ML17264A508
| ML17264A508 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 05/24/1996 |
| From: | Vissing G NRC (Affiliation Not Assigned) |
| To: | NRC (Affiliation Not Assigned) |
| References | |
| TAC-M94770, NUDOCS 9605300215 | |
| Download: ML17264A508 (57) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WAS NING TO N y 0 C 2055$ 4001 Hay 24, 1996 LICENSEE:
ROCHESTER GAS AND ELECTRIC COMPANY FACILITY:
R.
E.
GINNA NUCLEAR POWER PLANT
SUBJECT:
SUMMARY
OF MEETING WITH REPRESENTATIVES OF ROCHESTER GAS AND ELECTRIC COMPANY ON MAY 15,
- 1996, IN THE OFFICES OF THE NRC CONCERNING THE EXCHANGE OF NEUTRON FLUENCE CALCULATIONS TO SUPPORT AN AMENDMENT THAT WOULD CHANGE THE TECHNICAL SPECIFICATIONS RELATED TO PRESSURE/TEMPERATURE LIMITS (TAC NO. M94770) ntroductio To support the proposed amendment dated March 15,
- 1996, as supplemented April 22,
- 1996, the licensee had proposed the use of an estimated value of fluence at the exposed surface of the reactor vessel of 7 percent over previous values that were used.
Since this is an estimated value, the staff found that to support the amendment and the approval of the Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) it would be necessary to provide fluence values based on actual cat.I.'ovations.
The licensee contractor (Westinghouse Electric Company) had iIot been able to supply the necessary calculations in time to support the issuance of the approval of the licensee's RCS PTLR on a schedule tc support the restart of the plant.
Therefore, the licensee was requested by phone conversation of May '13, 1996, to bring the portions of the approved Westinghouse report that provided calculated value of fluence for discussion.
The attendees of the meeting are identified in Attachment 1.
The approved Westinghouse report providing the latest fluence calculations is provided as.
~oi scussio It was agreed that the Westinghouse report,, would be placed on the Docket and be reviewed by the staff to support the staff review of proposed PTLR and the proposed amendment of March 15,
- 1996, as supplemented April 22, 1996.
It appeared that the calculated neutron fluence values would be bounded by the estimated 7 percent over previous values
- and, thus, would satisfy the staff's review on an interim basis.
The new values appear to SOOOI'7 9605300215 960520 PDR ADOCK 05000244 PDR Izg RILE C~YE ~~
I impact other areas that were not considered in the Westinghouse report such as the chemistry factor.
Thus, the review would focus on an interim operation until a full review of all areas could be completed.
Docket No. 50-244 Guy. S. Vissing, Senior oject Manager Project Directorate I-I Division of Reactor Projects I/II Office of Nuclear Reactor Regulation Attachments:
1.
List.of Attendees 2.
Westinghouse Report Section 6.0, Radiation Analysis and Neutron Dosimetry cc w/attachs:
See next page
Rochester Gas and Electr ic Corporati on R.
E. Ginna Nuclear Power Plant CC:
Peter D. Drysdale, Senior Resident Inspector R.E.
Ginna Plant U.S. Nuclear Regulatory Commission 1503 Lake Road
- Ontario, NY 14519
'egional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. F. William Valentino, President New York State
- Energy, Research, and Development Authority 2 Rockefeller Plaza
- Albany, NY 12223-1253 Charlie Donaldson, Esq.
Assistant Attorney General New York Department of Law 120 Broadway New York, NY 10271 Nicholas S.
Reynolds Winston 8 Strawn 1400 L St.
N.W.
Washington, DC 20005-3502 Ms. Thelma Wideman
- Director, Wayne County Emergency Management Office Wayne County Emergency Operations Center 7336 Route 31
. Lyons, NY 14489 Ms. Mary Louise Meisenzahl Administrator, Monroe County Office of Emergency Preparedness.
ill West Fall Road, Room 11 Rochester, NY 14620 Dr. Robert C. Mecredy Vice President, Nuclear Operations Rochester Gas and Electric Corporation 89 East Avenue Rochester, New York 14649
I
LIST OF ATTENDEES G WITH REPR SENTATIVES OF C
GAS AN L
C R
C COMPANY 0
C R
G N TRON FLU NC CALCU ATIO S FO R
GINN UCL AR POWER P
AN ES T ON PRESSURE VESSE MAY 15 1996
~A Guy S. Vissing Jocelyn Mitchell Robert J. Giardina Tim Collins Lambros Lois Barry Elliot Melita P. Osborne Arnold Fero John Caren E. Dale McGarry James Adams George Wrobel
~AMIZ TIO NRR/PD I-1 NRR/PD I-1 NRR/TSB NRR/SRXB NRR NRR Westinghouse Tech Proj. Mgr.
NSD Westinghouse BNL NIST NIST Rochester Gas and Electric Attachment 1
SECTION 6.0 RADIATIONANALYSIS AND NEUTRON DOSIMETRY 6.1 Introduction Knowledge of the neutron environment within the reactor vessel and surveillance capsule geometry is required as an integral part of LWR reactor vessel surveillance programs for two reasons.
First, in order to interpret the neutron radiation induced material property changes observed in the test specimens, the neutron environment (energy spectrum, flux, fluence) to which the test specimens were exposed must be known.
Second, in order to relate the changes observed in the test specimens to the present and future condition of the reactor
- vessel, a relationship must be established between the neutron environment at various positions within the reactor vessel and that experienced by the test specimens.
The former requirement is normally met by employing a combination of rigorous analytical techniques and measurements obtained with passive neutron flux monitors contained in each of the surveillance capsules.
The latter information is generally derived solely from analysis.
The use of fast neutron fluence (E > 1.0 MeV) to correlate measured material property changes to the neutron exposure of the material has traditionally, been accepted for development of damage trend curves as well as for the implementation of trend curve data to assess vessel condition.
In recent years, however, it has been suggested that an exposure model that accounts for differences in neutron energy spectra between surveillance capsule locations and po'sitions 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 6-I Attachment 2
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-Vl 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.
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.
6.2 Discrete Ordinates Anal sis A plan view of the reactor geometry at the core midplane is shown in Figure 4-1.
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 4-1.
A plan view of a surveillance capsule holder attached to the thermal shield is shown in Figure 6-1.
The stainless steel specimen containers are approximately 1-inch square and approximately 38 inches in height.
The containers are positioned axially such that the 6-2
test, specimens are centered on the core midplane, thus spanning the central 3 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), $(E > 0.1 MeV), and dpa/sec) through the vessel wall. The neutron spectral information was required for the interpretation of neutron dosimetry withdrawn from the surveillance capsules as well as for the determination of exposure parameter ratios; i.e., [dpa/sec]/[$ (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 '/~T and i/~T locations.
The second set of calculations consisted of a series of adjoint analyses relating the fast neutron flux, $(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 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 6-3
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 burnup 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 4-1 and 6-1 was carried out in R,8 geometry using the DORT two-dimensional discrete ordinates code Version 2.7.3"" and 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 r'eference 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 2a margin 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 6-4
'ominal + 2o 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 Sorder of angular quadrature and the P, <<ross-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'eometric center of each surveillance capsule.
Again, these calculations were run in R,8 geometry to provide neutron source distribution importance functions for the exposure parameter of interest, in this case $(E > 1.0 MeV).
Having the importance functions and appropriate core source distributions, the response of interest c'ould be calculated as:
~(re) f=f J i(rB,E) S(rB,E) r dr dB dE r
0 E
where:
R(r,8) =
1(r,8,E)=
S(r,8,E)=
$(E > 1.0 MeV) at radius r and azimuthal angle 8.
Adjoint source importance function at radius r, azimuthal angle 8, and neutron source energy.E.
Neutron source strength at core location r,8 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 $(E > 0.1 MeV) were computed on a cycle specific basis by using [dpa/sec]/[$ (E > 1.0 MeV)] and [$(E > 0.1 MeV)]/[$(E > 1.0 MeV)7 ratios from the forward analysis in conjunction with the cycle specific $(E > 1.0 MeV) solutions from the individual adjoint evaluations.
6-5
The reactor core power distributions used in the plant specific adjoint calculations were taken'rom the fuel cycle design reports for the first twenty-five operating cycles and the upcoming twenty-sixth cycle of Ginna i'~'"'.
Selected results from the neutron transport analyses are provided in Tables 6-1 through 6-5.
The data listed in these tables establish the means for absolute comparisons of analysis and measurement for the capsule irradiation periods and provide the means to correlate dosimetry results with the corresponding exposure of the reactor vessel wall.
In Table 6-1, the calculated exposure parameters [$(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 6-2 for the reactor vessel inner radius.
Again, the three pertinetit exposure parameters are listed for the reference and Cycles 1 through 26 plant specific power distributions.
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 levels of 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 6-3, 6-4, and 6-5, 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 6-3 through 6-5.
6-6
'For ex'air)pie, the neutron flux (t)(E > 1.0 MeV) at the '/4T depth in the reactor v'essel wall along the 0'zimuth is given by:
4, )0')
= $(168.04, 0') F(172.26, 0 )
where:
(I)64 (0') =
Projected neutron flux at the '/4T position on the 0',azimuth.
(I)(168.04,0') =
Projected or calculated neutron flux at the vessel inner radius on the 0'zimuth.
F(172.25.0') =
Ratio of the neutron flux at the i/4T position to the flux at the vessel inner radius for the 0'zimuth.
This data is obtained from Table 6-3.
Similar expressions apply for exposure parameters expressed in terms of (I)(E > 0.1 MeV) and dpa/sec where the attenuation function F is obtained from Tables 6-4 and 6-5, respectively.
6.3 Neutron Dosimetr The passive neutron sensors included in the Ginna surveillance program are listed in Table 6-.
6.
Also given in Table 6-6 are the primary nuclear reactions and associated nuclear constants that were used in the evaluation of the neutron energy spectrum within the surveillance capsules and in the subsequent determination of the various exposure parameters of interest
{(I)(E > 1.0 MeV), (I)(E > 0.1 MeV), dpa/sec].
The relative locations of the neutron sensors within the capsules are shown in Figure 4-2.
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 6-6 does not yield a direct measure of the energy dependent neutron flux at the point of interest.
Rather, the activation or fission process is a measure of the integrated effect that the time and energy dependent neutron flux has on the target material over the course of the irradiation period.
An accurate assessment 6-7
of the average neutron flux level incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well known.
In particular. the following variables are of interest:
The measured specific activity of each>>ionitor, 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""@ ~'.
Following sample preparation and weighing, the activity of each monitor was determined by means of a lithium-drifted germanium, Ge(Li), gamma spectrometer.
The iriadiation history of the Ginna reactor was obtained from NUREG-0020, "Licensed Operating Reactors Status Summary Report," and plant personnel for the cycles I
through 25 operating period.
The irradiation history applicable. to the exposure of Capsules V, R, T, and S is given in Table 6-7.
Having the measured specific activities, the physical characteristics of the sensors, and the operating history of the reactor, reaction rates referenced to full power operation were determined from the following equation:
where:
A N()
N, F YP C [1-e
'] [e
'J mf 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.
6-8
Cl Pj Avera ~e core power level during irradiation period j (MW)
P, =
Maximum or reference power level of the reactor (MW).
C J
Calculated ratio of $(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 (I/sec).
t,.
=
Length of irradiation period j (sec).
t=
Decay time following irradiation period j (se<<).
and the summation is carried out over the total number of monthly intervals comprising the irradiation period.
In the equation describing the reaction rate calculation, the ratio [P)/[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 C,, which can be calculated for each fuel cycle using the adjoint transport technology discussed in Section 6.2, 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<ycle irradiation C> is normally taken to be 1.0.
However, for multiple cycle irradiations, particularly those employing low leakage fuel management, the additional C, 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 management 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 6-1.
Measured and saturated reaction product specific activities as well as the derived full power reaction rates are listed in Table 6-8.
The specific activities and reaction rates of the "'U sensors provided in Table 6-8 include corrections for "'U impurities, plutonium build-in, and gamma ray induced fissions.
Corrections for gamma ray induced fissions were also included in the t
specific activities and reaction rates for the "Np,sensors as well.
6-9'
Values of key fast neutron exposure parameters were derived from the measured rea<<tlon rates gsing the FERRET least squares adjustment <<ode'"'.
The FERRET approach used the measured reaction rate data, sensor reaction cross-sections, and a calculated trial spectrum a>>
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.
In general, the measured values f are linearly related to the flux $ by some response matrix f(BID),p A(s) $
(c) e where i indexes the measured values belonging to a single data set s, g designates the energy, group, and a delineates spectra that may be simultaneously adjusted.
For example, relates a set of measured reaction rates R; to a single spectrum $, by the multigroup reaction cross-section o,
. 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 code'"'.
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 6-10
I 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 ener'gy 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-VI data files, the covariance matrix for the input trial spectrum was constructed from the following relation:
M I=R +R RIP I
2 gg n
g g
gg where R. specifies 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:
P I = [1-8] 6 I + 8 e gg gg where:
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 6 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 8 is close to l.
Strong long range correlations (or anti-correlations) were justified based on information presented by R. E. Maerker'"'.
The uncertainties associated with the measured reaction rates included both statistical (counting) and systematic components.
The systematic component of the overall uncertainty accounts for counter efficiency, counter calibrations, irradiation history corrections. and corrections for competing reactions in the individual sensors.
Results of the FERRET evaluations of the Capsules V, R, T, and S dosimetry are given in Table 6-9.
The data summarized in this table include fast neutron exposure evaluations in terms of 4 (E > 1.0 MeV), C?(E > 0.1 MeV), and dpa.
Jn 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 6-10.
An examination of Table 6-10 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 6-11 in the FERRET 53 energy group structure.
In Table 6-12, 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.
6.4 Pro'ections of Reactor Vessel Ex osure 6-12
The best estimate exposure of the Ginna reactor vessel was developed using a combination ot'bsolute 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:
tE where:
C>,E=
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.
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 6-13
4(E > 1.0 MeV), 4(E > 0.1 MeV), and dpa, respectively.
Bias factors of this magnitude are fully consistent 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 va'riations 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.
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 6-14
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 %13%.
Based on this best estimate approach, neutron exposure projections at key locations on the reactor vessel inner radius are given in Table 6-13.
Along with the current (19.51 EFPY) exposure. projections are also provided for exposure periods of 28 EFPY, 32 EFPY, 42 II 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 24, 32, 42 and 48 EFPY were also employed.
Data based on both a 4(E ) 1.0 MeV) slope and a plant spe'cific dpa slope through the vessel wall are provided in Table 6-14.
In order to access RTN,>> vs.fluence curves, dpa equivalent fast neutron fluence levels for the
'/4T and '/~T positions were defined by the relations:
and dpa('/i T) dpa(OT)
Using this approach results in the dpa equivalent fluence values listed in Table 6-14.
In Table 6-15 updated lead factors are listed for each of the Ginna surveillance capsules.
Lead factor data based on the accumulated fluence through the Cycle 26 projection are provided for each remaining capsule.
6-15
FIGURE 6-1 PI.AN VIEW OF A REACTOR VESSEL SURUEILLANCE CAPSULE (is,z~..sz
)
( l2',22,32' CHARPY SPEC IMEhl THERMAL SHIELO 6-16
TABLE 6-1 CALCULATEDFAST NEUTRON EXPOSURE RATES AND IRON ATOM DISPLACEMENT RATES AT THE SURVEILLANCECAPSULE CENTER
~Cele No.
Reference IA IB 2
3 4
5 6
7 8
10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 13'.59e+
I I 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+11 1.063e+11 1.027e+II 9.149e+10 9.652e+10 1.030e+11 9.259e+10 9.133e+10 9.723e+10 9.691e+10 1.038e+11 9.322e+10 9.072e+10 8.436e+10 8.418e+10
$ (E > 1.0 MeV) (n/em=sec) f30 9.35e+10 7.051e+10 8.009e+10 8.096e+10 6.661e+10 7.266e+10 7.714e+10 8.240e+10 7.520e+10 8.433e+10 8.455e+10 7.865e+10 7.127e+10 7.241e+10 6.870e+10 7.053e+10 6.622e+10 6.054e+10 6.362e+10 6.261e+10 6.065e+10 6.228e+10 6.476e+10 6.575e+10 6.167e+10 5.884e+10 5.510e+10 5.299e+10 33'.83e+10 6.684e+10 7.141e+10 7.682e+IO 6.263e+10 7.217e+10 7.069e+10 7.397e+10 7.026e+10 8.033e+10 8.303e+10 7.751e+10 6.474e+10 6.357e+10 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 6-17
TABLE 6-1 cont'd CALCULATEDFAST NEUTRON EXPOSURE RATES AND IRON ATOM DISPLACEMENT RATES AT THE SURVEILLANCECAPSULE CENTER
~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 13 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+ll 5.149e+ll 4.706e+ll 4.737e+ll 5.220e+ll 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+11 3.438e+11 3.197e+11 3.191e+ll
$(E > 0.1 MeV) (n/cm -sec) 23'.22e+11 2.426e+ll 2.755e+11 2.785e+11 2.291e+11 2.500e+11 2.654e+11 2.835e+11 2.587e+11 2.901e+11 2.909e+11 2.706e+ll 2.452e+11 2.49le+11 2.363e+ll 2.426e+ll 2.278e+11 2.082e+1.1 2.189e+11 2.154e+11 2.086e+11 2.142e+ll 2.228e+11 2.262e+ll 2.121e+11 2.024e+11 1.896e+11 1.823e+11
~30 3.11e+11 2.353e+11 2.513e+11 2.704e+11 2.205e+Il 2.541e+ll 2.488e+ll 2.604e+11 2.473e+11 2.828e+11 2.923e+11 2.728e+ll 2.279e+11 2.238e+11 2.201e+11 2.438e+11 2.311e+11 2.095e+1.1 2.089e+11 2.054e+11 2.071e+11 2.028e+11 2.106e+11 2.132e+11 2.114e+11 1.935e+11 1.863e+11 1.766e+11 6-18
TABLE 6-1 cont'd CALCULATEDFAST NEUTRON EXPOSURE RATES AND IRON ATOM DISPLACEMENT RATES AT THE SURVEILLANCECAPSULE CENTER Cvcle No.
Reference IA IB 2
3 4
5 6
7 8
9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 130 2.83e-10 2.127e-10 2.535e-10 2.437e-10 2.037e-10 1.928e-10 2.360e-10 2.448e-10 2.123e-10 2.553e-10 2.418e-10 2.210e-10 2.225e-10 2.452e-10 1.893e-10 1.829e-10 1.629e-10 1.718e-10 1.833e-10 1.648e-IO 1.626e-10 1.731e-10 1.725e-10 1.847e-10 1.659e-10 1.615e-10 1.502e-10 1.498e-10 Displacement Rate (dpa/sec) 23'.59e-10 1.199e-10 1.362e-10 1.376e-10 1.132e-10 1.235e;10 1.311e-10 1.401e-10 1.278e-10 1.434e-10 1.437e-10 1.337e-10 1.212e-10 1.231e-10 1.168e-10 1.199e-10 1.126e-10 1.029e-10 1.082e-10 1.064e-10 1.031e-10 1.059e-10 1.101e-10 1.118 e-10 1.048e-10 1.000e-10 9.368e-I I 9.009e-I I 33'.52e-IO 1.150e-10 1.228e-I0 1.321e-10 1.077e-10 1.241e-10 1.216e-10 1.272e-10 1.208e-10 1.382e-10 1.428e-10 1.333e-10 1.113e-10 1.093e-10 1.075e-10 1.191e-10 1.129e-10 1.024e-10 1.021e-10 1.004e-10 1.012e-10 9.910e-I I 1.029e-10 1.042e-10 1.033e-10 9.454e-I I 9.103e-I I 8.627e-I I 6-19
TABLE 6-2 E.
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 iso.
Reference 1A 1B 2
3 4
5 6
7 8
10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 4.14e+09 4.022e+10 4.784e+10 4.590e+10 3.792e+10 3.488e+10 4.441e+10 4.505e+10 3.835e+10 4.828e+10 4.484e+10 4.092e+10 4.076e+10 4.746e+10 3.616e+10 3.404e+10 2.800e+10 3.078e+10 3.633e+10 3.008e+10 3.003e+10 3.127e+10
'3.007e+10 3.510e+10 3.039e+10 3.000e+,10 2.794e+10 2.885e+10 15'.20e+09 2.452e+10 2.909e+10 2.811e+10 2.348e+10 2.290e+10 2.720e+10 2.836e+10 2.488e+10 2.938e+10 2.816e+10
'.582e+10 2.570e+10 2.787e+10 2.258e+10 2.213e+10 2.003e+10 2.043e+10 2.170e+10 1.995e+10 1.956e+10 2.068e+10 2.080e+10 2.203e+10 1.995e+10 1.938e+10 1.803e+10 1.788e+10 30o 2.19e+09 1.680e+10 1.831e+10 1.926e+10 1.576e+10 1.796e+10 1.799e+10 1.90le+10 1.784e+10 2.014e+10 2.066e+10 1.930e+10 1.650e+10 1.635e+10 1.615e+10 1.735e+10 1.640e+10 1.482e+10 1.514e+10 1.496e+10 1.483e+10 1.475e+10 1.537e+10 1.553e+10 1.511e+10 1.406e+10 1.340e+10 1.275e+10 45'.83e+09 1.420e+10 1.517e+10 1.688e+10 1.374e+10 1.581e+10 1.501e+10 1.475e+10
'.459e+10 1.735e+10 1.820e+10 1.683e+10 1.543e+10 1.500e+10 1.355e+10 1.690e+10 1.664e+10 1.546e+10 1.380e+10 1.300e+10 1.366e+10 1.315e+10 1.338e+10 1.380e+10 1.408e+10 1.249e+10 1.226e+10 1.181e+10 6-20
TABLE 6-2 cont'd CALCULATEDAZIMUTHALVARIATIONOF FAST NEUTRON EXPOSURE RATES AND IRON ATOM DISPLACEMENT RATES AT THE REACTOR VESSEL CLAD/BASE METAL INTERFACE
$(E > 0.1 MeV) (n/em -sec)
~Cele Nn.
Reference IA IB 2
3 4
5 6
7 8
9 10 11 12 13 14 15'6 17 IN 19 20 21 22 23 24 25 26 3.25e+ I()
3.121e+11 3.712e+ll 3.562e+11 2.942e+11 2.706e+ll 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+ll 2.641e+11 2.173e+11 2.389e+11 2.819e+11 2.334e+11 2.330e+11 2.427e+11 2.333e+ll 2.724e+11 2.358e+ll 2.328e+11 2.168e+11 2.239e+11 15'.76e+10 2.116e+11 2.511e+11 2.426e+11 2.026e+11 1.976e+11 2.348e+11 2.447e+11 2.147e+ll 2.536e+11 2.430e+11 2.228e+11 2.218e+11 2.406e+11 1.948e+11 1.910e+11 1.729e+11 1.764e+11 1.872e+11 1.722e+II 1.688e+11 1.785e+ll 1.795e+11 1.901e+11 1.721e+II 1.672e+11 1.556e+11 1.543e+11 30'.90e+10 1.465e+11 1.597e+11 1.680e+11 1.374e+ll 1.566e+11 1.568e+11 1.658e+11 1.556e+ll 1.757e+11 1.802e+11 1.683e+ll 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 1.318e+11 1.226e+11 1.169e+11 1.112e+11 45 1.50e+10 1.164e+ll 1.244e+11 1.384e+ll 1.126e+ll 1.297e+11 1.230e+ll 1.210e+11 1.197e+ll 1.422e+11 1.492e+11 1.380e+11 1.266e+11 1.230e+11 1.111e+11 1.386e+ll 1.364e+11 1.268e+11 1.131e+11 1.066e+11 1.120e+11 1.079e+11 1.097e+11 1.132e+ll 1.155e+ll 1.024e+11 1.005e+11 9.687e+10 6-21
TABLE 6-2 cont'd CALCULATEDAZIMUTHALVARIATIONOF FAST NEUTRON EXPOSURE RATES AND IRON ATOM DISPLACEMENT RATES AT THE REACTOR VESSEL CLAD/BASE METAL INTERFACE
Displacement Rate (dpa/sec)
~Cele No.
Reference IA IB 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 1.17e-I I 1.138e-10 1.354e-10 1.299e-10 1.073e-10 9.870e-I I 1.257e-10 1.275e-10 1.085 e-10 1.366e-10
'.269e-10 1.158e-10 1.154e-10 1.343e-10 1.023e-10 9.633e-I I'.924e-I I 8.712e-I I 1.028e-10 8.513e-I I 8.498e-I I 8.850e-I I 8.510e-I I 9.933e-I I 8.599e-I I 8.491e-I I 7.906e-11 8.164e-I I 0
9.70e-12 7.430e-I I 8.815e-I I 8.517e-I I 7.113e-I I 6.938e-I I 8.242e-I I 8.593e-I I 7.538e-I I 8.903 e-I I 8.531e-I I 7.823e-I I 7.787e-I I 8.446 e-I I 6.841e-I I 6.706 e-I I 6.070e-I I 6.192e-I I 6.574e-I I 6.046e-I I 5.927e-I I 6.267e-I I 6.301e-I I 6.674e-I I 6.044e-I I 5.871e-I I 5.463e-I I 5.418e-I I 30'.70e-12 5.141e-I I 5.604e-I I 5.894e-I I 4.823e-I I 5.496 e-I I 5.503e-I I 5.818e-I I 5.460e-I I 6.164e-I I 6.322e-I I 5.904e-I I 5.048e-I I 5.002e-I I 4.941e-I I 5.308e-I I 5.019e-I I 4.534e-I I 4.632e-I I 4.577e-I I 4.538e-I I 4.514e-I I 4.702e-I I 4.752e-I I 4.624e-I I 4.301e-11 4.101e-I I 3.901e-I I 45'.36e-12 4.160e-I I 4.446e-I I 4.946e-I I 4.025e-I I 4.633e-I I 4.397e-I I 4.322e-I I 4.276e-I I 5.082e-I I 5.331e-I I 4.930e-I I 4.522e-I I 4.395e-I I 3.971e-I I
'4.952e-I I 4.875e-I I 4.530e-I I 4.042e-I I 3.809e-I I 4.002e-I I 3.854e-I I 3.920e-I I 4.045e-I I 4.127e-I I 3.660e-I I 3.592e-I I 3.461e-I I 6-22
TABLE 6-3 RELATIVE RADIALDISTRIBUTION OF $ (E > 1.0 MeV)
WITHINTHE REACTOR VESSEL WALL RADIUS
~cm 168.(>4 168.27 168.88 169.75 170.93 172.25 173.53 174.98 176.46 177.58 179.03 180.66 181.63 182.60 184.06 184.87
()'.000 0.987 0.940 0.862 0.754 0.639 0.540 0.444 0.362 0.308 0.250 0.196 0.169 0.144 0.110 0.101 AZIMUTHALANGLE 1 50 30'.0001.000 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 0.151 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 t/~T =
Base Metal-~/~T =
Base Metal ~/~T =
Base Metal Outer Radius =
168.04 cm.
172.25 cm.
176.46 cm.
180.66 cm.
184.87 cm.
6-23
TABLE 6-4 RELATIVERADIALDISTRIBUTION OF $(E > 0.1 MeV)
WITHINTHE REACTOR VESSEL WALL RADIUS
~crn 168.()4 168.27 168.88 169.75 170.93 172.25 173.53 174.98 176.46 177.58 179.03 180.66 181.63 182.60 184.06 184.87 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 AZIMUTHAL 1 50 1.()00 1.007 1.007 0.990 0.948 0.891 0.831 0.763 0.693 0.640 0.573 0.502 0.458 0.415 0.348 0.327 ANGLE
~0'.000 1.005 1.004 0.985 0.945 0.889 0.831 0.763 0.694 0.642 0.577 0.507 0.466 0.423 0.361 0.343 45n 1.000 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 KT =
Base Metal '/~T =
Base Metal Outer Radius =
168.04 cm.
172.25 cm.
176.46 cm.
180.66 cm.
184.87 cm.
6-24
TABLE 6-5 RELATIVE RADIALDISTRIBUTION OF dpa/sec 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 0'.000 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 0.187 0.173 AZIMUTHALANGLE 15" 30'.0001.000 0990
'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
'.220 0.206 0.208 45" 1.000 0.989 0.954 0.857 0.812 0.723 0.644 0.563 0.490 0.439 0.380 0.322 0.289 0.258 0.216 0.205 Note:
Base Metal Inner Radius =
Base Metal ~/4T =
Base Metal KT =
Base Metal '/~T =
Base Metal Outer Radius =
168.04 cm.
172.25 cm.
176.46 cm.
180.66 cm.
184.87 cm.
6-25
TABLE 6-6 NUCLEAR PARAMETERS USED IN THE EVALUATIONOF NEUTRON SENSORS Monitor Material Reaction of Interest Target Atom Fraction
Response
~Ran e
Fission Product Yield Half-life Copper Iron Nickel Uranium-238 Neptunium-237 Cobalt-Al
"'Cu (n,a)
"Fe (n,p)
"Ni (n,p)
"U (n,f)
'""Np (n,f)
"Co (n,q) 0.6917 0.0580 0.6827 0.9996 1.0000 0.0015 E > 4.7 MeV E> 10 MeV E>10MeV E >04 MeV E > 0.08 MeV E > 0.015 MeV 5.271 y 312.5 d 70.78 d 30.17 y 30.17 y 5.271 y 6.00 6.27 Note: "'U and "'Np monitors are cadmium shielded.
6-26
TABLE 6-7 iVIONTHLYTHERMALGENERATION DURING THE FIRST TWENTY-FIVE FUEL CYCLES OF THE GINNA REACTOR Cycle 1A Cycle 1B Cycle 2 Cycle 3 Month Thermal Gen.
MWt-hr Thermal Gen.
Month MWt-hr Thermal Gen.
Month MWt-hr Month Thermal Gen.
MWt-hr 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-71 Feb-71 Mar-71 0
0 435541 435541 435541 435541 435541 435541 435541 930964 860611 481017 830385 840563 831856 956228 27388 Apr-71 May-71 Jun-71 Jul-71 Aug-71 Sep-71 Oct-71 Nov-71 Dec-71 Jan-72 Feb-72 Mar-72 Apr-72 0
330899 831633 835525 922141 913338 957036 956391 941632 956804 955012 740009 655096 May-72 Jun-72 Jul-72 Aug-72 Sep-72 Oct-72 0
7722 818270 897991 910771 329771 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 381843 1053950 886()67 852420 942682 9 l4688 947502 906810 678681 945755 951662 818471 980073 923316 2908 0
0 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
'82048 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
'48096 1088328 872256 888480 870024 0
6-27
TABLE 6-7 cont'd MONTHLYTHERMALGENERATION DURING THE FIRST TWENTY-FIVE FUEL CYCLES OF THE GINNA REACTOR.
Cycle 8 Thermal Gen.
Month MWt-hr Cycle 9 Month Thermal Gen.
MWt-hr Month Thermal Gen.
MWt-hr Cycle 1()
Cycle 11 Thermal Gen.
Month MWt-hr 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
'ec-80 Jan-81 Feb-81 Mar 81 Apr-81 May-81 238944 1084200 1120680 1119528 1084296 1124472 72 1035552 1043136 1018488 1118424 604920
'0 Jun-81 Jul-81 Aug 81 Sep-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
Cycle 12 Cycle 13 Cycle 14 Cycle 15 Month May 82 Jun-82 Jul-82 Aug-82 Sep42 Oct-82 Nov-82 Dec-82 Jan-83 Feb-83 Mar43 Apr 83 May43 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 153007 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-85 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 6-28
TABLE 6-7 cont'd MONTHLYTHERMALGENERATION DURING THE FIRST TWENTY-FIVE FUEL CYCLES OF THE GINNA REACTOR Cycle 16 Cycle 17 Cycle 18 Cycle 19 Month Mar46 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 87 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 Cycle 20 Thermal Gen.
Month MWt-hr Cycle 21 Month Thermal Gen.
MWt-hr Month Thermal Gen.
MWt-hr Cycle 22 Month Thermal Gen.
MWt-hr Cycle 23 May-90 Jun-90 Jul-90 Aug-90 Sep-90 Oct-90 Nov-90 Dec-90 Jan-91 Feb-91 Mar-91 Apr-91 769318 1012320
, 1098566 1098437 959083 1093051 1056218 715030 1096111 988894 668083 0
May-91 Jun-91 JUI-91 Aug-91 Sep-91 Oct-91 Nov-91 Dec-91 Jan-92 Feb-92 Mar-92 Apr-92 599282 1065760 1100696 1022214 1049792 1103223 1067033 1102728 1106865 780610 868877 0
May-92 Jun-92, Jul-92 Aug-92 Sep-92 Oct-92 Nov-92 Dec-92 Jan-93 Feb-93 Mar-93 610650 948520 1097720 1101108 1059758 1103684 1045291 1106157 1100087 1004079 386820 Apr-93 May-93 Jun-93 Jul-93 Aug-93 Sep-93 Oct-93
'ov-93 Dec-93 Jan-94 Feb-94 Mar-94 91944 1059620 1068915 1105566 1088384 1068747 1106564 731946 1099588 1088034 998757 116893 6-29
TABLE 6-7 cont'd MONTHLYTHERMALGENERATION DURING THE FIRST TWENTY-FIVE FUEL CYCLES OF THE GINNA REACTOR 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-9Q Thermal Gen.
MWt-hr 9
934804 1040844 1085722 975736 1055684 1093897 1057617 1095856 1096191 1024993 947615 30568 6-30
TABLE 6-8 MEASURED SENSOR ACTIVITIES AND REACTION RATES SURVEILLANCECAPSULE V
"SATURATEDACTIVITIES AND REACTION RATES Reaction
"'Cu (n.u)
Co Top Top Middle Bottom Middle Bottom
~Fe (n,p) ~Mn W
1 R I S6 P7 W2 R3 S8 P9 "Ni (n,p) "Co Middle
'-"U (n,f) '"Cs Middle "N (,f) '"C Middle Measured Activity
+d)Qs~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
~de
<~~m 4.64e+05 4.25e+05 4.70e+05 5.lie+05 5.00e+06 5.2 le+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
{r~sat~om 6.79e-17 6.23e-17 6.88e-17 7.48e-17 7.60e-l5 7.9 le-15 6.7 le-l5 7.91e-15 7.67e-15 7.33e-15*
7.59e-15 7.89e-15 8.83e-15 4.81e-14 2.45e-13 6-31
TABLE 6-8 cont'd MEASURED SENSOR ACTIVITIES AND REACTION RATES SURVEILLANCE CAPSULE R
'SATURATED ACTIVITIES AND REACTION RATES Reaction
"'Cu (n,a) "'Co Top Top Middle Bottom Middle Bottom "Fe (n,p) '"Mn W 13 R 14 P 18 W 14 R 15 P 19 s'Ni (n p) "Co Middle "Co (n,y)
Co
'Top Top Middle Middle Bottom Middle Bottom
'Co (n;y)
Co (Cd)
Top Top Middle Middle Bottom Middle Bottom 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 I:24e+07 Saturated.
Activity
~ds
~~m 4.42e+05 3.96e+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
'1.38e+07 1.26e+08'.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 Reaction Rate Q)~sat~om 6.47e-17 5.80e-17 6.89e-17 6.89e-17 7.88e-15 7.51e-15 7.81e-15 7.54e-15 7.87e-15 8.56e-15 1.00e-14 8.00e-12 8.13e-12 7.66e-12 7.61e-12 7.61e-12 3.21e-12 3.18e-12 2.88e-12 3.34e-12 3.34e-12
'-"U (n f) '"Cs Middle.
4.32e+05 7.81e+06 5.15e-14
'~N (,f) 'Cs Middle 4.25e+06 7.68e+07 4.83e-l3 6-32
TABLE 6-8 cont'd MEASURED SENSOR ACTIVITIES AND REACTION RATES SURVEILLANCECAPSULE T SATURATED ACTIVITIES AND REACTION RATES Reaction
'Cu (n,a)
Co Top Top Middle Bottom Middle Bottom Fe (n,p)
Mn S 22
, P28 W 21 S 23 P 29 W 22 "Ni (n,p) "Co Middle "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 Middle
'-"Np (n,f) '"Cs Middle Measured Activity
~d>> ~m 1.60e+05 1.40e+05 1.66e+05 1.74e+05 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 Saturated Activity
~d>> v~m 3.52e+05 3.08e+05 3.65e+05 3.83e+05 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 Reaction Rate QrpsSat~om 5.10e-17 4.47e-17 5.29e-17 5.55e-17 5.13e-15 5.71e-15 5.85e-15 5.50e-15 5.61e-15 5.99e-15 7.19e-15 4.32e-12 4.17e-12 4.13e-l2 4.46e-l2 4.19e-12'.68e-12 1.57e-12 1.62e-12 1.76e-12 1.67e-12 3.53e-14 2.76e-13 6-33
F TABLE 6-8 cont'd MEASURED SENSOR ACTIVITIES AND REACTION RATES SURVEILLANCECAPSULE S
SATURATED ACTIVITIES AND REACTION RATES Reaction
"'Cu (n,a)
Co Top Top Middle Bottom Middle Bottom "Fe (n,p)
Mn P 31 "Ni (n,p) "Co Middle "Co (n,y)
Co
~ Top Top Middle Middle Bottom Middle Bottom
'"Co (n,y)
Co (Cd)
Top Top Middle Middle Bottom Middle Bottom 238U ( f) l37C Middle
'""Np (n,f) '"Cs Middle Measured Activity
~d)+s~m 2.06e+05 1.82e+05 1.98e+05 2.18e+05 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 1.45e+07 1.35e+07 1.40e+06
- l. I le+07 Saturated Activity
+d)~sv~m 3.07e+05 2.71e+05 2.95e+05 3.24e+05 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 2.0 le+07 4.64e+06 3.68e+07
'I Reaction Rate isa~tom 4.44e-17 3.93e-17 4.27e-17 4.70e-17 4.46e-15 5.82e-15 3.31e-12 3.46e-12 3.16e-lg 3.36e-12 3.22e-12 1.38e-12 1.32e-12 1.26e-12 1.39e-12 1.30e-12 3.06e-l4 2.3 le-13 6-34
TABLE 6-9
SUMMARY
OF NEUTRON DOSIiVIETRY RES.ULTS SURVEILLANCE CAPSULES V. R. T AND S Measured Flux and Fluence for Capsule V
~uantit
[n/cm -sec]
(E > 1.0 MeV)
(E > 0.1 MeV)
(E < 0.414 eV) dpa/sec Flux 1.129e+11 4.426e+11 2.314e+11 2.062e-10
~uantit
[n/cm ]
(E > 1.0 MeV) 4 (E >0.1 MeV)
(E < 0.414 eV) dpa Fluence 5.028e+l8 1.971e+19 1.031e+19 9.183e-03 Uncertaint 10c/c 21'/c 83%
15%
Measured Flux and Fluence for Capsule R
~uantiti
[n/cm'-sec]
(E > 1.0 MeV)
(E > 0.1 MeV)
(E < 0.414 eV) dpa/sec Flux 1.374e+11 5.892e+11 1.964e+11 2.606e-10
[n/em"]
(E > 1.0 MeV)
(E > O.l MeV)
Cl (E < 0.414 eV) dpa Fluence 1.105e+19 4.738e+19 1.579e+19 2.096e-02 Uncertaint 8%
15%
20%
11%
Measured Flux and Fluence for Capsule T quantity
[n/cm"-sec]
(E > 1.0 MeV)
(E > 0.1 MeV)
(E < 0.414 eV) dpa/sec Flux 8.611e+10 3.250e+11 1.080e+11 1.528e-10
~uantiti
'n/cm
]
(E > 1.0 MeV)
(E > 0.1 MeV)
(E < 0.414 eV) dpa Fluence 1.864e+19 7.035e+19 2.338 e+19 3.307e-02 Uncertaint 8%
15%
19%
10%
Measured Flux and Fluence for Capsule S
~uantiti
[n/cm'"-sec]
(E > 1.0 MeV)
(E > O.l MeV)
(E < 0.414 eV) dpa/sec Flux 6.982e+10 2.743e+ll 8.341e+10 1.268e-10
~uantiti
[n/cm']
(E > 1.0 MeV)
(E > 0.1 MeV)
(E < 0.414 eV) dpa Fluence 3.746e+19 1.472e+20 4.475e+19 6.803e-02 Uncertaint 8%
15%
19%
11%
6-35
TABLE 6-10 COMPARISON OF MEASURED AND FERRET CALCULATED REACTION RATES AT THE SURVEILLANCECAPSULE CENTER Surveillance Capsule V Reaction Rate (rps/nucleus)
"Cu (n,tr,)
Fe (n,p)
~8Nj (n p)
"'U (n,f) (Cd)
Measured 6.84e-17 7.58e-I 5 8.84e-15 3.94e-14 Adjusted Calc.
6.73e-17 7.4 le-15 9.36e-15 3.71e-14 M/C
~Ad'usted 1.02 1.02 0.94 1.06 Surveillance Capsule R Reaction Rate (rps/nucleus)
"'Cu (n,a)
'Fe (n,p)
Ni (n,p) 3sU (n,f) (Cd)
"'Np (n,f) (Cd)
Co (n,y)
"Co (n,y) (Cd)
Measured 6.51e-17 7.86e-15 1.00e-14 4.14e-14 4.75e-13 7.80e-12 3.19e-12 Adjusted
~atc.
6.44e-17 7.77e-15 1.04e-14 4.23e-14 4.38e-13 7.81e-12 3.19e-12 M/C
~A'acted 1.01 1.01 0.96 0.98 1.08 1.00 1.00 6-36
TABLE 6-10 cont'd COMPARISON OF MEASURED AND FERRET CALCULATED REACTION RATES AT THE SURVEILLANCECAPSULE CENTER Surveillance Capsule T Reaction Rate (rps/nucleus)
"'Cu (n,a)
Fe (n,p)
~'Ni (n,p)
U (n f) (Cd)
"'Np (n,f) (Cd)
Co (n,y)
"Co (n,y) (Cd)
Measured 5.10e-17 5.63e-15 7.19e-15 2.77e-14 2.72e-13 4.26e-12 1.66e-12 Adjusted Cute.
5.06e-17 5.58e-l5 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)
'Ni (n,p)
U (n,f) (Cd)
"'Np (n,f) (Cd)
"Co (n,y)
Co (n,y) (Cd)
Measured 4.34e-17 4.46e-15 5.82e-15 2.21e-14 2.27e-13 3.30e-12 1.33e-12 Adjusted Calc.
4.26e-17
'.47e-15 5.98e-15 2.25e-14 2.lie-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 6-37 s
TABLE 6-11 ADJUSTED NEUTRON ENERGY SPECTRUM AT THE CENTER OF SURVEILLANCE CAPSULE Capsule V
~Grou I
2 3
4 5
6 7
8 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 Energy
~MeV 1.73e+01 1.49e+01 1.35e+01 1.16e+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
'.35e+00 l.lie+00 8.2 le-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 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
~Grou ¹ 28 29 30 31 32 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 Energy
~MeV 9.12e-03 5.53e-03 3.35e-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 Flux n cm"-sec 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+ll Note: Tabulated energy levels represent the upper energy in each group.
6-38
TABLE 6-11 cont'd ADJUSTED NEUTRON ENERGY SPECTRUM AT THE CENTER OF SURVEILLANCECAPSULE Capsule R
~Grou 1
2 3
4 5
67' 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 Energy
~MeV 1.73e+Ol 1.49e+0 1
1.35e+Ol 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 l.lie+00 8.21e-01 6.39e-01 4.98e-01 3.88e-01 3.02e-01 1.83e-01
- l. 1 le-01 6.74e-02 4.09e-02 2.55e-02 1.99e-02 1.50e-02 Flux n em=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 8.0le+10 7.24e+10 5.47e+10 4.42e+10 2.67e+10 2.48e+10 1.62e+10 2.6le+10
~Grou ¹ 28 29 30 31 32 33 34 35 36 37 38 39 40'1 42 43 44 45 46 47 48 49 50 51 52 53 Energy
~MeV 9.12e-03
'.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 Flux n cm-'-sec 2.78e+10 2.94e+10
- 9. 14e+()9 8.74e+() 9 8.58e+() 9 2.54e+1()
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 3.51e+10 9.66e+10 Note: Tabulated energy levels repres'ent the upper energy in each group.
6-39
TABLE 6-11 cont'd P
ADJUSTED NEUTRON ENERGY SPECTRUM AT THE CENTER OF SURVEILLANCECAPSULE 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 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 l.lie+00 8.21e-01 6.39e-01 4.98e-01 3.88e 3.02e-01 1.83e-01
- l. I le-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 3.44e+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 34 35 36 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 Energy
~MeV 9.12e-03 5.53e-03 3;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 8.3 le-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 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.
640
TABLE 6-11 cont'd ADJUSTED NEUTRON ENERGY SPECTRUM AT THE CENTER OF SURVEILLANCECAPSULE Capsule S
~Grou ¹ 1
2 3
4 5
6 7
8 9
10ll 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 Energy
~(Me V 1.73e+01 1.49e+01 1.35e+01 1.16e+01 1.00e+01 8.6le+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
- l. 1 le+00 8.21e-01 6.39e-01 4.98e-01 3.88e-01 3.02e-01 1.83e-01
- l. 1 le-01 6.74e-02 4.09e-02 2.55e-02 1.99e-02 1.50e-02 Flux n em=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 1.45e+10 1.57e+10 2.69e+10 2.99e+10 3.19e+10 2.16e+10 2.96e+10 3.61e+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
~MeV 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 Flux n cm--sec 1.27e+10 1.34e+1()
4.15e+09 3.95e+09 3.85e+09 1.13e+1()
- l. I le+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.
TABLE 6-12 COMPARISON OF CALCULATEDAND MEASURED INTEGRATED NEUTRON EXPOSURE OF GINNA SURVEILLANCE CAPSULES V, R. T, AND S 4(E > 1.0 MeV)
[n/cm"]
C?(E > (). I MeV)
[n/cm"]
dpa CAPSULE V Calculated 5.864e+18 2.223e+19 1.044e-02 Measured 5.028e+18 1.971e+19 9.183e-03
~MC 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
~MC 1.09 1.23 1.16 4(E > 1.0 MeV)
[n/cm~]
4(E > 0.1 MeV)
[n/cm']
dpa CAPSULE T Calculated 1.669e+19 5.741e+19 2.837e-02 Measured 1.864e+19 7.035e+19 3.307e-02 1.12 1.23 1.17 4(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
~MC 1.05 1.17 1.1 I 6-42
0 TABLE 6-13 NEUTRON EXPOSURE PROJECTIONS AT KEY LOCATIONS ON THE REACTOR VESSEL CLAD/BASE METAL INTERFACE Best Estimate Ex osure 19.51 EFPY at the Reactor Vessel Inner Radius 450 4 (E > I.().iVleV) 4 (E > ().I MeV) dpa 2.32e+19 1.97e+20 6.88e-02 1.47e+19 1.39e+20 4.68e-02 1.05e+19 1.00e+20 3.36e-02 9.69e+18 8.38e+19 2.86e-02 Best Estimate Ex osure 28 EFPY at the Reactor Vessel Inner Radius 4 (E > 1.0 MeV) 4 (E > 0.1 MeV) dpa 0'.11e+19 2.65e+20 9.24e-02 15'.96e+19 1.86e+20 6.24e-02 30o 1.40e+19 1.34e+20 4.49e-02 45'.29e+19 1.13e+20 3.86e-02 Best Estimate Ex osure 32 EFPY at the Reactor Vessel Inner Radius 4 (E > 1.0 MeV) 4 (E > 0.1 MeV) dpa 3.49e+19 2.97e+20 1.03e-01 15'.20e+19 2.08e+20 6.98e-02 1.56e+19 1.50e+20 5.02e-02 45'.45e+19 1.27e+20 4.33e-02 Best Estimate Ex osure 42 EFPY at the Reactor Vessel Inner Radius 4 (E > 1.0 MeV) 4 (E > 0.1 MeV) dpa 00 4.42e+19 3.77e+20 1.31e-01 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 Ves el Inner Radiu 4 (E > 1.0 MeV) 4 (E > 0.1 MeV) dpa 0'.98e+19 4.25e+20 1.48e-01 15'.13e+19 2.96e+20 9.93e-02 30o 2.22e+19 2.13e+20 7.14e-02 45'.06e+19 1.82e+20 6.22e-02 6P3
.TABLE 6-14 NEUTRON EXPOSURE VALUES WITHINTHE GINNA REACTOR VESSEL FLUENCE BASED ON E > 1.0 MeV SLOPE 28 EFPY 4 E > 1.0 MeV n cm-30o 450 Surface
'/4 T T
3.11e+19 1.99e+19 6.10e+18 1.96e+19 1.27e+19 4.05e+18 1.40e+19 8.89e+18 2.8le+18 1.29e+19 8.38e+18 2.67e+18 32 EFPY 4 E > 1.0 MeV n cm Surface
'/4 T
-~/~ T 0'.49e+19 2.23e+19 6.83e+18 15'.20e+19 1.41e+19 4.53e+18 30'.56e+19 9.94e+18 3.14e+18 45'.45e+19 9.37e+18 2.98e+18 42 EFPY E > 1.0 MeV n cm~
Surface
'/4 T
~/~ T 0'.42e+19 2.83e+19 8.67e+18 15'.78e+19 1.79e+19 5.72e+18,
~00 1.98e+19 1.26e+19 3.97e+18 45'.83e+19 1.18e+19 3.77e+18 48 EFPY E > 1.0 MeV n cm Surface
'/4 T
~/~ T Po 4.98e+19 3.18e+19 9.77e+18 15' 3.13e+19 2.01e+19 6.44e+18 30'.22e+19 1.41e+19 4.47e+18 45'.06e+19 1.33e+19 4.25e+18
TABLE 6-14 cont'd NEUTRON EXPOSURE VALUES WITHINTHE GINNA REACTOR VESSEL FLUENCE BASED ON dpa SLOPE 28 EFPY 4 E > 1.0 MeV n cm-Surface
'/~ T NT 3.lie+19 2.22e+19 9.37e+18 15'.96e+19 1.43e+19 6.44e+18 30'.40e+19 1.00e+19 4.50e+18 45'.29e+19 9.36e+18 4.17e+18 32 EFPY 4 E > 1.0 MeV n cm Surface
'/~ T
~/~ T 0'.49e+19 2.48e+19.
1.05e+19 15 2.20e+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 E > 1.0 MeV n cm 15'0'5'urface
'/~ T
~/~ T 4.42e+19 3.15e+19 1.33e+19 2.78e+19 2.02e+19 9.11e+18 1.98e+19 1.42e+19 6.36e+18 1.83e+19 1.32e+19 5.90e+18 48 EFPY 4 E > 1.0 MeV n cm Surface
'/4 T
'/~ T 0'.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+18 450 2.06e+19 1.49e+19 6.64e+18
TABLE 6-15 UPDATED LEAD FACTORS FOR GINNA SURVEILLANCECAPSULES
~Ce cele V(~(
T(b(
R(c(
S(~l N(e(
plel Lead Factor 2.97 1.82 1.77 1.78 1.89 fa]
- 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.
6-46
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6-50
impact other areas that were not the chemistry factor.
Thus,'he until a full review of all areas Docket No. 50-244 considered in the Westinghouse report such as review would focus on an interim operation could be completed.
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