ML17240A231: Difference between revisions
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The plate representative material in the 300° surveillance capsule is not a heat specific match to material in the Enrico Fermi Unit 2 reactor vessel and therefore , direct use is not Together ... Shaping the Future of Electricity (5-1JtJif- | The plate representative material in the 300° surveillance capsule is not a heat specific match to material in the Enrico Fermi Unit 2 reactor vessel and therefore , direct use is not Together ... Shaping the Future of Electricity (5-1JtJif- | ||
µ (( ( PALO ALTO OFFICE 3.420 Hillview Avenue, Pala Alto, CA 9.430.4*1395 USA* 650.855.2000 *Customer Service 800.313.377.4 | µ (( ( PALO ALTO OFFICE 3.420 Hillview Avenue, Pala Alto, CA 9.430.4*1395 USA* 650.855.2000 *Customer Service 800.313.377.4 | ||
* www.epr i.com made of the surveillance data. There is no impact of the surveillance data on the Enrico Fermi Unit 2 P-T limit curves. It should be noted that the BWRVIP previously requested an extension to the 10 CFR 50 Appendix H surveillance capsule testing reporting requirements and committed to providing the results of the Hatch Unit 1 300° surveillance capsule b y August 31 , 2017. That request was approved in Reference | * www.epr i.com made of the surveillance data. There is no impact of the surveillance data on the Enrico Fermi Unit 2 P-T limit curves. It should be noted that the BWRVIP previously requested an extension to the 10 CFR 50 Appendix H surveillance capsule testing reporting requirements and committed to providing the results of the Hatch Unit 1 300° surveillance capsule b y August 31 , 2017. That request was approved in Reference | ||
: 2. Please also note that the enclosed report is non-proprietary and is available to the public by request to EPRI. If you have any questions on this subject p l ease call Steve Richter (Energy Northwest , BWRVIP Assessment Committee Chairman) at 509-377-4703. | : 2. Please also note that the enclosed report is non-proprietary and is available to the public by request to EPRI. If you have any questions on this subject p l ease call Steve Richter (Energy Northwest , BWRVIP Assessment Committee Chairman) at 509-377-4703. | ||
Sincerely , Andrew McGehee , EPRI , BWRVIP Chairman Tim Hanley , Exe l on Corporation , BWRVIP Chairman c: S. Ruffin , NRC-NRR M. Kirk , NRC-RES D. Odell , Exelon Corp. R. Carter , EPRI A. McGehee , EPRI C. Wirtz , EPRI ELECTRIC POWER RESEARCH INSTITUTE 2017 TECHNICAL REPORT BWRVIP-308NP: | Sincerely , Andrew McGehee , EPRI , BWRVIP Chairman Tim Hanley , Exe l on Corporation , BWRVIP Chairman c: S. Ruffin , NRC-NRR M. Kirk , NRC-RES D. Odell , Exelon Corp. R. Carter , EPRI A. McGehee , EPRI C. Wirtz , EPRI ELECTRIC POWER RESEARCH INSTITUTE 2017 TECHNICAL REPORT BWRVIP-308NP: | ||
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1-1 In tro du c tion 1.1 Implementation Requirements The results documented in this report will be utilized by the BWRVIP ISP and b y individual utilities to demonstrate compliance with 10CFR50 , Appendi x H , Reactor Vesse l Material Surveillance Program Requirements. | 1-1 In tro du c tion 1.1 Implementation Requirements The results documented in this report will be utilized by the BWRVIP ISP and b y individual utilities to demonstrate compliance with 10CFR50 , Appendi x H , Reactor Vesse l Material Surveillance Program Requirements. | ||
T h erefore, the implementation requirements of 10CFR50, Appendix H govern and the implementation requirements of Nuclear Energy Institute (NEI) 03-08, Guideline for t h e Management of Materials Issues [7], are not applicable. 1-2 2 MATERIALS AND T EST SPECIMEN DESCRIPTION The General Electric (GE) designed Hatch Unit 1 300° surveillance capsule was removed from the plant and shipped to MP Machinery and Testing, LLC (MPM) for analysis. | T h erefore, the implementation requirements of 10CFR50, Appendix H govern and the implementation requirements of Nuclear Energy Institute (NEI) 03-08, Guideline for t h e Management of Materials Issues [7], are not applicable. 1-2 2 MATERIALS AND T EST SPECIMEN DESCRIPTION The General Electric (GE) designed Hatch Unit 1 300° surveillance capsule was removed from the plant and shipped to MP Machinery and Testing, LLC (MPM) for analysis. | ||
The capsule conta i ned a total of two Charpy packets and three tensile tubes. The 300° surveillance capsule is an original plant capsule, and was irradiated in the plant since initial startup. This is the third surve i ll ance capsule to be removed from the Hatch Unit 1 reactor pressure vesse l (RPV) and tested. The 30° survei ll ance capsule was teste d by GE and the results are reported in Reference | The capsule conta i ned a total of two Charpy packets and three tensile tubes. The 300° surveillance capsule is an original plant capsule, and was irradiated in the plant since initial startup. This is the third surve i ll ance capsule to be removed from the Hatch Unit 1 reactor pressure vesse l (RPV) and tested. The 30° survei ll ance capsule was teste d by GE and the results are reported in Reference | ||
[8]. The 120° surveillance capsule was also tested by GE and the results are reported in Refer e nce [9]. Flux wires from the 30° position were removed after cycle 1 and were also tested by GE [10). 2.1 Dosimeters The dosimetry wires were located along the ends of the Charpy specimens within each of the Charpy packets during irradiation. | [8]. The 120° surveillance capsule was also tested by GE and the results are reported in Refer e nce [9]. Flux wires from the 30° position were removed after cycle 1 and were also tested by GE [10). 2.1 Dosimeters The dosimetry wires were located along the ends of the Charpy specimens within each of the Charpy packets during irradiation. | ||
Each of the two Charpy packets contained one high purity iron wire, one high purity copper wire, and one high purity nicke l wire for fluence eval u ation. Further detai l s on the exact wire locations during the irradiation are provided in the capsule open i ng discussion given in Section 2.3. A detai l ed discussion of the radiometric analysis of the capsu l e dosimetry wires is provided in Appendix A. 2.2 Test Materials The Hatch Unit 1 300° survei ll ance capsule Charpy V-notch specimen inventory, material descriptions, unirradiated (baseline) | Each of the two Charpy packets contained one high purity iron wire, one high purity copper wire, and one high purity nicke l wire for fluence eval u ation. Further detai l s on the exact wire locations during the irradiation are provided in the capsule open i ng discussion given in Section 2.3. A detai l ed discussion of the radiometric analysis of the capsu l e dosimetry wires is provided in Appendix A. 2.2 Test Materials The Hatch Unit 1 300° survei ll ance capsule Charpy V-notch specimen inventory, material descriptions, unirradiated (baseline) | ||
Charpy impact data, and previous l y measured data are summarized in this section of the report. 2.2.1 Capsule Loading Inventory The Hatch Unit 1 300° surveillance capsule inventory is provided in Tab l e 2-1. All of the capsule specimens , which include Charpy specimens , tensile specimens , and dosimeters , were recovered from the capsule basket. Testing was performed on the 24 Charpy specimens , and the dosimetry wires were co u nted and weighed to determine specific activities. All six of the tensile specimens (two base, two weld, and two h eat-affected zone [HAZ]) remain unteste d and are being held in reserve for future testing since there is no near-term use for the tensi l e data. The technical advan t age of storing the tensile specimens untested is that there will be options in the future for how these specimens will be used to obtain useful data. For example, the tensi l e specimen geometry is conducive to fabr i cation of sub-size Charpy as we ll as min i aturized Charpy V-notch specimens. | Charpy impact data, and previous l y measured data are summarized in this section of the report. 2.2.1 Capsule Loading Inventory The Hatch Unit 1 300° surveillance capsule inventory is provided in Tab l e 2-1. All of the capsule specimens , which include Charpy specimens , tensile specimens , and dosimeters , were recovered from the capsule basket. Testing was performed on the 24 Charpy specimens , and the dosimetry wires were co u nted and weighed to determine specific activities. All six of the tensile specimens (two base, two weld, and two h eat-affected zone [HAZ]) remain unteste d and are being held in reserve for future testing since there is no near-term use for the tensi l e data. The technical advan t age of storing the tensile specimens untested is that there will be options in the future for how these specimens will be used to obtain useful data. For example, the tensi l e specimen geometry is conducive to fabr i cation of sub-size Charpy as we ll as min i aturized Charpy V-notch specimens. | ||
Further , research is currently underway to develop testing methods which will enable the de t ermination of p l ane-strain | Further , research is currently underway to develop testing methods which will enable the de t ermination of p l ane-strain | ||
:fracture toughness data from Charpy-sized specimens. With these new technologies in view , there may also be a future need for static and/or dynamic tensile data for use in the calculation of :fracture toughness from experimental data obtained from 2-1 Materials and T es t Specimen D esc ription Charpy specimens. | :fracture toughness data from Charpy-sized specimens. With these new technologies in view , there may also be a future need for static and/or dynamic tensile data for use in the calculation of :fracture toughness from experimental data obtained from 2-1 Materials and T es t Specimen D esc ription Charpy specimens. | ||
Therefore , all of the tensile specimens have been placed into the archive storage so that they can be tested when necessary in the future. The broken Charpy specimen halves have been added to long-term archive storage for future use in miniature mechanical behavior specimen testing, reconstitution, chemistry analysis, and microstructural studies. As indicated in Table 2-1, there were two Charpy packets in the capsule, and each contained three dosimetry wires (one Fe wire, one Cu wire, and one Ni wire) and 12 Charpy specimens. | Therefore , all of the tensile specimens have been placed into the archive storage so that they can be tested when necessary in the future. The broken Charpy specimen halves have been added to long-term archive storage for future use in miniature mechanical behavior specimen testing, reconstitution, chemistry analysis, and microstructural studies. As indicated in Table 2-1, there were two Charpy packets in the capsule, and each contained three dosimetry wires (one Fe wire, one Cu wire, and one Ni wire) and 12 Charpy specimens. | ||
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: Fi g ur e 2-1 90 +/- 2 degre es 90 degrees +/- 10 minute s +/- 0.003 inche s 2.165 (+0.0 , -0.100) inch es +/- 0.039 inche s +/- 1 degree 0.010 +/- 0.001 inche s +/- 0.001 inche s 63 µ-inch on notched surface and opposite face; 4 µ-inch e l sewhere Draw i ng Showing the Charpy Test Specimen Geometry and ASTM E23 Permissib le V a riat i ons 2-3 Materials and T es t Sp eci m en D esc ription Figure 2-2 Photograph of the 300° Hatch Unit 1 Capsule (top) and a Magnified View of the External Identification Markings (bottom) Figure 2-2 shows the side of the surveillance capsule which faced the core. The identification code, "1 17C4671G3", was engraved near the hook. 2-4 Mat e rials and Test Specimen D esc ription Figure 2-3 Photograph of th e 300° Hatch Unit 1 Capsule Figure 2-3 s h ows the side of the surve ill ance capsule which faced the pressure vessel. T h e reac t or and ca p s u le co d es are seen near t he h ook. 2-5 Mat e rials and T e st Sp ec im e n D esc ription Figure 2-4 Photograph of the Inside of the 300° Hatch Unit 1 Capsule 2.2.2 Material Description The Hatch Unit 1 RPV is a 218-inch (5537mm) diameter BWRJ4 design. The pressure vessel construction was performed by Combustion Engineering to the 1965 Edition of the ASME Code, with Winter 1966 Addenda. The pressure vessel shell and head plate materials are ASME A533, Grade B , Class 1 low alloy steel. The nozzl es and closure flanges are A508 Class 2 low alloy steel, and the closure flange boltin g materials are ASME SA540 Grade B24 low alloy steel. The fabrication process employed quench and temper heat treatment immediately after hot forming, then submerged arc we lding and post-weld heat treatment. | : Fi g ur e 2-1 90 +/- 2 degre es 90 degrees +/- 10 minute s +/- 0.003 inche s 2.165 (+0.0 , -0.100) inch es +/- 0.039 inche s +/- 1 degree 0.010 +/- 0.001 inche s +/- 0.001 inche s 63 µ-inch on notched surface and opposite face; 4 µ-inch e l sewhere Draw i ng Showing the Charpy Test Specimen Geometry and ASTM E23 Permissib le V a riat i ons 2-3 Materials and T es t Sp eci m en D esc ription Figure 2-2 Photograph of the 300° Hatch Unit 1 Capsule (top) and a Magnified View of the External Identification Markings (bottom) Figure 2-2 shows the side of the surveillance capsule which faced the core. The identification code, "1 17C4671G3", was engraved near the hook. 2-4 Mat e rials and Test Specimen D esc ription Figure 2-3 Photograph of th e 300° Hatch Unit 1 Capsule Figure 2-3 s h ows the side of the surve ill ance capsule which faced the pressure vessel. T h e reac t or and ca p s u le co d es are seen near t he h ook. 2-5 Mat e rials and T e st Sp ec im e n D esc ription Figure 2-4 Photograph of the Inside of the 300° Hatch Unit 1 Capsule 2.2.2 Material Description The Hatch Unit 1 RPV is a 218-inch (5537mm) diameter BWRJ4 design. The pressure vessel construction was performed by Combustion Engineering to the 1965 Edition of the ASME Code, with Winter 1966 Addenda. The pressure vessel shell and head plate materials are ASME A533, Grade B , Class 1 low alloy steel. The nozzl es and closure flanges are A508 Class 2 low alloy steel, and the closure flange boltin g materials are ASME SA540 Grade B24 low alloy steel. The fabrication process employed quench and temper heat treatment immediately after hot forming, then submerged arc we lding and post-weld heat treatment. | ||
The post-weld heat treatment was typically 6 hours at 1150°F +/-25°F (621.1°C +/-13.9°C). The surveillance base metal specimens were machined from plate G-4804-2 from the beltline. The test plate was heat treated for 40 hours at l 150°F +/-25°F (621.1°C +/-13.9°C) to simulate the post weld he at treatment of the vessel. Specimens were machined from the Vi T and % T positions in the plate, in the longitudinal orientation (long axis parallel to the rolling direction). | The post-weld heat treatment was typically 6 hours at 1150°F +/-25°F (621.1°C +/-13.9°C). The surveillance base metal specimens were machined from plate G-4804-2 from the beltline. The test plate was heat treated for 40 hours at l 150°F +/-25°F (621.1°C +/-13.9°C) to simulate the post weld he at treatment of the vessel. Specimens were machined from the Vi T and % T positions in the plate, in the longitudinal orientation (long axis parallel to the rolling direction). | ||
The weld material and HAZ Charpy specimens were fabricated from trim-off pieces of plates G-4804-2 and G-4804-1 that were welded together with a weld identical to longitudinal seam we ld 1-308 in the RPV b eltline. The welded test plate for the weld and HAZ C harp y specimens received a heat treatment of 1150°F +/-25°F (621.1°C +/-13.9°C) for 40 hours to match the 2-6 Mat e rials and T e st Spe c im en D es cription fabri c ated condition of the RPV. The base metal orientation in the weld and RAZ specimens was longitudinal | The weld material and HAZ Charpy specimens were fabricated from trim-off pieces of plates G-4804-2 and G-4804-1 that were welded together with a weld identical to longitudinal seam we ld 1-308 in the RPV b eltline. The welded test plate for the weld and HAZ C harp y specimens received a heat treatment of 1150°F +/-25°F (621.1°C +/-13.9°C) for 40 hours to match the 2-6 Mat e rials and T e st Spe c im en D es cription fabri c ated condition of the RPV. The base metal orientation in the weld and RAZ specimens was longitudinal | ||
[8]. All specimens are stamped on the ends with the fabrication code listed on the Hatc h Unit 1 drawings. | [8]. All specimens are stamped on the ends with the fabrication code listed on the Hatc h Unit 1 drawings. | ||
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0.984 E q u ation is A+ B * (Tanh((T-TO) | 0.984 E q u ation is A+ B * (Tanh((T-TO) | ||
/(C+DD)J U pper Shelf Energy= 136.00 (Fixed) Lower Shelf E ner gy= 2.50 (Fixed) Tem p@30 ft-lbs=-6 I .50° F T e m p@_15 fl-l b s=-5 2.20° F Te m p@,50 ft-l b s=-28.80° F P l a nt: Ha t ch 1 Orientation | /(C+DD)J U pper Shelf Energy= 136.00 (Fixed) Lower Shelf E ner gy= 2.50 (Fixed) Tem p@30 ft-lbs=-6 I .50° F T e m p@_15 fl-l b s=-5 2.20° F Te m p@,50 ft-l b s=-28.80° F P l a nt: Ha t ch 1 Orientation | ||
: L T Mate ri al: SA533 Bt Ca ps u le: UNIRRA Heat: C41 1 4-2 Fluence: O.OOE+oOO n/c m' --r;l:J ..c -I ¢:: _, ;;..-. '"" Cl.I = :z > u 1 60 ._ 1 40 -120 ._ 100 -* . 80 -60 40 20 ._ 0 -300 CVGmph 6.02 Figure 2-5 n v / I /o T I I i6 / __/ -200 -100 0 100 200 300 Temperature | : L T Mate ri al: SA533 Bt Ca ps u le: UNIRRA Heat: C41 1 4-2 Fluence: O.OOE+oOO n/c m' --r;l:J ..c -I ¢:: _, ;;..-. '"" Cl.I = :z > u 1 60 ._ 1 40 -120 ._ 100 -* . 80 -60 40 20 ._ 0 -300 CVGmph 6.02 Figure 2-5 n v / I /o T I I i6 / __/ -200 -100 0 100 200 300 Temperature | ||
{° F) 02/23/20 17 Charpy Energy Plot for Plate Heat C4114-2 (LT) Un irradiated 2-1 0 ' --400 500 600 Pa ge 1/2 P lant: Hatc h 1 Orientati o n: LT Tempera tur e (0 F) 80 -4 0 -4 0 -40 1 0 I O 1 0 4 0 4 0 4 0 110 11 0 160 16 0 CVGraph 6.02 Fig u re 2-5 (Continued) | {° F) 02/23/20 17 Charpy Energy Plot for Plate Heat C4114-2 (LT) Un irradiated 2-1 0 ' --400 500 600 Pa ge 1/2 P lant: Hatc h 1 Orientati o n: LT Tempera tur e (0 F) 80 -4 0 -4 0 -40 1 0 I O 1 0 4 0 4 0 4 0 110 11 0 160 16 0 CVGraph 6.02 Fig u re 2-5 (Continued) | ||
Materials and T e st Specimen D escriptio n Ma t e rial: SA533Bl Ca psule: UN1RRA PLATE HEAT C4114-2 (HAl) Charpy V-Notch Data Input CVN Co mpu ted CVN 29.0 21.8 2 7.0 21.8 34.0 4 2.4 38.0 4 2.4 47 0 4 2.4 8 4.0 7 9.4 80.0 7 9.4 82.0 7 9.4 1 06.0 1 00.2 89.0 1 00.2 90.0 1 00.2 133.0 1 27.0 142.0 127 0 130.0 133.0 139.0 133.0 02/2 3/20 17 Heat: C 4114-2 F lu e nce: O.OOE+OOO n/cm* Differential 7.20 5.20 -8.45 -4.45 4.55 4.56 0.56 2.56 5.83 -1117 -1 0.17 602 1 5 02 -303 5.9 7 Page 212 Cha r py Energy Plot for Plate Heat C4114-2 (LT) Unirradiated 2-11 Materials and T e st Sp ec im e n D e s c ription Plant: Hatc h 1 Orientation | Materials and T e st Specimen D escriptio n Ma t e rial: SA533Bl Ca psule: UN1RRA PLATE HEAT C4114-2 (HAl) Charpy V-Notch Data Input CVN Co mpu ted CVN 29.0 21.8 2 7.0 21.8 34.0 4 2.4 38.0 4 2.4 47 0 4 2.4 8 4.0 7 9.4 80.0 7 9.4 82.0 7 9.4 1 06.0 1 00.2 89.0 1 00.2 90.0 1 00.2 133.0 1 27.0 142.0 127 0 130.0 133.0 139.0 133.0 02/2 3/20 17 Heat: C 4114-2 F lu e nce: O.OOE+OOO n/cm* Differential 7.20 5.20 -8.45 -4.45 4.55 4.56 0.56 2.56 5.83 -1117 -1 0.17 602 1 5 02 -303 5.9 7 Page 212 Cha r py Energy Plot for Plate Heat C4114-2 (LT) Unirradiated 2-11 Materials and T e st Sp ec im e n D e s c ription Plant: Hatc h 1 Orientation | ||
: LT --*-6 "-' = 0 *-= Q. --90 70 50 30 20 10 0 -300 CVGraph 6.02 Figure 2-6 UNIRRADIATED PLATE HEAT C4114-2 LE CV Grap h 6.0 2: H y pert>olic Tan g ent Curv e Prin l ed on 4/1 2/20 1 7 7: 45 AM A= 43.38 B = 42.38 C = 84.93 TO= -1 7.77 D = 0.00 Corre l a tion Coeffi c i e nt= 0.9 85 Equa1ion is A+ B * [Tanh ((T-TO)/(C+D D)J Upper Shelf LE.= 85.7 5 (F ixed) Low e r Sbelf L.E. = 1.00 (f i....,ed) | : LT --*-6 "-' = 0 *-= Q. --90 70 50 30 20 10 0 -300 CVGraph 6.02 Figure 2-6 UNIRRADIATED PLATE HEAT C4114-2 LE CV Grap h 6.0 2: H y pert>olic Tan g ent Curv e Prin l ed on 4/1 2/20 1 7 7: 45 AM A= 43.38 B = 42.38 C = 84.93 TO= -1 7.77 D = 0.00 Corre l a tion Coeffi c i e nt= 0.9 85 Equa1ion is A+ B * [Tanh ((T-TO)/(C+D D)J Upper Shelf LE.= 85.7 5 (F ixed) Low e r Sbelf L.E. = 1.00 (f i....,ed) | ||
I 1 9(o 7 J L/ I I T e mp@-15 70° F M a t e rial: SA533 Bl Ca psule: UNIRRA u 0 0 /"t> o( lo , I I . I I H eat: C4 114-2 Flu e n ce: O.OOE+-000 n/cm' *-* I I -200 -100 0 100 200 300 400 500 600 Temperature | I 1 9(o 7 J L/ I I T e mp@-15 70° F M a t e rial: SA533 Bl Ca psule: UNIRRA u 0 0 /"t> o( lo , I I . I I H eat: C4 114-2 Flu e n ce: O.OOE+-000 n/cm' *-* I I -200 -100 0 100 200 300 400 500 600 Temperature | ||
(° F) 04/12/2 0 1 7 Page 1/2 Lateral Expansion Plot for Plate Heat C4114-2 (LT) Un irradiated 2-12 Mate r ia l s and T e st S p eci m en D escrip ti on P lant: Ha tch 1 Orientation: | (° F) 04/12/2 0 1 7 Page 1/2 Lateral Expansion Plot for Plate Heat C4114-2 (LT) Un irradiated 2-12 Mate r ia l s and T e st S p eci m en D escrip ti on P lant: Ha tch 1 Orientation: | ||
L T Ma terial: SA5 33Bl Capsu le: UNIRRA Heat: C 4114-2 F lu e nce: O.OOE+OOO n/cm 2 UNIRRADIATED PLATE HEAT C4114-2 LE Charpy V-Notch Data Te mp e r a tur e(° F) In put L. E. C omput e d L. E. Differ e nti a l -80 23.0 16.9 6.1 0 -80 20.0 16.9 3.1 0 -40 25.0 32.5 -7.53 -4 0 29.0 32.5 -3.53 -40 35.0 32.5 2.47 1 0 57.0 56.8 0.24 10 58.0 56.8 1.24 1 0 59.0 56.8 2.2 4 4 0 66.0 68.4 -2.45 4 0 73.0 68.4 4.55 4 0 62.0 68.4 -6.45 11 0 85.0 81.8 3.24 11 0 89.0 81.8 7.2 4 16 0 83.0 84.5 -1.48 16 0 86.0 84.5 1.5 2 CVGraph 6.02 04/1 21201 7 Page 212 Figure 2-6 (Continued) | L T Ma terial: SA5 33Bl Capsu le: UNIRRA Heat: C 4114-2 F lu e nce: O.OOE+OOO n/cm 2 UNIRRADIATED PLATE HEAT C4114-2 LE Charpy V-Notch Data Te mp e r a tur e(° F) In put L. E. C omput e d L. E. Differ e nti a l -80 23.0 16.9 6.1 0 -80 20.0 16.9 3.1 0 -40 25.0 32.5 -7.53 -4 0 29.0 32.5 -3.53 -40 35.0 32.5 2.47 1 0 57.0 56.8 0.24 10 58.0 56.8 1.24 1 0 59.0 56.8 2.2 4 4 0 66.0 68.4 -2.45 4 0 73.0 68.4 4.55 4 0 62.0 68.4 -6.45 11 0 85.0 81.8 3.24 11 0 89.0 81.8 7.2 4 16 0 83.0 84.5 -1.48 16 0 86.0 84.5 1.5 2 CVGraph 6.02 04/1 21201 7 Page 212 Figure 2-6 (Continued) | ||
Lateral Expansion Plot for Plate Heat C4114-2 (LT) Un irradiated 2-13 Mat e rials and T e st Sp ec im en D esc ription 2.3 Capsule Opening As shown in Figures 2-2 through 2-4 , the 300° surveillance capsule consisted of a container holding two Charpy packets and three tensile tubes. Each Charpy packet contained 12 Charpy specimens. | Lateral Expansion Plot for Plate Heat C4114-2 (LT) Un irradiated 2-13 Mat e rials and T e st Sp ec im en D esc ription 2.3 Capsule Opening As shown in Figures 2-2 through 2-4 , the 300° surveillance capsule consisted of a container holding two Charpy packets and three tensile tubes. Each Charpy packet contained 12 Charpy specimens. | ||
The outside of the capsule had identification markings which could be clearly read. On one side, the capsule container was marked with the reactor and capsule codes. The reactor code matches the reactor code in Reference | The outside of the capsule had identification markings which could be clearly read. On one side, the capsule container was marked with the reactor and capsule codes. The reactor code matches the reactor code in Reference | ||
[8]. The capsule container was engraved with the marking "117C4671G3" on the side facing the core. Attention was paid to the location of the Charpy packets and specimen and dosimetry wires during disassembly of the capsule. The dosimetry wire location along the ends of the Charpy specimens is shown in Figure 2-7. Referring to the figure, the eight base metal specimens and four weld specimens in the G4 Charpy packet were installed at the bottom of the capsule and the eight HAZ specimens and four weld specimens in the G5 Charpy packet were installed in the top of the capsule. Specimen orientation within the packets can be seen in Figures 2-8 and 2-9. The dosimetry wires and Charpy specimens were placed in individually marked containers for positive identification throughout the work. During disassembly , the Charpy specimens in the G4 packet were found to be heavily oxidized (Figure 2-9) due to water in-leakage. Prior to testing , the Charpy specimens were carefully cleaned by hand lapping in a specially designed fixture. Every effort was made to preserve the ASTM E23 [12] dimensional tolerances and surface finish requirements. Despite all of the efforts to keep the test specimens within dimensional tolerances , some had experienced corrosive attack that resulted in an out-of-tolerance condition after cleaning. | [8]. The capsule container was engraved with the marking "117C4671G3" on the side facing the core. Attention was paid to the location of the Charpy packets and specimen and dosimetry wires during disassembly of the capsule. The dosimetry wire location along the ends of the Charpy specimens is shown in Figure 2-7. Referring to the figure, the eight base metal specimens and four weld specimens in the G4 Charpy packet were installed at the bottom of the capsule and the eight HAZ specimens and four weld specimens in the G5 Charpy packet were installed in the top of the capsule. Specimen orientation within the packets can be seen in Figures 2-8 and 2-9. The dosimetry wires and Charpy specimens were placed in individually marked containers for positive identification throughout the work. During disassembly , the Charpy specimens in the G4 packet were found to be heavily oxidized (Figure 2-9) due to water in-leakage. Prior to testing , the Charpy specimens were carefully cleaned by hand lapping in a specially designed fixture. Every effort was made to preserve the ASTM E23 [12] dimensional tolerances and surface finish requirements. Despite all of the efforts to keep the test specimens within dimensional tolerances , some had experienced corrosive attack that resulted in an out-of-tolerance condition after cleaning. | ||
After the hand lapping , the MPM Digital Optical Comparator (DOC) was used to measure the notch included angle, notch depth , and notch radius. A calibrated micrometer was used to measure the cross-sectional dimensions. The remaining ligament, or distance from the root of the notch to the back side of the specimen across the uncracked ligament , was found by subtracting the notch depth from the specimen width. The results of these measurement s are provided in Table 2-5. All of the specimens with an out-of-tolerance measurement are shown in the table. The data for the G5 packet, which did not l eak , are also shown in the tab l e for comparison. | After the hand lapping , the MPM Digital Optical Comparator (DOC) was used to measure the notch included angle, notch depth , and notch radius. A calibrated micrometer was used to measure the cross-sectional dimensions. The remaining ligament, or distance from the root of the notch to the back side of the specimen across the uncracked ligament , was found by subtracting the notch depth from the specimen width. The results of these measurement s are provided in Table 2-5. All of the specimens with an out-of-tolerance measurement are shown in the table. The data for the G5 packet, which did not l eak , are also shown in the tab l e for comparison. | ||
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The NRC has established guidelines in Regulatory Guide 1.190 [ 13] for determining best estimate values of flux, energy spectrum, and fluence for RPV d amage assessments using computationa l methods. These guidelines are not specifically intend e d for use in surveillance capsule evaluations; however , the guidelines provide suitable guida n ce to support the development of accurate neutron transport analysis models for surveillance capsule evaluations. | The NRC has established guidelines in Regulatory Guide 1.190 [ 13] for determining best estimate values of flux, energy spectrum, and fluence for RPV d amage assessments using computationa l methods. These guidelines are not specifically intend e d for use in surveillance capsule evaluations; however , the guidelines provide suitable guida n ce to support the development of accurate neutron transport analysis models for surveillance capsule evaluations. | ||
This section docume n ts the application of the modeling and analysis guidelines provided in [14] to determine the surveillance capsule accumulated irradiation (e.g., activation) and capsule specimen neutron fluence of the Hatch Unit 1 300° surveillance capsule flux wires. Additionally, the activation for the 30° flux wires , removed at the end of Cycle 1 , 30° surveillance capsule, removed at the end of Cycle 8, and 120° surveillance capsule, removed at the end of Cycle 16, was determined. | This section docume n ts the application of the modeling and analysis guidelines provided in [14] to determine the surveillance capsule accumulated irradiation (e.g., activation) and capsule specimen neutron fluence of the Hatch Unit 1 300° surveillance capsule flux wires. Additionally, the activation for the 30° flux wires , removed at the end of Cycle 1 , 30° surveillance capsule, removed at the end of Cycle 8, and 120° surveillance capsule, removed at the end of Cycle 16, was determined. | ||
The fast neutron fluence (E >1.0 MeV) was also calculated for all capsules at the time of removal. The fluence and activation values presented in this report were calculated using the RAMA Fluence Methodology | The fast neutron fluence (E >1.0 MeV) was also calculated for all capsules at the time of removal. The fluence and activation values presented in this report were calculated using the RAMA Fluence Methodology | ||
[14] (hereinafter referred to as "RAMA"). The specific activities predicted by RAMA are compared to the activity measurements reported in Appendix A. RAMA has been developed for the Electric Power Research Institute, Inc. (EPRl) and the BWRVIP for the purpose of calculating neutron fluence in Boiling Water Reactor (BWR) compo n ents. As prescribed in Regulatory Guide 1.190, RAMA has been benchmarked against industry standard benchmarks for both BWR and pressurized water reactor (PWR) designs. In additio n , RAMA has been compared with several plant-specific dosimetry measurements and reported fluences from several commercial operating reactors. The results of the benchmarks and comparisons to measurements show that RAMA accurately predicts specimen activities, RPV fluence, and vessel internal component fluence in all light-water reactor types. Under funding from E P RI and the BWRVIP, the RAMA methodology has been reviewed by the U.S. NRC and subsequently given generic approval for determining fast neutron fluence in BWR and PWR pressur e vessels [15 , 16] and BWR vessel internal components that include the core shroud and top guide [17]. 3-1 Ne utr o n Flu e n ce C a l c ul a tion 3.1 Description of the Reactor System This section provides an overview of the reactor design and operating data inputs that were used to develop the Hatch Unit 1 reactor fluence model. All reactor design and operating data inputs used to develop the model were plant-specific and were provided by the Hatch Unit 1 reactor operator, Southern Nuclear Operating Company, Inc. (SNOC). The inputs for the fluence geometry model were developed from design and as-built drawings for the RPV vessel internals, fuel assemblies, and containment regions. The reactor operating data inputs were de v eloped from core simulator data , when available, and other sources, that provided a historical accounting of how the reactor operated for Cycles 1 through 27. 3. 1. 1 Overview of the Reactor System Design Hatch Unit 1 is a General Electric BWR/4 class reactor lo cated near Bax l ey , GA. The reactor has a core loading of 560 fuel assemb l ies and began commercia l operation in December 1975 with a design rated power of 2436 MWt. In Cycle 17 , a power up-rate was achieved , raising the power to 2558 MWt. Two additional up-rates were achieved: | [14] (hereinafter referred to as "RAMA"). The specific activities predicted by RAMA are compared to the activity measurements reported in Appendix A. RAMA has been developed for the Electric Power Research Institute, Inc. (EPRl) and the BWRVIP for the purpose of calculating neutron fluence in Boiling Water Reactor (BWR) compo n ents. As prescribed in Regulatory Guide 1.190, RAMA has been benchmarked against industry standard benchmarks for both BWR and pressurized water reactor (PWR) designs. In additio n , RAMA has been compared with several plant-specific dosimetry measurements and reported fluences from several commercial operating reactors. The results of the benchmarks and comparisons to measurements show that RAMA accurately predicts specimen activities, RPV fluence, and vessel internal component fluence in all light-water reactor types. Under funding from E P RI and the BWRVIP, the RAMA methodology has been reviewed by the U.S. NRC and subsequently given generic approval for determining fast neutron fluence in BWR and PWR pressur e vessels [15 , 16] and BWR vessel internal components that include the core shroud and top guide [17]. 3-1 Ne utr o n Flu e n ce C a l c ul a tion 3.1 Description of the Reactor System This section provides an overview of the reactor design and operating data inputs that were used to develop the Hatch Unit 1 reactor fluence model. All reactor design and operating data inputs used to develop the model were plant-specific and were provided by the Hatch Unit 1 reactor operator, Southern Nuclear Operating Company, Inc. (SNOC). The inputs for the fluence geometry model were developed from design and as-built drawings for the RPV vessel internals, fuel assemblies, and containment regions. The reactor operating data inputs were de v eloped from core simulator data , when available, and other sources, that provided a historical accounting of how the reactor operated for Cycles 1 through 27. 3. 1. 1 Overview of the Reactor System Design Hatch Unit 1 is a General Electric BWR/4 class reactor lo cated near Bax l ey , GA. The reactor has a core loading of 560 fuel assemb l ies and began commercia l operation in December 1975 with a design rated power of 2436 MWt. In Cycle 17 , a power up-rate was achieved , raising the power to 2558 MWt. Two additional up-rates were achieved: | ||
in Cycle 19 to 2763 MWt, and in Cycle 22 to 2804 MWt. At the time of this fluence analysis , Hatch Unit 1 h a d completed 27 cycles of operation. Figure 3-1 illustrates the basic planar configuration of the Hatch Unit 1 reactor at an axial elevation near the reac t or core mid-plane. | in Cycle 19 to 2763 MWt, and in Cycle 22 to 2804 MWt. At the time of this fluence analysis , Hatch Unit 1 h a d completed 27 cycles of operation. Figure 3-1 illustrates the basic planar configuration of the Hatch Unit 1 reactor at an axial elevation near the reac t or core mid-plane. | ||
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The data calculated with core simulator codes represents the best-available information about the reactor core's operating history over the reactor's operating life. In this analysis, the core simulator data was processed by Trans Ware to generate state-point data files for input to the fluence model. The state-point files included three-dimensional data arrays that described core power distributions , fuel exposure distributions, fuel materials (isotopics), and coolant water densities. | The data calculated with core simulator codes represents the best-available information about the reactor core's operating history over the reactor's operating life. In this analysis, the core simulator data was processed by Trans Ware to generate state-point data files for input to the fluence model. The state-point files included three-dimensional data arrays that described core power distributions , fuel exposure distributions, fuel materials (isotopics), and coolant water densities. | ||
A separate neutron transport calculation was performed for each of the state-points tallied in Table 3-3. The calculated neutron flux for each state-point was combined with the appropriate power history data described in Section 3 .1.4.2 in order to provide an accurate accounting of the fast neutron fluence for the RPV and activation of the surveillance capsules. 3. 1.4. 3. 1 Beginning of Operation through Cycle 15 State-Points Cycles 1-15 required a state-point construction technique to approximate the missing "core simulator" data. In effect, power shapes, isotopics , water densities, and exposures are mapped onto a cycle-dominating fuel assembly type throughout the core. This also required addressing variances in the active fuel height for the loaded assemblies. | A separate neutron transport calculation was performed for each of the state-points tallied in Table 3-3. The calculated neutron flux for each state-point was combined with the appropriate power history data described in Section 3 .1.4.2 in order to provide an accurate accounting of the fast neutron fluence for the RPV and activation of the surveillance capsules. 3. 1.4. 3. 1 Beginning of Operation through Cycle 15 State-Points Cycles 1-15 required a state-point construction technique to approximate the missing "core simulator" data. In effect, power shapes, isotopics , water densities, and exposures are mapped onto a cycle-dominating fuel assembly type throughout the core. This also required addressing variances in the active fuel height for the loaded assemblies. | ||
To avoid spurious power mappings (e.g., lOxlO pin powers on 7x7 assemb l y), pin power data from other cycles was not used on the constructed cycles , and only nodal power distributions were applied. Any data still missing after attempting to use cycle-specific resources was based on Cycle 16, primarily for moderator densities. For Cycles 1-3 , traversing in-core probe (TIP) trace data was available to determine the axial power distribution. Additionally, the core design and operating data was previously characterized | To avoid spurious power mappings (e.g., lOxlO pin powers on 7x7 assemb l y), pin power data from other cycles was not used on the constructed cycles , and only nodal power distributions were applied. Any data still missing after attempting to use cycle-specific resources was based on Cycle 16, primarily for moderator densities. For Cycles 1-3 , traversing in-core probe (TIP) trace data was available to determine the axial power distribution. Additionally, the core design and operating data was previously characterized | ||
[18 , 19). Combined with assembly edge-to-near-edge power ratios and exposures, this information was expanded into state-points. | [18 , 19). Combined with assembly edge-to-near-edge power ratios and exposures, this information was expanded into state-points. | ||
For Cycles 4-12 , process computer exposure and void (E&V) data at the local power range monitor (LPRM) locations was expanded into cycle-average nodal datasets. Fuel assembly loadings and shuffles were tracked across each cycle to generate the exposure distribution. | For Cycles 4-12 , process computer exposure and void (E&V) data at the local power range monitor (LPRM) locations was expanded into cycle-average nodal datasets. Fuel assembly loadings and shuffles were tracked across each cycle to generate the exposure distribution. | ||
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The shroud is composed of several components, of which the following are modeled and discussed here: upper shroud , central shroud , and lower shroud. The shroud head is discu s sed with other above-core components. | The shroud is composed of several components, of which the following are modeled and discussed here: upper shroud , central shroud , and lower shroud. The shroud head is discu s sed with other above-core components. | ||
3-14 Ne utr o n Flu e n ce Cal c ulation 3.2.2.3.1 Upper Shroud The upper shroud wall mates the shr oud head to the top guide flan ge and central shroud wall. The wall protrudes beyond the central shroud regions and is modeled with cylindrical elements. | 3-14 Ne utr o n Flu e n ce Cal c ulation 3.2.2.3.1 Upper Shroud The upper shroud wall mates the shr oud head to the top guide flan ge and central shroud wall. The wall protrudes beyond the central shroud regions and is modeled with cylindrical elements. | ||
3.2.2.3.2 Central Shroud Axia lly , the shroud extends from the core support plate flange to the top guide flange. The central shro ud wall is modeled with cylindrical elements. | 3.2.2.3.2 Central Shroud Axia lly , the shroud extends from the core support plate flange to the top guide flange. The central shro ud wall is modeled with cylindrical elements. | ||
: 3. 2. 2. 3. 3 Lower Shroud Axially, the lowe r shroud extends from the bottom of the model to the core support plate flange. The lower shroud wa ll is modeled with conical elements to maintain the physical component's true s h ape. 3.2.2.4 Downcomer Region Model The d o wncomer region lies between the core shroud and the RPV. It is effective ly cylindrical in design , but with some geometrical complexities created by the presence of jet pumps and surveillance capsules in the region. The majority of the do wn comer region is modeled with cylindrical elements. | : 3. 2. 2. 3. 3 Lower Shroud Axially, the lowe r shroud extends from the bottom of the model to the core support plate flange. The lower shroud wa ll is modeled with conical elements to maintain the physical component's true s h ape. 3.2.2.4 Downcomer Region Model The d o wncomer region lies between the core shroud and the RPV. It is effective ly cylindrical in design , but with some geometrical complexities created by the presence of jet pumps and surveillance capsules in the region. The majority of the do wn comer region is modeled with cylindrical elements. | ||
The areas of the downcomer containing the jet pumps , tie rods , and specimen capsules are modeled with the appropriate geometry elements to represent their design features and to preserve their radial , azimutha l , and axial placement in the downcomer region. These structures are described further in the following subsections. | The areas of the downcomer containing the jet pumps , tie rods , and specimen capsules are modeled with the appropriate geometry elements to represent their design features and to preserve their radial , azimutha l , and axial placement in the downcomer region. These structures are described further in the following subsections. | ||
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However, by Cycle 8, an acceptable comparison developed, indicating that the approach remains valid. Excluding the Cycle 1 flux wires from the overall comparison yields a C/M ratio of 1.00 +/- 0.09. Best estimate fast fluence (E > 1. 0 Me V) was calculated for all removed capsules and the 3 0° survei ll ance capsule flux wire holder. Lead factors are determined and reported for all capsules. 3.3.1 Comparison of Predicted Activation to Plant-specific Measurements The comparison of predicted activation for the Hatch Unit 1 Cycle 1 , 8 , 16 , and 27 flux wires to measurements is presented in this subsection. | However, by Cycle 8, an acceptable comparison developed, indicating that the approach remains valid. Excluding the Cycle 1 flux wires from the overall comparison yields a C/M ratio of 1.00 +/- 0.09. Best estimate fast fluence (E > 1. 0 Me V) was calculated for all removed capsules and the 3 0° survei ll ance capsule flux wire holder. Lead factors are determined and reported for all capsules. 3.3.1 Comparison of Predicted Activation to Plant-specific Measurements The comparison of predicted activation for the Hatch Unit 1 Cycle 1 , 8 , 16 , and 27 flux wires to measurements is presented in this subsection. | ||
Fluence values are also calculated and reported in Section 3.3.2 for each of the capsule flux wires. 3.3.1.1 Cycle 1 30° Flux Wire Holder Activation Analysis Copper and iron flux wires were irradiated in the Hatch Unit 1 surveillance capsule flux wire holder at the 30° azimuth during the first cycle of operation. | Fluence values are also calculated and reported in Section 3.3.2 for each of the capsule flux wires. 3.3.1.1 Cycle 1 30° Flux Wire Holder Activation Analysis Copper and iron flux wires were irradiated in the Hatch Unit 1 surveillance capsule flux wire holder at the 30° azimuth during the first cycle of operation. | ||
The wires were removed after being irradiated for a total of 1.2 EFPY. Activation measurements were performed following irradiation for the following reactions | The wires were removed after being irradiated for a total of 1.2 EFPY. Activation measurements were performed following irradiation for the following reactions | ||
[10]: 63 Cu (n,a) 6°Co and 54 Fe (n , p) 54 Mn. The precise location of the individual wires within the surveillance capsule flux wire holder is not known; therefore, the activation calculations were performed at the center of the holder. Table 3-4 provides a comparison of the RAMA calculated specific activities and the measured specific activities for the flux wire specimens. | [10]: 63 Cu (n,a) 6°Co and 54 Fe (n , p) 54 Mn. The precise location of the individual wires within the surveillance capsule flux wire holder is not known; therefore, the activation calculations were performed at the center of the holder. Table 3-4 provides a comparison of the RAMA calculated specific activities and the measured specific activities for the flux wire specimens. | ||
The Cycle 1 total flu x wire average measured (C/M) value is 0.78 with a standard deviation of +/-0.03. 3-20 N e utr o n Fl u e n ce Ca l c ul a tion T a ble 3-4 Comparison of Specific Activities for Hatch Un i t 1 Cycle 1 30° Flux W i re Holder Wires (C/M) Flux W i res Measured Calculated Calculated vs. Standard (dps/g) (dps/g) Measured (C/M) Deviat i on (a) Iron I r on 1 1.10E+05 8.98E+04 0.82 -Iron 2 1.11E+0 5 8.98E+0 4 0.81 -Iron 3 1.12E+0 5 8.98E+0 4 0.80 -Av e rage --0.81 0.01 Copper C opper1 5.3 6E+0 3 4.0 4 E+0 3 0.75 -C opper 2 5.4 6E+0 3 4.0 4 E+0 3 0.7 4 -C opper 3 5.4 0E+03 4.04E+03 0.7 5 -Av e rage --0.75 0.01 Tota l Flux Wire --0.78 0.03 Aver a ge 3.3.1.2 Cycle 8 30° Surveillance Capsule Activation Analysis Copper, iron, and nickel flux wires were irra d iated in the Hatch Unit 1 s ur vei ll ance caps ul e at the 30° azimuth d uring t h e fir st 8 c y cles of operation. The wire s were removed after being irradiate d for a t o tal of 5.5 EF P Y. Activation meas u reme n ts were performed following irra d iatio n for the fo ll owing react i ons [8): 63 Cu (n , a) 6°Co , 54 Fe (n , p) 5 4 Mn , and 58 Ni (n , p) 58 Co. T h e precise l ocation of t h e indiv i dual wires within t h e survei ll ance capsule is not known; therefore , t h e activa t i on calc ul ations wer e performe d at the ce n ter of t h e caps ul e ho ld er. Tab l e 3-5 provi d es a comparison of t h e RAMA calcu l ate d specific activ it ies an d the meas u red spec i fi c activ i t i es for t h e survei ll ance cap s ule flux w ire s pecime n s. The Cycle 8 cap s u l e total fl u x wire average C/M v alue is 0.94 wi th a sta nd ard d ev i ation of +/-0.09. 3-21 N e utron Flu e n c e Ca l c ulatio n Table 3-5 Comparison of Specific Activities for Hatch Unit 1 Cycle 8 30° Surveillance Capsule Flux Wires (C/M) Flux Wires Measured Calculated Calculated vs. Standard (dps/g) (dps/g) Measured (C/M) Deviation (a) Iron Iron G1 7.62E+04 7.83E+04 1.03 -Iron G2 7.94E+04 7.83E+04 0.99 -Iron G3 7.5 1E+04 7.83E+04 1.04 -Average --1.02 0.03 Nickel Nickel G1 1.21 E+06 1.21 E+06 1.00 -Nickel G2 1.23E+06 1.21 E+06 0.98 -Nickel G3 1.23E+06 1.21E+06 0.98 -Average --0.99 0.01 Copper Copper G1 1.10E+04 9.00E+03 0.82 -Copper G2 1.10E+04 9.00E+03 0.82 -Copper G3 1.06E+04 9.00E+03 0.85 -Average --0.83 0.02 Total Flux Wire 0.94 0.09 Average --3.3.1.3 Cycle 16 120° Surveillance Capsule Activation Analysis Copper, iron, and nickel flux wires were irradiated in the Hatch Unit 1 surveillance capsule at the 120° azimuth during the first 16 cycles of operation. | The Cycle 1 total flu x wire average measured (C/M) value is 0.78 with a standard deviation of +/-0.03. 3-20 N e utr o n Fl u e n ce Ca l c ul a tion T a ble 3-4 Comparison of Specific Activities for Hatch Un i t 1 Cycle 1 30° Flux W i re Holder Wires (C/M) Flux W i res Measured Calculated Calculated vs. Standard (dps/g) (dps/g) Measured (C/M) Deviat i on (a) Iron I r on 1 1.10E+05 8.98E+04 0.82 -Iron 2 1.11E+0 5 8.98E+0 4 0.81 -Iron 3 1.12E+0 5 8.98E+0 4 0.80 -Av e rage --0.81 0.01 Copper C opper1 5.3 6E+0 3 4.0 4 E+0 3 0.75 -C opper 2 5.4 6E+0 3 4.0 4 E+0 3 0.7 4 -C opper 3 5.4 0E+03 4.04E+03 0.7 5 -Av e rage --0.75 0.01 Tota l Flux Wire --0.78 0.03 Aver a ge 3.3.1.2 Cycle 8 30° Surveillance Capsule Activation Analysis Copper, iron, and nickel flux wires were irra d iated in the Hatch Unit 1 s ur vei ll ance caps ul e at the 30° azimuth d uring t h e fir st 8 c y cles of operation. The wire s were removed after being irradiate d for a t o tal of 5.5 EF P Y. Activation meas u reme n ts were performed following irra d iatio n for the fo ll owing react i ons [8): 63 Cu (n , a) 6°Co , 54 Fe (n , p) 5 4 Mn , and 58 Ni (n , p) 58 Co. T h e precise l ocation of t h e indiv i dual wires within t h e survei ll ance capsule is not known; therefore , t h e activa t i on calc ul ations wer e performe d at the ce n ter of t h e caps ul e ho ld er. Tab l e 3-5 provi d es a comparison of t h e RAMA calcu l ate d specific activ it ies an d the meas u red spec i fi c activ i t i es for t h e survei ll ance cap s ule flux w ire s pecime n s. The Cycle 8 cap s u l e total fl u x wire average C/M v alue is 0.94 wi th a sta nd ard d ev i ation of +/-0.09. 3-21 N e utron Flu e n c e Ca l c ulatio n Table 3-5 Comparison of Specific Activities for Hatch Unit 1 Cycle 8 30° Surveillance Capsule Flux Wires (C/M) Flux Wires Measured Calculated Calculated vs. Standard (dps/g) (dps/g) Measured (C/M) Deviation (a) Iron Iron G1 7.62E+04 7.83E+04 1.03 -Iron G2 7.94E+04 7.83E+04 0.99 -Iron G3 7.5 1E+04 7.83E+04 1.04 -Average --1.02 0.03 Nickel Nickel G1 1.21 E+06 1.21 E+06 1.00 -Nickel G2 1.23E+06 1.21 E+06 0.98 -Nickel G3 1.23E+06 1.21E+06 0.98 -Average --0.99 0.01 Copper Copper G1 1.10E+04 9.00E+03 0.82 -Copper G2 1.10E+04 9.00E+03 0.82 -Copper G3 1.06E+04 9.00E+03 0.85 -Average --0.83 0.02 Total Flux Wire 0.94 0.09 Average --3.3.1.3 Cycle 16 120° Surveillance Capsule Activation Analysis Copper, iron, and nickel flux wires were irradiated in the Hatch Unit 1 surveillance capsule at the 120° azimuth during the first 16 cycles of operation. | ||
The wires were removed after being irradiated for a total of 14.2 EFPY. Activation measurements were performed following irradiation for the following reactions | The wires were removed after being irradiated for a total of 14.2 EFPY. Activation measurements were performed following irradiation for the following reactions | ||
[9]: 63 Cu (n , a) 6°Co, 54 Fe (n,p) 54 Mn, and 5 8 Ni (n,p) 5 8 Co. The precise location of th e in dividu a l w ires w ithin the surveillance capsule is not known; therefore, the activation calculations were performed at th e center of the capsule holder. Table 3-6 provides a comparison of the RAMA calculated specific activities and the measured specific activities for the surve illan ce capsule flux wire specimens. | [9]: 63 Cu (n , a) 6°Co, 54 Fe (n,p) 54 Mn, and 5 8 Ni (n,p) 5 8 Co. The precise location of th e in dividu a l w ires w ithin the surveillance capsule is not known; therefore, the activation calculations were performed at th e center of the capsule holder. Table 3-6 provides a comparison of the RAMA calculated specific activities and the measured specific activities for the surve illan ce capsule flux wire specimens. | ||
The Cycle 16 capsule total flux wire average C/M value is 1.04 with a sta ndard de viation of +/-0.08. 3-22 Neutro n Flu e n ce Calculation T a ble 3-6 Comparison of Specific Activities for Hatch Unit 1 Cycle 16 120° Surveillance Capsule Flux Wires (C/M) Flux Wires Measured Calculated Calculated vs. Standard (dps/g) (dps/g) Measured (C/M) Deviation (cr) Iron 9.55E+04 1.01 E+05 1.06 -Nick e l 1.18E+06 1.21 E+06 1.11 -Copper 1.54E+04 1.47E+04 0.96 -Total Flux Wire 1.04 0.08 Average --3.3.1.4 Cycle 27 300° Surveillance Capsule Activation Analysis Copp e r , iron, and nickel flux wires were irradiated in the Hatch Unit 1 surveillance capsule at the 3 0 0° azimuth during the first 27 cycles of operation. The wires were removed after being irradiated for a total of 32.0 EFPY. Activation measurements were performed following irradiation for the following reactions (See Appendix A): 63 C u (n ,a) 6°C o , 54 Fe (n,p) 54 Mn , and 58 N i (n,p) 58 Co. The precise location of the individual wires within the surveillance capsule is not known; therefore , the activation calculations were performed at the center of the capsule holder. Tab le 3-7 provides a comparison of the RAMA calculated specific activities and the measured specific activities for the surveillance capsule flux wire specimens. | The Cycle 16 capsule total flux wire average C/M value is 1.04 with a sta ndard de viation of +/-0.08. 3-22 Neutro n Flu e n ce Calculation T a ble 3-6 Comparison of Specific Activities for Hatch Unit 1 Cycle 16 120° Surveillance Capsule Flux Wires (C/M) Flux Wires Measured Calculated Calculated vs. Standard (dps/g) (dps/g) Measured (C/M) Deviation (cr) Iron 9.55E+04 1.01 E+05 1.06 -Nick e l 1.18E+06 1.21 E+06 1.11 -Copper 1.54E+04 1.47E+04 0.96 -Total Flux Wire 1.04 0.08 Average --3.3.1.4 Cycle 27 300° Surveillance Capsule Activation Analysis Copp e r , iron, and nickel flux wires were irradiated in the Hatch Unit 1 surveillance capsule at the 3 0 0° azimuth during the first 27 cycles of operation. The wires were removed after being irradiated for a total of 32.0 EFPY. Activation measurements were performed following irradiation for the following reactions (See Appendix A): 63 C u (n ,a) 6°C o , 54 Fe (n,p) 54 Mn , and 58 N i (n,p) 58 Co. The precise location of the individual wires within the surveillance capsule is not known; therefore , the activation calculations were performed at the center of the capsule holder. Tab le 3-7 provides a comparison of the RAMA calculated specific activities and the measured specific activities for the surveillance capsule flux wire specimens. | ||
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Charpy V-Notch Data Temperat ur e{° F) Input CVN Co mput e d CVN Differential | Charpy V-Notch Data Temperat ur e{° F) Input CVN Co mput e d CVN Differential | ||
-49 9.2 9.1 0.10 -2 17.7 21.8 -4.1 6 17 38.2 31.2 7.04 37 47.2 44.0 3.2 4 72 6 4.2 73.1 -8.99 12 0 1175 1 12.5 5.03 242 147.1 145.2 1.88 372 147.6 147.3 0.33 CVGrap h 6.02 05 1 1 2120 17 Page 212 Figure 5-1 (Continued) | -49 9.2 9.1 0.10 -2 17.7 21.8 -4.1 6 17 38.2 31.2 7.04 37 47.2 44.0 3.2 4 72 6 4.2 73.1 -8.99 12 0 1175 1 12.5 5.03 242 147.1 145.2 1.88 372 147.6 147.3 0.33 CVGrap h 6.02 05 1 1 2120 17 Page 212 Figure 5-1 (Continued) | ||
Irradiated Plate Heat C4114-2 Charpy Energy Plot (Hatch Unit 1 300° Capsule) (LT) 5-3 C h a r py T est R es ult s P l an!: Ha t c h 1 Ori e n ta tio n: LT ..-rl.> -*-6 '-' = 0 *-rl.> = -i.. ... 90 70 60 50 40 30 .... 20 .... 10 .... 0 -300 CV G nt ph 6.02 Figure 5-2 IRRADIATED PLATE HEAT C4114-2 LE (HAt-300) -200 CV G ra ph 6.02: H yperbo li c T a n ge nt Curve Print ed o n 5/1 2/20 1 7 2: 23 PM A= 4 2.45 B = 41.45 C = 86.4 0 TO = 56.94 D = 0.00 Co rrcl a tjon Coeffic i e nt = 0. 99 3 E qu a ti on is A+ B * [Tanh((T-TO)/(C+0 1))] U p pe r S h elf LE.= 8 3.90 (F i xed) L owe r S h e lf L.E. = 1.00 (Fi.-..:ed) T e mp.@3 5 m il s= 4 1.3 0° F Ma t e ri a l: SA533 Bl Ca psu le: 300 D EG / I I fo j I 0 ,,,,,----100 0 100 200 300 Temperature | Irradiated Plate Heat C4114-2 Charpy Energy Plot (Hatch Unit 1 300° Capsule) (LT) 5-3 C h a r py T est R es ult s P l an!: Ha t c h 1 Ori e n ta tio n: LT ..-rl.> -*-6 '-' = 0 *-rl.> = -i.. ... 90 70 60 50 40 30 .... 20 .... 10 .... 0 -300 CV G nt ph 6.02 Figure 5-2 IRRADIATED PLATE HEAT C4114-2 LE (HAt-300) -200 CV G ra ph 6.02: H yperbo li c T a n ge nt Curve Print ed o n 5/1 2/20 1 7 2: 23 PM A= 4 2.45 B = 41.45 C = 86.4 0 TO = 56.94 D = 0.00 Co rrcl a tjon Coeffic i e nt = 0. 99 3 E qu a ti on is A+ B * [Tanh((T-TO)/(C+0 1))] U p pe r S h elf LE.= 8 3.90 (F i xed) L owe r S h e lf L.E. = 1.00 (Fi.-..:ed) T e mp.@3 5 m il s= 4 1.3 0° F Ma t e ri a l: SA533 Bl Ca psu le: 300 D EG / I I fo j I 0 ,,,,,----100 0 100 200 300 Temperature | ||
{° F) 05/l 2/20 1 7 u 400 H ea t C4114-2 F lu en ce: nla 500 600 Page 112 Ir r adiated Plate Heat C4114-2 Lateral Expansion P l ot (Hatch Unit 1 300° Capsule) (LT) 5-4 Chmpy T est R es ults Plant: Hatc h 1 Orie n ta ti o n: LT Ma teri a l: SA533Bl Ca p s ule: 300 DEG Heat: C4114-2 F luen ce: n/a IRRADIATED PLATE HEAT C4114-2 LE (HAl-300) | {° F) 05/l 2/20 1 7 u 400 H ea t C4114-2 F lu en ce: nla 500 600 Page 112 Ir r adiated Plate Heat C4114-2 Lateral Expansion P l ot (Hatch Unit 1 300° Capsule) (LT) 5-4 Chmpy T est R es ults Plant: Hatc h 1 Orie n ta ti o n: LT Ma teri a l: SA533Bl Ca p s ule: 300 DEG Heat: C4114-2 F luen ce: n/a IRRADIATED PLATE HEAT C4114-2 LE (HAl-300) | ||
Charpy V-Notch Data Temperature(° F) InputL E. Computed LE. Differential | Charpy V-Notch Data Temperature(° F) InputL E. Computed LE. Differential | ||
-49 6.6 7.6 -1.00 -2 1 5.6 1 8.0 -2.43 17 3 0.7 2 4.7 6.0 0 37 33.1 33.0 0.10 72 44.0 4 9.4 -5.41 1 20 715 68.2 3.32 242 85.0 82.8 2.22 3 72 82.8 83.8 -1.0 4 C VGraph 6.02 0511 2/2 017 Page 212 Figure5-2 (Continued) | -49 6.6 7.6 -1.00 -2 1 5.6 1 8.0 -2.43 17 3 0.7 2 4.7 6.0 0 37 33.1 33.0 0.10 72 44.0 4 9.4 -5.41 1 20 715 68.2 3.32 242 85.0 82.8 2.22 3 72 82.8 83.8 -1.0 4 C VGraph 6.02 0511 2/2 017 Page 212 Figure5-2 (Continued) | ||
Irradiated Plate Heat C4114-2 Lateral Expansion Plot (Hatch Unit 1 300° Capsule) (LT) 5-5 C h a r py T e st R e sults 5-6 Table 5-1 Effect of Irradiation (E>1.0 MeV) on the Notch Toughness Properties T30* 30 ft-lb (40.7 J) T 50* 50 ft-lb (67 .8 J) T 3 5 mil* 35 mil CVN Upper Shelf Energy Transition Temperature Transition Temperature (0.89 mm) Lateral (USE) Material Expansion Temperature Identity Unirrad Irradiated ilT30 Unirrad Irradiated ilT50 Unirrad Irradiated ilT35mil Unirrad Irradiated Change OF OF OF OF OF OF OF OF OF ft-lb ft-lb ft-lb (oC) (oC) (oC) (oC) (oC) (oC) (oC) (oC) (oC) (J) (J) (J) C4114-2 -61.5 15.4 76.9 -28.8 44.8 73.6 -34.7 41.3 76.0 136.0 147.4 11.4 (LT orientation) (-51.9) (-9.2) (42.7) (-33.8) (7.1) (40.9) (-37.1) (5.2) (42.2) (184.4) (199.8) (15.4) Table 5-2 Comparison of Actual Versus Predicted Embrittlement Fluence RG 1.99 Rev. 2 RG 1.99 Rev. 2 Measured Shift 2 Predicted Ident i ty Material (E>1.0 MeV , x10 17 O F (OC) Predicted Shift 3 Shift+Margin 3*4 n/cm 2)1 OF (OC) OF (OC) C411 4-2 Hat c h U ni t 1 su rv e i lla n ce plate 1 3.8 76.9 (42.7) 4 0.9 (22.7) 7 4.9(41.6) (LT or i entat i on) l. Fluence value i s reported in Table 3-9. 2. The measured shift is taken from Tab l e 5-1. 3. Predicted shift= CF x FF, where CF is a Chemistry Factor taken from the base metal table in USNRC RG 1.99 , Rev. 2 [6], based on each material's Cu/N i co n te n t, and FF is F lu ence Factor, f0.28-0. l 0 log f, where f = tluence i n un i ts of 10 1 9 n/cm 2 (E > 1.0 MeV) specified. 4. Margin= 2./(G;2 + G t.2), where G; =the standard deviation on initial RT N D T (G; is taken to be 0°F), and G t. is the standard deviation on LlRT NDT (28°F for welds and l 7°F for base materials, except that G t. need not exceed 0.50 times the mean value of ilRT N o T). Thus, margin is defined as 34°F for plate materia l s and 56°F for we l d materials , or m argin equa l s shift (whichever is l ess), per Reg. Gui d e 1.99, Rev. 2. | Irradiated Plate Heat C4114-2 Lateral Expansion Plot (Hatch Unit 1 300° Capsule) (LT) 5-5 C h a r py T e st R e sults 5-6 Table 5-1 Effect of Irradiation (E>1.0 MeV) on the Notch Toughness Properties T30* 30 ft-lb (40.7 J) T 50* 50 ft-lb (67 .8 J) T 3 5 mil* 35 mil CVN Upper Shelf Energy Transition Temperature Transition Temperature (0.89 mm) Lateral (USE) Material Expansion Temperature Identity Unirrad Irradiated ilT30 Unirrad Irradiated ilT50 Unirrad Irradiated ilT35mil Unirrad Irradiated Change OF OF OF OF OF OF OF OF OF ft-lb ft-lb ft-lb (oC) (oC) (oC) (oC) (oC) (oC) (oC) (oC) (oC) (J) (J) (J) C4114-2 -61.5 15.4 76.9 -28.8 44.8 73.6 -34.7 41.3 76.0 136.0 147.4 11.4 (LT orientation) (-51.9) (-9.2) (42.7) (-33.8) (7.1) (40.9) (-37.1) (5.2) (42.2) (184.4) (199.8) (15.4) Table 5-2 Comparison of Actual Versus Predicted Embrittlement Fluence RG 1.99 Rev. 2 RG 1.99 Rev. 2 Measured Shift 2 Predicted Ident i ty Material (E>1.0 MeV , x10 17 O F (OC) Predicted Shift 3 Shift+Margin 3*4 n/cm 2)1 OF (OC) OF (OC) C411 4-2 Hat c h U ni t 1 su rv e i lla n ce plate 1 3.8 76.9 (42.7) 4 0.9 (22.7) 7 4.9(41.6) (LT or i entat i on) l. Fluence value i s reported in Table 3-9. 2. The measured shift is taken from Tab l e 5-1. 3. Predicted shift= CF x FF, where CF is a Chemistry Factor taken from the base metal table in USNRC RG 1.99 , Rev. 2 [6], based on each material's Cu/N i co n te n t, and FF is F lu ence Factor, f0.28-0. l 0 log f, where f = tluence i n un i ts of 10 1 9 n/cm 2 (E > 1.0 MeV) specified. 4. Margin= 2./(G;2 + G t.2), where G; =the standard deviation on initial RT N D T (G; is taken to be 0°F), and G t. is the standard deviation on LlRT NDT (28°F for welds and l 7°F for base materials, except that G t. need not exceed 0.50 times the mean value of ilRT N o T). Thus, margin is defined as 34°F for plate materia l s and 56°F for we l d materials , or m argin equa l s shift (whichever is l ess), per Reg. Gui d e 1.99, Rev. 2. | ||
Charpy Test R e sults Table 5-3 Percent Decrease In Upper Shelf Energy Fluence Measured Decrease in USE Predicted Decrease in USE 2 Identity Material (E>1.0 MeV , (%) (%) x10 1 7 n/cm 2)1 C 4 114-2 Hatch Un i t 1 3 13.8 --1 3.1 (LT orie n tation) surveillance plate 1. Fluence value is reported in Tab l e 3-9. 2. Based on the equations for Figure 2 of Reg. Guide 1.99, Rev. 2 [6] as provided in Reg. Guide 1.162 [26). 3. Value less than zero. 5-7 6 REFERENCES | Charpy Test R e sults Table 5-3 Percent Decrease In Upper Shelf Energy Fluence Measured Decrease in USE Predicted Decrease in USE 2 Identity Material (E>1.0 MeV , (%) (%) x10 1 7 n/cm 2)1 C 4 114-2 Hatch Un i t 1 3 13.8 --1 3.1 (LT orie n tation) surveillance plate 1. Fluence value is reported in Tab l e 3-9. 2. Based on the equations for Figure 2 of Reg. Guide 1.99, Rev. 2 [6] as provided in Reg. Guide 1.162 [26). 3. Value less than zero. 5-7 6 REFERENCES | ||
: 1. 10 CFR 50 , Ap p endices G (Fracture Toughness Requir e ment s) and H (Reactor Vessel Material Surveillance Program R equire ments), Federal Register , Volume 60 , No. 243 , dated December 19, 1995. 2. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code), Section XI, " Ru l es for In service Inspection of Nuclear Power Plant Components," Nonmandatory Appendix G, Fracture Toughness Criteria for Protection Against Failure. 3. ASTM E185-82, Standard Practice for Conducting Surveillance T ests for Light-Wat er Cooled Nuclear Pow er R eactor Vessels, E706 (IF), American Society for Testing and Materia l s, Philadelphia , PA , 1982. 4. BWRVIP-86 , Revision 1-A: BWR Vessel and Int erna ls Project , Updated BWR Integrated Surv e illanc e Program (ISP) Impl e m enta tion Plan. EPRI, Palo Alto, CA: 2012. 1025144. 5. 10 CFR 50, Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," August 28, 2007. 6. U.S. NRC Regulatory Guide 1.99 , " Radiation Embrittlement of Reactor Vessel Materials ," Revision 2, May 1988. 7. "Guideline for the Management of Materials Issues," NEI 03-08 , Nuclear Energy Institute, Washington, DC, Latest Edition. 8. "Edwin I. Hatch Nuclear Power Plant , Unit 1 , Reactor Pressure Vessel Surveillance Materia l s Testing and Fracture Toughness Analysis" , T. A. Caine, GE Nuclear Ene rgy , NEDC-30997 DRF Bl 1-00313 Class I, October, 1985. 9. "Plant Hatch Unit 1 RPV Surveillance Materials Testing and Analysis" , B. D. Frew, GE Nuclear Energy , GE-NE-Bl 100691-0lRl, March , 1997. 10. "Determination of Fast Neutron Flux Density and Neutron Fluence , Hatch 1 Power Station ," G.C. Martin, GE Nuclear Energy, GE Report 266-7801-02, January 26, 1978. 11. CV GRAPH, Hyperbolic Tangent Curve Fitting Program , Developed by A TI Consulting, Version 6.02, Apri l 2014. 12. ASTM Standard E23, Standard T est Methods for Notch Bar Impact Testing of Metallic Materials , ASTM International , West Conshohocken, PA, www.astm.org. 13. U.S. NRC Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001. 14. BWRVIP-126 , R ev ision 2: BWR Vessel Int e rnals Proje ct, RAMA Fluence Methodology Softwar e, Version 1.20. EPRI, Palo Alto , CA: 2010. 1020240. 6-1 R efe r e n ce s 15. L etter from William H. Bateman (U.S. NRC) to Bill Eaton (BWRVIP), Safety Evaluation of Proprietary EPRl Reports BWRVIP-114, -115, -117, and -121 and TWE-PSE-001-R-001 , dated May 13 , 2005. 16. " Safety Evaluation Report with Open Items Related to the License Renewal of Seabrook Station," Docket Number 50-443 , dated June 2012. 17. Letter from Matthew A. Mitchell (U.S. NRC) to Rick Libra (BWRVIP), " Safety Evaluation of Proprietary E PRl Report BWR Vessel and Int e rnals Project, E v aluation of Susquehanna Unit 2 Top Guide and Core Shroud Material Samples Using RAMA Fluence Methodology (BWRVIP-145 | : 1. 10 CFR 50 , Ap p endices G (Fracture Toughness Requir e ment s) and H (Reactor Vessel Material Surveillance Program R equire ments), Federal Register , Volume 60 , No. 243 , dated December 19, 1995. 2. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code), Section XI, " Ru l es for In service Inspection of Nuclear Power Plant Components," Nonmandatory Appendix G, Fracture Toughness Criteria for Protection Against Failure. 3. ASTM E185-82, Standard Practice for Conducting Surveillance T ests for Light-Wat er Cooled Nuclear Pow er R eactor Vessels, E706 (IF), American Society for Testing and Materia l s, Philadelphia , PA , 1982. 4. BWRVIP-86 , Revision 1-A: BWR Vessel and Int erna ls Project , Updated BWR Integrated Surv e illanc e Program (ISP) Impl e m enta tion Plan. EPRI, Palo Alto, CA: 2012. 1025144. 5. 10 CFR 50, Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," August 28, 2007. 6. U.S. NRC Regulatory Guide 1.99 , " Radiation Embrittlement of Reactor Vessel Materials ," Revision 2, May 1988. 7. "Guideline for the Management of Materials Issues," NEI 03-08 , Nuclear Energy Institute, Washington, DC, Latest Edition. 8. "Edwin I. Hatch Nuclear Power Plant , Unit 1 , Reactor Pressure Vessel Surveillance Materia l s Testing and Fracture Toughness Analysis" , T. A. Caine, GE Nuclear Ene rgy , NEDC-30997 DRF Bl 1-00313 Class I, October, 1985. 9. "Plant Hatch Unit 1 RPV Surveillance Materials Testing and Analysis" , B. D. Frew, GE Nuclear Energy , GE-NE-Bl 100691-0lRl, March , 1997. 10. "Determination of Fast Neutron Flux Density and Neutron Fluence , Hatch 1 Power Station ," G.C. Martin, GE Nuclear Energy, GE Report 266-7801-02, January 26, 1978. 11. CV GRAPH, Hyperbolic Tangent Curve Fitting Program , Developed by A TI Consulting, Version 6.02, Apri l 2014. 12. ASTM Standard E23, Standard T est Methods for Notch Bar Impact Testing of Metallic Materials , ASTM International , West Conshohocken, PA, www.astm.org. 13. U.S. NRC Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001. 14. BWRVIP-126 , R ev ision 2: BWR Vessel Int e rnals Proje ct, RAMA Fluence Methodology Softwar e, Version 1.20. EPRI, Palo Alto , CA: 2010. 1020240. 6-1 R efe r e n ce s 15. L etter from William H. Bateman (U.S. NRC) to Bill Eaton (BWRVIP), Safety Evaluation of Proprietary EPRl Reports BWRVIP-114, -115, -117, and -121 and TWE-PSE-001-R-001 , dated May 13 , 2005. 16. " Safety Evaluation Report with Open Items Related to the License Renewal of Seabrook Station," Docket Number 50-443 , dated June 2012. 17. Letter from Matthew A. Mitchell (U.S. NRC) to Rick Libra (BWRVIP), " Safety Evaluation of Proprietary E PRl Report BWR Vessel and Int e rnals Project, E v aluation of Susquehanna Unit 2 Top Guide and Core Shroud Material Samples Using RAMA Fluence Methodology (BWRVIP-145 | ||
)," dated February 7 , 2008. 18. " Core Design and Operating Data for Cycle 1 o f Hatch l ," EPRl, E PRl-NP-562, January 1979. 19. " Core Design and Operating Data for Cycl es 2 and 3 of Hatch 1 ," E PRl, E PRl-NP-562 , February 1984. 20. " BUGLE-96: | )," dated February 7 , 2008. 18. " Core Design and Operating Data for Cycle 1 o f Hatch l ," EPRl, E PRl-NP-562, January 1979. 19. " Core Design and Operating Data for Cycl es 2 and 3 of Hatch 1 ," E PRl, E PRl-NP-562 , February 1984. 20. " BUGLE-96: | ||
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A Fine-Group Cross Section Library Based on E NDF/B-VI Release 3 for Radiation Transport Applications | A Fine-Group Cross Section Library Based on E NDF/B-VI Release 3 for Radiation Transport Applications | ||
," RSICC Data Library Collection , DLC-184, December 1996. 22. BWRVIP-114-A: BWR Ve ssel and Int e rn a ls Proj ec t , RAMA Flu e n ce M e th o dolo gy Th e o ry Manual , EPRl , Palo Alto , CA: 2009. 1019049. 23. BWR VI P-121-A: BWR Ve s se l and Int e rnals Proj ec t , RAMA Flu e n ce M e thodolo gy Proc e dur es M a nual , EPRl , Palo Alto , CA: 2009. 1019052. 24. BWRVIP-115-A: BWR V e ss e l and Int e rn a l s Proj ec t , RAMA Flu e n ce M e th o dolo gy B e nchmark M a nual -E v aluation of R eg ulato ry Guid e 1.190 B e n c hmark Probl e ms , EPRl , Palo Alto , CA: 2009. 1019050. 25. BWRVIP-135 , R e vision 3: BWR V es s e l and Int e rnals Project , In teg rat e d Surv e illanc e Program (ISP) Data Sour ce Book and Plant Evaluation | ," RSICC Data Library Collection , DLC-184, December 1996. 22. BWRVIP-114-A: BWR Ve ssel and Int e rn a ls Proj ec t , RAMA Flu e n ce M e th o dolo gy Th e o ry Manual , EPRl , Palo Alto , CA: 2009. 1019049. 23. BWR VI P-121-A: BWR Ve s se l and Int e rnals Proj ec t , RAMA Flu e n ce M e thodolo gy Proc e dur es M a nual , EPRl , Palo Alto , CA: 2009. 1019052. 24. BWRVIP-115-A: BWR V e ss e l and Int e rn a l s Proj ec t , RAMA Flu e n ce M e th o dolo gy B e nchmark M a nual -E v aluation of R eg ulato ry Guid e 1.190 B e n c hmark Probl e ms , EPRl , Palo Alto , CA: 2009. 1019050. 25. BWRVIP-135 , R e vision 3: BWR V es s e l and Int e rnals Project , In teg rat e d Surv e illanc e Program (ISP) Data Sour ce Book and Plant Evaluation | ||
: s. EPRl , Palo Alto , CA: 2014. 3002003144. | : s. EPRl , Palo Alto , CA: 2014. 3002003144. | ||
: 26. U.S. NRC Re g ulatory Guide 1.162 , " Format and Content of Report for Thermal Annealin g of Reactor Pre ss ure Ve ss els ," February 1996. 6-2 A DOSIMETER ANA L YSIS A.1 Dos i meter Material D e scription The Hatch Unit 1 300° surveillance capsule dosimeter materia l s are pure meta l wires which were located within the surveillance capsu l e along t h e ends of the Charpy specimens. | : 26. U.S. NRC Re g ulatory Guide 1.162 , " Format and Content of Report for Thermal Annealin g of Reactor Pre ss ure Ve ss els ," February 1996. 6-2 A DOSIMETER ANA L YSIS A.1 Dos i meter Material D e scription The Hatch Unit 1 300° surveillance capsule dosimeter materia l s are pure meta l wires which were located within the surveillance capsu l e along t h e ends of the Charpy specimens. | ||
The wire types prov i ded for t h e Ha t c h Unit 1 surveillance program are iron , n i ckel, and copper. Each wire is nominally three inches (7.62 cm) long. Further discussion of the dosimeter cleanin g and mass measurements follows. A.2 Dosimeter Cleaning and Mass Measurem en t At the time the surveillance capsule Charpy packets were opened, the dosimeter wires were clean e d in an ultrasonic cleaner in an acetone bath and were wiped with acetone wetted wipes to remo v e loose contamination. Upon receipt at the radiometric laboratory, the wires were visually inspected under a low magnification optical microscope. | The wire types prov i ded for t h e Ha t c h Unit 1 surveillance program are iron , n i ckel, and copper. Each wire is nominally three inches (7.62 cm) long. Further discussion of the dosimeter cleanin g and mass measurements follows. A.2 Dosimeter Cleaning and Mass Measurem en t At the time the surveillance capsule Charpy packets were opened, the dosimeter wires were clean e d in an ultrasonic cleaner in an acetone bath and were wiped with acetone wetted wipes to remo v e loose contamination. Upon receipt at the radiometric laboratory, the wires were visually inspected under a low magnification optical microscope. | ||
| Line 436: | Line 436: | ||
Finally , Table A-5 presents the primary results of interest for flux and fluence determination. | Finally , Table A-5 presents the primary results of interest for flux and fluence determination. | ||
The specific activity units are in dps/mg, which normalizes the activity to dosimeter mass. The activities are specified for a useful reference date/time , which in this case is the Hatch Unit 1 plant shutdown date and time. This reference date/time was specified as February 8 , 2016, at 012:01:01 AM eastern standard time. A-2 Dosimeter A nal ysis G4Fe G4Fe Figure A-1 Hatch Unit 1 300° Capsule Packet G4 Fe Dosimeter Wire G4 Fe: Prior to Cleaning (left); and After Cleaning/Coiling (right) G4Cu G4Cu Fi g ure A-2 Ha t ch Unit 1 300° Capsule Packet G4 Cu Dosimeter Wire G4 Cu: Prior to Cleaning (left); and After Cleaning/Coiling (right) G4Ni G4Ni Figure A-3 Hatch Unit 1 300° Capsule Packet G4 Ni Dosimeter Wire G4 Ni: Prior to Cleaning (left); and After Cleaning/Coiling (right) A-3 Dosim eter Analysis G5Fe G5Fe Figure A-4 Hatch Unit 1 300° Capsule Packet GS Fe Dosimeter Wire GS Fe: Prior to Cleaning (left); and After Cleaning/Coiling (right) G5Cu G5Cu Figure A-S Hatch Unit 1 300° Capsule Packet GS Cu Dosimeter Wire GS Cu: Prior to Cleaning (left); and After Cleaning/Coiling (right) G5Ni G5Ni Figure A-6 Hatch Unit 1 300° Capsule Packet GS Ni Dosimeter Wire GS Ni: Prior to Cleaning (left); and After Cleaning/Coiling (right) A-4 I I I Dosi m eter A n a l ysis Table A-1 Hatch Unit 1 300° Capsule Charpy Packet Dosimeter Wire Masses Wire Dosi m eter ID Mass (mg) G4 Fe 149.17 G4 Cu 405.37 G4Ni 275.83 G5 Fe 162.60 G5 Cu 396.50 G5 Ni 298.71 Table A-2 G a mma Ray Spectrometer System (GRSS) Specifications System Component Description and/or Specifications Detector Canberra Model BE3830 Energy Resolution | The specific activity units are in dps/mg, which normalizes the activity to dosimeter mass. The activities are specified for a useful reference date/time , which in this case is the Hatch Unit 1 plant shutdown date and time. This reference date/time was specified as February 8 , 2016, at 012:01:01 AM eastern standard time. A-2 Dosimeter A nal ysis G4Fe G4Fe Figure A-1 Hatch Unit 1 300° Capsule Packet G4 Fe Dosimeter Wire G4 Fe: Prior to Cleaning (left); and After Cleaning/Coiling (right) G4Cu G4Cu Fi g ure A-2 Ha t ch Unit 1 300° Capsule Packet G4 Cu Dosimeter Wire G4 Cu: Prior to Cleaning (left); and After Cleaning/Coiling (right) G4Ni G4Ni Figure A-3 Hatch Unit 1 300° Capsule Packet G4 Ni Dosimeter Wire G4 Ni: Prior to Cleaning (left); and After Cleaning/Coiling (right) A-3 Dosim eter Analysis G5Fe G5Fe Figure A-4 Hatch Unit 1 300° Capsule Packet GS Fe Dosimeter Wire GS Fe: Prior to Cleaning (left); and After Cleaning/Coiling (right) G5Cu G5Cu Figure A-S Hatch Unit 1 300° Capsule Packet GS Cu Dosimeter Wire GS Cu: Prior to Cleaning (left); and After Cleaning/Coiling (right) G5Ni G5Ni Figure A-6 Hatch Unit 1 300° Capsule Packet GS Ni Dosimeter Wire GS Ni: Prior to Cleaning (left); and After Cleaning/Coiling (right) A-4 I I I Dosi m eter A n a l ysis Table A-1 Hatch Unit 1 300° Capsule Charpy Packet Dosimeter Wire Masses Wire Dosi m eter ID Mass (mg) G4 Fe 149.17 G4 Cu 405.37 G4Ni 275.83 G5 Fe 162.60 G5 Cu 396.50 G5 Ni 298.71 Table A-2 G a mma Ray Spectrometer System (GRSS) Specifications System Component Description and/or Specifications Detector Canberra Model BE3830 Energy Resolution | ||
<1.9 keV FWHM @ 1.33 MeV Detector Efficiency Relative to a 3 inch x 3 inch 33.3% at 1.3 MeV Nal Crystal Amplifier/Multichannel Analyzer Canberra DAS-1000 Computer System Intel i5-4460 CPU at 3.20 GHz , 16 GB Main Memory , 931 GB Hard Disk , 23-inch Monitor , HP LaserJet Printer S oftware Canberra Apex v 1.4 A-5 Dosimet e r Analysis Table A-3 Counting Schedule for Hatch Unit 1 300° Capsule Dosimeter Materials Dosimeter ID Count Start Date Count Start Time (EST) Count Duration (Live Time Seconds) G4 Fe 11/4/2016 2:40:44 PM 86,400 G4Cu 1 1/7/2016 6: S4: 00 AM 86,400 G4Ni 11/8/2016 8: 24: 1S AM 86,400 GS Fe 1 1/9/2016 9:4 0: S4 AM 86,400 GS Cu 11/10/2016 3: 21: 1 3PM 86 , 400 GS Ni 1 1/11/2016 S:3 7: 22 PM 86,400 Table A-4 Neutron-Induced Reactions of Interest Dosimeter Material Neutron-Induced Reaction Reaction Product Radionuclide I ron Fe 54 (n , p)Mn 54 Mn 54 Copper Cu 63 (n , a)Co 60 Co 6o Ni ckel Ni sa(n , p)Co sa co sa A-6 Dosi m e t er Ana l ysis Table A-5 Results of Hatch Unit 1 300° Capsule Radiometric Analysis Activity at Specific Activity at Activity Dosimeter ID Isotope ID Reference Reference Uncertainty Date/Timea | <1.9 keV FWHM @ 1.33 MeV Detector Efficiency Relative to a 3 inch x 3 inch 33.3% at 1.3 MeV Nal Crystal Amplifier/Multichannel Analyzer Canberra DAS-1000 Computer System Intel i5-4460 CPU at 3.20 GHz , 16 GB Main Memory , 931 GB Hard Disk , 23-inch Monitor , HP LaserJet Printer S oftware Canberra Apex v 1.4 A-5 Dosimet e r Analysis Table A-3 Counting Schedule for Hatch Unit 1 300° Capsule Dosimeter Materials Dosimeter ID Count Start Date Count Start Time (EST) Count Duration (Live Time Seconds) G4 Fe 11/4/2016 2:40:44 PM 86,400 G4Cu 1 1/7/2016 6: S4: 00 AM 86,400 G4Ni 11/8/2016 8: 24: 1S AM 86,400 GS Fe 1 1/9/2016 9:4 0: S4 AM 86,400 GS Cu 11/10/2016 3: 21: 1 3PM 86 , 400 GS Ni 1 1/11/2016 S:3 7: 22 PM 86,400 Table A-4 Neutron-Induced Reactions of Interest Dosimeter Material Neutron-Induced Reaction Reaction Product Radionuclide I ron Fe 54 (n , p)Mn 54 Mn 54 Copper Cu 63 (n , a)Co 60 Co 6o Ni ckel Ni sa(n , p)Co sa co sa A-6 Dosi m e t er Ana l ysis Table A-5 Results of Hatch Unit 1 300° Capsule Radiometric Analysis Activity at Specific Activity at Activity Dosimeter ID Isotope ID Reference Reference Uncertainty Date/Timea | ||
(µCi) Date/Time 1 (dps/mg) (%) G4 Fe 54 Mn 4.S9E-01 113.8S 2.20 G4Cu 6o co 2.2SE-01 20.S4 1.70 G4 Ni sa co 1.04E+01 139S.06 2.27 GS Fe 54 M n 4.99E-01 113.SS 2.20 GS Cu 6o co 2.22E-01 20.72 1.70 GS N i sa co 1.21 E+01 1498.78 2.27 1 Feb ru ary 8, 2016 at 12: 01 : 01 AM EST is the refe r ence date and time. A-7 Export Control Restrictions Access to and use of EPRI Int ellectua l Property is granted wi th the sci f i c understanding and requirement tha t re s ponsibility For ensuring full co mpl iance with all applicable U.S. and foreign export laws and lations is bei ng undertaken by you and your company. Thi s includes an ob lig ation to ensure that any individual receiving access hereunder who is not a U.S. citize n or permanent U.S. resident is per mitt ed ac c ess under applicable U.S. and foreign export laws and regulations. In the e v ent you are uncertain whether yo u or your company may la wfu ll y obtain a c cess to this EPRI Int el l ectua l Property , you acknowledge that it is your obligation to consult with your company's legal cou n sel to determine whether thi s access is la wful. Although EPRI may make available o n a ca s e-by-case bas is an informal assessment of the applicable U.S. export classif i cat i on for specific EPRI Int e lle c tual Property , you and your co mpan y acknowledge that this assessment is solely for informational purposes and not for reliance purposes. | (µCi) Date/Time 1 (dps/mg) (%) G4 Fe 54 Mn 4.S9E-01 113.8S 2.20 G4Cu 6o co 2.2SE-01 20.S4 1.70 G4 Ni sa co 1.04E+01 139S.06 2.27 GS Fe 54 M n 4.99E-01 113.SS 2.20 GS Cu 6o co 2.22E-01 20.72 1.70 GS N i sa co 1.21 E+01 1498.78 2.27 1 Feb ru ary 8, 2016 at 12: 01 : 01 AM EST is the refe r ence date and time. A-7 Export Control Restrictions Access to and use of EPRI Int ellectua l Property is granted wi th the sci f i c understanding and requirement tha t re s ponsibility For ensuring full co mpl iance with all applicable U.S. and foreign export laws and lations is bei ng undertaken by you and your company. Thi s includes an ob lig ation to ensure that any individual receiving access hereunder who is not a U.S. citize n or permanent U.S. resident is per mitt ed ac c ess under applicable U.S. and foreign export laws and regulations. In the e v ent you are uncertain whether yo u or your company may la wfu ll y obtain a c cess to this EPRI Int el l ectua l Property , you acknowledge that it is your obligation to consult with your company's legal cou n sel to determine whether thi s access is la wful. Although EPRI may make available o n a ca s e-by-case bas is an informal assessment of the applicable U.S. export classif i cat i on for specific EPRI Int e lle c tual Property , you and your co mpan y acknowledge that this assessment is solely for informational purposes and not for reliance purposes. | ||
You and your company knowledge that it i s s till the ob li ga t ion of you and your compa n y to make you r own assessment of the applicable U.S. export classif i cat i on and ensure comp lian c e accordi n gly. You and your compan y understand and acknowledge your ob ligati ons to make a prompt report to EPRI and the appropriate authorities r egarding any a cc e ss to or use of EPRI Intellectual Property hereunder that may be in vio lat ion of applicable U.S. o r foreign export law s or regulations. The Electric Power Research Institute, Inc. (EPRI , www.epri.com) conducts r e s ear c h and development relating to the generation , delivery and use o f electricity for the benefit of th e public. An ind epe ndent , nonprofit organ i za ti o n , EPRI brings together its s cientis t s and engineers as wel l as e x perts from academia and indu stry to help address challenges in elect ri c ity , including reliability , effi c iency , affordability, h ealth , safety and the e n vironmen t. EPRI me mb ers represent 90% o f the electric utility re v enue in the United States with international participation in 35 cou ntr ie s. EPRl's principal offices a nd laboratories are located in Palo Alto , Calif.; C harl otte , N.C.; Kno x vi ll e , Tenn.; and Lenox , Mass. Togethe r ... Shaping the Future of Electricity Program: BWR Ve s s el and Internal s Project © 20 1 7 El ect r ic P ow e r R ese ar c h I n s t i tut e (E P R I), In c. A ll right s re s er ve d. El ec tri c P ower Rese arch I n s t i tute , EPRI , and T OG ETHE R ... S H A P I N G T HE FUTURE O F E L E C TRICI TY are reg i s t e r e d se r v i ce mark s a l the E l e c tri c P ower R es ear ch I ns ti tu t e, In c. 3002010553 Electric Power Research Institute 3420 Hillview Avenue , Palo Alto, Ca l ifo rnia 94304-1338 | You and your company knowledge that it i s s till the ob li ga t ion of you and your compa n y to make you r own assessment of the applicable U.S. export classif i cat i on and ensure comp lian c e accordi n gly. You and your compan y understand and acknowledge your ob ligati ons to make a prompt report to EPRI and the appropriate authorities r egarding any a cc e ss to or use of EPRI Intellectual Property hereunder that may be in vio lat ion of applicable U.S. o r foreign export law s or regulations. The Electric Power Research Institute, Inc. (EPRI , www.epri.com) conducts r e s ear c h and development relating to the generation , delivery and use o f electricity for the benefit of th e public. An ind epe ndent , nonprofit organ i za ti o n , EPRI brings together its s cientis t s and engineers as wel l as e x perts from academia and indu stry to help address challenges in elect ri c ity , including reliability , effi c iency , affordability, h ealth , safety and the e n vironmen t. EPRI me mb ers represent 90% o f the electric utility re v enue in the United States with international participation in 35 cou ntr ie s. EPRl's principal offices a nd laboratories are located in Palo Alto , Calif.; C harl otte , N.C.; Kno x vi ll e , Tenn.; and Lenox , Mass. Togethe r ... Shaping the Future of Electricity Program: BWR Ve s s el and Internal s Project © 20 1 7 El ect r ic P ow e r R ese ar c h I n s t i tut e (E P R I), In c. A ll right s re s er ve d. El ec tri c P ower Rese arch I n s t i tute , EPRI , and T OG ETHE R ... S H A P I N G T HE FUTURE O F E L E C TRICI TY are reg i s t e r e d se r v i ce mark s a l the E l e c tri c P ower R es ear ch I ns ti tu t e, In c. 3002010553 Electric Power Research Institute 3420 Hillview Avenue , Palo Alto, Ca l ifo rnia 94304-1338 | ||
Revision as of 13:39, 26 April 2019
| ML17240A231 | |
| Person / Time | |
|---|---|
| Site: | Hatch, PROJ0704  |
| Issue date: | 08/24/2017 |
| From: | McGehee A O Electric Power Research Institute |
| To: | Orenak M D Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 2017-101, CAC MF7696 | |
| Download: ML17240A231 (84) | |
Text
ELECTRIC POWER RESEARCH INSTITUTE 2017-101 ________________
BWR Vessel & Internals Project (BWRVIP) August 24 , 2017 Document Control Desk U. S. Nuclear Regulatory Commission 11555 Rockville Pike Rockville, MD 20852 Attention:
Mr. Michael Orenak
Subject:
Project No. 704, BWRVIP-308NP:
BWR Vessel and Internals Project , Testing and Evaluation of the Hatch Unit 1 300° Surveillance Capsule
References:
- 1. M. Orenak to C. Pierce , " Edwin I. Hatch Nuclear Plant , Unit Nos. 1 and 2 , Issuance of Amendments Regarding the Relocation of the Pressure Temperature Limit Curves Following TSTF-419 (CAC NOS. MF6063 and MF6064)," March 23 , 2016. 2. M. Markley to A. McGehee and C. Pierce, "Edwin I. Hatch Nuclear Plant , Unit No. 1 -Request for Extension of Date to Submit Reactor Vessel Surveillance Capsule Summary Technical Report (CAC NO. MF7696)," December 5 , 2016. E nclosed are five (5) paper copies of the report " BWRVIP-308NP:
BWR Vessel and Internals P roject, Testing and Evaluation of the Hatch Unit 1 300° Surveillance Capsule ," EPRI Technical Report 3002010553 , August 2017. This report is being transmitted to the NRC for information only. T his report describes testing and evaluation of the Hatch Unit 1 300° surveillance capsule. These results will be used to monitor embrittlement as part of the BWRVIP Integrated Surveillance Program (ISP). The materials in the Hatch Unit 1 300° surveillance capsule are identified in BWRVIP-86 , Revision 1-A , as representative for target materials for the Hatch Unit 1 and Enrico Fermi Unit 2 reactor vessels. *
- The plate representative material in the 300° surveillance capsule is a heat specific match to material in the Hatch Unit 1 reactor vessel and therefore, direct use is made of the surveillance data. Hatch Unit l's Pressure-Temperature (P-T) limit curves are contained in an approved Pressure-Temperature Limits Report (PTLR) (Reference 1 ). The test results from the 300° surveillance capsule have been evaluated and it has been determined that the existing Hatch Unit 1 P-T limit curves are conservative.
The plate representative material in the 300° surveillance capsule is not a heat specific match to material in the Enrico Fermi Unit 2 reactor vessel and therefore , direct use is not Together ... Shaping the Future of Electricity (5-1JtJif-
µ (( ( PALO ALTO OFFICE 3.420 Hillview Avenue, Pala Alto, CA 9.430.4*1395 USA* 650.855.2000 *Customer Service 800.313.377.4
- www.epr i.com made of the surveillance data. There is no impact of the surveillance data on the Enrico Fermi Unit 2 P-T limit curves. It should be noted that the BWRVIP previously requested an extension to the 10 CFR 50 Appendix H surveillance capsule testing reporting requirements and committed to providing the results of the Hatch Unit 1 300° surveillance capsule b y August 31 , 2017. That request was approved in Reference
- 2. Please also note that the enclosed report is non-proprietary and is available to the public by request to EPRI. If you have any questions on this subject p l ease call Steve Richter (Energy Northwest , BWRVIP Assessment Committee Chairman) at 509-377-4703.
Sincerely , Andrew McGehee , EPRI , BWRVIP Chairman Tim Hanley , Exe l on Corporation , BWRVIP Chairman c: S. Ruffin , NRC-NRR M. Kirk , NRC-RES D. Odell , Exelon Corp. R. Carter , EPRI A. McGehee , EPRI C. Wirtz , EPRI ELECTRIC POWER RESEARCH INSTITUTE 2017 TECHNICAL REPORT BWRVIP-308NP:
BWR Vessel and Internals Project Testing and Evaluation of the Hatch Unit 1 300° Surveillance Capsule PRE P ARED UNDER THE NUCLEAR
- PR O GRAM BWRVIP-308NP:
BWR Vessel and Internals Project Testing and Evaluation of the Hatch Unit 1 300° Surveillance Capsule 3002010553 Final Report, August 2017 EPRI Project Manager N. Palm Work to deve l op this product was completed under the EPRI Nuclear Quality Assurance Program in compliance with 10 CFR 50 , Append ix B and 10 CFR 21 , 'NO ELECTRIC POWER RESEARC H IN STITUTE 3420 H illview Avenue , Palo Alto , Californ ia 94304-1338
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& Testing , LLC TransWare Enterprises Inc. THE TECHNICAL CONTENTS OF THIS DOCUMENT WERE PREPARED IN ACCORDANCE WITH THE EPRI QUALITY PROGRAM MANUAL THAT FULFILLS THE REQUIREMENTS OF 10 CFR 50 APPENDIX B , 1 0 CFR 21 , ANSI N45.2-1977 AND/ OR THE INTENT OF IS0-9001 (1994 ). CONTRACTUAL ARRANGEMENTS BETWEEN THE CUSTOMER AND E PRI MUST BE ESTABLISHED BEFORE QUALITY APPLICATION TO ASSURE FU L FILLMENT OF QUALITY PROGRAM REQUIREMENTS. NOTE For further i nformation about EPRI , call the EPRI C u stomer Assistance Center at 800.3 1 3.3774 or e-mail askep r i@epri.com. Electric Power Research Institute , EPRI , and TOGETHER ... SHAPING THE FUTURE OF ELECTRICITY are r egistered service ma r ks of the E l ectric Power Research Institute , Inc. Copyright© 20 1 7 Elect ri c Power Research Institute , Inc. All rights r eserved.
ACKNOW L EDGMENTS The following organizations prepared this report: E l ectr i c Power R esearch Institute (EPRI) 3420 H illview Ave. Palo A lto, CA 94304 Principal Investigators N. Pa l m E.Long MP M achinery & Testing , LLC 2161 S andy Drive State C ollege, PA 16803 Principal Investigator M. M a nahan, Sr. Trans Ware Enterprises Inc. 1565 Mediterranean Drive Sycamore , IL 60178 Principal Investigator J. Gla d den This r e port describes research sponsored by EPRI. This publication is a corporate document that should be cited in the literature in the following mann e r: BWRVIP-308NP
- BWR Vessel and Internals Proj ec t , T esting and Evaluation of th e Hatch Unit 1 300° Surv e illance Capsul e. EPRI, Palo Alto, CA: 2017. 3002010553. 111 PRODUCT DESCRIPTION In the late 1990s, a Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integrated Surveillance Program (ISP) was developed to improve the surveillance of the U.S. BWR fleet. This report describes testing and evaluation of the Edwin I. Hatch Nuclear Power Plant (HNPP) Unit 1 300° surveillance capsule. These results will be used to monitor embrittlement as part of the BWRVIP ISP. Background The BWRVIP ISP represents a major enhancement to the process of monitoring embrittlement for the U.S. fleet of BWRs. The ISP optimizes surveillance capsule tests while at the same time maximizing the quantity and quality of data; thus , resulting in a more cost-effective program. Neutr o n irradiation exposure reduces the toughness ofreactor vessel steel plates , welds, and forgings.
The BWRVIP ISP provides more representative data that can be used to assess embrittlement in reactor pressure vessel (RPV) beltline materials, and improve trend curves in the BWR range of irradiation conditions.
Obje c tive s
- To document the results of neutron dosimetry and Charpy V-notch toughness tests for the surveillance material (plate heat C4114-2) in the Hatch Unit 1 300° surveillance capsule.
- To compare the results with the embrittlement trend prediction of the U.S. Nuclear Regulatory Commission (U.S. NRC) Regulatory Guide 1.99 , Revision 2. Appr o ach The Hatch Unit 1 300° surveillance capsule had been irradiated in the reactor since plant startup. The surveillance capsule contained flux wires for neutron flux monitoring, Charpy V-notch impact test specimens, and tensile specimens. The project team removed the capsule from the reactor in 2016 and transported it to facilities for testing and evaluation. The team used dosim e try to gather information about the neutron fluence accrual of specimens from the capsule. They then performed a neutron transport calcu l ation in accordance with Regulatory Guide 1.190 and c o mpared it to the dosimetry results. Testing of Charpy V-notch specimens was performed accor d ing to the American Society for Testing and Materials (ASTM) standards.
Resu l ts The report includes capsule neutron exposure and Charpy V-notch test results for Hatch Unit 1 survei ll ance plate heat C4114-2. The project compared irradiated Charpy data to unirradiated data i n order to determine the shifts in Charpy index temperatures for the surveillance p l ate material due to irradiation.
For the surveillance plate, the measured shift is greater than the predic t ed shift+ margin using Regulatory Guide 1.99, Revision 2. Researchers also measured v
- flux wires , determined a fluence value for the 300° surveillance capsule, and calculated revised fluence values for the previous l y-tested 30° and 120° surveillance capsules. Applications , Value, and Use Results of this work will be used in the BWRVIP ISP that integrates individual BWR surveillance programs into a single program. The ISP provides high quality data to monitor BWR vessel embrittlement.
The ISP results in significant cost savings to the BWR fleet, and provides more accurate monitoring of embrittlement in BWR vessels. Keywo r ds BWR Charpy V-notch testing Mechanical properties Radiation embrittlement Reactor pressure vessel integrity Reactor vessel surveillance program Vl EPl21 ELECTRIC POWER RESEARCH INSTITUTE EXECUTIVE
SUMMARY
Deliverable Number: 3002010553 Product Type: Technical Report BWRVIP-308NP:
BWR Vessel and Internals Project: Testing and Evaluation of the Hatch Unit 1 300° Surveillance Capsule PRIMARY AUDIENCE:
Plant engineers responsible for reactor vessel integrity SECONDARY AUDIENCE:
Boiling Water Reactor Vessel and Internals Project (BWRVIP) Progr a m Owners KEY RESEARCH QUESTION The objectives of this project were:
- To withdraw and test the Hatch Unit 1 300° surveillance capsule per the approved test matrix of the BWRVIP Integrated Surveillance Program (ISP) (BWRVIP-86 , Revision 1-A),
- To document the results of neutron dosimetry and Charpy V-notch toughness tests for the surveillance material (plate heat C4114-2) per American Society for Testing and Materials (ASTM) E 185-82 , and determine capsule fluence per U.S. Nuclear Regulatory Commission (U.S. NRC) Regulatory Guide 1.190 ,
- To compare the results w ith embrittlement trend predictions of the U.S. NRC Regulatory Guide 1.99 , Revision 2. RESEARCH OVERVIEW The BWRVIP ISP combines individual BWR surveillance programs into a single program that monitors the reduction in toughness of reactor vessel steel plates , welds, and forgings as a result of neutron i rradiation exposure.
The Hatch Unit 1 300° surveillance capsule was withdrawn and tested per the schedule in BWRVIP-86 , Revision 1-A. The capsule had been irradi a ted in the reactor since plant startup , and contained flux wires for neutron flux monitoring , Charpy V-notch impact test specimens , and tensile specimens. The project team removed the capsule from the reactor in 2016 and transported it to facilities for testing and evaluation. The team u sed dosimetry to gather information about the neutron fluence accrual of the capsule speci m ens. They then performed a neutron transport calculation in accordance with Regulatory Guide 1.190 and compared it to the dosimetry results. Testing of Charpy V-notch specimens was p e rformed according to ASTM standards.
KEY FINDINGS
- The report includes capsule neutron exposure and Charpy V-notch test results for Hatch Unit 1 surveillance plate heat C4114-2.
- The project compared irradiated Charpy data to unirradiated data in order to determine the shifts in Charpy index temperatures for the surveillance plate material due to irradiation.
V ll EF=-121 ELECTR I C POWER RESEARCH I NSTITUTE EXECUTIVE
SUMMARY
- For the surveillance plate , the measured shift is greater than the predicted shift plus margin using Regulatory Guide 1.99 , Revision 2.
- Researchers measured flux wires and performed a fluence calculation to determine the fluence for the 300° surveillance capsule , and updated fluence values for the previously tested 30° and 120° surveillance capsules.
- Although the Hatch Unit 1 surveillance weld was tested, it was not evaluated because unirradiated baseline data i s not available and it is not a representative material under the ISP. WHY THIS MATTERS Results of this work will be used in the BWRVIP ISP , which is utilized by U.S. BWR fleet owners to satisfy the requ i rements of 10 CFR 50 Appendix G and Appendix H. The ISP provides high quality data to monitor BWR vessel embrittlement.
The ISP results in significant cost savings for the BWR fleet , and provides more accurate monitoring of embrittlement in BWR vessels. Plants for which the Hatch Unit 1 surveillance materials are assigned as the representative surveillance materials under the ISP must consider these test results in development of vessel integrity evaluations and plant operating limit curves. HOW TO APPLY RESULTS Instructions for use of ISP data are provided in the following technical report: BWRVIP-135 , Revision 3: BWR Vessel and Internals Project , Integrated Surveillance Program (ISP) Data Source Book and Plant Evaluations.
EPRI , Palo Alto , CA: 2014. 3002003144. LEARNING AND ENGAGEMENT OPPORTUNITIES
- The program plan for the BWRVIP ISP is described in the following technical report: BWRVIP-86 , Revision 1-A: BWR Vessel and Internals Project , Updated BWR Integrated Surveillance Program (ISP) Implementation Plan. EPRI , Palo Alto , CA: 2012. 1025144.
- Data collected under the ISP and instructions for use of the data are contained in the following technical report: BWRVIP-135 , Revision 3: BWR Vessel and Internals Project , Integrated Surveillance Program {ISP) Data Source Book and Plant Evaluations , EPRI , Palo Alto , CA: 2014. 3002003144. EPRI CONTACTS:
Nathan A. Palm , Principal Technical Leader , npalm@epr i.com PROGRAM: BWRVIP IMPLEMENTATION CATEGORY:
Category 1 -Regulatory Together ... Shap i ng the Future of Electricit y Electric Power Research Institute 3420 Hillview Avenue , Palo Alto , California 94304-1338
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CONTENTS ABSTRACT ................
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V EXECUTIVE
SUMMARY
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..... VII 1 INTRODUCTION
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..... 1-1 1.1 Implementation Requirements
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................ 1-2 2 MATERIALS AND TEST SPECIMEN DESCRIPTION
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2-1 2.1 Dosimeters
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............. 2-1 2.2 Test Materials
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......................... 2-1 2.2.1 Capsule Load i ng Inventory
...................................................................................... 2-1 2.2.2 Mater i al Descr i ption .................................................................................................
2-6 2.2.3 Chem i cal Composition
............................................................................................. 2-7 2.2.4 CVN Baseline Propert i es ......................................................................................... 2-8 2.3 Capsule Opening ........................................................................................................... 2-14 3 NEUTRON FLUENCE CALCULATION
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......... 3-1 3.1 Description of the Reactor System .................................................................................. 3-2 3.1.1 Overview of the Reactor System Design ................................................................. 3-2 3.1.2 Reactor System Mechanical Design lnputs ............................................................. 3-2 3.1.3 Reactor System Mater i al Compositions
.................................................................. 3-3 3.1.4 Reactor Ope r ating Data lnputs ................................................................................ 3-5 3.2 Calculation Methodology
................................................................................................. 3-9 3.2.1 Desc r iption of the RAMA Fluence Methodology
...................................................... 3-9 3.2.2 RAMA Geometry Model for the Hatch Unit 1 Reactor ................
........................... 3-10 3.2.3 RAMA Calculation Parameters
.............................................................................. 3-18 3.2.4 RAMA Neutron Source Calculation
....................................................................... 3-19 3.2.5 RAMA Fission Spectra .......................................................................................... 3-19 3.2.6 Parametric Sensitivity Analyses ............................................
................................ 3-19 I X
3.3 Surveillance
Capsule Activation and Fluence Results .................................................. 3-19 3.3.1 Comparison of Predicted Activation to Plant-specific Measurements
................... 3-20 3.3.2 Capsule Peak Fluence Calculations and Lead Factor Determinations
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3-25 3.4 Capsule Fluence Uncertainty Analysis .......................................................................... 3-25 3.4.1 Comparison Uncertainty
........................................................................................ 3-25 3.4.2 Analytic Uncertainty
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............................... 3-26 3.4.3 Combined Uncertainty
........................................................................................... 3-26 4 CHARPY TEST DATA ...........................
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4-1 4.1 Charpy Test Procedure
.................................................................................................. .4-1 4.2 Charpy Test Data for the 300° Capsule .................
........................................................ .4-2 5 CHARPY TEST RES UL TS .....................................................................................................
5-1 5.1 Analysis of Impact Test Results ........................................................
.............................. 5-1 5.2 Irradiated Versus Unirradiated CVN Properties
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5-1 6 REFERENCES
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6-1 A DOSIMETER ANALYSIS .........................................
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A-1 x LIST OF FIGURES Figure 2-1 Drawing Showing the Charpy Test Specimen Geometry and ASTM E23 Permissible Variations
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2-3 Figure 2-2 Photograph of the 300° Hatch Unit 1 Capsule (top) and a Magnified View of t h e External Identification Mark i ngs (bottom) ..........................................................
........... 2-4 Figure 2-3 Photograph of the 300° Hatch Unit 1 Capsule ....................................
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2-S Figure 2-4 Photograph of the Inside of the 300° Hatch Unit 1 Capsule ..................................... 2-6 Figure 2-S Charpy Energy Plot for Plate Heat C4114-2 (LT) Unirradiated
.............................. 2-10 Figure 2-6 Lateral Expansion Plot for Plate Heat C4114-2 (LT) Unirradiated
.......................... 2-12 Figure 2-7 Drawing of the Hatch Unit 1 300° Surveillance Capsule Contents Showing the Test Specimen Identification Markings ............................................................................ 2-1 S Figure 2-8 Photograph of the Inside of the GS Charpy Packet within the 300° Hatch Unit 1 Capsule ......................................................................................................................... 2-16 Figure 2-9 Photograph of the Inside of the G4 Charpy Packet within the 300° Hatch Unit 1 Capsule ....................................
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.............................................................. 2-16 Figure 3-1 Planar View of Hatch Unit 1 at the Core Mid-plane Elevation
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...................... 3-3 Figure 3-2 Planar View of the Hatch Unit 1 RAMA Quadrant Model at the Core Mid-plane Elevation
.......................................................................................................................... 3-11 Figure 3-3 Axial View of the Hatch Unit 1 RAMA Model ..................................................
........ 3-12 F i gure 4-1 Illustration of Digital Optical Comparator Measurement of Shear Fracture Area ........................................................................
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................................................ 4-2 Figure S-1 Irradiated Plate Heat C4114-2 Charpy Energy Plot (Hatch Unit 1 300° Capsule) (LT) ..................................................................................................................... S-2 F i gure S-2 Irradiated Plate Heat C4114-2 Lateral Expansion Plot (Hatch Unit 1 300° Capsule) (LT) ..................................................................................................................... S-4 F i gure A-1 Hatch Unit 1 300° Capsule Packet G4 Fe Dosimeter Wire G4 Fe: Prior to Cleaning (left); and After Cleaning/Coiling (right) ........................
...................................... A-3 Figure A-2 Hatch Unit 1 300° Capsule Packet G4 Cu Dosimeter Wire G4 Cu: Prior to Cleaning (left); and After Cleaning/Co i ling (right) .............................................................. A-3 Figure A-3 Hatch Unit 1 300° Capsule Packet G4 Ni Dosimeter Wire G4 Ni: Prior to Cleaning (left); and After Cleaning/Coiling (right) ..............
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.................................... A-3 Figure A-4 Hatch Unit 1 300° Capsule Packet GS Fe Dosimeter Wire GS Fe: Prior to C l eaning (left); and After Cleaning/Coiling (right) ..................................................
............ A-4 F i gure A-S Hatch Unit 1 300° Capsule Packet GS Cu Dosimeter Wire GS Cu: Prior to C l eaning (left); and After Cleaning/Coiling (right) ..................
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.................... A-4 Figure A-6 Hatch Unit 1 300° Capsule Packet GS Ni Dosimeter Wire GS Ni: Prior to C l eaning (left); and After Cleaning/Coiling (right) ........................................
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........ A-4 X I LIST OF TABLES Table 2-1 Hatch Unit 1 300° Surveillance Capsule Specimen lnventory 1 .................................. 2-2 Table 2-2 Best Estimate Chemistry of Available Data Sets for Plate Heat C4114-2 .................
2-7 Table 2-3 Unirradiated Longitudinal Charpy V-Notch Impact Test Results for Surveillance Base Metal (Heat C4114-2) Specimens from the Hatch Unit 1 Surveillance Program .............................
............................................................................ 2-8 Table 2-4 Baseline CVN Properties
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.......................................... 2-9 Table 2-5 Charpy Specimen Dimensional Measurements Compared to the ASTM E23 Dimensional Requirements after Specimen Cleaning ....................
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.................. 2-17 Table 3-1 Summary of Material Compositions by Region for Hatch Unit 1 ....................
............ 3-4 Table 3-2 Summary of Hatch Unit 1 Core Loading Inventory
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............... 3-6 Table 3-3 State-point Data for Each Cycle of Hatch Unit 1 ....................................................... 3-7 Table 3-4 Comparison of Specific Activities for Hatch Unit 1 Cycle 1 30° Flux Wire Holder Wires (C/M) .............................................................
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................... 3-21 Table 3-5 Comparison of Specific Activities for Hatch Unit 1 Cycle 8 30° Surveillance Capsule Flux Wires (C/M) ................................................................................................ 3-22 Table 3-6 Comparison of Specific Activities for Hatch Unit 1 Cycle 16 120° Surveillance Capsule Flux Wires (C/M) ..........................
...................................................................... 3-23 Table 3-7 Comparison of Specific Activities for Hatch Unit 1 Cycle 27 300° Surveillance Capsule Flux Wires (C/M) ..............
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............................. 3-24 Table 3-8 Comparison of Activities for Hatch Unit 1 Flux Wires ......................
........................ 3-24 Table 3-9 Calculated Capsule Fast Neutron Fluence and Lead Factors for Hatch Unit 1 ....... 3-25 Table 3-10 Hatch Unit 1 Surveillance Capsule Combined Uncertainty for Energy >1.0 MeV ................
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3-27 Table 4-1 Irradiated Charpy V-Notch Impact Test Results for Surveillance Base Metal Specimens (Heat C4114-2) from the Hatch Unit 1 300° Surveillance Capsule ................. 4-4 Table 4-2 Irradiated Charpy V-Notch Impact Test Results for Surveillance Weld Metal Specimens (Heat 1 P3571) from the Hatch Unit 1 300° Surve i llance Capsule ................... 4-4 Table 4-3 Irradiated Charpy V-Notch Impact Test Results for Surveillance HAZ Metal Specimens from the Hatch Unit 1 300° Surveillance Capsule .......................................... .4-5 Table 5-1 Effect of Irradiation (E>1.0 MeV) on the Notch Toughness Properties
...................... 5-6 Table 5-2 Comparison of Actual Versus Predicted Embrittlement..
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............... 5-6 Table 5-3 Percent Decrease In Upper Shelf Energy ..........
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........ 5-7 Table A-1 Hatch Unit 1 300° Capsule Charpy Packet Dosimeter Wire Masses ....................... A-5 Table A-2 Gamma Ray Spectrometer System (GRSS) Specifications
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........................... A-5 Table A-3 Counting Schedule for Hatch Unit 1 300° Capsule Dosimeter Materials
................. A-6 X lll Table A-4 Neutron-Induced Reactions of Interest..
................................................................... A-6 Table A-5 Results of Hatch Unit 1 300° Capsule Radiometric Analysis ................................... A-7 XIV 1 INTRODUCTION Test coupons of reactor vessel ferritic beltline materials are irradiated in reactor surveillance capsules to facilitate evaluation of vessel fracture toughness in vessel integrity evaluations.
The key values that characterize fracture toughness are the reference temperature of nil-ductility transition (RTNDT) and the upper shelf energy (USE). These are defined in 10CFR50, Appendix G [1] and in Appendix G of the ASME Boiler and Pressure Vessel Code,Section XI [2]. Appe n dix Hof 10CFR50 [1] and ASTM E185-82 [3] establish the methods to be used for testing of surveillance capsule materials.
In the late 1990s, the BWR Vessel and Internals Project (BWRVIP) initiated the BWRVIP Integrated Surveillance Program (ISP) [4], and the BWRVIP assumed responsibility for testing and e v aluation of ISP capsules.
The surveillance plate from the Edwin I. Hatch Nuclear Plant , Unit 1 (hereinafter, Hatch Unit 1) was designated as an "ISP representative surveillance material" to be tested by the ISP according to an approved capsule withdrawal and test schedule.
The surveillance weld from Hatch Unit 1 is not used in the ISP and, due to a lack of unirradiated (baseline) data, cannot be evaluated. This report addresses the withdrawal and testing of the Hatch Unit 1 300° surveillance capsule. The c a psule contained flux wires for neutron flux monitoring, Charpy V-notch impact test specimens, and tensile specimens.
The capsule was irradiated for 27 cycles of operation before it was removed in February 2016 and shipped to MP Machinery
& Testing, LLC for opening and testing of the Charpy V-notch surveillance specimens.
Evaluation of the fluence environment was conducted by Trans Ware Enterprises, Inc. Final evaluation of the Charpy test data and irradiated material properties and compilation of this report were performed by EPRI. The Charpy V-notch surveillance materials were tested per ASTM E185-82, and the information and the as s ociated evaluations provided in this report have been performed in accordance with the requirements of 10CFR50, Appendix B [5]. This r e port compares the irradiated material properties of surveillance plate heat C4114-2 to its unirradiated (e.g., baseline) properties.
The observed embrittlement (as characterized by the shift in the Charpy energy curve 30 ft-lb (41J) index temperature or 6T 30) is compared to that predicted by U.S. Nuclear Regulatory Commission (U.S. NRC) Regulatory Guide 1.99, Revis i on 2 [6]. Other BWRVIP ISP reports will integrate the results from the 300° surveillance capsule with the results from the Hatch Unit 1 30° surveillance capsule (withdrawn in 1984) and 1 2 0° surveillance capsule (withdrawn in 1996) for a broader characterization of embrittlement behavior.
1-1 In tro du c tion 1.1 Implementation Requirements The results documented in this report will be utilized by the BWRVIP ISP and b y individual utilities to demonstrate compliance with 10CFR50 , Appendi x H , Reactor Vesse l Material Surveillance Program Requirements.
T h erefore, the implementation requirements of 10CFR50, Appendix H govern and the implementation requirements of Nuclear Energy Institute (NEI) 03-08, Guideline for t h e Management of Materials Issues [7], are not applicable. 1-2 2 MATERIALS AND T EST SPECIMEN DESCRIPTION The General Electric (GE) designed Hatch Unit 1 300° surveillance capsule was removed from the plant and shipped to MP Machinery and Testing, LLC (MPM) for analysis.
The capsule conta i ned a total of two Charpy packets and three tensile tubes. The 300° surveillance capsule is an original plant capsule, and was irradiated in the plant since initial startup. This is the third surve i ll ance capsule to be removed from the Hatch Unit 1 reactor pressure vesse l (RPV) and tested. The 30° survei ll ance capsule was teste d by GE and the results are reported in Reference
[8]. The 120° surveillance capsule was also tested by GE and the results are reported in Refer e nce [9]. Flux wires from the 30° position were removed after cycle 1 and were also tested by GE [10). 2.1 Dosimeters The dosimetry wires were located along the ends of the Charpy specimens within each of the Charpy packets during irradiation.
Each of the two Charpy packets contained one high purity iron wire, one high purity copper wire, and one high purity nicke l wire for fluence eval u ation. Further detai l s on the exact wire locations during the irradiation are provided in the capsule open i ng discussion given in Section 2.3. A detai l ed discussion of the radiometric analysis of the capsu l e dosimetry wires is provided in Appendix A. 2.2 Test Materials The Hatch Unit 1 300° survei ll ance capsule Charpy V-notch specimen inventory, material descriptions, unirradiated (baseline)
Charpy impact data, and previous l y measured data are summarized in this section of the report. 2.2.1 Capsule Loading Inventory The Hatch Unit 1 300° surveillance capsule inventory is provided in Tab l e 2-1. All of the capsule specimens , which include Charpy specimens , tensile specimens , and dosimeters , were recovered from the capsule basket. Testing was performed on the 24 Charpy specimens , and the dosimetry wires were co u nted and weighed to determine specific activities. All six of the tensile specimens (two base, two weld, and two h eat-affected zone [HAZ]) remain unteste d and are being held in reserve for future testing since there is no near-term use for the tensi l e data. The technical advan t age of storing the tensile specimens untested is that there will be options in the future for how these specimens will be used to obtain useful data. For example, the tensi l e specimen geometry is conducive to fabr i cation of sub-size Charpy as we ll as min i aturized Charpy V-notch specimens.
Further , research is currently underway to develop testing methods which will enable the de t ermination of p l ane-strain
- fracture toughness data from Charpy-sized specimens. With these new technologies in view , there may also be a future need for static and/or dynamic tensile data for use in the calculation of :fracture toughness from experimental data obtained from 2-1 Materials and T es t Specimen D esc ription Charpy specimens.
Therefore , all of the tensile specimens have been placed into the archive storage so that they can be tested when necessary in the future. The broken Charpy specimen halves have been added to long-term archive storage for future use in miniature mechanical behavior specimen testing, reconstitution, chemistry analysis, and microstructural studies. As indicated in Table 2-1, there were two Charpy packets in the capsule, and each contained three dosimetry wires (one Fe wire, one Cu wire, and one Ni wire) and 12 Charpy specimens.
A drawing of the Charpy test specimen is shown in Figure 2-1 for reference.
Photographs of the capsule are given in Figures 2-2 through 2-4. The markings on the outside of the capsule, including the reactor code and the capsule code were recorded and verified.
2-2 Table 2-1 Hatch Unit 1 300° Surveillance Capsule Specimen lnventory 1 Number of Number of Charpy Specimens Flux Wires Charpy Relative Packet Vertical No.2 Base Weld HAZ Fe Cu Ni Position Highest G5 0 4 8 1 1 1 Charpy Packet in the Basket Lowest G4 8 4 0 1 1 1 Charpy Packet in the Basket 1. The surveillance program includes tensile specimens, but the tensile specimens were not tested. Four of the tensile specimens for this capsule we re l ocated at axial positions below Charpy packet GS and above Charpy packet G4. Two of the tensile specimens for this capsule were l ocated below Charpy packet G4. 2. The packet numbers in this table are organized by axial position in the capsule with packet G4 at the lowest ele va tion in the reactor and packet GS at the highest elevation in the reactor.
Mat e rials and T es t Specimen D escrip tion R \ 0.394 O.ITT9-+ ---f-I I -I 0.3 94 ASTM E23 [2 0] permissible variations shall be as follows: Notch length to edge: Adjacent sides shall b e at: Cross-sectional dimen sio n s: Length of specimen (L): Ce ntering of notch (L/2): Angle of notch: R a diu s of notch: Notch depth: Finish requirements
- Fi g ur e 2-1 90 +/- 2 degre es 90 degrees +/- 10 minute s +/- 0.003 inche s 2.165 (+0.0 , -0.100) inch es +/- 0.039 inche s +/- 1 degree 0.010 +/- 0.001 inche s +/- 0.001 inche s 63 µ-inch on notched surface and opposite face; 4 µ-inch e l sewhere Draw i ng Showing the Charpy Test Specimen Geometry and ASTM E23 Permissib le V a riat i ons 2-3 Materials and T es t Sp eci m en D esc ription Figure 2-2 Photograph of the 300° Hatch Unit 1 Capsule (top) and a Magnified View of the External Identification Markings (bottom) Figure 2-2 shows the side of the surveillance capsule which faced the core. The identification code, "1 17C4671G3", was engraved near the hook. 2-4 Mat e rials and Test Specimen D esc ription Figure 2-3 Photograph of th e 300° Hatch Unit 1 Capsule Figure 2-3 s h ows the side of the surve ill ance capsule which faced the pressure vessel. T h e reac t or and ca p s u le co d es are seen near t he h ook. 2-5 Mat e rials and T e st Sp ec im e n D esc ription Figure 2-4 Photograph of the Inside of the 300° Hatch Unit 1 Capsule 2.2.2 Material Description The Hatch Unit 1 RPV is a 218-inch (5537mm) diameter BWRJ4 design. The pressure vessel construction was performed by Combustion Engineering to the 1965 Edition of the ASME Code, with Winter 1966 Addenda. The pressure vessel shell and head plate materials are ASME A533, Grade B , Class 1 low alloy steel. The nozzl es and closure flanges are A508 Class 2 low alloy steel, and the closure flange boltin g materials are ASME SA540 Grade B24 low alloy steel. The fabrication process employed quench and temper heat treatment immediately after hot forming, then submerged arc we lding and post-weld heat treatment.
The post-weld heat treatment was typically 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> at 1150°F +/-25°F (621.1°C +/-13.9°C). The surveillance base metal specimens were machined from plate G-4804-2 from the beltline. The test plate was heat treated for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> at l 150°F +/-25°F (621.1°C +/-13.9°C) to simulate the post weld he at treatment of the vessel. Specimens were machined from the Vi T and % T positions in the plate, in the longitudinal orientation (long axis parallel to the rolling direction).
The weld material and HAZ Charpy specimens were fabricated from trim-off pieces of plates G-4804-2 and G-4804-1 that were welded together with a weld identical to longitudinal seam we ld 1-308 in the RPV b eltline. The welded test plate for the weld and HAZ C harp y specimens received a heat treatment of 1150°F +/-25°F (621.1°C +/-13.9°C) for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> to match the 2-6 Mat e rials and T e st Spe c im en D es cription fabri c ated condition of the RPV. The base metal orientation in the weld and RAZ specimens was longitudinal
[8]. All specimens are stamped on the ends with the fabrication code listed on the Hatc h Unit 1 drawings.
2.2.3 Chemical
Composition Table 2-2 details the best estimate average chemistry values for plate heat C4114-2 surveillance material.
Chemical compositions are presented in weight percent. If there are multiple measurements on a single specimen, those are first averaged to yield a single value for that specimen, and then t h e different specimens are averaged to determine the heat best estimate.
The best estimate chemistry values for the surveillance weld material are not provided because , as stated in Section 1.0 of this report, the weld material is not used in the ISP and cannot be evaluated.
Table 2-2 B e st Estimate Chemistry of Available Data Sets for Plate Heat C4114-2 Cu (wt%) Ni (wt%) p (wt%) s (wt%) Si (wt%) Specimen ID Source 0.11 0.68 --0.21 DEM 0.11 0.57 --0.19 D4L 0.12 0.71 --0.2 2 CUB Reference 9 0.13 0.77 0.015 -0.26 D14 0.13 0.78 0.014 -0.26 CUP 0.12 0.68 0.014 -0.19 CUD 0.13 0.7 0.010 0.013 0.28 Baseline CMTR Reference 9 0.12 0.70 0.013 0.013 0.23 +-Best Estimate Average 2-7 Materia l s and Test Sp e cimen Description 2.2.4 CVN Baseline Properties Ta bl e 2-3 c on ta ins t h e un irra di a t e d C h a rp y d ata for t h e C411 4-2 s u rve ill ance p l a t e mate ri a l. As state d in Sec tion 1.0 of t h is re po rt, u ni rra d iate d C h arpy d a t a is n ot avai l a bl e for t h e survei ll ance we ld m a t eria l. 2-8 Table 2-3 Unirradiated Longitudinal Charpy V-Notch Impact Test Results for Surveillance Base Metal (Heat C4114-2) Specimens from the Hatch Unit 1 Surveillance Program Base Unirradiated:
Heat C4114-2 , Longitudinal Test Lateral Specimen Temperature Impact Energy Expansion Percent S h ear ID O F (oC) ft-lb (J) mils (mm) (%) 1 -80 (-62.2) 29.0 (39.3 2) 2 3.0 (0.58) 1 2 -8 0 (-6 2.2) 27.0 (36.61) 20.0 (0.5 1) 1 3 -4 0 (-40.0) 34.0 (46.10) 25.0 (0.64) 20 4 -40 (-40.0) 38.0 (51.52) 29.0 (0.74) 20 5 -40 (-40.0) 47.0 (63.72) 35.0 (0.89) 25 6 10 (-1 2.2) 84.0 (113.89) 57.0 (1.4 5) 70 7 10 (-1 2.2) 80.0 (108.47) 58.0 (1.4 7) 50 8 10 (-1 2.2) 82.0 (1 11.1 8) 59.0 (1.50) 50 9 4 0 (4.4) 106.0 (143.72) 66.0 (1.68) 80 10 4 0 (4.4) 89.0 (120.67) 73.0 (1.85) 70 11 4 0 (4.4) 90.0 (122.0 2) 62.0 (1.57) 70 12 110 (4 3.3) 133.0 (180.3 2) 85.0 (2.16) 1 00 13 1 10 (43.3) 142.0 (192.5 3) 89.0 (2.26) 100 1 4 16 0 (7 1.1) 1 30.0 (1 7 6.2 6) 83.0 (2.11) 1 00 15 160 (71.1) 139.0 (188.4 6) 86.0 (2.18) 1 00 Mat e rials and T es t Sp ec im en D e s c ription For the surveillance plate material , the baseline test data were fit to a hyperbolic tangent curve using the computer program CVGRAPH [11]. Figure 2-5 shows the fitted Charpy energy data curve for the unirradiated plate. Figure 2-6 shows the fitted lateral expansion curve for the unirradiated plate. Table 2-4 summarizes the unirradiated (baseline)
Charpy notch properties (index temperatures) of plate heat C4114-2. In this table and throughout this report, T 3o is the 30 ft-lb (41 J) transition temperature; T s o is the 50 ft-lb (68 J) transition temperature; T 3 smil is the 35 mil (0.89 mm) lateral expansion temperature
- and USE is the average energy absorption at full shear fracture appearance. Table 2-4 Baseline CVN Properties Material T3o Tso T3sm i1 Upper Shelf Identity Material OF (OC) OF (OC) OF (OC) Energy (USE) ft-lb (J) C4114-2 (LT Hatch Un it 1 -61.5 -28.8 -34.7 136.0 Orientation)
Surve illanc e Plate (-51.9) (-33.8) (-37.1) (184.4) 2-9 Mat e r i a l s and T e st Spe ci men D escr iption PLATE HEAT C4114-2 (HAl) CV Graph C>.Cl2: H y perbo l ic T a n ge nt Curve Printed on 2/23/2 017 3 : 52 PM A= 69.25 B = 66. 75 C = 86.30 TO = -3.28 D = 0.00 Corre l atio n Coefficient=
0.984 E q u ation is A+ B * (Tanh((T-TO)
/(C+DD)J U pper Shelf Energy= 136.00 (Fixed) Lower Shelf E ner gy= 2.50 (Fixed) Tem p@30 ft-lbs=-6 I .50° F T e m p@_15 fl-l b s=-5 2.20° F Te m p@,50 ft-l b s=-28.80° F P l a nt: Ha t ch 1 Orientation
- L T Mate ri al: SA533 Bt Ca ps u le: UNIRRA Heat: C41 1 4-2 Fluence: O.OOE+oOO n/c m' --r;l:J ..c -I ¢:: _, ;;..-. '"" Cl.I = :z > u 1 60 ._ 1 40 -120 ._ 100 -* . 80 -60 40 20 ._ 0 -300 CVGmph 6.02 Figure 2-5 n v / I /o T I I i6 / __/ -200 -100 0 100 200 300 Temperature
{° F) 02/23/20 17 Charpy Energy Plot for Plate Heat C4114-2 (LT) Un irradiated 2-1 0 ' --400 500 600 Pa ge 1/2 P lant: Hatc h 1 Orientati o n: LT Tempera tur e (0 F) 80 -4 0 -4 0 -40 1 0 I O 1 0 4 0 4 0 4 0 110 11 0 160 16 0 CVGraph 6.02 Fig u re 2-5 (Continued)
Materials and T e st Specimen D escriptio n Ma t e rial: SA533Bl Ca psule: UN1RRA PLATE HEAT C4114-2 (HAl) Charpy V-Notch Data Input CVN Co mpu ted CVN 29.0 21.8 2 7.0 21.8 34.0 4 2.4 38.0 4 2.4 47 0 4 2.4 8 4.0 7 9.4 80.0 7 9.4 82.0 7 9.4 1 06.0 1 00.2 89.0 1 00.2 90.0 1 00.2 133.0 1 27.0 142.0 127 0 130.0 133.0 139.0 133.0 02/2 3/20 17 Heat: C 4114-2 F lu e nce: O.OOE+OOO n/cm* Differential 7.20 5.20 -8.45 -4.45 4.55 4.56 0.56 2.56 5.83 -1117 -1 0.17 602 1 5 02 -303 5.9 7 Page 212 Cha r py Energy Plot for Plate Heat C4114-2 (LT) Unirradiated 2-11 Materials and T e st Sp ec im e n D e s c ription Plant: Hatc h 1 Orientation
- LT --*-6 "-' = 0 *-= Q. --90 70 50 30 20 10 0 -300 CVGraph 6.02 Figure 2-6 UNIRRADIATED PLATE HEAT C4114-2 LE CV Grap h 6.0 2: H y pert>olic Tan g ent Curv e Prin l ed on 4/1 2/20 1 7 7: 45 AM A= 43.38 B = 42.38 C = 84.93 TO= -1 7.77 D = 0.00 Corre l a tion Coeffi c i e nt= 0.9 85 Equa1ion is A+ B * [Tanh ((T-TO)/(C+D D)J Upper Shelf LE.= 85.7 5 (F ixed) Low e r Sbelf L.E. = 1.00 (f i....,ed)
I 1 9(o 7 J L/ I I T e mp@-15 70° F M a t e rial: SA533 Bl Ca psule: UNIRRA u 0 0 /"t> o( lo , I I . I I H eat: C4 114-2 Flu e n ce: O.OOE+-000 n/cm' *-* I I -200 -100 0 100 200 300 400 500 600 Temperature
(° F) 04/12/2 0 1 7 Page 1/2 Lateral Expansion Plot for Plate Heat C4114-2 (LT) Un irradiated 2-12 Mate r ia l s and T e st S p eci m en D escrip ti on P lant: Ha tch 1 Orientation:
L T Ma terial: SA5 33Bl Capsu le: UNIRRA Heat: C 4114-2 F lu e nce: O.OOE+OOO n/cm 2 UNIRRADIATED PLATE HEAT C4114-2 LE Charpy V-Notch Data Te mp e r a tur e(° F) In put L. E. C omput e d L. E. Differ e nti a l -80 23.0 16.9 6.1 0 -80 20.0 16.9 3.1 0 -40 25.0 32.5 -7.53 -4 0 29.0 32.5 -3.53 -40 35.0 32.5 2.47 1 0 57.0 56.8 0.24 10 58.0 56.8 1.24 1 0 59.0 56.8 2.2 4 4 0 66.0 68.4 -2.45 4 0 73.0 68.4 4.55 4 0 62.0 68.4 -6.45 11 0 85.0 81.8 3.24 11 0 89.0 81.8 7.2 4 16 0 83.0 84.5 -1.48 16 0 86.0 84.5 1.5 2 CVGraph 6.02 04/1 21201 7 Page 212 Figure 2-6 (Continued)
Lateral Expansion Plot for Plate Heat C4114-2 (LT) Un irradiated 2-13 Mat e rials and T e st Sp ec im en D esc ription 2.3 Capsule Opening As shown in Figures 2-2 through 2-4 , the 300° surveillance capsule consisted of a container holding two Charpy packets and three tensile tubes. Each Charpy packet contained 12 Charpy specimens.
The outside of the capsule had identification markings which could be clearly read. On one side, the capsule container was marked with the reactor and capsule codes. The reactor code matches the reactor code in Reference
[8]. The capsule container was engraved with the marking "117C4671G3" on the side facing the core. Attention was paid to the location of the Charpy packets and specimen and dosimetry wires during disassembly of the capsule. The dosimetry wire location along the ends of the Charpy specimens is shown in Figure 2-7. Referring to the figure, the eight base metal specimens and four weld specimens in the G4 Charpy packet were installed at the bottom of the capsule and the eight HAZ specimens and four weld specimens in the G5 Charpy packet were installed in the top of the capsule. Specimen orientation within the packets can be seen in Figures 2-8 and 2-9. The dosimetry wires and Charpy specimens were placed in individually marked containers for positive identification throughout the work. During disassembly , the Charpy specimens in the G4 packet were found to be heavily oxidized (Figure 2-9) due to water in-leakage. Prior to testing , the Charpy specimens were carefully cleaned by hand lapping in a specially designed fixture. Every effort was made to preserve the ASTM E23 [12] dimensional tolerances and surface finish requirements. Despite all of the efforts to keep the test specimens within dimensional tolerances , some had experienced corrosive attack that resulted in an out-of-tolerance condition after cleaning.
After the hand lapping , the MPM Digital Optical Comparator (DOC) was used to measure the notch included angle, notch depth , and notch radius. A calibrated micrometer was used to measure the cross-sectional dimensions. The remaining ligament, or distance from the root of the notch to the back side of the specimen across the uncracked ligament , was found by subtracting the notch depth from the specimen width. The results of these measurement s are provided in Table 2-5. All of the specimens with an out-of-tolerance measurement are shown in the table. The data for the G5 packet, which did not l eak , are also shown in the tab l e for comparison.
The out-of-tolerance notch depth condition for spec im en CUK from the G4 packet is not considered important because the remaining li gament meets E23 requirements.
Specimens CUU , CY6, CY2 , CU4 (base metal specimens) and D45 (weld metal specimen) exhibited remaining ligament tolerance measurements (Note: These specimens were tested as-is and , when curve-fit, exhibited no anomalous results. See Sections 4 and 5.). 2-14 Materials and T es t Spe c imen D esc ription 0 r-HAZ HAZ U.A7 UA? IUA7 HAZ UA7 U47 Weld Weld Welc Weld V-notcb facing up \0 VI QO v r--D v 3 dosimetry wires in Chlllpy packet 000 0 n6' D4' DE o's' D4' n6' D6 D4' rib D3 D'i D°I' v .. 0. 0 L c y p y p A B L 2" L D I/ HAZ I HAZ -']3 Weld I If Weld -G2 0 r-Weld V-notcb facing up '° "<:" Weld Welc Weld Base Base Base Base !Base Base Base Base {)o v r--D I/ 3 dosimetry wires in Chlllpy packet 000 0 rib r)j D4 D2 cij cY o'i' cu n'i' cY cr cij v . *0*. K D "5 "K 4 6 7 K B 2 u u I/ I Base Base _1 \ Added white paint to identify orientation on all tensile specimens
-Fig u re 2-7 Drawing of the Hatch Unit 1 300° Surveillance Capsule Contents Showing the Test Spe c imen Identific a tion Markings 2-15 M a t e ria l s and T e st Sp ec i m e n D es cription Figure 2-8 Photograph of the Inside of the GS Charpy Packet within the 300° Hatch Unit 1 Capsule. Figure 2-9 Photograph of the Inside of the G4 Charpy Packet within the 300° Hatch Unit 1 Capsule. 2-16 Ma t e r i a l s and Test Sp e c im en L ) 2-5 py Specimen Dimensional Measurements Compared to the ASTM E23 Dimensional Requirements after S pecimen Cleaning Meets Meets Remaining E23 Meets Meets Notch Notch E123 Remaining Specimen Specimen Cross-Radius Radius E23 Angle Angle E23 Depth Depth Notch Ligament Ligament Thickness Width Section Side 1 Side 2 Radius Side 1 Side2 Angle Side 1 Side2 Depth Side 1 S i de2 Material (inches) (inches) Req.1 (inches) (inches) Req. 2 (degrees) (degrees)
Req. 3 (inches) (inches) Req. 4 (inches) (inches) Packet G4 Base 0.3940 0.3 940 yes 0.00901 0.00912 yes 45.78 45.85 yes 0.0804 0.0795 no 0.3136 0.3145 Base 0.3940 0.3940 yes 0.00908 0.00906 yes 45.67 45.78 yes 0.0797 0.0798 yes 0.3143 0.3142 Base 0.3938 0.3941 yes 0.00906 0.00903 yes 45.35 45.94 yes 0.0797 0.0795 yes 0.3144 0.3146 Weld 0.3943 0.3937 yes 0.0098 0.0095 yes 44.23 44.40 yes 0.0791 0.0793 yes 0.3146 0.3144 Weld 0.3941 0.3939 yes 0.0096 0.0100 yes 44.89 44.95 yes 0.0789 0.0791 yes 0.3150 0.3148 Weld 0.3944 0.3938 yes 0.0093 0.0090 yes 44.04 45.01 yes 0.0798 0.0797 yes 0.3140 0.3141 Base 0.3944 0.3940 yes 0.0094 0.0092 yes 44.73 44.16 yes 0.0804 0.0812 no 0.3136 0.3128 Base 0.3944 0.3940 yes 0.0096 0.0098 yes 45.12 45.10 yes 0.0800 0.0797 yes 0.3140 0.3143 Base 0.3939 0.3941 yes 0.0094 0.0097 yes 44.82 44.91 yes 0.0800 0.0801 no 0.3141 0.3140 Base 0.3940 0.3941 yes 0.0096 0.0098 yes 44.73 45.30 yes 0.0802 0.0808 no 0.3139 0.3133 Weld 0.3943 0.3939 yes 0.0103 0.0102 yes 44.55 45.26 yes 0.0803 0.0800 no 0.3136 0.3139 Base 0.3942 0.3939 yes 0.0108 0.0102 yes 44.18 44.98 yes 0.0796 0.0805 no 0.3143 0.3134 and T e st Sp e cim en D e s c ription I ipecimen D imensional Measurements Compared to the ASTM E23 Dimensional Requirements after Specimen Cleaning (continued)
Meets E23 Cross-Specimen Specimen Section Radius Material Thickness Width Req.1 Side 1 (inches) (inches) (inches) HAZ 0.3940 0.3938 yes 0.0092 HAZ 0.3939 0.3 937 yes 0.0090 HAZ 0.3940 0.3936 yes 0.0090 HAZ 0.3941 0.3935 yes 0.0095 HAZ 0.3941 0.3938 yes 0.0092 HAZ 0.3941 0.3938 yes 0.0091 HAZ 0.3940 0.3937 yes 0.0092 HAZ 0.3940 0.3938 yes 0.0092 Weld 0.3942 0.3939 yes 0.0095 Weld 0.3941 0.3939 yes 0.0091 Weld 0.3944 0.3940 yes 0.0093 Weld 0.3943 0.3939 yes 0.0095 23 requires the cross section dimensions to be 0.394 +/- 0.003 inches. 23 requires the notch radius to be 0.0100 +/- 0.001 inches. 23 require s the included angle to be 45 +/- 1 degree. 23 requires the notch depth to be 0.079 +/- 0.001 inche s. 23 requires the ligament length to be 0.315+/-0.001 inches. Radius Side 2 (inches) 0.0093 0.0093 0.0090 0.0095 0.0092 0.0093 0.0091 0.0093 0.0095 0.0099 0.0093 0.0093 Meets Meets E 2 3 E23 Notch Notch Radius Angle Angle Angle Depth Depth Req. 2 Side 1 Side 2 Req. 3 Side 1 Side 2 (degrees) (degrees) (inches) (inches)
Packet GS yes 44.35 45.28 yes 0.0787 0.0785 yes 44.23 44.32 yes 0.0782 0.0781 yes 45.17 44.44 yes 0.0787 0.0790 yes 45.37 44.58 yes 0.0788 0.0787 yes 45.14 45.38 yes 0.0787 0.0785 yes 45.50 44.69 yes 0.0781 0.0786 yes 44.84 45.34 yes 0.0787 0.0789 yes 44.50 45.10 yes 0.0787 0.0786 yes 45.04 45.07 yes 0.0791 0.0793 yes 44.54 45.04 yes 0.0791 0.0782 yes 45.17 45.60 yes 0.0789 0.0790 yes 45.23 44.95 yes 0.0789 0.0791 Meets E 1 23 N o tch Remaining Rer Depth Ligament Req. 4 Side 1 s (inches) (ir yes 0.3151 0 yes 0.3155 0 yes 0.3149 0 yes 0.3147 0 yes 0.3151 0 yes 0.3157 0 yes 0.3150 0 yes 0.3151 0 yes 0.3148 0 yes 0.3148 0 yes 0.3151 0 yes 0.3150 0 3 NEUTRON FLUENCE CALCULATION The surveillance capsule was placed in the Hatch Unit 1 reactor's 300° surveillance capsule holder prior to Cycle 1 and was removed following Cycle 27 for a total irradiation period of 32.0 EFPY. The surveillance capsule included copper , iron, and nickel flux wire dosimetry specimens for w h ich activation measurements were taken. Eva l uation of the sur v eillance capsule specimens requires know l edge of the neutron irradiation environment.
The neutron flux density , neutron energy spectrum, and neutron fluence are required at the surveillance capsule location.
The NRC has established guidelines in Regulatory Guide 1.190 [ 13] for determining best estimate values of flux, energy spectrum, and fluence for RPV d amage assessments using computationa l methods. These guidelines are not specifically intend e d for use in surveillance capsule evaluations; however , the guidelines provide suitable guida n ce to support the development of accurate neutron transport analysis models for surveillance capsule evaluations.
This section docume n ts the application of the modeling and analysis guidelines provided in [14] to determine the surveillance capsule accumulated irradiation (e.g., activation) and capsule specimen neutron fluence of the Hatch Unit 1 300° surveillance capsule flux wires. Additionally, the activation for the 30° flux wires , removed at the end of Cycle 1 , 30° surveillance capsule, removed at the end of Cycle 8, and 120° surveillance capsule, removed at the end of Cycle 16, was determined.
The fast neutron fluence (E >1.0 MeV) was also calculated for all capsules at the time of removal. The fluence and activation values presented in this report were calculated using the RAMA Fluence Methodology
[14] (hereinafter referred to as "RAMA"). The specific activities predicted by RAMA are compared to the activity measurements reported in Appendix A. RAMA has been developed for the Electric Power Research Institute, Inc. (EPRl) and the BWRVIP for the purpose of calculating neutron fluence in Boiling Water Reactor (BWR) compo n ents. As prescribed in Regulatory Guide 1.190, RAMA has been benchmarked against industry standard benchmarks for both BWR and pressurized water reactor (PWR) designs. In additio n , RAMA has been compared with several plant-specific dosimetry measurements and reported fluences from several commercial operating reactors. The results of the benchmarks and comparisons to measurements show that RAMA accurately predicts specimen activities, RPV fluence, and vessel internal component fluence in all light-water reactor types. Under funding from E P RI and the BWRVIP, the RAMA methodology has been reviewed by the U.S. NRC and subsequently given generic approval for determining fast neutron fluence in BWR and PWR pressur e vessels [15 , 16] and BWR vessel internal components that include the core shroud and top guide [17]. 3-1 Ne utr o n Flu e n ce C a l c ul a tion 3.1 Description of the Reactor System This section provides an overview of the reactor design and operating data inputs that were used to develop the Hatch Unit 1 reactor fluence model. All reactor design and operating data inputs used to develop the model were plant-specific and were provided by the Hatch Unit 1 reactor operator, Southern Nuclear Operating Company, Inc. (SNOC). The inputs for the fluence geometry model were developed from design and as-built drawings for the RPV vessel internals, fuel assemblies, and containment regions. The reactor operating data inputs were de v eloped from core simulator data , when available, and other sources, that provided a historical accounting of how the reactor operated for Cycles 1 through 27. 3. 1. 1 Overview of the Reactor System Design Hatch Unit 1 is a General Electric BWR/4 class reactor lo cated near Bax l ey , GA. The reactor has a core loading of 560 fuel assemb l ies and began commercia l operation in December 1975 with a design rated power of 2436 MWt. In Cycle 17 , a power up-rate was achieved , raising the power to 2558 MWt. Two additional up-rates were achieved:
in Cycle 19 to 2763 MWt, and in Cycle 22 to 2804 MWt. At the time of this fluence analysis , Hatch Unit 1 h a d completed 27 cycles of operation. Figure 3-1 illustrates the basic planar configuration of the Hatch Unit 1 reactor at an axial elevation near the reac t or core mid-plane.
All of the radial regions of the reactor that are required for fluence projections are shown. Beginning at the center of the reactor and projecting outward, the regions include: the core region , including control rod locations and fuel assembly locations (fuel l ocations are shown only for the 0° to 90° quadrant); core reflector region (bypass water); centra l shrou d wall; downcomer water region , including the jet pumps; RPV wall; cavity between RPV and insu l ation; insulation; ca v ity region between the insu l ation and biological shie ld; and biological shield (concrete wall). 3.1.2 Reactor System Mechanical Design Inputs The mechanical design inputs that were used to construct the Hatch Unit 1 fluence geometr y mo d el included as-built and nominal design dimensional data. As-bui lt data for the reactor components and regions of the reactor system is always preferred when constructing specific models; however , as-built data is not a l ways available.
In these situations , nominal design information is used. For the Hatch Unit 1 fluence model , the predominant dimensional information u sed to construct the fluence model was nominal design data. As-built data was used for the following dimensions:
- Core shroud radii Another important component of the fluence ana l ysis is the accurate description of the surveillance capsules in the reactor. It is shown in Figure 3-1 that three surveillance capsules were initially installed in the Hatch Unit 1 reactor. The capsules were attached radially to the inside surface of the RPV (looking o u tward from the core region) at the 30°, 120°, and 300° a z imuths. 3-2 Ne utr o n Flu e n ce Cal c ulation 330° 315° 3 00° + + + + + + + + 270° + + + + + + + + 240° 225° 210* Notes: Th i s draw i ng i s not to scale. Figure 3-1 F = Fuel bundle locations. (Locations shown only for the northeast quadrant.) + = Contro l rod locations. + + + + + + + + Reactor North o* + + + + f, if! * ;+; ;+f + f, ;+, ;+, * + + + + + + +
+ + + + + + + + + + + + + + + + + + + + + 150° Cavity Regions 1eo* Biological Shield Planar View of Ha tc h U nit 1 at the C or e Mid-plan e Elevatio n 3. 1.3 Reactor Sy stem M aterial Compos it io n s Surveillance Capsule Shroud Repair Tie Rod Core Reflector Downcomer 90° Reactor East Reactor P r essure Vessel and Clad Thermal Insulation 135° Jet Pump Assembly Each region of the re a ctor is comprised of materials that include reactor fuel , steel, water , insulation, concrete , and air. Accurate material information is essential for the fluence evaluation as the material compositions determine the scattering and absorption of neutrons throughout the reactor system and, thus, affect the determination of neutron fluence in the reactor components.
Tab l e 3-1 provi d es a summary of the materia l s for the various components and regions of the Hatch U nit 1 reactor. The material attributes and compositions (i.e., material densities and isotopic concentrations) for the steel , insulation, concrete, and air are assumed to remain constant for the operating life of the reactor. The attributes of the fuel compositions in the reactor core region change continuously during an operating cycle due to changes in power level , fuel burnup, control rod movements , and changing moderator density levels (voids). Due to the dynam i cs of the fuel attributes with reactor operation , several state-point data sets are used to descr i be the operating states of the reactor for each operating cycle. 3-3 Neu tr o n Flue n ce Calcula t ion Table 3-1 Summary of Materia l Compositions by Region for Hatch Unit 1 Region Material Composition Control Rods Stainless Steel , B 4 C Control Rod Guide Tubes Stainless Steel Core Support Plate Stainless Steel Core Support Plate Rim Bolts Stainless Steel Fuel Support Piece Stainless Steel Fuel Assembly Lower Tie Plate 'stain l ess Steel , Zircaloy , l nconel Reactor Core 235 LJ , 23B LJ , 239 Pu , 24D Pu , 2 4 1 Pu, 2 42 Pu , O tuel, Zircaloy Reactor Coolant I Moderator Water Core Reflector Water Fuel Assemb l y Upper T i e P l a t e Sta i nless Steel , Zir c a l oy , lnco n el Top G ui de S t a i nless S t eel Core Spray Sparger Pipes Stainless Steel Core Spray Sparger Flow Areas Water Shroud Stainless Steel Downcomer Reg i on Wate r Jet Pump R i ser and M i xer Flow Areas Water Jet Pump R i ser and Mixer Metal Sta i nless Stee l Jet Pump R i ser B r ace and Pads S t a i nless Steel Jet Pump Hold Down Beams lnconel Jet Pump Hold Down Brackets Stainless Steel Surve i llance Capsu l e Spec i mens Carbon St e e l Shroud R e pa ir T i e Rod Sta inl ess Steel , l n conel Reacto r Pressure Vessel Clad Stainles s S t eel Reac t or Pressure Vessel Wall Carbon Steel Reactor Pressure Vessel Nozzle Forgings Carbon St e el Cavity Regions A i r (Nitrogen) Insulation Clad Stainless Steel Insulat i on S t a i n l ess Ste el , Air Biological Shie l d Clad Ca r bon S teel Bio l ogical Sh i eld Wal l Re i nforced Conc r e t e 3-4 N e utron Flu e n c e Cal c ulation 3. 1.4 Reactor Operating Data Inputs An accurate evaluati o n of reactor vessel and component fluence requires an accurate accounting of the reactor's operating history. The primary reactor operating parameters that affect the determination of fast neutron fluence in light-water reactors include reactor power levels, core powe r distributions, c oolant water density distributions, and fuel material (isotopic) distributions.
3.1.4.1 Core Loadi n g It is common in BWRs that more than one fuel assembly design may be loaded in the reactor core i n any given operating cycle. For fluence evaluations , it is important to account for the fuel assembly designs that are loaded in the core in order to accurately represent the neutron source distribution at the co r e boundaries (i.e., peripheral fuel locations and the top and bottom fuel elevations).
Several different fuel assembly designs were loaded in the Hatch Unit 1 reactor during the period inclu d ed in this evaluation. Table 3-2 provides a summary of the fuel mechanical designs loaded in the reactor core for each evaluated operating cycle. The cycle core loading provided by SNOC was u s ed to identify t he fuel assembly designs in each cycle and their location in the core loading inventory.
For each cycle , appropriate fuel assembly models were used to build the reactor core regio n of the Hatch Unit 1 fluence model. 3.1.4.2 Power History Data Reactor power history is the measure of reactor power levels and core expo s ure on a continual or per i odic basis. For this fluence and activation evaluation , the power history for the Hatch Unit 1 reactor was developed from power history inputs provided by SNOC. The power history data showed that Hatch Unit 1 started commercial operation with a design rated thennal power of 2436 MWt. In Cycle 17 , a power up-rate was achieved , raising the power to 2558 MWt. Cycle 19 began operation at 2763 MWt. Cycle 22 and beyond operated at 2804 MWt. The power history data for Hatch Unit 1 included daily power levels for most cycle s. For cycles where daily power hi s tories were unavailable, average power levels were constructed based on exposure accumulation.
This data was used to calculate the capsule and vessel fluence. Periods of reactor shutdown due to refueling outages and other e v ents were also accounted for in the model. The power history data was verified by comparing the calculated energy production in effective full power years with power production records provided by SNOC. Tabl e 3-3 list s the accumulated EFPY at the end of e ach c y cle for this fluence e v aluation.
3-5 Neutro n Fluence Ca l culatio n Table 3-2 Summary of Hatch Unit 1 Core Loading Inventory 7x7 8x8 Designs 9x9 Designs 10x10 Designs Designs Dominant Cycle GE3 GE4 GES/6 GE7 GES GE9 GE13 ANF92 GE12 GE14 Peripheral Design 1 560 GE3 2 468 92 GE3 3 300 92 168 GE3 4 136 92 332 GE3 5 92 468 GE3 6 132 428 GE3 7 76 484 GE3 8 64 436 60 GE3 9 76 321 163 GE3 10 229 331 GE5/6 11 41 515 4 GE5/6 12 4 44 108 4 4 GE7 13 264 108 184 4 GE7 14 100 108 348 4 GE8 15 4 4 5 507 4 GE8 16 1 555 4 GE9 17 55 3 3 4 GE9 18 365 191 4 GE9 19 176 380 4 GE 9 20 560 GE 13 21 336 224 GE13 22 126 434 GE 1 3 23 560 GE1 4 24 560 GE14 25 560 GE 14 26 560 GE 14 27 560 GE 14 3-6 N e utron Fluenc e Ca l cula ti o n Table 3-3 S t ate-point Data for Each Cycle of Hatch Unit 1 C ycle N u mber Number of State-Rated Thermal Power Accumulated EFPY Points (MWt)1 1 12 2436 1.2 2 3 2436 1.8 3 8 2436 2.6 4 1 2436 3.6 5 1 2436 3.9 6 1 2436 4.4 7 1 2436 4.9 8 1 2436 5.5 9 1 2436 6.2 10 1 2436 7.1 11 1 2436 8.1 12 1 2436 9.3 13 2 2436 10.4 14 10 2436 11.7 15 13 2436 12.9 16 11 2436 14.2 17 12 2558 15.6 18 12 2558 16.9 19 12 2763 18.2 20 12 2763 19.6 21 15 2763 21.3 22 19 2804 23.0 23 25 2804 24.8 24 21 2804 26.6 25 22 2804 28.4 26 18 2804 30.2 27 19 2804 32.0 1. The rated th erma l power i s li sted fo r each cycle. The ac tual power l evels we r e u sed fo r the individual p oi nt ca lcul atio n s for Cycles 1-2 7. 3-7 Neutro n Flu ence Calculation 3.1.4.3 Reactor State-Point Data Modem core simulator codes (e.g., PANACl 1) provide the preferred data used as input into fluence calculations. Core sim ulat or data was provided by SNOC to characterize the historical operating conditions of Hatch Unit 1. A total of 255 state-points were used to represent the operating history for the first 27 operating cycles of Hatch Unit 1. The state-points and corresponding data can be characterized into two categories regarding quantity and quality: for Cycles 1-15, minimal or unreliable data was available; for Cycles 16-27 , PANAClO and PANACl 1 core simulator data was available.
The data calculated with core simulator codes represents the best-available information about the reactor core's operating history over the reactor's operating life. In this analysis, the core simulator data was processed by Trans Ware to generate state-point data files for input to the fluence model. The state-point files included three-dimensional data arrays that described core power distributions , fuel exposure distributions, fuel materials (isotopics), and coolant water densities.
A separate neutron transport calculation was performed for each of the state-points tallied in Table 3-3. The calculated neutron flux for each state-point was combined with the appropriate power history data described in Section 3 .1.4.2 in order to provide an accurate accounting of the fast neutron fluence for the RPV and activation of the surveillance capsules. 3. 1.4. 3. 1 Beginning of Operation through Cycle 15 State-Points Cycles 1-15 required a state-point construction technique to approximate the missing "core simulator" data. In effect, power shapes, isotopics , water densities, and exposures are mapped onto a cycle-dominating fuel assembly type throughout the core. This also required addressing variances in the active fuel height for the loaded assemblies.
To avoid spurious power mappings (e.g., lOxlO pin powers on 7x7 assemb l y), pin power data from other cycles was not used on the constructed cycles , and only nodal power distributions were applied. Any data still missing after attempting to use cycle-specific resources was based on Cycle 16, primarily for moderator densities. For Cycles 1-3 , traversing in-core probe (TIP) trace data was available to determine the axial power distribution. Additionally, the core design and operating data was previously characterized
[18 , 19). Combined with assembly edge-to-near-edge power ratios and exposures, this information was expanded into state-points.
For Cycles 4-12 , process computer exposure and void (E&V) data at the local power range monitor (LPRM) locations was expanded into cycle-average nodal datasets. Fuel assembly loadings and shuffles were tracked across each cycle to generate the exposure distribution.
Cycle 16 data was a l so used to replace the remaining missing information.
For Cycle 13-14, PANACEA 8 output was available.
This data was based on ear l y core nodal simulator techniques and was missing the fuel types and isotopics. This information was determined and inserted into the state-point files to complete the fluence model inputs. For Cycle 15 , PANACEA 9 data was available. Similar to the PANACEA 8 data, this data was missing the fuel types and isotopics; thus , this information was determined and inserted into the state-point files. 3-8 Ne utron Flu e n ce Cal c ulation 3.1.4.3.2 Cycle 16 through Cycle 27 State-Points For Cycle 16 and beyond, the state-points were selected from hundreds of exposure points that were c alculated with the P ANAC 11 core simulator code. The hundreds of exposure points were evalu a ted and grouped into exposure ranges in order to reduce the number of transport calculations required to perform the fluence evaluation. Several criteria were used in the dete rm ination of the exposure ranges, including evaluations of core thermal powers, core flows, core p ower profiles, and control rod patterns.
In determining exposure ranges , it is assumed that there w ill be at least one exposure step in that range that would accurately represent the average opera t ing conditions of the reactor over that range. This sing l e exposure step is then referred to as the "state-point." Table 3-3 shows the number of state-points used for each cycle in this fluen c e eva luation. 3.2 Calculation Methodology This section provides an overview of the Hatch Unit 1 fluence model that was developed for the RAMA Fluence Methodology software [14]. The RAMA fluence model for Hatch Unit 1 is a plant-s pecific model that was constructed from the design inputs described in Section 3 .1. 3.2.1 Description of the RAMA Fluence Methodology The RAMA Fluence Methodology (RAMA) is a system of comp ut er codes , a data library , and an uncert a inty methodology that det ermines best-estimate fluence in light-water RPVs an d vessel comp o nents. The primary co de s that comprise the RAMA methodology include model builder codes, a particle transport code, and a fluence calculator code. The data library contains nuclear cross-sections and response functions that are needed for each of the codes. The uncertainty metho d ology is used to determine the uncertainty and bias in the best-estimate fluence calculated by the software.
The primary inputs for RAMA are mechanical design parameters and reactor operating history data. The mechanical design inputs are obtained from plant-specific design drawings , which includ e as-built measurements when availab le. The reactor operating history data is obta ined from r ea ctor core simu l ator codes, system heat balance ca l culations, daily operating logs, and cycle summary reports that describe the operating cond ition s of the reactor over its operating lifetim e. The primary outputs from RAMA ca l culations are neutron flux, neutron fluence, dosim etry activation, and an uncertainty evaluation. The model builder codes consist of geometry and material processor codes that generate input for the particle transport code. The geometry model builder code uses mechanical design inputs an d meshi ng specifications to generate three-dimensional combinatorial geometry models of the reactor. The material processor code uses reactor operating data input s to process fuel materials, structural materials , and water densities that are consistent with the geometry meshing generated by the g eometry model builder code. The pa rt icle transport code performs three-dimensional neutron flux calculations using a multigroup, deterministic, particle transport theory method with anisotropic scattering.
The primar y inputs prepared by the user for the transport code include the geometry and material data generated by the model builder codes and numerical integration and convergence parameters for the ite ra tive transport calculation.
The transport solver utilizes a three-dimensional method of 3-9 Ne utron Flu e n ce Calculation characteristics technique coup led with a genera l geometry modeling capability (combinatorial solid geometry). The coupling of general geometry with a deterministic transport solver provides a flexible, accurate, and efficient too l for calculating neutron flux in light -water RPVs and vessel components.
The primary output from the transport code is the neutron flux in multigroup form. The fluence calculator code determines fluence and activation in the RPV and vessel components over specifie d periods of reactor operation.
The primary inputs to the fluence calculator include the multigroup neutron flux from the transport code, response functions for the various materials in the reactor, reactor power levels for the operating periods of interest, the specification of which components to evaluate, and the energy ranges of interest for evaluating neutron fluence. The fluence calculator includes treatments for isotopic production and decay that are required to calculate specific activities for irradiated materials.
The reactor operating history is generally represented with several reactor state-points that represent the various power levels and core power shapes generated by the reactor over the life of the plant. These detailed state-points are combined with the daily reactor power l evels to produce accurate estimates of the fluence and activations accumulated in the p l ant. The uncertainty methodology provides an assessment of the overall accuracy of the RAMA Fluence Methodology. Variances in the dimensional data, reactor operating data, dosimetry measurement data, and nuclear data are evaluated to determine if there is a statistically significant bias in the calculated results that might affect the determination of the best-estimate fluence for the reactor. The plant-specific results are a l so weighted with comparative results from experimental benchmarks and other plant ana l yses and analytica l uncertainties pertaining to the methodology to determine if the plant-specific model under evaluation is statistically acceptable as defined in Regulatory Guide 1.190 [ 13]. The RAMA nuclear data library contains atomic mass data, nuclear cross-section data , and response functions that are needed in the material processing, transport, fluence, and reaction rate calculations.
The cross-section data and response functions are based on the BUGLE-96 nuclear data library [20] and the VITAMIN-B6 data library [21]. The RAMA Fluence Methodo l ogy is described in the Theory Manual [22]. The general procedures for using the methodology are presented in the Procedures Manual [23]. 3.2.2 RAMA Geometry Model for the Hatch Unit 1 Reactor Section 3 .1 describes the design inputs that were provided by SNOC for the Hatch Unit 1 reactor fluence evaluation. These design inputs were used to develop a plant-specific, three-dimensional computer model of the Hatch Unit 1 reactor for the RAMA Fluence Methodology. Figures 3-2 and 3-3 provide general illustrations of the primary components, structures, and regions developed for the Hatch Unit 1 fluence model. Figure 3-2 shows the planar configuration of the reactor model at an elevation corresponding to the reactor core mid-plane elevation.
Figure 3-3 shows an axial configuration of the reactor model. Note that the figures are not drawn to scale. They are intended only to provide a perspective for the layout of the model, and specifically how the various components, structures, and regions lie relative to the reactor core region (i.e., the neutron source). 3-10 Ne utron Flu e n ce Cal c ulation The figures are intended only to provide a general overview of the model and do not include illustrations of the geometry meshing developed for the model, because such detail is beyond the scope of this document.
The following subsections provide an overview of the computer models that were developed for the various compone n ts, structures , and coolant flow regions of the Hatch Unit 1 reactor. 0 1' ' 157.44 (399.89 cm! :::::f-------157.06 (398.94 cm 97.75 (248.29 cm) 88 775 (225.49 cm) ---1---87.275 (221.68cm) --I t-----18 17 16 t---+--+-....,...-+--+-+--+--+-+--+--+-+--ll 15 t---+--+-+--t--+-+--+----t-+--t----1,__
....... ....,. jj = 14 .... _. _.___.._.....__.___._....__.___,_....__.___..__...__..__
....... ""+-___ _ 90 14 15 16 17 18 19 20 21 22 23 24 Shroud Pressure Vessel Clad Notes: Th i s drawing is not to scale. Core Reflector D im ensions are given i n inches (cm). Figure 3-2 Planar View of the Hatch Unit 1 RAMA Quadrant Model at the Core Mid-plane Elevation 3-11 N e utron Flu e n ce Cal c ulation l:::: :J I::::! L:::-I I _ j l .J l -Sli*ou*""'" Rom -., . . ...... StvoudHead T op Gu id e. F u el Upper Tie Pla t e and Plefiu m Regtons Reactor Core !! c: .2 CJ) .. I 8 :1 "C 0. E -.; ..., L_j § .s:; ,i l (J) ! ! i : ' w 8oUom of R AMA Model Note: T h is drawing i s not t o scale. Figure 3-3 Axial View of the Hatch Unit 1 RAMA Model 3-1 2 ! 0 I N2 Nozzle N e utron Flu e n c e Cal c ulation 3.2.2.1 Geometry Model RAMA uses a generalized three-dimensional geometry modeling system that is based on a comb i natorial solid geometry (CSG), which is mapp ed to a Cartesian coordinate system. In this ana ly s is, an axial plane of the reactor model is defined by the (x,y) coordinates of the modeling system and the axial e l evation at which a plane exists is defined along a perpendicular z-a xis of the modeling system. Thus , any point in the reactor model can be addressed by specifying the (x,y,z) coordinates for that point. Figur e s 3-1 and 3-2 illustrate a planar cross-section view of the Hatch Unit 1 reactor design at an axia l e levation corresponding to the reactor core mid-plane elevation.
It is shown for this one elevat i on that the reactor design is a complex geometry composed of various combinations of rectan g ular, cylindrical, and wedge-shaped bodies. When the reactor is viewed in three dimen s ions, the varying heights of the diff erent components, structures, and regions create additional geometry modeling complexities.
An accurate representation of these geometrical compl e xities in a predictive computer model is essential for calculating accurate, best-estimate fluenc e in the RPV, the vessel internals, and the surrounding structures.
Figure s 3-2 and 3-3 provide general illustrations of the planar and axial geometry comp l exities that are represented in the Hatch Unit 1 fluence model. For comparison purposes , the planar view illustrated in Figure 3-2 corresponds to the same core mid-plane elevation illustrated in Figure 3-1. The computer model for Hatch Un it 1 assumes azimuthal quadrant symmetry in the planar dimension.
Figure 3-2 illustrates the quadrant geometry that was modeled in this analysis. In terms of the modeli n g coordinate system, the "northeast" quadrant of the geometry is represented in the model. The 0° azimuth, which has a " north" designation, corresponds to the 0° azimuth referen c ed in the plant drawings for the RPV. Degrees are incremented clockwise.
Thus , the 90° azimuth is designated as the "east" direction.
All other components, structures, and regions have been appropriately mirror reflected or rotated to this quadrant based upon their relationship to the pressur e vesse l orientation to ensure that the fluence is appropriate l y calculated relative to the neutron source (i.e., the core region). Although symmetry is a modeling consideration, the results present e d in this report for the different components and structures are given at their correct azimuths in the plant. Figure 3-3 illustrates the axial configuration of the primary components, structures, and regions in the fluence model. For discussion purposes, the same components , structures , and regions shown in the planar view of Figure 3-2 are also illustrated in Figure 3-3. Figure 3-3 shows that the axia l height of the fluence model spans from a lower elevation below the recirculation inlet nozzle t o above the core shroud head. This axia l height covers all areas of the RPV that are expected to exceed a fluence threshold of I.OE+ 17 n/cm 2 prior to the reactor's end oflicensed operation.
As previously noted, Figures 3-2 and 3-3 are not drawn precisely to scale. They are only intended to provide a perspective of how the various components, structures, and regions of the reactor are positioned relative to the reactor core region (i.e., the neutron source) and each other. The following subsections provide details on the modeling of individual components, structures, and regions. Please refer to the figures for visual orientation of the components and regions described in the following subsections.
3-13 Ne utron Flu e n ce Cal c ulation 3.2.2.2 Reactor Core and Core Reflector Models The reactor core contains the nuclear fuel that is the source of the neutrons that irradiate all components and structures of the reactor. The core is surrounded by a shroud structure that serves to channel the reactor coolant through the core region during reactor operation.
The region between the core and the core shroud is the core reflector , and it contains coolant. The reactor core geometry is rectangular in design and is modeled with rectangular elements to preserve its shape in the analysis.
The core reflector region interfaces with the rectangular shape of the core region and the curved shape of the core shroud. It is , therefore, modeled using a combination of rectangular and cylindrical elements.
The core region is centered in the RPV and is characterized in the analysis w ith two fundamental fuel zones: interior fuel assemblies and peripheral fuel assemblies.
The peripheral fuel assemblies are the primary contributors to the neutron source in the fluence calculation.
There is a sharp power gradient across these assemblies that requires consideration because these assemblies are loaded at the core edge where neutron leaka g e from the core is greatest.
To account for the power gradient , the peripheral fuel assemblies are sub-meshed with additional rectangular and triangular elements that preserve the pin-wise details of the fuel assembly geometry and power distribution.
The interior fuel assemblies make a lesser contribution to the reactor fluence and are, therefore, modeled in various homogenized forms in accordance with their contributions to the reactor fluence. For computational efficiency, homogenization treatments are used in the interior core region primarily to reduce the number of mesh regions that must be solved in the transport calculation.
The meshing configuration for each fuel assembly location in the core region is determined by parametric studies to ensure an accurate estimate of fluence throughout all regions of the reactor system. Each fuel assembly design , whether loaded in the interior or peripheral locations in the core, is represented with four axial material zones: the lower tie plate and end plug zone, the fuel zone , the fuel upper plenum zone , and the upper tie plate/end plug zone. The structural materials in the top and bottom nozzles for each unique assembly design are represented in the model to address the shielding effects that these materials have on the components above and below the core region. The fuel z one contains the nuclear fuel and structural materials for the fuel assemblies.
The materials for each fuel assembly are unique , and are modeled to accommodate the different fuel designs of each cycle. The materials are incorporated into the model using reactor operating data from core simulator codes. The upper plenum region captures fission gases during reactor operation. From an isotopic standpoint , the core is modeled using quadrant symmetry.
For the 30°, 120°, and 300° surveillance capsule evaluations , as well as the peak RPV fluence calculations, the northeast fuel quadrant is used. 3.2.2.3 Core Shroud Model The core shroud is a canister-like structure that contains the reactor core and channels the reactor coolant and steam produced by the core into the steam separators.
The shroud is composed of several components, of which the following are modeled and discussed here: upper shroud , central shroud , and lower shroud. The shroud head is discu s sed with other above-core components.
3-14 Ne utr o n Flu e n ce Cal c ulation 3.2.2.3.1 Upper Shroud The upper shroud wall mates the shr oud head to the top guide flan ge and central shroud wall. The wall protrudes beyond the central shroud regions and is modeled with cylindrical elements.
3.2.2.3.2 Central Shroud Axia lly , the shroud extends from the core support plate flange to the top guide flange. The central shro ud wall is modeled with cylindrical elements.
- 3. 2. 2. 3. 3 Lower Shroud Axially, the lowe r shroud extends from the bottom of the model to the core support plate flange. The lower shroud wa ll is modeled with conical elements to maintain the physical component's true s h ape. 3.2.2.4 Downcomer Region Model The d o wncomer region lies between the core shroud and the RPV. It is effective ly cylindrical in design , but with some geometrical complexities created by the presence of jet pumps and surveillance capsules in the region. The majority of the do wn comer region is modeled with cylindrical elements.
The areas of the downcomer containing the jet pumps , tie rods , and specimen capsules are modeled with the appropriate geometry elements to represent their design features and to preserve their radial , azimutha l , and axial placement in the downcomer region. These structures are described further in the following subsections.
3.2.2.5 Tie Rod Model Four shroud repair tie rods are installed in the downcomer region of the Hatch Unit 1 reactor. The tie rods are located at 45°, 135°, 225°, and 315°, which correspond to an azimuthal quadrant symmetric location of 45° (modeled).
The tie rod shaft was approximated as a solid cylinder, while the upper and lo wer mounting brackets were modeled as combinations of cylindrical and wedge e l ements. 3.2.2.6 Jet Pump Model There a r e ten jet pump assemblies in the downcomer region of Hatch Unit 1 which provide the main recirculation flow for the core. For the fast fluence model, the jet pumps are modeled at azimuths 30°, 60°, and 90° in the downcomer region. When symmetry is applied to the model, the 30° lo cation represents the jet pump assemblies that are posit ion ed azimuthally at 30°, 150°, 210°, a n d 330°; the 60° loc ation represents those at 60°, 1 20°, 2 40°, and 300°; and the 90° location represents the jet pump assemblies at 90° and 270°. The jet pump model includes representations for the riser, mixer, and diffuser pipes; nozzles; ram's head; riser inlet pipe; and riser brace yoke, l eaves, and pads. The jet pump assembly design is modeled using cylindrical elements for the jet pump riser and mixer pipes. The mixer nozzles, adapters, and diffusers are modeled as stepped shells to represent the axia ll y-varying radii. The riser pipe is correctly situated between the mixer pipes. The riser brace assemb l y model includes two lea f structures that attach to the yoke and pad elements. The jet pump 3-15 Neutron Flu e nce Calculation assembly includes hold down beams and brackets, built with rectangular elements, which are attached to the ram's head. 3.2.2. 7 Surveillance Capsule Model Section 3 .1 de scri bes the three surveillance capsules installed in the Hatch Unit 1 reactor. The capsules are positioned near the inner surface of the pressure vessel wall. The surveillance capsules are rectangular in design , and thus are not easily implemented in the otherwise cylindrical elements of the downcomer region model. It can be observed in Figure 3-1 that the capsules are of small dimensions in the planar geometry and they reside a long distance from the core region, yielding an effectively small view factor at that distance (viz., the view factor indicates any reasonably planar or lightly curved shape of a small width is indistinguishable).
Based on these factors, the otherwise rectangular shape of the surveillance capsules can be reasonably approximated in the model with arc elements.
The surveillance capsule model also includes a representation for the downcomer water that surrounds the capsule on all sides. The surveillance capsules are modeled behind the jet pump riser pipes at the 30° and 60° azi muths based on the as-built dimensions of the lower capsule mounting bracket. When symmetry is applied to the model, the 30° location represents the capsule installed at 30°, while the 60° location represents the capsules at 120° and 300°. The surveillance capsules are modeled at their correct axial position relative to the core region. The surveillance capsules cover approximately nine percent of the total core height. 3.2.2.8 Reactor Pressure Vessel Model The RPV and vessel cladding lie outside the downcomer region, with both modeled using cylindrical elements.
The cladding-pressure vessel interface is a key location for RPV fluence calculations and is preserved in the model. This interface defines the inside s urface (OT) for the pressure vesse l base metal where the RPV fluence is calculated.
Hatch Unit 1 ha s cladding only on the inside surface of the pressure vessel wall. Representations of the forgings for the recirculation inlet (N2) nozzles are also included in the model out to the biological shield radius. The no zz le representations are modeled in their true conical elements to preserve their basic design features.
It is noted that the Hatch Unit 1 RPV ha s two wall thicknesses.
The thickness variation occurs just above the nozzle forgings and is modeled as a stepped outer radius. This ste p was modeled as a discrete jump as opposed to the angular contour of the physical vessel. 3.2.2.9 Thermal Insulation Model The thermal insulation lies in the cavity region outside the pressure vessel wa ll. The insulation is composed of three layers: an inner cladding, a center insulating layer composed of an arra y of foils, and outer cladding. With the foil layer homogenized , the insulation is cylindrical in design and follows th e contour of the pressure vesse l wall. It is modeled wit h cylindrical elements. 3.2.2.10 Inner and Outer Cavity Models The cavity region lies between the pressure vessel and biological shie ld structures. As described in the previous subsection, the insulation lies in the cavity region, thus creating two cavity regions. The inner cavity region lies between the vessel and the in s ulation. The outer cavity 3-16 N e utron Flu e n ce Cal c ulation regio n lies between t he insulation and biological shie ld cladding.
The boundaries of the cavity regions follow the contours of the pressure vessel, vessel insu lati on, and biological shie ld. The cav i ty regions are m o deled with cylindrical elements.
3.2.2.11 Biological Shield Model The biological shield (concrete) defines the outer most region of the fluence model. The biological shield for the modeled elevation is cylindrical in design and is modeled with cylindrical elements. Cladding is modeled on the inside and outside surfaces of the biological shield. 3.2.2.12 Above-Core Component Models Figure 3-3 includes illustrations of other components and regions that lie above the reactor core regio n. The predominant above-core components represented in the model include the top guide, core s p ray spargers, and shroud head. 3.2.2.12.1 Top Guide The top guide component lies above the core region. The top guide is appropriately modeled by including representat i ons for the vertical fuel assembly parts and top guide plates. The upper fuel assem b ly parts that extend into the top guide region are modeled in three axial segments: the fue l rod plenum, fuel rod u pper end plugs, and fuel assembly upper tie plate. The fuel assembly parts and to p guide plates a re modeled with rectangu l ar elements.
3.2.2.1 2.2 Core Spra y Sparger The core spray spargers include upper and lower sparger annulus pipes and inlet piping. The core spray s parger annulus pipes are appropriately represented as torus structures in the model. The sparge r annulus pipes reside inside the upper shroud wall above the top guide. The spargers are modeled as pipe-like structures and include a representation ofreactor coolant inside the pipes. The model includes an accurate representation of the horizontal pipe extending from the penetration in the shro ud , and the vertical pipe extending upwards to the top of the model. 3.2.2.12.3 Shroud Head The fluence model also includes an accurate representation of the shroud head. It is modeled with spherica l elemen t s and includes the penetration and piping for the steam separator stand pipes. T he stand pipes are modeled with cylindrical elements.
3.2.2.13 Below-Core Component Models Figure 3-3 includes ill ust rations of other components and regions that lie below the reactor core region. The fuel support piece, core support plate, and core inlet regions appropriately include a represe n tation of the cruciform control rod below the core region. The lower fuel assemb l y parts include representations for the fuel rod lower end plugs, lower tie plate, and nose piece. The below-core components are modeled with rectangular elements with the exception of the core support plate. The core support plate is modeled using both rectangular and cylindrical elements to provide an appropriate representation of that component.
Core support plate rim bolts are not explicit l y modeled. 3-17 N e utron Flu e nce Calculation 3.2.2.14 Summary of the Geometry Modeling Approach To summarize the reactor modeling proce ss, there are several key features that allow the reactor de s ign to be accurately repr ese nted for RPV and capsule fluence evaluations.
Following is a s ummary of some of the key features of the model.
- Rectangular, cylindrical, spherical, conical, and wedge elements are mixed in the model in order to provide an accurate geometrical representation of the components and regions in the reactor.
- The reactor core geometry is modeled with rectangular elements to represent its actual shape in the reactor. The fuel assemblies in the core region are also sub-meshed with additional rectangular elements to represent the pin cell regions in the assemblies.
- A combination of rectangular and cylindrical elements are used to describe the transition parts between the rectangular core region and the cylindrical outer core regions.
- Cylindrical and wedge elements are us e d to model the components and regions that extend outward from the core region (core shroud, downcomer , RPV, etc.).
- The surveillance capsules are modeled at their as-built radial, azimuthal, and elevational po s itions behind the jet pumps in the downcomer region. The capsules are surrounded by water on all sides.
- The above-core region includes accurate representations of the top guide, core spray spargers, shroud head and steam separator stand pipes.
- The below-core region includes appropriate representations for the fuel support piece, core support plate, core inlet regions, cruciform control rods , and control rod guide tubes.
- The biological shield and cladding are appropriately represented as cylindrical elements , with cladding on the inside and outside surfaces of the concrete shield. 3.2.3 RAMA Calculation Parameters The RAMA transport code uses a three-dimensional deterministic transport method based on the Method of Characteristics (MOC) to calculate the neutron flux. The accuracy of the transport method is based on a numerical integration technique that uses ray-tracing to characterize the geo metry , anisotropy treatments to determine the directional flow of particles , and convergence parameters to determine the overall accuracy of the flux solution between iterations.
The code allows the user to specify values for each of these parameters.
The primary input parameters that control the ray-tracing calculation are the distanc e betwe en parallel rays in the planar and axial dimensions, the distance that a particle is tracked when a reflective boundary is encountered, and the number of equally spaced angles in polar coordinates for tracking the particles.
Plant-specific values are determined for each of the parameters.
The angular quadrature for determining ray trajectories is specified as S10, which provides an acceptable compromise between computational accuracy and performance.
The RAMA transport calculation employs a treatment for anisotropy that is based on a Legendre expansion of the scattering cross-sections.
By default , the RAMA transport calculation use s the maximum order of expansion that is available for each nuclide in the RAMA nuclear data 3-18 Ne utron Flu e n ce Cal c ulation library. For the actinide and zirconium nuclides, a P s expansion of the scattering cross-sections is used. For all other nuclides, a P 1 expansion of the scattering cross-sections is used. The overall accuracy of the neutron flux calculation is determined using an iterative technique to converge the flux iterations. The convergence criterion used in the evaluation was determined by parametric study to be 0.01, which provides an asymptotic solution for this model. 3.2.4 RAMA Neutron Source Calculation RAMA calculates a unique neutron source distribution for each transport calculation using the input relative power density factors for the fuel region and data from the RAMA nuclear data library. The source distribution changes with fuel burnup; thus, the source is determined using core-specific three-dimensional isotopics and power distributions at frequent intervals thro u ghout a cycle. For the fluence model, the peripheral fuel assemblies are modeled to preserve the power gradient a t the core edge that is formed from the pin-wise source distributions in these fuel a s semblies. 3.2.5 RAMA Fission Spectra RAMA calculates a weighted fission spectrum for each transport calculation that is based on the relative contributions of 23 5 U, 2 38 U , 2 39 Pu, 240 Pu, 2 41 Pu, and 2 4 2 Pu isotopes.
The fission spectra for these i sotopes are derived from the BUGLE-96 nuclear data library [20]. 3.2.6 Parametric Sensitivity Analyses Several sensitivity analyses were performed to evaluate the stability and accuracy of the RAMA transport calculation for the Hatch Unit 1 model. Mesh sizes and numerical integration parameters were among the items evaluated.
3.3 Surveillance
Capsule Activation and Fluence Results This section documents the fluence and activation results for the Hatch Unit 1 reactor. The activation resu lts also form the basis for the validation and qualification of the app l ication of the RAMA Fluence Methodology to Hatch Unit 1 in accordance with the requirements of U.S. NRC Regulatory Guide 1.190 [ 13]. Four flux wire activation analyses were performed for the Hatch Unit 1 reactor. Flux wires were remov e d from the 30° surveillance capsule flux wire holder and analyzed at the End of Cycle (EOC) 1 (irradiated for 1.2 EFPY); surveillance capsule flux wires were removed at EOC 8 from the 30° surveillance capsule (irradiated for 5.5 EFPY); surveillance capsule flux wires were removed at EOC 16 from the 120° surveillance capsule (irradiated for 14.2 EFPY); and surve illance capsule flux wires were removed at EOC 27 from the 300° surveillance capsule (irradiated for 32.0 EFPY). Details of the dosimetry specimens and ana l ysis are presented in Section 3.3.1. During the evaluation process, it was observed that the calculated activity underestimated the dosimetry measurements by for the Cycle 1 flux wires. As discussed in Section 3.1.4.3, simu l ator data was no t available for Cycles 1-15. The replacement operating data was constructed by taking representative fuel power shapes and mapping the shapes to the dominant Hatch Unit 1 fuel design of each cycle. This approach likely limits the accuracy of the early 3-1 9 N e utron Flu e nc e Cal c ulation cycle calculations.
However, by Cycle 8, an acceptable comparison developed, indicating that the approach remains valid. Excluding the Cycle 1 flux wires from the overall comparison yields a C/M ratio of 1.00 +/- 0.09. Best estimate fast fluence (E > 1. 0 Me V) was calculated for all removed capsules and the 3 0° survei ll ance capsule flux wire holder. Lead factors are determined and reported for all capsules. 3.3.1 Comparison of Predicted Activation to Plant-specific Measurements The comparison of predicted activation for the Hatch Unit 1 Cycle 1 , 8 , 16 , and 27 flux wires to measurements is presented in this subsection.
Fluence values are also calculated and reported in Section 3.3.2 for each of the capsule flux wires. 3.3.1.1 Cycle 1 30° Flux Wire Holder Activation Analysis Copper and iron flux wires were irradiated in the Hatch Unit 1 surveillance capsule flux wire holder at the 30° azimuth during the first cycle of operation.
The wires were removed after being irradiated for a total of 1.2 EFPY. Activation measurements were performed following irradiation for the following reactions
[10]: 63 Cu (n,a) 6°Co and 54 Fe (n , p) 54 Mn. The precise location of the individual wires within the surveillance capsule flux wire holder is not known; therefore, the activation calculations were performed at the center of the holder. Table 3-4 provides a comparison of the RAMA calculated specific activities and the measured specific activities for the flux wire specimens.
The Cycle 1 total flu x wire average measured (C/M) value is 0.78 with a standard deviation of +/-0.03. 3-20 N e utr o n Fl u e n ce Ca l c ul a tion T a ble 3-4 Comparison of Specific Activities for Hatch Un i t 1 Cycle 1 30° Flux W i re Holder Wires (C/M) Flux W i res Measured Calculated Calculated vs. Standard (dps/g) (dps/g) Measured (C/M) Deviat i on (a) Iron I r on 1 1.10E+05 8.98E+04 0.82 -Iron 2 1.11E+0 5 8.98E+0 4 0.81 -Iron 3 1.12E+0 5 8.98E+0 4 0.80 -Av e rage --0.81 0.01 Copper C opper1 5.3 6E+0 3 4.0 4 E+0 3 0.75 -C opper 2 5.4 6E+0 3 4.0 4 E+0 3 0.7 4 -C opper 3 5.4 0E+03 4.04E+03 0.7 5 -Av e rage --0.75 0.01 Tota l Flux Wire --0.78 0.03 Aver a ge 3.3.1.2 Cycle 8 30° Surveillance Capsule Activation Analysis Copper, iron, and nickel flux wires were irra d iated in the Hatch Unit 1 s ur vei ll ance caps ul e at the 30° azimuth d uring t h e fir st 8 c y cles of operation. The wire s were removed after being irradiate d for a t o tal of 5.5 EF P Y. Activation meas u reme n ts were performed following irra d iatio n for the fo ll owing react i ons [8): 63 Cu (n , a) 6°Co , 54 Fe (n , p) 5 4 Mn , and 58 Ni (n , p) 58 Co. T h e precise l ocation of t h e indiv i dual wires within t h e survei ll ance capsule is not known; therefore , t h e activa t i on calc ul ations wer e performe d at the ce n ter of t h e caps ul e ho ld er. Tab l e 3-5 provi d es a comparison of t h e RAMA calcu l ate d specific activ it ies an d the meas u red spec i fi c activ i t i es for t h e survei ll ance cap s ule flux w ire s pecime n s. The Cycle 8 cap s u l e total fl u x wire average C/M v alue is 0.94 wi th a sta nd ard d ev i ation of +/-0.09. 3-21 N e utron Flu e n c e Ca l c ulatio n Table 3-5 Comparison of Specific Activities for Hatch Unit 1 Cycle 8 30° Surveillance Capsule Flux Wires (C/M) Flux Wires Measured Calculated Calculated vs. Standard (dps/g) (dps/g) Measured (C/M) Deviation (a) Iron Iron G1 7.62E+04 7.83E+04 1.03 -Iron G2 7.94E+04 7.83E+04 0.99 -Iron G3 7.5 1E+04 7.83E+04 1.04 -Average --1.02 0.03 Nickel Nickel G1 1.21 E+06 1.21 E+06 1.00 -Nickel G2 1.23E+06 1.21 E+06 0.98 -Nickel G3 1.23E+06 1.21E+06 0.98 -Average --0.99 0.01 Copper Copper G1 1.10E+04 9.00E+03 0.82 -Copper G2 1.10E+04 9.00E+03 0.82 -Copper G3 1.06E+04 9.00E+03 0.85 -Average --0.83 0.02 Total Flux Wire 0.94 0.09 Average --3.3.1.3 Cycle 16 120° Surveillance Capsule Activation Analysis Copper, iron, and nickel flux wires were irradiated in the Hatch Unit 1 surveillance capsule at the 120° azimuth during the first 16 cycles of operation.
The wires were removed after being irradiated for a total of 14.2 EFPY. Activation measurements were performed following irradiation for the following reactions
[9]: 63 Cu (n , a) 6°Co, 54 Fe (n,p) 54 Mn, and 5 8 Ni (n,p) 5 8 Co. The precise location of th e in dividu a l w ires w ithin the surveillance capsule is not known; therefore, the activation calculations were performed at th e center of the capsule holder. Table 3-6 provides a comparison of the RAMA calculated specific activities and the measured specific activities for the surve illan ce capsule flux wire specimens.
The Cycle 16 capsule total flux wire average C/M value is 1.04 with a sta ndard de viation of +/-0.08. 3-22 Neutro n Flu e n ce Calculation T a ble 3-6 Comparison of Specific Activities for Hatch Unit 1 Cycle 16 120° Surveillance Capsule Flux Wires (C/M) Flux Wires Measured Calculated Calculated vs. Standard (dps/g) (dps/g) Measured (C/M) Deviation (cr) Iron 9.55E+04 1.01 E+05 1.06 -Nick e l 1.18E+06 1.21 E+06 1.11 -Copper 1.54E+04 1.47E+04 0.96 -Total Flux Wire 1.04 0.08 Average --3.3.1.4 Cycle 27 300° Surveillance Capsule Activation Analysis Copp e r , iron, and nickel flux wires were irradiated in the Hatch Unit 1 surveillance capsule at the 3 0 0° azimuth during the first 27 cycles of operation. The wires were removed after being irradiated for a total of 32.0 EFPY. Activation measurements were performed following irradiation for the following reactions (See Appendix A): 63 C u (n ,a) 6°C o , 54 Fe (n,p) 54 Mn , and 58 N i (n,p) 58 Co. The precise location of the individual wires within the surveillance capsule is not known; therefore , the activation calculations were performed at the center of the capsule holder. Tab le 3-7 provides a comparison of the RAMA calculated specific activities and the measured specific activities for the surveillance capsule flux wire specimens.
The Cycle 27 capsule total flux wire average C/M value is 1.05 with a standard deviation of +/-0.08. 3-23 Neutron F l uence Calculation Table 3-7 Comparison of Specific Activities for Hatch Unit 1 Cycle 27 300° Surveillance Capsule Flux Wires (C/M) Measured Calculated Calculated vs. Standard Flux Wires (dps/mg) (dps/mg) Measured (C/M) Deviation (a) Iron Iron G4 1.14E+02 1.19E+02 1.0 4 -Iron GS 1.14E+02 1.19E+02 1.04 -Average --1.04 0.00 N i ckel Nickel G4 1.4 0E+03 1.63E+03 1.1 7 -Nickel GS 1.SOE+03 1.63E+03 1.0 9 -Average --1.13 0.06 Copper Copper G4 2.0SE+01 2.00E+01 0.97 -Coppe r GS 2.07E+01 2.00E+01 0.96 -Average --0.97 0.01 Total Flux Wire 1.05 0.08 Average --3.3.1.5 Surveillance Capsule Activation Analysis Summary Ta ble 3-8 presents a s u mmary of th e to t a l average ca l c ul ated-to-m easured result of spec i fic activ iti es fo r a ll Hatc h Unit 1 fl u x wires. Com bin ing a ll fl u x wires (co pp er, iron, and n i ckel), the tota l average C/M is 0.94 wit h a stan d ard d eviat i on of +/-0.13 an d 1.00+/-0.09 ex cludi ng Cycle 1. Table 3-8 Comparison of Activities for Hatch Unit 1 Flux Wires Dosimeter Number of Calculated vs. Standard Deviation Measurements Measured (C/M) (a) 30° Flux W i re (EOC 1) 6 0.78 0.03 30° Capsule (EOC 8) 9 0.94 0.09 120° Capsule (EOC 16) 3 1.04 0.08 300° Capsule (EOC 27) 6 1.0S 0.08 Total 24 0.94 0.13 Total (Exe. EOC 1) 18 1.00 0.09 3-24 N e utron Flu e nce Calcu lati o n 3.3.2 Capsule Peak Fluence Calculations and Lead Factor Determinations Best-estimate fast neutron fluence was calculated for each of the surveillance capsules originally installed in the Hatc h Unit 1 reactor. Of the three original capsu l es, all have been removed. The fluen c e for the capsu l es i s reported at the time of their removal. Additionally, th e lead factor for each c apsule is calculated by dividing the peak c aps ule flu ence by the resp ect ive pe ak RPV fluen c e. The results of these calculations are pre sente d in Table 3-9. Table 3-9 Calculated Capsule Fast Neutron Fluence and Lead Factors for Hatch Unit 1 Capsule Fast RPV Peak Time of EFPY at Capsule Removal Removal Fluence Fast Fluence Lead Factor (n/cm 2) (n/cm 2) 30° EOCB 5.5 EFPY 2.33E+17 3.70E+17 0.63 120° EOC16 14.2 EFPY 5.79E+17 9.14E+17 0.63 300° E OC27 32.0 EFPY 1.38E+18 1.97E+18 0.70 3.4 Capsule Fluence Uncertainty Analysis This section presents the combined uncertainty analysis and bias determination for the Hatch U nit 1 capsule fluence eva luation. The combined uncertainty is comprised of the activation comp a risons de ve loped in Section 3.3 and an analytical uncertainty factor developed in this sec tio n. When combined, t hese components provide a ba sis for determinin g the overall uncertain ty (lcr) and b ias in the capsule fluence for this analysis. The r e quirement s for determining the combined unc erta inty and bias for li ght-water reactor fluenc e eva luation s a r e provided in Re g ulatory Guide 1.190. The method implem e n te d for determinin g the combined uncert ainty and bias for reactor component fluenc e is described in the RAMA Theory Man ual [22]. Re ga rding the determination of a bia s in the fluence, R eg ulatory Guide 1.190 provides that an adjustment to the calculated fluence for bias effects is needed if a stat isticall y significant bi as exists in the fluence computation.
T he re s ult s present e d in this section s ho w that the combined unc ertai nty (lcr) for the Hatch Unit 1 300° s u rvei llance caps ule flue nce evaluation is+/- 10.5% a nd that no adjustment for bia s effects i s required to the calculated capsule fluence reported in Section 3.3 of this r eport. The follo wi ng s ub sec t ion s describ e the comparison uncertainties , the det erminat ion of the analytic uncertainty , a nd the determination of the overall combined uncertainty and bias for the Hatch Unit 1 capsule fluen ce evaluation. 3.4. 1 Comparison Uncertainty Comp a rison uncert ainty factors are d etermined b y comparing calculated activities wit h activity measur eme nts. For capsule fluence eva luation s, two comparison uncertainty factors are considered
- an operating reactor comparison unc ertainty factor and a benchm a rk comparison uncertainty factor. The de term ination of a comparison unc erta inty factor ba se d on measurements involves the combination of two m eas urem e nt components.
One component is th e variat ion in 3-25 Neutro n Flu e nce Calculation the comparison of the calculated-to-measured (C/M) activity ratio and the other accounts for the uncertainty introduced by the measurement process. 3.4.1.1 Operating Reactor Comparison Uncertainty The operating reactor, or plant-specific, comparison uncertainty is determined by combining the standard deviation for the activity comparisons with the measurement uncertainty for the plantspecific activity measurements.
3.4.1.2 Benchmark Comparison Uncertainty The benchmark/simulator comparison uncertainty used in the Hatch Unit 1 uncertainty analysis is based on a set of industry standard simulation benchmark comparisons.
Specifically, the VENUS-3 and PCA RAMA benchmarking resu lt s [24] are included for comparison.
3.4.2 Analytic
Uncertainty The calculational models used for fluence analyses are comprised of numerous analytical parameters that have associated uncertainties in their values. The uncertainty in these parameters are tested for their contribution to the overall fluence uncertainty.
The uncertainty values for the geometry parameters are based upon uncertainties in the dimensional data used to construct the plant geometry model. The uncertainty values for the material parameters are based upon uncertainties in the material densities for the water and nuclear fuel materials and the compositional makeup of typical steel materials. The uncertainty val ues for the fission source parameters are based upon uncertainties in the fuel exposure and power factors for the fuel assemblies loaded on the core periphery.
The transport method used in the fluence analysis employs a fission source calculation that accounts for the relative contributions of the uranium and plutonium fissile isotopes in the fuel and the relative power density of the fuel in the reactor. Both fission source parameters are derived directly from information calculated by three-dimensional core sim ulator codes. The uncertainty values for the nuclear cross-section parameters are based upon uncertainties in the number densities for the predominant nuclides that make up the reactor materials.
The uncertainty parameters for the fluence model inputs are based upon geometry meshing and numerical integration parameters used in the neutron flux transport calculation.
The process for determining the geometry meshing and numerical integration parameters involves the sensitivity studies described in Section 3.2.6 and in the RAMA Procedures Manual [23]. 3.4.3 Combined Uncertainty The combined uncertainty for the capsule fluence evaluation is determined with a weighting function that combines the analytic, plant-specific comparison, and benchmark comparison uncertainty factors developed in Sections 3 .4.1 and 3 .4.2. Table 3-10 shows that the combined uncertainty (lcr) determined for the Hatch Unit 1 300° surveillance capsule fluence is 10.5% for energy E > 1.0 Me V and contains the uncertainty results for the other capsules.
As no bias term exists, it is not necessary to adjust the RAMA predicted 3-26 Ne utron Flu e nc e Cal c ulation capsu l e fluence in this analysis. The combined uncertainty for all capsules is within the limits prescribed in U. S. NRC Regulatory Guide 1.190 (:'.S 20% ). Table 3-10 Hatch Unit 1 Surveillance Capsule Combined Uncertainty for Energy >1.0 MeV Capsule Combined Bias 1 Combined Uncertainty (1 a) 30° 0.0% 10.4% 120° 0.0% 10.5% 300° 0.0% 10.5% 1. The bias terms are Jess than their constituent uncert a inty values , concluding that no statistically-significant bias exists. 3-27 4 CHARPY TEST DATA 4.1 Charpy Test Procedure Charpy impact tests were conducted in accordance with American Society for Testing and Mate r ials (ASTM) Standards El85-82 [3] and E23-02 [1 2]. The 1982 version of E185 has been reviewed and approved by the NRC for surveillance capsule testing applications.
This standard references ASTM E23. The tests were conducted using a Tinius Olsen Testing Machine Company, Inc. Model 84 impact test machine with a 300 ft-lb (406.75 J) energy capacity.
The Model 84 is equipped with a dial gage as we ll as the MPM optical encoder system for accurate absor b ed energy measurement.
The machine is also equipped with an instrumented striker , so a total of three independent measurements of the absorbed energy were made for every test. In all cases, the optical encoder measured energy was reported as the impact energy. The optica l encoder energy is much more accurate than the analog dial. The optical encoder can reso l ve the energy to within 0.0 4 ft-lbs (0.054 J), whereas, for the dial, the resolution is around 0.25 ft-lbs (0.34 J). The im pact e nergy was corrected for windage and friction for each test performed.
The velocity of the striker at impact was nominally 1 8 ft/s (5.49 mis). The MPM encoder system measures the exact impact ve loci ty for every test. Calibration of the machine was verifie d as specified in ASTM E23, and verification specimens were obtained from the National Institute for Standards and Techno lo gy (NIST) and tested in accor d ance wit h the standard.
The ASTM E23 procedure for specimen temperature control using an in-situ heating and cooling system was followed.
The advantage of using th e MPM in-situ heating and coo lin g technology is that each specimen is thermally conditioned right up to the instant of impact. Thermal losses associated w ith liqui d bath systems, such as those resulting from transfer of a specimen from a liquid bath to the test machine, are comp l etely eliminated.
Each specimen was held at the desired test temperature for at least 5 minutes prior to testing, and the fracture process zone temperature was held to w ith in +/-1.8°F (+/-1°C) up to the instant of strike. Precision calibrated tongs were used for specimen centering on the test machine. Latera l expansion (LE) was determined from measurements made with a lateral expansion gage. The lateral expansion gage was calibrated using precision gage blocks which are traceable to NIST. The percentage of shear fracture area was determined by integrating the ductile and brittle fracture areas using the MPM Digital Optical Comparator (DOC) image analysis system. As shown in Figure 4-1, each fracture surface image is captured, o utli ned to delineate the brittle area, and out lin ed to define the outer ductile fracture region. The DOC software then performs a pixel area integration and automatica ll y calculates the shear fracture area. This method for shear area d e t ermination is the most accurate method given in ASTM E23, and is far superior to the commonly used photograph comparison method. The number of Charpy specimens for measurement of the transition regio n and upper she l f was limited. Therefore, the choice of test temperatures was very important.
Prior to testing, the 4-1 Charpy T e st Data Charpy energy-temperature curve was predicted using embrittlement models and previous data. The first test was then conducted near the middle of the transition region , and test temperature decisions were then made based on the test results. Overall, the goal was to perform two or three tests on the upper she lf , and to use the remaining specimens to characterize the 30 ft-lb (41 J) index. This approach was successful and the transition region and upper shelf energy are well defined. Figure 4-1 Illustration of Digital Optical Comparator Measurement of Shear Fracture Area Fir s t , th e Brittl e F rac tur e Ar ea is Outlin e d (with in g r e en li n e). N ex t , th e Out e r Du ct il e Fractur e Ar e a i s Outlined (wi thin r e d lin e). Finall y, t h e S o ftw are Int eg rat es th e Are as a nd Cal c ul a t es the P e r ce nt Sh ear Fra c tur e Area. 4.2 Charpy Test Data for the 300° Capsule A total of eight irradiated base, eight irradiated weld, and eight irradiated HAZ metal specimens were tested over the transition region temperature range and on the upper shelf. The data are summarize d in Tables 4-1through4-3.
In addition to the energy absorbed by the specimen during impact , the measured lat eral expansion values and the percentage shear fracture area for each test specimen are listed in the tables. The Charpy energy was acquired from the optical encoder signa l and has been corrected for windage an d friction in accordance w ith ASTM E23. The impact energy is the energy required to initiate and propagate a crack in the Charpy spec im en. The optical encoder and the dial cannot correct for tossing energy or losses in the test machine, and therefore this small amount of additiona l energy, if present, may be included in the 4-2 Charpy T e st Data data for some tests. The instrumented striker energy does not include tossing energy or machine vibration energy since the energy, in this case, is measured only during a few milliseconds of contact between the striker and specimen.
Based on comparison between the instrumented striker energy and the optical encoder energy, it has been shown that the tossing energy, and other losses, are small for most tests. The l ateral expansion is a measure of the transverse plastic deformation produced by the contact edge of the striker during the impact event. Lateral expansion is determined by measuring the maximum change o f specimen thickness along the sides of the specimen.
Lateral expansion is a measure of the ductility of the specimen.
In accordance with ASTM E23, the lateral expansion for some specimens , which could be broken after the impact test, should not be reported as broken since the lateral expansion of the unbroken specimen is less than that for the broken specimen.
Therefore, when these conditions exist, the value listed is the unbroken measurement and a footnote is included to identify these specimens.
All of the 300° surveillance capsule specimens that did not separate during the test could be broken by hand under the ASTM E23 requirements.
The percentage of shear fracture area is a direct quantification of the transition in the fracture mode s as the temperature increases. All metals with a body centered cubic (BCC) lattice structure, such as ferritic pressure vessel materials , undergo a transition in fracture modes. At low test temperatures , a crack propagates in a brittle manner and cleaves across the grains. As the temperature increases, the percentage of shear (or ductile) fracture increases. This temperature range is referred to as the transition region and the fracture process is mixed mode. As the temperature increases further, the fracture process is eventually completely ductile (i.e., no brittle component) and this t emperature range is referred to as the upper shelf region. 4-3 C h a rp y T est D ata 4-4 Table 4-1 Irradiated Charpy V-Notch Impact Test Results for Surveillance Base Metal Specimens (Heat C4114-2) from the Hatch Unit 1 300° Surveillance Capsule Base Irradiated
- Heat C4114-2, Longitudinal , 300° Capsule Test Lateral Specimen Temperature Impact Energy Expansion Percent Shear ID OF (oC) ft-lb (J) m i ls (mm) (%) 018 -48.8 (-44.9) 9.1 6 (1 2.4) 6.6 (0.17) 7.4 CY2 -1.5 (-18.6) 17.6 8 (24.0) 15.6 (0.4 0) 15.8 cuu 17.4 (-8.1) 38.20 (5 1.8) 30.7 (0.78) 22.1 CTU 36.9 (2.7) 47.20 (6 4.0) 3 3.1 (0.8 4) 25.9 CU 4 71.6 (22.0) 64.15 (87.0) 44.0 (1.1 2) 39.7 CY6 119.7 (48.7) 117.5 2 (159.3) 7 1.5 (1.82) 76.4 017 242.4 (116.9) 147.10 (199.4) 85.0 (2.16) 100.0 CUK 372.0 (188.9) 147.59 (200.1) 8 2.8 (2.10) 100.0 Table 4-2 Irradiated Charpy V-Notch Impact Test Results for Surveillance Weld Metal Specimens (Heat 1 P3571) from the Hatch Unit 1 300° Surveillance Capsule Weld Irradiated
- 300° Capsule Test Late r al Specimen Temperature Impact Energy Expansion Percent Shear ID OF (oC) ft-lb (J) mils (mm) (%) OOK 72.3 (22.4) 8.13 (11.0) 6.9 (0.18) 11.4 045 107.4 (41.9) 2 2.95 (31.1) 20.5 (0.52) 17.6 010 127.2 (52.9) 42.1 3 (57.1) 2 6.6 (0.68) 3 9.2 030 150.3 (65.7) 36.83 (4 9.9) 29.4 (0.75) 3 7.0 02K 201.9 (9 4.4) 54.68 (74.1) 4 1.5 (1.05) 5 6.6 02L 247.8 (119.9) 80.3 4 (108.9) 54.3 (1.3 8) 77.6 OOL 314.8 (157.1) 90.0 2 (1 22.0) 62.2 (1.58) 1 0 0.0 032 37 1.3 (18 8.5) 8 3.2 9 (11 2.9) 50.0 (1.27) 100.0 C h arpy T e s t Dat a Table 4-3 I r radiated Charpy V-Notch Impact Test Results for Surveillance HAZ Metal Specimens from t h e Hatch Unit 1 300° Surveillance Capsule HAZ Irradiated:
Longitudinal, 300° Capsule Test Lateral Specimen Temperature Impact Energy Expansion Percent Shear ID OF (oC) ft-lb (J) mils (mm) (%) D6A -61.6 (-52.0) 10.50 (14.24) 8.3 (0.21) 14.2 D6P -1.7 (-18.7) 31.06 (4 2.1 1) 25.8 (0.66) 32.3 D6L 72.7 (22.6) 28.39 (38.49) 25.0 (0.64) 34.3 D4B 1 17.7 (47.6) 80.39 (108.99) 57.5 (1.46) 64.0 D4C 1 52.6 (67.0) 103.68 (140.57) 66.1 (1.68) 77.8 DEY 243.5 (117.5) 78.21 (106.04) 61.6 (1.56) 100.0 D5P 334.2 (167.9) 96.80 (131.24) 70.1 (1.78) 100.0 D4Y 373.5 (189.7) 123.67 (167.67) 82.6 (2.10) 100.0 4-5 5 CHARPY TEST RESULTS 5.1 Analysis of I mpact Test Results For a n alysis of the Charpy test data , the BWRVIP ISP has selected the hyperbolic tangent (tanh) funct i on as t h e statistical curve-fit too l to model the transition temperature toughness data. A hyperbolic tangent curve-fitting program named CVGRAPH [11] was used to fit the C ha rpy V-notch (CVN) ene r gy and lateral expansion data. Analysis methodology (e.g., definition of upper fixed shelf and lower shelf) followed the BWRVIP conventions established for analysis of all ISP data [25]. The impact energy curve-fit from CVGRAPH for surveillance plate heat C4114-2 is provided in Figure 5-1 and the lateral expansion curve-fit is provided in Figure 5-2. As discussed in Sect i on 1.0, the surveillance weld from Hatch Unit 1 is not used in the ISP and unirradiated (baseline)
Charpy data is not available.
Therefore, the Charpy data for the surveillance weld from the 300° surveillance capsule were not fit. HAZ results are not used in the BWRVIP ISP; th u s, the HAZ data were also not fit. For the analysis of Charpy energy test data , lower she l f energy was fixed at 2.5 ft-lbs (3.4 J). Upper shelf energy was fixed at the average of all test energies exhibiting shear greater than or equal to 95%, consistent with ASTM Standard E185-82 [3]. For analysis of the lateral expansion test data, the lower shelf was fixed at 1.0 mils; the fixed upper shelf was defined as the average of t h e l ateral expansion test data points exhibiting shear greater than or equal to 95%, consistent with the approach used for upper shelf energy. 5.2 Irradiated Versus Unirradiated CVN Properties Tab le 5-1 summarizes the T3o [30 ft-lb (41 J) Transition Temperature], T3smiI [35 mil (0.89 mm) Latera l Expansion Temperature], T so [50 ft-lb (68 J) Transition Temperature], and Upper Shelf Energy for the unirradiated and irradiated materials and shows the change (shift) from baseline values. The unirradiated values of T 3o and Ts o were taken from the CV GRAPH fit provided in Figure 2-5; the unirradiated va lu e of T3 s miI was taken from the CV GRAPH fit provided in Figure 2-6. The irradiated values are from the inde x temperatures determined in Figures 5-1 and 5-2. Table 5-2 provides a comparison of the measured T3o shift to the predicted shift for plate heat C4114-2. Predicted shift is based on the formula provided in Regulatory Position 1.1 of Reg. Gui d e 1.99 , Rev. 2 [6] as shown in Note 3 to Tab l e 5-2. The fluence was input as 1.38 x 10 1 8 n/cm 2 , as reported in Ta ble 3-9 for the 300° surveillance capsule. The measured shift for the survei ll ance plate is greater than the va lue expected (e.g., the measured shift is greater than predicted shift+ margin). Measured percent decrease in USE is presented in Tab l e 5-3 and compared to the percent decrease predicted by Regulatory Position 1.2 and Figure 2 of Reg. Guide 1.99, Rev. 2. The measured percent decrease in USE for the survei ll ance plate is less than the predicted percent decrease.
5-1 C h arpy T est R esu l ts IRR A DIAT E D PL A TE H EA T C411 4-2(H A1-300) CVG r a ph 6.02: H yperbol i c T a n ge nt Curve Print ed o n 5/12/20172
- 1 6 PM A= 74.92 B = 72.42 C = 80.28 TO= 7 3.5 8 D = 0.0 0 Corre l a ti o n Coeffic i e nt= 0.996 E qu a ti on i s A+ B * [Tan h ((T-TO)/(C+Dn)J U p per S h elf E n e r gy= 14 7.35 (Fixed) Lowe r S h e lf Ene r gy= 2.50 (FL-..:ed)
T e m p@J O ft-lb s= 1 5.40° F Te mp@.3 5 ft-lb s= 23.80° F T em p@5 0 ft-l b s= 44.80° F 5-2 Pl an t: Hatc h 1 Ori en t a tion: L T ,-..... .c -I ¢:: -;;..... i... Q,) = z ;;;;... u 160 140 -120 .... 100 .... 80 -60 .... 40 .... 20 3 00 CVGrn ph 6.02 Figure 5-1 -200 Mate ri al: S A533Bl Ca p su l e: 300 DEG I l / k I 7 n .,..--100 0 100 2 00 3 00 Tempera t ure(° F) 05/12/2017 ..... -. . . . 4 00 Irradiated Plate Heat C4114-2 Charpy Energy Plot (Hatch Unit 1 300° Capsule) (LT) Heat C4114-2 Flu ence: n/a 500 600 Page 1/2 Charpy T es t R es ult s Plant: Ha tch 1 Orie ntati o n: LT Material: SA533Bl Ca p s ule: 300 DEG Heat: C4114-2 F lu e nce: nla IRRADIATED PLATE HEAT C4114-2 (HAl-300)
Charpy V-Notch Data Temperat ur e{° F) Input CVN Co mput e d CVN Differential
-49 9.2 9.1 0.10 -2 17.7 21.8 -4.1 6 17 38.2 31.2 7.04 37 47.2 44.0 3.2 4 72 6 4.2 73.1 -8.99 12 0 1175 1 12.5 5.03 242 147.1 145.2 1.88 372 147.6 147.3 0.33 CVGrap h 6.02 05 1 1 2120 17 Page 212 Figure 5-1 (Continued)
Irradiated Plate Heat C4114-2 Charpy Energy Plot (Hatch Unit 1 300° Capsule) (LT) 5-3 C h a r py T est R es ult s P l an!: Ha t c h 1 Ori e n ta tio n: LT ..-rl.> -*-6 '-' = 0 *-rl.> = -i.. ... 90 70 60 50 40 30 .... 20 .... 10 .... 0 -300 CV G nt ph 6.02 Figure 5-2 IRRADIATED PLATE HEAT C4114-2 LE (HAt-300) -200 CV G ra ph 6.02: H yperbo li c T a n ge nt Curve Print ed o n 5/1 2/20 1 7 2: 23 PM A= 4 2.45 B = 41.45 C = 86.4 0 TO = 56.94 D = 0.00 Co rrcl a tjon Coeffic i e nt = 0. 99 3 E qu a ti on is A+ B * [Tanh((T-TO)/(C+0 1))] U p pe r S h elf LE.= 8 3.90 (F i xed) L owe r S h e lf L.E. = 1.00 (Fi.-..:ed) T e mp.@3 5 m il s= 4 1.3 0° F Ma t e ri a l: SA533 Bl Ca psu le: 300 D EG / I I fo j I 0 ,,,,,----100 0 100 200 300 Temperature
{° F) 05/l 2/20 1 7 u 400 H ea t C4114-2 F lu en ce: nla 500 600 Page 112 Ir r adiated Plate Heat C4114-2 Lateral Expansion P l ot (Hatch Unit 1 300° Capsule) (LT) 5-4 Chmpy T est R es ults Plant: Hatc h 1 Orie n ta ti o n: LT Ma teri a l: SA533Bl Ca p s ule: 300 DEG Heat: C4114-2 F luen ce: n/a IRRADIATED PLATE HEAT C4114-2 LE (HAl-300)
Charpy V-Notch Data Temperature(° F) InputL E. Computed LE. Differential
-49 6.6 7.6 -1.00 -2 1 5.6 1 8.0 -2.43 17 3 0.7 2 4.7 6.0 0 37 33.1 33.0 0.10 72 44.0 4 9.4 -5.41 1 20 715 68.2 3.32 242 85.0 82.8 2.22 3 72 82.8 83.8 -1.0 4 C VGraph 6.02 0511 2/2 017 Page 212 Figure5-2 (Continued)
Irradiated Plate Heat C4114-2 Lateral Expansion Plot (Hatch Unit 1 300° Capsule) (LT) 5-5 C h a r py T e st R e sults 5-6 Table 5-1 Effect of Irradiation (E>1.0 MeV) on the Notch Toughness Properties T30* 30 ft-lb (40.7 J) T 50* 50 ft-lb (67 .8 J) T 3 5 mil* 35 mil CVN Upper Shelf Energy Transition Temperature Transition Temperature (0.89 mm) Lateral (USE) Material Expansion Temperature Identity Unirrad Irradiated ilT30 Unirrad Irradiated ilT50 Unirrad Irradiated ilT35mil Unirrad Irradiated Change OF OF OF OF OF OF OF OF OF ft-lb ft-lb ft-lb (oC) (oC) (oC) (oC) (oC) (oC) (oC) (oC) (oC) (J) (J) (J) C4114-2 -61.5 15.4 76.9 -28.8 44.8 73.6 -34.7 41.3 76.0 136.0 147.4 11.4 (LT orientation) (-51.9) (-9.2) (42.7) (-33.8) (7.1) (40.9) (-37.1) (5.2) (42.2) (184.4) (199.8) (15.4) Table 5-2 Comparison of Actual Versus Predicted Embrittlement Fluence RG 1.99 Rev. 2 RG 1.99 Rev. 2 Measured Shift 2 Predicted Ident i ty Material (E>1.0 MeV , x10 17 O F (OC) Predicted Shift 3 Shift+Margin 3*4 n/cm 2)1 OF (OC) OF (OC) C411 4-2 Hat c h U ni t 1 su rv e i lla n ce plate 1 3.8 76.9 (42.7) 4 0.9 (22.7) 7 4.9(41.6) (LT or i entat i on) l. Fluence value i s reported in Table 3-9. 2. The measured shift is taken from Tab l e 5-1. 3. Predicted shift= CF x FF, where CF is a Chemistry Factor taken from the base metal table in USNRC RG 1.99 , Rev. 2 [6], based on each material's Cu/N i co n te n t, and FF is F lu ence Factor, f0.28-0. l 0 log f, where f = tluence i n un i ts of 10 1 9 n/cm 2 (E > 1.0 MeV) specified. 4. Margin= 2./(G;2 + G t.2), where G; =the standard deviation on initial RT N D T (G; is taken to be 0°F), and G t. is the standard deviation on LlRT NDT (28°F for welds and l 7°F for base materials, except that G t. need not exceed 0.50 times the mean value of ilRT N o T). Thus, margin is defined as 34°F for plate materia l s and 56°F for we l d materials , or m argin equa l s shift (whichever is l ess), per Reg. Gui d e 1.99, Rev. 2.
Charpy Test R e sults Table 5-3 Percent Decrease In Upper Shelf Energy Fluence Measured Decrease in USE Predicted Decrease in USE 2 Identity Material (E>1.0 MeV , (%) (%) x10 1 7 n/cm 2)1 C 4 114-2 Hatch Un i t 1 3 13.8 --1 3.1 (LT orie n tation) surveillance plate 1. Fluence value is reported in Tab l e 3-9. 2. Based on the equations for Figure 2 of Reg. Guide 1.99, Rev. 2 [6] as provided in Reg. Guide 1.162 [26). 3. Value less than zero. 5-7 6 REFERENCES
- 1. 10 CFR 50 , Ap p endices G (Fracture Toughness Requir e ment s) and H (Reactor Vessel Material Surveillance Program R equire ments), Federal Register , Volume 60 , No. 243 , dated December 19, 1995. 2. American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),Section XI, " Ru l es for In service Inspection of Nuclear Power Plant Components," Nonmandatory Appendix G, Fracture Toughness Criteria for Protection Against Failure. 3. ASTM E185-82, Standard Practice for Conducting Surveillance T ests for Light-Wat er Cooled Nuclear Pow er R eactor Vessels, E706 (IF), American Society for Testing and Materia l s, Philadelphia , PA , 1982. 4. BWRVIP-86 , Revision 1-A: BWR Vessel and Int erna ls Project , Updated BWR Integrated Surv e illanc e Program (ISP) Impl e m enta tion Plan. EPRI, Palo Alto, CA: 2012. 1025144. 5. 10 CFR 50, Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," August 28, 2007. 6. U.S. NRC Regulatory Guide 1.99 , " Radiation Embrittlement of Reactor Vessel Materials ," Revision 2, May 1988. 7. "Guideline for the Management of Materials Issues," NEI 03-08 , Nuclear Energy Institute, Washington, DC, Latest Edition. 8. "Edwin I. Hatch Nuclear Power Plant , Unit 1 , Reactor Pressure Vessel Surveillance Materia l s Testing and Fracture Toughness Analysis" , T. A. Caine, GE Nuclear Ene rgy , NEDC-30997 DRF Bl 1-00313 Class I, October, 1985. 9. "Plant Hatch Unit 1 RPV Surveillance Materials Testing and Analysis" , B. D. Frew, GE Nuclear Energy , GE-NE-Bl 100691-0lRl, March , 1997. 10. "Determination of Fast Neutron Flux Density and Neutron Fluence , Hatch 1 Power Station ," G.C. Martin, GE Nuclear Energy, GE Report 266-7801-02, January 26, 1978. 11. CV GRAPH, Hyperbolic Tangent Curve Fitting Program , Developed by A TI Consulting, Version 6.02, Apri l 2014. 12. ASTM Standard E23, Standard T est Methods for Notch Bar Impact Testing of Metallic Materials , ASTM International , West Conshohocken, PA, www.astm.org. 13. U.S. NRC Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," March 2001. 14. BWRVIP-126 , R ev ision 2: BWR Vessel Int e rnals Proje ct, RAMA Fluence Methodology Softwar e, Version 1.20. EPRI, Palo Alto , CA: 2010. 1020240. 6-1 R efe r e n ce s 15. L etter from William H. Bateman (U.S. NRC) to Bill Eaton (BWRVIP), Safety Evaluation of Proprietary EPRl Reports BWRVIP-114, -115, -117, and -121 and TWE-PSE-001-R-001 , dated May 13 , 2005. 16. " Safety Evaluation Report with Open Items Related to the License Renewal of Seabrook Station," Docket Number 50-443 , dated June 2012. 17. Letter from Matthew A. Mitchell (U.S. NRC) to Rick Libra (BWRVIP), " Safety Evaluation of Proprietary E PRl Report BWR Vessel and Int e rnals Project, E v aluation of Susquehanna Unit 2 Top Guide and Core Shroud Material Samples Using RAMA Fluence Methodology (BWRVIP-145
)," dated February 7 , 2008. 18. " Core Design and Operating Data for Cycle 1 o f Hatch l ," EPRl, E PRl-NP-562, January 1979. 19. " Core Design and Operating Data for Cycl es 2 and 3 of Hatch 1 ," E PRl, E PRl-NP-562 , February 1984. 20. " BUGLE-96:
C oupled 47 Neutron , 20 Gamma-Ray Group Cros s Section Library Deriv e d from ENDF/B-VI for LWR Shielding and Pressur e Vessel Dosimetry Applications
," RSIC C Data Library Collection , DL2C-185 , March 1996. 21. " VITAMIN-B6:
A Fine-Group Cross Section Library Based on E NDF/B-VI Release 3 for Radiation Transport Applications
," RSICC Data Library Collection , DLC-184, December 1996. 22. BWRVIP-114-A: BWR Ve ssel and Int e rn a ls Proj ec t , RAMA Flu e n ce M e th o dolo gy Th e o ry Manual , EPRl , Palo Alto , CA: 2009. 1019049. 23. BWR VI P-121-A: BWR Ve s se l and Int e rnals Proj ec t , RAMA Flu e n ce M e thodolo gy Proc e dur es M a nual , EPRl , Palo Alto , CA: 2009. 1019052. 24. BWRVIP-115-A: BWR V e ss e l and Int e rn a l s Proj ec t , RAMA Flu e n ce M e th o dolo gy B e nchmark M a nual -E v aluation of R eg ulato ry Guid e 1.190 B e n c hmark Probl e ms , EPRl , Palo Alto , CA: 2009. 1019050. 25. BWRVIP-135 , R e vision 3: BWR V es s e l and Int e rnals Project , In teg rat e d Surv e illanc e Program (ISP) Data Sour ce Book and Plant Evaluation
- s. EPRl , Palo Alto , CA: 2014. 3002003144.
- 26. U.S. NRC Re g ulatory Guide 1.162 , " Format and Content of Report for Thermal Annealin g of Reactor Pre ss ure Ve ss els ," February 1996. 6-2 A DOSIMETER ANA L YSIS A.1 Dos i meter Material D e scription The Hatch Unit 1 300° surveillance capsule dosimeter materia l s are pure meta l wires which were located within the surveillance capsu l e along t h e ends of the Charpy specimens.
The wire types prov i ded for t h e Ha t c h Unit 1 surveillance program are iron , n i ckel, and copper. Each wire is nominally three inches (7.62 cm) long. Further discussion of the dosimeter cleanin g and mass measurements follows. A.2 Dosimeter Cleaning and Mass Measurem en t At the time the surveillance capsule Charpy packets were opened, the dosimeter wires were clean e d in an ultrasonic cleaner in an acetone bath and were wiped with acetone wetted wipes to remo v e loose contamination. Upon receipt at the radiometric laboratory, the wires were visually inspected under a low magnification optical microscope.
There was evidence of oxidation indic a ting the need for chemical etching and further cleaning. This was accomplished by soaking the Fe wire segments in a 4N solution of hydrochloric acid until the oxidation was etched from the surface. Similarly , the Cu and Ni wires were immersed in a 2N solution of nitric acid. The wires were then rinsed with distilled water, wiped once more with ethanol, and then allowed to dry in air at room temperature.
The wires then exhibited a clean, shiny appearance. Figures A-1 throu g h A-6 show low-power magnifications of the dosimetry wires as they were found prior to cleani n g, and after cleaning and coiling. Although the corro s ive attack on the 04 packet wire was si g nificant, there was sufficient wire length availab l e to perform counting under the standard proce d ures. The total mass of eac h wire was measured using a Mettler Toledo XS 105DU analytical digital balanc e. Tab l e A-1 lists the results of these measurements , as well as the identification assigned to each dosimeter.
The dosimeter identifications were assigned as the packet ID containing the dosimeter wire and type of dosimeter material.
As pre v iously mentioned , the wires were tightly coiled for subsequent counting and weighing.
Each wire was wrapped around a thin metal rod to form a coil of approximately 0.5 inch (12.7 mm) diameter or less, which yields a good approximation to a point source geometry at the distanc e the dosimeter wires are placed from the gamma detector.
The coi l ed wire segments were presse d firmly against a hard surface to flatten t h e coi l to yield t h e best counting geometry.
A.3 Radiometric Analysis Radiometric analysis was performed using high resolution gam ma emission spectroscopy. In this method , gamma emiss i ons from the dosimeter materials are detected and quantified using solidstate gamma ray detectors and computer-based signal processing and spectrum analysis. The specifications of the gamma ray spectrometer system (GRSS) are l isted in Table A-2. The GRSS A-1 Dosimeter Anal y sis features a hyper pure germanium (HPGe) detector that is housed in a lead-copper shield to reduce background count rates. Standard background subtraction procedures were used. GRSS calibration was performed using a National Institute for Standards and Technology (NIST) traceable mixed gamma quasi-point source. The Canberra analysis software provides the capability for energy resolution and efficiency calibration using specified standard source information.
Calibration information is stored on magnetic disk for use by the spectrographic analysis software package. Since detector efficiency depends on the source-detector geometry, a fixed-reproducible geometry must be selected for the gamma spectrographic analysis of the dosimeter materials.
For the dosimeter wires , the counting geometry was that of a quasi-point source (coiled wire) placed five inches (12. 7 cm) vertically from the top surface of the detector shell. In this way, extended sources up to 0.5 inch (1.27 cm) can be analyzed with a good approximation to a point source. The coiled wires were well within the area needed to approximate a point source geometry.
The HPGe detector was calibrated for efficiency using the NIST traceable source. The accuracy of the efficiency calibration was checked using a gamma spectrographic analysis of the NIST traceable mixed gamma source. The isotopes contained in the source emit gamma rays which span the energy response of the detector for the dosimeter materials.
These measurements show that the efficiency calibration is providing a valid measurement of source activity.
The acceptance criteria for these measurements are that the software must yield a valid isotopic identification, and that the quantified activity of each correctly identified isotope must be within the uncertainty specified in the source certification.
Validation of system performance was made prior to starting the counting tasks , and upon completion of all counting work for Hatch Unit 1. The counting system performance was acceptable in each case, indicating that the counting system properties did not change during the course of the counting procedure.
Table A-3 shows the counting schedule established for this work. There was no requirement for order of counting since the dosimeter materials still contained sufficient quantities of activation products to allow accurate radio assay. Counting times were more than sufficient to achieve the desired statistical accuracy for gamma emissions of interest in all cases. Neutrons interact with the constituent nuclei of the dosimeter materials producing radionuclides in varying amounts depending on total neutron fluence, its energy spectrum, and the nuclear properties of the dosimeter materials.
Table A-4 lists the reactions of interest and their resultant radionuclide products for each element contained in the dosimeters.
These are threshold reactions involving an n-p or n-a interaction.
Finally , Table A-5 presents the primary results of interest for flux and fluence determination.
The specific activity units are in dps/mg, which normalizes the activity to dosimeter mass. The activities are specified for a useful reference date/time , which in this case is the Hatch Unit 1 plant shutdown date and time. This reference date/time was specified as February 8 , 2016, at 012:01:01 AM eastern standard time. A-2 Dosimeter A nal ysis G4Fe G4Fe Figure A-1 Hatch Unit 1 300° Capsule Packet G4 Fe Dosimeter Wire G4 Fe: Prior to Cleaning (left); and After Cleaning/Coiling (right) G4Cu G4Cu Fi g ure A-2 Ha t ch Unit 1 300° Capsule Packet G4 Cu Dosimeter Wire G4 Cu: Prior to Cleaning (left); and After Cleaning/Coiling (right) G4Ni G4Ni Figure A-3 Hatch Unit 1 300° Capsule Packet G4 Ni Dosimeter Wire G4 Ni: Prior to Cleaning (left); and After Cleaning/Coiling (right) A-3 Dosim eter Analysis G5Fe G5Fe Figure A-4 Hatch Unit 1 300° Capsule Packet GS Fe Dosimeter Wire GS Fe: Prior to Cleaning (left); and After Cleaning/Coiling (right) G5Cu G5Cu Figure A-S Hatch Unit 1 300° Capsule Packet GS Cu Dosimeter Wire GS Cu: Prior to Cleaning (left); and After Cleaning/Coiling (right) G5Ni G5Ni Figure A-6 Hatch Unit 1 300° Capsule Packet GS Ni Dosimeter Wire GS Ni: Prior to Cleaning (left); and After Cleaning/Coiling (right) A-4 I I I Dosi m eter A n a l ysis Table A-1 Hatch Unit 1 300° Capsule Charpy Packet Dosimeter Wire Masses Wire Dosi m eter ID Mass (mg) G4 Fe 149.17 G4 Cu 405.37 G4Ni 275.83 G5 Fe 162.60 G5 Cu 396.50 G5 Ni 298.71 Table A-2 G a mma Ray Spectrometer System (GRSS) Specifications System Component Description and/or Specifications Detector Canberra Model BE3830 Energy Resolution
<1.9 keV FWHM @ 1.33 MeV Detector Efficiency Relative to a 3 inch x 3 inch 33.3% at 1.3 MeV Nal Crystal Amplifier/Multichannel Analyzer Canberra DAS-1000 Computer System Intel i5-4460 CPU at 3.20 GHz , 16 GB Main Memory , 931 GB Hard Disk , 23-inch Monitor , HP LaserJet Printer S oftware Canberra Apex v 1.4 A-5 Dosimet e r Analysis Table A-3 Counting Schedule for Hatch Unit 1 300° Capsule Dosimeter Materials Dosimeter ID Count Start Date Count Start Time (EST) Count Duration (Live Time Seconds) G4 Fe 11/4/2016 2:40:44 PM 86,400 G4Cu 1 1/7/2016 6: S4: 00 AM 86,400 G4Ni 11/8/2016 8: 24: 1S AM 86,400 GS Fe 1 1/9/2016 9:4 0: S4 AM 86,400 GS Cu 11/10/2016 3: 21: 1 3PM 86 , 400 GS Ni 1 1/11/2016 S:3 7: 22 PM 86,400 Table A-4 Neutron-Induced Reactions of Interest Dosimeter Material Neutron-Induced Reaction Reaction Product Radionuclide I ron Fe 54 (n , p)Mn 54 Mn 54 Copper Cu 63 (n , a)Co 60 Co 6o Ni ckel Ni sa(n , p)Co sa co sa A-6 Dosi m e t er Ana l ysis Table A-5 Results of Hatch Unit 1 300° Capsule Radiometric Analysis Activity at Specific Activity at Activity Dosimeter ID Isotope ID Reference Reference Uncertainty Date/Timea
(µCi) Date/Time 1 (dps/mg) (%) G4 Fe 54 Mn 4.S9E-01 113.8S 2.20 G4Cu 6o co 2.2SE-01 20.S4 1.70 G4 Ni sa co 1.04E+01 139S.06 2.27 GS Fe 54 M n 4.99E-01 113.SS 2.20 GS Cu 6o co 2.22E-01 20.72 1.70 GS N i sa co 1.21 E+01 1498.78 2.27 1 Feb ru ary 8, 2016 at 12: 01 : 01 AM EST is the refe r ence date and time. A-7 Export Control Restrictions Access to and use of EPRI Int ellectua l Property is granted wi th the sci f i c understanding and requirement tha t re s ponsibility For ensuring full co mpl iance with all applicable U.S. and foreign export laws and lations is bei ng undertaken by you and your company. Thi s includes an ob lig ation to ensure that any individual receiving access hereunder who is not a U.S. citize n or permanent U.S. resident is per mitt ed ac c ess under applicable U.S. and foreign export laws and regulations. In the e v ent you are uncertain whether yo u or your company may la wfu ll y obtain a c cess to this EPRI Int el l ectua l Property , you acknowledge that it is your obligation to consult with your company's legal cou n sel to determine whether thi s access is la wful. Although EPRI may make available o n a ca s e-by-case bas is an informal assessment of the applicable U.S. export classif i cat i on for specific EPRI Int e lle c tual Property , you and your co mpan y acknowledge that this assessment is solely for informational purposes and not for reliance purposes.
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