ML17309A606
| ML17309A606 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 06/30/1996 |
| From: | Boyle D, Perock J, Terek E WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML17264A598 | List: |
| References | |
| WCAP-14684, NUDOCS 9609180203 | |
| Download: ML17309A606 (176) | |
Text
WESTINGHOUSE NON-PROPRIETARY CLASS 3 WCAP-1 4684 R. E. Ginna Heatup and Cooldown Limit Curves For Normal Operation Ed Terek John D. Perock June 1996 Work Performed Under Shop Order RBXP-139 Prepared by the Westinghouse Electric Corporation for the Rochester Gas and Electric Corporation Approved:
J F
D. E. Boyle, Manag r
Reactor Equipment & Materials Engineering WESTINGHOUSE ELECTRIC CORPORATION Systems and Major Projects Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 1996 Westinghouse Electric Corporation All Rights Reserved
~ ~
PREFACE This report has been technically reviewed and verified by:
P. A. Grendys Sections 1, 2, 4, 5, 6, and 7 T. M. Lloyd Section 3
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R. E. Ginna Heatup and Cooidown Limit Curves June 1996
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TABLE OF CONTENTS LIST OF TABLES LIST OF FIGURES Vl INTRODUCTION PURPOSE RADIATIONANALYSISAND NEUTRON DOSIMETRY
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3.1 Introduction 3.2 Discrete Ordinates Analysis.........
3.3 Neutron Dosimetry 3A Projections of Reactor Vessel Exposure.......
4 7
CRITERIA FOR ALLOWABLEPRESSURE-TEMPERATURE RELATIONSHIPS 49
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CALCULATIONOF ADJUSTED REFERENCE TEMPERATURE............
52 HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMITCURVES....
63 REFERENCES.....
80 E. Ginna Heatup and Cooldown LimitCurves June 1996
LIST OF TABLES Calculated Fast Neutron Exposure Rates and Iron Atom Displacement Rates at the Surveillance Capsule Center...
L 16 3-2 Calculated Azimuthal Variation of Fast Neutron Exposure Rates and Iron Atom Displacement Rates at the Reactor Vessel Clad/Base Metal interface 19 3-3 Relative Radial Distribution of $(E > 1.0 MeV) within the Reactor Vessel Wall...
22 3-4 Relative Radial Distribution of $(E > 0.1 MeY) within the Reactor Vessel Wall...
23 3-'5 Relative Radial Distribution of dpa/sec within the Reactor Vessel Wall........
24 3-6
.Nuclear Parameters used in the Evaluation of Neutron Sensors.............
25 3-7 Monthly Thermal Generation During the First Twenty-Five Fuel Cycles of the Ginna Reactor 26 3-8 Measured Sensor Activities and Reaction Rates Surveillance Capsules V, R, T, and S Saturated Activities and Reaction Rates..
3-9 Summary of Neutron Dosimetry Results Surveillance Capsules V, R, T, and S..
37 3-10 Comparison of Measured and FERRET Calculated Reaction Rates at the Surveillance Capsule Center 38 3-11 Adjusted Neutron Energy Spectrum at the Center of Surveillance Capsules....
40 3-12 Comparison of Calculated and Measured Integrated Neutron Exposure of Gin'na Surveillance Capsules V, R, T, and S 3-13 Neutron Exposure Projections at Key Locations on the Reactor Vessel Clad/Base Metal Interface.. ~..........
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45 3-14 Neutron Exposure Values within the Ginna Reactor Vessel..............
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46 3-15 Updated Lead Factors for Ginna Surveillance Capsules.......
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48 5-1 Summary of the Peak Pressure Vessel Neutron Fluence Values used for the Calculation of ARTValues......
53 R. E. Ginna Heatup and Cooldown Limit Curves June 1996
LIST OF TABLES (Continued) 5-2 Measured Integrated Neutron Exposure of the R. E. Ginna Surveillance Capsules 53 5-3 Measured 30 ft-Ib Transition Temperature Shifts of the Beltline Materials Contained in the Surveillance Program 54 5-4 Reactor Vessel Beltline Material Unirrdiated Toughness Properties......
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55 5<A Compilation of Copper and Nickel Weight Percent Values for the Ginna Surveillance Program Weld Metal 55 5-5 Calculation of Chemistry Factors using Ginna Surveillance Capsule Data..
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56 5-6 Summary of the Ginna Reactor Vessel Beltline Material Chemistry Factors.....
57 5-7 Summary of the Calculated Fluence Factors Used for the Generation of the 24, 28, 32 and 40 EFPY Heatup and Cooldown Curves 57 5-8 Calculation of the ART Values for the 1/4T Location and 24 EFPY.....
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58 5-9 Calculation of the ART Values for the 3/4T Location and 24 EFPY......
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58 5-10 Calculation of the ART Values for the 1/4T Location and 28 EFPY........
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59 5-11 Calculation of the ART Values for the 3/4T Location and 28 EFPY..........
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59 5-12 Calculation of the ART Values for the 1/4T Location and 32 EFPY........
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60 5-13 Calculation of the ART Values for the 3/4T Location and 32 EFPY ~.........
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60 5-14 Calculation of the ART Values for the 1/4T Location and 40 EFPY.... ~......
61 5-15 Calculation of the ART Values for the 3/4T Location and 40 EFPY...........
61 5-16 Summary of the Limiting ART Values Used in the Generation of the Ginna Heatup/Cooldown Curves 62 6-1 R. E. Ginna 24 EFPY Heatup Curve Data Points 6-2 R. E. Ginna 24 EFPY Cooldown Curve Data Points.
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72 73 R. E. Ginna Heatup and Cooidown Limit Curves June 1996
LIST OF TABLES (Continued) 6-3 R. E. Ginna 28 EFPY Heatup Curve Data Points 74 6-4 R. E. Ginna 28 EFPY Cooldown Curve Data Points......................
75 6-5 R. E. Ginna 32 EFPY Heatup Curve Data Points 76 6-6 R. E. Ginna 32 EFPY Cooldown Curve Data Points..
77 6-7 R. E. Ginna 40 EFPY Heatup Curve Data Points.....
78 6-8 R. E. Ginna 40 EFPY Cooldown Curve Data Points..
79 R. E. Ginna Heatup and Cooldown Limit Curves June 1996
LIST OF FIGURES 0
3-1 Arrangement of Surveillance Capsules in the R. E. Ginna Reactor Vessel....
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14 3-2 Plan View of a Reactor Vessel Surveillance Capsule 15 6-1 R. E. Ginna Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100'F/hr) Applicable to 24 EFPY (Without Margins for Instrumentation Errors) 64 6-2 R. E. Ginna Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 20, 40, 60 and 100'F/hr) Applicable to 24 EFPY (Without Margins for Instrumentation Errors) 65 6-3 R. E. Ginna Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100'F/hr) Applicable to 28 EFPY (Without Margins for Instrumentation Errors) 66 6-4 R. E. Ginna Reactor Coolant System Cooldown Limitations (Coo!down Rates of 0, 20, 40, 60 and 100'F/hr) Applicable to 28 EFPY (Without Margins for Instrumentation Errors).
67 6-5 '. E. Ginna Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100'F/hr) Applicable to 32 EFPY (Without Margins for Instrumentation Errors) 68 6-6 R. E. Ginna Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 20, 40, 60 and 100'F/hr) Applicable to 32 EFPY (Without Margins for Instrumentation Errors)..
69 6-7 R. E. Ginna Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100'F/hr) Applicable to 40 EFPY (Without Margins for Instrumentation Errors) 70 6-8 R. E. Ginna Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 20, 40, 60 and 100'F/hr) Applicable to 40 EFPY (Without Margins for Instrumentation Errors) 71
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R. E. Ginna Heatup and Cooidown Limit Curves June 1996
INTRODUCTlON Heatup and cooldown limit curves are calculated using the adjusted RTNpT (reference nil-ductilitytemperature) corresponding to the limiting beltline region material of the reactor vessel.
The adjusted RT>>y of the limiting material in the core region of the reactor vessel is determined by using the unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced BRTNpT and adding a margin. The unirradiated RT>> is designated as the higher of either the drop weight nil-ductilitytransition temperature (NDTT) or the temperature at which the material exhibits at feast 50 ft-Ib of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60 F.
'T>>
increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RTNp7 at any time period in the reactor's life, b,RT>> due to the radiation exposure associated with that time period must be added to the unirradiated RT>>(IRT>>).
The extent of the shift in RT>> is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels.
The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials""'. Regulatory Guide 1.99, Revision 2, is used for the calculation of Adjusted Reference Temperature (ART) values (lRT>>~+
b,RT>>~ + margins for uncertainties) at the 1/4T and 3/4T locations, where T is the thickness of the vessel at the beltline region measured from the clad/base metal interface.
The most limiting ART values are used in the generation of heatup and cooldown pressure-temperature limitcurves.
R. E. Ginna Heatup and Cooldown Limit Curves June 1996
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PURPOSE The Rochester Gas and Electric Corporation contracted Westinghouse to re-analyze the vessel fluences using the latest approved methodology contained in WCAP-14040-NP-A, Revision 2"',and to generate new heatup and cooldown curves for 24, 28, 32 and 40 EFPY.
The heatup and cooldown curves were generated without margins for instrumentation errors.
The curves include a hydrostatic leak test limitcurve from 2485 psig to 1500 psig and include pressure-temperature limits for the vessel flange regions per the requirements of 10 CFR Part 50, Appendix G'".
The purpose of this report is to present the calculations and the development of the R. E.
Ginna heatup and cooldown curves for 24, 28, 32 and 40 EFPY. This report documents the neutron fluence evaluation, the calculated adjusted reference temperature (ART) values following the methods of Regulatory Guide 1.99, Revision 2,"'or all beltline materials and the development of the heatup and cooldown pressure-temperature limitcurves for normal operation.
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R. E. Ginna Heatup and Cooldown Limit Curves June 1996
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3 RADIATIONANALYSISAND NEUTRON DOSIMETRY
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3.1 Introduction Knowledge of the neutron environment within the reactor vessel and surveillance capsule geometry is required as an integral part of LWR reactor vessel surveillance programs for two reasons.
First, in order to interpret the neutron radiation induced material property changes observed in the test specimens, the neutron environment (energy spectrum, flux, fluence) to which the test'specimens were exposed must be known. Second, in order to relate the changes observed in the test specimens to the present and future condition of the reactor vessel, a relationship must be established between the neutron environment at various positions within the reactor vessel and that experienced by the test specimens.
The former requirement is normally met by employing a combination of rigorous analytical techniques and measurements obtained with passive neutron flux monitors contained in each of the surveillance capsules.
The tatter information is generally derived solely from analysis.
The use of fast neutron fluence (E > 1.0 MeV) to correlate measured material property changes to the neutron exposure of the material has traditionally been accepted for development of damage trend curves as well as for the implementation of trend curve data to assess vessel condition.
In recent years, however, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and positions within the vessel wall could lead to an improvement in the uncertainties associated with damage trend curves as well as to a more accurate evaluation of damage gradients through the reactor vessel wall.
Because of this potential shift away from a threshold fluence toward an energy dependent damage function for data correlation, ASTM Standard Practice E853, "Analysis and Interpretation of Light Water Reactor Surveillance Results," recommends reporting displacements per iron atom (dpa) along with fluence (E > 1.0 MeV) to provide a data base for future reference.
The energy dependent dpa function to be used for this evaluation is specified in ASTM Standard Practice E693, "Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom." The application of the dpa parameter to the assessment of embrittlement gradients through the thickness of the reactor vessel wall has already been promulgated in Revision 2 to Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials."
This section provides the results of the neutron dosimetry evaluations performed in conjunction with the analysis of test specimens contained in surveillance Capsules V, R, T, and S, withdrawn at the end of the first, third, ninth, and twenty-second fuel cycles, respectively.
This update. is based on current state-of-the-art methodology and nuclear data including recently released neutron transport and dosimetry cross-section libraries derived from the ENDF/B-VI data base.
This report provides a consistent up-to-date neutron exposure data base for use in evaluating the material properties of the Ginna reactor vessel.
R. E. Ginna Heatup and Cooldown Limit Curves June 1996
In each of the capsule dosimetry evaluations, fast neutron exposure parameters in terms of neutron fluence (E > 1.0 MeV), neutron fluence (E > 0.1 MeV), and iron atom displacements (dpa) are established for the capsule irradiation history. The analytical formalism relating the measured capsule exposure to the exposure of the vessel wall is described and used to project the integrated exposure of the vessel wall.
Also, uncertainties associated with the derived exposure parameters at the surveillance capsules and with the projected exposure of the reactor vessel are provided.
3.2 Discrete Ordinates Analysis A plan view of the reactor geometry at the core midplane is shown in Figure 3-1. Six irradiation capsules attached to the thermal shield are included in the reactor design to constitute the reactor vessel surveillance program.
The capsules are located at azimuthal angles of 57', 67', 77, 237', 247', and 257'elative to the core cardinal axis as shown in Figure 3-1. A plan view of a surveillance capsule holder attached to the thermal shield is shown in Figure 3-2. The stainless steel specimen containers are approximately 1-inch square and approximately 38 inches in height.
The containers are positioned axially such that the test specimens are centered on the core midplane, thus spanning the central three feet of the 12 foot (141.4 in) high reactor core.
From a neutronic standpoint, the surveillance capsules and associated support structures are significant. The presence of these materials has a marked effect on both the spatial distribution of neutron flux and the neutron energy spectrum in the water annulus between the thermal shield and the reactor vessel.
In order to determine the neutron environment at the test specimen location, the capsules themselves must be included in the analytical model.
In performing the fast neutron exposure evaluations for the surveillance capsules and reactor vessel, two distinct sets of transport calculations were carried out. The first, a single computation in the conventional forward mode, was used primarily to obtain relative neutron energy distributions throughout the reactor geometry as well as to establish relative radial distributions of exposure parameters
($(E > 1.0 MeV), g(E > 0.1 MeV), and dpa/sec) through the vessel wall. The neutron spectral information was required for the interpretation of neutron dosimetry withdrawn from the surveillance capsules as well as for the determination of exposure parameter ratios; i.e., [dpa/sec]/[f(E > 1.0 MeV)], within the reactor vessel geometry.
The relative radial gradient information was required to permit the projection of measured exposure parameters to locations interior to the reactor vessel wall, i.e., the ~/4T and %T locations.
The second set of calculations consisted of a series of adjoint analyses relating the fast neutron flux, g(E > 1.0 MeV), at surveillance capsule positions and at several azimuthal locations on the reactor vessel inner radius to neutron source distributions within the reactor core. The source importance functions generated from these adjoint analyses provided the basis for all absolute exposure calculations and comparison with measurement.
These R. E. Ginna Heatup and Cooldown Limit Curves June 1996
importance functions, when combined with fuel cycle specific neutron source distributions, yielded absolute predictions of neutron exposure at the locations of interest for each cycle of irradiation. They also established the means to perform similar predictions and dosimetry evaluations for all subsequent fuel cycles.
It is important to note that the cycle specific neutron source distributions utilized in these analyses included not only spatial variations of fission rates within the reactor core but also accounted for the effects of varying.neutron yield per fission and fission spectrum introduced by the build-up of plutonium as the bumup of individual fuel assemblies increased.
The absolute cycle specific data from the adjoint evaluations together with the relative neutron energy spectra and radial distribution information from the reference forward calculation provided the means to:
1 -
Evaluate neutron dosimetry obtained from surveillance capsules, 2 -
Relate dosimetry results to key locations at the inner radius and through the thickness of the reactor vessel wall, 3 -
Enable a direct comparison of analytical prediction with measurement, and 4 -
Establish a mechanism for projection of reactor vessel exposure as the design of each new fuel cycle evolves.
The forward transport calculation for the reactor model summarized in Figures 3-1 and 3-2 was carried out in R,G geometry using the DORT two-dimensional discrete ordinates code Version 2.7.3'4'nd the BUGLE-93 cross-section library. The BUGLE-93 library is a 47 energy group ENDF/B-Vl based data set produced specifically for light water reactor applications.
In these analyses anisotropic scattering was treated with a P, expansion of the scattering cross-sections and the angular, discretization was modeled with an S, order of angular quadrature.
The core power distribution utilized in the reference forward transport calculation was derived from statistical studies of long-term operation of Westinghouse 2-loop plants.
Inherent in the development of this reference core power distribution is the use of an out-in fuel management strategy; i.e., fresh fuel on the core periphery.
Furthermore, for the peripheral fuel assemblies, the neutron source was increased by a 2'argin derived from the statistical evaluation of plant to plant and cycle to cycle variations in peripheral power.
Since it is unlikely that any single reactor would exhibit power levels on the core periphery at the nominal
+ 2a value for a large number of fuel cycles, the use of this reference distribution is expected to yield somewhat conservative results.
All adjoint calculations were also carried out using an S, order of angular quadrature and the P, cross-section approximation from the BUGLE-93 library. Adjoint source locations were chosen at several azimuthal locations along the reactor vessel inner radius as well as at the geometric center of each surveillance capsule.
Again, these calculations were run in R,G R. E. Ginna Heatup and Cooldown Limit Curves June 1996
geometry to provide neutron source distribution importance functions for the exposure parameter of interest, in this case f(E > 1.0 MeV).
L Having the importance functions and appropriate core source distributions, the response'of interest could be calculated as:
R(1;e)
= fff IirB,E) S(r8,E) rd'r c8 dE r
0 E
where:
R(r,e) =
l(r,e,E)=
S(r,e E)=
$(E > 1.0 MeV) at radius r and azimuthal angle e.
Adjoint source importance function at radius r, azimuthal angle e, and neutron source energy E.
Neutron source strength at core location r,e and energy E.
Although the adjoint importance functions used in this analysis were based on a response function defined by the threshold neutron flux $(E > 1.0 MeV), prior calculations"'ave shown that, while the implementation of low leakage loading patterns significantly impacts both the magnitude and spatial distribution of the neutron field, changes in the relative neutron energy spectrum are of second order. Thus, for a given location the ratio of [dpa/sec]/[$ (E > 1.0 MeV)] is insensitive to changing core source distributions.
In the application of these adjoint importance functions to the Ginna reactor, therefore, the iron atom displacement rates (dpa/sec) and the neutron flux $(E > 0.1 MeV) were computed on a cycle specific basis by using [dpa/sec]/[f(E > 1.0 MeV)] and [g(E > 0.1 MeV)]/[g(E> 1.0 MeV)] ratios from the forward analysis in conjunction with the cycle specific f(E > 1.0 MeV) solutions from the individual adjoint evaluations.
The reactor core power distributions used in the plant specific adjoint calculations were taken from the fuel cycle design reports for the first twenty-five operating cycles and the upcoming twenty-sixth cycle of Ginna ~~~ ~'.
Selected results from the neutron transport analyses are provided in Tables 3-1 through 3-5.
The data listed. in these tables establish the. means for absolute comparisons of analysis and measurement for the capsule irradiation periods and provide the means to correlate dosimetry results with the corresponding exposure of the reactor vessel wall.
In Table 3-1, the calculated exposure parameters [)(E > 1.0 MeV), $(E > 0.1 MeV), and dpa/sec] are given at the geometric center of the three azimuthally symmetric surveillance capsule positions (13, 23', and 33') for both the reference and the plant specific core power distributions.
The plant specific data, based on the adjoint transport analysis, are meant to establish the absolute comparison of measurement with analysis.
The reference data derived from the forward calculation are provided as a conservative exposure evaluation against which plant specific fluence calculations can be compared.
Similar data are given in Table 3-2 for the reactor vessel inner radius.
Again, the three pertinent exposure parameters are listed for the reference and Cycles 1 through 25 plant specific power distributions.
R. E. Ginna Heatup and Cooldown Limit Curves June 1996
It is important to note that the data for the vessel inner radius'were taken at the clad/base
'(.,
metal interface, and thus, represent the maximum predicted exposure teveis oi the vessel plates and welds.
Radial gradient information applicable to $(E > 1.0 MeV), $(E > 0.1 MeV), and dpa/sec is given in Tables 3-3, 3P, and 3-5, respectively.
The data, obtained from the reference forward neutron transport calculation, are presented on a relative basis for each exposure parameter at several azimuthal locations.
Exposure distributions through the vessel wall may be obtained by normalizing the calculated or projected exposure at the vessel inner radius to the gradient data listed in Tables 3-3 through 3-5.
For example, the neutron flux g(E > 1.0 MeV) at the ~/4T depth in the reactor vessel wall along the 0'zimuth is given by:
Qq/4/0') = $(168.04, 0') +172.25, 0')
where:
tIt>>( 0') =
y(168.04,0 ) =
~ F(172.25,0') =
Projected neutron flux at the ~/4T position on the 0'zimuth.
Projected or calculated neutron flux at the vessel inner radius on the 0'zimuth..
Ratio of the neutron flux at the ~/4T position to the flux at the vessel inner radius for the 0'zimuth. This data is obtained from Table 3-3.
Similar expressions apply for exposure parameters expressed in terms of $(E > 0.1 MeV) and dpa/sec where the attenuation function F is obtained from Tables 3-4 and 3-5, respectively.
3.3 Neutron Dosimetry The passive neutron sensors included in the Ginna surveillance program are listed in Table 3-
- 6. Also given in Table 3-6 are the primary nuclear reactions and associated nuclear constants that were used in the evaluation of the neutron energy spectrum within the surveillance capsules and in the subsequent determination of the various exposure parameters of interest
[$(E > 1.0 MeV), tIt(E > 0.1 MeV), dpa/sec].
The relative locations of the neutron sensors within the capsules are shown in WCAP-13902"". The iron, nickel, copper, and cobalt-aluminum monitors, in wire form, were placed in holes drilled in spacers at several axial levels within the capsules.
The cadmium shielded uranium and neptunium fission monitors were accommodated within the dosimeter block located near the center of the capsule.
The use of passive monitors such as those listed in Table 3-6 does not yield a direct measure of the energy dependent neutron flux at the point of interest.
Rather, the activation or fission process is a measure of the integrated effect that the time and energy dependent neutron flux has on the target material over the course of the irradiation period.
An accurate assessment of the average neutron flux level incident on the various monitors may be derived from the R. E. Ginna Heatup and Cooidown Limit Curves June 1996
activation measurements only if the irradiation parameters are well known.
In particular, the following variables are of interest:
The measured specific activity of each monitor, The physical characteristics of each monitor, The operating history of the reactor, The energy response of each monitor, and The neutron energy spectrum at the monitor location.
The specific activity of each of the neutron monitors was determined using established ASTM procedures'I '~~'.
Following sample preparation and weighing, the activity of each monitor was determined by means of a lithium-drifted germanium, Ge(Li), gamma spectrometer.
The irradiation history of the Ginna reactor was obtained from NUREG-0020, "Licensed Operating Reactors Status Summary Report," and plant personnel for the cycles 1 through 25 operating period. The irradiation history applicable to the exposure of Capsules V, R, T, and S is given in Table 3-7.
Having the measured specific activities, the physical characteristics of the sensors, and the
'operating history of the reactor, reaction rates referenced to full power operation were determined from the following equation:
R-where:
R No F
Pi Pfei =
C/
Reaction rate averaged over the irradiation period and referenced to operation at a core power level of P, (rps/nucleus).
Measured specific activity (dps/gm).
Number of target element atoms per gram of sensor.
Weight fraction of the target isotope in the sensor material.
Number of product atoms produced per reaction.
Average core power level during irradiation period j (MW).
Maximum or reference power level of the reactor (MW).
Calculated ratio of g(E > 1.0 MeV) during irradiation period j to the time weighted average $(E > 1.0 MeV) over the entire irradiation period.
Decay constant of the product isotope (1/sec).
Length of irradiation period j (sec).
Decay time following irradiation period j (sec).
and the summation is carried out over the total number of monthly intervals comprising the irradiation period.
R. E. Ginna Heatup and Cooldown Limit Curves June 1996
In the equation describing the reaction rate calculation, the ratio [PP[P,] accounts for month by month variation of reactor core power level within any given fuel cycle as well as over multiple fuel cycles.
The ratio Ci, which can be calculated for each fuel cycle using the adjoint transport technology discussed in Section 3.2, accounts for the change in sensor reaction rates caused by variations in flux level induced by changes in core spatial power distributions from fuel cycle to fuel cycle. For a single cycle irradiation Ci is normally taken to be 1.0.
However, for multiple cycle irradiations, particularly those employing fow leakage fuel management, the additional Ci term should be employed.
The impact of changing flux levels for constant power operation can be quite significant for sensor sets that have been irradiated for many cycles in a reactor that has transitioned from non-low leakage to low leakage fuel m'anagement or for sensor sets contained in surveillance capsules that have been moved from one capsule location to another.
For the irradiation history of Capsules V, R, T, and S, the flux level term in the reaction rate calculations'was developed from the plant specific analysis provided in Table 3-1. Measured and saturated reaction product specific activities as well as the derived full power reaction rates are listed in Table 3-8.
The specific activities and reaction rates of the ~U sensors provided in Table 3-8 include corrections for ~U impurities, plutonium build-in, and gamma ray induced fissions.
Corrections for gamma ray induced fissions were also included in the specific activities and reaction rates for the Np sensors as well.
Values of key fast neutron exposure parameters were derived from the measured reaction rates using the FERRET least squares adjustment code'~'.
The FERRET approach used the measured reaction rate data, sensor reaction cross-sections, and a calculated trial spectrum as input and proceeded to adjust the group fluxes from the trial spectrum to produce a best fit (in a least squares sense) to the measured reaction rate data.
The "measured" exposure parameters along with the associated uncertainties were then obtained from the adjusted spectrum.
In the FERRET evaluations, a log-normal least squares algorithm weights both the a priori values and the measured data in accordance with the assigned uncertainties and correlations.
ln general, the measured values f are linearly related to the flux g by some response matrix A:
P~~)
P A(4 ~(~)
g where i indexes the measured values belonging to a single data set s, g designates the energy group, and u delineates spectra that may be simultaneously adjusted.
For example, R. E. Ginna Heatup and Cooldown Limit Curves June 1996
relates a set of measured reaction rates R, to a single spectrum $, by the multigroup reaction cross-section c@. The log-normal approach automatically accounts for the physical constraint of positive fluxes, even with large assigned uncertainties.
In the least squares adjustment, the continuous quantities (i.e., neutron spectra and cross-sections) were approximated in a multi-group format consisting of 53 energy groups.
The trial input spectrum was converted to the FERRET 53 group structure using the SAND-II codet~'.
This procedure was carried out by first expanding the 47 group calculated spectrum into the SAND-II 620 group structure using a SPLINE interpolation procedure in regions where group boundaries do not coincide. The,620 point spectrum was then re-collapsed into the group structure used in FERRET.
The sensor set reaction cross-sections, obtained from the ENDF/B-Vl dosimetry file'~', were also collapsed into the 53 energy group structure using the SAND-II code.
In.this instance, the trial spectrum, as expanded to 620 groups, was employed as a weighting function in the cross-section collapsing procedure.
Reaction cross-section uncertainties in the form of a 53 x 53 covariance matrix for each sensor reaction were also constructed from the information contained on the ENDF/B-Vl data files. These matrices included energy group to energy group uncertainty correlations for each of the individual reactions.
However, correlations between cross-sections for different sensor reactions were not included.
The omission of this additional uncertainty information does not significantly impact the results of the adjustment.
Due to the importance of providing a trial spectrum that exhibits a relative energy distribution close to the actual spectrum at the sensor set locations, the neutron spectrum input to the FERRET evaluation was taken from the center of the surveillance capsule modeled in the reference forward transport calculation.
While the 53 x 53 group covariance matrices applicable to the sensor reaction cross-sections were developed from the ENDF/B-Vl data files, the covariance matrix for the input trial spectrum was constructed from the following relation:
2 Mggl Rn
+ Rg RgI Pggl where Rspecifies an overall fractional normalization uncertainty (i.e., complete correlation) for the set of values.
The fractional uncertainties R~ specify additional random uncertainties for group g that are correlated with a correlation matrix given by:
where:
P I = [1-0] 6 I + 0 e R. E. Ginna Heatup and Cooldown Limit Curves June 1996
The first term in the correlation matrix equation specifies purely random uncertainties, while the second term describes short range correlations over a group range y (8 specifies the strength of the latter term). The value of 5 is 1 when g = g'nd 0 otherwise.
For the trial spectrum used in the current evaluations, a short range correlation of y = 6 groups was used.
This choice implies that neighboring groups are strongly correlated when 6 is close to 1.
Strong long range correlations (or anti-correlations) were justified based on information presented by R. E. Maerkeri'~. The uncertainties associated with the measured reaction rates included both statistical (counting) and systematic compohents.
The systematic component of the overall uncertainty accounts for counter eNciency, counter calibrations, irradiation history corrections, and corrections for competing reactions in the individual sensors.
Results of the FERRET evaluations of the Capsules V, R, T, and S dosimetry are given in Table 3-9. The data summarized in this table include fast neutron exposure evaluations in terms of 4 (E > 1.0 MeV), 4(E > 0.1 MeV), and dpa.
In general, excellent results were achieved in the fits of the adjusted spectra to the individual measured reaction rates.
The measured and FERRET adjusted reaction rates for each reaction are given in Table 3-10. An examination of Table 3-10 shows that, in all cases, reaction rates calculated with the adjusted spectra match the measured reaction rates to better than 8%. The adjusted spectra from the least squares evaluation is given in Table 3-11 in the FERRET 63 energy group structure.
'n Table 3-12, absolute comparisons of the measured and calculated fluence at the center of each capsule are presented.
The results for the Capsules V, R, T, and S dosimetry evaluations (M/C ratios of 1.03 for 4(E > 1.0 MeV)) are consistent with results obtained from similar evaluations of dosimetry from other reactors using methodologies based on ENDF/B-Vl cross-sections.
3.4 Projections of Reactor Vessel Exposure The best estimate exposure of the Ginna reactor vessel was developed using a combination of absolute plant specific transport calculations and all available plant specific measurement data.
In the case of Ginna, the measurement data base consists of the four surveillance capsules discussed in this report.
Combining this measurement data base with the plant specific calculations, the best estimate vessel exposure is obtained from the following relationship:
+~Fr. = <+m where:
4~~ ~ =
The best estimate fast neutron exposure at the location of interest.
The plant specific measurement/calculation (M/C) bias factor derived from the surveillance capsule dosimetry data.
R. E. Ginna Heatup and Cooldown Limit Curves June 1996
@Cate.
The absolute calculated fast neutron exposure at the location of interest.
The approach defined in the above equation is based on the premise that the measurement data represent the most accurate plant specific information available at the locations of the dosimetry; and, further that the use of the measurement data on a plant specific basis essentially removes biases present in the analytical approach and mitigates the uncertainties that would result from the use of analysis alone.
That is, at the measurement points the uncertainty in the best estimate exposure is dominated by the uncertainties in the measurement process.
At locations within the reactor vessel wall, additional uncertainty is incurred due to the analytically determined relative ratios among the various measurement points and locations within the reactor vessel wall.
For Ginna, the derived plant specific bias factors were 1.03, 1.13, and 1.08 for 4(E > 1.0 MeVj, 4(E > 0.1 MeV), and dpa, respectively.
Bias factors of this magnitude are fullyconsistent with experience using the BUGLE-93 cross-section library.
The use of the bias factors derived from the measurement data base acts to remove plant specific biases associated with the definition of the core source, actual vs. assumed reactor dimensions, and operational variations in water density within the reactor.
As a result, the overall uncertainty in the best estimate exposure projections within the vessel wall depends on the individual uncertainties in the measurement process, the uncertainty in the dosimetry location, and, in the uncertainty in the calculated ratio of the neutron exposure at the point of interest to that at the measurement location.
The uncertainty in the derived neutron flux for an individual measurement is obtained directly from the results of a least squares evaluation of dosimetry data.
The least squares approach combines individual uncertainty in the calculated neutron energy spectrum, the uncertainties in dosimetry cross-sections, and the uncertainties in measured foil specific activities to produce a net uncertainty in.the derived neutron flux at the measurement point. The associated uncertainty in the plant specific bias factor, K, derived from the M/C data base, in turn, depends on the total number of available measurements as well as on the uncertainty of each measurement.
In developing the overall uncertainty associated with the reactor vessel exposure, the positioning uncertainties for dosimetry are taken from parametric studies of sensor position performed as part a series of analytical sensitivity studies included in the qualification of the methodology.
The uncertainties in the exposure ratios relating dosimetry results to positions within the vessel wall are again based on the analytical sensitivity studies of the vessel thickness tolerance, downcomer water density variations and vessel inner radius tolerance.
Thus, this portion of the overall uncertainty is controlled entirely by dimensional tolerances associated with the reactor design and by the operational characteristics of the reactor.
R. E. Ginna Heatup and Cooidown Limit Curves June 1996
e 0
0
13 The net uncertainty in the bias factor, K, is combined with the uncertainty from the analytical sensitivity, study to define the overall fluence uncertainty at the reactor vessel wall. In the case of Ginna, the derived uncertainties in the bias factor, K, and the additional uncertainty from the analytical sensitivity studies combine to yield a net uncertainty of 212.9%.
Based on this best estimate approach, neutron exposure projections at key locations on the reactor vessel inner radius are given in Table 3-13. Along with the current (19.51 EFPY) exposure, projections are also provided for exposure periods of 28 EFPY, 32 EFPY, 42 EFPY, and 48 EFPY.
Projections for future operation were based on the assumption that the average exposure rates during the upcoming cycle 26 (18 month fuel cycle operation) irradiation period would continue to be applicable throughout plant life.
In the calculation of exposure gradients within the reactor vessel wall for the Ginna reactor vessel, exposure projections to 28, 32, 42 and 48 EFPY were also employed.
Data based on both a 4(E > 1.0 MeV) slope and a plant specific dpa slope through the vessel wall are provided in Table 3-14.
ln order to access RT>> vs fluence curves, dpa equivalent fast neutron fluence levels for the
~/4T and %T positions were defined by the relations:
yy q y(0q dp (
7) dpa(07) and gpss q y(pq dpa(e/47) dpa(07)
Using this approach results in the dpa equivalent fluence values listed in Table 3-14.
In Table 3-1'5 updated lead factors are listed for each of the Ginna surveillance capsules.
Lead factor data based on the accumulated fluence through the Cycle 26 projection are provided for each remaining capsule.
E. Ginna Heatup and Cooldown Limit Curves June 1996
FIGURE 3-1 ARRANGEMENTOF SURVEILLANCECAPSULES IN THE R. E. GINNA REACTOR VESSEL R (removed) 270'EACTOR VESSEL THERMALSHIELD CAPSULE (TYP) 10 10 570 180'o S (removed)
T (removed)
V (removed)
R. E. Ginna Heatup and Cooidown Limit Curves June 1996
FIGURE 3-2 PLAN VIEW OF A REACTOR VESSEL SURVEILLANCECAPSULE CHARPY SPECIMEN THERMAL SHIEU3 R. E. Ginna Heatup and Cooldown LimitCurves June 1996
0 0
TABLE 3-1 CALCULATEDFAST NEUTRON EXPOSURE RATES AND IRON ATOM DISPLACEMENT RATES AT THE SURVEILLANCECAPSULE CENTER
$(E > 1.0 MeV) (n/crn'-sec)
~Cele No.
Reference 1A 1B 2
3 5
6 7
8 9
10 1I 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 130 1.59e+11 1.195e+11 1.424e+11 1.369e+11 1.144e+11 1.083e+11 1.326e+11 1.375e+11 1.193e+11 1.434e+11 1.359e+11 1.242e+11 1.250e+11 1.377e+1 1 1.063e+11 1.027e+11 9.149e+1 0 9.652e+1 0 1.030e+1 1 9.259e+10 9.133e+10 9.723e+10 9.691e+10 1.038e+11 9.322e+10 9.072e+10 8.436e+10 8A18e+10 230 9.35e+10 7.051e+1 0 8.009e+10 8.096e+10 6.661e+10 7.266e+10 7.714e+10 8.240e+1 0 7.520e+10 8.433e+10 8.455e+10 7.865e+10 7.127e+10 7.241e+10 6.870e+10
?.053e+10 6.622e+10 6.054e+10 6.362e+10 6.261e+10 6.065e+10 6.228e+1 0 6.476e+10 6.575e+10 6.167e+10 5.884e+1 0 5.51 Oe+10 5.299e+10 33'.83e+10 6.684e+1 0 7.141 e+1 0 7.682e+1 0 6.263e+10 7.21 7e+1 0 7.069e+1 0 7.397e+1 0 7.026e+1 0 8.033e+1 0 8.303e+1 0 7.751e+1 0 6.474e+1 0 6.357e+1 0 6.252e+10 6.926e+10 6.567e+10 5.953e+10 5.936e+10 5.835e+10 5.883e+10 5.762e+10 5.982e+10 6.057e+10 6.004e+10 5.497e+10 5.293e+10 5.016e+10 R. E. Ginna Heatup and Cooldown LimitCurves June 1996
17 TABLE3-1 cont'd CALCULATEDFAST NEUTRON EXPOSURE RATES AND tRON ATOM DlSPLACEMENT RATES AT THE SURYEILLANCECAPSULE CENTER
$(E > 0.1 MeY) {n/cm'-sec)
~Cele No.
Reference 1A 1B 2
3 4
5 6
7 8
9 10 1l 12 13 14 15 16 17 18 19 20 2l
,22 23 24 25 26 6.02e+11 4.529e+11 5.398e+11 5.189e+11 4.337e+11 4.105e+11 5.025e+11 5.211e+11 4.520e+11 5.435e+11 5.149e+11 4.706e+11
'.737e+11 5.220e+11 4.030e+11 3.893e+11 3.468e+11 3.658e+11 3.904e+11 3.509e+11 3.462e+11 3.685e+11 3.673e+11 3.933e+11 3.533e+1 1 3A38e+1 1 3.197e+11 3.191e+11 230 3,22e+1 'l 2.426e+11 2.755e+1 1 2.785e+11 2.291e+11 2.500e+11 2.654e+11 2.835e+11 2.587e+11 2.901 e+11 2.909e+11 2.706e+11 2:452e+11 2.491e+11 2.363e+11 2.426e+11 2.278e+ l1 2.082e+11 2.189e+11 2.1 54e+11 2.086e+11 2.142e+11 2.228e+11 2.262e+11 2.121 e+11 2.024e+11 1.896e+11 1.823e+11 3.1 1 e+1 1 2.353e+11 2.5'l 3e+11 2.704e+11 2.205e+11 2.541 e+11 2.488e+11 2.604e+11 2.473e+11 2.828e+11 2.923e+ l1 2.728e+11 2.279e+11 2.238e+11 2,201e+11 2.438e+11 2.31 1 e+1 1 2.095e+11 2.089e+11 2.054e+11 2.071 e+11 2.028e+11 2.106e+11 2.132e+11 2.114e+11 1.935e+11 1.863e+11 1.766e+11 I
R. E. Ginna Heatup and Cooidown LimitCurves June 1996
18 TABLE 3-1 cont'd CALCULATEDFAST NEUTRON EXPOSURE RATES AND IRON ATOM DISPLACEMENT RATES AT THE SURVEILLANCECAPSULE CENTER Displacement Rate (dpa/sec)
~Cele No.
Reference 1A 1B 2
3 4
5 6
7 8
9 10 11 12 13 14 l5 16
'I7 18 19 20 21 22 23 24 25 26 130 2.83e-10 2.127e-1 0 2.535e-1 0 2.437e-1 0 2.037e-1 0 1.928e-1 0 2.360e-1 0 2.448e-1 0 2.123e-1 0 2.553e-1 0 2.418e-1 0 2.21 Oe-1 0 2.225e-1 0 2.452e-1 0 1.893e-1 0 1.829e-1 0 1.629e-1 0 1.718e-1 0 1.833e-1 0 1.648e-1 0 1.626e-1 0 1.731 e-1 0 1.725e-1 0 1.847e-1 0 1.659e-1 0 1.615e-1 0 1.502e-1 0 1.498e-1 0 230 1.59e-10 1.199e-1 0 1.362e-1 0 1.376e-1 0 1.132e-1 0 1.235e-10 1.311e-10 1.401 e-1 0 1.278e-10 1.434e-1 0 1.437e-1 0 1.337e-1 0 1.212e-1 0 1.231 e-10 1.168e-10 1 ~199e-10 1 ~126e-10 1.029e-1 0 1.082e-10 1.064e-1 0 1.031e-1 0 1.059e-1 0 1.101e-1 0 1.118e-1 0 1.048e-1 0 1.000e-1 0 9.368e-11 9.009e-11 33'.52e-10 1.150e-1 0 1.228e-1 0 1.321 e-1 0 1.077e-1 0 1.241e-1 0 1.21 6e-1 0 1.272e-1 0 1.208e-1 0 1.382e-1 0 1.428e-1 0 1.333e-1 0 1.113e-1 0 1.093e-1 0 1.075e-1 0 1.191e-10 1.129e-10 1.024e-10 1.021 e-1 0 1.004e-1 0 1.012e-10 9.910e-11 1.029e-1 0 1.042e-10 1.033e-1 0 9.454e-11 9.103e-11 8.627e-1 1
R. E. Ginna Heatup and Cooldown Limit Curves June 1996
'ABLE 3-2 CALCULATEDAZIMUTHALVARIATIONOF FAST NEUTRON EXPOSURE RATES AND IRON ATOM DISPLACEMENT RATES AT THE REACTOR VESSEL CLAD/BASE METALINTERFACE
$(E > 1.0 MeV) (n/cm'-sec)
~Cele No.
Reference 1A 18 2
3 5
6 7
8 9
10 12
'I3 14 15 I6 l7 18
'f9 20 21 22 23 24 25 26 po 4.14e+09 4.022e+1 0 4.784e+10 4.590e+1 0 3.792e+10 3ASSe+10 4A41e+10 4.505e+10 3.835e+10 4.828e+1 0 4.484e+10 4.092e+1 0 4.076e+1 0 4.746e+10 3.61 6e+1 0 3,404e+10 2,800e+10 3.078e+1 0 3.633e+1 0 3.008e+1 0 3.003e+1 0 3.127e+1 0 3.007e+1 0 3.51 Oe+1 0 3.039e+1 0 3.000e+1 0 2.794e+1 0 2.8S5e+1 0 15'.20e+09 2A52e+10 2.909e+10 2.811e+10 2.348e+1 0 2,290e+10 2.720e+1 0 2.836e+1 0 2.488e+1 0 2.938e+1 0 2.S1 6e+1 0 2.582e+1 0 2.570e+1 0 2.787e+1 0 2.258e+1 0 2.21 3e+1 0 2.003e+1 0
.2.043e+1 0 2.170e+1 0 1.995e+1 0 1.956e+10 2.06Se+10 2.080e+10 2.203e+10 1.995e+10 1.938e+10 1.803e+10 1.788e+10 30'.19e+09 1.680e+1 0 1.831 e+10 1.926e+1 0 1.576e+1 0 1.796e+10 1.799e+1 0 1.901 e+1 0 1.784e+1 0 2.014e+1 0 2.066e+1 0 1.930e+1 0 1.650e+1 0 1.635e+1 0 1.615e+1 0 1.735e+1 0 1.640e+1 0 I.4S2e+1 0 1.514e+1 0 1.496e+1 0 1.483e+10 1.475e+1 0 1.537e+1 0 1.553e+10 1.511e+1 0 1.406e+1 0 1.340e+10 1.275e+1 0 45O 1.83e+09 1.420e+10 1.517e+10 1.688e+10 1.374e+10 1.581e+10 1.501e+10 1.475e+10 1.459e+10 1.735e+10 1.820e+10 1.683e+10 1.543e+1 0 1.500e+10 1.355e+10 1.690e+ I0 1.664e+10 1.546e+10 1.380e+1 0 1.300e+10 1.366e+10 1.31 5e+1 0 1.338e+10 1.380e+10 1.408e+1 0 1.249e+10 1.226e+1 0 1.181e+1 0 R. E. Ginna Heatup and Cooldown Limit Curves June 1996
20 TABLE3-2 cont'd CALCULATEDAZIMUTHALVARIATIONOF FAST NEUTRON EXPOSURE RATES
'AND IRON ATOM DISPLACEMENT RATES AT THE REACTOR VESSEL CLAD/BASE METALINTERFACE
$(E > 0.1 MeV) (n/cm'-sec)
~Cele No.
Reference 1A 1B 2
3 4
5 6
7 8
9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 00 3.25e+10 3.121 e+11 3.712e+ I I 3.562e+11 2.942e+11'.706e+11 3.446e+11 3.496e+11 2.976e+11 3.746e+11 3.480e+11 3.176e+11 3.163e+11 3.683e+11 2.806e+11 2.641 e+11 2.1 73e+1 1 2.389e+11 2.81 9e+11 2.334e+11 2.330e+11 2.427e+11 2.333e+11 2.724e+11 2.358e+11 2.328e+11 2.168e+11 2.239e+11 15o 2.76e+10 2.116e+11 2.511e+11 2.426e+11
'.026e+11 1.976e+11 2.348e+11 2A47e+11 2.147e+11 2.536e+11 2.430e+11 2.228e+11 2.21 8e+11 2.406e+11 l.948e+ I 1 1.910e+11 1.729e+11
'1.764e+11 1.872e+1 1 1.722e+ I 1 1.688e+11 1.785e+11 1.795e+11 1.901e+11 1.721 e+1 1 1.672e+11 1.556e+11 1.543e+11 30'.90e+10 1.465e+11 1.597e+11 1.680e+11 1.374e+11 1,566e+11 1.568e+11 1.658e+11 1.556e+11 1.757e+11 1.802e+11 1.683e+11 1.438e+11 1.425e+11 1.408e+11 1.513e+11 1.430e+11 1.292e+11 1.320e+11 1.304e+11 1.293e+11 1.286e+11 1.340e+11 1.354e+11 1.31 8e+11 1.226e+11 1.169e+11 1.112e+11 45'.50e+10 1.164e+11 1.244e+11 1.384e+11 1.126e+11 1.297e+11 1.230e+11 1.210e+11 1.197e+11 1.422e+11 1.492e+11 1.380e+11 1.266e+11 1.230e+11 1.111e+11 1.386e+11 1.364e+11 1.268e+11 1.1 31 e+1 1 1.066e+11 1.120e+11 1.079e+11 1.097e+11 1.1 32e+11 1.155e+11 1.024e+11 1.005e+11 9.687e+10 R. E. Ginna Heatup and Cooidown Limit Curves June 1996
I
)
TABLE3-2 cont'd CALCULATEDAZIMUTHALVARIATIONOF FAST NEUTRON EXPOSURE RATES AND IRON ATOM DISPLACEMENT RATES AT THE REACTOR VESSEL CLAD/BASE METALINTERFACE Displacement Rate (dpa/sec)
~Cele No.
Reference 1A 1B 2
3 4
5 6
7 8
9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 00 1.17e-11 1.138e-1 0 1.354e-10 1.299e-10 1.073e-1 0 9.870e-11 1.257e-1 0 1.275e-1 0
~
1.085e-1 0 1.366e-1 0 1.269e-1 0 1.158e-10 1.154e-10 1.343e-1 0 1.023e-1 0 9.633e-11 7.924e-11 8.71 2e-11 1.028e-1 0 8.513e-11 8.498e-11 8.850e-11 8.51 Oe-11 9.933e-11 8.599e-11 8.491e-11 7.906e-11 8.164e-11 15o 9.70e-12 7.430e-11 8.815e-11 8.517e-11 7.113e-11 6.938e-11 8.242e-11 8.593e-11 7.538e-11 8.903e-11 8.531 e-11 7.823e-11 7.787e-11 8.446e-11 6.841 e-11 6.706e-11 6.070e-11 6.192e-11 6.574e-11 6.046e-11 5.927e-11 6.267e-11 6.301 e-11 6.674e-11 6.044e-11 5.871e-11
'5.463e-11 5.41 8e-11 30'.70e-12 5.141 e-11 5.604e-11 5.894e-11 4.823e-11 5.496e-11 5.503e-11 5.818e-11 5.460e-11 6.164e-11 6.322e-11 5.904e-1 1 5.048e-11 5.002e-11 4.941 e-11 5.308e-11 5.019e-11 4.534e-11 4.632e-11 4.577e-11 4.538e-11 4.514e-11 4.702e-11 4.752e-11
.4.624e-11
- 4.301 e-11 4.101 e-11 3.901e-11 45'.36e-12 4.160e-11 4A46e-11 4.946e-11 4.025e-11 4.633e-11 4.397e-11 4.322e-11 4.276e-11 5.082e-11 5.331 e-11 4.930e-11 4.522e-11 4.395e-11 3.971 e-1 1 4.952e-11 4.875e-11 4.530e-11 4.042e-11 3.809e-11 4.002e-11 3.854e-11 3.920e-11 4.045e-1 1 4.127e-1 1 3.660e-11 3.592e-11 3.461e-11 R. E. Ginna Heatup and Cooldown LimitCurves June 1996
22 TABLE3-3 RELATIVERADIALDISTRIBUTION OF $(E > 1.0 MeV)
WITHINTHE REACTOR VESSEL WALL RADIUS
, ~cm 168.04 168.27 168.88 169.75 170.93 172.25 173.53 174.98 176.46 177.58 179.03 180.66 181.63 182.60 184.06 184.87 00 1.000 0.987 0.940 0.862 0.754 0.639 0.540 OA44 0.362 0.308 0.250 0.196 0.169 0.144 0.110 0.101 AZIMUTHALANGLE 15'0 1.000 1.000 0.987 0.985 0.942 0.937 0.865 0.857 0.757 0.749 0.644 0.636 0.546 0.539 0.451 0.444 0.370 0.363 0.317 0.311 0.259 0.253 0.206 0.201 0.179 0.175 0.154 0.15 I 0.122 0.120 0.113 0.112 45'.000 0.987 0.942 0.866 0.760 0.647 0.550 0.454 0.372 0.318 0.260 0.206 0.178 0.154 0.122 0.113 Note:
Base Metal Inner Radius =
Base Metal ~/4T =
Base Metal ~AT =
Base Metal %T =
Base Metal Outer Radius =
168.04 cm.
172.25 cm.
176.46 cm.
180.66 cm.
184.87 cm.
R. E. Ginna Heatup and Cooldown Limit Cuwes June 1996
TABLE3-4 RELATIVERADIALDISTRIBUTIONOF g(E > 0.1 MeV)
WITHINTHE REACTOR VESSEL WALL RADIUS
~cm 168.04 168.27 168.88 169.75 170.93 172.25 173.53 174.98 176.46 177.58 179.03 180.66 181.63 182.60 184.06 184.87 00 1.000 1.005 1.002 0.980 0.934 0.873 0.809 0.736 0.662 0.606 0.536 0.461 0.416 0.369 0.298 0.276 AZIMUTHALANGLE 15 30'.0001.000 1.007 1.005 1.007 1.004 0.990 0.985 0.948 0.945 0.891 0.889 0.831 0.831 0.763 0.763 0.693 0.694 0.640 0.642 0.573 0.577 0.502 0.507 0.458 0.466 0.415 0.423 0.348 0.361 0.327 0.343 45'.OOO 1.007 1.008 0.992 0.953 0.899 0.841 0.773 0.703 0.650 0.582 0.509 0.465 0.421 0.357 0.339 Note:
Base Metal Inner Radius =
Base Metal ~/4T =
Base Metal ~/2T =
.Base Metal ~/4T =
Base Metal Outer Radius =
168.04 cm.
172.25 cm.
176.46 cm.
180.66 cm.
184.87 cm.
R. E. Ginna Heatup and Cooldown Limit Curves June 1996
24 TABLE 3-5 RELATIVE RADIALDISTRIBUTION OF dpa/sec WITHINTHE REACTOR VESSEL WALL RADIUS
~cm 168.04 168.27 168.88 169.75
'I70.93 172.25 173.53 174.98 176.46 177.58
'I79.03 180.66 181.63 182.60 184.06
.184.87 Note:
Base Base Base Base Base 00 I.OOO 0.988 0.951 0.889 0.804 0.712 0.630 0.547 0.472 0.420 0.360 0.301 0.267 0.234 O. I 87 0.173 Metal Inner Radius =
Metal ~14T =
Metal ~AT =
Metal ~/4T =
Metal Outer Radius =
AZIMUTHALANGLE 15'0'.000 1.000 0.990 0.988 0.955 0.950 0.896 0.889 0.814 0.805 0.726 0.716 0.648 0.638 0.568 0.558 0.495 0.486 0.445 0.436 0.386 0.379 0.328 0.322 0.296 0.291 0.264 0.261 0.219 0.220 0.206 0.208 168.04 cm.
172.25 cm.
176.46 cm.
180.66 cm.
184.87 cm.
45'.OOO 0.989 0.954 0.857 0.812 0.723 0.644 0.563 0.490 OA39 0.380 0.322 0.289 0.258 0.216 0.205 R. E. Ginna Heatup and Cooidown Limit Curves June 1996
25 TABLE3-6 NUCLEAR PARAMETERS USED IN THE EVALUATIONOF NEUTRON SENSORS Monitor Material Copper Iron Nickel Uranium-238 Neptunium-237 Cobalt-Al Reaction of Interest
~Cu (n,u)
~Fe (n,p)
~NI (n,p)
~U (n,f)
Np (n,f)
"Co (n,y)
Target Atom Fraction 0.6917 0.0580 0.6827 0.9996 1.0000 0.0015
Response
~Ran e
E > 4.7 MeV E > 1.0 MeV E > 1.0 MeV E > 0.4 MeV E > 0.08 MeV E > 0.015 MeV Product Half-life 5.271 y 312.5 d 70.78 d 30.17 y 30.17 y 5.271 y Fission Yield 6.00 6.27 Note: ~U and ~Np monitors are cadmium shielded.
R. E. Ginna Heatup and Cooldown LimitCurves June 1996
TABLE 3-7 MONTHLYTHERMALGENERATION DURING THE FIRST TWENTY-FIVE FUEL CYCLES OF THE GINNA REACTOR Cycle 1A Cycle 1B Cycle 2 Cycle 3 Month Nov-69 Dec-69 Jan-70 Feb-70 Mar-70 Apr-70 May-70 Jun-70 Jul-70 Aug-70 Sep-70 Oct-70 Nov-70 Dec-70 Jan Feb-71 Mar-71 Thermal Gen.
MWt-hr 0
0 435541 435541 435541 435541 435541 435541 435541 930964 860611 481017 830385 840563 831856 956228
.27388 Month Apr-71 May-71 Jun-71 Juf-71 Aug-71 Sep-71 Oct-71 Nov-71 Dec-71 Jan-72 Feb-72 Mar-72 Apr-72 Thermal Gen.
MWt-hr 0
330899 831633 835525 922141 913338 957036 956391 941632 956804 955012 740009 655096 Month May-72 Jun-72 Jul-72 Aug-72 Sep-72 Oct>>72 Thermal Gen.
MWt-hr 0
7722 818270 89799I 910771 329771 Month Nov-72 Dec-72 Jan-73 Feb-73 Mar-73 Apr-73 May-73 Jun-73 Jul-73 Aug-73 Sep-73 Oct-73 Nov-73 Dec-73 Jan-74 Feb-74 Mar-74 Thermal Gen.
MWt-hr 381843 1053950 886067 852420 942682 914688 947502 906810 678681 945755 951662 818471 980073 923316 2908 0
0 R. E. Ginna Heatup and Cooldown LimitCurves June 1996
27 TABLE3-7 cont'd MONTHLYTHERMALGENERATlON DURING THE FIRST TWENTY-FIVE FUEL CYCLES OF THE GINNA REACTOR Cycle 4 Cycle 5 Cycle 6 Cycle 7 Month Apr-74 May-74 Jun-74 Jul-74 Aug-74 Sep-74 Oct-74 Nov-74 Dec-74 Jan-75 Feb-75 Mar-75 Apr-75 Thermal Gen.
MWt-hr 89688 739986 582048 710424 895176 992088 1034808 562206 1102170 1123848 1018003 325920 0
Month May-75 Jun-75 Jul-75 Aug-75 Sep-75 Oct-75 Nov-75 Dec-75 Jan-76 Feb-76 Mar-76 Thermal Gen.
MWt-hr 154344 624696 1088160 1115328 1086864 1047312 1085784 1036296 532560 0
0 Month Apr-76 May-76 Jun-76 Jul-76 Aug-76 Sep-76 Oct-76 Nov-76 Dec-76 Jan-77 Feb-77 Mar-77 Apr-77 Thermal Gen.
MWt-hr 262656 667032 1069992 1079064 151512 747336 331608 1087128 1055472 1108248 1013808 1119720 541872 Month May-77 Jun-77 Jul-77 Aug-77 Sep-77 Oct-77 Nov-77 Dec-77 Jan-78 Feb-78 Mar-78 Apr-78 Thermal Gen.
MWt-hr 195504 1081512 810480 973440 1083960 1116840 948096 1088328 872256 888480 870024 0
R. E. Ginna Heatup and Cooldown Limit Curves June 1996
28 TABLE 3-7 cont'd MONTHLYTHERMALGENERATION DURING THE FIRST TWENTY-FIVE FUEL CYCLES OF THE GINNAREACTOR Cycle 8 Month Thermal Gen.
MWt-hr Cycle 9 Month Thermal Gen.
MWt-hr Cycle 10 Thermal Gen.
Month MWt-hr Month Thermal Gen.
MWt-hr Cycle 11 May-78 Jun-78 Jul-78 Aug-78 Sep-78 Oct-78 Nov-78 Dec-78 Jan-79 Feb-79 245784 1082184 1107864 1081872 1079232 1120344 1078368 1066896 1116480 290064 Mar-79 Apr-79 May-79 Jun-79 Jul-79 Aug-79 Sep-79 Oct-79 Nov-79 Dec-79 Jan-80 Feb-80 Mar-80 Apr-80 312 856560 1111296 1085088 212952 952248 1084848 1089960 1059936 486144 1127568 1055352 935736 0
May-80 Jun-80 Jul-80 Aug-80 Sep-80 Oct-80 Nov-80 Dec-80 Jan-S1 Feb-81 Mar-81 Apr-81 May-81 238944 1084200 1120680 1119528 1084296 1124472 72 1035552 1043136 1018488 1118424 604920
. 0 Jun-81 Jul-S1 Aug-81'ep-81 Oct-81 Nov-81 Dec-81 Jan-82 Feb-82 Mar-82 Apr-82 335016 1110480 1112424 1087728 1122552 1028280 1121904 881568 0
0 0
R. E. Ginna Heatup and Cooidown Limit Curves June 1996
29 I'ABLE 3-7 cont'd MONTHLYTHERMALGENERATION DURING THE FIRST TWENTY-FIVE FUEL CYCLES OF THE GINNAREACTOR Cycle 12 Cycle 13 Cycle 14 Cycle 15 I
May-82 Jun-82 Jul-82 Aug-82 Sep-82 Oct-82 Nov-82 Dec-82 Jan-83 Feb-83 Mar-83 Apr-83 May-83 Thermal Gen.
MWt-hr 87336 1078152 1118424 1087944 866400 352944 1082784 1121448 996120 1018944 906408 0
0 Month Jun-83 Jul-83 Aug-83 Sep-83 Oct-83 Nov-83 Dec-83 Jan-84 Feb-84 Mar-84 Apr-84 Thermal Gen.
MWt-hr 1S3007 1114224 1126776 1025808 1123392 1088808 1121328 1114992 1027536 64896 0
Month May-84 Jun-84 Jul-84 Aug-84 Sep-84 Oct-84 Nov-84 Dec-84 Jan-85 Feb-85 Mar-85 Thermal Gen.
MWt-hr 160440 1002072 1123896 1106304 1089792 1127040 1084032 1126992 1127616 964344 27648 Month Apr-85 May-85 Jun-8S Jul-85 Aug-85 Sep-85 Oct-85 Nov-85 Dec-85 Jan-86 Feb-86 Thermal Gen.
MWt-hr 614232 1126704 1034760 1126464 1126776 1010352 1116312 1042512 1119144 1056672 198456 II R. E. Ginna Heatup and Cooidown LimitCurves June 1996
TABLE 3-7 cont'd MONTHLYTHERMALGENERATION DURING THE FIRST TWENTY-FIVE FUEL CYCLES OF THE GINNAREACTOR Cycle 16 Cycle 17 Cycle 1S Cycle 19 Month Mar-86 Apr-86 May-86 Jun-86 Jul-86 Aug-86 Sep-86 Oct-86 Nov-86 Dec-86 Jan-87 Feb-87 Thermal Gen.
MWt-hr 280056 1091832 1129992 1080912 1035216 1123488 1093752 1075584 1040304 1121160 1088064 171240 Month Mar-87 Apr-S7 May-87 Jun-87 Jul-87 Aug-87 Sep-87 Oct-87 Nov-87 Dec-87 Jan-88 Feb-88 Thermal Gen.
MWt-hr 708461 1088688 1088112 1089638 1126248 1125559 1057654 1128811 1091532 1113677 1029730 125990 Month Mar-88 Apr-88 May-88 Jun-88 Jul-88 Aug-88 Sep-88 Oct-88 Nov-88 Dec-88 Jan-89 Feb-89 Mar-89 Apr-89 Thermal Gen.
MWt-hr 223258 1072037 1108301 939036 1044756 1113811 1068655 1125024 1088028 1104055 1061697 1006661 556975 0
Month May-89 Jun-89 Jul-89 Aug-89 Sep-89 Oct-89 Nov-89 Dec-89 Jan-90 Feb-90 Mar-90 Apr-90 Thermal Gen.
MWt-hr 13848 878174 1040174 710578 1080422 1122288 1074098 1107108 1105469 1012205 808550 0
R. E. Ginna Heatup and Cooidown Limit Curves June 1996
31 TABLE 3-7 cont'd MONTHLYTHERMALGENERATION DURING THE FIRST TWENTY-FIVE FUEL CYCLES OF THE GINNAREACTOR Cycle 20 Cycle 21 Cycle 22 Cycle 23
'onth May-90 Jun-90 Jul-90 Aug-90 Sep-90 Oct-90 Nov-90 Dec-90 Jan-91 Feb-91 Mar-91 Apr-91 Thermal Gen.
MWt-hr 769318 1012320 1098566 1098437 959083 1093051 1056218 715030 1096111 988894 668083 0
Month May-91 Jun-91 Jul-91 Aug-91 Sep-91 Oct-91 Nov-91 Dec-91 Jan-92 Feb-92 Mar-92 Apr-92 Thermal Gen.
MWt-hr 599282 1065760 1100696 1022214 1049792 1103223 1067033 1102728 1106865 780610 868877 0
Month May-92 Jun-92 Jul-92 Aug-92 Sep-92 Oct-92 Nov-92 Dec-92 Jan-93 Feb-93 Mar-93 Thermal Gen.
MWt-hr 610650 948520 1097720 1101108
'1059758 1103684 1045291 1106157 1100087 1004079 386820 Month
'I Apr-93 May-93 Jun-93 Jul-93 Aug-93 Sep-93 Oct-93 Nov-93 Dec-93 Jan-94 Feb-94 Mar-94 Thermal Gen.
MWt-hr 91944 1059620 1068915 1105566 1088384 1068747 1106564 731946 1099588 1088034 998757 116893 R. E. Ginna Heatup and Cooldown LimitCurves June 1996
32 TABLE 3-7 cont'd MONTHLYTHERMALGENERATION DUMNGTHE FIRST TWENTY-FIVE FUEL CYCLES OF THE GINNAREACTOR Cycle 24 Cycle 25 Month Apr-94 May-94 Jun-94 Jul-94 Aug-94 Sep-94 Oct-94 Nov-94 Dec-94 Jan-95 Feb-95 Mar-95 Thermal Gen.
MWt-hr 283107 1019528 960449 1103548 615679 1060583 1115487 1081351 1115511 1095892 1000530 911453 Month Apr-95 May-95 Jun-95 Jul-95 Aug-95 Sep-95 Oct-95 Nov-95 Dec-95 Jan-96 Feb-96 Mar-96 Apr-96 Thermal Gen.
MWt-hr 9
934804 1040844 1085722 975736 1055684 1093897 1057617 1095856 1096191 1024993 947615 30568 R. E. Ginna Heatup and Cooldown Limit Curves June 1996
33 TABLE 3-8 MEASURED SENSOR ACTIVITIESAND REACTION RATES SURVEILLANCECAPSULE V SATUIMTEDACTIVITIESAND REACTION RATES i
~ Reaction
'Cu (n,or) ~Co Top Top Middle Bottom Middle Bottom Fe (n,p) ~Mn Wl Rl S6 P7 W2 R3 S8 P9
'¹ (np) 'Co Middle
"'U (n,g '"Cs Middle Np (nf)'s Middle Measured Activity
~ds/ ~~m 7.38e+04 6.77e+04 7.48e+04 8.13e+04 2.47e+06 2.57e+06 2.18e+06 2.57e+06 2.04e+06 1.95e+06 2.02e+06 2.10e+06 2.38e+07 2.30e+05 1.23e+06 Saturated Activity
~ds/ ~~m 4.64e+05 4.25e+05 4.70e+05 S.l le+05 5.00e+06 5.21e+06 4.42e+06 5.21e+06 4.13e+06 3.95e+06 4.09e+06 4.26e+06 6.51e+07 7.29e+06 3.90e+07 Reaction Rate
$~s/at~om 6.79e-17 6.23e-17 6.88e-l7 7.48e-l7 7.60e-15 7.91e-15 6.7 le-15 7.9le-l5 7.67e-15 7.33e-15 7.59e-15 7.89e-l5 8.83e-15 4.81e-14 2.45e-13 R. E. Ginna Heatup and Cooldown Limit Curves June 1996
34 TABLE 3-8 cont'd MEASURED SENSOR ACTIVITIESAND REACTION RATES SURVEILLANCECAPSULE R SATURATED ACTIVITIESAND REACTION RATES Reaction
'Cu (n,u) ~Co Top Top Middle Bottom Middle Bottom
~Fe (n,p) ~Mn W 13 R 14 P 18 W 14 R 15 P 19 Ni (n,p) 5'Co Mddle Co (n,y) ~Co Top Top Middle Mddle Bottom Middle Bottom
'o (n,y) ~Co (Cd)
Top Top Middle Mddle Bottom Middle Bottom
'U (n,f)'s Mddle Np (n,f) '37Cs Middle Measured Activity
+d)~s/ ~m 1.08e+05 9.68e+04 1.15e+05 1.15e+05 2.08e+06 1.98e+06 2.06e+06 1.63e+06 1.70e+06 1.85e+06 5.83e+06 3.09e+07 3.14e+07 2.96e+07 2.94e+07 2.94e+07 1.19e+07 1.18e+07 1.07e+07 1.24e+07 1.24e+07 4.32e+05 4.25e+06 Saturated Activity
~ds/ ~~m 4.42e+05
'.96e+05 4.70e+05 4.70e+05 5.19e+06 4.94e+06 5.14e+06 4.07e+06 4.24e+06 4.62e+06 7.38e+07 1.26e+08 1.28e+08 1.21e+08 1.20e+08 1.20e+08 4.87e+07 4.83e+07 4.38e+07 5.07e+07 5.07e+07 7.8le+06 7.68e+07 Reaction Rate Q)s/ato~m 6.47e-17 5.80e-l7 6.89e-17 6.89e-17 7.88e-15 7.51e-15 7.81e-15 7.54e-15 7.87e-15 8.56e-l5 1.00e-l4 8.00e-12 8.13e-12 7.66e-l2 7.61e-12 7.61e-12 3.2le-12 3.18e-12 2.88e-12 3.34e-12 3.34e-12 5.15e-14 4.83e-13 R. E. Ginna Heatup and Cooldown Limit Curves June 1996
TABLE 3-8 cont'd MEASURED SENSOR ACTIVITIESAND REACTION RATES SURVEILLANCECAPSULE T SATURATED ACTIVHlES AND REACTION RATES Reaction "Cu (n,u) ~Co Top Top Mddle Bottom Mddle Bottom Measured Activity
~ds/ ~~m 1.60e+05 1.40e+05 1.66e+05 1.74e+05 Saturated Activity
~ds/~m 3.52e+05 3.08e+05 3.65e+05 3.83e+05 Reaction Rate
$~s/a~tom 5.10e-17 4.47e-l7 5.29e-17 5.55e-17
)
'Fe (n,p) ~Mn S 22 P 28 W 21 S 23 P 29 W 22 "Ni (n,p) "Co Mddle "Co (n,y) ~Co Top Top Middle Middle Bottom Middle Bottom "Co (n,y) ~Co (Cd)
Top Top Middle Middle Bottom Middle Bottom
"'U (n,f) '"Cs Mddle Np (n,f) '~~Cs Middle 1.14e+06 1.27e+06 1.30e+06 1.01e+06 1.03e+06 1.10e+06 8.62e+05 3.17e+07 3.06e+07 3.03e+07 3.27e+07 3.07e+07 1.21e+07 1.13e+07 1.16e+07 1.26e+07 1.20e+07 7.4le+05 6.09e+06 3.38e+06 3.76e+06 3.85e+06 2.99e+06 3.05e+06 3.26e+06 5.30e+07 6.98e+07 6.73e+07 6.67e+07 7.20e+07 6.76e+07 2.66e+07 2.49e+07 2.55e+07 2.77e+07 2.64e+07 5.36e+06 4.40e+07 5.13e-15 5.71e-15 5.85e-l5 5.50e-l5 5.61e-l5 5.99e-15 7.19e-l5 4.32e-12 4.17e-l2 4.13e-12 4.46e-12 4.19e-12 1.68e-12 1.57e-l2 1.62e-12 1.76e-12 1.67e-12 3.53e-14 2.76e-13 R. E. Ginna Heatup and Cooldown Limit Cuwes June 1996
36 TABLE 3-8 cont'd MEASURED SENSOR ACTIVITIESAND REACTION RATES SURVEILLANCECAPSULE S
SATURATED ACTIVlTIESAND REACTION RATES Reaction
'Cu (n,u) ~Co Top Top Middle Bottom Middle Bottom Measured Activity
~ds/~m 2.06e+05 1.82e+05 1.98e+05 2.18e+05 Saturated Activity
~ds/~m 3.07e+05 2.71e+05 2.95e+05 3.24e+05 Reaction Rate
$rps/a~tom 4.44e-17 3.93e-l7 4.27e-l7 4.70e-17 Fe (n,p)
Mn P 31 Ni (n,p) ~~Co Mddle
$9C
(
~) 60C Top Top Middle Middle Bottom Mddle Bottom
'~Co (n,y)
Co (Cd)
Top Top Middle Mddle Bottom Mddle Bottom U (n,f) '~ Cs Middle
~'Np (n,f) '3'Cs Mddle 1.62e+06 8.51e+06 3.55e+07 3.71e+07 3.39e+07 3.60e+07 3.45e+07 1.43e+07 1.37e+07 1.31e+07
'.45e+07 1.35e+07 1.40e+06 1.11e+07 2.94e+06 4.29e+07 5.28e+07 5.52e+07 5.05e+07 5.36e+07 5.14e+07 2.13e+07 2.04e+07 1.95e+07 2.16e+07 2.01e+07 4.64e+06 3.68e+07 4.46e-15 5.82e-15 3.3le-l2 3.46e-12 3.16e-12 3.36e-12 3.22e-12 1.38e-12 1.32e-12 1.26e-12 1.39e-12 1.30e-l2 3.06e-14 2.31e-13 R. E. Ginna Heatup and Cooldown LimitCurves June 1996
37
'ABLE 3-9
SUMMARY
OF NEUTRON DOSMETRY RESULTS SURVEILLANCECAPSULES V, R, T AND S Measured Flux and Fluence for Capsule V
~uantit
[n/cm~-sec]
$ (E> 1.0 MeV) g (E > 0.1 MeV) g (E < 0.414 eV) dpa/sec
~uantit
[n/cm~]
1.129e+11 4
(E > 1.0 MeV) 4.426e+11 4 (E > 0.1 MeV) 2.314e+11 4 (E < 0.414 eV) 2.062e-10 dpa Fluence 5.028e+18 1.97le+19 1.03le+19 9.183e-03 Uncertain 10%
21%
83%
15%
Measured Flux and Fluence for Capsule R quantity
[n/cm'-sec]
$ (E> 1.0MeV)
$ (E > 0.1 MeV) g (E < 0.414 eV) dpa/sec Flux 1.374e+11 5.892e+11 1.964e+11 2.606e-lO
~usntit
[n/cm ]
4 (E > 1.0MeV) 4 (E>0.1MeV) 4 (E <0.414 eV) dpa Fluence 1.105e+19 4.738e+19 1.579e+19 2.096e-02 Uncertain 8%
15%
20%
11%
Measured Flux and Fluence for Capsule T
~uantiti
[n/cm -sec]
g (E> 1.0 MeV) f (E> 0.1 MeV)
$ (E< 0.414 eV) dpa/sec Flux 8.61 le+10 3.250e+11 1.080e+11 1.528e-10
~uantit~
[n/cm ]
4 (E> 1.0MeV) 4 (E>0.1 MeV)
Ci (E < 0.414 eV) dpa Fluence 1.864e+19 7.035e+19 2.338e+19 3.307e-02 Uncertain 8%
15%
19%
10%
Measured Flux and Fluence for Capsule S
~uan~ttr
[n/cm~-sec]
$ (E > 1.0MeV)
$ (E >0.1 MeV)
$ (E < 0.414 eV) dpa/sec Flux 6.982e+10 2.743e+ll 8.341e+10 1.268e-10
~uan~ttr
[n/cm']
4 (E > 1.0 MeV) 4 (E > 0.1 MeV)
(E < 0.414 eV) dpa Fluence 3.746e+19 1A72e+20 4.475e+19 6.803,e-02 Uncertain 8%
15%
19%
11%
R. E. Ginna Heatup and Cooldown Limit Curves June 1996
38 TABLE 3-10 COMPARISON OF MEASURED AND FERRET CALCULATED REACTION RATES AT THE SURVEILLANCECAPSULE CENTER Surveillance Capsule V Reaction Rate (rps/nucleus)
Adjusted Measured Cate.
M/C
~Ad'usted
~'Cu (n,a)
Fe (n,p)
Ni (n,p)
U (n,f) (Cd) 6.84e-17 7.58e-l5 8.84e-15 3.94e-14 6.73e-17 7.41e-15 9.36e-15 3.7le-14 1.02 1.02 0.94 1.06 Surveillance Capsule R Reaction Rate (rps/nucleus) 6'Cu (n,a)
~Fe (n,p)
'8Ni (n,p)
~'U (n,f) (Cd) 7Np (n,f) (Cd)
'~Co (n,y)
'o (n,y) (Cd)
Measured 6.51e-17 7.86e-15 1.00e-14 4.14e-l4 4.75e-13 7.80e-12 3.19e-l2 Adjusted Cate.
6A4e-17 7.77e-15 1.04e-14 4.23e-14 4.38e-13 7.8le-12 3.19e-12 M/C
~Ad usted 1.01 1.01 0.96 0.98 1.08 1.00 1.00 R. E. Ginna Heatup and Cooldown LimitCurves June 1996
39 TABLE3-10 cont'd COMPAMSON OF MEASURED AND FERRET CALCULATED REACTION RATES AT THE SURVEILLANCECAPSULE CENTER Surveillance Capsule T Reaction Rate (rps/nucleus)
"Cu (n,u)
~Fe (n,p)
Ni (n,p)
U (n,f) (Cd)
'Np (n,f) (Cd)
'~Co (ng)
'Co (n,y) (Cd)
Measured 5.10e-l7 5.63e-15 7.19e-l5 2.77e-14 2.72e 4.26e-12 1.66e-12 Adjusted Cele.
5.06e-17
. 5.58e-15 7.40e-15 2.79e-14 2.54e-13 4.26e-12 1.66e-12 M/C
~Ad'usted 1.01 1.01 0.97 0.99 1.07 1.00 1.00 Surveillance Capsule S Reaction Rate (rps/nucleus)
"Cu (n,a)
Fe (n,p) s¹ (np)
~ U (n,f) (Cd)
~'Np (n,f) (Cd)
Co (n,y)
"Co (n,y) (Cd)
Measured 4.34e-17 4.46e-15 S.82e-lS 2.21e-14 2.27e-13 3.30e-12 1.33e-12 Adjusted Calc.
4.26e-17 4.47e-15 5.98e-15 2.25e-l4 2.11e-13 3.31e-12 1.33e-12 M/C
~Ad'usted 1.02 1.00 0.97 0.98 1.08 1.00 1.00 R. E. Ginna Heatup and Cooldown LimitCurves June 1996
40 TABLE 3-11 ADJUSTED NEUTRON ENERGY SPECTRUM AT THE CENTER OF SURVEILLANCECAPSULES
~Grou Energy QMeV~
Capsule V Flux
~Ge Energy
~MeV Flux n/cm~-sec 1
2 3
4 5
6 7
8 9
10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 1.73e+01 1.49e+01 1.35e+01 1.16e+Ol 1.00e+01 8.61e+00 7.41e+00 6.07e+00 4.97e+00 3.68e+00 2.87e+00 2.23e+00 1.74e+00 1.35e+00 1,11e+00 8.21e-01
'.39e-01 4.98e-01 3.88e-01 3.02e-01 1.83e-01 1.lie-01 6.74e-02 4.09e-02 2.55e-02 1.99e-02 1.50e-02 7.97e+06 1.73e+07 6.54e+07 1.82e+08 4.16e+08 7.40e+08 1.80e+09 2.81e+09 5.99e+09 7.14e+09 1.35e+10 1.76e+10 2.34e+10 2.48e+10 4.23e+10 4.70e+10 5.04e+10 3.43e+10 4.79e+10 5.89e+10 5.45e+10 4.22e+10 3.52e+10 2.19e+10 2.09e+10 1.40e+10 2.32e+10 28 29 30 31 32 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 9.12e-03 5.53e-03 3.35e-03 2.84e-03 2.40e-03 2.03e-03 1.23e-03 7.49e-04 4.54e-04 2.75e-04 1.67e-04 1.01e-04 6.14e-05 3.73e-05 2.26e-05 1.37e-05 8.31e-06 5.04e-06 3.06e-06 1.86e-06 1.13e-06 6.83e-07 4.14e-07 2.51e-07 1.52e-07 9.24e-08 2.52e+10 2.71e+10 8.57e+09 8.32e+09 8.29e+09 2.50e+10 2.50e+10 2.44e+10 2.25e+10 2.41e+10 2.52e+10 2.49e+10 2.47e+10 2.45e+10 2.40e+10 2.32e+10 2.28e+10 2.27e+10 2.26e+10 2.24e+10 1.99e+10 1.85e+10 3.40e+10 3.70e+10 4.03e+10 1.20e+11 Note: Tabulated energy levels represent the upper energy in each group.
R. E. Ginna Heatup and Cooldown Limit Curves June 1996
TABLE3-11 cont'd ADJUSTED NEUTRON ENERGY SPECTRUM AT THE CENTER OF SURVEILLANCECAPSULES Capsule R
~Grou 1
2 3
4 5
6 7
8 9
10 11 12 13 14 15 16 17 18
. 19 20 21 22 23 24 25 26 27 Energy QMeV~
1.73e+01 1.49e+01 1.35e+01 1.16e+01 1.00e+01 8.61e+00 7.4le+00 6.07e+00 4.97e+00 3.68e+00 2.87e+00 2.23e+00 1.74e+00 1.35e+00 1.1 le+00 8.21e-01 6.39e-01 4.98e-01 3.88e-01 3.02e-01 1.83e-01 1.lie-01 6.74e-02 4.09e-02 2.SSe-02 1.99e-02 1.SOe-02 Flux n/cm2-sec 7.58e+06 1.63e+07 6.14e+07 1.70e+08 3.93e+08 7.08e+08 1.76e+09 2.84e+09 6.29e+09 7.80e+09 1.54e+10 2.08e+10 2.91e+10 3.24e+10 5.75e+10 6.52e+10 7.06e+10 4.78e+10 6.62e+10 8.01e+10 7.24e+10 5.47e+10 4.42e+10 2.67e+10 2.48e+10 1.62e+10 2.61e+10
~Grou 28 29 30 31 32 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 48 50 51 52 53 Energy
~e~V 9.12e-03 5.53e-03 3.35e-03 2.84e-03 2.40e-03 2.03e-03 1.23e-03 7.49e-04 4.54e-04 2.75e-04 1.67e-04 1.0le-04 6.14e-05 3.73e-05 2.26e-05 1.37e-05 8.31e-06 5.04e-06 3.06e-06 1.86e-06 1.13e-06 6.83e-07 4.14e-07 2.51e-07 1.52e-07 9.24e-08 Flux n/cm -sec 2.78e+10 2.94e+10 9.14e+09 8.74e+09 8.58e+09 2.54e+10 2.50e+10 2.40e+10 2.17e+10 2.30e+10 2.23e+10 2.36e+10 2.38e+10 2.38e+10 2.35e+10 2.29e+10 2.27e+10 2.26e+10 2.26e+10 2.25e+10 1.99e+10 1.76e+10 3.15e+10 3.31e+10 3.Sle+10 9.66e+10 Note: Tabulated energy levels represent the upper energy in each group.
)
R. E. Ginna Heatup and Cooldown Limit Curves June 1996
42 TABLE 3-11 cont'd ADJUSTED NEUTRON ENERGY SPECTRUM AT THE CENTER OF SURVEILLANCECAPSULES Capsule T
~Grou 1
2 3
4 5
6 7
8 9
10 11 12 13 14 15 16 17 18-19 20 21 22 23 25 26 27 Energy
~e~V 1.73e+01 1.49e+01 1.35e+01 1.16e+01 1.00e+Ol 8,61e+00 7.41e+00 6.07e+00 4.97e+00 3.68e+00 2.87e+00 2.23e+00 1.74e+00 1.35e+00 1.11e+00 8.21e-01 6.39e-01 4.98e-01 3.88e-01 3.02e-01 1.83e-01 1.11e-01 6.74e-02 4.09e-02 2.55e-02 1.99e-02 1.50e-02 Flux n/cm -sec 6.40e+06 1.37e+07 5.05e+07 1.39e+08 3.17e+08 5.59e+08 1.39e+09 2.16e+09 4.52e+09 5.27e+09 1.02e+10 1.33e+10 1.79e+10 1.92e+10 3.26e+10 3.57e+10 3.77e+10 2.53e+10 3A4e+10 4.15e+10 3.70e+10 2.78e+10 2.25e+10 1.35e+10 1.26e+10 8.29e+09 1.35e+10
~Grou ¹ 28 29 30 31 32 33 3,4 35 36 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 Energy QMeV~
9.12e-03 5.53e-03 3.35e-03 2.84e-03 2.40e-03 2.03e-03 1.23e-03 7.49e-04 4.54e-04 2.75e-04 1.67e-04 1.0 le-04 6.14e-05 3.73e-05 2.26e-05 1.37e-05 8.3le-06 5.04e-06 3.06e-06 1.86e-06 1.13e-06 6.83e-07 4.14e-07 2.51e-07 1.52e-07 9.24e-08 Flux
'/cm~-sec 1.44e+10 1.52e+10 4.69e+09 4.50e+09 4.43e+09 1.32e+10 1.30e+10 1.24e+10 1.13e+10 1.19e+10 1.16e+10 1.22e+10 1.22e+10 1.22e+10 1.20e+10 1.17e+10 1.16e+10 1.16e+10 1.16e+10 1.16e+10 1.04e+10 9.06e+09 1.55e+10 1.72e+10 1.91e+10 5.62e+10 Note: Tabulated energy levels represent the upper energy in each group.
R. E. Ginna Heatup and Cooldown'Limit Curves June 1996
43 TABLE 3-11 cont'd ADJUSTED NEUTRON ENERGY SPECTRUM AT THE CENTER OF SURVEILLANCECAPSULES Capsule S
~Grou 1
2 3
4 5
6 7
8 9
10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 Energy
~MeV 1.73e+01 1.49e+Ol 1.35e+01 1.16e+01 1.00e+01 8.61e+00 7.41e+00 6.07e+00 4.97e+00 3.68e+00 2.87e+00 2.23e+00 1.74e+00 1.35e+00 1.lie+00 8.21e-01 6.39e-01 4.98e-01 3.88e-01 3.02e-OI 1.83e-01 1.lie-01 6.74e-02 4.09e-02 2.5Se-02 1.99e-02 1.50e-02 Flux n/cm~-sec 5.49e+06 1.17e+07 4.31e+07 1.17e+08 2.66e+08 4.64e+08 1.13e+09 1.73e+09 3.62e+09 4.24e+09 8.14e+09 1.07e+10 IA5e+10 1.57e+10 2.69e+10 2.99e+10 3.19e+10 2.16e+10 2.96e+10 3.6le+10 3.25e+10 2.46e+10 1.99e+10 1.20e+10 1.12e+10 7.35e+09 1.19e+10
~Grou 28
'29 30 31 32 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 Energy QMeV~
9.12e-03 5.53e-03 3.35e-03 2.84e-03 2.40e-03 2.03e-03 1.23e-03 7.49e-04 4.54e-04 2.75e-04 1.67e-04 1.01e-04 6.14e-05 3.73e-05 2.26e-05 1.37e-05 8.31e-06 5.04e-06 3.06e-06 1.86e-06 1.13e-06 6.83e-07 4.14e-07 2.5le-07 1.52e-07 9.24e-08 Flux n/cm -sec 1.27e+10 1.34e+10 4.15e+09 3.95e+09 3.85e+09 1.13e+10 1.11e+10 1.04e+10 9.36e+09 9.82e+09 9.16e+09 1.00e+10 1.01e+10 1.02e+10 1.02e+10 9.92e+09 9.89e+09 9.93e+09 9.94e+09 9.91e+09 8.80e+09 7.60e+09 1.30e+10 1.39e+10 1.50e+10 4.16e+10 Note: Tabulated energy levels represent the upper energy in each group.
R. E. Ginna Heatup and Cooldown Limit Curves June 1996
44 TABLE 3-12 COMPARISON OF CALCULATEDAND MEASURED INTEGRATED NEUTRON EXPOSURE OF GINNASURVEILLANCECAPSULES V, R, T, AND S re(E > 1.0 MeV)
[n/cm~]
4(E > 0.1 MeV)
[n/cm2]
dpa CAPSULE V Calculated 5.864e+18 2.223e+19 1.044e-02 Measured 5.028e+18 1.971e+19 9.183 e-03 0.86 0.89 0.88 4(E > 1.0 MeV)
[n/cm~]
4(E > 0.1 MeV)
[n/cm~]
dpa CAPSULE R Calculated 1.013e+19 3.839e+19 1.803e-02 Measured 1.105e+19 4.738e+19 2.096e-02 1.09 1.23 1.16 C(E > 1.0 MeV) fn/cm~]
4(E > 0.1 MeV)
[n/cm~]
dpa CAPSULE T Calculated 1.669e+19 5.74le+19 2.837e-02 Measured 1.864e+19 7.035e+19 3.307e-02 1.12 1.23 1.17 rlr(E > 1.0 MeV)
[n/cm ]
4(E > 0.1 MeV)
[n/cm',]
dpa CAPSULE S Calculated 3.575e+19 1.259e+20 6.150e-02 Measured 3.746e+19 1.472e+20 6.803e-02 1.05 1.17 1.11 R. E. Ginna Heatup and Cooldown Limit Curves June 1996
45 TABLE3-13 NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS ON THE REACTOR VESSEL CLAD/BASEMETALINTERFACE Best Estimate Ex osure 19.51 EFP at the Reactor Vessel Inner Radius 4 (E > 1.0 MeV) 4 (E > 0.1 MeV) dpR PO 2.32e+19 1.97e+20 6.88e-02 15'.47e+19 1.39e+20 4.68e-02 30'.05e+19 1.00e+20 3.36e-02 45'.69e+18 8.38e+19 2.86e-02 Best Estimate Ex osure 28 EFP at the Reactor Vessel Inner Radius po C (E > 1.0 MeV) 3.11e+19 4 (E > 0.1 MeV) 2.65e+20 dpR 9.24e-02 15o 1.96e+19 1.86e+20 6.24e-02 30'.40e+19 1.34e+20 4.49e-02 45'.29e+19 1.13e+20 3.86e-02 Best Estimate Ex osure 32 EFP at the Reactor Vessel Inner Radius 4 (E > 1.0 MeV) 4 (E > O.l MeV) dpa PO 3.49e+19 2.97e+20 1.03e-01 15'.20e+19 2.08e+20 6.98e-02 30'.56e+19 1.50e+20 5.02e-02 45'.45e+19 1.27e+20 4.33e-02 Best Estimate Ex osure 42 EFP at the Reactor Vessel Inner Radius 4 (E > 1,0MeV) 4 (E > 0.1 MeV) dpa PO 4.42e+19 3.77e+20 1.31e-pl 15'.78e+19 2.63e+20 8.83e-02 30'.98e+19 1.89e+20 6.34e-02 45'.83e+19 1.61e+20 5.51e-02 Best Estimate Ex osure 48 EFPY at the Reactor Vessel Inner Radius 4 (E > 1.0 MeV) 4 (E > 0.1 MeV) dpa PO 4.98e+19 4.25e+20 1.48e-pl 15'.13e+19 2.96e+20 9.93e-02 30'.22e+19 2.13e+20 7.14e-02 45'.06e+19 1.82e+20 6.22e-02 R. E. Ginna Heatup and Cooldown Limit Curves June 1996
46 TABLE3-14 NEUTRON EXPOSURE VALUES WITHINTHE GINNAREACTOR VESSEL FLUENCE BASED ON E > 1.0 MeV SLOPE 28 EFPY 4
> 1.0 MeV n/cm Surface
'/4 T 3/4 T 00 3.11e+19 1.99e+19 6.10e+18 15'.96e+19 1.27e+19 4.05e+18 30'.40e+19 8.89e+18 2.81e+18 45'.29e+19 8.38e+18 2.67e+18 32 EFPY 4
> 1.0 MeV n/cm Surface
'/4 T s/4 T 00 3.49e+19 2.23e+19 6.83e+18 15'.20e+19 1.41e+19 4.53e+1S 30'.56e+19 9.94e+1S 3.14e+18 45'.45e+19 9.37e+18 2.98e+18 42 EFPY 4
> 1.0 Me n/cm~
Surface
'/4 T 3/4 T po 4.42e+19 2.83e+19 8.67e+18 15 2.78e+19 1.79e+19 5.72e+18 30 1.98e+19 1.26e+19 3.97e+18 45'.83e+19 1.18e+19 3.77e+18 48 EFPY 4
> 1.0 MeV n/cm Surface
~/4 T 3/4 T po 4.98e+19 3.18e+19 9.77e+1S 15'.13e+19 2.01e+19 6A4e+18 30'.22e+19 1.4le+19 4.47e+18 45'.06e+19 1.33e+19 4.25e+18 R. E. Ginna Heatup and Cooldown Limit Curves June 1996
(g TABLE3-14 cont'd A
NEUTRON EXPOSURE VALUES WITHINTHE GINNAREACTOR VESSEL FLUENCE BASED ON dpa SLOPE 28 EFPY 4
> 1.0 MeV n/cm Surface
'/4 T
~/~ T po 3.1le+19 2.22e+19 9.37e+18 15'.96e+19 1.43e+19 6A4e+18 30'.40e+19 1.00e+19 4.50e+18 45'.29e+19 9.36e+18 4.17e+18 32 EFPY 4
> 1.0 Me n/cm~
Surface
'/4 T 3/4 T po 3.49e+19 2.48e+19 1.05e+19 15'.20e+19 1.59e+19 7.21e+18 30'.56e+19 1.12e+19 5.03e+18 45'.45e+19 1.05e+19 4.66e+18 42 EFPY 4
> 1.0 MeV n/cm~
Surface r/4 T
~/4 T po 4.42e+19 3.15e+19 1.33e+19 15'.78e+19 2.02e+19 9.11e+18 30'.98e+19 1.42e+19 6.36e+18 1.83e+19 1.32e+19 5.90e+18 48 EFPY 4
> 1.0 MeV n/cm~
Surface t/4 T
~/4 T po 4.98e+19 3.55e+19 1.50e+19 15'.13e+19 2.27e+19 1.03e+19 30'.22e+19 1.59e+19 7.16e+18 2.06e+19 1.49e+19 6.64e+18 R. E. Ginna Heatup and Cooldown Limit Curves June 1996
48 TABLE 3-15 UPDATED LEADFACTORS FOR GINNA SURVEILLANCECAPSULES
~Ca sule VÃ Ttb)
Rtc)
S(dj Ntc) pfe)
Lead Factor 2.97 2.99 1.82 1.77 1.78 1.89
[a] - Withdrawn at the end of Cycle 1B.
[b] - Withdrawn at the end of Cycle 3.
= [c] - Withdrawn at the end of Cycle 9.
[d] - Withdrawn at the end of Cycle 22.
[e] - Capsules remaining in the reactor.
R. E. Ginna Heatup and Cooldown Limit Curves June 1996
49 4
CRlTERlA FOR ALLOWABLEPRESSURE-TEMPERATURE RELATIONSHIPS N
ot3)
Appendix 6 to 10 CFR Part 50, Fracture Toughness Requirements specifies fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary of light water nuclear power reactors to provide adequate margins of safety during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. The ASME Boiler and Pressure Vessel Code forms the basis for these requirements.
Section XI, Division 1, "Rules for Inservice Inspection of Nuclear Power Plant Components", Appendix 6'~', contains the conservative methods of analysis.
The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, K, for the metal temperature at that time. K is obtained from the reference fracture toughness curve, defined in Appendix 6 of Section Xl of the ASME Code.
The Kcurve is given by the following equation:
K -2878 1233 el
- where, K= reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductilitytemperature RT>>
Therefore, the governing equation for the heatup-cooldown analysis is defined in Appendix 8 of the ASME Code as follows:
(2)
- where, K, = stress intensity factor caused by membrane (pressure) stress Krr stress intensity factor caused by the thermal gradients K= function of temperature relative to the RT>>T of the material C =
2.0 for Level A and Level B service limits C =
1.5 for hydrostatic and leak test conditions during which the reactor core is not critical R. E. Ginna Heatup and Cooldown Limit Curves June 1996
50 At any time during the heatup or cooldown transient, K~ is determined by the metal temperature at the tip of a postulated flaw at the 1/4T and 3/4T location, the appropriate value for RT>>and the reference fracture toughness curve.
The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, K, for the reference flaw are computed.
From Equation 2, the pressure stress intensity factors are obtained and, from these, the allowable pressures are calculated.
For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference flaw of Appendix 6 to the ASME Code is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the.thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates.
Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations.
From these relations, composite limitcurves are constructed for each cooldown rate of interest.
The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw. During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel inner diameter.
This condition, of course, is not true for the steady-state situation.
It follows that, at any given reactor coolant temperature, the hT (temperature) developed during cooldown results in a higher value of Kat the 1/4T location for finite cooldown rates than for steady-state operation.
Furthermore, if conditions exist so that the increase in Kexceeds K, the calculated allowable pressure during cooldown will be greater than the steady-state value.
Ml I
The above procedures are needed because there is no direct control on temperature at the 1/4T location and, therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a coo!down ramp. The use of the composite curve eliminates this problem and ensures conservative operation of the system for the entire cooldown period.
Three separate calculations are required to determine the limitcurves for finite heatup rates.
As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as finite heatup rate conditions assuming the presence of a 1/4T defect at the inside of the wall. The heatup results in compressive stresses at the inside surface that alleviate the tensile stresses produced by internal pressure.
The metal temperature at the crack tip lags the coolant temperature; therefore, the K for the 1/4T crack during heatup is lower than the K for the 1/4T crack during steady-state conditions at the same coolant temperature.
During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower Kvalues do not offset each other, and the pressure-temperature curve based on steady-state R. E. Ginna Heatup and Cooldown Limit.Curves June 1996
conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered.
Therefore, both cases have to'be analyzed in order to ensure that at any coolant temperature the tower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.
The second portion of the heatup analysis concerns the calculation of the pressure-temperature limitations for the case in which a 1/4T flaw located at the 1/4T location from the outside surface is assumed.
Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present.'hese thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp.
Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.
Following the generation of pressure-temperature curves for both the steady state and finite heatup rate situations, the final limitcurves are produced by constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data.
At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration.
The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the inside to the outside, and the pressure limitmust at all times be based on analysis of the most critical criterion.
10 CFR Part 50, Appendix G addresses the metal temperature of the closure head flange and vessel flange regions.
This rule states that the metal temperature of the closure flange regions must exceed the material unirradiated RT>>, by at least 120'F for normal operation when the pressure exceeds 20 percent of the preservice hydrostatic test pressure (3106 psig""), which is 621 psig for R. E. Ginna.
The limiting unirradiated RT>>, of -52 F"" occurs in the vessel flange of the R. E. Ginna reactor vessel, so the minimum allowable temperature of this region is 68'F at pressures greater than 621 psig.
However, this limitdoes not impact the curves presented in Figures 6-1 through 6-8.
R. E. Ginna Heatup and Cootdown Limit Curves June 1996
52 5
CALCULATIONOF ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99, Revision 2, the adjusted reference temperature (ART) for each material in the beltline region is given by the following expression:
ART=lnitialRTN~+KRT~Margin Initial RTN>> is the reference temperature for the unirradiated material as defined in paragraph
'NB-2331 of Section III of the ASME Boiler and Pressure Vessel Code'~'.
if measured values of initial RT>>, for the material in question are not available, generic mean values for that class of material may be used if there are sufficient test results to establish a mean and standard deviation for the class of material.
d RT>> is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows:
t
~RT CF~g(om~.<alogy To calculate ART>> at any depth (e.g., at 1/4T or 3/4T), the following formula must first be used to attenuate the fluence at the specific depth.
(4)
(5) where x inches (vessel beltline thickness is 6.5 inchest"t) is the depth into the vessel wall measured from the vessel clad/base metal interface.
The resultant fluence is then placed in Equation 4 to calculate the 8 RT>> at the specific depth.
The Westinghouse Radiation Engineering and Analysis group evaluated the vessel fluence projections and the results are presented in Section 3 of this report. The evaluation used the ENDF/B-VI scattering cross-section data set.
This is consistent with the methods presented in WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown LimitCurves"~'.
Table 5-1 contains the reevaluated fluence values used to calculate the ART values for all beltline materials in the Ginna reactor vessel.
Additionally, the surveillance capsule fluence values are presented in Table 5-2.
R. E. Ginna Heatup and Cooldown Limit Cuives June 1996
53 TABLE5-1 Summary of the Peak Pressure Vessel Neutron Fluence Values used for the Calculation of ART Values (E > 1.0 MeV, n/cm')
EFPY Surface 1/4 T 3/4 T 24 2.74 x 10" 1.86 x1O" 8.50 x 10'8 32 40 3.>> x1O" 3.49 x 10'.23 x 10" 2.11 x 1O" h
2.36 x 10'.87 x 10" 9.66 x1O" 1.O8 x1O" 1.S1 x1O"'ABLE 5-2 Measured Integrated Neutron Exposure of the R. E. Ginna Surveillance Capsules
'Capsule V
R S
Fluence 5.028 x 10" n/cm'E > 1.0 MeV) 1.105 x 10" n/cm'E > 1.0 MeV) 1.864 x 10" n/cm'E > 1.0 MeV) 3.746 x 10" n/cm~ (E > 1.0 MeV)
Margin is calculated as, M = 2 * (o,'+ o~')'~. The standard deviation for the initial RT>>
margin term, o is O'F when the initial RTNpy is a measured value, and 17'F when a generic value is used.
The standard deviation for the ERTpy margin term, o is 17'F for plates or forgings when surveillance capsule data is not used and 8.5'F for plates or forgings when surveillance capsule data is used.
For welds, a, is 28 'F when surveillance capsule data is not used and 14'F when surveillance capsule data is used.
a~ need not exceed one-half the mean value of BRTNpy Per the request of the Rochester Gas and Electric Corporation', the margin term to be used for the circumferential weld Seam SA-847 when using surveillance capsule data is 48.3'F.
R. E. Ginna Heatup and Coofdown LimitCurves June 1996
Contained in Table 5-3 is a summary of the Measured 30 ft-Ib transition temperature shifts of the beltline.materials contained in the surveillance program"".
TABLE5-3 Measured 30 ft-lb Transition Temperature Shifts of the Beltline Materials Contained in the Surveillance Program l
Material Capsule Measured 30 ft-Ib Transition Temperature Shift Lower Shell Forging 125P666 V
R 25oF 25oF 30'F 420F Intermediate Shell Forging 125S255 V
R OOF OOF OOF S
60'F Surveillance Weld Metal V
R 140oF 165'F 150oF S
205oF R. E. Ginna Heatup and Cooldown Limit Curves June 1996
55 Table 5-4 Contains a summary of the weight percent of copper, the weight percent of nickel and the initial RT>>, of the beltline materials, vessel flanges and the'surveillance program weld metal. The weight percent values of Cu and Ni given in Table 5-4 were used to generate the calculated chemistry factor (CF) values based on Tables 1 and 2 of Regulatory Guide 1.99, Revision 2, and presented in Table 5-6. Table 5-5 provides the calculation of the CF values based on surveillance capsule data, Regulatory Guide 1.99, Revision 2, Position 2.1.
TABLE5-4 Reactor Vessel Beltline Material Unirradiated Toughness Properties Material Description Closure Head Flange Vessel Flange Intermediate Shell Forging 125S255 Lower Shell Forging 125P666 Circumferential Weld Seam SA Cu (%)
0.07 0.05 0.25 Ni (%)
0.69 0.69 0.56 Initial RTNoT ('F)
-75
-52 20 40
-4.8 TABLE5-4A Compilation of Copper and Nickel Weight Percent Values for the Ginna Surveillance Program Weld Metal Reference 51 51 54 54 54 54 54 Average Weight % Cu 0.23 0.22 0.25 0.25 0.27 0.22 0.23 0.23 0.21 0.25 0.236 Weight % Ni 0.56 0.50 0.57 0.52 0.58 0.45 0.50 0.50 0.46 0.53 0.517 R. E. Ginna Heatup and Coo!down Limit Curves June 1996
56 TABLE5-5 Calculation of Chemistry Factors using Ginna Surveillance Capsule Data Material Lower Shell Forging 125P666 (Tangential)
Capsule V
Fluence"'.028 x 10" 1.105 x10>>
1.864 x 10>>
3.746 x10'F 0.8081 1.0279 1.1706 1.3418 hRT~ar 25'F 25oF 30OF 42oF FF 'RT~
20.20F 25 7OF 35.1'F 56.4'F FF2 0.6530 1.0566 1.3703 1.8004 SUM:
137.4'F 4.8803 Chemistry Factor = 137.4 ~ 4.8803 = 28.2'F Intermediate Shell Forging 125S255 (Tangential)
R 5.028 x 10'.105 x 10>>
1.864 x 10>>
3.746 x10>>
0.8081 1.0279 1.1706 1.3418 04F 00F 60OFSUM'4F 04F 80 5'F 80.5'F 0.6530 1.0566 1.3703 1.8004 4.8803 Chemistry Factor = 80.5 ~ 4.8803 = 16.5'F Circ.
Weld Seam SA-847"~
V 5.028 x 10" 1.105 x 10>>
1.864 x 10>>
3.746 x 10>>
0.8081 1.0279 1.1706 1.34'I8 149.7'F 176.40F 160 4'F 219.1'F SUM:
121.0'F 181.3'F 187.8 F 294.0'F 784.1'F 0.6530 1.0566 1.3703 Chemistry Factor = 784.1 ~ 4.8803 = 160.7'F (a)
Fluence values are in units of nlcm2, E > 1.0 MeV.
(b)
The BRTNpy values given in this Table have been multiplied by a ratio factor of 1.069.
The ratio factor was calculated per Regulatory Guide 1,99, Revision 2 as follows:
Ratio Factor = CF,
~
~ ~ = 170.4'F 159;4'F
= 1.069 R. E. Ginna Heatup and Cooldown LimitCurves June 1996
57 TABLE5-6 Summary of the Ginna Reactor Vessel Beltline Material Chemistry Factors Material Inter. Shell Forging 125P666 Lower Shell Forging 125S255 Circ. Weld Seam SA-847 Reg. Guide 1.99, Rev. 2, Position 1.1 CFs 44oF 31'F 170 4'F Reg. Guide 1.99, Rev. 2, Position 2.1 CFs 16.5'F 28.2'F 160.7'F Contained in Table 5-7 is a summary of the fluence factors (FF) used in the calculation of adjusted reference temperatures for the Ginna reactor vessel beltline materials.
TABLE5-7 Summary of the Calculated Fluence Factors Used for the Generation of the 24, 28, 32 and 40 EFPY Heatup and Cooldown Curves EFPY 1/4 T FF 3/4T FF 24 28 32 1 ~17 1.20 1.23 0.95 1.00 1.02 40 1.28 1.08 The adjusted reference temperature (ART) must be calculated for 24, 28, 32 and 40 EFPY for each beltline material at the 1/4T and 3/4T locations.
In addition, ART values must be calculated per Regulatory Guide 1.99, Revision 2, Position 1.1 (RG 1.99, R2, P1.1) and Regulatory Guide 1.99, Revision 2, Position 2.1 (RG 1.99, R2, P2.1).
Contained in Tables 5-8 through 5-15 is the calculation of the ART values used for the generation of the heatup and cooldown curves.
R. E. Ginna Heatup and Cooldown Limit Curves June 1996
58 TABLE5-8 Calculation of the ART Values for the 1/4T Location and 24 EFPY Material Method FF Margin IRT~
ART Inter. Shell Forging 125S255 RG 1.99, R2, P1.1 RG 1.99, R2, P2.1 16.5'F 1.17 51 5OF 1.17 19.3'F OF 17'F 20OF 20'F 106'F 56'F Lower Shell Forging 125P666 RG 1.99, R2, P1.1 RG 1.99, R2, P2.1 31'F 28.2'F 1.17 1.17 36 3OF 33 OOF 17oF 4Q'F 40oF 110'F 90'F Circ. Weld Seam SA-847 RG 'I.99, R2, P1.1 170.4'F RG 1.99, R2, P2.1 160.74F 1.17 188.0'F 1.17 199.40F 56oF 48.3'F A.8'F A.8'F 251'F 232'F TABLE5-9 Calculation of the ART Values for the 3/4T Location and 24 EFPY Material Method CF hRT~
Margin ART Inter. Shell Forging 125S255 RG 1.99, R2, P1.1 RG 1.99, R2, P2.1 44'F 0.95 16.5OF 0.95 41.80F 15 7OF 15.7'F 200F 20'F 964F 51OF Lower Shell Forging 125P666 Circ. Weld Seam SA-847 RG 1.99, R2, P1.1 RG 1.99, R2, P2.1 RG 1.99, R2, P1.1 RG 1.99, R2, P2.1 31'F 0.95 28.2'F 0.95 170 4oF 0 95 160.7'F 0.95 26.80F 17'F 161.9'F 56oF 152.7'F 48.3'F 29.5'F 29 O'F 4Q'F 99OF 404F 84oF A O' 196'F A 8OF 213OF R. E. Ginna Heatup and Cooldown Limit Curves June 1996
h
59 TABLE5-10 Calculation of the ART Values for the 1/4T Location and 28 EFPY Material Method FF B Rior Margin IRTvor ART Inter. Shell Forging 125S255 RG 1.99, R2, P1.1 RG 1.99, R2, P2.1 16.5'F 1.20 52.8'F 1.20 19.8'F 34oF 17'F 20oF 20'F 107oF 57'F Lower Shell Forging 125P666 Circ. Weld Seam SA-847 RG 1.99, R2, P1.1 RG 1.99, R2, P2.1 RG 1.99, R2, P1.1 RG 1.99, R2, P2.1 31'F 28.2'F 170.4oF 160.7oF 1.20 37.2'F 1.20 192.8'F 1.20 33.8'F 1.20 204.5'F 40'F 17'F 40'F 56'F
<.8oF 48 3oF A 8oF 111'F 91'F 256'F 236'F TABLE5-11 Calculation of the ART Values for the 3/4T Location and 28 EFPY Material Method CF Margin ART Inter. Shell Forging 125S255 Lower Shell Forging 125P666 RG 1.99, R2, P1.1 RG 1.99, R2, P2.1 RG 1.99, R2, P1.1 RG 1.99, R2, P2.1 16.5'F 31oF 28.20F 44 O' 1.00 16.5'F 1.00 31.0'F 1.00 28.2'F 34oF 16 5'F 31.0'F 17oF 20'F 20oF 40oF 40'F 98oF 53oF 102'F 85oF Circ. Weld Seam SA-847 RG 1.99, R2, P1.1 RG 1.99, R2, P2.1 170 4oF 160 7'F 1.00 170.4'F 160.7'F 56'F 8oF 48.3'F A.8'F 222oF 204'F R. E. Ginna Heatup and Cooldown Limit Curves June 1996
0
60
'ABLE 5-12 Calculation of the ART Values for the 1/4T Location and 32 EFPY Material inter. Shell Forging 125S255 Method RG 1.99, R2, P1.1 RG 1.99, R2, P2.1 CF 1.23 16.5'F 1.23 BRTNor 54.14F 20.3'F Margin tRTNm ART 20'F 108'F 204F y
574F Lower Shell Forging 125P666 Circ. Weld Seam SA-847 RG 1.99, R2, P1.1 RG 1.99, R2, P2.1 RG 1.99, R2, P1.1 RG 1.99, R2, P2.1 31'F 28.2'F 170 44F 123 1.23 1.23 160.7'F 1.23 38.1'F 34 7'F 209.64F l74F 564F 197.7'F 48.3'F 40'F 112'F 40'F 92'F A 84F 241'F
<.8'F 261'F
'ABLE 5-13 Calculation of the ART Values for the 3/4T Location and 32 EFPY Material Method FF Margin IRTNOT ART inter. Shell Forging 125S255 RG 1.99, R2, P1.1 RG 1.99, R2, P2.1 444F 16.5'F 1 02 44 O' 16.84F 16.8'F 204F 204F 99'F 544F Lower Shell Forging 125P666 RG 1.99, R2, P1.1 RG 1.99, R2, P2.1 314F 28.2'F 1.02 31.6'F 28.8'F 31.6'F 17'F 40'F 40'F 1034F 86'F Circ. Weld Seam SA~7 RG 1.99, R2, P2.1 160.7'F RG 1.99, R2, P1.'f 170.4'F 1.02 173.8'F 1.02 '63.9'F 564F
<.84F 48.3'F
-4.84F 2254F 207'F R. E. Ginna Heatup and Cooldown LimitCurves June 1996
l
.4 J
61 TABLE5-14 Calculation of the ART Values for the 1/4T Location and 40 EFPY Material Method 6RTgpz Margin ART Inter. Shell Forging 125S255 Lower Shell Forging 125P666 Circ. Weld Seam SA-847 RG 1.99, R2, P1.1 RG 1.99, R2, P2.1
'G 1.99, R2, P1.1
.'RG 1.S9, R2, P2.1 RG 1.S9, R2, P1.1 RG 1.99, R2, P2.1 44'F 16.5'F 31OF 28.2'F 170.4'F 160.7'F 1 28 563oF 1.28 21.1'F 1.28 '97'F 1.28 36.1'F 1.28 218.1'F 1.28 205.7'F 34OF 17'F 34OF 17'F 564F 48.3'F 20'F 20OF 40'F 400F A 8OF A 8oF 110'F 58'F 114'F 93'F 269'F 249'F TABLE5-15 Calculation of the ART Values for the 3/4T Location and 40 EFPY Material Method CF FF Margin IRTvpr ART Inter. Shell Forging 125S255 RG 1.99, R2, P1.1 RG 1.99, R2, P2.1 44'F 16.5'F 1.08 47.5'F 1.08 17.8'F 34OF 17'F 20'F 20'F 102'F 55OF Lower Shell Forging 125 P666 Circ. Weld Seam SA-847 RG 1.99, R2, P1.1 31'F RG 1.99, R2, P2.1 28.2'F RG 1.99, R2, P2.1 160.7'F RG 1.99, R2, P1.1 170.4'F 1.08 33.5'F 1.08 30.5'F 1.08 184.0'F 1.08 173.6'F 33 5'F 40OF 17oF 40'F 56'F A.8'F 48.3'F
-4.8'F 107'F 88'F 235'F 217'F 0
R. E. Ginna Heatup and Cooldown LimitCurves'une 1996
,)
The circumferential weld seam SA-847 is the limiting beltline material for all heatup and cooldown curves to be generated.
Contained in Table 5-16 is a summary of the limiting ARTs to be used in the generation of the Ginna reactor vessel heatup and cooldown curves.
TABLE5-16 Summary of the LimitingART Values Used in the Generation of the Ginna Heatup/Cooldown Curves EFPY 1/4 T Limiting ART 3/4 LimitingART 24 232'F l96'F 236'F 204oF 32 241'F 207'F 40 249'F 217'F R. E. Ginna Heatup and Cooldown Limit Curves June 1996
':l~gcgge
\\
63 6
HEATUP AND COOLDOWN PRESSURE-TEMPERATURE LIMITCURVES Pressure-temperature limit curves for normal heatup and cooldown of the primary reactor coolant system have been calculated for the pressure and temperature in the reactor vessel beltline region using the methods'~'iscussed in Section 3.0, 4.0 and 5.0 of this report.
This approved methodology is also presented in WCAP-14040-NP-A, dated January 1996'".
Figures 6-1, 6-3, 6-5 and 6-7 present the heatup curves without margins for possible instrumentation errors using heatup rates of 60 and 100 F/hr. These curves are applicable to 24, 28, 32 and 40 EFPY for the R. E. Ginna reactor vessel.
Additionally, Figures 6-2, 6-4, 6-6 and 6-8 present the cooldown curves without margins for possible instrumentation errors using cooldown rates of 0, 20, 40 60 and 100'F/hr. These curves are also applicable to 24, 28, 32 and 40 EFPY for the R. E. Ginna reactor vessel.
Allowable combinations of temperature and pressure for speciTic temperature change rates are below and to the right of the limitlines shown in Figures 6-1 through 6-8. This is in addition to other criteria which must be met before the reactor is made critical, as discussed below in the following paragraphs.
The reactor must not be made critical until pressure-temperature combinations are to the right of the criticality limit line shown in Figures 6-1, 6-3, 6-5 and 6-7. The straight-line portion of the criticality limitis at the minimum permissible temperature for the 2485 psig inservice hydrostatic test as required by Appendix G to 10 CFR Part 50. "The governing equation for the hydrostatic test is defined in Appendix G to Section XI of the ASME Code'~'s follows:
1><i
<ia (6)
- where, K,
is the stress intensity factor covered by membrane (pressure) stress, K~= 26.78+ 1.233 e ""~"
T is the minimum permissible metal temperature, and RT>>T is the metal reference nil-ductilitytemperature The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in Reference 50. The pressure-temperature limits or core operation (except for low power physics tests) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40'F higher than the minimum permissible temperature in the corresponding pressure-temperature curve for heatup and coo!down calculated as described in Section 4.0 of this report. The vertical line drawn from these points on the pressure-temperature curve, intersecting a curve 40 F higher than the pressure-temperature limitcurve, constitutes the limitfor core operation for the reactor vessel.
Figures 6-1 through 6-8 define all of the above limits for ensuring prevention of nonductile failure for the R. E. Ginna reactor vessel.
The data points for the heatup and cooldown pressure-temperature limit curves shown in Figures 6-1 through 6-8 are presented in Tables 6-1 through 6-8.
R. E. Ginna Heatup and Cooidown Limit Curves June 1996
64 MATERIALPROPERTY BASIS LIMITINGMATERIAL: CIRCUMFERENTIALWELD SA-847 LIMITINGART VALVESAT 24 EFPY:
3/4T, 232'F 3/4T, 196'F 2500 m 2250
~ 2000 688423I080688 LEAK TEST LIMIT
~
I
~
(
(
~
(
(
~
I
~
I 1750 1500 1250 1000 750 500 250 UNACCEPTABLE OPERATION HBATUP RATE UP TO BO F/Hr.
HEATUP RATE UP TO IOO P/Hr.
CRITICALITY LIMIT BASED ON INSERVICE HYDROSTATIC TEST TEMPERATURE (383 P)
POB THE SERVICE PERIOD UP TO 24 '
EPPY
~
~
i
~
(
(
i ACCEPTABLE OPERATION ('
50 100 150 200 250 300 350 400 450 500 Iudicated Temperature (Deg.Fj FIGURE 6-1 R. E. Ginna Reactor Coolant System Heatup Limitations {Heatup Rates of 60 and 100'F) Applicable to 24 EFPY {Without Margins for Instrumentation Errors)
R. E. Ginna Heatup and Cooldown Limit Curves June 1996
65 MATERIALPROPERTY BASIS LIMITINGMATERIAL: CIRCUMFERENTIALWELD SA-847 LIMITINGART VALUES AT 24 EFPY:
1/4T, 232'F 3/4T, 196'F 2500 2250
~ W" 2000 1750 SSS42SI9SQSSS I
I
~
~
I
~
I I
I I
~
UNACCEPTABLE OPERATION
~
~
~
~
1500 1250 1000 750 500 250 COOLDOlfN RATES F/Hr.
0 20 40 00 100 i
~
i I
ACCEPTABLE OPERATION
~
I
~
0
.0 50 100 150 200 250 500 350 400 450 500 Indicated Temperature (Deg.Fj FIGURE 6-2 R. E. Ginna Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 20 40 60 and 100'F/hr) Applicable to 24 EFPY (Without Margins for Instrumentation Errors)
R. E. Ginna Heatup and Cooldown Limit Curves, June 1996
66 MATERIALPROPERTY BASIS LIMITINGMATERIAL: CIRCUMFERENTIALWELD SA-847 LIMITINGART VALUES AT 28 EFPY:
1/4T, 236'F 3/4T, 204'F 2500 588588i252352
~
I I
~
~
I ~,
I 1
~
I 2250
~ W 2000 LBAK TBST LIMIT g
~:
s
~
I i
1750 1500 1 250 1000 HBATUP RATB UP TO 00 P/Hr.
UNACCEPTABLE OPERATION s
I
~
~
ACCEPTABLE OPERATION 750 500 250 0
HBATUP RATE UP TO 100 P/Hr.
CRITICALITY LIXIT BLSED ON INSERVICE HYDROSTATIC TEST TEMPERATURE (357 F)
FOR THE SERVICE PERIOD UP TO 28 0
EFPY 50 100 1'50 200 250 300 350 400 450 500 Indicated Temp.erature (Beg.Fj FIGURE 6-3 R. E. Ginna Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100'F) Applicable to 28 EFPY (Without Margins for Instrumentation Errors)
R. E. Ginna Heatup and Cooldown LimitCurves June 1996
67 MATERIALPROPERTY BASIS LIMITINGMATERIAL: CIRCUMFERENTIALWELD SA-847 LIMITINGART VALUES AT 28 EFPY:
1/4T, 236'F 3/4T, 204 F 2500
~ ~,
I I
I SSSSjjjj42S2SS2
~
I I
~
~
I
~
I
~
I
~
j I
2250
~~
2000 1750 I
j I
~
I I
j I,
I I
~
I I
i
~
I UNACCEPTABLE OPERATION 1500 1250 1000 ACCEPTABLE OPERATION I~,
~
750 500 250 COOLDOlfN RATES F/Hr.
0 20 40 00 100 i:
~
I
~
I 0
I 0
50 100 150 200 250 300 350 400 450 500 Indicated Temperature (Deg.Fj FIGURE 6-4 R. E. Ginna Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 20 40 60 and 100'F/hr) Applicable to 28 EFPY (Without Margins for Instrumentation Errors)
R. E. Ginna Heatup and Cooldown Limit Curves June 1996
68 MATERIALPROPERTY BASIS LIMITINGMATERIAL: CIRCUMFERENTIALWELD SA-847 LIMITINGART VALUES AT 32 EFPY:
1/4T, 241'F 3/4T, 207'F 2500 4542588588425 m 2250
~ pk
~ '2000
~
~
LB AK TB T. L I)t 1250 1500 UN'ACCEPTABLE OPERATION
~
}
1250 1000 HBATUP RATB To 5'0 Z/Hr.
ACCEPTABLE OPERATION 750 500 250 0
0 HBATUP RATB UP TO 100 F/Hr.
CRITICALITY LI)tlT EASED ON INSERVICE HYDROSTATIC TEST TE)IPERATURE (882 F)
FOR THE SERVICE PERIOD UP TO 8' EFPT 50 100 150 200 250 300 350 400 450 500 Indicated Temperature (Deg.F)
FIGURE 6-5 R. E. Ginna Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100'F) Applicable to 32 EFPY (Without Margins for Instrumentation Errors)
R. E. Ginna Heatup and Cooldown Limit Curves June 1996
MATERIALPROPERTY BASIS
(
LIMITINGMATERIAL: CIRCUMFERENTIALWELD SA-847 LIMITINGART VALUES AT 32 EFPY:
1/4T, 241'F 3/4T, 207'F 2500
~
~
.'I i
4S42SSSSSS42S i
i
~
~
I
~
I 2250
~~
2000
~
~
~
I
~
I
~
I I
~
I I
! i i
i I
g
~
1750 I
UNACCEPTABLE OPERATION 1500 1250 1000 750 500 250 c 0 0 L D 0 'rr N RATBS F/Hr.
0 20 40 eo 100 i:
I
~:
s I
~
I
~
ACCEPTABLE OPERATION 0
i 0
50 100 150 200 250 300'50 400 450 500 Indicated Temperature (Deg.F}
FIGURE 6-6 R. E. Ginna Reactor Coolant System Cooldown Limitations (Cooldown Rates of
~
0, 20 40 60 and 100'F/hr) Applicable to 32 EFPY (Without Margins for Instrumentation Errors)
R. E. Ginna Heatup and Cooidown Limit Curves June 1996
70 MATERIALPROPERTY BASIS LIMITINGMATERIAL: CIRCUMFERENTIALWELD SA-847 LIMITINGART VALUES AT 40 EFPY:
1/4T, 249'F 3/4T, 217'F 2500 I
2809525885bbh
~ 2250
~ W
~ 2000 LEAK TE ST L I%IT 1750 UNACCEPTABLE OPERATION 1500 1250 1000 750 500 250 HEATUP RATE UP TO 00 P/Hr.
HBATUP RATE UP TO 100 P/Hr.
CRITICLLITY LIMIT BLEED ON INSERVICE HYDROSTLTIC TEST TEHPERLTURE (870 P)
POR THE SERVICE PERIOD UP TO hb
~ 0 EPPY I
i I
~
ACCEPTABLE OPERATION 0
0 50 100 150 200 250 300 350 400 450 500 Indicated.Temperature (Beg.Fj FIGURE 6-7 R. E. Ginna Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100'F) Applicable to 40 EFPY (Without Margins for Instrumentation Errors)
R. E. Ginna Heatup and Coo!down Limit Curves June 1996
MATERIALPROPERTY BASIS LIMITINGMATERIAL: CIRCUMFERENTIALWELD SA-847 LIMITINGART VALUES AT 40 EFPY:
1/4T, 24S'F 3/4T, 2 I7 F 2500 2250
~~
2000 2599525555554 t
t t
I t
I
~
i t
~
g
~
1750 UNACCBPTABLB OPERATION 1500 1250 s
1000 ACCEPTABLE OPERATION 750 500 250 COOLDOlfN RATES F/Hr.
0 20 40
.50 100 0
0 I
i 50 100 150 200 250 300 350 400 450 500 indicated Temperature (Deg.Fj
'IGURE 6-8 R. E. Ginna Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 20 40 60 and 100'F/hr) Applicable to 40 EFPY (Without Margins for Instrumentation Errors)
R. E. Ginna Heatup and Cooldown Limit Curves June 1996
72 TABLE 6-1 R. E. Ginna 24 EFPY Heatup Curve Data Points Heatup Guv 60F T
60 65 70 75 80 85 90 95 100 105 110 I IS 120 125 130 135 140 145 150 155 160 165 170 175.
180 185 190 195 200 205 210 215 220 225 230
, 235 240 245 250 255 260 265 270 275 280 285 290 295 300 305 310 315 320 325 330 335 340 345 350 355 360 365 370 375 380 es Cri P
507 507 507 507.
507 507 507 508 509 512 515 518 522 526 530 535 541 547 553 560 567 575 584 593 603 614 625 638 651 665 681 697 715 734 754 772 791 811 833 856 881 908 937 968 1002 1038 1077 1118 1163 1211 1262 1312 1364 1420 1479 1543 1612 1685 1764 1848 1938 2035 2138 2248 2365 ticaL Li T
353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 355
. 360 365 370 375 380 385 390 395 400 405 410 415 420 Ilut P
0 525 516 511 508 507 507 508 509 512 SIS 518 522 526 530 535 541 547 553 560 567 575 584 593 603 614 625 638 651 665 681 697 715 734 754 772 791 811 833 856 881 908 937 968 1002 1038 1077 1118 1163 1211 1262 1312 1364 1420 1479 1543 1612
'I685 1764 1848 1938 2035 2138 2248 2365 lOOF T
60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 155 160 165 170 175 180 185 190 195 200 205 210 215 220 225 230 235 240 245
'50 255 260 265 270 275 280 285 290 295 300 305 310 315 320 325 330 335 340 345 350 355 360 365 370 375 380 385 Cri P
478 A78 478 478 478 478 478 478 478 478 478 478 480 482 485 488 491 495 500 505 511 517 524 531 539 548 557 567 578 590 603 616 631 647 664 682 702 723 746 770 796 825 855 887 922 960 1000 1044 1090 1140 1193 1250 1312 1378 1448 1524 1590 16S7
'729 1807 1889 1977 2072 2173 2280 2395 tical. Limit T
353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353 353
'53 353 353 355
'60 365 370 375 380 385 390 395 400 405 410 415 420 425 Test Limit T
P 296
)500 Leak P
0 524 511 501 493 487 483 480 478 478 478 478 480 482 485 488 491 495 500 505 511 517 524 531 539 548 557 567 578 590 603 616 631 647 664 682 702 723 746 770 796 825 855 887 922 960 1000 1044 1090 1140 1193 1250 1312 1378 1448 1524 1590 1657 1729 1807 1889 1977 2072 2173 2280 2395 304 1600 315 1750.
322 1850 330 2000 353 2485 R. E. Ginna Heatup and Cooldown Limit Curves June 1996
73 TABLE6-2 R. E. Ginna 24 EFPY Cooldown Curve Data Points Steady Stat T
60 65 70 75 80 85 90 95 100 105 110 115 120 125 130
~
135 140 145 150 155 160 16$
170 175 180 185 190 195 200 20$
210 215 220 225 230 235 24$
250 255 260 265 270 275 280 28S 290 295 300 30$
310 315 320 32$
330 335 340 345 350 355 360 36$
370
'75 P
540 S42 544 545 547
$49 SSI 554 556 559 562 565 568 572 S76 580 584
$89 594 599 605 612 619 626 634 642 6$2 661 672 683 696 709 723 738 754 772 791 811 833 856 881 908 937 968 1002 1038 1077 1118 1163 1211 1262 1318 1377 1440 15ts 1581 1660
'744 1834 1931 2034 2144 2262 2388 20F T
60 65 70 75 80 85 90 95 100 105 I IO 115 120 125 130 135 140 145 150 155 160 16$
17D 175 180 185 190 195 200 205 21D 215 220 225 23D 235 240 245 250 255 260 26$
270 275 280 285 290 295 300 30$
P 515 516 518 519 521 523 525 528 530 533 536 539 542
$46 550 554 558 563 568
$74 580 587 594 601 609 618 628 638 649 661 673'87 701 717 734 752 772 793 816 840 866 89$
925 957 993 1030 1070 1114 1160 1211 40F T
60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 155 160 165 170 175 180 185 190 195 200 20$
210 215 220 225 230 235 240 245 250 255 260 265 270 275 280 285 290 295 300 P
489 490 492 493 495 497 499 501 504 506 509 512 516 519 523 528 532 537 542 S48 554 561 568
$76 585 594 603 614 625 637 651 665 680 696 714 733 753 799 825 852 881 913 947 983 1023 106$
1110 1159 T
P 60, 462 65 ~
463 465 466 468 470 472 474 477 70 75 80
&5 90 95 100 120 125 130 135 140 145 150 155 160 165 170 175 180 185 190 195 200 205 210 215 220 225 230 235 240 245 250 255 RO 265 270 275 280 285 290 295 300 489 493 497
'01 506 511 516 S22 529 S36 543 551 S60 569 579 S90 602 615 628 643 659 676 694 714 73$
7$8 783 809 838 869 902 937 975 1016 1061 1108 1159 105.
480 110
. 483 115 486 IODF T
60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 l40 145 150 155 160 165 170 175 180 185 190 195 200 20$
'10 215 220 225 230 235 240 245 250 255 260 265 270 275 280 2I5 290 295 P
408 48 410 412 414 415 418 420 422 425 428 431 435 438 442 447 452 457 463 469 476 483 491 500 S09 519 S30 542 554 568 583 599 616 634 654 676 699 751 780 811 845 881 920 962 1007 1056 1108 R. E. Ginna Heatup and Cooldown Limit Curves June 1996
74 TABLE 6-3 R. E. Ginna 28 EFPY Heatup Curve Data Points Heatup Curv 60F 60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 155 160 165 170 175 180 185 190 195 200 205 210 215 220 225 230 235 240 245 250 255 260 265 270 275 280 285 290 295 300 305 310 315 320 325
'30 335 340 345 350 355 360 365 370 375 380 385 C
P 502 502 ritical. Li T
357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357
~ 357 357 357 357 357 357 357 357 357 357
~357 357 357 357 357 357 357 357 357 357 357 357 3S7 357 360 365 370 375 380 385 390 395 400 405 410 415 420 425 502 502 502 503 504 511 515 518 522 527 532 537 542 555 570 578 587 596 618 629 642 656 671 686 703 741 763 785 810 836 861 887 914 944 975
)009 1046 1085 1127 1172 1221 1273 1322 1375 143) 1492 1557 1626 1701 1780 1865 1957 2054 2159 2270 2389 m)t P
0 521 513 507 504 502 502 503 504 506 509 511 515 518 522 527 532 537 542 548 555 562 570 578 587 596 607 618 629 642 656 671 686 703 722 741 763 785 810 836 861 887 914 944 975
)009 1046 1085 1127 1172 1221 1273 1322 1375 1431 1492 1557 1626 1701 1780 1865 1957 2054 2159 2270 2389 100F T
60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 ISS 160 165 170 175 180 185 190
)95 200 205 210 215 220 230 235 240 245 250 255 260 265 270 275 280 285 290 295 310 3)5 320 325 330 335 340 345 350 355 360
- 365, 370 375 380 385 390 Cri P
472 472 472 472 472 472 472 472 472 472 472 473 474 476 478 480 483 487 491 496 500 506 512 518 526 533 541 550 560 570 582 594 607
. 621 636 652 670 688 709 730 754 779 806 835 866 899 935 974 1015 1059 1107 1158 12)3 1272 1335 1402 1475 1552 1635 1724 1819 1906 1996 2091 2193 2302 2419 tical. Umit T
357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 357 3S7 357 357 360 365 370 375 380 385 390 395 400 405 410 415 420 425 430 P
0 521 508 497 489 483 479 476 474 473 472 473 474 476 478 480 483 487 491 496 500 506 512 518 526 533 541 550 560 570 582 594 607 621 636 652 670 688 709 730 754 779 806 835 866 899 935 974
- 10) 5 1059
)107 1158 1213 1272 1335 1402 1475 1552 1635 1724 1819 1906 1996 209) 2193 2302 24)9 Leak Tert Limit T
P 300 1500 308 1600 319 1750 326 1850 334 2000 357 2485 R. E. Ginna Heatup and Cooldown Limit Curves June 1996
75 TABLE 6-4 R. E. Ginna 28 EFPY Cooldown Curve Data Points C
P 539 541 542 544 546 547 550 552 554 557 559 562 565 569 572 576 581 585 590 595 601 607 613 620 628 635 644 653 664 674 686 698 712 726 741 758 776 795 815 837 861 887 914 944 975 1009 1046 1085 1127 1172 1221 1273 1329 1389 1454 1523.
1597 1676 1762 1853 1951 2055 2167 2286 2414 Cooldown C Steady Set T
60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 155 160 165 170 175 180 185 190 195 200 205 210 215 220 225 230 235 240 245 250 255 260 265 270 275
. 280 285 290 295 300 305 310 315 320 325 330 335 340 345 350 355 360 365 370
'75 380 20F 60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 ISS 160 165 170 175 180 185 190 195 200 205 210 215 220 225 230 235 240 245 2$0 255 260 265 270 275 280 285 290 295 300 305 P
513 51$
516 518 520 522 524 526 528 T
60 6$
70 75 80 85 90 95 100 539 543 546 550 555 559 564 569 575 581 588 595 603 611 620 630 651 663 676 720 738 756 776 797 820 845 900 931 964 1000 1038 1079 1123 1170 120 125 130 135 140 145 150 155 160 165 170 175 180 185 190 195 205 210 215 220 235 240 245 250 255 260 265 270 275 280 285 290 295 300 305 531 105 533 110 536 115 P,
487 489 490 492 493 495 497 499 T
60 65 70 75 80
&5 90 95 510 513 516 520 SD 528 533 538 543 549 556 HS 120 125 130 135 140 145 ISO 155 160 165 562 170 570 578 586 595 605 616 628 718 737 758 780 804 830 858 887 919 954 991 1031 1074 1120 1170 175 180 185 190 195 210 215 220 230 240 245 250 2$5 260 265 270 275 280 285 290 295 300 305 501 100 504 105.
507 110 P
461 462 463 465 467 468 470 472 475 477 480 483 486 lOOF T
60 65 70 75 85 90 95 100 105 110 115 120 506 512 517 523 530 537 544 552 561 571 581 617 631 662 679 698 718 739 763 788 815 844 875 908 944 983 1025 1070 1118 1170 145 150 155 160 165 170 175 180 185 190 195 205 210 215 220 225 230 235 240 245 250 255 260 265 270 275 280 285 290 295 300 490 125 493 130 497 135 410 412 414 415 417 420 422 425 428 431 435 439 443 447 453 458 464 470 477 485 493 501 511 521 532 544 557 S71 586 602 619 638 658 680 704 729 756 786 818
'52 888 928 971 1016 1066 1119 R. E. Ginna Heatup and Cooldown Limit Curves June 1996
76 TABLE 6-5 R. E. Ginna 32 EFPY Heatup Curve Data Points Hcatup C 60F T
60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 155 160 165 170 175 180 185 190 195 200 205 210 215 220 225 230 235 240 245 250 255 260 265 270 275 280 285 290 295 300 305 310 315 320.
325 330 335 340 345 350 355 360 365 370 375 380 385 390 Cri P
501 501 501 501 501 501 501 501 502 504 507 509 512 516 520 524 529 533 539 545 551 558 565 573 581 590 600 611 622 634 647 661 677 693 710 729 749 771 795 815 837 861 S87 914 944 975 1009 1046 1085 1127 1172 1221 1273 1322 1375 1431 1492 1557 1626 1701 1780 1865 1956 2054 2159 2270 2389 tical. Li T
362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362
'62-362 362 362 362 362 362 362 362 365 370 375 380 385 390 395 405 410 415 420 425 430 mtt P
0 520 512 506 501 501 501 512 516 520 533 539 551 558 565 573 581 611 622 634 647 661 677 693 710 729 749 771 795 815 837 861 887 914 944 975 1085 1127 II72 1221 1273 1322 1375 1431 1492 1557 1626 1701 1780 1865 1956 2054 2159 2270 2389 100F T
60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 155 160 165 170 175 180 185 190 195 200 205 210 215 220 225 230 235 240 245 250 255 260 265 270 275 280 285 290 295 300 305 310 315 320 325 330 335 340 345 350 355 360 365 370 375 380 385 390 395 Cri P
471 471 471 471 471 471 471 471 471 471 471 471 472 473 475 478 481 484 488 492 497 502 508 514 521 528 536 544 554 564 574 586 598 612 626 642 659 677 696 717 739 763 789 817 846 878 913 950 989 1032 1077 1126 1179 1235 1295 1360 1430 1504 1583 1669 1744 1822 1906 1995 2091 2193 2302 2418 tical. Limit 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 362 365 370 375 380 385 390 395 400 405 410 415 420 425 430 435 P
0 520 507 496 488 482 477 474 472 471 471 471 472 473 475 478 481 484 488 492 497 502 508 514 521 528 536 544 554 564 574 586 598 612 626 642 659 677 696 717 739 763 789 817 846 878 913 950 989 1032 1077 1126 1179 1235 1295 1360 1430 1504 1583 1669 1744 1822 1906 1995 2091 2193 2302 2418 Leak Test Limit T
P 305 1500 313 1600 324 1750 331 1850 339 2000 362 2485 R. E. Ginna Heatup and Cooldown Limit Curves June 1996
77 TABLE 6-6 R. E. Ginna 32 EFPY Cooldown Curve Data Points Steady Stat T
60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 155 160 165 170 175 180 185 190 195 200 205 210 215 220 225 230 235 240 245 250 255 260 265 270 275 280 285 290 295 300 305 310
,315 320 325 330 335 340 345 350 355 360 365 370 375 380 385 P
538 539 541 542 544 546 547 550 552 554 557 559 562 565 569 572 576 581 585 590 595 601 607 613 620 628 635 644 653 664 674 686 698 712 n6 741 758 776 795 815 837 861 887 914 944 975 1009 1046 1085 1127 1ln 1221 1273 1329 1389 1454 1523 1597 1676 1762 1853 1951 2055 2167 2286 2414 Cooldown Guv 20F T
60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 155 160 165 170 175 180 185 190 195 200 205 210
'15 220 225 230 2'40 245 250 255 260 265 270 275 280 285 290 295 300 305 310 P
512 513 515 516 518 520 521 523 526 528 530 533 536 539 543 546 550 554 559 564 569 575 581
'88 595 603 611 620 629 640 651 663 676 689 704 720 737 756 776 797 820 845 872 900 931 964 1000 1038 1079 1123 1170 T
60 65 70 75 80
&5 90 95 100 105 110 115 120 125 130 135 140 145 150 ISS 160 165 170 175 180 185 190 195 200 205 210 215 220 225 230 235 240 245 250 255 260 265 270 275 280 285 290 295 300 305 310
'P 486 487 488 490 491 493 495 497 499 501 504 506 509 513 516 520 524 528 533 538 543 549 555 562 570 578 586 595 605 616 627 640 653 667 683 699 717 737 757 780 804 829 857 887 919 954 991 1031 1074 1120 1169 IOOF T
60 65 70 75 80 85 90 95 100 463 465 466 468 470 472 75 80
&5 90 95 100 105 110 115 120 474 477 480 483 105.
, 110 115 120 125 130 486 489 125 130 135 140 145 150 ISS 160 165 170 175 180 185 190 195 200 205 210 215 220 225 230 235 240 245 250 255 260 265 270 275 280 285 290 295 300 305 310 135 493 145 150 155 160 165 170 175 180 185 190 195 200 501 506 Sll 517 523 529 536 544 552 561 571 581 205 210 215 220 617 630 645
- 225, 661 230 679
. 235 697 240 245 250 255 260 265 270 275 280 285 290 295 300 30S 718 739 762 787 814 843 875 908 944 983 1025 1070 1118 1170 T
P 60 459 65 460 70 462 P
405 406 407 408 410 411 413 415 417 419 422 424.
427 431 434 438 442 447 452 457 463 470 476 484.
492 501 510 520 531 543 556 570 585 601 619 637 658 680 703 729 756 785 817 851 888 928 970 1016 1066 1119 R. E. Ginna Heatup and Cooidown Limit Curves June 1996
78 TABLE 6-7 R. E. Ginna 40 EFPY Heatup Curve Data Points Heatup C 60F T
60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 155 160 165 170 175 180 185 190 195 200 205 210 215 220 225 230 235 240 245 250 255 260 265 270 275 280 285 290 295 300 305 310 315 320 325 330
'335 340 345 350 355 360 365 370 375 380 385 390 395 400 ritical. Li C
P 496 496 496 496 496 496 496 496 497 498 500 503 505 508 511 515 519 523 528 533 538 544 550 557 564 572 580 589 599 610 621 633 646 660 676 692 709 728 748 770 793 819 845 871 897 925 956 988 1023 1061 1101 1145 1191 1241 1293 1343 1397 1455 1517 1584
)655 1732 1813 1901 1995 2095 2202 2316 2438 T
370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 375 380 385 390 395 400 405 4)0 415 420 425 430 435 440 mtt P
0 516 508 502 498 496 496 496 497 498 500 503 505 508 Sll 515 5)9 523 528 533 538 544 550 557 564
~
572 580 589 '
599 610 621 633 646 660 676 692 709 728 748 770 793 819 845 871 897 925 956 988 1023 1061 110) 1145 1191 1241 1293 1343 1397
)455'517 1584 1655 1732
) 813 190) 1995 2095 2202 2316 2438 T
60 65 70 75 80 85 90 95 100 105 110 115 120
)25 130 135 140 145 150 155 160 165 170 175 180 185 190 195 205 210 215 220 225 230 235 240 260 265 270 275 280 285 290 310 315 320 325 330 335 340 345 350 355 360 365 370 375 380 385 390 395 400 405 Cri P
465 465 465 465 465 465 465 465 465 465 465 465 466 467 46&
.470 472 475 478 482 486 490 495 500 506 512 519 527 535 543 552 562 573 585 597 610 625 640 657 675 694 715 737 761 787 814 844 876 9)0 947 987 1029 1074
)123 1176 1232 1292 1356 1426 1500 1579 1664 1755 1852 1940 2032 2130 2235 2346 2466 tical. Limit T
370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 370 375 380 385 390 395 400 405 410 415 420 425 430 435 440 445 P
0 516 503 492 484 478 473 470 467 466 465 465 466 467 468 470 472 475 478 482 486 490 495 500 506 512 519 527 535 543 552 562 573 585 597 610 625 640 657 675 694 715 737 76) 787 814 844 876 910 947 987 1029 1074 1123 1176 1232 1292 1356 1426 1500.
1579 1664 1755 1852 1940 2032 2130 2235 2346 2466 Leak Test Limit T
P 313 1500 321
)600 332 1750 339 1850 347 2000 370 2485 R. E. Ginna Heatup and Cooldown Limit Curves June 1996
79 TABLE 6-8 R. E. Ginna 40 EFPY Coo!down Curve Data Points Cooldown C Steady Stat T
60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 155 160 165 170 175 180 185 190 195 200 205 210 215 220 225 230 235 240 245 2$0 255 260 265 270 275 280 285 290 295 300 305 310 315 320 325 330 335 340 345 350 355 360 365 370
'375 380 385 390 395 P
536 537 538 541 543 545 546 548 550 553 555 558 560 563 567 570 574 578 582 587 592 597 603 609 616 623 631 639 648 657 668 679 691 703 717 732 748 765 783 803 824 847 871 897 925 956 988 1023 1061 1101 1145 1191 1241 1295 1352 1414 1481 1552 1628 1710 1797 1891 1992 2099 2214 20F T
60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 155 160 165 170 175 180 185 190 195 200 205 210 215 220 225 230 235 240 245 250 255 260 265 270 275 280 285 290 295 300 305 310 315 320 P
510 511 512 514 515 517 518 520 522 524 526 529 531 534 537 540 544
$48 552 556 561 566
$71.
577 584 590 598 606 614 623 633 644 655 668 681 695 710 727 745 764 784 806 830 855 883 912 944 978 1015 1054 1096 1141 1190 40F T
60 65 70 75 80 85 90 95 100 105 110 115 120 125 130 135 140 145 150 ISS 160 165 170 175 180 185 190 195 200 205 210 215 220 225 230 235 240 245 2$0 255 260 265 270 275 280 285 290 295 300 305 310 315 320 P
484 485 486 487 490 492 493 495 497 499 502 504 507 510 514 517 521 525 530 534 540 545 551 558 565 572 581 589 599 609 620 632 645 658 673 689 706 725 744 766 789 814 840 869 900 933 968 1048 Ib92 1139 1190 T
60 65 70 75 80 85 90 95 100 105.
,. 110 115 120 125 130 140 145 150 155 160 165 170 175 180 185 190 195 200 205 210 215 220 225 230 235 240 245 2$0 255 260 265 270 275 280 285 290 29$
300 305 310 315 P
457 458 459 460 462 463 465 466 468 470 472 475 477 480 483 487 490 494 498 503 508 513 519 525 532 539 547 555 564
$74 585
$96 608 622 636 651 668 686 705 726 748 772 798 825 855 887 922 959 999 1042 1089 H38 lOOF 60 65'0 75 P
403 403 405 95 100 105 110 115 120 411 413 415 417 419 422 145 150 155 160 165 170 175 180 185 190 195 200 205 210 215 220 225 230 235 240 245 250 255 260 265 270 275 280 285 290 295 300 30$
310 315 439 443 448 453 459 471 478 486 494 504 513 524 535 547 561 575 591 607 625 645 666 688 713 739 767 797 830 865 903 944 988 1035 1086 1141 125 425 130 428 135 431 140 '35 R. E. Ginna Meatup and Cooldown Limit Curves June 1996
80 REFERENCES Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials",
U.S. Nuclear Regulatory Commission, May, 1988.
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Andrachek, et al., January 1996.
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RSIC Computer Code Collection CCC-543, "TORT-DORT Two-and Three-Dimensional Discrete Ordinates Transport, Version 2.7.3", May 1993.
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W. L. Baldewicz, et. al., "The Nuclear Design and Core Management of the R. E. Ginna Nuclear Reactor
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westinghouse Proprietary Class 2]
10 W. L. Baldewicz, et. al., "The Nuclear Design and Core Management of the R. E. Ginna Nuclear Reactor
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pNestlnghouse Proprietary Class 2]
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PNestinghouse Proprietary Class 2]
I 13 M. A. Mann, "The Nuclear Design and Core Management of the R. E. Ginna Nuclear Reactor - Cycle 5", WCAP-8514, March 1975.
PNestinghouse Proprietary Class 2]
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81 15 1
C. Watarumi, et. al., "The Nuclear Design and Core Management of the R. E. Ginna Nuclear Reactor
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pNestinghouse Proprietary Class 2]
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S. Srinilta, et. al., "The Nuclear Design and Core Management of the R. E. Ginna Nuclear Reactor
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23 24 D. J. Krieg, et. al., "The Nuclear Design and Core Management of the R. E. Ginna Nuclear Reactor
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PNestinghouse Proprietary Class 2]
/
S. Srinilta, et. al., "The Nuclear Design and Core Management of the R. E. Ginna Nuclear Reactor
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- Cycle 25", WCAP-14290, April 1995.
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R. E. Ginna Heatup and Cooldown Limit Curves June 1996
82
'1 ASTM Designation E482-89, "Standard Guide for Application of Neutron Transport Methods for Reactor Vessel Surveillance", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1993.
ASTM Designation E560-84, "Standard Recommended Practice for Extrapolating Reactor Vessel Surveillance Dosimetry Results", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1993.
32 ASTM Designation E693-79, "Standard Practice for Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom (dpa)", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1993.
33 ASTM Designation E706-87, "Standard Master Matrix for Light-Water Reactor Pressure Vessel Surveillance Standard", in ASTM Standards,'Section 12, American Society for Testing and Materials, Philadelphia, PA, 1993.
ASTM Designation E853-87, "Standard Practice for Analysis and Interpretation of Ught-Water Reactor Surveillance Results", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1993.
35 II ss37'STM Designation E261-90, "Standard Practice for Determining Neutron Fluence Rate, Fluence, and Spectra by Radioactivation Techniques", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1993.
ASTM Designation E262-86, "Standard Method for Determining Thermal Neutron Reaction and Fluence Rates by Radioactivation Techniques", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1993.
ASTM Designation E263-88, "Standard Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Iron", in ASTM.Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1993.
38 39 ASTM Designation E264-92, "Standard Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Nickel", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1993.
ASTM Designation E481-92, "Standard Method for Measuring Neutron-Fluence Rate by Radioactivation of Cobalt and Silver", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1993.
40 ASTM Designation E523-87, "Standard Test Method for Measuring Fast-Neutron Reaction Rates by Radioactivation of Copper", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1993.
41
~(
42 ASTM Designation E704-90, "Standard Test Method for Measuring Reaction Rates by Radioactivation of Uranium-238", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1993.
ASTM Designation E705-90, "Standard Test Method for Measuring Reaction Rates by Radioactivation of Neptunium-237", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1993.
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,83 43 45 46 47 48 49 ASTM Designation E1005-84, "Standard Test 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, 1993.
F. A. Schmittroth, "FERRET Data Analysis Core", HEDL-TME79-40, Hanford Engineering Development Laboratory, Richland, WA, September 1979.
W. N. McElroy, S. Berg and T. Crocket, "A Computer-Automated Iterative Method of Neutron Flux Spectra Determined by Foil Activation", AFWL-TR-741, Vol. I-IV,Air Force Weapons Laboratory, Kirkland AFB, NM, July 1967.
RSIC Data Library Collection DLC-178, "SNLRMLRecommended Dosimetry Cross-Section Compendium", July 1994.
EPRI-NP-2188, "Development and Demonstration of an'Advanced Methodology for LWR Dosimetry Applications", R. E. Maerker, et al., 1981.
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MSE-REME-0280, "R. E. Ginna Heatup/Cooldown Curves at 20 EFPY for Normal Operation", Letter from P. A. Grendys (W) to Mr George Wrobel (RGE), Dated April 26, 1996.
O 50.
52 53 1989 Section III, Division 1 of the ASME Boiler and Pressure Vessel Code, Paragraph NB-2331, "Material for Vessels".
WCAP-13902, "Analysis of Capsule S from the Rochester Gas and Electric Corporation R.
E. Ginna Reactor Vessel Radiation Surveillance Program", Dated December 1993.
"R. E. Ginna Nuclear Power Plant RCS Pressure and Temperature Limits Report",
Revision 1, Fax from Mr. Ron Jaquin (RGE) to Mr E. Terek ~, Dated April 17, 1996 WCAP-7924-A, "Basis for Heatup and Cooldown Limit Curves", W. S. Hazelton, et al.,
April 1975.
RGE-96-503, "Your P.O. 0 CP-64647-C-RD, Chemical Analysis of Eight Weld Metal Samples", from Mr. S. M. Sconce (W) to Mr Ron Jaquin (RGE),Dated February 27, 1996.
R. E. Ginna Heatup and Cooldown Limit Curves June 1996
~I'
~
I
~
Attachment IX Response to NRC Questions Concerning Previous LTOP Analysis (Revised from that provided in Enclosure 3 to June 3, 1997 letter to NRC)
Provide a discussion of the dynamic and static head effects and how these were accounted for in the analyses for the mass addition and heat addition cases.
For the dynamic head effect, consider the effect of all RCPs and RHR pumps that are allowed to operate. Ifyou are limiting operation of such pumps in the LTOP
'egion, provide a discussion of such controls.
The RELAP model used for the analysis automatically accounts for dynamic and static head effects.
The pressure sensors for the LTOPs actuation are taken from the RCP suction leg volumes, consistent with their location in the plant.
For Appendix G pressure limits the pressure is taken at the lowest downcomer node in the reactor vessel.
This results in the highest pressure in the reactor vessel and is therefore conservative.
For RHR pressure, the highest pressure in the RHR system is used (pump discharge).
Cases chosen for the LTOPs analysis utilize running pumps consistent with the operating conditions, and conservatisms as described below.
Mass Addition Cases For the mass addition cases, two conditions bound all operating conditions:
SI Pump injecting with a 1.1 in'ent on the RCS, no RCPs running.
3 Charging Pumps injecting with both RCPs running.
In all three mass addition cases, the RHR system is considered isolated.
Isolation of the RHR system is conservative in that the RHR relief valve (designed to handle the full flowrate of 3 charging pumps) is isolated, and isolating the RHR system provides a smaller volume for the injected mass to expand into, thus resulting in a higher RCS pressure.
Case 1 is run at 60'F and 212'F, bounding the conditions for which a vent can be established in the RCS.
RCPs cannot be run when the RCS is vented.
Case 2 is run for RCS temperatures above 60'F, for cases when protection is provided by the pressurizer PORVs.
Results of these cases are attached.
Heat Addition For the heat addition cases, several scenarios were evaluated in order to determine the limiting cases, allowing adequate conservatism but not being unreasonably conservative which could result in unnecessary operation restrictions, as discussed below.
From an Appendix G standpoint not attaching the RHR system results in the most limiting pressure transient, since the RHR system would remove some of the heat being added from the start of an RCP. It was, however, necessary to model the RHR system in order to demonstrate adequate protection of the RHR system from overpressurization (assuming shutoff head for the RHR pumps for all cases while conservative would result in undesirable operating constraints).
In order to ensure bounding conservatism for Appendix G, the heat exchangers were modeled as being inactive.
In order to determine the number of running RHR pumps, several sensitivities were run to determine minimum and maximum flowrates within the constraints of cooldown limits, considering both minimum and maximum decay heat removal requirements.
The results indicate that for RCS temperatures below 280'F two RHR pumps may be in operation, above 280'F one pump would be in operation.
Therefore the analyses assumes two pumps running for cases below 280'F, and one pump running above 280'F. The impact of these conditions on the transient is not significant since the pump curves for the RHR pumps is relatively flat.
Additionally, it should be noted that the heat addition case was not considered in the original design basis for overpressurization protection of the RHR system.
Inadvertent start of an RCP takes several distinct actions'y the operator which were not considered credible.
Therefore, consideration of this transient for RHR
~
~
~
~
~
~
~
overpressure protection is in and of itself added conservatism.
2.
Justify the use of 85'F as the limiting low temperature for LTOP analyses (both heat and mass addition), in light of your curves extending to 60'F; or provide analyses at the lower value.
Transients were run at 85'F due to the fact that until recently curves were only available down to 85'F.
Additional cases have been run at 60'F.
For the heat addition cases utilizing a higher temperature for the initiating temperature results in a higher peak pressure.
This is due to the fact that the change in density of water per degree fahrenheit increases as the temperature increases.
For the mass addition case a
lower temperature is more limiting, due to the higher density of the RCS liquid.
Again, the change in density is less that 0.5% and willhave a negligible impact.
Results of these cases are attached.
e I
Justify the use of 430 psig for PORV liftsetpoint in the analysis.
List the parameters that may affect this setpoint (e.g., Instrument uncertainty and others) and show that the proposed limitof 411 psig is sufficient to protect the Appendix G curves.
The analysis assumes a setpoint of 430 psig. It demonstrates, with this assumed setpoint, that all limits are met (with margin).
The instrument uncertainty for the actuation channels has been calculated to be 16.95 psig.
When added to the proposed 411 psig limit, an acceptable analytical value of ) 427.95 psig is arrived at.
The analyzed 430 psig setpoint is therefore conservative.
The submittal indicates that instrument uncertainty need not be accounted for in Appendix G analyses.
This in incorrect.
Revise methodology to correct this statement.
The methodology document states that instrument uncertainty is accounted for (page 3.3 item q).
The initial analysis done utilizing the methodology did account for instrument uncertainty, however a statement in the calculation was made that it did not need to be accounted for for Appendix'G cases.
This was consistent with the previously approved methodology.
The current methodology requires'instrument uncertainty to be accounted for and therefore all future analyses will account for it.
No revision to the methodology is required.
Justify the use of the RHR system for cooling in the heat addition analysis.
Discuss the effect of this configuration on the analysis in terms of differences in peak pressures that would be obtained had the RHR system been assumed not operating.
See response to question 1.
No credit is taken for cooling from the RHR system.
You assumed no RCS flow in the mass addition analysis.
You must account for dynamic head effects for this analyses from both the RCPs and RHR pumps that may be operating.
Did you account for the dynamic head effect separately?
See discussion in response to question 1.
In Section 6.2 you state that in the future you can credit the instrument uncertainty ifneeded.
Instrument uncertainty must be accounted for.
Therefore, this statement is incorrect.
Correct this statement in your methodology.
See discussion under question 4.
Instrument uncertainty is included in the methodology and must be accounted for in all future analysis.
l
TS 3.5.3, "ECCS-MODE 4," requires that one train of ECCS be operable in Mode 4. This appears to be in conflict with the LTOP TS 3.4.12 which required that all SI pumps be incapable of injecting into the RCS.
Explain how Ginna meets these two TSs when in the LTOP region and in Mode 4.
See last paragraph of LCO bases for LCO 3.5.3 (page B3.5-26).
Our ECCS requirements specify the capability of injection within 10 minutes which specifically addresses the need to place the SI pumps in pull-stop for LTOP.
9.
TS 3.4.12, "Low Temperature Overpressure Protection (LTOP) System," allows an SI pump to be capable of injecting into the RCS ifthe RCS is depressurized and an RCS vent of 1.1 square inches is established.
Provide the analysis for this configuration with the new P/T curves.
Calculation of the pressure limitfor the limiting mass addition cases when protection is provided by a 1.1 in'ent are provided in the updated LTOP analysis and results are attached.
10.
Justify the use of Table 3 for RCP start profile and 1 second opening time for the PORV.
Acceleration times for the RCPs are dependent upon pump loading, electrical system impedance, and voltage levels.
For the LTOP heat addition cases a faster accelerating time results in more limiting transients since the heat addition occurs over a shorter time period.
A computer model (ETAP) was developed and benchmarked based on vendor pump performance curves and actual current, voltage, and power measurements taken during RCP starts during the 1991 and 1994 outages.
Conservatisms employed. in the model include:
Pum Loadin 1.
The density of RCS fluid is taken at 350'F (lowest density) for all cases.
This minimizes load and therefore results in a faster acceleration.
2.
Flowrates were maximized in the RCS.
Again, this decreases load and results in a faster acceleration.
S stem Im cdance 1.
The offsite power impedance varies with changes in transmission system configuration.
For the analysis, the configuration with the least impedance was used.
A conservatively low impedance for this configuration was used.
S stem Volta e With Ginna Station off-line, the voltage level on the 4160V busses is reduced.
A conservatively high voltage above measured voltages was chosen (4300V).
Utilizing the above conservatisms computer runs were made for all RCPs (A, B, Spare) with acceleration times ranging from 17.4 to 18.6 seconds.
For the LTOPs analysis the fastest acceleration, 17.4 seconds was chosen.
For the PORV, a stroke time of 1 second is chosen in the analysis.
The acceptance criteria in periodic test procedure PT-2.6.5-SD specifies a maximum stroke time of.
0.62 seconds.
Actual stroke times recorded during the test have been consistently on the order of 0.5 seconds or less. Therefore, the analytical stroke time is conservative.
0
~ '
LTOP Cases Summary of Results Type/Case Mass Addition SI pump SI pump 3 Charging RCS Temp, 'F 60 212 60
¹ of RCPs running Limit":,
psia 608.7 780.3 608.7 RV Peak, psia 413.5 396.7 587.4 RHR limit, psia 674.7 674.7 674;7 RHR Peak, psia 542.3 525.9 664.0 Heat Addition RCP Start RCP Start RCP Start 60 280 320 1 start 1 start 1 start 608.7 1116.9 1529.4 551.3 569.3 563.8
~
674.7 674.7 674.7 650.0 663.7 655.7 110% of Appendix G limitper code case N-514.
i~
560 520 4SO 440 400 360 FIGURE 7 CASE 2 MASS ADDITIONCASE PRIMARYTEMPERATURE 60'F PRIMARYPRESSURE 329.7 PSIA
< C}-}ARGINGPUMP ONE RC PUMP RUNNING 3 GPM SEAL LEAKAGE ll O
U Oa G) 9)
320 0
IO 20 30 40 TIME IN SECONDS 50 60 70 SO
480 P
400 20U 320 FIGURE 8 CASE 3 MASS ADDITIONCASE PRIMARYTEMPERATURE GPoF PRIMARYPRESSURE 14.7 PSIA 1 Sl PUMP STARTED NO RC PUMP RUNNING 1.1 SQ.INCH VENT NO SEAL LEAKAGE rl 0D O'a CD 9) 240 160 80 0
0 50 TIME IN SECONDS l25 150
i J
3 l.
4SO 400 320 FIGURE 9 CASE 3 MASS ADDITIONCASE PRIMARYTEMPERATURE 6PoF PRIMARYPRESSURE 14.7 PSIA q SI PUMP STARTED NO RC PUMP RUNNING
$ SQ.INCH VENT NO SEAL LEAKAGE n
O Cla 1
CD 01 160 80 0
0 50 100 125 TIME IN SECONDS 150 175 1
'44 CD CD
480 400 320 FIGURE 10 CASE 4 MASS ADDITIONCASE PRIMARYTEMPERATURE 2'I2oF PRIMARYPRESSURE 14.7 PSIA
. ) St PUMP STARTED NO RC PUMP RUNNING
).q SQ.INCH VENT NO SEAt LEAKAGE Tl O
0U CD Q) 160 80 0
0 50 TIME IN SECONDS 150 175 CD I
tv 44 CO.
fu C)
I C) 44
480 400 320 240 160 FIGURE 11 CASE 4 MASS ADDITIONCASE PRIMARYTEMPERATURE 212'F PRIMARYPRESSURE 14.7 PSIA 1 Sl PUIUIP STARTED NO RC PUMP RUNNING 1.1 SQ.INCH VENT NO SEAL LEAKAGE n
OD 0
0U (D
80 0
0 50 75 100 TIMEIN SECONDS 150 175 Co fu CO fu CD
)
C)
$60 FIGUR CASE 5 HEATADDITIONCASE PRIMARYTEMPERATURE 60oF PRIMARYPRESSURE 329.7 PSIA NO VENT, NO Sl, NO CHARGING PUMP ONE RC PUMP STARTED 2000 GPM RHR l1 0
0D CD Dl 0
2O 0
520 480 440 400 360 320 20 40 60 Transient time in secs 80 CO 0
I CD I
C)
Lpt
0
~\\
4
~\\
c4 680 640 600 560 S20 480 FIGURE 13 CASE g
. HEATADDITIONCASE PRIMARYTEMPERATURE 60'F PRIMARYPRESSURE 329.7 PSIA NO VENT, NO Sl, NO CHARGING PUMP ONE RC PUMP STARTED 2000 GPM RHR LEGEND
~
RHR-PUMP A
~
RHR-PUMP B Tl O
OD 0
rD 440 20 40 60 80 Transient time in secs
FIGURE 14 CASE g HEATADDITIONCASE PRIMARYTEMPERATURE 60'F PRIMARYPRESSURE 329.7.PSIA NO VENT, NO Sl, NO CHARGING PUMP ONE RC PUMP STARTED 2000 GPM RHR
~I O
Q 0A l04 96 88-80 72-64 S6 0
20 40 60 Transient time in secs 80 100 Il z
0'a rp 9)
CO cr' CQ C)
I C)
D0 U
Q 0
ll0 l00 90
&0 20 60 FIGURE 15 CASE 5 HEATADDITIONCASE PRIMARYTFMPERATURE 60'F PRIMARYPRESSURE 329.7 PSIA NO VENT, NO Sl, NO CHARGING PUMP ONE RC PUMP STARTED 2000 GPM RHR Il 0
0 fD 0) 50 0
20 40 60
&0 10Q Transient time in secs
O
~\\
O0 U
R x10 0
12 8
4 0
FIGURE CASE 5 HEATADDITIONCASE PRIMARYTEMPERATURE 60 F PRIMARYPRESSURE 329.7 PSIA NO VENT, NO Sl, NO CHARGING PUMP ONE RC PUMP STARTED 2000 GPM RHR Tl O
O CD 0)
-4 20 40 60 Transient time in sees 80
400 U
Pl p,
-800 0O U -1200
~ -MM FIGURE CASF 5 HEATADDITIONCASE PRIMARYTEMPERATURE 60'F PRIMARYPRESSURE 329.7 PSIA NO VENT, NO Sl, NO CHARGING PUMP ONE RC PUMP STARTED 2000 GPM RHR
'T1 0
OU CD
-2000
-24QO 20 60 40 Transient time in secs 80 I
PJ CO C)
I C)
112 108 104 100 96 92 FIGUR CASE 5 HEATADDITIONCASE PRIMARYTEMPERATURE 60'F PRIMARYPRESSURE 329.7 PSIA NO VENT, NO Sl, NO CHARGING PUMP ONE RC PUMP STARTED 2000 GPM RHR rl O
OD CDI 88 20 40 60 Transient time in secs 80 100
FIGURE 19 CASE 5 HEATADDITIONCASE PRIMARYTEMPERATURE 60'F PRIMARYPRESSURE 329.7 PSIA NO VENT, NO Sl, NO CHARGING PUMP ONE RC PUMP STARTED 2000 GPM RHR 115 110 105 100 95 90 85 0
20 60 40 Transient time in secs 80 Tl O
U OU 6) 9)
CO 0
l CX) tD
CASE 5 FIGURE 20 HEATADDITIONCASE PRIMARYTEMPERATURE 60'F PRIMARYPRESSURE 329.7 PSIA NO VENT, NO Sl, NO CHARGING PUMP ONE RC PUMP STARTED 2000 GPM RHR 0
9)
(D (D
Ol C) 60 50 40 30 20 10 20 40 60 Transient time in secs 80 100 Tl O
C)D (D
9)
CO I
fV CO Pv C)
I C)
560 FIGURE CASE 6 HEATADDITIONCASE PRIMARYTEMPERATURE 85oF PRIMARYPRESSURE 32S.7 PSIA NO VENT, NO Sl, NO CHARGING PUMP ONE RC PUMP STARTED 2000 GPM RHR Tl OD 0U CD 520 480 440 400 360 320 10 20 30 Transient time in secs 50 C3 l
fV I
C)
680 FIGUR CASE 6 HEATADDITIONCASE
~ PRIMARYTEMPERATURE 85oF PRIMARY PRESSURE 329.7 PSIA NO VENT, NO Sl, NO CHARGING PUMP ONE RC PUMP STARTED 2000 GPM RHR li 0
Oa 6) f0 520 480 10 20 30 Transient time in secs LEGEND
~
RHR.PUMP A
lt RHR.PUMP B 40 50 i
PJ tV I"
C)
'44
pf
~1 00 U
O 128 120 112 104 96 88 FIGUR CASE 6 HEATADDITIONC PRIMARYTEMPERATURE 86oF PRIMARYPRESSURE 329.7 PSIA NO VENT, NO Sl, NO CHARGING PUMP ONE RC PUMP STARTED 2000 GPM RHR Tl O
U OD
~a (0
80 10 20 30 Transient time in secs 50 Q'
COfv I
FIGUR CASE 6 HEATADDITIONCASE PRIMARYTEMPERATURE 85oF PRIMARY PRESSURE 329.7 PSIA NO VENT, NO Sl, NO CHARGING PUMP ONE RC PUMP. STARTED 2000 GPM RHR
~\\
00 O
R 2
U 0g 1C0 130 120 110 100 90 80 10 20 30 Transient time in secs SO Tl 0
0a (D
9)
CO 0
I CO I
lyl
xl0 20 l6 FIGURE 25 CASE 6 HEATADDITIONCASE PRIMARYTEMPERATURE 85oF PRIMARYPRESSURE 329.? PSIA NO VENT, NO Sl, NO CHARGING PUMP ONE RC PUMP STARTED 2000.GPM RHR rl "0
OU 1
(D PD O
~I 00 O
12 8
4 0
l0 20 30 SO Transient time in secs
FIGURE 2 CASE 6 HEATADDITIONCASE PRIMARYTEMPERATURE 85oP PRIMARYPRESSURE 329.7 PSIA NO VENT, NO Sl, NO CHARGING PUMP ONE RC PUMP STARTED 2000 GPM RHR 0
Oa CD P) 0O
-S00 U -1200 P
-Moo
-2000
-2400 10 20 30 Transient time in secs.
50 0
1 R)
Co fV C)
I tD
140 FIGURE CASP6 HEATADDITIONCASE PRIMARYTEMPERATURF85'F PRIMARYPRESSURE 329.7 PSIA NO VENT, NO Sl, NO CHARGING PUMP ONE RC PUMP STARTED 2000 GPM RHR Q
C4ul z
Pl zOO 136 132 128 124 120 116 10 20 30 Transient time in secs 50 rl 0
OU CD 1
FV CO fM C)
I CO
Q f-cU z
2 lV 2
A "O
txl 140 136 132 128 124 120 116 0
FIGU CASE 6 HEATADDITION PRIMARYTFMPERATURE 85'F PRIMARYPRESSURE 329.7 PSIA NO VENT, NO Sl, NO CHARGING PUMP ONE RC PUMP STARTED 2000 GPM RHR 10 20 30 Transient time in secs 50
=ll QD 0o tD fD CQ l
RJ
'44 CO M
C)
I CO V4
120 100 80 60 FIG 9
CASE 6 HEATADDITIONCASE PRIMARYTEMPERATURE 85'F PRIMARYPRESSURE 329.7 PSIA NO VENT, NO Sl, NO CHARGING PUMP ONE RC PUMP STARTED 2000 GPM RHR Il OD OD (D
Ql 20 0
0 10 20 30 Transient time in secs 50
550 500 450 350
~ 0 FIGURE 30 CASE 7 HEATADDITIONCASE PRIMARYTEMPERATURE 280oP PRIMARYPRESSURE 329.7 PSIA NO VENT, NO Sl, NO CHARGING PUMP ONE RC PUMP STARTED 2000 GPM RHR 30 20 lo Transient time in secs S0 Tl O
0U CD Ql CO I
hJ CX) fM CO I
C)
FIG 31 CASE 7 HEATADDITIONCASE PRIMARYTEMPERATURE 280'F PRIMARYPRESSURE 329.7 PSIA NO VENT, NO Sl, NO CHARGING PUMP ONE RC PUMP STARTED 2000 GPM RHR 650 600 550 I
II I
PI III LEGEND
~
RHR-PUMP A
~
RHR-PUMP B n
O OU
~o CD 10 20 30 Transient time in secs 40
~ ~
FIGU CASF 7 HEATADDITIONCASE PRIMARYTEMPERATURE 280'F PRIMARYPRESSURE 329.7 PSIA NO VENT, NO Sl, NO CHARGING PUMP ONE RC PUMP STARTED-2000 GPM RHR
~\\
00 Pl Ql U
0 3 IQ 305 300 295 290 285 280 10 I
I I
I I
2Q 30 Transient time in secs 50 ll O
U OaI 0)
CO I
CO bJ CD I
CO
~I 00 U
0 320 3I2 304 296 288 280 10 20 30
'I Transient time in secs FIGURE 33 CASE 7 HEATADDITIONCASE PRIMARYTEMPERATURE 280oF PRIMARYPRESSURE 32S.7 PSIA NO VENT, NO SI, NO CHARGING PUMP ONE RC PUMP STARTED 2000 GPM RHR 50
'Tl O
OD CD Gl I
bJ fV CD I
CD
xlQ 20 16 FIGURE 34 CASE 7 HEATADDITIONCASE PRIMARYTEMPERATURE 280'F PRIMARYPRESSURE 329.7 PSIA NO VENT, NO Sl, NO CHARGING PUMP ONE RC PUMP STARTED 2000 GPM RHR n
0 0U 6) 12 8
4 0
10 30 Transient time in secs CO 1
fV CO (D
tD
FIG 35 CASE 7 HEATADDITIONCASE PRIMARYTEMPERATURE 280'F PRIMARYPRESSURE 329.7 PSIA NO VENT, NO SI, NO CHARGING PUMP ONE RC PUMP STARTED 2000 GPM RHR rl 0
OU CD p,
-800 00 U -1200 Pl E
-16OO
-2000
-2400 0
10 20 30 Transient time in secs So