (E < 0.68 ev) 2 It should be noted that the uncertainties listed for the upper energy ranges extend down to the lower range. Thus, the 58% group uncertainty in the second range is made up of a 30% uncertainty with a 0.9 short range correlation and a range of 6, and a second part of magnitude 50% with a-0.5 correlation and a range of 3.
These input uncertainty assignments were based on prior experience in using the FERRET least squares adjustment approach in the analysis of neutron dosimetry from surveillance capsule, reactor cavity, and benchmark irradiations. The values are liberal enough to permit adjustment of the input spectrum to fit the measured data for all practical applications.
3-15
i imm k
6 4
a- a
6 SECTION 4.0 l
RESULTS OF NEUTRON TRANSPORT CALCULATIONS ,
As noted in Section 3.0 of this report, data from the forward transport calculations performed for the Connecticut Yankee reactor were used in evaluating dosimetry from both reactor cavity and surveillance capsule irradiations as well as in relating the results of these evaluations to the neutron exposure of the pressure vessel wall, in this section, the key data extracted from these forward calculations are presented and their relevance to the dosimetry evaluations and vessel exposure projections is discussed. The reader should recall that the results of the transport calculation with the thermal shield in place should be used in all :
absolute comparisons with surveillance capsule dosimetry; whereas, the results from the computation without the thermal shield and with the Cycle 16 core power distribution are appropriate for direct comparison with the results of the Cycle 16 measurements.
Data from the forward calculations pertinent to cavity and surveillance capsule sensor evaluations are provided in Tables 4.1-1 through 4.1-5.
In Table 4.1-1, the calculated neutron energy spectra applicable to sensor locations at 0.0, 7.5, 37.5, and 45.0 degrees relative to the core cardinal axes are listed. These data represent _the a priori spectra used as the starting guess in the FERRET least squares adjustment evaluations of the cavity sensor sets. On a relative basis these calculated energy distributions establish a baseline against which adjusted spectra may be l compared; and, when coupled with cycle specific absolute calculated results, provide an analytical prediction of absolute neutron spectra at the sensor set locations for each irradiation period. In Table 4.1-2, the calculated neutron sensor reaction rates associated with the spectra from Table 4-1 are provided along with the reference exposure rates in terms of
( (E > 1.0 MeV), p gE < 0.1 HeV) and dpa/sec. Also listed are the associated exposure rate ratios calculated for each of the cavity sensor set locations.
4-1
._.m ,. _ , . _.r -, - , _ _ . ,. - _ _ ,,_ . . . - . , . -,. ._. . ,
l
- The reference reaction rates, exposure rates, and exposure rate ratios were used in conjunction with absolute cycle specific calculations to provide calculated sensor set reaction rates and to project sensor set exposures in terms of p-(E > 0.1 MeV) and.' d pa/sec for each irradiation period, In addition, the ratios of U238(1,f)/U238(n,f) and Np237(7,f)/Np237(n,f) were used to make photo-fission corrections to measured reaction rates in the U238 and Np237 fission monitors prior to
- use in the FERRET adjustment procedure.
In Table 4.1-3, the calculated neutron energy spectra at the geometric center of surveillance capsules located at 43.5 degrees relative to the-core cardinal axes are listed. in Table 4.1-4, the calculated neutron i sensor reaction rates and exposure rate ratios associated with the spectra from Table 4.1-3 are provided along with the calculated exposure rates in '
terms of p (E > 1.0 MeV), p (E < 0.1 MeV) and dpa/sec. Again, these data are applicable to the geometric center of the surveillance capsules. These tabulated data were used in the surveillance capsule dosimetry evaluations and exposure calculations in the same fashion as was the case for the cavity sensor sets.
As noted earlier in this report, surveillance capsule dosimetry ;
evaluations aiso require spatial gradtent corrections to be applied to 1 measured reaction rates in sensors dispersed throughout the capsule.
In the case of the Connecticut Yankee surveillance _ capsules, neutron sensors were positioned within the specimen array as shown in Figure 2.2-1. In Table 4.1-5, gradient correction factors- applicable to the variods dosimetry locations are provided for each sensor reaction. These factors were used in a multiplicative fashion to relate measured reaction-rates to the corresponding value at the geometric center of the capsules.
Data from the forward calculation pertinent to absolute comparisons with measurement are provided in Tables 4.1-6 through 4.1-11. In Tables 4,1-6 '
through 4.1-8 exposure parameters applicable to the core midplane elevation are listed for the Cycle 1-15 calculation with the thermal snield in place, for the Cycle 1-15 calculation with the thermal shield 4-2
)
l
~_ .- _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _
' ~
removed,-and _for the Cycle 16 calculation with the thermal-shield
. removed.
In Table 4.1-6,- the calculated azimuthal distributions of fast neutron !
flux (E > 1.0 MeV), fast neutron flux (E > 0.1 MeV), and iron atom displacement rate (dpa) are listed for the cavity sensor locations.
Similar data. applicable to the pressure vessel clad / base metal interface and the geometric center of the surveillance capsules are given in _ Tables 4.1-7 and 4.1-8, respectively. The data provided in these tables are meant to be used in the absolute comparisons of calculations with midplane capsule and cavity dosimetry results.
Radial gradient information for p (E > 1.0 MeV), p (E > 0.1 Mev),
and dpa/sec is given in Tables 4.1-9, 4.1-10, and 4.1-11, respectively.
These data are presented on a relative basis for each exposure parameter at the 0, 10, 20, 40, and 45 degree azimuthal locations, Exposure rate distributions within the vessel wall can be obtained by normalizing the calculated or projected exposure at the vessel inner radius to the gradient data given in Tables 4.1-9 through 4.1-11.
In regard to the calculated results for the configuration based on the Cycles 1-15 average power distribution with the thermal shield removed, it should be noted that the data are provided solely for use in projecting exposure results beyond the end of Cycle 16. To date, this particular configuration has not been run in the reactor, but the results are represtntative of exposure rates that would be expected if continued use of non-low leakage fuel management were to be continued with the thermal shield removed- and a zirconium clad core were to replace the currently installed stainless steel clad core. As noted later in Section 8.0 of this report a commitment to future operation with low leakage fuel management has not been made at this time. Therefore, the use of the non-low leakage results for future projections are warranted. A future implementation of low leakage fuel management could significantly reduce the end of life exposure projections for the. pressure vessel materials.
4-3
TABLE 4.1-1 CALCULATED REFERENCE NEUTRON ENERGY SPECTRA AT CAVITY SENSOR SET LOCATIONS' 1825 MWt; Fa = 1.0 LOWER AZIMVTHAL ANGLE LOWER AZIMUTHAL ANGLE ENERGY ENERGY
_(Mevi. 0 DEG 7.5 DEG 37.5 DG 45 DEG (Mqyl. 0 DEG J.5 DEG 37.5 DG 45 DEG 1.42E+1 4.62E+5 4.12E+5 1.98E+5 1.94E+5 2.97E-1 4.66E+9 4.21E+9 1.46E+9 1.35E+9 1.22E41 1.66E46 1.49E+6 7.03E+5 6.83E+5 1.83E-1 4.44E+9 4.02E+9 1.43E+9 1.32E+9 1.00E+1 4.59E+6 4.08E+6 1.77E+6 1.70E+6 1.llE-1 5.40E+9 4.90E+9 1.76E+9 1.62E+9 8.61E+0 7.43E+6 6.60E+6 2.74E+6 2.62E+6 6.74E-2 3.37E+9 3.07E+9 1.12E+9 1.04E+9 7.41E+01.05E+7 9.28E46 3.65E46 3.47E+6 4.09E-2 2.56E+9 2.34E+9 8.77E+8 8.12E48
- 6.07E+0 2.00E+7 1.76E+7 6.64E+6 6.28E46 3.18E-2 6.54E+8 6.00E+8 2.32E48 2.15E+8 4.97E+0 2.33E+7 2.06E+7 7.34E+6 6.8SE+6 2.61E-2 2.07E+8 1.91E+8 7.61E+7 7.04E+7 3.68E+0 4.08E+7 3.59E+7 1.21E+7 1.12E+7 2.42E-2 2.09E+9 1.90E+9 6.82E+8 6.29E+8
- 3.01E+0 3.99E+7 3.52E+7 1.16E+7 1.07E+7 2.19E-2 1.62E+9 1.48E+9 5.32E+8 4.91E+8 2.73E40 3.47E+7 3.05E+7 9.76E+6 9.01E+6 1.50E-2 2.14E+9 1.96E+9 7.24E+8 6.69E+8 2.47E+0 4.38E+7 3.87E+7 1.23E+7 1.14E+7 7.10E-3 1.61E+9 1.48E+9 5.84Et8 5.40E+8 2.37E+0 2.23E+7 1.96E+7 6.20E+6 5.71E+6 3.36E-3 1.69E+9 1.57E+9 6.33E+8 5.87E+8 2.35E+0 8.03E+6 7.02E+6 2.12E+6 1.95E+6 1.59E-3 1.26E+9 1.17E+9 4.84E+8 4.50E+8 2.23E+0 4.09E+7 3.60E+7 1.10E+7 1.01E+7 4.54E-4 1.75E+9 1.62E+9 6.89E+8 6.41E+8 1.92E+0 1.03E+8 9.13E+7 2.90E+7 2.67E+7 2.14E-4 7.76E+8 7.23E+8 3.12E+8 2.90E+8 1.65E+0 1.81E+8 1.60E+8 4.99E+7 4.58E+7 1.01E-4 9.01E+8 8.40E+8 3.64E+8 3.39E+8 1.35E+0 2.80E+8 2.48E+8 7.81E+7 7.17E+7 3.73E-5 1.18E+9 1.10E+9 4.81E+8 4.48E+8 1.00E+0 8.58E+8 7.62E+8 2.42E+8 2.22E+8 1.07E-5 1.34E+9 1.25E+9 5.52E+8 5.14E+8 8.21E-1 1.00E+9 8.92E+8 2.87E+8 2.63E+8 5.04E-6 7.05E+8 6.57E+8 2.89E+8 2.69E+8 7.43E-1 3.61E+8 3.24E+8 1.08E+8 9.89E+7 1.86E-6 7.72E+8 7.20E+8 3.15E+8 2.93E+8 6.08E-1 2.50E+9 2.24E+9 7.39E+8 6.77E+8 8.76E-7 4.44E+8 4.13E+8 1.79E+8 1.67E+8 4.98E-1 2.40E+9 2.16E+9 7.29E+8 6.69E+8 4.14E-7 3.33E+8 3.09E+8 1.32E+8 1.23E+8 3.69E-1 2.70E+9 2.44E+9 8.46E+8 7.79E+8 1.00E-7 4.95E+8 4.58E+8 1.87E+8 1.73E+8 0.00 1.20E+9 1.llE+9 4.29E+8 3.94E+8 NOTE
- The upper energy of group 1 is 17.33 Mev.
4-4
TABLE 4.1-2 CALCULATED NEUTRON SENSOR REACTION RATES AND EXPOSURE RATES AT THE CAVITY SENSOR SET LOCATIONS 1825 MWt; Fa = 1.0 AZIMUTHAL-ANGLE 0.0 DEG 7.5 DEG 37.5 DEG 45.0 DEG Reaction Rate (ros/ nucleus)
Cu63(n,a) 1.12E-18 9.95E-19 4.07E-19 3.89E-19 Ti46(n,p) 1.29E-17 1.14E 4.40E-18 -4.17E-18 Fe54(n,p) 6.88E-17 6.08E-17 2.14E-17 2.00E-17 NiS8(n.p) 9.62E-17 8.49E-17 2.95E-17 2.75E-17 U238(n,f) (Cd) 4.50E-16 3.97E-16 1.29E-16 1.20E-16 Np237(n,f) (Cd) 1.06E-14 9.40E-15 3.15E-15 2.90E-15=
CoS9(n,7) 1.22E-13 1.13E-13 4.73E-14 4.39E-14 CoS9(n,7) (Cd) 7.88E-14 7.32E-14 3.16E-14 2.94E-14 U238(7,f) 2.25E-17 2.09E-17 8.88E-18 8.lJE-18 Np237(7,f) 6.47E-17 5.92E-17 2.55E-17 2.29E-17 -
Neutron' Flux (n/cn81 p (E > 1.0 MeV) 1.79E+09 1.59E+09 5.07E+08 4.67E+08
( (E > 0.1 MeV) 2.59E+10 2.33E+10 8.08E+09 7.44E+09
, doa/sec Displacement rate 8.95E-12 7.21E-12 2.48E-12 2.29E-12 Exoosure Rate Ratios p(E > 0.1)/p(E > 1.0) 14.5 14.7 15.9 15.9
[dpa/sec]/p(E > 1.0) 5.00E-21 4.53E-21 4.89E-21 4.90E-21 U238(7,f)/U238(n,f) 0.0501 0.0527 0.0688 0.0675 Np237(7,f)/Np237(n,f) 0.0061 0.0063 0.0081 0.0079 4-5
~
TABLE 4.1 ,
CALCULATED REFERENCE NEUTRON ENERGY SPECTRA AT SURVEILLANCE CAPSULE LOCATIONS 1825 MWt; Fa 1.0 NEUTRON NEUTRON LOWER FLUX LOWER FLUX ENERGY (n/cm2-sec) ENERGY (n/cm2-sec)
(Mevi 43.5 DEG (Mev) 43.5 DEG 14.19 1.60E+07 0.183- 2.28E+10 12.21 5.77E+07 0.111 2.24E+10 10.00 1,98E+08 6.74E-02 1.75E+10 8.61 3.61E+08 4.09E-02 1.26E+10 7.41 5.94E+08 3.18E-02 3.84E+09 6.07 1.33E+09 2.61E-02 2.28E+09 4.97 1.78E+09 2.42E-02 4.37E+09 3.68 3.32E+09 2.19E-02 3.58E+09 3.01 2.76E+09 1.50E-02 5.44E+09 2.73 2.21E+09 7.10E-03 9.15E+09 2.47- 2.57E+09 3.36E-03 1.20E+10 2.37 1.26E+09 1.59E-03 1.24E+10 2.35 3.77E+08 4.54E-04 1.79E+10 2.23 1.86E+09 2.14E-04 9.34E+09 1.92 4.89E+09 1.01E-04 1.02E+10 1.65- 6.23E+09- 3.73E-05 1.38E+10 1.35 9.44E+09 1.07E-05 1.64E+10 1.00 2.01E+10 5.04E-06 9.15E+09 0.821 1.45E+10 1.86E-06 1.06E+10.
0.743 6.84E+09 8.76E-07~6.80E+09 0.608 2.20E+10 4.14E-07 5.67E+09 0.498 1.77E+10 1.00E-07 1.26E+10-0.369 2.11E+10 0.00 3.45E+10 0.297 2.06E+10 NOTE: The upper energy of group 1 is 17.33 Mev.
4-6
.. 3 TABLE 4.1-4 CALCULATED NEUTRON SENSOR REACTION RATES AND EXPOSURE RATES AT THE CENTER OF THE SURVEILLANCE CAPSULES
'1825 MWt, f, = 1.0 A7IMUTHAL ANGLE 43.5 DEGREES Reaction Rate (ros/ nucleus) ;
Cu63(n.o) 5.93E-17 Fe54(n.p) 4.45E-15 NiS8(n,p) 5.90E-15 U238(n,f) (Cd) 2.01E-14 Np237(n,f) (Cd) 1.67E-13 CoS9(n,7) 2.04E-12 CoS9(n,7)- (Cd) 9.20E-13 U238(7,f) 6.04E-16 Np237(7,f) ).68E-15 NeutronFlux(n/cnd1' '
p (E > 1.0 MeV) 5.95E+10 p (E > 0.1 MeV) 2.llE+11 doa/sec Displacement rate 1.03E-10 i
L l
Exposure Rate Ratios
( p(E'> 0.1)/p(E > 1.0) 3.55 l (dpa/sec)/p(E > l'.0) 1.73E-21 i
U238(7,f)/U238(n,f) 3.00E-02 Np237(7, f)/Np237(n, f) 1.01E-02 4-7
TABLE ~4.1-5 RADIAL GRADIENT CORRECTIONS FOR SENSORS CONTAINED IN' CONNECTICUT. YANKEE INTERNAL SURVEILLANCE CAPSVLES' RADIAL LOCATION (cm) 183.77 184.14 184.77 43.5 DEGREE CAPSULE Cu63(n,o) 0.934 1.000 1.109 fe54(n p) 0.921 1.000. 1.138.
NiS8(n,p). 0.918 1.000 1.143 U238(n,f) (Cd) 1.000 Np237(n,f) (Cd) 1.000 "
CoS9(n,7) 1.176 1.000 0.707 CoS9(n,7) (Cd) 0.962 1.000 1.040 f
4-8
TABLE 4.1-6
$UMMARY Of CALCULATED EXPOSURE RATES AT THE LOCATION OF MIDPLANE CAVITY SENSOR SETS; 1825 MWt, Fa = 1.15 THETA flux (n/cm2-sec)
(Dea) (E > 1.0) (E > 0.1) doa/sec CYCLE l-15 WITH THERMAL SHIELD 0.0 1.05E+09 2.17E+10 6.84E-12 7.5 9.21E+08 1.94E+10 6.12E-12 37.5 2.69E+08 5.92E+09 1.87E-12 45.0 2.44E+08 5.38E409 1.70E-12 CYCLE l-15 WITHOUT THERMAL SHIELD 0.0 1.98E+09 2.90E+10 9.49E-12 7.5 1.75E+09 2.61E+10 8.53E-12 37.5 5.60E+08 9.02E+09 2.94E-12 45.0 5.15E+08 8.31E+09 2'.70E-12 CYCLE 16 WITHOUT THERMAL SHIELD 0.0 1.59E409 2.36E+10 7.72E-12 7.5 1.45E+09 2.17E+10 7 llE-12 37.5 5.81E+08 9.19E+09 3.00E-12 45.0 5.41E+08 8.57E+09 2.79E-12 4-9 i
TABLE 4.1-7
SUMMARY
OF CALCULATED EXPOSURE RATES AT THE PRESSURE VESSEL-CLAD / BASE METAL INTERFACE; 1825 MWt, a F - 1.15-THETA Flux (n/cm2-sec)
(Dea) (E > 1.0) -(E > 6'.1) doa/ set _
CYCLE l-15 WITH THERMAL SHIELD 0.0 6.65E+10 2.05E+11 1.15E-10 7.5 5.72E+10 1.76E+11 9.90E-11 10.0 4.90E+10 1.51E+11' 8.49E-Il 20.0 2.29E+10 6.70E+10 3.91E-Il 37.5 1.28E+10 3.59E+10 2.16E-11 40.0 1.20E+10 3.32E+10 2.01E-11 45.0 1.14E+10 3.14E+10 1.91E-Il CYCLE 1-15 WITHOUT THERMAL SHIELD 0.0 1.18E+11 2.55E+11 1.82E-10 7.5 1.03E+11 2.21E+11 1.56E-10 10.0 8.82E+10 1.69E+11 1.35E-10 20,0 4.50E+10- 9.57E+10- 6.98E-11 37.5 2.54E+10 5.31E+10 3.96E-Il 40.0 2.38E+10 4.97E+10 3.70E 45.0 2.32E+10 4.78E+10 3.61E CYCLE 16 WITHOUT THERMAL SHIELD 0.0 9.19E+10 1.98E+11 1.40E-10 ,
7.5 8.48E+10- 1.82E+11 1.29E-10 10,0 7.53E+10 1.62E+11 1.15E-10 20.0 4.37E+10 9.25E+10 6.76E-11 37.5 2.69E+10- 5.60E+10 4.17E-11 40.0 2.53E+10 .5.27E+10 3,93E-Il-45.0 2.47E+10 5.llE+10 3.85E-Il l~
4-10
TABLE 4.1-8
SUMMARY
OF CALCULATED EXPOSURE RATES AT THE CENTER OF SURVEILLANCE CAPSULES; 1825 MWt, Fa - 1.15 Flux (n/cm2-sec)
(E > 1.0) (E > 0.1) doa/see CAPSULE A 5.69E+10 2.02E+11 9.90E-10 CAPSULE F 6.19E+10 2.20E+11 1.08E-11 CAPSULE H 6.72E+10 2.39E+11 1.17E-ll CAPSULE D 6.73E+10 2.39E+11 1.17E-11 CAPSULE E 6.82E+10 2.42E+11 1.19E-Il 4-11
TABLE 4.1-9 RELATIVE RADIAL DISTRIBUTION OF NEUTRON FLUX (E > 1.0 MeV)
WITHIN THE PRESSURE VESSEL WALL RADIUS AZIMUTHAL ANGLE (cm) 0. DEGB11.S 15 DEGREES 30 DEGREES 45 DEGREES 168.04 1.000 1.000 1.000 1.000 168.27 0.986 0.988 0.987 0.988 168.88 0.972 0.976 0.975 0.975 169.75 0.922 0.929 0.927 0.928 170.89 0.841 0.850 0.847 0.850 172.17 0.734 0.744 0.740 0.745 173.49 0.621 0.631 0.627 0.633 174.90 0.519 0.530 0.525 0.531 176.30 0.426 0.437 0.431 0.438 177.50 0.348 0,359 0.353 0.359 178.91 0.291 0.302 0.296 0.302 180.42 0.235 0.247 0.242 0.246 181.51 0.187 0.197 0.192 0.196 182.60 0.156 0.167 0.163 0.165 183.90 0.129 0.141 0.137 0.139 184.55 0.101 0.113 0.111 0.112 Note: Base Metal Inner Radius - 168.04 cm.
1/4 T Location - 172.17 cm.
1/2 T Location - 176.30 cm.
3/4 T Location - 180.42 cm.
Base Metal Outer Radius = 184,55 cm.
4-12
c .
TABLE 4.1-10 RELAYlVE RADIAL DISTRIBUTION OF NEUTRON FLUX (E > 0.1 MeV)
WITHIN THE PRESSURE VESSEL WALL RADIUS AZIMUTHAL ANGLE (cm) 9 DEGREES 15 DEGREES 3_Q_ DEGREES 45 DEGREES 163.04 1.000 1.000 1.000 1.000 168.27 1.005 1.006 1.007 1.007 168.88 1.009 1.013. 1.014 1.014 169.75 1.006 1.012 1.014 1.015 170.89 0.981 0.992 0.996 0.998 172.17 0.934 0.949 0.955 0.959 173.49 0.872 0.891 0.898 0.903 174.90 0.803 0.827 0.834 0.840 176.30 0.729 0.757 0.764 0.769 177.50 0.657 0.687 0.695 0.700 178.91 0.595 0.629 0.638 0.641 180.42 0.524 0.562 0.572 0.572 181.51 0.452 0.492 0.503 0.500 182.60 0.400 0.442 0.454 0.450 183.90 0.347 0.393 0.407 0.400 184.55 0.282 0.331 0.350 0.343
'l Note: Base Metal Inner Radius = 168.04 cm.
1/4 T Location - 172.17 cm.
1/2 T Location = 176.30 cm.
3/4 T Location = 180.42 cm.
Base Metal Outer Radius = 184.55 cm.
4-13
. g TABLE 4.1-11 RELATIVE RADIAL DISTRIBUTIONS OF IRON DISPLACEMENT RATE (dpa)
WITHIN THE PRESSURE VESSEL WALL RADIUS AZIMUTHAL ANGLE (cm) 0 DEGREES 15 DEGREES 30 DEGREES 45 DEGREES 168.04 1.000 1.000 1.000 1.000 163.27 0.988 0.990 0.990 0.990 168.88 0.977 0.981 0.980 0.980 169.75 0.937 0.945 0.943 0.943 170.89 0.874 0.886 0.882 0.883 172.17 0.790 0.806 0.801 0.803 173.49 0.701 0.720 0.713 0.715 174.90 0.617 0.639 0.630 0.632 176.30 0.536 0.561 0.552 0.553 177.50 0.465 0.431 0.482 0.482 178.91 0.409 0.437 0.429 0.428 180.42 0.350 0.380 0.372 0.370 181.51 0.294 0.324 0.318 0.314 182.60 0.256 0.287 0.282 0.277 183.90 0.219 0.251 0.249 0.243 184.55 0.177 0.210 0.211 0.205 Note: Base Metal Inner Radius = 168.04 cm.
1/4 T Location = 172.17 cm.
1/2 T Location - 176.30 cm.
3/4 T Location - 180.42 cm.
Base Metal Outer Radius - 184.55 cm.
4-14 l
l
SECTION 5.0-EVALUATIONS OF SVRVEILLANCE CAPSULE DOSIMETRY In this section, the results of the evaluations of the five neutron sensor sets withdrawn as a part of the Connecticut Yankee Reactor Vessel
= Materials Surveillance Program are presented. The capsule designation, location within the reactor, and time of withdrawal of each of these dosimetry sets were as follows:
AZIMUTHAL WITHDRAWAL -IRRADIATION CAPSULE ID LOCATION TIME- TIME (EFPS)
A 43.5 DEGREES END OF CYCLE 1 5.33E+07 F 43.5 DEGREES END OF CYCLE 4 7.67E+07 H 43.5 DEGREES END OF CYCLE 7 2.40E+08 0 43.5 DEGREES END OF CYCLE 10 3.43E+08 E 43.5 DEGREES END OF CYCLE 15 5.20E+08 x 5.1 - Measured Reaction Rates With the exception of Capsules A and F, radiometric counting of each of these data sets was accomplished by Westinghouse using the procedures discussed in Section 3.0 of this report. The measured specific activities are included in Appendix A to this-report. Radiometric counting of the L sensors from Capsule A and F, on the other hand, was carried out by the Battelle Memorial Institute [19, 20]. However, in this case, the measured I,pecific activities were not published.
The-irradiation history of the Connecticut Yankee reactor during.the first
- 15. fuel cycles is also listed in Appendix A. The irradiation-history was obtained from NUREG-0020, " Licensed Operating Reactors Status Summary
'~
. Report" for the applicable operating periods. In addition to.the reactor power history, for the multiple cycle irr::diations (Capsules F, H, D, and I
5-1 L
I
4 S), the flux' level adjustment factors for each cycle are also tabulated in.
Appendix A.
Based on the irradiation history and associated flux level adjustment factors, the individual sensor characteristics, and the measured specific activities, reaction rates averaged over the appropriate irradiation periods and referenced to a core power level of 1825 MWt were computed for the sensor sets-removed from Capsules H, 0, and E. In the case of-Capsules A and f, reaction rates were developed directly from the derived neutron flux and spectrum averaged reaction cross-sections reported in References 38 and 39. The computed reaction rates for the multiple _ foil sensor sets from each of the five internal surveillance capsules are provided in Table 5.1-1.
In regara to the data listed in Table 5.1-1, the fission rate measurements for the U-238 sensors include corrections for U-235 impurities, for the build-in of Plutonium isotopes during the long irradiations, and for the effects of 7,f reactions. Likewise, the fission rate measurements for the Np-237 include adjustments for 7,f reactions occuring'over the-respective irradiation periods.
5.2 - Results of the Least Squares Adjustment Procedure The results of the application of the least squares adjustment procedure to the five sets of surveillance capsule dosimetry are provided in _ Tables 5.2-1 through 5.2-5. In these tables, the derived exposure experienced by each capsule along with data illustrating the fit of-both the a priori and adjusted spectra to the measurements are given. 'Also included in the-
, tabulations are the la uncertainties associated with each of the de.-ived exposure rates.
In regard to the comparisons listed _ in Tables 5.2-1 through 5.2-5, it should be_ noted that the columns labeled "a priori calc" were obtained by normalizing the neutron spectral data from Table 4.1-3 to the measured 5-2
Fe-54 (n,p) reaction rates from each sensor set as discussed in Section 3.0. Thus, the comparisons illustrated in Tables 5.2-1 through 5.2-5 indicate only the degree to which the relative neutron energy spectra matched the measured sensor data before and after adjustment. These data are not meant to provide an absolute comparison of calculation and measurement, Absolute comparisons are discussed in Section 7.0 of this report.
5.3 - Consistency Check on least Squares Adjustment Results Due to the sparcity of measured reaction rates for several of the internal surveillance capsule dosimetry sets (Type 1 Capsules such as H and F did t contain fission monitors and, since the specimens were re-encapsulated for further irradiation, no iron data were available from Capsule E), the evaluations described in this section represent a difficult test for the spectrum adjustment approach. Therefore, as a consistency check on the dosimetry results provided in Tables 5.2-1 through 5.2-5 the exposure values derived from the least squares adjustment for each capsule dosimetry set were compared with corresponding values developed using the spectrum averaged cross-section method.
In the spectrum averaged cross-section evaluations, cross-sections were developed from the analysis of a reference reaction rate set derived from the data given in Table 5.1-1. The reference set was developed by first normalizing the data listed in Table 5.1-1 to an Fe-54 (n,p) Mn-54 reaction rate of 4.55E-15 rps/ nucleus and then linearly averaging the five data sets to provide a reference set for use in the spectrum adjustment procedure described in Section 3.0 of this report. In the case of Capsule E no iron data were available for the normalization procedure. Therefore, in the generation of the reference data set, it was assumed that the
[Fe-54 (n,p)/U-238 (n,f)] reaction rate ratio for Capsule E was the same as that measured for Capsule 0; and the normalization of the Capsule E dosimetry set proceeded accordingly.
5-3
v e The normalized data sets for each of the surveillance capsules-are listed in Table 5.3-1 and the averaged reference reaction rate data set is given in Table 5.3-2.
Results of the FERRET evaluation of the reference reaction rate set are provided in Table 5.3-3. As was the case with each of the individual capsule evaluations, in Table 5.3-3 the derived exposure rates as well as data illustrating the fit of the adjusted spectrum to the measured data are listed. From the data listed in Table 5.3-3, spectrum averaged cross-sections referenced to the p (E > 1.0 MeV) threshold flux were computed for each reaction using both the a priori and adjusted spectra and are listed in Table 5.3-4. Also included in Table 5.3-4 are the exposure parameter ratios required to calculate p (E > 0.1 MeV),
dpa/sec, and p (E < 0.414 eV) given the values of 4 (E > 1.0 MeV) derived from the spectrum averaged cross-section approach.
Neutron exposure results based on the spectrum averaged cross-section approach are provided in Table 5.3-5 for each of the surveillance capsules; and comparisons of those results with the-exposure parameters derived using the least squares adjustment methodology are given in Table 5.3-6.
An examination of Table 5.3-6 indicates excellent agreement between the the results of the least squares adjustment approach and the spectrum averaged cross-section evaluation for all fast neutron exposure ,
parameters. The agreement in the thermal flux assessment is not quite as good, but is still well within the lo uncertainties associated with the FERRET evaluations. Overall the comparisons illustrated in Table 5.3-6 demonstrate that, in spite of the sparcity of measured reaction rate data for some of the capsules, an excellent degree of consistency is observed between the two evaluation methods. This observation is also in keeping with prior experience with surveillance capsule dosimetry analyses for other reactors. Therefore, based on these comparisons, there is sufficient confidence in the least . squares adjustment results for these capsules to specify the exposure parameters computed with the FERRET code-as representative of the individual capsule exposures.
5-4 l
TABLE 5.1-1 SJMMARY OF REACTION RATES DERIVED FROM MULTIPLE F0lt SENSOR SETS WITHDRAWN FROM INTERNAL SURVElllA.4CE CAPSULES REACTION RAJE (ros/ nucleus)
CAPSULE CAPSULE CAPSULE CAPSULE CAPSULE PEACTION A F H D E
- Cu63(n,o)Co60 4.64E-17 5.00E-17 4.94E-17 6.07E-17
- Fe54(n,p)Mnb4
- NiS8(n,p)CoS8 3.43E-15 5.35E-15 4.68E-15 4.84E-15 5.14E-15 7.34E-15
]
U238(n,f)Csl37 2.63E-14 2.74E-14 Np237(n,f)Csl37 1.48E-13 1.51E-13 2.02E-13
- CoS9(n,y)Co60 2.77E-12 CoS9(n,y)Co60 1.42E-12 1.36E-12 1.35E-12
- - Bare foil, all others were cadmium covered 5-5
' a.
TA8LE 5.2-1 DERIVED EXPOSURE RATES FROM SURVEILLANCE CAPSULE A 00SIMETRY' WITHDRAWN AT THE END OF F'JEL CYCLE 1 -
A PRIORI ADJUSTED la
, PARAMETER VALUE VALUE UNCERTAINTY Flux (E > 1.0 Mev) 4.48E+10 5.llE+10- 10% _
Flux (E > 0.1 Mev) 1.59E+11 1.83E+11 15%
Flux (E < 0.414 ev) 3.52E+10 3.76E+10 82%
Flux (Total) 3.29E+11 3.67E+11 20% I dpa/sec 7.78E-11 8.87E-Il 11%'
COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES CAPSULE A EVALUATION REACTION RATE (ros/ nucleus) C/M A PRIORI ADJUSTED REACTION MEASURED CALC. CALC. A PRIORI ADJUSTED Fe-54 (n,p) 3.43E-15 3.35E-15 3.58E-15 0.98 1.04 Ni-58 (n,p) 5.35E-15 4.45E-15 5 14E-15
. 0.83 0.96 Np-237 (n,f) (Cd) 1.48E-13 1.26E-13 1.46E-13 0.85 0.99 5-6
L ~,
TABLE 5.2-2 DERIVED EXPOSURE RATES FROM SURVEILLANCE CAPSULE F 00SIMETRY
' WITHDRAWN AT THE END'0F FUEL CYCLE 4 A PRIORI ADJUSTED la PARAMETER - . -VALUE__ _VALUE UNCERTAINTY Flux (E > 1.0 Mev) 6.12E+10 6.90E+10 15%
Flux (E > 0.1 Mev) 2.17E+11 2.40E+11 24%
Flux (E < 0.414 ev) 4.81E+10 4.58E+10 82%
Flux (Total) 4.50E+11 4.64E+11 24%
dpa/sec 1.06E-10 1.17E-10 18%
1 COMPARIS0N OF MEASURED AND CALCULATED SENSOR REACTION RATES CAPSULE F EVALUATION REACTION RATE (ros/ nucleus) C/M A PRIORI ADJUSTED REACTION MEASURED . CALC,__ CALC. A PRIORI ADJUSTED Cu-63 (n,a). 4.64E 6.10E-17 4.77E-17 1.31 1.03 Fe-54 (n,p) 4.68E-15 4.58E-15 4.56E-15 0.98- 0.97 5-7 -
TABLE 5.2 DERIVED EXPOSURE RATES FROM SURVEILLANCE CAPSULE H DOSIMETRY WITHDRAWN AT THE END OF FUEL CYCLE 7 A PRIORI ADJUSTED la PARAMETER VALUE VALUE UNCERTAINTY Flux (E > 1.0 Mev) 6.33E+10 7.14E+10 15%
Flux (E > 0.1 Mev) 2.24E+11 2.54E+11 23%
Flux (E < 0.414 ev) 4.97E+10 5.54E+10 20%
Flux (Total) 4.65E+11 5.29E+11 18%
dpa/sec 1.10E-10 1.23E-10 18%
9 COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES CAPSULE H EVALUATION REACTION RATE (rns/ nucleus) C/M A PRIORI ADJUSTED REACTION MEASURED CALC. -CALC. A PRIORI ADJUSTED Cu-63 (n,a) 5.00E-17 6.30E-17 5.12E-17 1.26 1.02 Fe-54 (n,p) 4.84E-15 4.73E-15 4.74E-15 0.98 0.98 Co-59 (n,7) 2.77E-12. 2.17E-12 2.77E-12 0.78 1.00 C0-59 (n,y) (Cd) 1.42E 9.77E-13 1.42E 0.69 1.00 l
5-8
F ',
TABLE 5.2-4 DERIVED EXPOSURE RATES FROM SURVEILLANCE CAPSULE D DOSIMETRY WITHDRAWN AT THE END OF FUEL CYCLE 10 A PRIORI ADJUSTED la PARAMETER VALUE VALUE UNCERTAINTY Flux (E > 1.0 Mev) 6.72E+10 7.07E+10 8%
Flux (E > 0.1 Mev) 2.38E+11 2.17E+11 14%
Flux (E < 0.414 ev) 5.28E+10 5.28E+10 82%
Flux (Total) 4.94E+11 4.62E+11 17%
dpa/sec 1.17E-10 1.13E-10 10%
i COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES i CAPSULE D EVALUATION i
REACTION RATE (ros/ nucleus) C/M A PRIORI ADJUSTED REACTION MEASURED CALC, CALC. A PRIORI ADJUSTED Cu-63 (n,o) 4.94E-17 6.70E-17 5.14E-17 1.36 1.04-Fe-54 (n,p) 5.14E-15 5.03E-15 5.09E-15 0.98' '0.99 Ni-58 (n p) 7.34E-15 6.67E-15 7.11f-15 0.91 0.97
-U-238 (n,f) (Cd) 2.63E 2.27E-14 2.44E-14 0.86 0.93 Np-237 (n,f)-(Cd) 1.51E-13 1.89E-13 1.71E-13 1.25 1.13 Co-59 (n,y) (Cd) 1.36E-12 1.04E-12 1.35E-12 0.76 1.00 l
5-9
TABLE 5.2-5 DERIVED EXPOSURE RATES FROM SURVEILLANCE CAPSULE E D0SIMETRY WITHDRAWN AT THE END OF FUEL CYCLE 15 A PRIORI ADJUSTED la PARAMETER VALUE VALUE UNCERTAINTY Flux (E > 1.0 Mev) 6.99E+10 7.63E+10 9%
Flux (E > 0.1 Mev) 2.48E+11 2.65E+11 15%
Flux (E < 0.414 ev) 5 .50E+10 5.58E+10 82%-
Flux (Total) 5.14E+11 5.45E+11 17%
dpa/sec 1.21E-10 1.30E-10 10%
COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES CAPSULE E EVALUATION REACTION RATE (ros/ nucleus) . C/M A PRIORI ADJUSTED REACTION MEASURED CALC, CALC. A PRIORI ADJUSTED Cu-63 (n,o) 6.07E-17 6.97E-17 6.15E 1.15 1.01 U-238 (n,f) (Cd) 2.74E-14 2.37E-14 2.55E-14 0.86 0.93 Np-237 (n,f) (Cd) 2.02E-13 _1.97E-13 _2.09E-13 0.98 1.03 Co-59 (n, y) (Cd) 1.35E-12 1.08E-12 1.35E-12 0.80 1.00- '
5-10 l
TABLE 5.3-1 NORMALIZED-NEUTRON SENSOR REACTION RATES CORRECTED'TO THE CAPSULE CENTER FOR CAPSULES E, D, H F, AND A REACTION RATE (rps/ nucleus)
REACTION .(APSULE E CAPSULE D CAPSULE H CAPSULE F CAPSULE A Cu-63(n,a)Co-60 5.16E-17 4.37E-17 4.70E-17 4.51E-17 Fe-54(n,p)Mn-54 4.55E-15 4.55E-15 4.55E-15 4.55E-15 4.55E-15 Ni-58(n.p)Co-58 6.50E-15 7.10E-15 U-238(n,f)Cs-137-(Cd) 2.33E-14 2.33E-14 Np-237(n,f)Cs-137 (Cd) 1.72E-13 1.34E-13 1.96E-13 Co-59(n,1)Co-60 2.60E-12 Co-59(n,1)Co-60 (Cd) 1.15E-12 1.20E-13 1.33E-12 TABLE 5.3 REFERENCE REACTION RATE DATA SET FOR SURVE!LLANCE CAPSULES POSITIONED AT 43.5 DEGREES REFERENCE REACTION RATE REACTION (ros/ nucleus)
Cu-63(n.o)Co-60 4.69E-17 Fe-54(n p)Mn-54 4.55E-15 Ni-58(n.p)Co-58 6.80E-15 U-238(n,f)Cs-137 (Cd) 2.33E Np-237(n,f)Cs-137 (Cd). 1.67E-13 Co-59(n,1)Co-60 2.60E-12 Co-59(n,1)Co-60 (Cd) 1.23E-13 5-11
1:
. e TABLE 5.3-3 DERIVED EXPOSURE RATES FROM THE DOSIMETRY EVALUATION OF THE REFERENCE REACTION RATE DATA SET A PRIORI ADJUSTED la PARAMETER VALUE VALUE MCERTAINTY Flux (E > 1.0 Mev) 5.94E+10 6.74E+10 8%
Flux (E > 0.1 Mev) 2.llE+11 2.22E+11 15%
Flux (E < 0.414 ev) 4.67E+10 5.58E+10 19%
Flux (Total) 4.37E+11 4.69E+11 13%
dpa/sec 1.03E-10 1.llE-10 10%
COMPARIS0N OF MEASURED AND CALCULATED SENSOR REACTION RATES' REFERENCE REACTION RATE DATA SET EVALUATION REACTION RATE (ros/ nucleus) C/M A PRIORI ADJUSTED REACTION MEASURED _ CALC. CALC. A PRIORI ADJUSTED Cu-63 (n,a) 4.69E-17 5.93E-17 4.84E-17 1.26 1.03 Fe-54 (n,p) 4.55E-15 4.45E-15 4.59E-15 0.98 1.01-Ni-58 (n.p) 6.80E-15 5.90E-15 6.54E-15 0.87 0.96 U-238 (n,f) (Cd) 2.33E-14 2.01E-14 2.25E-14 0.86 0.97=
Np-237 (n,f) (Cd) 1.67E-13 1.68E-13' l.77E-13 1.00 1.06-Co-59 (n,1) 2.60E-12 2.04E-12 2.60E-12 0.79 1.00 Co-59 (n,1) (Cd) 1.23E-12 9.19E-13 1.23E-12 0.75 1.00-2
's-12 l
_ - _ ---- -- - \
p .
TABLE 5.3-4 CALCULATED SPECTRUM AVERACED REACTION CROSS-SECTIONS AND EXPOSURE PARAMETER RATIOS AT THE 43.5 DEGREE
~ SURVEILLANCE CAPSULE LOCATION-i CROSS-SECTIONS (barns)
REACTION _A PRIORI SPECTRUM ADJUSTED SPECTRUM h Cu-63 (n,o) 0.000998 0.000718- d Fe-54 (n,p) 0.0749 0.0681 Ni-58 (n,p) 0.0993 0.0971- !
U-238-(n,f) (Cd) 0,338 0.334 Np-237 (n,f) (Cd) 2.83 2.63 REACTION RATE RATIO A PRIORI ADJUSTED -
SPECTRUM SPECTRUM
[dpa/sec]/[ Flux (E > 1.0 Mev)] 1.74E 1.65E [ Flux (E > 0.1 Mev)]/[ Flux (E > 1.0 Hev)] 3.55. 3.30
[ Flux (E < 0.414 ev)]/[ Flux-(E > 1.0 Mev)] 0.786 0.828-
'5-13
~
TABLE 5.3-5 FAST NEUTRON EXPOSURE PARAMETERS DERIVED USING THE SPECTRUM AVERACED REACTION CROSS-SECTION APPROACH NEUTRON FLUX (E > 1.0 MeV) [n/cm2-sec]
REACTION CAPSULE E CAPSULE D CAPSULE H CAPSULE F CAPSULE A Cu-63(n,a)Co-60 8.45E+10 6.88E+10 6.96E+10 6.46E+10 Fe-54(n.p)Mn-54 7.55E+10 7.llE+10 6.87E+10 5.04E+10 Ni-58(n,p)Co-58 7.56E+10 5.51E+10 U-238(n,f)Cs-137 (Cd) 8.20E+10 7.87E+10 Np-237(n,f)Cs-137 (Cd) 7.68E+10 5.74t+10 5.63E+10 AVERAGE 8.llE+10 7.12E+10 7.04E+10 6.67E+10 5.39E+10 FLUX (E > 1.0 MeV) R.llE+10 7.12E+10 7.04E+10 6.67E+10 5.39E+10 FLUX (E > 0.1 MeV) 2.68E+11 2.35E+11 2.32E+11 2.20E+11 1.78E+11
, dpa/sec 1.34E-10 1.17E-10 1.16E-10 1.10E-10 8.89E-ll FLUX (E < 0.414 ev) 6.72E+10 5.90E+10 5.83E+10 5.52E+10 4.46E+10 1
5-14
4 h a- - J A u -w # 4 L e .
I TABLE 5'.3-6 I COMPARISON OF EXPOSURE PARAMETERS' DERIVED USING THE FERRET APPROACH WITH RESULTS_USING-THE SPECTRUM AVERAGED CROSS-SECTION METHOD _
REACTION CAPSULE E CAPSULE D CAPSULE _B CAPSULE F CAPSULE A NEUTRON FLUX (E > 1.0 MeV) (n/cm2-sec)
FERRE1 7.63E+10 7.07E+10 7.14E+10 6.90E+10 5.llE+10 Sigma Avg 8.llE+10 7.12E+10 7.04E+10- 6.67E+10 5.39E+10
[ FERRET]/[ Sigma Avg] 0.941 0.993 1.014 1.034 0.948 NEUTRON FLUX (E > 0.1 HeV) (n/cm2-sec]
FERRET 2.65E+11 2.17E+11 2.54E+11 2.40E+11 1.83E+11 Signa Avg 2.68E+11 2.35E+11 2.32E411 2.20E+11 1.78E+11 (FERRET]/[ Sigma Avg] 0.989 0.923 1.095 1.091 1.028 dpa/sec FERRET 1.30E-10 1.13E-10 1.23E-10 1.17E-10 8.87E-ll Sigma Avg 1.34E-10 1.17E-10 1.16E-10 1.10E-10 8.89E-ll (FERRET]/(Sigma Avg] 0.970 0.966 1.060- 1.064 0.998' NEUTRON FLUX (E < 0.414 eV) (n/cm2-sec)
FERRET 5.58E+10 5.28E+10 5.54C+10 4.58E+10 3.76E+10 Sigma Avg 6.72E+10 5.90E+10 5.83E+10 -5.52E+10 4.46E+10
[ FERRET]/[ Sigma Avg] 0,830 0.895 0.950 0.830 0.843 5-15
SECTION 6.0 i
. EVALUA110NS Of REACTOR CAVITY DOSlHETRY In this section, the results of the evaluations of all neutron sensor sets l
irradiated since the inception of the Reactor Cavity Heasurement Program- ,
are presented. At Connecticut Yankee the program was initiated at-the ;
Leginning of fuel Cycle 16; and, to date, has included measurement evaluations at the conclusion of that first irradiation cycle. 1he ;
evaluation of the dosimetry was accomplished using the methodology ,
discussed in Section 3.0, resulting in an accurate data base defining the exposure of the reactor vessel wall.
6.1 - Cycle 16 Results 6.1.1 - Measured Reaction rates During the Cycle 16 irradiation, six multiple foil sensor sets and seven stainless steel gradient chains were deployed in the reactor cavity as -
depicted in figures 2.1-1 and ?.1-2. The capsule identifications '
associated witn each of the multiple foil sensor sets were ar.-follae s (1):
CAPSUtE IDENTIFICATION _
AZlMUTH -CORE CORE CORE ,
(dearees) TOP MIDPLANE BOTTOM 0 A B C 7.5 D_ l 37.5 E L 45- f- 3 E
I L The contents of each of these irradiation capsules is specified in l Reference 1 and,- for completeness,- is also included in Appendix _ B to this report.
6-1 -
l l
<+-e ~ -----,~~w.,-,._,--,...--_,.,,..--.---w-.--.--. *.e, -....,-~.--*-c. -,,-.,-.-,r.i-un .,-- . ,- ,-+ # .. - -,u , ~---v,,-e-.,c.,m-,ey
O The irradiation history of the Connecticut Yankee reactor during Cycle 16 is also listed in Appendir B. The irradiation history was obtained from NUREG-0020, " Licensed Operating Reactors Status Summary Report" for the appilcable operating period. Based on this reactor operating history, the individual sensor characteristics, and the measured specific activities given in Appendix B, cycle average reaction rates referenced to a coro power level of 1825 MWt were computed for each multiple foil sensor and ,
nradient chain segment.
The computed reaction rates for the multiple foil sensor sets, including ;
radiometric foils and solid state track recorders, irradiated during Cycle 16 are provided in Table 6.1-1. Corresponding reaction rate data from the the seven stainless steel gradient chains are recorded in Tablas 6.1-2 through 6.1-4 for the Fe-54 (n.p), Ni-58 (n,p), and Co-59 (n,y) reactions, respectively.
In regard to the data listed in Table 6.1-1, the Fe-54 (n.p) reaction rates represent an average of the bare and cadmium covered measurements for each capsule. Likewise, the U-238 (n,f) reaction rates were obtained by averaging the results of the radiometric foil and solid state track recorder data. In addition, the fission rate measurements . include corrections for U-235 impurities and the effects of 1,f reactions in=
the U-238 sensors as well for the effects of 1,f reactions in the Np-237 monitors.
6.1.2 - Results of the least Squares Adjustment Procedure The results of the application of the least squares adjustment procedure to the six sets of multiple foil measurements obtained from the Cycle 16 irradiation are provided in Tables 6.1-5 through 6.1-10. In these tables, the derived exposure experienced at each sensor set location along with data illustrating the fit of both the a priori and adjusted spectra to the measurements are given. Also included in the tabulations are the la ,
uncertainties associated with each of the derived exposure rates.
6-2
> 'g In regard to the comparisons listed in Tables 6.1-5 through 6.1-10, it should be noted that the columns labeled "a priori calc
- were obtained by normalizing the neutron spectral data from Table 4.1-1 to the measured fe-54 (n.p) reaction rates from each sensor set as discussed in Section 3.0. Thus, the comparisons illustrated in Tables 6.1-5 through 6.1-10 indicate only the degree to which the relative neutron energy spectra matched the measured data before and after adjustment. These data are not meant to provide hn absolute comparison of calculation and measurement.
Absolute comparisons are discussed in Section 7.0 of this report.
Complete traverses of fast neutron exposure rates in the reactor cavity were developed by combining the results of the least squares adjustment of-the multiple foil data with the fe-54 (n,p) and Ni-58 (n.p) reaction rate measurements from the gradient chains. The gradient data were employed to establish relative axial distributions over the measurement range and these relative distributions were then normalized to the FERRET results from the midplane sensor sets to produce axial distributions of exposure rates in terms of & (E > 1.0 MeV), p (E > 0.1 MeV), and dpa/sec in the reactor cavity.
The resultant axial distributions of ( (E > 1.0 MeV),
p (E-> 0.1 MeV), and dpa/sec are given in Tables 6.1-11, 6.1-12, and 6.1-13, respectively. The distributions of ( (E > 1.0 MeV) are depicted graphically in figures 6.1-1 through 6.1-7. In these graphical presentations, the multifoil points represent the explicit results of the FERRET evaluations, while the remaining data depict _the normalized results from the gradient chains.
6-3
o
)
TABLE 6.1-1 SUHKARY Of REACTION RATES DERIVED FROM MULTIPLE T0ll SENSOR SETS .;
I 1RRADIATED DURING CYCLE 16 I
RLALIJON RATE (ros/ nucleus) _
CAPSULE CAPSULE CAPSULE CAPSULE CAPSULE CAPSULE REACTION B D E F A 'C 1 CuG3(n,u) 7.45E-19 7.30E-19 3.51E-19 3.46E-19 2;39E-19 3.25E-19 Ti46(n.p) 1.13E-17 1.llE-17 5.18E-18 4.91E-18 3.70E-18-5.03E-18 Fe54(n.p) 6.37E-17 6.27E-17 2.58E-17 2.50E-17 2.00E-17 2.71E-17 '
NiS8(n.p) _9.69E-17 9.38E-17 3.81E-17 3.64E-17 3.39E-17 4.44E-17 U238(n,f) 3.58E-16 3.40E-16 1.44E-16 1.22E-16 1.32E-16_1.42E-16 Np237(n,f) 1.02E-14 8.13E-15 3.40E-15 2.77E-15 3.llE-15 4.09E-15
- CoS9(n,1) 2.15E-13 2.08E-13 1.01E-13 2.57E-13 3.82E-14 1.07E-13 a CoS9(n,1) 9.63E-14 9.10E-14 4.18E-14 1.06E-13 3.16E-14 4.38E-14
- U235(nf) 2.92E-12 2.66E-12 1.230-12 1.08E-12 3.00E-13 1.12E-12 U235(nf) 5.47E-13 4.66E-13 1.95E-13 2.02E-13 1.48E-13 2.27E-13
- - Bare foil, all others were cadmium covered h
i a
l-6-4 F
r .,
TABLE 6.1-2 fe-54 (n.p) REACTION RATES DERIVED FROM THE STAINLESS STEEL GRADIENT CHAINS 1RRADIATED DURING CYCLE 16 FEET REACTION RATE (ros/nucleust FROM MIDPLANE O DEG 7.5 DEG 37.5 DG 45 DEG 90 DEG 180 DEG 270 DEG
+6.0 5.84E-18 6.40E-18 2.78E-18 3.44E-18 5.62E-18 6.18E-18 7.46E-18
+5.0 1.97E-17 8.68E-18 8.68E-18 2.05E-17 2.10E-17 2.33E-17
+4.0 4.45E-17 4.01E-17 1.70E-17 1.58E-17 4.50E-17 4.42E-17 4.68E-17
+3.0 5.84E-17 5.32E-17 2,18E-17 2.06E-17 5.73E-17 5.84E-17 6.23E-17
+2.0 5.95E-17 5.73E-17 2.55E-17 2.46E-17 6.06E-17 6.45E-17 6.57E-17
+1.0 6.18E-17 6.12E-17 2.60E-17 2.30E-17 6.57E-17 6.57E-17 6.73E-17 0.0 5.90E-17 6.45E-17 6.62E-17
-1.0 5.95E-17 4.92E-17 2.41E-17 2.39E-17 6.23E-17 5.95E-17 6.79E-17
-2.0 5.90E-17 5.26E-17 2.36E-17 2.34E-17 5.20E-17 7.07E-17 6.12E-17
-3.0 5.41E-17 4.68E-17 2.34E-17 1.95E-17 6.01E-17 6.62E-17 6.45E-17
-4.0 4.19E-17 4.02E-17 1.55E-17 1.50E-17 4.17E-17 4.22E-17 4.78E -5.0 2.24E-17 7.46E-18 7.12E-18 2.32E-17 2.46E-17 2.19E-17
-6.0 4.90E-18 4.llE-18 2.53E-18 2.41E-18 5.84E-18 7.01E-18 5.28E-18 6-5 4
4 TABLE 6.1-3
-Ni-58 (n p) REACTION RATES DERIVED FROM THE STAINLESS _ STEEL GRADIENT CHAINS 1RRADIAlED DURING I.YCLE 16 FEET REACTION RATE (rns/ nucleus)
FROM MIDPLANE O DEQ_ 7.5 DEG,37.5 DG .45 DEG- 90 DEG 180 DEQ_ 270 DEG
+6.0 1.04E-17 9.47E-18 5.08E-18 4.70E-18 9.42E-18 1.05E-17 1.12E-18 45.0 3.31E-17 1.31E-17 1.31E-17.3.38E-17 3.41E-17 3.77E-17 t
+4.0 6.59E-17 6.38E-17 2.40E-17 2.42E-17 6.59E-17 7.llE-17 7.85E-17 43.0 S.68E-17 8.370-17 3.33E-17 3.19E-17 8.58E-17 9;31E-17 9.63E-17
+2.0 9.15E-17 9.21E-17 3.50E-17 3.37E-17 9.26E-17 9.94E-17 9.73E-17
+1.0 9.21E-17 8.74E-17 3.75E-17 3.02E-17 9.78E-17 9.63E-17 9.99E-17
- 0.0 8.63E-17 9.89E-17 1.08E-16
-1.0 8.84E-17 8.06E-17 3.40E-17 3.18E-17 8.84E-17 8.79E-17 1.00E -2.0 9.26E-17 8.95E-17 3.47E-17 3.13E-17 9.63E-17 1.07E-16 1.03E-16 ,
-3.0 9.00E-17 7.85E-17 3.39E-17 2.86E-17 8.26E-17 9.05E-17 9.31E-17
-4.0 6.90E-17 5.44E-17 2.46E-17 2.27E-17 6.96E-17 7.38E-17 7.69E-17 ,
-5.0 3.44E-17 1.20E-17 1.14E-17 3.61E-17 4.52E-17 3.92E-17
-6.0 9.63E-18 7.22E-18 4.01E-18 3.47E-18 1.06E-17 1.06E-17 1.llE-17
+
l- 6-6 l
l:
l
.- q
- l TABLE 6.1-4 -j Co-59 (n,1) REACTION RALES DERIVED FROM THE STAINLESS STEEL GRADIENT CHAINS IRRADIATED DURING CYCLE 16 FEET REACTION RATE (ros/nucloud___
FROM l
1 MIDPLANE - 0 DEG 7.5 DEG 37.5 DG - 45 DEG- 90 DEG -180 DEG 270 DEG ,
+6.0 1.94E-14 2.05E-14 1.08E-14 1.07E-14 1.91E-14 2.03E-14 2.28E-14 l
+5.0 3.76E-14 1.82E-14 1.75E-14 3.63E-14 3.78E-14 4.20E-'14 j
+4.0 5.68E-14 5.97E-14 2.72E-14-2.58E-14 5.78E-14 5.92E-14 6.53E-14 .
0 1 E 9- 14 9 9 4 .0 E
+1.0 1.18E-13 1.18E-13 5.40E-14 5.07E-14 1.17E-13 1.24E-13 1.35E-13 0.0 1.88E-13 1.97E-13 2.12E-13
-1.0 2.llE-13 2.07E-13 9.26E-14 8.83E-14 2.18E-13 2.28E-13 2.39E-13 -
-2.0 2.06E-13 2.00E-13 8.83E-14 8.22E-14 2.llE-13 2.23E-13 2.32E-13
-3.0 1.83E-13 1.79E-13 7.70E-14 7.23E-14 1.90E-13 2.02E-13-2.09E-13
-4.0 1.41E-13 1.36E-13 S.92E-14 5.59E-14 1.48E-13 1.57E-13 1.60E-13
-5zo 8.60E-14 3.87E-14 3.60E-14 9.44E-14 1.01E-13 1.02E-13 -
-6.0 4.47E-14 4.31E-14 2.14E-14 2.01E-14 4.64E-14 5.26E-14'5.45E ,
+
G-7 v & < , , - ---,+-- , ,.+-. , m-,'v-,-,.,,w, w mwew,aw,,,- ,,,w,-- -n ,-,we.,- .,~g n- , - . , , 4 n- ,m_,. n,.- ge- .,
I I TABLE 6.1 DERIVED EXPOSURE RATEk FROM THE CAPSULE 8 0051MElRY EVALUATION 0 DEGREE AZIMUTH - CORE-MIDPLANE A PRIORI ADJUSTED 10 PARAMETER VALUE VALUE UNCERTAINTY flux (E > 1.0 Mev) 1.64E+09 1.72E+09 8%
Flux (E > 0.1 Hev) 2.38E+10 2.50E+10 16%
Flux (E < 0.414 ev) 1. Su+ 09 4.80E409 18%
flux (Total) 5.51E+10 6.82E+10 15%
dpa/sec 8.22E-12 8.72E-12 15%
COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 0 DEGREE AZIMUTH - CORE MIDPLANE REACTION RATE (rnJ/ nucleus) C/M A PRIORI ADJUSTED REACTION MEASURED CALC. CALC. A PRIORI ADJUSTED Cu-63 (n.o) 7.45E-19 l.03E-18 7.86E-19 1.38 1.05 Ti-46 (n,p) 1.13E-17 1.18E-17 1.llE-17 1.05 0.98 Fe-54 (n.p) 6.37E-17 6.32E-17 6.27E-17 0.99 0.98 Ni-58 (n p) 9,69E-17 8.83E-17 9.30E-17 0.91 0.96-U-238 (n,f) (Cd) 3.58E-16 4.13E-16 4.18E-16 1.15 1.17.
Np-237 (n,f) (Cd) 1.02E-14 9.71E-15 1.03E-14 0.95 1.01 Co-59 (n,7) 2.15E-13 1.12E-13 2.27E-13 0.52 1.06 Co-59 (n,y) (Cd) 9.63E-14 7.23E-14 9.46E-14 0.75 0.98 U-235 (n,f) 2.92E-12 8.43E-13 2.49E-12 0.29 0.85 U-235 (n,f) (Cd) 5.47E-13 2.94E-13 5.26E-13 0.54 0.96 6-8
l TABLE 6.1-6 !
DERIVED EXPOSURE RATES FROM 1HE CAPSULE D 00SIMETRY EVALVATION 7.5 DEGREE AZIMUTH - LORE MIDPLANE A PRIORI ADJUSTED la ;
PARAMETER VALUE VALUE UNCERTAINTY Flux (E > 1.0 Hev) 1.63E409 1.55E409 8%
Flux (E > 0.1 Mev) 2.39E+10 2.06E410 17% I Flux (E < 0.414 ev) 1.59E+09 4.62E+09 18%
Flux (Total) 4.88E+10 5.20E+10 13%
dpa/sec 7.39E-12 6.53E-12 14%
COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 7.5 DEGREE AZIMUTH - CORE HIDPLANE REACTION RATE fros/ nucleus) C/M $
A PRIORI ADJUSTED REACTION MEASURED CALC. CALC. &_ERlqRL ADJUSTED Cu-63 (n,a) 7.30E-19 1.02E-18 7.72E-19 1.40 1.06 Ti-46 (n.p) 1.llE-17 1.170-17 -1.09E-17 1.05 0.98 Fe-54 (n,p) 6.27E-17 6.23E-17 .6.13E-17 0.99 . 0.98 Ni-58 (n.p) 9.38E-17 8.70E-17 9.00E 0.93 0.96 U-238 (n,f) (Cd) 3.40E-16 4.07E-16 3.90E-16 1.20 1.15 Np-237 (n,f) (Cd) 8.13E-15 9.64E-15 8.48E-15 1.19 1.04 Co-59 (n,1) 2.08E-13 1.16E-13 2.18E-13 0.56 1.05 Co-59 (n,7) (Cd) 9.10E-14 7.50E-14 8.97E-14 0.82 0.99 U-235 (n,f) 2.66E-12 8.56E-13 2.33E-12 0.32 - 0.88 U-235-(n,f) (Cd) 4.66E-13 2.90E-13 4.49E-13 0.62 10.96 6-9
.t t w wt t f*w -v e .w,. y er y. ,gr- -mfy r- Hew- -wy- gw y ,w- -
y W -
.g .. -..ye -
, ,\
TABLE 6.1-7 DERIVED EXPOSURE RATES FROM THE CAPSULE E 00SIMETRY EVALUATION 37.5 DEGREE AZlHUTf! . CORE HIDPLANE A PRIORI ADJUSTED 10 PARAMETEIL _YALUL VALUE 1/NCERTAINTY Flux (E > 1.0 Mev) 6.00C+08 6.15E408 8% <
flux (E > 0.1 Mov) 9.56E409 8.79E+09 17%
Flux (E < 0.414 ev) 7.20E408 2.25E409 17% -;
flux (Total) 2.10E410 2.28E410 13%
dpa/sec 2.93E-12 2.76E-12 14% ;
i COMPARISON Of MEASURED AND CALCULATED SENSOR REACTION RATES 37.5 DEGREE AZlMU1H - CORE HIDPLANE EEACTION RATE fros/DycleusL C/M A PRIORI ADJUSTED REACTION- HEASURED . CALC.__ CALC. A PRIORI 6DJUSTED Cu-63 (n.of 3.51E-19 4.81E-19 3.71E-19 1.37 1.06-Ti-46 (n.p) 5.18E-18 5.21E-18 5.05E-18 1.01 0.97 fe-54 (n.p) 2.58E-17 2.53E-17 2.55E-17 0.98 0.99 Ni-58 (n,p) 3.81E-17 3.49E-17 3.68E-17 0.92 0.97-U-238'(n.f) (Cd) 1.44E-16 1.53E-16 1.57E-16 1.07 1.09 Np-237 (n,f) (Cd) 3.40E-15 3.73E-15 3.52E-15 1.10 1.04 Co-59 (n,y) 1.01E-13 5.60E-14 1.04E-13 0.55 1.03 Co-59-(n.1) (Cd) 4'18E-14 3.74E-14 4.14E . 0.89 0.99 U-235 (n f)' l.23E-12 3.94E-13 1.llE-12 :0.32 0.91-U-235 (n,f) (Cd)- 1.95E-13 1.410-13 1.90E-13 - 0.72 0.97 L
6-10 L
l l .:
TABLE 6.1-0 j
-DERIVED EXPOSURE RATES FROM 1HE CAPSULE f DOSIMETRY EVALUATION 45 DEGREE AZIMUTH - CORE MIDPLANE A PRIORI ADJUSTED la
_. PARAMETER VALUE _.yf_Ly(_ UNCERTAlflIX ,
Flux (E > 1.0 Mev) 5.70E+08 5.29[408 8% .
Flux (E > 0.1 Mev) 9.08E+09 7.26E409 17% ,
Flux (E < 0.414 ev) 6.83E408 2.19E+09 18% l Flux (Total) 2.00E+10 2.05E+10 13%
dpa/sec 2.79E-12 2.31E-12 14%
COMPARISON OF MEASURED AND CALCULATED SENSOR REACTION RATES 45 DEGREE AZIMUTH - CORE MIDPLANE REACTION RATE fros/ nucleus) C/M A PRIORI ADJUSTED
'RfACTION Bff_1E[D CALC. CALC. $_.ERl0H1 ADJUSTED Cu-63 (n.o)- 3.46E-19 4.74E-19 3.66E-19 1.37 1.06-Ti-46 (n,p) 4.91E-18 5.09E-18 -4.83E -1.04 0.98 Fe-54 (n.p) 2.50E-17 2.44E-17 2.43E-17 0.98 0.97 Ni-58 (n.p) 3.64E-17 3.36E-17. 3.49E-17 0.92 0.96 U-238 (n,f) (Cd) 1.22E-16 1.46E 1.39E-16 1.20 -1.14 Np-237 (n,f) (Cd) 2.77E-15 3.54E-15 2.92E-15 1.28- 1.06 00-59 (n,y) 1.03E-13 5.36E-14 1.03E-13 0.52 1.00 -
Co-59 (n,7) (Cd) 4.24E-14 3.59E-14 4.26E-14 0.85 1.00~
U-235.(n,f) -1.08E-12 3.75E-13 1.07E-12 0.35 0.99 .
U-235 (n,f) (Cd) 2.02E-13 1.35E-13 1.92E-13 0.67 0.95 6-11
(
TABLE 6.1-9 DERlVED EXPOSURE RATES FROM THE CAPSULE A 00SIMElRY EVALUATION O DEGREE AZIMUTH - CORE TOP A PRIORI ADJUSTED la '
PARAMETER VALUE . VALUE UNCERTAINTY -
flux (E > 1.0 Mev) 5.16E408 5.79E408 8% _
flux (E > 0.1 Mov) 7.45E409 7.70E409 16%
flux (E < 0.414 ev) 4.84E408 3.82E+08 29%
Flux (Total) 1.73E+10 1.92E410 15% ;
dpa/sec 2.58E-12 2.70E-12 14% i t
COMPARISON Of MEA $URED AND CALCULATED SENSOR REACTION RA1ES 0 DEGREE AZIMU1H - CORE TOP e
_Rf.AL.110N RATE fros/ nucleus) C/M A PRIORI ADJUSTED.
REACTION MEASURED CALC. CALC. .a_ PRIORI ADJUSTED Cu-63 (n.o) 2.39E-19 3.23E-19 2.51E-19 1.35 1.05-Ti-46-(n.p) 3.70E-18 3.72E-18 3.62E-18 1.00 0.98 fe-54 (n.p) 2.00E-17 1.98E-17 :2.04E-17 0.99 1.02 Ni-58 (n.p) 3.39E-17 2.77E-17 3.19E 0.82 0.94 U-238 (n.f) (Cd) 1.32E-16 1.30E-16 1.42E-16_ 0.98 1.08 .
Np-237 (n,f) (Cd). 3.llE-15 3.05E-15 3.22E-15 '0.98 1.04 Co-59 (n,7) 3.82E-14 3.52E-14 ,4.04E-14 0.92 1.06 C0-59 (n,1) (Cd) 3.16E-14 2.27E-14 3.04E-14 0.72 - 0.96 U-235 (n,f) 3.00E-13 2.650-13 2.76E-13 0.88 0.92 ,
U-235 (n,f) (Cd) 1.48E-13 9.21E-14 1.45E-13 0.62 0.98L '
6-12
TABLE 6.1-10 DERIVED EXPOSURE RATES FROM THE CAPSULE C DOSIMETRY EVALUATION 0 DEGREE AZIMUTH - CORE BOTTOM A PRIORI ADJUSTED 10 -
PARAMETER _VALUE -VALUE UNCERTAINTY Flux (E > 1.0 Mev) 6.98E+08 7.08E+08 8%
Flux (E > 0.1 Mev) 1.01E+10 9.96E409 16%
Flux (E < 0.414 ev) 6.55E+08 2.27E+09 18%
Flux (Total) 2.34E+10 2.80E+10 14%
dpa/sec 3.49E-12 3.51E-12 15%
C L
COMPARIS0N Of MEASURED AND CALCULATED SENSOR REACTION RATES 0 DEGREE AZIMUTH - CORE BOTTOM REACTION RATE Iros/ nucleus) ,
C/M A PRIORI ADJUSTED REACTION MEASURED CALC. CALC. A PRIOR 1 ADJUSTED Cu-63 (n a) 3.25E-19 4.37E-19 3.44E-19 1.35 1.06 Ti46 (n.p) 5.03E-18 5.03E-18 4.92E-18 1.00- 0.98-Fe-54 (n.p) 2.71E-17 2.68E-17 2.70E-17 0.99 1.00 Ni-58 (n.p) 4.44E-17 3.75E-17 4.16E-17 0.84 0.94 U-238 (n,f) (Cd) 1.42E-16 1.75E-16 1.74E-16 1.24 1.23 Np-237 (n,f) (Cd) 4.09E-15 4.13E-15 4.13E-15 1.01 1.01 Co-59 (n,7) 1.07E-13 4.77E-14 1.06E-13 0.45 0.99 Co-59 (n,7) (Cd) 4.38E-14 3.07E-14 4.41E-14 0.70 1.01 U-235 (n,f) 1.12E-12 .3.58E-13 1.12E-12 0.32 1. 00 --
U-235 (n,f) (Cd) 2.27E-13 1.25E-13 2.13E-13 0.55 0.94 i 6-13
- - - _ ~ . , , ,-
TABLE 6.1-11 TAST NEUTRON fLVX (E > 1.0 MeV) AS A FUNCTION Of AX1AL POSITION WITHIN THE REACTOR CAVITY CYCLE 16 IRRADIATION AZIMUTHAL ANGLE HEIGHT (ft) _Q.0 DEG 7.5 DEG 37.5 DEG 45.0 DEG 90 DID 180 DEG 270 DEG
+6.0 1.98E+08 1.79E408 7.09E407 5.27E407 1.98E+08 2.11E+08 2.18E+08 45.0 5.79E408 5.81E+08 2.30E403 1.98E+08 6.44E+08 6.87E+08 7.08E 08
+4.0 1.24E+09 1.12E+09 4.44E408 3.72E+08 1.24E+09 1.32E409 1.36E409
+3.0 1.62E+09 1.46E+09 5.81E+08 4.93E+08 1.62E409 1.73E409 1.78E+09 >
+2.0 1.75E409 1.58E+09 6.26E408 5.28E408 1.75E+09 1.86E409 1.92E+09
+1.0 1.77E+09 1.59E+09 6.32E+08 5.14E+08 1.77E+09 1.88E+09 1.94E409 0.0 1.72E+09 1.55E+09 6.15E+08 5.29E+08 1.72E+09 1.83E+09 1.89E409
-1.0 1.67E+09 1.51E409 5.97E408 5.44E+08 1.67E+09 1.78E409 1.84E+09
-2.0 1.72E+09 1.55E409 6.14E+08 5.38E+08 1.71E+09 1.83E+09 1.89E409
-3.0 1.60L+09 1.44E+09 5.73E+08 4.99E+08 1.60E409 1.71E+09 1.76E+09
-4.0 1.21E409 1.09E409 4.32E+08 3.82E+08 1.21E+09 1.29E+09 1.33E409
-5.0 7.08E+08 5.80E408 2.30E+081.98E+08 6.43E408 6.86E408 7.07E+08
-6.0 1.7]E+08 1.54E+08 6.12E+07 6.10E407 1.71E408 1.83E+08 1.88E408 t
6-14
TABLE 6.1-12 FAST NEV1RON FLUX (E > 0.1 MeV) AS A FUNCTION OF AXIAL POSITION WITHIN-THE REACTOR CAVITY CYCLE 16 IRRADIATION AZIMUTHAL ANGLE HEIGHT (ft) 0.0 DEG 7.5 DEG 37.5 DEG 45.0 DEG 90 DEG 180 DEG .270 DEG
+6.0 2.88E409 2.37E+09 1.01E+09 8.37E+08 2.88E+09 3.07E+09 3.17E+09 45.0 7.70E+09 7.72E+09 3.29E+09 2.72E+09 9.36E+09 9.99E+09 1.03E+10
+4.0 1.80E+10 1.49E+10 6.34E+09 5.24E+09 1.80E+10 1.92E+10 1.98E+10 -
+3.0 2.36E+10 1.94E+10 8.30E+09 6.85E+09 2.30E+10 2.52E+10 2.59E+10
+2.0 2.54E+10 2.10E+10 8.94E+09 7.38E+09 2.54E+10 2.71E+10 2.79E+10
+1.0 2.57E+10 2.12E+10 9.04E+09 7.46E409 2.57E+10 2.74E+10 2.82E+10 4 0.0 2.50E+10 2.06E+10 8.79E+09 7.26E+09 2.50E+10 2~67E+10 2.75E+10
-1.0 2.43E+10 2.00E+10 8.54E+09 7.05E+09 2.43E+10 2.59E+10 2.67E+10
-2.0 2.49E+10 2.06E+10 8.77E+09 7.25E+09 2.49E+10 2.66E+10 2.74E+10
-3,0 2.33E+10 1.92E+10 8.19E+09 6.76E+09'2.33E+10 2,48E+10 2.56E+10
-4.0 1.76E+10 1.45E+10 6.18E409 5.10E+09 1.76E+10 1.87E+10 1.93E+10
-5.0 9.96E+09 7.70E+09 3.29E+09 2.71E+09 9.34E+09 9.97E+09 1.03E+10
-6.0 2.49E+09 2.05E+09 8.75E+08 7.23E+08 2.49E+09 2.65E+09-2.74E+09 t
r 5
6-15 T
,a t TABLE 6.1-13 IRON ATOM DISPLACEMENT RATE (dpa/sec) AS A FUNCTION Of AX1AL POSITION WITHIN THE REAC10R CAVITY CYCLE 16 IRRADIATION AZIMUTHAL ANGLE HEIGHT
, -(ft) 0.0 DEG 7.5 DEG 37.5 DEG 45.0 DEG 90 DEG 180 DEG _2]D_QLG 46.0 1.01E-12 7.53E-13 3.18E-13 2.66E-13 1.00E-12 1.07C-12 1.10E-12 45.0 2.70E-12 2.450-12 1.03E-12 8.66E-13 3.26E-12 3.48E-12 3.59E-12
+4.0 6.29E-12 4.71E-12 1.99E-12 1.67E-12 6.29E-12 6.71E-12 6.92E-12'
+3.0 8.230-12 6.16E-12 2.61E-12 2.18E-12 8.22E-12 8.78E-12 9.050-12
+2.0 8.87E-12 6.64E-12 2.81E-12 2.35E-12 8.86E-12 9.45E-12 9.75E-12
+1.0 8.96E-12 6.71E-12 2.84E-12 2.37E-12 8.95E-12 9.55E-12 9.85E-12 0.0 8.72E-12 6.53E-12 2.76E-12 2.31E-12 8.71E-12 9.30E-12 9.58E-12
-1.0 8.47E-12 6.34E-12 2.68E-12 2.24E-12 8.46E-12 9.03E-12 9.31E-12
-2.0 8.70E-12 6.52E-12 2.75E-12 2.31E-12 8.69E-12 9.28E-12 9.56E-12
-3.0 8.12E-12 6.08E-12 2.57E-12 2.15E-12 8.12E-12 8.66E-12 8.93E -4.0 6.13E-12 4.59E-12 1.94E-12 1.62E-12 6.12E-12 6.53E-12 6.74E-12
-5.0 3.5]E-12 2.44E-12 1.03E-12 8.64E-13 3.26E-12 3.48E-12 3.58E-12
-6.0 8.68E-13 6.50E-13 2.75E-13 2.30E-13 8.67E-13 9.25E-13 9.54E-13 6 ;
- i i
FIGURE 6.1-1 i
i FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION- l ALONG THE O DEGREE TRAVERSE IN THE REACTOR CAVITY ,
CYCLE 16 IRRADIATION ;
1 i
i 1.0 E + 10 tr 8 C rc. We Id - '
+ Cir:. Weld c0 <
E h
x 1.0E+09
.2
. 7
(( i
\
~-
LL "r 2. _. ..
C / M -- - '
O / .\' '
y /j
-4 y
- d f
v
~ ,
1.0E+08 G 5 -4 3 2 -1 0 1 2 3 45 G -7 Distance From Core Midplane (ft)
+ GRADIENT H MULTIFOIL-6-17 l;
a u FIGURE 6.1-2 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AX1AL POSITION ALONG THE 7.5 DEGREE fRAVERSE IN THE REACTOR CAVITY CYCLE 16 IRRADIATION
'I 1.0E+10 >
TT . -
r 8 C rc. Woldt -
Cir:. Welt s .
^
I E
u , :. ...
g x 1.0E+09
'~ y m -
P 3
G~ / .\ :
y ,e c
O / \
g ;j< ~, . ,
y j /: .
y' < g\
1 y ,' m J
' ' c f
- N:
1.0E+08 6 5 -4 -
321 0- 1- 2 3 -4 5 6 7 Distance From Core Midplane (ft)
+ GRADIENT H MULTIFOIL -
W Y
6-18 V
w, cm,,,- we,, >>m,w,- -,--,------o,,,or v.-.- N , n -.m v . r ar ,y,w-- -, m--w, w& -,nu- ,, e, --y v.e- +m ~
4 FIGURE 6.1-3 FAST-NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION Of AXIAL POSITION ALONG THE 37.5 DEGREE TRAVERSE IN THE REACTOR CAVITY CYCLE 16 IRRADIATION 1.0E+09
.fr 2 O 'If ll-
$ -/ \
ja f , .-
h; C
g 1.0E+08 ,
E . I ,
8 vir::. w etr g .C Ic. We id . m.
^
1.0E+07
-7654 21 0 1 2 3 4 5 6 7 Distance From Coro Midplane (ft)-
+ GRADIENT a MULTIFOIL 6-19
- ~ a , . - - - ,- , - - - - -- - ,r ~. , , , ~ ,.-n . - ~ .
. _ - . _ - ._ . _ . _ _ _ . ~ . _ _ . _ _ . _ _ . . _ . . _ . _ - .
4
,- i FIGURE 6.1-4 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 45 DEGREE TRAVERSE IN THE REACTOR CAVITY !
CYCLE 16 IRRADIATION ,
F 1.0E+09 -,
g e ,
m ,
g , -f 4 i j '
/ "
h ;
-R -
c w ~
a ,
x 1.0E+08 , .
E / 1.
h 4
Ulr3. = %V5113
)
z G irc. We id4 r
n .'
1.0E+07 54 321 0 1. 3 4 5 6 7-Distance From Core Midplane (ft) .
+ GRADIENT a MULTIFOIL i
r 6-20 ;
FIGURE 6.1-5 ;
i FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AX1AL POSITION ALONG THE 90 DEGREE TRAVERSE IN THE REACTOR CAVITY ,
r,YCLE 16 IRRADIATION
[
?
1.0E+10
- l fj C rc; We'ld '- Cir::. Welt s
E =
.sz y ,
5 :-
x 1.0E+09
')n ..
( m 2 l \
- u. n n
/ \
e O I N y ,
ll - ..
. \. .
g fp ,
t 4 -
.g
+- ,,
1.0E+08 6 -5 3 -21 0 1 :2 3 4 5 6 .7 Distance From Core Midplane (ft)
+ GRADIENT 6-21
FIGURE 6,1-6 FAST NEUTRON FLUX (E > 1.0 MeV).AS A FUNCTION OF AXIAL POSITION ALONG THE 180 DEGREE TRAVERSE IN THE REACTOR CAVITY CYCLE 16 IRRADIATION 1.0E+10 g ,m 8 C rc. Werid ' Cir::: Weld x 1.0E+09 ,
[ _
/:-
2 u.
.l v .'n
/ Y e l "'
O /
g /,' o 1\
g
+
/ ,
.1 w , . -
4 1.0E+08 -
7 -654 -3 21 0 1 23- 4 5- 6 7-Distance From Core Midplane (ft)
+ GRADIENT l 6-22
i.
L FIGURE 6.1-7 FAST NEUTRON FLUX (E > 1.0 MeV) AS A FUNCTION OF AXIAL POSITION ALONG THE 270 DEGREE TRAVERSE IN Tile REACTOR CAVITY CYCLE 16 1RRA0!AT10N 1.0E+10 77
$ C rc. We>ld - > ' '
Circ.-Weld-x 1.0E+09
)..
~,
Ny
.2 . l ',.
LL T n C # \ '
O / \- f f f' -\
- g ,.
p ; n .
- bc .
1.0E+08
-7 G -5 4 -3 2 -1 0 1 2 3 -4 5 6- Distance From Core Midplane (ft) r
+ GRADIENT 6
6-23 '
SECTION 7.0 i COMPARISON Of CALCULATIONS WITH MEASUREMENTS 5 3
in order to develop accurate neutron exposure profiles at the inner diameter -i and through the thickness of the pressure vessel wall, the measurement results '
provided in Sections 5.0 and 6.0 must be combined with analytically determined spatial gradients provided in Section 4.0. In essence, this approach accepts -
the measurement results as the best available exposure rate information for the irradiation period in question and assumes that the analytically ;
determined radial distribution functions provide accurate representations of the spatial gradients that exist among the measurement locations and the points of interest within-the pressure vessel wall. This approach is analagous to the common practice of normalizing a cycle specific forward neutron transport calculation to available measurements from either surveillance capsule or reactor cavity dosimetry programs.
An indication of the acceptability of this method of exposure determination can be gained by an absolute comparison of the results of neutron transport calculations with all measured results applicable to a given reactor. These comparisons quantify the biases that may exist due to the transport methodology, reactor _modeling, and/or reactor operating characteristics over the respective irradiation periods; and, furthermore, demonstrate the degree-of consistency among the measurements obtained from different geometric locations and varying irradiation intervals.
In this section, comparisons of the measurement results from surveillance capsule and reactor cavity dosimetry with corresponding analytical predictions at the measurement locations are presented. These comparisons are provided on two levels, in the first instance, predictions of fast neutron exposure rates in terms of 4 (E > 1.0 MeV), d (E > 0.1 MeV), and dpa/sec are compared with the results of the FERRET least squares adjustment procedure; while, in the second case, calculations of individual sensor' reaction rates are compared directly with the measured data from the counting laboratories. It is shown 7-1 v- r -- gr r hegr Fg-y- v y ,p wyty y e- e ! e w>-Pa-pm g- e-
that these two levels of comparison yield consistent and similar results, indicating that the least squares adjustment methodology is producing accurate exposure results and that the calculation / measurement comparisons are yielding accurate bias factors that can be applied to neutron transport calculations performed for the Connecticut Yankee reactor. !
1 7.1 Comparison of Least Squares Adjustment Results with Cal;olation In Table 7.1-1, comparisons of calculated and measured exposure rates for the five surveillance capsule dosimetry sets and for the four reactor midplane dosimetry sets irradiated during Cycle 16 are given. In all j cases, the calculated values were based on the fuel cycle specific exposure calculations averaged over the appropriate irradiation period.
That is, the Capsule A values apply to Cycle 1, the Capsules F. H, 0,- and E values represent averages over Cycles 1 through 4, I through 7, 1l through 10, and I through 15, respectively; and, the cavity measurements r directly apply to Cycle 16.
An examination of Table 7.1-1 indicates that, considering all of the available core midplane data, the calculated exposure rates compare to the measurement results by ratios of 0.958, 1.022, and 1.023 for p (E > 1.0 MeV), d (E > 0.1 HeV), and dpa/sec, respectively. The standard deviations associated with each of the 9 sample data sets were 6.8%, 8.9%, and 10.1%, respectively.
7.2 Comparisons of Measured and Calculated Sensor Reaction Rates ,
in Table 7.2-1, calculation / measurement ratios for each fast neutron sensor reaction rate'from the surveillance capsule and reactor cavity irradiations are listed. This tabulation provides a direct comparison, on an absolute basis, of calculation and measurement prior to the wpplication of the least squares adjustment procedure as represented in the FERRET .
evaluations.
7-2 i
_. . . a _._ a _._.,___.__.u.._ - - . _
An examination of Table 7.2-1 shows consistent vehavior for all reactions and all measurement points. .The standard deviations observed for the six fast neutron reactions range from 3% to 13% on an individual reaction basis; whereas, the overall average C/M ratio for the entire data set has an associated 10 standard deviation of approximately 11%.
Furthermore, the average C/M bias of 1.004 observed in the reaction rate comparisons is in excellent agreement with the values of 0.958, 1.022, and 1.023 observed in the exposure rate comparisons shown in Table 7.1-1.
j 7-3
- . _ . _ . . . _ . _ . . _ _ _ . _ . . _ . . . . _ . _ . _ . _ _ _ . _ _ _ . _ , . _ - . _ . - . _ . _ . _ _ . _ _ . . _ _ _ - , _ . _ _ . _ . . ~
4~ t TABLE 7.1-1 COMPARISON Of HEASURED AND CALCULATED EXPOSURE RATES FROM SURVE!LLANCE CAPSULE AND CAV11Y DOSlHE1RY 1RRADIAT10NS
{ASTNEUTRONFLUX(E>1.0_NeV) p (n/, n2-sec)
_ MEASURED-- .IALCUL ATED - C/M 43.5 DEGREE CAPSULE (A) 5.llE410 5.69E410 1.114 43.5 DEGREE CAPSULE (f) 6.90E410 6.19E+10 0.897 .:
43.5 DEGREE CAPSULE (H) 7.14E+10 6.72E4!0 0.941 43.5 DEGREE CAPSULE (D) 7.07E410 6.73E+10 0.952 43.5 DEGREE CAPSULE (E) 7.63E+10 6.82E410 0.894 0.0 DEGREE CAVITY (CY16) 1.72E+09 1.59E409 0.924 7.5 DEGREE CAVITY (CYl6) 1.55E409 1.45E+09 0.935 37.5 DEGREE CAVITY (CY16) 6.15E408 5.81E408 0.945 45.0 DEGREE CAVITY (CYl6) 5.29E408 5.41E408 i.023 AVERAGE C/M RATIO 0.958-la VARIATION 0.065 l
l 7-4 1
t
. TABLE 7.1-1 (Continued)
COMPARIS0N Of HEASURED AND CALCULATED EXPOSURE RATES TROM SURVEILLANCE CAPSULE AND CAVITY DOSIMETRY IRRADIATIONS FAST NEUTRON FLUX (E > 0.1 HeV) p [n/cm2-sec) i MEASURED _ _-CALCULATE (L C/M .
43.5 DEGREE CAPSVLE (A) 1.83E+11 2.02E+11 1.104 l 43.5 DEGREE CAPSVLE (f) 2.40E+11 2.20E+11 0.917 ;
43.5 DEGREE CAPSULE (H) 2.54E+11 2.39E+11 0.941 43.5 DEGREE CAPSULE (D) 2.17E411 2.39E+11 1.101 -
43.5 DEGREE CAPSULE (E) 2.65E+11 2.42E+11 0.913 0.0 DEGREE CAVITY (CYl6) 2.50E410 2.36E+10 0.944 7.5 DEGREE CAVITY (CY16) 2.06E+10 2.17E+10 1.053 37.5 DEGREE CAVITY (CY16) 8.79E409 9.19E409 1.046 45.0 DEGREE CAVITY (CY16) 7.26E409 8.57E409 1.180-AVERAGE C/M RATIO 1.022 la VARIATION 0.091 l
7-5
, ., ~ , , - . . . - - - , ' . . , . . , '
-,,,,..,-.n-,
, . , , , , . . . , ,,,,,,-ae,. _,,,, ,.,,._,,e ,n,,,. ,----.m,n,,, - -.w,
. ,.. - w-..-a, y -
Y El j/ TABLE 7el-1 (Continued).
-3 COMPARISON OF HEASURED AND CALCULATED EXPOSURE RATES FROM SURVEILLANCE CAPSULE AND CAVITY DOSlHETRY IRRADIATIONS 1RON ATOM DISPLACEMENT RATE -'
[dpa/sec)
MEASURED CALCULATE.Q_ C/M 43.5 DEGREE CAPSULE (A) 8.87E-Il 9.90E-10 1.116 43.5 DEGREE CAPSULE (F) 1.17E-10 1.08E-10 -0.923 43.5 DEGREE CAPSULE (H) 1.23E-10 1.17E-10. 0.951 J 43.5 DEGREE CAPSULE (D) 1.13E-10 1.17E-10 1.035-43.5 DEGREE CAPSULE (E) 1.30E-10 1.19E-10 0.915 0.0 DEGREE CAVITY (CY16) 8.72E 7.72E-12 -0.885 7.5 DEGREE CAVITY (CY16) 6.53E-12 7.llE-12 1.089 37.5 DEGREE CAVITY (CY16) 2.76E-12 3.00E-12' l.087-45.0 DEGREE CAVITY (CY16) 2.31E-12 2.79E-12 1.208' AVERAGE C/M RATIO 1.023 la VARIATION 0.103 1
7-6
~
TABLE-7.2-1 COMPARISON OF MEASURED AND CALCULATED NEUTRON SENSOR REACTION RATES FROM' SURVEILLANCE CAPSULE AND CAVITY DOSIMETRY IRRADIATIONS c
Cu63(n.a) Ti46(n.0) Fe54(n.0) NiS8(n._ gl1Ein fl No237(n;f)
CAPSULES A 1.240 1.056 1.081 F 1.083 0.990 H 1.091 1.037 0 1.106 0.980 0.910 0.864 1,254' E 0.912 0.844 0.952 CYl6 CAVITY 0.0 DEGREE 1.016 1.014 0.959 0.882 1.121 0.923 7.5 DEGREE 0.944 0.937 0.884 0.825 1.067 1.054 37.5 DEGREE 1.019 0.973 0.951 0.887 1.032 1.062-45.0 DEGREE 1.001 0.984 0.927 0.875 1.187 1.213 AVERAGE C/M 1.022 0.977 0.996 0.906 .l.019 1.077- :
lo 0.065 0.028 0.101 0.072 0.126 0.113 TOTAL C/M RATIO 1.004 la VARIATION 0.106 7-7
SECTION 8.0 BEST ESTIMATE NEUTRON EXPOSURE Of PRESSURE VESSEL MATERIALS In this section the measurement results provided in-Sections 5.0 and-6.0-are combined with the results of the neutron transport calculations-
-described in Section 4.0 to establish a mapping of the best estimate neutron exposure of the beltline region of the Connecticut Yankee reactor-pressure vessel through the completion of Cycle 15.- Based on the continued use of the Cycle 1-15 average, non-low leakage fuel loading pattern, projections of future vessel exposure to 32 effective full power years of operation are also provided. In addition to the spatial mapping.
over the beltline region, data pertinent to the maximum exposure experienced by the beltline circumferential welds as well as for the.
nozzle shell course, intermediate and lower shell plates, and the associated longitudinal welds are highlighted.
8.1 Exposure Distributions Within the Beltline Region In essence, an approach using analytically determined gradient information to extrapolate measurement results to locations.of interest within the-pressure vessel is based on the assertion that the measured values of exposure rates in the reactor cavity represent.the best available. neutron -
_ flux data-for the irradiation period in question- and, further,- on.the-assumption that the analytically determined radial distribution functions provide accurate representations of the spatial gradients that exist among the measurement locations and points of interest within the pressure vessel wall. This method is analagous to the common practice of normalizing a cycle specific forward neutron transport calculation to available measurements from either. surveillance capsule or reactor cavity dosimetry programs.
This approach provides accurate assessments of vessel exposure with
. associated uncertainties for periods of operation during which continuous 8-1
monitoring has occured. In the case of Connecticut Yankee, the cavity dosimetry program providing a complete spatial mapping of a sector of the beltline region of the pressure vessel-was installed at the start of Cycle-
- 16. Additional monitoring was limited to the five scheduled surveillance capsule withdrawals described in preceding sections of this report. The dosimetry data from these capsules provide measurement information at a-single point within the reactor geometry for the five extended' irradiation periods, but cannot be used to establish a verification of the exposure of the vessel at azimuthal locations far removed from the measurement point..
Therefore, in order to establish a baseline exposure of the pressure vessel applicable to the onset of the reactor cavity measurement program, all available core midplane measured data were combined with fuel cycle specific transport calculations to provide best estimate exposures for the first 15 cycles of operation. The reactor cavity measurements were then used directly to provide the continuous monitoring capability for Cycles 16 and beyond.
8.1.1 Baseline Exposure at the End of Cycle 15 In Table 7.1-1, comparisons of calculated and measured exposure rates for.
the five surveillance capsule dosimetry sets and for the four cavity- ,
dosimetry sets that were located on the core midplane are given. From Table 7.1-1, it was noted that, considering all of the midplane data, the calculated exposure values underpredicted measurement by factors of 0.958 for p (E > 1.0 MeV) and overpredicted measurement by factors of 1.022 and 1.023 for d (E > 0.1 MeV), and dpa/sec, respectively. The corresponding 10 standard deviations in these averages of the nine sample data sets were 6.8%, 8.9%, and 10.1%.
In developing the best estimate baseline exposure for the Connecticut Yankee reactor pressure vessel these ratios were employed as bias factors to scale the cycle specific neutron transport calculations documented in Section 4.0 of this report. In particular, the following bias _ factors were employed to establish the baseline exposures cf the vessel wall:
.8-2
_ ~ , , . . . _ _ _
3 c -'.
M/C BIAS t (E > 1.0 MeV) 1.044
+ (E > 0,1 MeV) 0.978 dpa 0.978-8.1.2 Exposure Accrued During Cycle-16 To assess the incremental exposure resulting from the Cycle 16 irradiation, the measured results from the reactor cavity multiple foil sensor sets were directly extrapolated to the vessel clad / base metal interface using the analytically derived gradient data from Section 4.0 of' this report. The axial gradient chain measurements were, of course, employed to develop the axial traverse along the vessel wall. The-extrapolated results applicable to the vessel inner surface are then combined with the Cycles 1-15 exposure to establish the best estimate exposure accrued by the reactor vessel- through the end of Cycle 16.
The end of Cycle 16 best estimate exposures at the pressure vessel clad / base metal. interface are provided in Tables 8.1-1 through 8.1-5 for 4 (E > 1.0 MeV),_in Tables 8.1-6 through 8.1-10 for 4 (E > 0.1 MeV), and in Tables 8.1-11 through 8.1-15 for dpa. In these data tables, exposures are presented as a function of axial position for five azimuthal locations around the circumference of the vessel. From these tabulations, the locations of maximum exposure of the various materials comprising the beltline region can easily be determined. Exposure distributions through the vessel wall can be developed.by normalizing the . surface exposures from Tables 8.1-1 through 8.1-15 to the appropriate radial. distribution functions given in Section 4.0 of this report.
This. exposure information,- applicable through the end of Cycle 16,- was derived from an extensive set of measurements and assures that embrittlement gradients can be established with a minimum uncertainty.
Further, as the monitoring program continues and additional data become-available, the overall plant specific data base for Connecticut Yankee 8-3 4
- J will expand resulting-in reduced uncertainties and an improved accuracy in the assesment of vessel condition.
8.1.3 Projection of Future Vessel Exposure At the end of Cycle 16, the Connecticut Yankee reactor had accrued 17.5 ;
effective full power years (EFPY) of operation.- In order to establish a framework for the assessment of future vessel condition, exposure projections to 32 EFPY tre included in Tables 8.1-1 through 8.1-15 in addition to the plant specific exposure assessments .through the end of.
Cycle 16.
These temporal extrapolations into the future were based on the assumption that a zirconium clad core using the average core power distribution derived for Cycles 1-15 was representative of all future fuel cycles.
That is, that future fuel designs would not incorporate low leakage fuel management concepts using burned fuel on the core periphery at the maximum flux locations. Future commitment to a low leakage fuel management approach would, of course, significantly reduce these exposure projections for future operation. However, quantification of the impact of low leakage operation depends on time of implementation as well as on the specifics of the fuel assemblies located on the periphery of the reactor core. Flux reductions approaching a factor of 2.0 have been achieved using conventional low leakage fuel management at a number of operating pl ants .
The projected exposure levels presented in this section can be used as a guide in assessing strategies for future vessel exposure management;'i.e.,
would conversion to Icw leakage fuel management prove to be cost.
beneficial and , if so, when should implementation take place?. The validity of these projections for future operation will be confirmed via the continued cavity monitoring program.
8-4 -
) *
-i TABLE 8.1-1
SUMMARY
OF BEST ESTIMATE FAST NEUTRON (E-> 1.0 MeV) EXPOSURE PROJECTIONS.FOR THE BELTLINE REGION OF THE CONNECTICUT YANKEE REACTOR PRESSURE VESSEL - 0 DEGREE AZIMUTHAL ANGLE HEIGHT 4 (E > 1.0 MeV) (n/cm2]
+6.0 -4.52E+18 1.llE+19
+5.0 1,46E+19 3.59E+19
+4.0 2.82E+19 6.91E419
+3.0 3.70E+19 9.07E+19
+2.0 3.97E+19 9.74E+19
+1.0 4.01E+19 9.82E+19 0.0 3.90E+19 9.57E+19
-1.0 3.80E+19 -9.32E+19
-2.0 3.90E+19 9.57E+19
-3.0 3.63E+19 8.90E+19
-4.0 2.75E+19 6.73E+19
-5.0 1.46E+19 3.58E+19
-6.0 3.90E+18 9.57E+18 Note: Height is provided relative to the axial midplane of the active core.
I l 8-5 l-
TABLE 8.1-2
SUMMARY
OF BEST ESTIMATE FAST NEUTRON (E > 1.0 MeV) EXPOSURE PROJECTIONS FOR THE BELTLINE REGION OF THE CONNECTICUT YANKEE REACTOR PRESSURE VESSEL - 10 DEGREE AZIMUTHAL ANGLE HEIGHT + (E > 1.0 MeV) [n/cm2]
+6.0 3.36E+18 8.23E+18
+5.0 1.09E+19 2.67E+19 __
+4.0 2.10E+19 5.15E+19 t +3.0 2.75E+19 6.75E+19
+2.0 2.95E+19 7.24E+19
+1.0 2.98E+19 7.31E+19 0.0 2.90E+19 7.12E+19
-1,0 2.83E+19 6.93E+19
-2.0 2.90E+19 7.12E+19
-3.0 2.70E+19 6.62E+19
-4.0 2.04E+19 5.01E+19
-5.0 1.09E+19 2.66E+19
-6.0 2.90E+18 7.12E+18 Note: Height is provided relative to the axial midplane of the active core.
8-6
p ,
TABLE 8.1-3
SUMMARY
OF BEST ESTIMATE FAST NEUTRON (E > 1.0 HeV) EXPOSUREL PROJECTIONS FOR THE BELTLINE REGION OF THE CONNECTICUT YANKEE.
- REACTOR PRESSURE VESSEL - 20 DEGREE AZIMUTHAL ANGLE HEIGHT 4 (E > 1.0 MeV) [n/cm2]
+6.0 1.60E+18 4.09E+18
+5.0 5.18E+18 1.32E+19
+4.0 1.00E+19 2.55E+19
+3.0 1.31E+19 3.35E+19
+2.0 1.41E+19 3.59E+19
+1.0 1.42E+19 3.63E+19 0.0 1.38E+19 3.53E+19
-1.0 1.35E+19 3.44E+19
-2.0 1.38E+19 3.53E+19
-3.0 1.29E+19 3.29E+19
-4.0 9.73E+18 2.49E+19
-5.0 5.17E+18 1.32E+19
-6.0 1.38E+18 3.53E+18 Note: Height is provided relative to the axial
! midplane of the active core.
- l l
l I
i t 8-7 l
d
~
TABLE 8.1-4
SUMMARY
OF BEST ESTIMATE FAST NEUTRON (E > 1.0 MeV) EXPOSURE-PROJECTIONS FOR THE BELTLINE REGION OF THE CONNECTICUT' YANKEE REACTOR PRESSURE VESSEL - 40 DEGREE AZIMUTHAL ANGLE HEIGHT + (E > 1.0 MeV) [n/cm2]
+6.0 8.45E+17 2.16E+18
+5.0 2.74E+18 7.01E+18
+4.0 5.28E+18 1.35E+19
+3.0 6.93E+18 1.77E+19
+2.0 7.44E+18 1.90E+19
+1.0 7.50E+18 1.92E+19 0.0- 7.31E+18 1.87E+19
-1.0 7.12E+18 1.82E+19
-2.0 7.31E+18 1.87E+19
-3.0 6.80E+18 1.74E+19
-4.0 5.14E+18 1.31E+19
-5.0 2.73E+18 6.99E+18
-6.0 7.31E+17 1.87E+18 Note: Height is provided relativeoto the axial midplane of the active core.
8-8
a- .
14 TABLE 8.1
SUMMARY
OF BEST ESTIMATE FAST NEUTRON (E > 1.0'MeV) EXPOSURE PROJECTIONS FOR THE BELTLINE REGION OF THE-CONNECTICUT-YANKEE REACTOR PRESSURE VESSEL - 45 DEGREE AZIMUTHAL ANGLE HEIGHT 4 (E > 1.0 MeV) [n/cm2)
+6.0 8.05E+17 2.09E+18
+5.0 2.61E+18 6.77E+18
+4.0 5.03E+18 1.31E+19
+3.0 6.60E+18 1.71E+19
+2.0 7.08E+18 1.84E+19
+1.0 7.14E+18 1.85E+19 0.0 6.96E+18 1.81E+19
-1.0 6.78E+18- 1.76E+19
-2.0 6.96E+18 1.81E+19
-3.0 6.48E+18 1.68E+19
-4.0 4.90E+18 1.27E+19
-5.0 2.60E+18 6.76E+18
-6.0 6.96E+17- 1.81E+18 Note: Height is provided relative to the axial midplane of the active core.
-s 8-9 4 1
.)
TABLE 8.1-6'-
SUMMARY
OF BEST ESTIMATE-FAST NEUIRON (E >:0.1 MeV) EXPOSURE-PROJECTIONS FOR THE DELILINE REGION OF THE CONNECTICUT YANKEE
- REACTOR PRESSURE VESSEL - 0 DEGREE AZIMUTHAL ANGLE-HEIGHT + (E > 0.1 MeV) [n/cm2)
+6.0 1.27E+19 2.60E+19
+5.0 4.13E+19 8.41E+19
+4.0 7.95E+19 1.62E+20
+3.0 1.04E+20 2.13E+20
+2.0 1.12E+20 2.28E+20
+1.0 1.13E+20 2.30E+20 0.0 1.10E+20 2.24E+20
-1.0 1.07E+20 2.19E+20
-2.0 1.10E+20 2.24E+20
-3.0 1.02E+20 2.09E+20
-4.0 7.74E+19 1.58E+20
-5.0 4.12E+19 8.39E+19 l -6.0 1.10E+19 2.24E+19 Note: Height is provided relative to the axial midplane of the active core.
l
- l. 8-10 t
TABLE-8.1-7'
SUMMARY
OF BEST ESTIMATE FAST NEUTRON (E > 0.1 HeV) EXPOSURE PROJECTIONS FOR-THE_ BELTLINE REGION OF THE CONNECTICUT YANKEE REACTOR PRESSURE VESSEL - 10 DEGREE AZIMUTHAL ANGLE HEIGHT 4 (E > 0.1 MeV) [n/cm2]
+0.0 9.43E+18 1.82E+19
+5.0 3.06E+19 5.89E+19
+4.0 5.89E+19- 1 14E+20
+3.0 7.73E+19 1.49E+20.
+2.0 8.29E+19 1.60E+20
+1.0 8.36E+19 1.61E+20 0.0 8.15E+19 1.57E+20
-1.0 7.94E+19- 1.53E+20
-2.0 8.15E+19 1.57E+20
-3.0 7.58E+19 1.46E+20
-4.0 5.73E+19 1.llE+20
-5.0 3.05E+19 5.88E+19
-6.0 8.15E+18 1.57E+19 Note: Height is provided relative to the axial midplane of the active core.
8-11
TABLE-8.1-8
SUMMARY
Of BEST ESTIMATE FAST NEUTRON (E > 0.1 MeV) EXPOSURE PROJECTIONS FOR THE BELTLINE REGION OF THE CONNECTICUT YANKEE REACTOR PRESSURE VESSEL - 20 DEGREE AZIMUTHAL ANGLE HEIGHT 4 (E > 0.1 MeV) [n/cm2]
+6.0 4.27E+18 9.22E+18
+5.0 1.38E+19 2.99E+19
+4.0 2.67E+19 5.76E+19
+3.0 3.50E+19 7 56E+19
+2.0 3.75E+19 8.11E+19
+1.0 3.79E+19 8.18E+19 0.0 3.69E+19 7.97E+19
-1.0 3.59E+19 7.77E+19
-2.0 3.69E+19 7.97E+19
-3.0 3.43E+19 7.42E+19 ,
-4.0 2.60E+19 5.61E+19
-5.0 1.38E+19 2.98E+19
-6.0 3.69E+18 7.97E+18 Note: Height is provided relative to the axial midplane of the active core.
L j ..
8-12
TABLE 8.1-9
SUMMARY
OF BEST ES11 MATE FAST. NEUTRON (E-> 0.1_MeV) EXPOSURE PROJECTIONS FOR THE BELTLINE-REGION OF THE CONNECTICUT YANKEE REACTOR PRESSURE VESSEL - 40 DEGREE AZIMUTHAL ANGLE HEIGHT 4 (E > 0.1 MeV) [n/cm2)
+6.0 2.14E+18 4,71E+18
+5.0 6.93E+18 1.53E+19
+4.0 1.34E+19 2.94E+19
+3.0 1.75E+19 3.86E+19
+2.0 1.88E+19 4.15E+19
+1.0 1.90E+19 4.18E+19 0.0 1.85E+19 4.07E+19
-1.0 1.80E+19 3.97E+19
-2.0 1.85E+19 4.07E+19
-3.0 1.72E+19 3.79E+19
-4.0 1.30E+19 2.87E+19
-5.0 6.92E+18 1.52E+19
-6.0 1.85E+18 4.07E+18 Note: Height is provided relative to the axial midplane of the active core,
, 4 8-13
TABLE 8.1-10
SUMMARY
OF BEST ESTIMATE FAST NEUTRON (E > 0.1 MeV) EXPOSURE PROJECTIONS FOR THE BEL 1LINE REGION OF THE CONNECTICllT. YANKEE-REACTOR PRESSURE VESSEL - 45 DEGREE AZIMUTHAL ANGLE-HEIGHT 4 (E > 0.1 MeV) [n/cm2)
+6.0 2.02E+18 4.50E+18
+5.0 6.56E+18 1.46E+19
+4.0 1.27E+19 2.81E+19
+3.0 1.66E+19 3.69E+19
+2.0 1.78E+19 3.96E+19
+1.0 1.80E+19 4.00E+19
]
0.0 1.75E+19 3.89E+19
-1.0 1.71E+19 3.79E+19
-2.0 1.75E+19 3.89E+19
-3.0 l'.63E+19 3.62E+19
-4.0 1.23E+19 2.74E+19
-5.0 6.55E+18 1.46E+19
-6.0 1.75E+18 3.89E+18 Note: Height is provided relative to the axial midplane of the active core.
i l
l 8-14 l
TABLE 8.1-11 L
SUMMARY
OF BEST ESTIMATE' IRON ATOM DISPLACEMENT
~
PROJECTIONS FOR THE BELTLINE REGION OF THE-CONNECTICUT YANKEE REACTOR PRESSURE VESSEL - 0 DEGREE AZIMUTHAL ANGLE ,
V HEIGHT IRON DISPLACEMENTS [dpa)
(ft) 17,5 EFPY 32 EFPY 46,0 7.25E-03 1.67E-02 -
+5.0 2.35E-02 5.40E-02
+4.0 4.53E-02 1.04E-01
+3.0 5.95E-02 1.37E-01
+2.0 6.38E-02 1.47E +1.0 6.44E-02 1.48E-01 0.0 6.27E-02 1.44E-01
-1.0 6.11E-02 1.40E-01
-2.0 6.27E-02 1.44E-01
-3.0 5.84E-02 1.34 E-01
-4.0 4.41E-02 1.01E-01
-5.0 2.35E-02 5.39E-02
-6.0 6.27E-03 1.44E-02 Note: Height is provided reletive to the axial midplane of the active core, 8-15 r 9
h TABLE 8.1-12
SUMMARY
OF BEST ESTIMATE IRON ATOM DISPLACEMENT PROJECTIONS FOR THE BELTLINE REGION OF'THE CONNECTICUT YANKEE REACTOR PRESSURE VESSEL - 10 DEGREE AZIMUTHAL ANGLE-HEIGHT IRON DISPLACEMENTS [dpa)
(ft) 17.5 EFPY 32 EFPY 46.0 5.39E-03 1.24E-02
+5.0 1.75E-02 4.01E-02
+4.0 3.37E-02 7.73E-02 :
+3.0 4.42E-02 1.01E-01
+2.0 4.74E-02 1.09E-01 41.0 4.79E-02 1.10E-01 0.0 4.66E-02 1.07E-01
-1.0 4.54E-02 1.04E-01
-2.0 4.66E-02 1.07E-01 1
-3.0 4.34E-02 9.95E-02
-4.0 3.28E-02 7.52E-02
-5.0 1.74E-02 4.00E-02
-6.0 4.66E-03 1.07E-02 Note: Height is provided relative to the axial midplane of the active' core.
8-16
v- ,
TABLE 8.1-13
SUMMARY
OF BEST ESTIMATE IRON ATOM DISPLACEHFNT o
PROJECTIONS FOR.THE BELTLINE REGION OF THE-CONNECTICUT YANKEE REACTOR PRESSURE VESSEL - 20 DEGREE AZIMUTHAL ANGLE 1
HEIGHT IRON DISPLACEMENTS [dpa)
+6.0 2.54E-03 6.15E-03
+5.0 8.22E-03 1.99E-02
+4.0 1.58E-02 3.84E-02
+3.0 2.08E-02 5.04E-02
+2.0 2.23E-02 5.41E-02
+1.0 2.25E-02 5.46E-02 0.0 2.19E-02 5.32E-02
-1.0 2.14E-02 5.18E-02
-2.0 2.19E-02 5.32E-02
-3,0 2.04E-02 4.95E-02
-4.0 1.54E-02 3.74E-02
-5.0 8.20E-03 1.99E-02
-6.0 2.19E-03 5.32E-03 i Note: He19 ht is provided relative to the axial midplane of the active core.
l l
i i
8-17 a
i
r
- (
TABLE 8.1-14
SUMMARY
OF BEST ESTIMATE IRON ATOM DISPLACEMENT PROJECTIONS FOR THE BELTLINE REGION OF THE CONNECTICUT YANKEE-REACTOR PRESSURE VESSEL - 40 DEGREE AZIMUTHAL ANGLE HEIGHT IRON DISPLACEMENTS [dpa)
+6.0 1.32E-03 3.24E-03
+5.0 4.28E-03 1.05E-02
+4.0 8.25E-03 2.02E-02
+3.0 1.08E-02 2.65E-02
+2.0 1.16E-02 2.85E-02
+1.0 1.17E-02 2.87E-02 0.0 1.14E-02 2.80E-02
-1.0 1.11E-02 2.73E-02
-2.0 1.14E-02 2.80E-02
-3,0 1.06E-02 2.61E-02
-4.0 8.04E-03 1.97E-02
-5.0 4.27E-03 1.05E-02
-6.0 1.14E-03 2.80E-03 Note: Height is provided relative to the axial midplane of the active core.
8-18
L- ,
TABLE 8.1-15
SUMMARY
OF BEST ESTIMATE IRON ATOM DISPLACEMENT PROJECTIONS-FOR=THE BELTLINE REGION OF THE CONNECTICUT YANKEE REACTOR PRESSURE VESSEL - 45 DEGREE AZIMUTHAL ANGLE HEIGHT 1RON DISPLACEMENTS (dpa) .
+6.0 1.26E-03 3.13E-03
+5.0 4.07E-03 1.01E-02
+4.0 7.86E-03 1.95E-02
+3.0 1.03E-02 2.56E-02
+2.0 1.llE-02 2.75E-02
+1.0 1.12E-02 2.78E-02 0.0 1.09E-02 2.70E-02
-1.0 1.06E-02 2.63E-02
-2.0 1.09E-02 2.70E-02
-3.0 1.01E-02 2.52E-02
-4.0 7.65E-03 1.90E-02
-5.0 4.07E-03 1.01E-02 <
-6.0- 1.09E-03 2.70E-03 Note: Height is provided realative to the axial midplane of the active core.
i 8-19
8 8.2 Exposure of Specific Beltline Materials As shown in. Figures 2.1-1 and 2.1-2, the beltline region of the Connecticut Yankee reactor pressure vessel is comprised of a nozzle shell course, an intermediate shell course, and a lower shell course, each consisting of three shell plates and three longitudinal welds.
Circumferential welds join the three shell courses to form the highly irradiated portion of the reactor vessel wall.
The upper and lower circumferential welds are centered 54.975 inches above the axial midplane of the active core and 60.525 inches below the axial midplane of the' active core, respectively. The intermediate shell course is positioned between these two circumferential welds, in each of the l three shell courses, longitudinal welds are located at azimuthal angles of ,
10, 20, and 40 degrees relative to the core cardinal axes and exhibit the same axial extent as the respective shell courses. The maximum neutron exposure experienced by each of these beltline materials can be extracted from the data provided in Tables 8.1-1 through 8.1-15, i
The current (End of Cycle 16) and projected maximum exposures.of.the l beltline circumferential welds are 1.isted in Table 8.2-1. In this table, the weld exposure is expressed in terms of f (E > 1.0 MeV),
4 (E > 0.1 MeV), and dpa. Similar data applicable to the intermediate shell course is provided in Table 8.2-2. Data pertinent to the plates comprising the nozzle shell course and lower shell courses are listed in '
Table 8.2-3. The maximum exposure of all longitudinal welds is provided in Table 8.2-4.
The maximum integrated exposure of each of the beltline materials is also illustrated graphically in Figures 8.2-1 through 8.2-9. In Figures 8.2-1 through 8.2-3, the maximum exposure of the intermediate shell plates is ,
presented as a function of azimuthal angle for exposures expressed in terms of f (E > 1.0 MeV), 4 (E > 0.1) MeV, and dpa, respectively.- 1 Similar data applicable to the upper circumferential weld and nozzle shell course is provided in Figures B.2-4 through 8.2-6 and data applicable to i
8-20 P
1 .- .
the lower circumferential weld and the lower shell. course are shown in Figures 8.2-7fthrough 8.2-9.
In. regard to Figures 8.2-1 through 8.2-9, the data depicted at the 10, 20, L and 40 degree azimuthal locations represent maximum exposures of the individual longitudinal welds present in each of the shell courses, Specific locations of each of these longitudinal welds as well.as the azimuthal span of the various plates comprising each shell course can be obtained from Figure 2,1-1.
8-21
,~*
i TABLE 8.2-1 MAXIMUM FAST NEUTRON EXPOSURE OF CONNECTICUT YANKEE BELTLINE CIRCUMFERENTIAL WELDS 4- (E > 1.0 MeV) In/cm2]
AZIMUTHAL UPPER WELD ' LOWER WELD ANGLE 17.5 EFPY 32.0 EFPY lime _F.[EX 32.0 EFPY 0 DEGREES 2.03E+19 4.98Ee19 1.41E+19 3.47E+19 10 DEGREES 1.51E+19 3.71E+19 1.06E+19 2.57E+19 i 20 DEGREES 7.20E+18 1.84E+19 5.00E+18 1.28E+19 40 DEGREES 3.81E+18 9.74E+18 2.64E+18 6.77E+18 45 DEGREES 3.63E+18 9.43E418 2.52E+18 6.54E+18 4 (E > 0.1 MeV) in/cm21 AZIMUTHAL UPPER WELD LOWER WELD- '
ANGLE 17.5 EFPY 11u0 EFPY 17.5 EFPY 32.0 EFPY 0 DEGREES 5.73E+19 1.17E+20 3.99E+19 8.12E+19 10 DEGREES 4,25E+19 8.20E+19 2.95E+19 5.69E+19 20 DEGREES 1.92E+19 4.15E+19 1.34E+19 2.88E+19 40 DEGREES 9.65E+18 2.12E+19 6.70E+18 1.47E+19 45 DEGREES 9.14E+18 2.03E+19 .6.34E+19 1.41E+19 IRON ATOM DISPLACEMENTS Idea]
AZIMUTHAL UPPER WELD LOWER WELD ANGLE 17.5 EFPY 32.0 EFP1 17.5 EFPY 32.0 EFPY 0 DEGREES 3.27E-02 7.50E-02 2.27E-02 5.22E-02 10 DEGREES 2.43E-02 5.57E-02 1.68E-02 3.87E-02 20-DEGREES 1.14E-02 2.77E-02 7.94E-03 1.93E-02 40 DEGREES 5.95E-03 1.46E-02 4.13E-03 1.02E-02 45 DEGREES 5.66E-03 1.40E-02 3.94E-03 9.78E-03 8-22
- '{ a TABLE 8.2-2 MAXIMUM FAST NEUTRON EXPOSURE OF CONNECTICUT YANKEE
!NTERMEDIATE SHELL PLATES i_(E > 1.0 MeV) In/cm21 AZIMUTHAL ANGLE 17.5 EFPY 32.0 EFPY 0 DEGREES 4.01E+19 9.82E+19 10 DEGREES 2.98E+19 7.31E+19 20 DEGREES 1.42E+19 3.63E+19 40 DEGREES 7.50E+18 1.92E+19 45 DEGREES 7.14E+18 1.85E+19 4-(E > 0.1 MeV) In/cm21 AZIMUTHAL ANGLE 17.5 EFPY 32.0 EFPY 0 DEGREES 1.13E+20 2.30E+20 10 DEGREES 8.36E+19 1.61E+20 20 DEGREES 3.79E+19 8.19E+19 40 DEGREES 1.90E+19 4.18E+19 45 DEGREES 1.80E+19 4.00E+19 IRON ATOM DISPLACEMENTS Idoal AZIMUTHAL' ANGLE 17.5 EFPY 32.0 EFPY 0 DEGREES- 6.44E-02 1.40E-01 10 DEGREES 4.79E-02 1.10E-01 20 DEGREES 2.25E-02 5.46E-02 40 DEGREES 1.17E-02 2.87E-02 45 DEGREES 1.12E-02 2.78E-02 8-23
TABLE 8.2-3 MAXIMUM FAST NEUTRON EXPOSURE OF CONNECTICUT YANKEE N0ZZLE COURSE AND LOWER SHELL PLATES f (E > 1.0 MeV) In/cm&J AZIMUTHAL N0ZZLE COURSE LOWER PLATES ANGLE 17.5 EFPY 32.0 EFPY 17.5 EFPY- 32.0 EFPY 0 DEGREES 2.03E+19- 4.98E+19 1.41E+19 3.47E+19 10 DEGREES 1.51E+19 3.71E+19 1.06E+19 2.57E+19 20 DEGREES 7.20E+18 1.84E+19 5.00E+18 1.28E+19 40 DEGREES 3.81E+18 9.74E+18 2.64E+18 6.77E+18 45 DEGREES 3.63E+18 9.43E+18 2.52E+18 6.54E+18 Y
4 (E > 0.1 MeV) In/cm21 AZIMUTHAL N0ZZLE COURSE LOWER PLATES ANGLE 17.5 EFPY 32.0 EFPY 17.5 EFPY 32.0 EFPY 0 DEGREES 5.73E+19 1.17E+20 3.99E+19 8.12E+19 10 DEGREES 4.25E+19 8.20E+19 2.95E+19 5.69E+19 20 DEGREES 1.92E+19 4.15E+19 1.34E+19 2,88E+19 40 DEGREES 9.65E+18 2.12E+19 6.70E+18 1.47E+19 45 DEGREES 9.14E+18 2.03E+19 6.34E+19 1.41E+19 IRON ATOM DISPLACEMENTS Tdoa]'
AZIMUTHAL N0ZZLE COURSE LOWER PLATES.
ANGLE 17.5 EFPY 32.0 EFPY 17.5 EFPY 32.0 EFPY 0 DEGREES 3.27E-02 7.50E-02 2.27E-02 5.22E-02 10 DEGREES 2.43E-02 5.57E-02 1.68E-02 3.87E-02' 20 DEGREES 1.14E-02 2.77E-02 7.94E-03 1.93E 40 DEGREES 5.95E-03 1.46E-02 4.13E-03 1.02E 45 DEGREES 5.66E-03 1.40E-02 3.94E-03 9.78E-03 8-24
V ..
TABLE 8.2-4 MAXIMUM FAST NEUTRON EXPOSURE OF CONNECTICUT YANKEE BELTLINE LONGITUDINAL WELDS 4 (E > 1.0 MeV) In/cm2)
AZIMUTHAL N0ZZLE COURSE INTERMEDIATE COURSE LOWER COURSE.
ANGLE 17.5 EFPY 32,0 FFPY 17.5 EFPY 32.0 EFPY 17.5 EFPY 32.0 EFPY 10 DEGREES 1.51E+19 3.71E+19 2.98E+19 7.31E+19 1.06E+19 2.57E+19 20 DEGREES 7.20E+18 1.84E+19 1.42E+19 3.63E+19 5.00E+18 1.28E+19 40 DEGREES 3.81E+18 9.74E+18 7.50E+18 1.92E+19 2.64E+18 6.77E+18 4 (E > 0.1 MeV)'in/cm2)
AZIMUTHAL N0ZZLE COURSE INTERMEDIATE COURSE LOWER COURSE ANGLE 17.5 EFPY 32.0 EFPY 17.5 EFP1 32.0 EFPY 17.5 EFPY 32.0 EFPY 10 DEGREES 1.51E+19 3.71E+19 8.36E+19 J.61E+20 1.06E+19 2.57E+19 20 DEGREES 7.20E+18 1.84E+19 3.79E+19 8.19E+19 5.00E+18- 1.28E+19 40 DEGREES 3.81E+18 9.74E+18 1.90Ea19 4.18E+19 2.64E+18 6.77E+18 IRON ATOM DISPLACEMENTS Idoa]
AZIMUTHAL N0ZZl.E COURSE INTERMEDIATE COURSE LOWER COURSE
__ ANGLE' 17.5 EFPY 32.0 EFPY 17.5 EFPY 32.0 EFPY 17.5 EFPY 32.0 EFPY 10 DEGREES 1.51E+19 3.71E+19 4'.79E-02 1.10E-01 1.06E+19- 2.57E+19 20 DEGREES 7.20E+18 1.84E+19 2.25E-02 5.46E-02 5.00E+18 1.'28E+19 40 DEGREES 3.81E+18 .9.74E+18 1.17E-02 2.87E-02 2.64E+18' 6.77E+18 8-25 t
FIGURE 8'.2-1
. MAXIMUM FAST NEUTRON
- FLUENCE (E > 1.0 MeV) AS A FUNCTION OF AZIMUTHAL ~
. ANGLE AT THE-INNER RADIUS OF THE INTERNEDIATE SHELL PLATES' 1.0E+20: ~
NL x
-T.
g .N .
- E '7N '
. Ok N
{ K 'N T ,
K N -
=
8 . ..
m1 ,0E+19 N_ ~
k n a
C O
2: . u. ,~ , >
c -
3 j +,
s
, ~ , . 4 y
. ms
' \ _
~
1,0E+18
^ '
O 5 10 -15 20 25 30- 35 40. 45-Azimuthal Angle (Degrees)
L i
l 8-26.
1.
FIGURE 8.2-2 MAX 1HUMFASTNEUTR0flFLVENCE(E>0.1HeV)ASAFUNCTIONOfAZlHUTHAL ANGLE AT THE INNER RADIUS Of THE 'NTERhi.DIATE 5 HELL PLATES 1.0 E + 21 -
N b
s i,x w
u
% s N
gE 1.0E+20b 7 _,, _ 'm
- u. -
,s N. - -- _ . .
f a N
$ m ._
1.0E+19 -
0 5 10 15 20 25 30 35 40 45 Azimuthal Angle (Degrees)
+ 17.5 EFPY +- 32.0 EFPY 8-27 ]
r!GURE 8.2-3 KAXIMUM 1RON ATOM blSPLACEMEN15 [dpa) AS A FUNCTION Of AZIMUTHAL A!.GLE AT THE INNER RADIUS Of THE IN1ERMEDIATE SHELL PLATES 1.0E+00 o.
8.
2 5
E '
w im 5 1.0E 01 N y _
6 "- N
- x. m E -- N -- -
_. N 2
< N '
w -
vm 8 ' s
& N
--m 1.0E-02 ",
0 5 10 15 20 25 30 35 40 45 Azimuthal Angle (Degrees)
+ 17.5 EFPY + 32.0 EFPY 8-28 I
FIGURE 8.2-4 HAXIMUM FAST NEUTR0tl FLUENCE (E > 1.0 MeV) AS A FUNCTION OF AZIMUTHAL ANGLE AT THE INNER RADIUS OF THE UPPER CIRCUMFERENTIAL WELD AND THE N0ZZLE SHELL COURSE 1.0E+20 g ^ x.
E o N s, c
e' w ,
N- '1 8
g 1.0E+19
\ s N ~
g . _ _
8 %_
g ,
1,0 E + 18 '
'I O 5 10 15 20 25 30 35 40 45 Azimuthal Angle (Degrees)
FIGURE 8.2-5 MAX 1 HUM FAST NEUTRON' FLUENCE (E > 0.1 HeV) AS A FUNCTION Of AZlHUTHAL ANGLE AT THE INNER RADIUS Of THE UPPER CIRCUMfERENTIAL WELD AND THE N0ZZLE SHELL COURSE 1.0 E + 21 R
E U
' 1.0 E + 20"'
s ~ ~ '" bw_ --
8 0% ' ' , .
x N
g ' , .
~
2 % ~ . ,,
1,0 E + 19 %
~5 1,0E+18 0 5 10 15 20 25 30 35 40 45 Azimuthal Angle (Degrees)
+ 17.5 EFPY + 32.0 EFPY B-30 1
FIGURE 8.2-6 MAXIMUM 1RON ATOM blSPLACEMENTS [dpa) AS A FUNCTION OF AZlHUTHAL ANGLE AT THE INNER RADIUS OF THE UPPER CIRCUMFERENTIAL WELD AND N0ZZLE SHELL COURSE 1.0E+00 2
E b 1.0E 01 g " - -
.$ n- N a mw '
E y 1.0E 02 s'A ,_
_="
c __
- c ,,
2:
1.0E-03 0 5 10 15 20 25 30 35 40 45
( Azimuthal Angle (Degrees)
FIGURE 8.2-7 MAXIMUM FAST NEUTROM FLUENCE (E > 1.0 HeV) AS A FUNCTION OF AZlHUTHAL ANGLE AT THE INNER RADIUS OF THE LOWER CIRCUHFERENTIAL WELD AND THE LOWER SHELL COURSE 1.0E+20 k? l i "#%
E 'A , ,j A. j, ,.
( hy , .,
1.0E+19 K -
E x ,
.-n C N '
O "N s N m g NH,
. 1.0E+18 -
0 5 10 15 20 25 30 35 40 45 Azimuthal Angle (Degrees)
FIGURE 8.2-8 MAX 1 HUM FAST NEUTR0!( FLUENCE (E > 0.1 HeV) AS A FUNCTION OF AZIMUTHAL ANGLE AT THE INNER RADIUS OF 1HE LOWER CIRCUMFERENTIAL WELD AND THE LOWER SHELL COURSE ji 1.0 E + 20 .
^
~
R e' ' N N E N
$ N N m 5 \
N E 1.0E+19 -
.2 -
u.
, ,-u 5
s
- E 1.0E+18 0 5 10 15 20 25 30 35 40 45 Azimuthal Angle (Degrees)
1 FIGURE 8.2-9 HAX1 HUM 1RON ATOM DISPLACEMENTS [dpa) AS A FUNCTION OF AZIMUTHAL ANGLE AT THE INNER RADIUS OF THE LOWER CIRCUHFERENTIAL WELD AND LOWER SHELL COURSE ,
1.0E+00 l
$m h1.0E-01 (
m ..
.$ N O w N_
-<$ 1.0E-02
\ %_
mn '
C %
g - - -
~. -
1.0E-03 '
0 5 10 15 20 25 30 35 40 45 Azimuthal Angle (Degrees)
=
s..... ..
j .
8.3 Uncertainties in Exposure Projections The overall uncertainties associated with the exposure rater and integrated exposures determined for Connecticut Yankee stem from two basic
. sources; the accuracy of the neutron flux measurements at the sensor set locations and the accuracy of the radial gradient projections derived from the use of the transport code. Based on the least squares adjustment approach used in the FERRET analyses the lo uncertainties in the measured data were as follows:
la UNCERTAINTY CAPSULE CAVITY Flux (E > 1.0 MeV) 8-15% 8%
Flux (E > 0.1 MeV) 14-24% 17%
dpa/sec 10-18% 14%
These values represent uncertainties derived from the reaction rate measurements and from the least squares fit of the output spectrum to the measured data. As additional data is obtained from the ongoing measurement program, the knowledge of the neutron spectra at the measurement locations will increase and the uncertainties in the measured exposure parameters will be reduced somewhat.
Since the ultimate goal of the cavity measurement program is the evaluation of the exposure of the vessel itself, an additional uncertainty associated with the ability to translate results from the measurement locations to the points of interest within the vessel must be included along with the measurement uncertainties listed above. Information pertinent to this extrapolation uncertainty has been obtained from benchmarking studies using the Westinghouse neutron transport methodology and from several comparisons of power reactor internal surveillance capsule dosimetry and reactor cavity dosimetry for which the irradiation history of all sensors was the same.
8-35
9 6
Based on these benchmarking evaluations the uncertainty or bias associated with the calculated slope through the steel vessel was estimated to be approximately 5% for all e.tposure parameters. Thus, the total uncertainty associated with projections at the clad / base uetal interface is estimated to be as follows for each exposure parameter of interest.
lo_VNCERTAINTY VESSEL 1R flux (E > 1.0 MeV) 15%
flux (E > 0.1 MeV) 23%
dpa/sec 19%
Use of these values represents the bounding 10 uncertainties for vessel exposure, since with penetration iato the vessel wall the extrapolation uncertainty lessens until at the outer surface the overall uncertainty reverts simply to the measurement uncertainty. Again, as more data are accumulated from both reactor cavity and surveillance capsule dosimetry sets, the extrapolation uncertainty will also be reduced resulting in higher levels of accuracy in the vessel exposure projections.
8-36
SECTION 9.0 REFEREllCES
- 1. Anderson, S. L. and fero, A. H., " Reactor Cavity Neutron Measurement Program for Connecticut Yankee Atomic Power Company Haddam Neck Plant,"
WCAP-12549, August 1989.
- 2. Yanichko, S. E., et. al., " Connecticut Yankee Reactor Vessel Radiation Surveillante Program," WCAP-7036, April 1967.
- 3. Soltesz, R. G. , et. al., " Nuclear Rocket Shielding Methods, Modification, Updating, and input Data Preparation - Volume 5 - Two Dimensional Discrete Ordinates Transport Technique," WANL-PR-(LL)-034, I.ugust 1970.
- 4. SAILOR RSIC DATA LIBRARY COLLECTION DLC-76, " Coupled Self-Shielded, 47 Neutron, 20 Gamma Ray, P3, Cross Section Library for Light Water Reactors.
- 5. ASTM Designation E706-87, " Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standards," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa.,1989
- 6. ASTM Designation E853-87, " Standard Practice for Analysis and Interpretation of Light -Water Reactor Surveillance Results," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa. ,1989.
- 7. ASTM Designation E261-77, " Standard Method for Determining Neutron flux, Fluence, and Spectra by Radioactivation Techniques," in ASTM Standards, Section 12 American Society for Testing and Materials, Philadelphia, Pa.
1989.
9-1 l
l i
I
e
- 8. ASTM Designation E262-86, " Standard Method for Measuring Thermal Neutron flux by Radioactivation Techniques," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa.1989.
- 9. ASTM Designation E263-88, " Standard Method for Determining fast Neutron flux Density by Radioactivation of Iron,* in AS1H Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa.,1989.
- 10. ASTM Designation E264-87, " Standard Method for Determining fast Neutron flux Density by Radioactivation of Nickel," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa.,1989.
a
- 11. ASTM Designation E481-86, " Standard Method for Measuring Neutron flux Density by Radioactivation of Cobalt and Silver," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia. Pa.,
1989.
- 12. ASTM Designation E523-87, " Standard Method for Determining fast Neutron Flux Density by Radioactivation of Copper," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa.,1989,
- 13. ASTM Designation E704-84, " Standard Method for Measuring Reaction Rates by Radioactivation of Uranium-238," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa., 1989.
- 14. ASTM Designation E705-84, " Standard Method for Determining fast Neutron flux Density by Radioactivation of Neptunium-237," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa.,
1989.
- 15. AS1H Designation E1005-84, " Standard Method for Application and Analysis of Radiometric Monitors for Reactor Vessel Surveillance," in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, Pa. ,1989.
9-2
- 16. Schmittroth, E. A., " FERRET Data Analysis Code", HEDL-THE-79-40, Hanford Engineering Development Laboratory, Richland, Washington, September 1979.
- 17. McElroy, W. N., et. al., 'A Computer-Automated Iterative Method of Neutron flux Spectra Determined by foil Activation," AfWL-TR-67-41. Volumes I-IV, Air force Weapons Laboratory, Kirkland AFB, NM, July 1967.
- 18. Macrker, R. E. as reported by Stallman, F. W., " Workshop on Adjustment Codes and Uncertainties - Proc. of the 4th ASTM / EURATOM Symposium on Reactor Dosimetry," NUREG/CP-0029, NRC, Washington, D.C., July 1982.
- 19. Ireland, D. R., et. al., " Examination and Evaluation of Capsule A for the Connecticut Yankee Reactor Pressure Vessel Surveillance Program",
Battelle Memorial Institute Report, October 30, 1970.
- 20. Perrin, J. S., et. al., " Examination and Evaluation of Capsule f for the Connecticut Yankee Reactor Pressure Vessel Surveillance Program",
Battelle Memorial Institute Report, October March 30, 1972.
9-3
APPENDIX A SPECIFIC ACilVITIES AND 1RRADIAT!0N HISTORY OF SENSORS FROM SURVEILLANCE CAPSULES H, D, AND E in this appendix, the irradiation history as extracted from NVREG-0020 and the measured specific activities of radiometric sensors irradiated in surveillance Capsules H, D, and E are provided. ,
The irradiation history of capsules withdrawn to date was as follows:
CYCLE NO. STARTUP SHVTDOWN COMMENT 1 10/01/67 04/17/70 CAPSVLE A WITHDRAWN 2 06/26/70 04/16/71 3 05/26/71 06/10/72 4 07/14/72 07/08/73 CAPSULE F WITHDRAWN 5 12/14/73 05/17/75 6 07/01/75 05/18/76 7 08/18/76 10/14/77 CAPSULE H WITHDRAWN 8 12/01/77 01/27/79 9 03/12/79 05/03/80 10 07/27/80 09/26/81 CAPSVLE D WITHDRAWN 11 11/12/81 01/22/83 12 04/11/83 08/01/84 13 11/09/84 01/04/86 14 05/10/86 07/18/87 15 03/26/88 09/02/89 CAPSULE E WITHDRAWN RET. CORE POWER - 1825 MWt The monthly thermal generation applicable to the Connecticut Yankee reactor is prov'ided on pages A-2 and A-3. Pages A-4 through A-6 contain the measured specific activities of sensors removed from Capsules H, 0, and E.
A-1
MONTHLY THERMAL GENERATION DURING THE flRST FIFTEEN FUEL CYCLES OF THE CONNECTICUT YANKEE REACTOR THERMAL THERMAL THERMAL THERHAL GENERATION GENERATION GENERATION GENERATION MQtilB _(MW-hr) MQMIB (HW-hr) MQJilli (MW-bri tiQ!ild _IMW-br) 10/67 183334 5/71 176431 12/74 1305172 7/78 1338395 11/67 458883 6/71 1077628 1/75 1346827 8/78 1331075 12/67 1068302 7/71 1183774 2/75 1151632 9/78 1283751 1/68 731800 8/71 1195500 3/75 1257899 10/78 1356276 2/68 1022594 9/71 1258689 4/75 1302962 11/78 1296933 3/68 60807 10/71 1260876 5/75 663735 12/78 1353375 4/68 373279 11/71 1319133 6/75 0 1/79 1138678 5/68 1085167 12/71 1303127 7/75 1207342 2/79 0 6/68 869054 1/72 1320342 8/; 5 1348784 3/79 759242 7/68 927058 2/72 1221797 9/75 1283524 4/79 1311567 8/68 721120 3/72 1348820 10/75 1353019 5/79 1314982 9/68 872314 4/72 1300120 11/75, 1304298 6/79 1300611 10/68 1114992 5/72 1301311 12/75 1183964 7/79 1138807 11/68 922092 6/72 372870 1/76 1300149 8/79 1355853 12/68 953074 7/72 591844 2/76 1259935 9/79 1231704 1/69 1091961 8/72 1313482 3/76 1344491 10/79 983672 2/69 973577 9/72 1121490 4/76 1181768 11/79 1310839 3/69 1206750 10/72 1350380 5/76 777717 12/79 1356111 4/69 477375 11/72 1166735 6/76 0 1/80 1356098 5/69 584061 12/72 1350313 7/76 518160 2/80 1200939 6/69 827562 1/73 1345262 8/76 1308570 3/80 1301936 7/69 937842 2/73 1105055 9/76 1274249 4/80 1145726 8/69 968428 3/73 1354896 10/76 1348482 5/80 85371 9/69 1058024 4/73 1294180 11/76 1304841 6/80 0 10/69 1201267 $/73 1339215 12/76 1342614 7/80 107212 11/69 1050895 6/73 549134 1/77 1351509 8/80 1185534 12/69 1323581 7/73 269383 2/77 1220694 9/80 1159150 1/70 1349932 8/73 0 3/77 1351182 10/80 1355070 2/70 1116667 9/73 0 4/77 1305897 11/80 1184795 3/70 981894 10/73 0 5/77 1351705 12/80 1328413 4/70 513871 11/73 0 6/77 1313095 1/81 1294197 5/70 0 12/73 504037 7/77 1354885 2/81 1167760 6/70 48987 1/74 1282676 8/77 999830 3/81 1349920 7/70 1248974 2/74 1167558 9/77 1280708 4/81 1294385 8/70 1133392 3/74 960432 10/77 604102 5/81 1336224 9/70 1054033 4/74 387823 11/77 0 6/81 1310216 10/70 1288120 5/74 1309656 12/77 972672 7/81 1244286 11/70 1284678 6/74 1236073 1/78 1114568 8/81 1245695 12/70 1312583 7/74 1346981 2/78 1128610 9/81 1051704 1/71 1349689 8/74 1348241 3/78 1206607 10/81 0 2/71 1202303 9/74 1272515 4/78 1284573 11/81 707933 3/71 1293857 10/74 1252667 5/7P 1321761 12/81 1095543 4/71 604048 11/74 1302968 5/;8 1100600 1/82 1248639 A-2
(
MONTHLY THERMAL GENERATION DURING THE flRST FIFTEEN IUEL CYCLES Of THE CONNECTICUT YANKEE REACTOR THERMAL THERMAL THERMAL THERMAL GENERATION GENERA 110N GENERATION GENERATION MONTH (MW _ttt)_ MONTH _mW-bri MONTH _(MW-brL MON 111 (MW-br) 2/82 887511 1/84 1353244 12/85 1244703 11/87 0 3/82 1327330 2/84 1264967 1/86 117556 12/87 0 4/82 1265930 3/84 1331459 2/86 0 1/88 0 5/82 1350787 4/84 1308920 3/86 0 2/88 0 6/82 1041904 5/84 1325577 4/86 0 3/88 82124 7/82 13?.7547 6/84 1185816 5/86 548577 4/88 1221711 8/82 1355453 7/84 1086170 6/86 823856 5/88 71941 9/82 916859 8/84 2590 7/86 445484 6/88 1279110 10/82 1355400 9/84 0 8/86 1084490 7/88 1355051 11/82 1083232 10/84 0 9/86 1234016 8/88 1348665 12/82 1355642 11/84 605570 10/86 1341900 9/88 1261130 1/83 921585 12/84 1350310 11/86 1288394 10/88 1346916 2/83 0 1/85 1354784 12/86 380276 11/88 1310047 3/83 0 2/85 1218633 1/87 1353623 12/88 1353862 4/83 734900 3/85 1152057 2/87 1225119 1/89 1351734 5/83 1342815 4/85 1284517 3/87 13330d9 2/89 1222835 6/83 1199531 5/85 1265872 4/87 10978'2 3/89 689533 7/83 1337082 6/85 1298329 5/87 13398il 4/89 1218778 8/83 1354702 7/85 1334672 6/87 12923:1 5/89 1345820 9/83 1303109 8/85 1235549 7/87 74'648 6/89 930971 10/83 1356900 9/85 1221235 8/87 0 7/89 1351483 11/83 1307356 10/85 1356063 9/87 0 8/89 1311603 12/83 1349090 11/85 962315 10/87 0 9/89 44000 A-3
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Westingtoane Electric Coricration ,
Advarced Friergy systerts - krm1 pics) Iworstar/
RtmDT Walt ) fill Sits T w eett1 042 OrigiMtore t'.Torck (d) tr.er7/ Center (170)
S.t..hde s son Ot) rr.er7j renter (470) Picened: 1/09/91 l't;crted: 2/1/91 (TRIES Cr 1!N.YSISj tcaiMtrys cor.nacticutt Yar.ket E-capoule tweeti Or191Mt4r t.ab. I4siteter (Dec.12,1990) 1D tacple a Material fuelide c*p @ g
- 2 mirra M137:t! in;":CRt, U-238 $1-245 U-230 ca-337 1.59D+03 +/- 1.0t401 11-237 91-144 1.p-227 - c .127 9.64R+03 +/- 4.f t+C1 WIRES '
Alco(Cd) $1-247 AICo do-(0 1.1DPa04 4/- 5.St+C1 Cu 91 248 C\t Co-40 2.100+02 */-
- 6te00 Cu 91-249 cu co-60 2.0 R+02 +/- 1.90400 Cu 91-250 Cu Co-40 2.24E4C2 4/- 1.7t4C0 t%30'N 91-211 Co-40 4. 64D+ C 2 ~ +/- B.0t400
%"t:K1AL $1-251 pn-54 1.1Et+C2 +/- 3.7t*00 91-251 Co-59 7.23E401 4/- 7.St400 Ibte: I6cntity aM locution of Wirta not availt.ble.
TisDion r.cnitors (braas) were twiled aM cutsMe surface was creded
- ~ ~ . - -
P a rke
- Results are iri units of dra/(rg of Iceiteter t'eterie.11.
M. File: 14242 Refe(cnocal Lab.BocAf 51 Noe 2 # f.
Prrrerbrest A-512, A-513, A-ife /.
AnalyetiW:f,17C Agprcued ' ',d#
A-6
_ E
Since surveillance Capsules H D, and E were irradiated for multiple fuel cycles, the flux adjustment factors, Cj, defined in Section 3.0 were employed in the reaction rate calculations for the individual sensor sets.
The quantity Cj is defined as the calculated ratio nf ( (E > 1.0 MeV) during irradiation period j to the time weighted average p (E > 1.0 MeV) over the entire irradiation period.
The values of C3 used in the evaluation of the Connecticut Yankee surveillance capsules were as follows*
Cj CAPSVLE H CAPS.t)]lj (APSULE E CYCLE 1 0.847 0.840 C.834 CYCLE 2 1.091 1.089 1. ,
CYCLE 3 1.077 1.076 1.062 CYCLL 4 1.082 1.080 1.066 CYCLE 5 0.942 0.941 0.928 CYCLE 6 1.043 1.042 1.028 CYCLE 7 1.060 1.058 1.044 CYCLE 8 0.951 0.938 CYCLE 9 1.055 1.041 CYCLE 10 1.006 0.993 CYCLE 11 1.000 CYCLE 12 1.003 CYCLE 13 1.040 CYCLE 14 1.050 CYCLE 15 1.040 A-7
APPitiDIX D MEASUR[0 SPEClfl0 AC11V11Y AtlD IRRADIA110!1 HISTORY Of R[ACT CAVITY EtiSOR SE15 in this appendix, the irradiation history as extracted from tiUREG-0020 and the measured specific activities of radiometric sensors irradiated in the reactor cavity during Cycle 16 are provided.
The irradiation history of Cycle 16 was as follows:
llCLL.ll CYCLE STAR 1VP 08/30/90 CYCLE SHU1DOWN 10/17/91 REf. 00RL POWER 1825 MWL 1HERMAL GENERAT10t1 M0ti1H _(liW-Ar)_
8/90 79812 9/90 631758 10/90 1163430 11/90 591778 12/90 1338960 1/91 1352288 2/91 1220284 3/91 724179 4/91 1307387 5/91 1349367 6/91 1204636 7/91 1354411 8/91 1329182 9/91 1306314 10/91 721859 TOTAL 15675645 B-1 r _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
The irradiation capsule loading diagram and the measured specific activities of the radiometric monitors from the Cycle 16 irradiation are provided on pages B-3 through B-12. For the multiple foil sensor sets, the individual foil 10 can be correlated with the capsule loading diagrt.ms provided in Section 6.1.1 in order to determine the location of the foil within the reactor cavity during irradiation.
B-2
CONTENTS OF HULTIPLE F0ll SENSOR SETS CYCLE 16 1RRADIAT10N CAPSULE ID BARE OR RADIOMETRIC MONITOR ID and CADMlUM SSTR POSITION- .jilllLDID._ [1 tii G I.1 @ [.0 V-223 PACKAG1 A-1 B BA -- -- -- -- AA --
W15-5B A-2 Cd A AA K AA AA A A --
A-3 Cd -- -- -- -- -- -- --
W15-5C B-1 B BB -- -- -- --
AB --
W15-1B B-2 Cd B AB L AB AB B B --
B-3 Cd -- -- -- -- -- -- --
WIS-lC C-1 B BC -- -- -- -- AC --
W15-68 C-2 Cd C AC H AC AC C C --
C-3 Cd -- -- -- -- -- -- --
W15-6C D-1 B BD -- -- -- -- AD --
W15-2B D-2 Cd D AD N AD AD D D --
D-3 Cd -- -- -- -- -- -- --
W15-2C E-1 B BE -- -- -- --
AE --
W15-3B E-2 Cd E AE 0 AE AE E E --
E-3 Cd -- -- -- -- -- -- --
W15-3C F-1 B BF -- -- -- --
AF --
W15-4B F-2 Cd F AF P AF AF F V --
F-3 Cd -- -- -- -- -- -- --
W15-4C B-3
l Weetiintruse Mvarced Programs l Pnoft i Aralytical Latoratory - Waltz Mill Site l Recpestl 14520 Originator: S. L. Arderson Padiation Dqineering & Analysis Peoeived: 12/10/91 Westirntouse Electric Corporation Poported: 4/10/92
[PISit3 (F APPLYSIS]
CCtHITICUT YAtTJ:E Fr>CIUt CAVITt DOSIMm7 Lab Dositeter (0 2/19/92)
Poil ID Sarplet Material tuclide dps/trg 2 sigma EA 91-2071 re Mn-54 4.88E+00 +/- 8.0E-02 A 91-2072 re Mn-54 5.42E+00 +/- 7.6E-02 BB 91-2080 re Mn-54 1.67E+01 +/- 1.6E-01 D 91-2081 re Mn-54 }.62E+01 +/- 1.3E-01 IC 91-2009 re Mn-54 6.73E+00 +/- 1.1E-01 C 91-2090 re Pti-54 7.28E+00 +/- 9.0E-02 DD 91-2098 re Mn-54 1.63E+01 +/- 1.6E-01 D 91-2099 re Mn-54 1.61E+01 +/- 1.4E41 BE 91-2107 re Mn-54 6.72E+00 +/- 1.0E-01 E 91-2108 Fe Mn-54 6.64E+00 +/- 8.9E-02 BF 91-2116 re Mn-54 6.54E@ 0 +/- 9.8E-02 P 91-2117 re Mn-54 6.39E+00 +/- 8.4E-02 AA 91-2073 Ni Co-58 6.55EM 1 +/- 9.4E-01 AB 91-2082 Ni Co-58 1.87E+02 +/- 1.6E+00 AC 91-2091 Ni Co-58 8.57E+01 +/- 1.1E+00 AD 91-2100 Ni Co-58 1.81E+02 +/- 1.6E+00 AE 91-2109 Ni Co-58 7.35E+01 +/- 1.0E+00 Ar 91-2118 Ni co-58 7.03E+01 +/- 9.7E-01 4
AA 91-2077 A1Co Co-60 2.13E+02 +/- 2.0E+00 A 91-2078 A1Co Co-60 1.76E+02 +/- 1.9E+00 AB 91-2086 A1Co Co-60 1.20E+03 +/- 8.0E+00 B 91-2087 A1Co Co-60 5.37t+02+/- 5.4E+00 AC 91-2095 A1Co Co-60 5.97Ed 2 +/- 4.5E+00 C 91-2096 A1Co co-60 2.44E+02 +/- 2.2E+00 AD 91-2104 A1Co co-60 1.16EM3 +/- 7.7E+00 D 91-2105 A1Co fo-60 5.07E+02 +/- 3.2E+00 AE 91-2113 A1Co Co-60 5.60E+02 +/- 2.8E+00 E 91-2114 A1Co C&-60 2.33E+02 +/- 1.9E+00 Ar 91-2122 A1Co Co-60 1.43EM3 +/- 8.7E+00 r 91-2123 A1Co Co-60 5.92E+02 +/- 5.6E+00 Pesarks:
AL File: 14520
References:
Lab Bookf 49 pages 267-268.
Proce&res: A-524.
Analyst: WIF, ME , ITC A @ rove $
( '
4y g I
B-4
l Westirghouse Advanced Programs l REKRT l Analytical Laboratory - Waltz Mill Site l Pequesti 14520 Originator S. L. Arvierson Radiation Engineering & Analysis Received: 12/18/91 Westirghouse Electric Corporation reported: 4/10/92 (RESUL'Is T A!ALYSIS1 cottarrIcirr YArr.cE mcim CAvnv tcmMEm Lab Dositeter (f 2/19/92)
Foil ID Sarplet Material ILiclide 4*hg 2 sigma A 91-2079 0-238 Ca-137 6.13E-01 +/- 2.93C-02 D 91-2088 U-238 Cs-137 1.84E+00 +/- 8.24E-02 C 91-2097 U-238 Ca-137 6.79E-01 +/- 3.23E-02 D 91-2106 D-238 Ca-137 1.80E+00 +/- 6.70E-02 E 91-2115 0-238 Ca-137 7.59E-01 +/- 5.91E-02 V 91-2124 0-238 Ca-137 5.85E-01 +/- 2.36E-01 A 91-2079 0-238 Ru-103 3.01E+00 +/- 5.00E-02 B 91-2088 D-238 Ru-103 8.39E+00 +/- 1.33C-01 C 91-2097 0-238 Ru-103 4.45E+00 +/- 5.75E-02 D 91-2106 D-238 Ru-103 8.26E+00 +/- 1.41E-01 E 91-2115 0-238 Ru-103 3.06E+00 +/- 7.83E-02 V 91-2124 D-238 Ru-103 2.82E+00 +/- 4.44E-02 A 91-2079 0-238 tr-95 6.00E+00 +/- 6.72E-02 B 91-2088 D-238 2r-95 1.81E+01 +/- 1.81E-01 C 91-2097 D-238 2r-95 7.07E+00 +/- 7.49E-02 D 91-2106 0-238 tr-95 1.79E+01 +/- 1.64E-01 E 91-2115 U-238 tr-95 7.02E400 +/- 1.37E-01 V 91-2124 D-238 tr-95 5.88E+00 +/- 5.76E-02 AA 91-2075 Ti Sc-46 1.23E+00 +/- 1.61E-02 AD 91-2084 Ti Tc-46 3.76E+00 +/- 2.82E-02 AC 91-2093 T1 Sc-46 1.67E+00 +/- 1.89PA 2 AD 91-2102 Ti Sc-46 3.69E+00 +/- 2.80E-02 AE 91-2111 Ti Sc-46 1.72E+00 +/- 1.92E-02 AF 91-2120 T1 Sc-46 1.63E+00 +/- 1.89E-02 K 91-2074 Cu Co-60 1.80E-01 +/- 4.32E-03 L 91-2083 Cu Co-60 5.62E-01 +/- 5.81L-03 M 91-2092 Cu Co-60 2.45E-01 +/- 4.60E-03 N 91-2101 Cu Co-60 5.51E-01 +/- 7.73E-03 0 91-2110 Cu co-60 2.65E-01 +/- 5.33E-03 P 91-2119 Cu Co-60 2.61E-01 +/- 4.12E-03 Remarks:
AL rile: 14520
References:
Lab Bookf49 pages 267-268. -
Proceiltes: A-524. i Analyst: WIF, MRK, ITC Approved:
B-5
e l Westinghouse Aovanced Energy $ystems REPORT l Analytical Laboratory Waltz Mill $tte Requestf 14520 Originator: $. L. Anderson Radiation Engineering & Analysil Received: 12/18/91 Westinghouse Electric Corporation Reported: 3/4/92 (RESULT 50FANALY$15)
CONNETICUT YANKEE REACTOR CAVITY D0$1 METRY 8ead Chain Tag 10: O deg., $.10 reet [< ........-.... dps/mg of chain 9 2/19/92 ..............>)
from Lab ....... Mn.$4 ....... .......Co58....... ....... Co.60 .......
Midplane Samplef dps/mg 2 sigma dps/mg 2 sigma dps/mg 2 sigma
+6.0 91 2064A 1.0$E+00 +/. 1.$E 01 1.90E+00 +/ 1.9E.01 4.13E+01 +/ 2.7E-01
+5.0 91 2064B*
+4.0 91 20640 7.99E400 +/. 6.0E.01 1.26E+01 +/. 6.3E 01 1.21E+02 +/- B.=E 01
+3.0 91 20640 1.05E+01 +/. 7.0E.01 1.66E+01 +/ 7.1E.01 1.58E402 +/. 9.6E.01
+2.0 91 2064E 1.07E+01 +/. 7.2E.01 1 75E+0! +/ 8.2E.01 1.92E+02 +/. 1.lE+00
+1.0 91 2064F 1.11E+01 +/. 1.0E+00 1.76E+01 +/. 1.2[+00 2.51E+02 +/ 1.$E400
+0.0 91 2064G*
.l.0 91 2064H 1.07E+01 +/. 1.8E+00 1.69E+01 +/. 1.8E+00 4.50E+02 +/. 2.82+00 2.0 91 20641 1.06E+01 +/ 1.6t+00 1.77E+01 +/ 2.0E+00 4.3BE+02 +/ 2.8E+00 3.0 91 2064J 9.73E+00 +/. 1.5E400 1.72E+01 +/ 1.8E400 3.90E+02 +/ 2.6E+00 4.0 91 2064K 1.53E+00 +/ 9.9E.01 1.32E+01 +/- 1.![+00 3.00E+02 */. 1.6E+00 5.0 91 2064L*
6.0 91 2064M 8.80E.01 +/. 3.3E.01 1.84E+00 +/. 3.7E.01 9.51E+01 +/. 7.4E.01 Remarks:
- Location of dosimetry block Al File: 14520 / -
References:
Lab Bookf49 pages 207 215 Lab Bookf51 pages 31 ,
/
Procedures: A 524, 6 [(/(f/j)h/"
' , s.
Analyst: WTF TRK, MRK, FRC Approved: __. /
/
B-6
____ - . _ _ _ _ _ _ _ _ . _ _ . _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ' " - - - - - - - " - - - - - - - - - - - - - - ~ - - - - - - - - - ' - - - - - - - - -
l =
i Westin house Advanced Energy Systems l REPORT l Analyt cal Laboratory . Waltz Mill Site l Requesti 14520 Originator: 5. L. Anderson Radiation Engineering & Analysis Received: 12/18/91 Vestinghouse Electric Corporation Reported: 3/4/92
[ RESULT 5 0F ANALY515)
CONNET! CUT YANKEE REACTOR CAVITY 0051 METRY Bead Chain Tag 10: 7.5 degrees 5115 Feet (< ............. dps/mg of chain 0 2/19/92 .........-...>]
from Lab ....... Mn.54 ....... ....... Co.58 ....... ....... Co.60 .......
Midplane Samplef dps/mg 2 sigma dps/mg 2 sigma dps/mg 2 sigma
+6.0 91 2065A 1.15E+00 +/ 1.5E.01 1.81E+00 +/. 1.6E 01 4.36E+01 +/. 2.8E.01
+5.0 91 20658 3.54E+00 +/. 2.6E.01 6.33E+00 +/. 2.6E.01 8.01E+00 +/. 3.9E.01
+4.0 91 2065C 7.21E+00 +/. 5.7E.01 1.22E+01 +/ 6.BE 01 1.27E+02 +/. 8.6E.0)
+3.0 91 20650 9.56E+00 +/. 6.6E.01 1.60E+0! +/. 7.8E.01 1.66t+02 +/ 9.9E.01
+2.0 91 2065E 1.03E401 +/. 7.3E 01 1.76E401 +/. 8.0E 01 1.96E+02 +/ 1.lE+00
+1.0 91 2065F 1.10E+01 +/. 1.1E+00 1,67E+01 +/.1.0E+00 2.52E+02 +/. 1.5E+00
+0.0 91 2065G*
.l.0 91 2065H 8.84E+00 +/. 1.6E+00 1.54E+01 +/ 1.7E+00 4.40E+02 +/. 2.BE+00
-2.0 91 20651 9.45E400 +/.1.4E400 1.71E+0) +/. 1.7E+00 4.26E+02 +/. 2.7t+00 3.0 91 2065J B.42E+00 +/ 1.EE.00 1.50E+01 +/. 1.7E+00 3.81E+ 02 +/ 2.6E+00 4.0 91 2065K 7.23E+00 +/. 9.6E+01 1.04E+01 +/. 9.5E.01 2.89E+02 +/. 1.6E+00 5.0 91 2065L 4.02E+00 +/. 5.9E.01 6.58E+00 +/. 6.4E.01 1.03E+02 +/ 1.0E+00 6.0 91 2065M 7.39E.01 +/. 3.0E.01 1.38E+00 +/. 4.2E.01 9.17E+01 +/ 7.3E 01 Remarks:
- Location of dosimetry block Al File: 14520 - 1
References:
Lab Bookf49 pages 207 215; Lab Bookf51 pages 31 [
Procedures: A.524. ,
/v/h'I I y Analyst: Wif, TRK MRK, FRC Approved: I
/
B-7 t _ _____-_-____- _ __ - - - - - -- _ - --- _ - - - - - -- - -- -
+*
Westinghouse tdvanced Ener y Systems 1 REPORT Analytical laboratory . Wa tz Hill Site l Requesti 14520 Originator: 5. L. Anderson Radiation Engineering & Analysis Received: 12/18/91 Vestinghouse Flectric Corporation Peported: 3/4/92
[RESULTSOFANALY$15)
CONNETICUT YANrEE REACTOR CAVITY 0051 METRY Bead Cha*4 Tag 10: 37.5 degreest 5 1 30 Feet from Lab
\<..............
....... Hn.54 .......
dps/mg of chain t t/19/92 ............+.>)
...... Co 58 ....... ....... to.60 .......
Hidplane Sample # dps/mg 2 sigma dp:/mg 2 sigma dps/mg 2 sigma
+6.0 91 2066A 5.00E.01 +/ 9.9E 02 9.71E 01 t/. 1.1E 01 2.29E+01 +/. 2.0E 01
+5.0 91 20668 1.55E+00 +/. 1.6E 01 2.50E+00 +/. 1.7E.01 3.88t+01 +/. 2.7E 01
+4.0 91 2066C 3.06E+00 +/ 2.3E 01 4.58t+00 +/. 2.3E.01 5.80E+01 +/ 3.3E.01
+3.0 91 20660 3.92E400 +/. 2.4E.01 6.36E+00 +/. 2.6E 01 7.30E+01 +/- 3.7E.01
+2.0 91-2066E 4.58E+00 +/ 3.BE 01 6.69E+00 +/. 3.8E.01 8.70E+01 +/ 5.lt.01
+1.0 91 2066F 4.67E+00 +/. 5.4E.01 7.16E+00 +/. 5.7E.01 1.15E+02 +/. 8.2E 01
+0.0 91 2066G*
.l.0 91 2066H 3.33E+00 +/. 5.3E.01 6.50E+00 +/ 6.6E.01 1.97E402 +/. 1.1[+00 2.0 91 20061 4.24E+00 +/. 6.8E.01 6.64E+00 +/ 6.8E.01 1.BBE+02 +/. 1.1E+00 3.0 91 2066J 4.21E+00 +/- 6.1E.01 6.48t+00 +/. 6.5E.01 1.64E+02 +/. 9.9E.01 4.0 91 2006K 2.79E+00 +/. 3.4E.01 4.70E+00 +/ 3.8E 01 1.26E+02 +/ 6.0E.01
.$.0 91 2066L 1.34E+00 +/. 3.0E.01 2.30E+00 +/. 3.0E.01 8.23E+01 +/. 4.9E 01 6.0 91-2066M 4.54E 01 +/. 1.2E 01 7.66E.01 +/. 1.2E.01 4.56E+01 +/- 2.5E 01 Remarks:
- Location of fiosimetry block AL File: 14520
References:
Lab Bookf49 pages 207 215; Lab Bookf51 pages 31.
Procedures: A.524
, Analyst: WTF, TRK, HRK, FRC Approved:
g g[
g fp 4 &N t
/
B-8
Westinghouse Advanced inergy Systems ,
REPORT Analytical Laboratory . Waltz Hill Site Requesti 14520
............................................... 1 Oricinator: S. L. Anderson Radiation Engineering & Analysis Westinghouse Electric Corporation Received: 12/1B/91
....................................................................... Reported: 3/4/92'
[RESULTS OF ANALYS15]
CONNETICUT YANKEE REACTOR CAVITY 00SIMETRY Bead Chain Tag 10: 45.0 degrees; S.1 45 feet [< - -. . . .-
dps/mg of chain 9.2/19/92 - '
from Lab Midplane Samplel
.--.. Hn.54 .......
dps/mg 2 sigma
....... Co 58 . - ..
dps/mg 2 sigma
.....- Co 60 .......)
......... .......... dps/mg 2 sigma
+6.0 91 2067A 6 lBE 0! +/ 7.7E 02
+5.0 91 20678 8.99E 01 +/ 7.3E 02 2.28E+01 +/. 1.3E 01 1.56E+00 +/ 1.5E 01 2.50E+00 +/ 1.5E.01 3.72E+01 +/- 2.6E 01
+4.0 91 2067C 2.84E+00 +/ 2.lE 01
+3.0 91 2067D 4.63E+00 +/. 2.3E 01 5,50E+01 +/- 3.2E 01 3.70E+00 +/- 2.2E 01 6.10 +00 +/. 2.7E-01 5.72E+0! +/ 3.5E 01
+2.0 91-2067E 4.41E+00 +/- 3.6E.01
+1.0 91-2067F 6.44E+00 +/ 3.3E 01 7.62E+01 +/- 4.7E.01 4.13E400 +/ 2.2E.01 3.77E+00 +/- 3.lE-01 1.08E+02 +/ 4.0E 01
+0.0 91 2067G*
1.0 91 2067H 4.29E+00 +/ 4.5E 01
-2.0 91 20671 6.07E+00 +/ 4.3E 01 1.88E+02 +/ 7.4E 01 4.21L+00 +/. 4.7E-01 5.99E+00 +/ 4.3E 01 1.75E+02 +/ 7.2E 01 3.0 91 2067.) 3.50E+00 ,/ 3.9E 01 4.0 91 2067K 5.46E+00 +/. 4.0E 01 1.54E+02 +/ 6.7E 01 2.69E+00 +/ 3.1E 01 4.33E+00 +/ 3.5E 01 1.19E+02 +/- 5.9E Ol 5.0 91-2067L 1.28E+00 +/ 1.4E 01
-6.0 91 2067H 2.17E+00 +/- 1.5E 01 7.67E+01 +/ '2.7E 01 4.33E-01 +/ 4.9E.02 6.63E-01 +/. 5.2E.02 4.27E+0! +/ 9.9E 02 Remarks:
- Location of dosimetry bMk Al File: 14520 5
References:
Lab Bookf49 pages 207 215, Lab Bookl51 pages 31 r / -
Procedures: A 524. i Analyst: WTF TRK, MRK, FRC Approved: l i f(, tf/[ 1,4 s I
B-9
+*
l Westinghouse Advanced En6r y Systems RE PORT l Analytical Laboratory Wa tz Mill Site Requesti 14520 Originator: S. L. Anderson Radiation Engineering & Analysis Received: '?/.8/91 Westinghouse Electric Corporation Reported- '
92
[RESULTSOFANALYSIS)
CONNETICUT YANKEE RFACTOR CAVITY DOSIMETRY Bead Chain Tag 10: 90 degrees; S 1 90 feet [< -...-- .. dps/mg of chain 9 2/19/92 ---
>]
from Lab ....... Mn.54 .. -=.- .... Co 58 - - -- - Co 60 - -
Midplane Samplef dps/mg 2 sigma dps/mg 2 sigma dps/mg 2 sigma ,
+6.0 91 2068A 1.01E+00 +/ 1.lE-01 1.80E400 +/. 1.2E-01 4.06E+0! +/- 1.9E.01
+5.0 91 2068B 3.69E+00 +/- 2.4E-01 6.46E+00 +/. 3.0E-01 7.73E+01 +/- 3.8E 01
+4.0 91-2008C 8.08E+00 +/- 5.9E.01 1.26E+01 +/- 6.5E-01 1.23E+0? +/. 8.6E.0k
+3.0 91 20683 1.03E+01 +/ 7.0E.01 1.64E+f. t' 7.9E-01 1.60E+02 +/- 9.7E 01
+2.0 91 206BE 1.09E+01 +/- 6.7E+00 1.77E+o? r/- 7.8E-01 7.94E+02 +/- 1.lE+00
+1.0 91 2068F 1.18E+0) +/- 1.0E+00 1.87E+() + '- 1.lE+00 2.48E+02 +/- 1.5E+00
+0.0 91 2068G 1.06E+0! +/- 1.6E+00 1.65E+U >/- 2.lE+00 4.01E+02 +/- 2.1E+00
-1.0 91 2068H 1.12E401 +/- 1.6E+00 1.69E+01 +/. 1.8E+00 4.63E+02 +/- 2.9E+00 2.0 91 20681 9.34E+00 +/ 1.5E+00 1.84E+01 +/ 1.9E+00 4.49E+02 +/. 2.8E,00
-3.0 91 2068J 1.08E+01 +/- 1.5E+00 1.58E+01 +/- 1.6E+00 4.05E+02 +/ 2.7E+00
-4.0 91-2008K 7.49E+00 +/- 9.9E.01 1.33E+01 +/- 1.lE+00 3.15E+02 +/ 1.7E+00 5.0 91-2068L 4.17E+00 +/- 6.7E-01 6.91E+00 +/ 6.6E-01 2.01E+02 +/- 1.lE+00
-6.0 91 2068M 1.05E+00 +/- 3.9E 01 2.02E+00 +/- 4.2E 01 9.87E+01 +/. 7.6E 01 Remarks:
- Location of dosimetry block AL File: 14520
References:
Lab Bookf49 pages 207 215; Lab Bookf51 pages 31
=
[
Procedures: A-524. q i/j,lg( f ,
[t/F xt, 4-Analyst: WTF, TRK, MRK, FRC Approved: / 1 a
B-10 1
u +
Wett ingt.ouse Advanced Ener y Systems l REPORT Analytical Laboratory - Wa tr Mill Site Requesti 14520
.............................................l Originator: S. L. Anderson Radiation Engineering & Analysis Westinghouse Electric Corporation Received: 12/18/91
.............................................................eported:R 3/4/92
[RESULTS OF ANALYt!S}
CONNETICUT YANKEE REACTOR CAVITY DOSIMETRY Bead Chain Tag ID: 180 degrees; 5 1 180 Feet [<.........-.... dps/mg of chain 9 2/19/92 from Lab --- -. Mn 54 --- --
....>]
Midplane Sample # dps/rg 2 sigma
-....- Co-58 ....... ......- Co.60 . ....
dps/ng 2 sigma dps/mg 2 sigma
+6.0 91 2069A 1.11E+00 +/- 1.7E-01
+5.0 2.00E+00 +/- 2.0E-01 4.32E+01 +/ 2.8E 01 91-20698 3.77E+00 +/ 4.5E 01
+4.0 6.52E+00 +/ 4.7E.01 8.05E+01 +/ 6.9E.01 91 2069C 7.95E+00 +/- 6.3E-01 1.36E+01 +/- 7.2E 01
+3.0 91-20690 1.05E+01 +/- 6.3E 01 1.26E+02 +/ 8.6E-01
+2.0 1.78E+01 +/- 7.8E.01 1.64E+^2 +/- 9,9E.01 91 2069E 1.16E+01 +/- 5.7E.01 is 90E+01 +/. 6,5E.01
+1.0 91 2069F 2.00E+02 +/- 7.7E-01 1.18E+0) +/- 1.0E+00 1.84E+01 +/- 1.lE+00 2.65E+02 +/- 1.5E+00
+0.0 91 2069G 1.16E+01 +/- 1.6E,00 1.0 91-2069H 1.89E+01 +/ 1.9E+00 4.19E+02 +/- 2.7E+00 1.07E+01 +/ 1.7E+00 1.6BE401 +/- 1.8E+00 4.85E+02 +/- 2.9E+00 2.0 91-20691 1.27E+01 +/ 1.7E+00 3.0 2.05E+01 +/- 2.0E+00 4. 74E+02 +/- 2.9E+00 91-2069J 1.19E+01 +/ 1.8E+00 1.73E+01 +/- 1.9E+00
-4.0 91 2069K 4.31E402 +/- 2.8E400 7.58E+00 +/ 9.2E-01 1.41E+01 +/- 1.2[+00 3.34E+02 +/- 1.7E+00 5.0 91 2069L 4.42E+00 +/- 6.6E-01
-6.0 8.65E+00 +/- 8.9E-01 2.14E+02 +/- 1.lE+00 91-2069H 1.26E+00 +/- 4.5E.01 2.03E+00 +/- 4.8E.01 1.12E+02 +/ 8.lE 01 i
l b
l l
l Remarks:
- Location of dosimetry block Al Flie: 14520 j/
References:
Lab Bookf49 pages 207-215; Lab Bookf51 pages 31 Procedures: A 524
'/ /
Analyst: WTF, TRK, MRK, FRC Approved: [4 (/ 44e7 UA4[-
/
B-ll i
P
+ +
l Westingh use Advanced Ener y Systems REPORT l Analytic 1 Laboratory Wa 12 Mill Site Requesti 14520 Originator: S. L. Anderson Radiation Engineering & Analysis Received: 12/18/91 Westinghouse Electric Corporation Reported: 3/4/92
[RESULTS OF ANALYSIS)
CONNETICUT YANKEE REACTOR CAY!TY 00SIMETRY Bead Chain Tag 10: 270 degrees; S 1 270 feet [<.- . -........ dps/mg of chain 9 2/19/92 -.-....... - >) ~
from Esb ....... Kn-54 . -- - .
-. Co 58 .-- -
- . -.- Co-60 .......
Midplane Sample # dps/mg 2 sigma dps/mg 2 sigma dps/mg 2 sigma
+6.0 91 2070A 1.34E+00 +/. 1.8E-01 2.15E+00 +/- 1.8E-01 4.86E+01 +/. 3.0E.01
+5.0 91 20708 4.lBE+00 +/- 5.0E.01 7.20E+00 +/. 5.0E.01 8.93E+01 +/. 7.2E.01
+4.0 91 2070C 8.42E+00 +/- 6.3E.01 1.50E+01 +/ 7.0E.01 1.39E+0? +/. 9.0E.01
+3.0 91 20700 1.12E+01 +/- 7.4E-01 1.84E+01 +/. B.2E-01 1.78E+02 +/. 1.0E+00
+2.0 91 2070E 1.18E+0) +/ 7.9E 01 1.86E+01 +/- 8.6E.01 2.17E+02 +/ 1.lE+00
+1.0 51 2070F 1.21E+01 +/ 9.2E Gi 1.91E+01 +/- 1.lE+00 2.87E+02 +/ 1.6E+00
+0.0 91 2070G l-19E+01 +/- 1.7E+00 2.06E+0! +/. 2.3E+00 4.52E+02 +/- 2.8E+00
-1.0 9e2070H 1.22E+01 +/ 1.8t+00 1.92E+01 +/ 2.lE+00 5.08E+02 +/. 3.0E+00
-2.0 91 20701 1.10E+01 +/ 1.6E+00 1.96E+01 +/- 2.4E+00 4.93E+02 +/. 3.0E+00 3,0 91 2070J 1.16E+01 +/ 1.3E+00 1.78E+01 +/. 1.4E+00 4.44E+02 +/. 2.0E+00 4.0 91 2070K 8.59E+00 +/- 1.lE,00 1.47E+01 +/. 1.2E+00 3.40E+02 +/. 1.7E+00 5.0 91 2070L 3.94E+00 +/. 6.0E-01 7.49E+00 +/ 6.9E.01 2.18E+02 +/ 1.lE+00 6.0 91-2070M 9.49E-01 +/- 3.5E-01 2.12E+00 +/. 4.8E.01 1.16E+02 +/- 8.2E.01 4
Remarks:
- Location of dosimetry block Al File: 14520
References:
Lab Bookf49 pages 207 215; Lab Bookf51 pages 31 Procedures: A-524.
-Analyst: WTF, TRK, MRK, FRC Approved: 1g t
/
Af/$ fp/
/
B-12 l
_ __ -__ _- . - _ _ _ - - --