NL-04-0073, Fluence Calculations and Methodology for the Unit I Updated Analysis of Core Shroud Vertical Welds
| ML040350557 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 01/30/2004 |
| From: | Sumner H Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-04-0073 | |
| Download: ML040350557 (24) | |
Text
H.L Sumner, Jr. Southern Nuclear Vice President Operating Company, Inc.
Hatch Project Post Office Box 1295 Birmingham, Alabama 35201 Tel 205.992.7279 SOUTHERNZ January 30, 2004 COMPANY Energy to Serve Your World's Docket No.: 50-321 NL-04-0073 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant Fluence Calculations and Methodology for the Unit I Updated Analysis of Core Shroud Vertical Welds Ladies and Gentlemen:
By letter dated November 14, 2003, Southern Nuclear Operating Company (SNC) provided to the NRC the results of the updated analysis of the Edwin l. Hatch Nuclear Plant Unit I flawed Core Shroud Vertical welds.
Subsequently, during a phone conversation in December 2003, the NRC requested a copy of the calculation and a description of the methodology. A copy of the calculation is enclosed. This Hatch Unit I- specific calculation is not considered proprietary.
However, the methodology, which is referenced in the enclosed calculation, is described in BWRVIP proprietary documents previously transmitted to the NRC. Those documents are listed below.
As noted in the NRC Agencywide Documents Access and Management System (ADAMS), the B MR VIP-115: BWVR Vessel and Internals Project,RAAfA Fluence Methodology Benchmnark Manutal - Evaluation of Regulatory Guide 1.190 Benchmnark Problenms (reference 3 of the enclosed calculation) was transmitted to the NRC on June 26, 2003 (Accession Number ML031820253). The B WiT VIP-] 17: BIER Vessel and InternalsProject,RAMA Fluence AMethodology PlantApplication- Susquehanna Unit 2 Surveillance CapsuleFluence Evaluationfor Cycles 1-S (reference 4 of the enclosed calculation) was transmitted to the.NRC on August 5, 2003 (Accession Number ML032230313). The BJ1RVIP-114: BIVR Vessel and InternalsProject,RAAMA Fluence AMethodolog Th7eony Manual was transmitted on June 11, 2003 (Accession Number ML031640194). An Errata to BJVRVI'P-I14 was sent on August 21, 2003 (Accession Number ML032380140). Additionally, the BVR VIP-121: BHR Vessel and Internals Project,RAMA FluenceMethodology ProceduresMfanual was transmitted on October 29, 2003 (Accession Number ML033040182).
U. S. Nuclear Regulatory Commission NL-04-0073 Page 2 This letter contains no NRC commitments. If you have any questions, please advise.
Sincerely, H. L. Sumner, Jr.
HLS/whc/daj
Enclosure:
TransWare Report Number TWE-HATCHI-001-R-001, Revision 0 cc: Southern Nuclear Operating Company Mr. J. B. Beasley, Jr., Executive Vice President Mr. G. R. Frederick, General Manager - Plant Hatch Document Services RTYPE: CHAO2.004 U. S.Nuclear Regulatory Commission Mr. L. A. Reyes, Regional Administrator Mr. S. D. Bloom, NRR Project Manager - Hatch Mr. D. S. Simpkins, Senior Resident Inspector - Hatch
EVALUATION OF >1.0 MEV FLUENCE IN THE HATCH 1 SHROUD TransWare Report Number: TWE-HATCHl-001-R-001 Revision Number: 0 Prepared By:
Dean B. Jones TransWare Enterprises Inc.
5450 Thornwood Dr., Ste. M San Jose, California 95123 Prepared For:
Southern Nuclear Operating Company, Inc.
PO Box 1295 Birmingham, AL 35201 September 10, 2003
TWE-HATCH-OO1-R-OO1 Revision 0 Page 2 of 22 DISCLAIMER OF WARRANTIES AND LIMITATION OF LIABILITIES TRANSWARE ENTERPRISES INC. MAKES NO REPRESENTATIONS OR WARRANTIES WITH RESPECT TO THE CONTENTS HEREOF AND SPECIFICALLY DISCLAIMS ANY IMPLIED WARRANTIES OF FITNESS OR MERCHANTABILITY FOR ANY PARTICULAR PURPOSE. IN NO EVENT SHALL TRANSWARE ENTERPRISES INC. BE LIABLE FOR ANY LOSS OF PROFIT OR ANY OTHER COMMERCIAL DAMAGE, INCLUDING BUT NOT LIMITED TO SPECIAL, CONSEQUENTIAL OR OTHER DAMAGES.
ORGANIZATION THAT PREPARED THIS DOCUMENT:
TRANSWARE ENTERPRISES INC.
5450 THORNWOOD DRIVE, SUITE M SAN JOSE, CA 95123 K>
i, 7WE-HATCH-OO1-R-OO1 Revision 0 Page3 of 22 K> Purpose The purpose of the calculation documented in this report is to estimate the >1.0 MeV neutron fluence in the Hatch 1 shroud structure. The regions of interest in the shroud are locations where cracks have been documented in the Core Shroud Weld Identification Layout diagram [1].
Objective The objective of the calculation is to estimate the >1.0 MeV neutron fluence at observed locations in the Hatch 1 shroud where cracking has been observed. The calculated fluence will be compared to the fracture toughness threshold of 1.0 x 1021 n/cm2 .
Model Description For this study, a model shall be developed to calculate estimated neutron fluence values for energy >1.0 MeV at specified locations in the Hatch 1 core shroud structure. The locations of interest in the shroud are the azimuthal, elevation, and radial locations where cracks have been visually verified and documented in the Core Shroud Weld Identification Layout diagram [1].
Figures 1 and 2 illustrate the vertical and horizontal crack locations evaluated in this report.
Figure 1 identifies the cracks in azimuthal angles 0 to 180 degrees and Figure 2 identifies the cracks in angles 180 to 360 degrees.
The neutron fluence calculations were performed using the RAMA Fluence Methodology software [2] that is under development by EPRI and the BWRVIP. The RAMA methodology has been developed and benchmarked [3, 4, 5] in accordance with the intent of U.S. NRC Regulatory Guide 1.190 [6]. The methodology has been submitted to the U.S. NRC for review. As of the date of this report, the methodology has not been approved for safety-related calculations.
The RAMA Fluence Methodology computer model assumes azimuthal symmetry in the reactor geometry. The reactor is represented by a north-northeast octant symmetric model that ranges from 0 to 45 degrees, where 0 degrees corresponds to the north compass direction specified in the reactor plan drawings. Figure 3 illustrates a top view of the reactor model cross section that corresponds to the core mid-plane axial elevation.
The computer model used in this analysis omits the regions outside the reactor pressure vessel, i.e., the cavity, mirror insulation, and biological shield regions. The assumption has no affect on the shroud calculation, as the downcomer region provides an adequate boundary condition for the analysis. The purpose of the assumption was used to allow the three-dimensional calculations to be performed within the allotted time for the analysis. When time perrmits, the model can be expanded to include these regions.
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TVE-HATCH-OO-R-OOJ Revision 0 Page 6 of 22 To do the analysis, the azimuthal positions of the cracks shown in Figures 1 and 2 were translated to the symmetrical azimuthal positions of the model that is illustrated in Figure 3.
Table 1 identifies the cracks and provides the reactor actual and translated azimuthal angles for the model.
00 433.59 al -
392.97 cm-450 316.61 cm -
308.039 an 307.88 cn 294.0051 cm -
280.3526 an 279.5588 cm 248.8118 cm -
225.425 cm -
221.615 cm -
198.12 a -
26 25 24 23 22 21 20 19 18 17 16 15
= 14 ii = 14 15 16 17 18 19 20 21 22 23 24 25 26 Figure 3 Top View of the Hatch 1 Computer Model
( ( TWE-HATCH-OQI( 01 Revision 0 Page 7 of 22 Table 1 List of Shroud Cracks ih f Go'c"a"
".6 ml M L V04-01 Inside 2" 324 36 5.08 cm ( 2.0") 807.415 - 812.495 cm ( 317.9 - 319.9")
0 V05-01 Outside 13.5" 0
47 43 .34.29 cm ( 13.5") 722.325 - 756.615 cm ( 284.4 - 297.9")
V05-02 Outside 2" 0
47 43 5.08 cm ( 2.0") 660.095 - 665.175 cm ( 259.9 - 261.9")
V05-03 Outside 3/4" 0
47 43 1.905 cm ( 0.75") 648.030 - 649.935 cm ( 255.1 - 255.9 )
V05-04 Outside 1A" 47 43 0.635 cm ( 0.25") 603.580 - 604.215 cm ( 237.6 - 237.9 ):-
0 V05-05 Outside 2" 47 43 - 5.08 cm ( 2.0") 594.055 - 599.135 cm ( 233.9 - 235.9")
0 V05-06 Outside 1.5" 47 43 3.81 cm ( 1.5") 575.005 - 578.815 cm ( 226.4 - 227.9")
0 V05-07 Outside 1.5" 47 43 3.81 cm ( 1.5") 554.025 - 557.835 cm ( 218.1 -j 219.6 ")
V05-08 Inside 3/4"4- 0 53 - 37 1.905 cm ( 0.75") 675.970 - 677.875 cm ( 266.1 - 266.9")
V05-09 Outside 1.5" I. 53 37 3.81 cm ( 1.5") 554.025 - 557.835 cm ( 218.1 - 219.6")
V06-01 Inside `. 0 - 228 42 - 1.27 cm ( 0.5") -709.625 - 710.895 cm ( 279.4 - 279.9")
V06-02 Outside 11.5" 0 -233 37. 29.21 cm 11.5") 778.205 - 807.415 cm ( 306.4 - 317.9")
V06-03 Outside 25.5" 0 232 38 64.77 cm ( 25.5") 710.895 - 775.665 cm ( 279.9 - 305.4")
V06-04 Outside 5.25" 0
232 38 13.335 cm ( 5.25") -669.620 - 682.955 cm ( 263.6 - 268.9")
V06-05 Inside 2" 232 38 5.08 cm ( 2.0") -557.835 - 562.915 cm ( 219.6 - 221.6").
VO-01 Outside ta2 - 322 38 1.27 cm ( 0.5") 550.215 - 551.485 cm ( 216.6 - 217.1).
H05-01, Inside 10" I 228.5-231.5 38.5-41.5 25.4 cm (10.0") 557.835 cm ( 219.6-)
TWE-HATCH-OO-R-OOJ Revision 0 Page 8 of 22 The reactor geometry and reactor operating data for the fluence calculation were taken from design inputs generated by Brookhaven National Laboratory (BNL) [5]. BNL generated the design inputs for the purpose of benchmarking computer methods against neutron fluence measurements that were taken from the Hatch 1jet pump riser brace pads at the end of operating cycle 19. The operating data was specifically prepared to evaluate the brace pad measurements that were taken about 30 cm (12 inches) above the core mid-plane elevation. As a result, some of the operating data was prepared specifically to represent the operating conditions near the core mid-plane elevation. Following are discussions regarding modeling assumptions used in the jet pump riser brace pad model and how they might affect this analysis.
- The >1.0 MeV neutron fluence calculated for the Hatch 1 shroud is based on the RAMA Fluence Methodology computer model developed in [5]. The objective of the work was to calculate the >1.0 MeV neutron fluence at the jet pump riser brace pad locations. The brace pads are located at elevation 742.95 cm (292.5 inches), which is the centerline elevation for the pads. The pads are about 11.43 cm (4.5 inches) in height. The elevation of the brace pads is within about 30 cm (about 12 inches) of the core axial mid-plane, specified as elevation 712.6224 cm (280.56 inches). The results of the RAMA Hatch 1 benchmark show that the RAMA model over-predicted the >1.0 MeV fluence by about 17%. Thus, some conservatism exists in the current RAMA Hatch 1 model at the elevations near the jet pump riser brace pads.
- The jet pump riser brace pads are situated at about 30 cm (12 inches) above the core mid-plane elevation. This allowed certain assumptions to be made in the preparation of the Hatch 1 geometry model and reactor operating data. These assumptions are documented in [5]. Specifically, the reactor operating data for the Hatch 1 benchmark was prepared to reflect the operating conditions near the core mid-plane elevation. Operating data for off mid-plane elevations is not accurately represented. Consequently, the operating data that was prepared for the brace pad elevation is not fully applicable for the Hatch 1 shroud crack analysis that spans the full axial height of the reactor core region.
- The reactor operating history is modeled using 19 cycles of reactor operating data. Each cycle is characterized with one data set. This implies that a cycle-averaged, core axial power shape is used in the flux/fluence calculations. The axial power shape of a reactor can vary significantly throughout a cycle, therefore, the data used in this analysis may not provide an accurate representation of the core power for the shroud fluence evaluation.
The affect of this approximation is not known.
- The core power distribution is grossly approximated for the first 12 operating cycles.
Fairly detailed core power distribution data is provided for cycles 13 through 19.
Following are discussions on the power distribution data used in the model:
- The axial powers for the fuel assemblies in cycles 1 through 4 assumed that the active fuel height was 365.76 cm (144.0"). The axial powers for these cycles were provided in four 91.44 cm (36.0") axial segments.
TWE-HATCH-OO1-R-OO1 Revision 0 Page9 of 22
- The axial powers for the fuel assemblies in cycles 5 through 12 assumed that the active fuel height wa-s381.0 cm (150.0"). The axial pow-ers for these cycles were provided in three 91.44 cm (36.0') and one 106.68 cm (42.0") axial segments.
- The axial power data provided for cycles 1 through 12, which are represented with 4 axial power segments, represents a gross approximation of the core axial power shapes. Axial segments 2 and 3, which span the middle elevations in the core region (621.1824 cm to 804.0624 cm), provide a fair approximation of the power shape (i.e.,
+/-10%) around the core mid-plane elevation. Axial segments 1 and 4, which represent the powers in the bottom and top axial segments, respectively, are poor approximations because the average powers in these axial segments do not accurately represent the power gradients at the core boundaries. This is particularly important when evaluating the fluence in the lower regions near elevation H-5 (557.8348 cm). It is estimated that the shroud fluence could be over-estimated by 200-300% at the lower elevations, thus, providing a conservative result at these elevations.
- The axial powers for the fuel assemblies in cycles 13 through 19 assumed that the active fuel height was 381.0 cm (150.0"). The axial powers for these cycles were provided in twenty-five 15.24 cm (6.0") axial segments. Cycles 13 through 19 data is represented with 25 axial power segments and provides the best representation of axial power distribution data that is commonly available from core design methods.
- The core operating data provided for the jet pump riser brace pad benchmark assumed a constant moderator void fraction of about 40% in all regions of the reactor core. A moderator void fraction of 40% is representative of the core average voids around the core mid-plane elevation. Actual moderator coolant densities through the core axial height range from slightly sub-cooled at core entry to about 80% voids at core exit. The moderator void fraction in the peripheral assemblies, which are the primary neutron source for the determination of the shroud fluence, will be slightly lower than the core average due to different flow orificing in the peripheral fuel locations and lower power generation in the peripheral fuel assemblies. Thus, it is expected that the calculated neutron flux' will be under-estimated in' the upper core elevations due to higher than normal moderator densities and over-estimated in the lower core elevations due to lower than normal water densities.
- The core operating data that was generated for the jet pump'iser brace pad benchmark assumed a computer model that is based on traditional discrete ordinate transport methods. These methods employ correlations (approximations) that assume a fission spectrum for the rea'ctor'core that is based on core exposure. The RAMA Fluence Methodology employs a more-accurate neutron source treatment that uses actual fuel uranium and plutonium number densities in the core region to calculate the fission spectrum and neutron source. In order to perform the jet pump riser brace pad analysis with RAMA, it was necessary to generate fuel isotopic data using a fuel assembly lattice code. The effect of using the calculated fuel isotopics data instead of fuel exposure data to obtain the fission spectrum is not known.
TWE-HATCH-OO1-R-OOJ Revision 0 Page 10 of 22
- The reactor operating data provided for the jet pump riser brace pad benchmark omitted plant-specific data for the ex-core regions. Of particular importance is the water densities used in the core reflector, downcomer, and jet pump pipes. Generic water density information was taken from [7] and was used instead of plant-specific water densities.
The effect of this assumption cannot be assessed until plant-specific water density inputs are made available.
Axial elevations for the cracks in the Hatch 1 shroud range from about 550.2 cm (about 216.62")
at the lower elevation to about 812.5 cm (about 319.88") at the upper elevation. Figure 4 illustrates the axial elevations in a side view of the core shroud.
Cracks at the upper elevation are about 69.5 cm (27.38") above the jet pump brace pad elevation.
Because of the assumptions used in the preparation of the Hatch 1 operating data, the results at the upper elevations may be non-conservative relative to the brace pad elevation.
Cracks at the lower elevation are within about 30 cm (12 inches) of the bottom of active fuel.
Because of the assumptions used in the preparation of the Hatch 1 operating data, the results at the lower elevations should be more conservative than reported at the brace pad elevations. Note that there is no measurement data to support any statements regarding conservatisms in the model at elevations away from the core mid-plane.
For this analysis, the Hatch 1 shroud spanning the axial height of the core was meshed into 45 azimuthal arcs, 25 axial segments, and 3 radial annuluses, for a total of 3,375 mesh regions. The RAMA computer model is illustrated in Figure 5. Although twenty-five axial segments are illustrated; only segments 1 through 22 were calculated. This provided coverage of all axial crack locations and allowed the computer calculations to be performed within the time allotted for the analysis.
Calculations The >1.0 MeV and >0.1 MeV neutron fluences were estimated at the vertical and horizontal crack locations in the Hatch 1 shroud through the end of cycle 20. The fluence was estimated using cycle-specific reactor operating data through cycle 19 that is described in [5]. Operating data for cycle 20 was not specifically available. Cycle 20 data was, therefore, approximated using the reactor core power shapes from cycle 19 and the operating history for cycle 20.
Volume-average neutron fluence was calculated for each of the 3,375 mesh regions of the shroud model that is illustrated in Figure 5. Interpolation/extrapolation techniques were used to determine the fluence at the locations of interest, specifically at the inner and outer surfaces of the shroud wall and at the end points of the cracks.
TWE-HATCH-OO1-R-OO1 Revision 0 Page 11 of 22 987.7552cm (.388) -I - _
910.7424 cm (35.561 soe.4572cm (356.881) H2 -
25 89.8552cm (33.881) K - -
895.5024 cm p52.581 24 880.2824 em (348.561 23 s8.0224 cm (340.581 22 849.7824 cm (34.581 21 834.5424cm (328.58)
-20 819.3024cm (322.561 812.80 Cm (-3201) 19 807,4152cm (317.88) "4 - 804.0624cm (316.581 l8
. .~~~~~~~~~~~. 788.8224cm (310.581 17 773.S824cm (304.68) 75e.3424 cm (298.58)
Brace Pad .742.95 cm -.4. 743.1024 cm (292.58)
D 15 727M824cm (288.56) 13 Capsule b - 712.6224cm (280.581s t12 897.3824cm (274.56 11 682.1424 cm (268.561 10 6e6.9024cm (262.58) 09 e51.8824 cm (258.58) 08 836.4224 cm (250.581 07 621.1824 cm (24.581
- 08 605.9424 cm (238.58 05 690.7024 cm (232.s5) 04 575.4624 cm c(22.881 03 57.8348 cm (219.82 H- 50.0224 cm (220.581 551.18CM (-2171 -y- 02 54.9824 cm (214.56) 01 Bocnm OfAc1veFuel - 529.7424 cm (208.5) 466.3948cm (153.521 MA J
Figure 4 Side View Showing Elevations of the Hatch 1 Shroud
TWE-HATCH-OO-R-OO1 Revision 0 Page 12 of 22
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TWE-HATCH-OO1-R-OO1 Revision 0 Page 13 of 22
<>~' Table 2 lists the >1.0 MeV and >0.1 MeV fluences for the vertical and horizontal cracks identified in Table 1.Fluence values are given for the end points of each of the reported cracks.
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Table 2 Fluence Estimates for the Shroud Cracks Elevations (inches) >1.0 MeV Fluence >0.1 MeV Fluence Crack ID Degrees Bottom Top Bottom Top Bottom Top V04-01 Inside 2" 324 317.88 - 319.88 8.95E+20 8.68E+20 1.74E+21 1.68E+21 V05-01 Outside 13.5" 47 284.38 - 297.88 7.61 E+20 7.49E+20 1.64E+21 1.61 E+21 V05-02 Outside 2" 47 259.88- 261.88 7.55E+20 7.58E+20 1.62E+21 1.63E+21 V05-03 Outside 33/4" 47 255.13 - 255.88 7.42E+20 *7.44E+20 1.60E+21 1.60E+21 V05-04 Outside Y4" 47 237.63 - 237.88 6.40E+20 6.42E+20 1.38E+21 1.38E+21 V05-05 Outside 2" 47 233.88 - 235.88 6.20E+20 6.31 E+20 1.33E+21 1.36E+21 V05-06 Outside 1.5" 47 226.38 - 227.88 5.75E+20 5.86E+20 1.23E+21 1.26E+21 V05-07 Outside 1.5" 47 218.12 - 219.62 4.77E+20 4.98E+20 1.02E+21 1.07E+21 V05-08 Inside 3/4" 53 266.13- 266.88 1.23E+21 1.23E+21 2.42E+21 2.42E+21 V05-09 Outside 1.5" 53 218.12- 219.62 3.92E+20 -4.09E+20 8.36E+20 8.73E+20 V06-01 Inside W" 228 279.38 - 279.88 1.45E+21 1.45E+21 2.88E+21 2.87E+21 V06-02 Outside 11.5" 233 306.38 - 317.88 5.85E+20 5.21 E+20 1.26E+21 1.12E+21 V06-03 Outside 25.5" 232 279.88 - 305.38 6.95E+20 6.64E+20 1.49E+21 1.43E+21 yw....' V06-04 Outside 5.25" 232 263.63 - 268.88 6.95E+20 7.00E+20 1.49E+21 1.50E+21 V06-05 Inside 2" 232 219.62- 221.62 9.35E+20 9.86E+20 1.85E+21 1.95E+21 VOB-01 Outside Y2" 322 216.62- 217.12 4.12E+20 4.24E+20 8.78E+20 9.03E+20 H05-01, Inside 10" 228.5 219.62 1.01 E+21 1.98E+21 231.5 219.62 9.66E+20 1.90E+21 Table 2 shows that vertical cracks V05-08 and V06-01 and horizontal crack H05-01 exceed the fracture toughness threshold of 1.0 x 1021 n/cm2 for the >1.0 MeV fluence. Each of these cracks' are on the inside surface of the shroud wall. Vertical cracks V04-01 and V06-05, which are also on the inside' surface, are approaching the toughness threshold with values ranging from 8.68 x 1020 n/cm 2 to 9.86 x 1020 n/cm 2. Several cracks a're identified on the outside surface of the shroud wall, however, none of the outer surface cracks show violations of the fracture toughness threshold.
All of the cracks except V05-09 and V08-01 for the >0.1 MeV fluence are shown to have exceeded the fracture toughness threshold. The two exceptions are cracks that are on the outside surface of the shroud wall, although they are approaching the toughness threshold.
The cracks appear to be situated around specific azimuths in the reactor. Reviewing the reactor core configuration shown in Figure 3, it is observed that the cracks are nearly aligned with the fuel bundles that are closest to the shroud wall. This should be expected since these locations are h'-' closest to the neutron source.
TWE-HATCH-OO1-R-OO1 Revision 0 Page 14 of 22 Vertical Cracks Figures 6 through 11 contain axial plots of the >1.0 MeV and >0.1 MeV neutron fluences for the vertical welds in the Hatch 1 shroud.
Figure 6 shows the axial >1.0 MeV and >0.1 MeV neutron fluences for the vertical cracks at azimuth 53 degrees (near V-5) between the elevations H-4 and H-5. Both the inside and outside wall fluences are shown.
Figure 7 shows the axial >1.0 MeV and >0.1 MeV neutron fluences for the vertical crack at azimuth 47 degrees (near V-5) between the elevations H-4 and H-5. Only the outside wall fluence is shown.
Figure 8 shows the axial >1.0 MeV and >0.1 MeV neutron fluences for the vertical cracks at azimuths 228 and 233 degrees (near V-6) between the elevations H-4 and H-5. The inside wall fluence is given for the 228 azimuth position and the outside wall fluence is given for the 233 azimuth position.
Figure 9 shows the axial >1.0 MeV and >0.1 MeV neutron fluences for the vertical cracks at azimuth 232 degrees (near V-6) between the elevations H-4 and H-5. Both the inside and outside wall fluences are shown.
Figure 10 shows the axial >1.0 MeV and >0.1 MeV neutron fluences for the vertical cracks at azimuth 322 degrees (near V-8) between the elevations bottom of active fuel and H-5. Only the outside wall fluence is shown.
Figure 11 shows the axial >1.0 MeV and >0.1 MeV neutron fluences for the vertical cracks at azimuth 324 degrees (near V-4) between the elevations H-4 and top of active fuel. Only the inside wall fluence is shown.
Horizontal Cracks Figure 12 shows azimuthal plots of the >1.0 MeV and >0.1 MeV neutron fluences for the H-5 weld . Azimuthal fluences are shown for both the inside and outside wall surfaces.
TWE-HATCH-OOI-R-OOI Revision 0 Page 15 of 22 Hatch 1 Shroud >1.0 MeV Neutron Fluence at End of Cycle 20 Near Vertical Weld V-5, Between Elevatlons H-4 and H-5 895.50 865.02 834.54 8040D6-o 773.58 743.10 -
, 682.14 -
651.66 __ _ _ _ _ _
62t118 - _ Am s 590.70 5. Dogs Azimuth.Outside
..-.-.53 56022 Z lOEOE 21Threshold 529.74 2.50E+20 5.00E+20 7.50E420 tOOE421 t25E+21 t50E421 175E421 2.00E421 225E+21 2.OE421
>1.0 MeV Neutron Fluence Hatch 1 Shroud >0.1 MeV Neutron Fluence at End of Cycle 20 Near Vertical Weld V-5, Between ElevatIons H-4 and H-5 895.50 53 Degs Azimuth, Inside 865.02 . . . . . .53 Degs Azimuth, Outside 834.54 - 1.OE+21 Threshold 804.06 Q 773.58 -
C 743.10 -
> 712.62 11 682.14 J
- 651.66 621.18 -
590.70-560.22 529.74 2.50E+20 5.00E+20 7.50E+20 1.WOE+21 1.25E+21 1.50E+21 1.75E+21 2.OOE+21 2.25E+21 2.50E+21
>0.1 MeV Neutron Fluence Figure 6 Hatch 1 Shroud Axial Neutron Fluence at Azimuth 53 Degrees
TWVE-HATCH-001-R-001 Revision 0 Page 16 of 22 Hatch 1 Shroud >1.0 MeV Neutron Fluence at End of Cycle 20 Near Vertical Weld V-5, Between Elevations H-4 and H-5 895.50 -
865.02 -
834.54 -
E804.06-E c 773.58 C 743.10
' 712.62 iii 682.14 651.66 621.18 590.70 / D _ t_
560.22 /_
560.22 ~ ~~~~~~~-1.OE+21 Threshold 529.74 2.50E+20 5.OOE+20 7.50E+20 1.00E+21 1.25E+21 1.50E+21 1.75E+21 2.OOE+21 2.25E+21 2.50E+21
>1.0 MeV Neutron Fluence Hatch 1 Shroud >0.1 MeV Neutron Fluence at End of Cycle 20 Near Vertical Weld V-5, Between Elevations H-4 and H-5 895.50 l- 47 Degs Azimuth, Outside 865.02 -
-1.OE+21 Threshold 834.54 -
E 804.06 c 773.58 C 743.10
( 712.62 M 682.14 c 651.66 -
621.18 -
590.70 560.22 529.74 2.50E+20 5.OOE+20 7.50E+20 1.OOE+21 1.25E+21 1.50E+21 1.75E+21 2.OOE+21 2.25E+21 2.50E+21
>0.1 MeV Neutron Fluence Figure 7 Hatch 1 Shroud Axial Neutron Fluence at 47 Degrees
TWE-HATCH-001-R-001 Revision 0 Page 17 of22 Hatch 1 Shroud >1.0 MeV Neutron Fluence at End of Cycle 20 Near Vertical Weld V-6, Between Elevations H-4 and H-5 895.50 -
865.02 -
834.54 -
~804.06 U
773.58 -
743.10 -
712.62 W 682.14 621.18 -
/ / ~~~~~~228 Degs Azimuth, Inzside 590.70- 590.70
/ * / s * - w-~~~....
233 Degs Azimuth, Outside 560.22 - _ - t0E+21 Threshold 529.74 2.50E+20 5.00E+20 7.50E+20 1.00E+21 1.25E+21 1.50E+21 1.75E+21 2.00E+21 2.25E+21 2.50E+21
>1.0 MeV Neutron Fluence Hatch 1 Shroud >0.1 MeV Neutron Fluence at End of Cycle 20 Near Vertical Weld V-6, Between Elevations H-4 and H-5 895.50 * - 228 Degs Azimuth, Inside 865.02 -- .
233 Degs Azimuth, Outside 834.54 -_ 1.0E+21 Threshold 804.06 _
773.58-c 743.10-
.2 712.62-LIl 682.14 651.66 -
621.18-590.70-560.22 - -
529.74 7.50E+20 1.00E+21 1.25E+21 1.50E+21 1.75E+21 2.OOE+21 2.25E+21 2.50E+21 2.75E+21 3.OOE+21
>0.1 MeV Neutron Fluence Figure 8 Hatch 1 Shroud Axial Neutron Fluence at 228 and 233 Degrees
TEVE-HATCH-OO1-R-OO1 Revision 0 Page 18 of 22 Hatch 1 Shroud >1.0 MeV Neutron Fluence at End of Cycle 20 Near Vertical Weld V-6, Between Elevations H-4 and H-5 0
(U a) wU
'a 529.74 1 I 2.50E+20 5.OOE+20 7.50E+20 1.00E+21 1.25E+21 1.50E+21 1.75E+21 2.00E+21 2.25E+21 2.50E+21
>1.0 MeV Neutron Fluence Hatch 1 Shroud >0.1 MeV Neutron Fluence at End of Cycle 20 Near Vertical Weld V-6, Between Elevations H-4 and H-5 895.50 - n 2 232 Degs Azimuth, Inside 865.02 - . . . . . . *232 Degs Azimuth, Outside 834.54 - 1.OE+21 Threshold 804.06-c 773.58 O 743.10
> 712.62-Eu 682.14-651.66-621.18 590.70 560.22 - .
529.74 7.50E+20 1.00E+21 1.25E+21 1.50E+21 1.75E+21 2.OOE+21 2.25E+21 2.50E+21 2.75E+21 3.OOE+21
>0.1 MeV Neutron Fluence Figure 9 Hatch 1 Shroud Axial Neutron Fluence at 232 Degrees
TWE-HATCH-OO1-R-OOI Revision 0 Page 19 of 22 Hatch 1 Shroud >1.0 MeV Neutron Fluence at End of Cycle 20 Near VerticaI I .I....
Weld V-8, Between Elevations BOAF and H-5 895.50 _
-322 Degs Azimiuth Outside 865.02 *_
834.54 _- 1.0E+21 Threst ;old 804.06 c 773.58 O 743.10 a 712.62 Lu 682.14
< 651.66 621.18 -
590.70 56022 -
529.74 t.
2.00E+2 4.W0E+2 6.OOE+2 8.WOE+2 1.WOE+2 1.20E+2 1AOE+2 1.60E+2 1.80E+2 2.OOE+2 0 0 0 0 1 1 1 1 1 1
>1.0 MeV Neutron Fluence Hatch 1 Shroud >0.1 MeV Neutron Fluence at End of Cycle 20 Near Vertical Weld V-8, Between Elevations BOAF and H-5 895.50
- 322 Dogs Azimuth, Outside 80 865.02-834.54 - 1.0E+21 Threshold l 804.06 E
c 773.58
= 743.10
> 712.62 wU 682.14
- 651.66 621.18 590.70 56022 -
529.74 .
2.OOE+2 4.OOE+2 6.00E+2 8.WOE+2 1.OOE+2 1.20E+2 1.40E+2 1.60E+2 1.80E+2 2.0012+2 0 0 0 0 1 1 1 1 1 I
>0.1 MeV Neutron Fluence Figure 10 Hatch 1 Shroud Axial Neutron Fluence at 322 Degrees
TWE-HATCH-OOI-R-001 Revision 0 Page20 of 22 Hatch 1 Shroud >1.0 MeV Neutron Fluence at End of Cycle 20 Near Vertical Weld V4, Between Elevations H-4 and TOAF 895.50 865.02 834.54
\1 E 804.06 U
c 773.58 o 743.10 0
CO712.62 C)
LU 682.14 651.66 621.18 590.70
- 324 Degs Azimuth, Inside 560.22 1.0E+21 Threshold 529.74 2.0(DE+2 4.OOE+2 6.OOE+2 8.00E+2 1.00E+2 1.20E+2 1.40E+2 1.60E+2 1.80E+2 2.OOE+2 0 0 0 0 1 1 1 1 1 1
>1.0 MeV Neutron Fluence Hatch 1 Shroud >0.1 MeV Neutron Fluence at End of Cycle 20 Near Vertical Weld V-4, Between Elevations H-4 and TOAF 895.50 865.02 834.54 804.06 773.58 o 743.10
' 712.62 LU 682.14
- 4651.66 621.18 590.70
-324 Degs Azimuth, Inside 560.22 - 1.OE+21 Threshold 529.74 2.00E+204.00E+206.OOE+208.OOE+20 1.OOE+21 1.20E+21 1.40E+21 1.60E+21 1.80E+21 2.OOE+21
>0.1 MeV Neutron Fluence Figure 11 Hatch 1 Shroud Axial Neutron Fluence at 324 Degrees
i TWE-HATCH-OO1-R-OO1 Revision 0 Page21 of 22 Hatch 1 Shroud >1.0 MeV Neutron Fluence at End of Cycle 20 Azimuthal Fluence at H-5 Elevation 120E+21 1.10E+21 - Inside Wall Surface
. . . . Outside. Wall Surface 1.00E+21 *e 1.0E+21 Threshold 9.00E+20 -
c 8.00E+20 -
U-c 7.OOE+20 -
- 6.00E+20 -
> 5.WOE+20 /
o 4.OOE+20 /
A 3.OOE+20 , -. . .- .....
2.00E+20 -......----.... " .... -.,,..
11.OOE+20 O.OOE+00 0 5 10 15 20 25 30 35 40 45 Azimuthal Angles In Degrees Hatch 1 Shroud >0.1 MeV Neutron Fluence at End of Cycle 20 Azimuthal Fluence at H-5 Elevation 2.20E+21 2.00E+21 -Inside Wall Surface
. . . .Outside.0 Wall Surface 1.80E+21 . 1.OE+21 Threshold CD 1.60E+21 L 1.40E+21 -
2 1.20E+21 -
z 1.00E+21 I 0 8.OOE+20 d 6.00E+20 --
4.OOE+20 -
2.00E+20 O.OOE+OO 0 5 10 15 20 25 30 35 40 45 Azimuthal Angles In Degrees Figure 12 Hatch I Shroud Azimuthal Neutron Fluence at H-5 Elevation
TWE-HATCH-OO-R-OO1 Revision 0 Page 22 of 22 References
- 1. Core Shroud Weld Identification Layout for the E. I. Hatch Unit 1 reactor, transmitted to Dean Jones of TransWare Enterprises Inc. from Robin Dyle of Southern Nuclear via email, May 23, 2003.
- 3. BWRVIP-115: BWR Vessel andInternals Project, RAMA Fluence Methodology Benchmark Manual - Evaluationof Regulatory Guide 1.190 Benchmark Problems, EPRL Palo Alto, CA:
2003 1008063.
- 4. BWRVIP-117: BWR Vessel andInternals Project, RAMA Fluence Methodology Plant Application - Susquehanna Unit 2 Surveillance CapsuleFluence Evaluationfor Cycles 1-5, EPRI, Palo Alto, CA: 2003 1008065.
- 5. "RAMA Analysis of the Edwin I. Hatch 1 Jet Pump Riser Brace Pad Neutron Dosimetry Measurements", BNL Contract 49733, April 30, 2003.
- 6. "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence",
Nuclear Regulatory Commission Regulatory Guide 1.190, March 2001.
- 7. J. F. Carew, et al., "PWR and BWR Pressure Vessel Fluence Calculation Benchmark Problems and Solutions", NUREG/CR-6115, Brookhaven National Laboratory, BNL-NUREG-52395, September 2001.