ML051230443
| ML051230443 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 02/28/2005 |
| From: | Sitaraman S, Wu T General Electric Co |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| DRF 0000-0031-6251, NRC-05-0033 GE-NE-0000-0031-6254-R1-NP, Rev 1 | |
| Download: ML051230443 (28) | |
Text
ENCLOSURE 3 TO NRC-05-0033 TRANSMITTAL OF THE RPV NEUTRON FLUX EVALUATION REPORT GE REPORT, GE-NE-0000-031-6254-R1-NP,
'¶DTE ENERGY FERMI-2 ENRGY CENTER NEUTRON FLUX EVALUATION,"
FEBRUARY 2005 NON-PROPRIETARY VERSION l
GE Nuclear Energy General Eleddc Company 3901 Casde Heyne Road, VMimington, North Carolina 28401 GE-NE-0000-003 1-6254-RI-NP.
DRF 0000-0031-6251 Revision 1 Class I February 2005 DTE Energy Fermi-2 Energy Center Neutron Flux Evaluation Principal Contributor:
T.Wu Principal Verifier:
S. Sitaraman tIN.
GE-NE-00O-0031-6254-R1-NP NON-PROPRIETARY VERSION IMPORTANT NOTICE REGARDING THE CONTENTS OF THIS REPORT Please Read Carefully A. Disclaimer The only undertakings of the General Electric Company (GE) respecting information in this document are contained in the contract between the company receiving this document and GE. Nothing contained in this document shall be construed as changing the applicable contract. The use of this information by anyone other than a customer authorized-by GE to have this document, or for any purpose other than that for which it is intended, is not authorized. With respect to any unauthorized use, GE makes no representation or warranty, and assumes no liability as to the completeness, accuracy or usefulness of the information contained in this document, or that its use may not infringe privately owned rights.
B. Non-Proprietary Notice This is a non-proprietary version of the document GE-NE-0000-0031-6254-RI, which has the proprietary information removed. Portions of the document that have been removed are indicated by an open and closed bracket as shown here ((
].
i
GE-NE-0000-0031 -6254-RI-NP NON-PROPRIETARY VERSION'-
TABLE OF CONTENTS Page ACRONYMS AND ABBREVJATIONS..................
1.0 INTRODUCTION
........ 1.
1.1 BACKGROUND
I 1.2 SCOPE 1
2.0 METHODS AND ASSUMPTIONS......
2.1 FLUX CALCULATION METHODOLOGY
.2 2.2 SELECTION OF REPRESENTATIVE PRE-EPU CORE.....................................
n3 2.3 INPUTS AND ASSUMPTIONS
.3 2.3.1
'Core Loading...........
3 2.3.2 (RO) Model 3
2.3.3 (RZ) Model 4
2.3.4 Material Composition and CoolantDensity 5
2.3.5 Neutron Source Distribution 5
2.4
-BIAS ANDUNCERTAINTY..
6 3.0 FLUX RESULTS... _____. 1 3.1 NEUTRON FLUX AT RPV INSIDE SURFACE.12 3.2 NFUTRON FLUX AT SURVEILLANCE CAPSULE LOCATION..................
12 3.3 NEUTRON FLuX AT SHROUD INSIDE SURFACE.
1 3 3.4 NEUTRON FLUX AT TOP GUIDE AND CORE PLATE
.13
- 3.5 OBSERVATIONS.13 4.0 NEUTRON FLUENCE...............
5.0 REFERENCES
22 ii
GE-NE-00000031-6254-R1-NP
-. NON-PROPRIETARY VERSION...
v LIST QF TABLES Table Title Page Table 2-1 Design Input Data for Flux Calculation..................................
7 LIST OF FIGURES Title Fi zure Pace Figure 2-1.
Figure 2-2:
Figure 2-3.
Figure 3-1.
Figure 3-2.
Figure 3-3.
Figure 3-4.
Figure 4-1.
Figure 4-2.
Core Layout and Vessel Internal Components..................................................
9 Schematic View of (RZ) Model.................................................
10 Relative Cycle Energy at Core Midplane.................................................
11 Azimuthal Distribution of Fast Neutron Flux at RPV Inside Surface at Peak Elevation................................................
1 5 Axial Distribution of Relative Fast Neutron Flux at RPV Inside Surface........... 16 Azimuthal Distribution of Fast Neutron Flux at Shroud Inside Surface at Peakl Elevation.........
17 Axial Distribution of Relative Fast Neutron Flux at Shroud Inside Surface....... 18 Axial Distribution of Fast Neutron Fluence at RPV Inside Surface at the Peak Azimuth.......
20 Axial Distribution of Fast Neutron Fluence at Shroud Inside Surface at the Peak Azimuth.......
21 iiiI
GE-NE-0000-003 1-6254-RI-NP NON-PROPRMErARY VERSION ACRONYMS AND ABBREVIATIONS BAF Bottom of Active Fuel BOC Beginning of Cycle CLTP Current Licensed Thermal Power ECP Engineering Computer Program ENDF Evaluated Nuclear Data File EFPY Effective Full-Power Year EOC End of Cycle EPU Extended Power Uprate GE General Electric GE-NE General Electric Nuclear Energy IASCC Irradiation Assisted Stress Corrosion Cracking ID
-Inside Diameter LTR Licensing Topical Report MWt Megawatt Thermal NRC Nuclear Regulatory Commission OD Outside Diameter ORNL Oak Ridge National Laboratory RG Regulatory Guide RPV Reactor Pressure Vessel
(
iv
GE-NE-0000-003 1 -6254-Ri-NP NON-PROPRIETARY VERSION
1.0 INTRODUCTION
1.1 Background
Neutron irradiation of the reactor pressure vessel (RPV) causes reduction in material ductility and creates structural embrittlement at higher operating temperatures. 'The effect is particularly significant when impurities such as nickel, copper, or phosphorus are present in noticeable levels, as is commonly true for the RPV steel. Therefore determination of neutron fluence level is one of the first steps toward RPV fracture toughness evaluations.
Irradiation by fast neutrons (with energies greater than 1 MeV) can also be a concern with respect to irradiation assisted stress corrosion cracking (IASCC) for reactor internal components such as core shroud, top guide, core plate, etc. Crack growth evaluation for these components also requires adequate determination of neutron fluence level.
Neutron flux, or fluence rate, can be determined through radiochemical analysis of the surveillance flux wire samples; which are made of iron, copper, or nickel, sealed inside a capsule and held in place by a holder near the RPV inside surface. The calculated ratio of surveillance sample flux to the RPV peak flux defines a lead factor. This lead factor can be applied to the sample dosimetry data to determine the RPV peakl flux.
Neutron flux can also be determined by solving the Boltzmann neutron transport equation with the discrete ordinates method or Monte Carlo simulation. When appropriate bias is applied to the calculation results, these numerical methods can be used to obtain the best-estimate flux distribution.
1.2 Scope GE has performed an RPV. flux evaluation for the DTE Energy Fermi-2 plant to support a Fermi-2 Extended Power Uprate (EPU). Detailed flux calculations were performed for a representative pre-EPU cycle at the Current Licensed'Thermal Power (CLTP) of 3430 MWt and for an EPU core that is representative of future cycles at the target power level of 3952 MWt.
The NRC approved GE fluence methodology was used for these flux calculations. In addition, fluence distributions at the Effective-Full-Power Years (EFPY) of interest were evaluated based on the calculated CLTP and EPU flux.' distributions, in' conjunction with the cycle-dependent energy generation data. The objective of this report is to document the flux/fluence evaluation.
I
GE-NE-00000031-6254-RI-NP NON-PROPRIETARY VERSION 2.0 METHODS AND ASSUMPTIONS 2.1 Flux Calculation Methodology The current GE methodology for neutron flux calculation is documented in a Licensing Topical Report (LTR) NEDC-32983P-A [1], which was approved by the U.S.
NRC in the Safety Evaluation Report for referencing in licensing actions [2]. In general, GE's methodology described in the LTR adheres to the guidance in Regulatory Guide 1.190 [3] for neutron flux evaluation.
((I The flux calculations are performed with DORTGO1V, which is a controlled version of the ORNL DORT computer code in the GE Engineering Computation Program (ECP) library (4]. The NRC has approved the use of DORT as part of the GE methodology [2].
The cross section data used in the DORT calculation are processed with the nuclear cross-section processing package in the GE ECP. The basic cross section data library is the 80-group MATXS6D library, which was based on Version V of the Evaluated Nuclear Data File (ENDF/B-V) and was generated by the Los Alamos National Laboratory. This 80-group library has been revised to include upgraded cross sections from ENDFIB-VI for the important components of BWR neutron flux calculation - oxygen, hydrogen, and individual iron isotopes - to meet the guidelines of Regulatory Guide 1.190, Regulatory Position 1.1.2.
)) A P3 truncation of the Legendre polynomial expansion, which meets Regulatory Position 1.1.2 in RG 1. 190, is used to approximate the anisotropy in the differential scattering cross sections. The approach
-discussed here is consistent with the approved LTR methodology.
2
GE-NE-0000-0031-6254-RI-NP NON-PROPRIETARY VERSION 2.2 Selection of Representative Pre-EPU Core It was agreed between GE and DTE Energy that a single representative core would be selected to be the basis for the pre-EPU flux calculations. In order to determine the cycle that would be used, a series of flux calculations were performed for Cycles 1-9 based on the core data retrieved from the GE Engineering Data Bank.
The peak RPV ID flux was calculated in each case and used as the basis for selecting the representative pre-EPU cycle.
The results of calculated peak RPV ID fluence for Cycles 1-9 indicate that Cycle 6 is the suitable choice to represent the pre-EPU cycles. Using Cycle 6 as the representative cycle for the pre-EPU cycles would impose a conservatism of about 10% on the pre-EPU fluence.
2.3 Inputs and Assumptions The operating condition assumed for this analysis is based on the specific core design data of the cycles of interest. Pertinent core data were retrieved and processed from GE's Engineering Data Bank.
These data include the three-dimensional descriptions of fuel bundle configuration, initial uranium mass, fuel exposure, and exposure-dependent water density (or void fraction) data.
The RPV and shroud dimensions, jet pump dimensions/configuration, and the configuration and location of surveillance capsule holder are verified data through the Design Input Request (DIR) and DTE Energy response/concurrence. These data are summarized in Table 2-1.
2.3.1 Core Loading The Cycle 6 core contains 92 bundles of GE6, 47 bundles of GE9, and 625 bundles of GE11. The active fuel length is 150 inches for GE6 and GE9, and 146 inches for GE1.
Fuel bundles in the outermost row contain only GE6 fuel. The axial flux shape for the RPV and shroud will be mostly'impacted by these peripheral bundles.
The El.] cycle contains a full core load of GE14 fuel. The active fuel length is 150 inches.
2.3.2 (RO) Model Figure 2-1 shows a quadrant of the core and the vessel internal components that are relevant to the flux calculation. ((
3
GE-NE-0000-0031-6254-RI-NP NON-PROPRIErARY VERSION 1))
In the angular coordinate 0, the mesh size is 1/2 degree per mesh interval, except for the two boundary nodes, which is 1/4 degree. For a core quadrant, a total of 181 fine meshes are used in the 0-direction..This model is more detailed than the minimum requirement of 40 intervals per octant stipulated in Section 1.3.1 of RG 1.190.
Radial meshes vary in sizes. Sufficient fine mesh steps are provided to simulate the outer boundary profiles of the core. ((
- ] The radial mesh size for each region has been selected to conform to the minimum requirements of RG 1.190, Section 1.3.1. [
The (re) calculation is performed for the first quadrant of the reactor. The model includes three sets of jet-pump/riser, which are centered at 300, 60°, and 90°. It also includes the 300 capsule, which is completely shadowed by the jet-pump riser.
The azimuthal distribution of RPV flux in the first quadrant is expected to be representative of other quadrants because of the quadrant symmetry in the power distribution. The RPV flux at 0 and 180 degrees is expected to be higher than that at 90 and 270 degrees, due to the.absence ofjet pumps.
2.3.3 (RaZ) Model 4
GE-NE-0000-0031-6254-RI-NP
- . i.
NON-PROPRIETARY VERSION
)) The radial mesh size for each region has been selected to conform to the minimum requirements of RG 1.190, Section 1.3.1.
2.3.4 Material Composition and Coolant Density Material compositions in each calculation model are treated as homogeneous mixtures. The volume fractions of solid material in each core region are calculated based on specific fuel bundle design data. ((
-], which is NRC approved [5].
[i i
2.3.5 Neutron Source Distribution Spatial distribution of the neutron source density is simulated based on relative cycle-integrated energy production in conjunction with the fission energy and fission yield data.
5
I GE-NE-0000-0031-6254-RI-NP NON-PROPRIETARY VERSION
))
Figure 2-3 shows the bundle-dependent relative cycle energy (normalized to the core average) at core midplane elevation (axial node 13) for Cycle 6 and the EPU cycle.. These discrete data in the (x,y) plane are processed to form the neutron source distribution in the (r,O) meshes for the (rO) calculation. The source distribution for the (rz) calculation is generated from a similar process.
2.4 BiAs and Uncertainty
))
6
GE-NE-0000-0031-6254-Ri-NP NON-PROPRIETARY VERSION Table 2-1 Design Input Data for Flux Calculation Parameter:
Unit Data (A) Surveillance Capsule Basket Thickness inch 0.75 Width inch 7
Length inch 9.5 Azimuths degree 30, 120, 300 Radial Position of Basket Center, from inch 0.9375 RPV Clad Elevation, Bottom of Basket inch 283.563 (B) RPV (Beltline Region)
RPV Base Metal ID inch 254 RPV Thickness inch 6.125 RPV Clad Thickness, inch 0.3125 (C) Shroud (Active Fuel Region)
Shroud ID inch 203.125 Shroud Thickness inch 2"
(D) Jet Pump Components Recirc. Inlet Nozzles Azimuths degree 30, 60, 90, 120, 150, 210, 240, 270,300, 330 Jet Pump Riser Pipe OD inch 10.71 Jet Pump Riser Pipe ID inch 10.136 Jet Pump Inlet Mixer OD inch 8.734 Jet Pump Inlet Mixer ID inch 8.18 Radial Location of Riser from Core inch 112.28 Center Radial Offset, Mixer to Riser inch 0
Distance between Mixers, centerline inch 29.5 to centerline 7
GE-NE-O-00031-6254-RI-NP NON-PROPRIETARY VERSION Table 2-1 Design Input Data for Flux Calculation (Cont.)
Parameter
-Unit-Dat.
(E) Top Guide Beam Lower Elevation inch 364.62 Upper Elevation inch 380.69 (F) Core Plate Elevation of top of core plate support inch 191.12 flange Upper Elevation inch 211.31 Thickness inch 2
(G) Other Core Data Thermal Power MWt 3430 (CLTP) 3952 (EPU)
Elevation of Bottom of Active Fuel inch 216.313 (BAF)
Active Fuel Length inch 150. (GE6, GE9, GE14) 146 (GEI 1) 8
GE-NE-000O-003 1-6254-RI-NP NON-PROPRIETARY VERSION Figure 2-1. Core Layout and Vessel Internal Components 00 J
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GE-NE-0000-0031-6254-RI-NP NON-PROPRIETARY VERSION Figure 2-2: Schematic View of (RZ) Model
((
JI.
.10
GE-NE-OO00 31-6254-RI-NP NON-PROPRIETARY VERSION Figure 2-3. Relative Cycle Energy at Core Midplane EPU Cycle:
JA 1
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GE-NE-000-003 1-6254-RI-NP NON-PROPRIETARY VERSION 3.0 FLUX RESULTS This section documents the flux results from the calculation of the representative pre-EPU core and the EPU core. Per agreement with DTE Energy, a factor of 1.1 has been applied to the calculated EPU flux values to cover any potential non-conservatism owing to the use of the equilibrium cycle core design developed for the EPU project.
3.1 Neutron Flux at RPV Inside Surface The calculated peak fast flux at the RPV inside surface is shown below:
Parameter Elev tion Aiimu Flux (Inches above
)
(n/cm2 -s)
JBAF)
RPV ID peak flux (>1.0 MeV) - EPU 85.1 64.0
.1.01E9 RPV Dpeak flux (>1.O MeV) - CLTP 101.6 26.5 8.73E8 Figure 3-1 shows the azimuthal flux distribution -at the peak elevation for the first quadrant of the RPV. The effects of inelastic scattering by steel in the jet-pump components are clearly displayed in Figure 3-1, where the flux depression occurs in regions shadowed by metal components. Fluxes for other quadrants are expected to be similar due to quadrant symmetry of the core loading as well as the arrangement of vessel internal components.
Figure 3-2 shows the axial flux variation at the RPV inside surface. The ratio of peak flux to midplane (75 inches above the BAF) flux is 1.06 for CLTP and 1.01 for EPU. These axial adjustment factors have been incorporated in the calculation of the peak flux.
3.2 Neutron Flux at Surveillance Capsule Location The calculated fast flux and lead factor at the capsule location is shown below:
CapsuzeNo.
Azimuh Flux LedFactor (0)
(nlcm 2.-s) 1, 2, and 3 - EPU 30, 120, and 300 1.06E9 1.05 1, 2, and 3 - CLTP 30, 120, and 300.
9.11E8 1.04 12
GE-NE-0000-003 1-6254-RI-NP NON-PROPRIETARY VERSION 3.3 Neutron Flux at Shroud Inside Surface The calculated peak fast flux at the shroud inside surface is shown below:
Parameter e-,Eletion-Aimuth Fux (Inches above
(
(n/cms)
-BAF)Y Shroud ID peak flux (>1.O MeV) - EPU 78.4 65.5 2.78E12 Shroud ID peak flux (>1.O MeV)- CLTP 109.5 24.5 2.37E12 Figure 3-3 shows the azimuthal flux distribution at the core midplane elevation for the first quadrant of the shroud. Fluxes for other quadrants are expected to be similar due to quadrant symmetry of the core loading.
Figure 3-4 shows the axial flux variation at the shroud inside surface. The ratio of peak flux to midplane (75 inches above the BAF) flux is 1.10 for CLTP and 1.00 for EPU.
These axial adjustment factors have been incorporated in the calculation of the peak flux.
3.4 Neutron Flux at Top Guide and Core Plate The estimated bounding fast fluxes at the top guide and core plate are shown below:
Parameter
.'Flux (nlcm2 )
Top Guide bounding flux (>1.O MeV) - EPU 1.06E13 Top Guide bounding flux (>1.0 Mey) - CLTP 5.81El2 Core Plate bounding flux (>1.0 MeV) - EPU 7.74E11 Core Plate bounding flux (>1.0 MeV) - CLTP 6.32E1 I
[1
' ]
3.5 Observations Compared to the CLTP flux, EPU flux increase is approximately 16% at the RPV ID and capsule locations, which is approximately the same as the percent change in power from CLTP to EPU. In the case of the shroud the increase is 17% going from CLTP to EPU. The peak elevation for the shroud and RPV flux distribution shifts towards the core midplane for the EPU, due to the core load of GE14 fuel with part-length rods.
- 13
GE-NE-0000-003 1-6254-Ri-NP NON-PROPRIETARY VERSION The calculated EPU lead factor of 1.05 is not significantly different from the CLTP lead factor of 1.04. Variation of the lead factor is expected due to change in the axial and azimuthal flux distributions that affect both the capsule flux and the peak RPV flux.
.14
GE-NE-0000-0031-6254-RI-NP NON-PROPRIETARY VERSION Figure 3-1. Azimuthal Distribution of Fast Neutron Flux at RPV Inside Surrace at Peak Elevation 1.2E+09
.0E+09 -
-~8.OE+08-6.0E+08-Z A n=.AR 0
10 20 30 40 50 Azimuth (degrees) 60 70 80 90 15
GE-NE-0000-0031-6254-RI-NP NON-PROPRIETARY VERSION I
iI I
Figure 3-2. Axial Distribution of Relative Fast Neutron Flux at RPV Inside Surface X
MIL 4,
o.o I
I I
I
.I
-25 0
25 50.
75 100 125 150 175 Elevation Above BAF (inches) 16
GE-NE-0000-003 1-6254-Ri -NP NON-PROPRIETARY VERSION Figure 3-3. Azimuthal Distribution of Fast Neutron Flux at Shroud Inside Surface at Peak Elevation
'At x
4-z 0
10 20 30 40 50 Azimuth (degrees) 60 70 80 90 17 r
GE-NE-0000-0031-6254-RI-NP NON-PROPRIETARY VERSION Figure 3-4. Axial Distribution of Relative Fast Neutron Flux at Shroud Inside Surface 1.1
- 1.
I I
I I = =
X m
A, m.a 1.0 0,9 0.8 0.7 0.6 0.5.
0.2 RPV ID, Pre-EPU 0 1
-f --RPV ID, Post EPU
=
-25 0
25 50 75 100 125 150 175 ElevatIon Above BAF (Inches) 18
GE-NE-0000-0031-6254-R1-NP NON-PROPRIETARY VERSION 4.0 NEUTRON FLUENCE The flux results reported in this evaluation are used to calculate the fluences for the target EFPY of interest. Based on the cycle power and cycle energy generation data provided by DTE Energy, the total energy generated for Cycles 1-10 is 15,089 GWd. This is equivalent to 12.04 EFPY at the CLTP of 3430 MWt. The flux values at CLTP will be used for this portion of the plant lifetime for the calculation of the fluence. The remainder of the lifetime is conservatively assumed to be operating at the EPU power level.
Using the notation 'FCLTP as the flux at CLTP and 4)EPU as the flux for EPU, fluence can be calculated as 32-EFPY Fluence = (DPcL-
- 12.04 + 4Epu
- 19.96)*(365.25*24*3600) n/cm2.
24-EFPY Fluence = (OcLTP
- 12.04 +.()EPU
- 11.96)*(365.25*24*3600) n/cm2.
The calculated fluence values are shown below:
Parameter i.32-EFJPY Fuence 24-EFPY Fluence (n/cm2)
(n/cm2)
RPV ID peak fluence (>1.0 MeV) 9.68E17 7.13E17 Shroud ID peak fluence (>1.0 MeV) 2.65E21 1.94E21 Top guide bounding fluence (>1.0 MeV) 8.87E21 6.20E21 Core plate bounding fluence (>1.0 MeV) 7.28E20 5.32E20 Girth weld (elevation 28.3 125 inches above 6.23E17 4.66E17 BAF) peak fluence (>1.0 MeV)
Figure 4-1 shows the axial fluence distribution at the peak azimuth at the RPV inside surface.
Two sets of fluence data are presented: at the end of 40 years (32 EFPY) and at intermediate time of 24 EFPY, which will be used for the P-T curve evaluation. Figure 4-2 shows similar data for the shroud inside surface.
Figure 4-1 shows that, at the end of 40-year plant life, the region of fast fluence greater than 1.0E17 n/cm2 extends from 5.8 inches below and 158.4 inches above BAF elevation. For the 24 EFPY, the region extends from 3.5 inches below and 155.2 inches above BAF elevation.
19
GE-NE-0000-0031.6254-R1-NP NON-PROPRIETARY VERSION Figure 4-1. Axial Distribution of Fast-Neutron Fluence at RPV Inside Surface at the Peak Azimuth 1.2E+18 1.OE+18 8.0E417 6.OE+17 4.OE+1 7 5e U.
i z
2.OE+17 O.OE+00
-25 0
25 50 75 100 125 Elevation Above BAF (Inches) 150 175 20
GE-NE-0000-003 1 -6254-RI-NP NON-PROPRIETARY VERSION Figure 4-2. Axial Distribution of Fast Neutron Fluence at Shroud Inside Surface at the Peak Azimuth I
3.0E+21 2.5E+21
_ 2.0E+21 C
c 1.5E+21 in z
1.OE+21-_
5.0E+20 -
O.OE+00 -
-25 0
25 50 75 100 125 Elevation Above BAF (Inches) 150 175 21
GE-NE-OOO-003 1-6254-RI-NP NON-PROPRIETARY VERSION
5.0 REFERENCES
- 1. NEDC-32983P-A, Rev. 1, "Licensing Topical Report, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluations," December 2001.
- 2. Letter, S. A. Richards (USNRC) to J. F. Klapproth, "Safety Evaluation forNEDC-32983P, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation (TAC No. MA9891)," MFN 01-050, September 14, 2001.
- 3. Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," U.S. NRC, March 2001.
- 4. CCC-543, "TORT-DORT Two-and Three-Dimensional Discrete Ordinates Transport Version 2.8.14," Radiation Shielding Information Center (RSIC), January 1994.
- 5. Letter, S. A. Richards (USNRC) to G. A. Watford, "Amendment 26 to GE Licensing Topical Report NEDE-2401 1-P-A, GESTAR II - Implementing Improved GE Steady-State Methods (TAC No. MA648 1)," November 10, 1999.
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