PNC-TN9410-89-089:3.95MB
Contents
1. NETWORK 1 2 3 NETWORK Fig 11 2 1 MK II OOH E2. 63 345 155 5 9410 89 089 cee a Kes NNN exes Dux Ex al o 5 T e o gt s 2 2 MMN CS gt s NINN 3 REX 5 EX Mr So Ss omy 2 So Fs ese r s ss ENS tcu E 654 C Y ww as as 555 pco 2 5 Row 4 Row 3
3. Fig 8 1 2 5 21 7 g 309 psi g CHO 52 C 7 10m PH 6 8 PH 500 1 WCAP 7828 RADIOLOGICAL CONSEQUENCES OF A FUEL HANDLING ACCIDENT December 1971 123 5 9410 89 089 FAC
4. 4 63 430 5 9410 89 089 Table 4 4 1 Comparison between experiment and calculation results for EX 2 displacement at the pads displ t ee ispl t 1 t Load the pads 680 68 OIL6NS Strain on the wrapper tube maximum displacement 05 570mm 0 i 1170mm O i 2370mm 1471 1170mm 1765 2370mm 13510 200mm
5. SR 5 9410 89 089 2 21 RRP OLE 1 H 3A 7 100 MW 17 3 31 70 1 17 100 MW 7 100 MW 17
6. LiNbO 8 Fig 12 1 1 63 344 169 02T Temporary Established GL 3090 Rotating Plug GL 6100 Electical Boiler T C Fig 12 1 1 Wire Rope 4 P Connector mu p 22 11 34 Shaft Sealing Device Sedi
7. mM J 2 75 000 MWd Z t Table 8 1 1 J2 Table8 1 1 Fissile Material Weight of JOYO J 2 Fuel kg Sub Assembly 118 5 9410 89 089 3 METHUSELAH 5
8. 3 m 1 X BEES Na 2 Na KCI 3 1 4 Nas O NaOH K SRH Fig 9 3 2 EMER LUC Kcl 20 wt O wt 10 2 x Na O 0 569 o wt 89 1 0 980 COC IKQI KCI INaOH Na O NaOH 63 378 mi 5 9410 89 089 Sample Holder Sample Aerosol Inert Gas AK VA Beryllium plate 0 25 mmt Body SUS304 70mm 8 x125mmH Fig 9 3 1 Special Sample Chamber Intensity 10 00 21 20 C INa20 IK C 08 INaQH IKCE 1 00 8 00 0 80 6 00 INa O TIKC
9. Fig 5 1 1 5 1 4 2 2 Table 5 1 1 4 17 X 63 334 SN9410 89 089 BEN X X x 90 00 10000 OPENING 2 0 0 00 48 0 64 12 00 80 00 pra 9600 10400 CONTROL SIGNAL TN 400 5 38 6 75 8 13 9 51 10 89 12 26 13 64 15 02 16 40 17 77 1915 VANE SIGNAL rT 4 00 15 52 SIGNAL Fig 5 1 1 1 Vane damper characteristic 1988 12 7 E x 2 80 00 100 00 60 00 OPENING 5 50
10. ARR 1319 1184 4 49 Inquiries about copyright and reproduction should be addressed to Technical Cooperation Section Technology Management Division Japan Nuclear Cycle Development Institute 4 49 Muramatsu Tokai mura Naka gun Ibaraki 319 1184 Japan Japan Nuclear Cycle Development Institute fi 18 9410 89 089 Vol 5 No 4 1 3 1 3
11. 5095238 131 1 Fuel Center Release Temperature Factor of I 22 10107412 2200 888 9 Fuel Value Sinter Pellet Vibro compacted Low Burn up Fuel 1 A W Thorley et al Fission and Corrosion Product Behaviour in Primary Circuit of LMFBR s a Status Revieew of Working Being Undertaken in the UK Int Conf on Fission and Corrosion Product Behaviour in Primary Circuits of LMFBR s Dimitrongrad USSR Septemder 1975 p117 122 J SN9410 89 089 6 500
12. 1 31 7 MK He ATR 1 1 2 2 17
13. FPGS 3 5 80 KW FPGS 3 5 63 423 100 559410 89 089 5 45KKW Ns 5 8 gt Power Q kw 10 20 30 40 50 60 70 80 90 100 64 m nr Fig 6 10 1 Calculation Power of Every Case 101 SN9410 89 089 6 11 NEUTRON FLUENCE EVALUATION AMIR 4 DOSIMETER SETS Sigit Asmare Santa KAA Gr i Outline Irradiation of 4 subassembly has been conduct
14. MK I CRDM 3 Na Na Ar 3 3 CRD Flux Tilting 17 18 iC19 20 MK T BER DOR 2 SG FPS
15. MIMIR N2 2 BAR QO 1 15 10 2 1 DHX 2 GD 3 Table 11 7 1 Table 11 7 2 3 Sm AE 1 10 D
16. Np CRD CRD 2 t min h mm R 3 68378 1072 7 9 56315 x 1073 6 24062 10 0 lt t lt 130 hb h 130 lt BOC Fig 3 7 1 CRD LOL CRD Np EOC RE
17. 5 9410 89 089 31 17 47 5157 FRAEN KOA ICH E Gr 7 A amp 301 MK 17 17 0 10 0 09 _0 08 0 07 0 06 0 05 Reactivity Ak Zk 0 04 0 03 0 02 0 01 20 30 40 50 60 70 80 90 100 110 120 130 140 150 Doubling time sec Fig Fig 3 1 1 Relation between Doubling time and Reactivity 63 354 SN9410 89 089 32 MK I 17 4 7 SLE f SR 8 Cr 1 MK I 17 100 MW 2 1 6 350mm
18. 63 425 an 160 11 5 MK ee 63 431 162 11 6 MK R 1 1 164 117 MK 1 2 166 12 et 168 121 oe 63 344 169 122 CGCS eee 63 391 17 1 12 3 3 oon 63 394 172 12 4 ICRP Rub 26 BAOREN ee 63 412 420 448 176 SN9410 89 089 1 1 3 31
19. 12 95 924 ERE 1 HA RCs 63 433 19 Table4 5 1 Comparison between experiment and calculation results for 3 1 49 14 98 99 1441 188 5 230 6 277 4 Table 4 5 2 Comparison between experiment and calculation results for EX 3 CASE 2 680 68 0196 5 ONd SN9410 89 089 mm Node 34901 20 Load 19 18 17 16 15 14 13 cross s ction 12 1
20. 1 002 166 E PNC SN9410 89 089 Tableil 7 1 Condition of Best Case Primary System Secordary System Set up DHX Secondary Pump Flow Flow Outlet temperature after coast down after coast down at abnormal running Natural A Loop 310 C Circulation B Loop 330 C Table 11 7 2 Compare with thermal transient condition of Loss of Power Event Inlet Outlet 4 4 41 4 4 AT At sec CC sec Power R V Case Loss of eva 1 Estimate Calculated Point Case side PNC SN9410 89 089 12 3 11 168 SN9410 89 089 121 Gr FBR
21. 17 JOYDAS 800 B 001 B004 2 1MW 1 2 1 1 90 MW B 1 21 16 27 2 208 100 MW 1 23 8 25 3 1 26 16 30 63 362 379 145 SN9410 89 089 10 2 5 Gr 1
22. 144 SN9410 89 089 10 1 JOYDAS PURE Gr 710 JOYDAS JOYDAS 1 OTS JOYDAS Cii
23. 8 63 435 135 96 63 436 137 97 63 431 139 98 NaBF FCNaF 63 438 141 10 asses pa s Z Sep a as CR aV paway 144 10 1 JOYDAS FAFA A DEED RIE IT UY 63 362 379 145 10 2 e 63 439 146 10 3 148 10 4 150 1 1 MK IMACE 0 4 63 339 350 152 112 5 63 345 155 113 ic Rb e VERE RE 63 885 157 5 9410 89 089 114 MK
24. 1 63 435 2 6 100ml IC 54g 20 Ae 10 000 Ag 100mll CI 3 added 46 Na added 4g 5 0 0 5 0 20 found peak height mm 8 7 100 0 101 1 101 1 102 2 101 1 92 0 not measurement 4 X Na 2000 peak height 8 0 VEC 10 000 4g 1 000 4g peak height 1 000 g
25. MRELORMAH 6 08 Node 20 029 10 6 08 i Node 20 10 0 Node 10 181 3 Node 20 559410 89 089 mm Node 2440 20 19 18 17 2410 1170 0 Measurement Point A Temperature Experiment Analysis Model Sleeve 2 Deflection 530 Fig 4 4 11 Subassembly Model 5 8 46 0 1 2 3 4 5 6 7 8 9 ail ls E di 2 Load Load cell cell 81 58 upper pad pper pad 163 04 lower lower pad 244 68 326 09 407 77 489 36 570 84 652 43 734 01 0 4 core sabassembly 57 8 blanket subassmbly 5 inner 6 8 outer 0 reflector Fig 4 4 1 2 Arrangement of ten subassemelies in a single row 680 68 OIF6NS 559410 89 089 18 3 81 5 Restraint frame Subassembly 0 9 0 44 LN Fig 4 4 1 3 Relation between subassembly and resiraint frame position 99 No C fel 0 mm Node 2 440 1 830 1 220 610 Fig 4 4 2 0 20 o 19 18 17 16 15 14 13 12 11 10 9 e ceo Qo O Experiment Calculation 60 40 20 0 20 40 60 Comparison
26. 0 6 wg 100 ml 3 YY 63 435 135 SN9410 89 089 0 6 4g 100 ml 3 5 gg 100m 1 68 22 0 6 g 100ml 4 10 22 6 4 1 Peak Height 0 0 13022 9 0 0 1282 2 Conc 8 5330 1077 X peak Height 0 0794 Conc 8 3250 107 Xpeak Area 0 2191 Fig 9 5 1 Calibration Curve 2 Table9 5 1 Test of Detection Limit g 100m F found peak height mm ND ND 6 5 12 56 25 10 8 2 54 Not Detect 3 Tadle9 5 2 Test of Determination Errors added 100m 2 F found 1000 F added g 100m F found g 100m 5 0 4 98 0 6 0 6 0 6 0 6 0 6 udis 5 9410 89 089 96
27. XL 5 4 amp HzSO 9 NH SO 5 183 5 9410 89 089 Fig 9 4 1 SPC 8 14 Reaction Flask 134 Size Cap Flask mm SN9410 89 089 95 Gr 1 C1 IC SN9520 88 015 F
28. FPGS 3 5 2 63 10 11 35 3 1 1 2 2 1 2 5 m r IHX 7 2 N 3 13 376 26mi hr AE etc 5 33 653 10m hr 45 0 R K 2 FITTING 0 49 15 KW Nz
29. X URP URP 63 413 Table4 3 1 Calculation results 1 displacement at the pads Load on the pads mm pivot point pivot point pivot point 1 7 1370 mn 1170 2370 10 20 10 20 pivot 1 26 i i pivot ond ivot point pivot point from 1220m 2440 ane 1220 mn from 2440 10 10 4 20 ES Node 10 Node 20 Node 10 Node 10 680 68 0176 5 displacement mm from 2370 mm EX1B EX1Bi 14 02 14 31 position Table 4 3 2 Calculation results EX 1B maximum displacement maximum displacement displacement at the pads Load on r ww 3 B km pivot point Node 20 ivot point om 2440 Node 20 pivot point from 1170 from 2370mm Node 10 14 31 Node 20 Node 10 pivot point 14 02 Gron 1220 680 68 OIT6NS 5 9410 89 089 mm Node 2440 20 2370 E112 19 18 111 1
30. 75 000 MWd Zt 1 FP 425 2 PANDA 1 Om 442m 26m X 17m 1971 6 16 3 10 7 28 1075 rem 4 6 1071 rem 1 2 5 X 107 121 5 9410 89 089 5
31. 63 443 5 909410 89 089 2 200 CRD Thermal Contraction DB Non corrected Value Corrected Value Present Method Corrected Value 2 195 New Method Excess Reactivity 2 2 190 0 10 20 30 40 50 60 70 80 90 100 110 120 130140 Time from the C R Latch min 2 670 CRD Thermal Contraction O Non corrected Value x Corrected Value Present Method Corrected alue 2 665 New Method X Excess Reactivity 5 4 k k 2 660 0 10 20 30 40 50 60 70 80 290 100 110 120 130 140 Time from the C R Latch min Fig 3 7 1 Time Dependency of Excess Reactivity 5 9410 89 089 38 MK HL 18 1 Hm E 63 315 FE MK 18 17 C5J B6 PFD402 18 17 1 8
32. 2 We PRICY CI CDF 1 FP S 3 FP 2 2 110 4 FREER ORR 3
33. SO Hs 50 5ml 50 10 ml heating test counts Peak hight heating test counts Peak hight CV ae lt a mm EXE E ETUR LIE BENI EL GA reps pes per 4 505 last tk 6 50 5 ml RUN 1 225 H SO 10 m1 RUN 7 H SO 5 mi RUN 1 20 ZAFT oko CN 15 mm 1 gt 5mm LC 2 63 392 132 539410 89 089 2 2 H SO CH SO
34. 63 405 109 5 9410 89 089 73 2 FFD CG l H 2 CURT IR O ESSE 2 2 1 2 EREDO 1 ic 1 L 3 4 Kr 100 20 3 3 6x10 cps DO 1 2 8
35. J 1 J 2 Fig 8 1 1 J 1 J 2 ELON 119 5 9410 89 089 EFFECTIVE MULTIPLICATION FACTOR mem Mk I FUEL WATER FUEL VOID 92 FUEL 71 12 0 8 FUEL VOID 41 42 Cees 269 0 5 227727777777 0 3 15 20 25 30 35 LATTICE PITCH Fig 8 1 1 keff Calculated for JOYO 1 and Mk I Fuel Arranged Square in the Storage Pool 120 5 9410 89 089 4 1 g 2 1
36. CGCS HTTR HTTR IAEA PIE Na 1 2 Na Ar R amp D 5 9410 89 089 3
37. 18 0 1 Zkk 0147 kk 8 AK KK AK KK MAGI MAGI Calc E nes Calc Cycle 63 377 SN9410 89 089 37 SHER 8 CRD CRD CRD Np OM 2
38. NaBF NaF WARE Gr 1 63 435 63 436 63 437 LOL NaBF NaBF NaF 2 ik NaBF 2 LC 100 LT 1 3 5 1928 Fig 9 8 1 4 5 1
39. 1 6 3 5 5 MK I 60 2 5 MK 6 X MK 48 3 100 MW 70 4 82 5 6 MK 31cy 63 298 X 3 Fig 11 5 1 6 MK 33 cy 35 cy 81 63 431 162 691 Excess Reactivity k 11 5 1 Preparation for Core 81 3
40. 3x10 mSv 4E 1 mSy 9 x 10 mSv 4 50 mSv 4 x 10 mSv 50 mSv 9 63 15 tc 1 2 2 4 7 Hi ALI 63 20 1 2 ALI 63 20 1
41. 2 5 ug ml 100m1l 0 2ml peak hejght peak area 100 ml Kk 0 04 0 12 m 5 Ag 100 ml 3 1 peak Height peak Area 10 4g 100m 1 Fig 9 5 1 2 0 5 wg 100m1i OBA 12 56 8 0 6 2 54 95 Table 9 5 1 3 7 v RBE b 4g100ml 1 689 0 6 4g 100ml DRA 4 10 8 12 gt Table 9 5 2 4 1 peak height peak area peak area Kit Y peak height 2 0 5 zg 100ml
42. 5 63 436 137 SN9410 89 089 S T R 138 SN9410 89 089 97 AAR Gr 1 63 435 RO 48 HA 63 436 h C1 2 5 100ml 2g NaOH A 5 Ag 15ml
43. EEPD EOC 100MW b CITATION RZ MK 25 Power Neutron Power 185 0 MeV fission 63 442 103 SN9410 89 089 R HL 7 CORE Flux Neutron Power 2 0 4 0 b Fig 6 12 1 1 2 200 300 BOC EOC 0 2 12 CYCLE
44. 17 1 17 Fig 5 4 1 Fig 5 4 2 GE 16 Sid Fig 5 4 3 3 M 63 380 SN9410 89 089 o 1 o 2A 15 y 2B AN TS 1 2 3 4 5 6 7 8 8 10 1 12 13 14 15 16 17 CYCLE No GAIN MARGIN Fig 5 4 1 Personal History of Dump Heat Exchangers Gain Margin A TEMP O 1A 6 2A 18 28 VANE 9 1 LN NA TITRE 2 3 4 5 6 7 8 8 10 11 12 13 14 15 16 17 CYCLE TEMP DEG Fig 5 4 2 Personal History of Dump Heat Exchangers Vane Opening and Atmosphere Temperature SN9410 89 089 1 2
45. Found WR 115 107 145x10 4 0 107 32x10 4 2g RAT CHE 1 mg ko kk 96 22 SSS 63 437 139 PNC SN9410 89 089 98 25 5 4 NaBF NaF 140 SN9410 89 089 98
46. Gr 1 MK L SMIR 4 SMIR 4 1 0 MeV 0 1 MeV 2 12 SMIR 4 MK IF 1 1988 8 10 1983 9 28 7162 MWd 2 SMIR 4 787 3 Fe Cu Co Al Ti Nu V 3 ik SMIR 4 END B V 2 DOT 3 5 NEUPAC JLOG cm 4 Fable 6 7 1 Table 6 7 2 Fig 6 7 1
47. Orcs Ofv t Mou 3 MK I 17 MK 17 Fig 10 2 1 2 amp VAX 63 439 146 5 9410 89 089 Sampling Time Interval 1 0 Control Rod Reactivity Reactivity by Kinetic Equation o Burn up Reactivity O Feedback Reactivity A Reactor Power MW CYCLE i7 89 02 15 Fig 10 2 1 Data at Reactivity Characterstics test by Step response method Sampling Time Interval 2 0 5 Control Reactivit n Reactivity by Kinetic Equation i B Reactivit A Feedback Reactivity X Reactor Bower 2 MW 100 7 90 6 P 2 9 80 4 3 10 2 Md 60 mE 0 Pr PT
48. KK GEYSER DUMP GEYSER DUMP iz ES GEYSER DUMP 4 DBS GEYSER DUMP DBS 63 454 150 5 9410 89 089 1 MK MK 15 539410 89 089 1 1 MK Gr 2 2 5 3A3
49. 4 151 107 2 75 9 1 051 10 2 33 6 318 x 105 Standard Deviation in Percent 63 330 331 559410 89 089 62 Gr 1 MK I CA 02 L 2 1 MK 1 MK 12 cycle 2 Fe Cu 3 ENDE B V 3 5 NEUPAC dpa sec gt 17 4 m Table
50. EXAMPLE C ZE Criticality Analysis with Fuel Pins in Water USA 1 014 Lattice Performance Analysis on Cluster Type Fuel DCA 1 004 1 016 Core Performance Analysis for FUGEN Performance Test 1 000 1 009 Initial 3 Cycle 0 990 1 003 METHUSELAH CE 0 990 1 016 780 2 0 5 7203 81 04 1981 2 HACHIYA Y et al Lattice Parameter Measurements on Cluster Type Fuel for Advanced Thermal Reactor J Nucl Sci Technol 13 11 618 1976 3 J 1979 4 H et al Core Management Analysis of the Fugen Heavy Water Moderated Plutonium Uranium Mixed Oxide Reactor Nucl Sci Eng 87 361 1984 117 5 9410 89 089 2 0 3 fissile material fertile material
51. 1 0 table 4 1 2 3 BEACON 8 A 0 9 20 1 SZ 1 Table 4 1 3 3 J BRE ORME 0 2 SA 570 mm SA 2430 mm 0 3 20 SN9450 88 003 4 1 No 0 3 63 395 PNC SN9410 89 089 Table4 1 1 Wrapper tube flat to flat temperature difference directi
52. 2 17 MAGI 13 17 MAGI 13 16 018 17 0 247 kk 0 147 TO 3 17 MAGI iD 17 15 16 4 m MAGI
53. C RRR AISA HE 2 IPTE get cu A B AMIR REAR CSMIR CMIR RRNA C SHMIR MK 1 SHE 5 A B C 2 2 1 3C2 CSMIR 582 505 6A4 604 TL GB BR tA 6C6 6D6 CMIR 1 1 U Z SHMIR 5 5 6 5 2 Ja e 4 gt SN9410 89 089 22 1 3 1 1 17 S A 3 FFD 2
54. 2 Al Be V 2mg 10 8 He oA He 10 He atoms 5 SUA 4 SMIR 16 He 5 5 J 6 5 cm ORIGEN79 He Al 7 89 x 10 atoms B 867 1017 atoms 2 189 10 5 V 810x10 He 1012 atoms 63 418 426 432 559410 89 089 6 10 Gr 1 Ns
55. kg 239 24 n i66 49 100 000 100 000 2 928 61 080 55639 90 710 90 951 44 m 40 2 119 2 161 637 252 6 587 552 551 228 23 22 100 000 100 000 752 a 292 91 060 15 171 411 156 999 156 418 230 155 000 102 100 000 1 392 id MES 17 572 72 089 91 00 x 5 201 145 056 156 156 584 177 315 213 215 m 25 416 28 681 43 884 92 92 876 Fm 48 831 96 971 156 584 157 877 100 000 100 000 46 253 im 947 76 E ss 0 i 100 000 100 000 22210 92 876 TEE 719 157 817 IER 24 100 000 100 000 EM 33 91 284 x 2 685 20 2 78 156000 156 404 163 210 s ECCE N 680 68 0196 5 RR 18 0 1239 17 870 62 7 771 218 25 641 97 X 19 5 x 233 d 2 746 13 214 159 21 263 256 34 477 11 347 at MN 11 347 556 11 465 nui OR kg 239 241
56. 2 1 OMW 250 C Predict Value of Excess Excess Reactivity Calculated MAGI Reactivity BOC 3 370 0 0 2 1 763 SAk Ek Calculated Value by 2 on _ maoa xw 1 396 x 1079 1 395 x 10 1 395 x 107 4777 109 3 779 x 10 3 781 1075 GO 16th Cycle 17th Cycle 18th Cycle 4 487213 107 4 502384 107 4 452631 1077 63 450 SN9410 89 089 4 SN9410 89 089 41 IAEA Stage 2 EX 2 BEACON Gr 1 IAEA Stage 2 EX 2 BEACON 2 1 De table 4 1 1 2
57. 1 Na Na I Na 500 C 800 C 2 63 336 3 Na 3 63 336 Na Na Table 9 2 1 Fig 9 2 1 Bredig Allan 2 log 8 8 557 3524 800 C 661 4 C log S 9 751 4640 661 4 C 500 C CC S ppm T Na K 2 661 4 C 2 661 4 C Bredig 63 347 128 5 9410 89 089 5 ppm 900 800 1700 600 500 400 105 104 103 102
58. 2 fa 17 Table 3 3 1 Fig 3 3 1 Fig 3 3 4 Table 3 3 1 Summary of Measured Reactivity Coefficient 17th Cycle Total Power MWd 6838 1989 1 18 4 1 Core Average Burn up MWd Zt BOC 2 24 x 10 3 17 x 104 1 2 3 C R Total worth Excess Reactivity Ak Zk BOC 2 665 at OMW 250 C Isoth 1 ture Coefficient ermal Temperature Coefficien x 10 7 BOC C Flow Rate Reactivity Coefficient 250 x 10 BOC SAk k Flow 1 25 x 107 EOC 0 930 Power Reactivity Coefficient BOC EOC 0 100 MW Average BAk k MW 5 26 x 1073 3 83 x 1073 Rated Power Np Ineffective 2 54 2 33 ea 8 Burn up Reactivity Coefficient Zero Power Rated Power X 107 AkZ k 1 Calculated Value by JOYDAS 2 Calculated Value by MAGI 3 Sodium Temperature 250 C R Equality 63 367 370 393 415 428 SN9410 89 089 17 17 BOC 1 18
59. Reacton Vessel Measurment Points 003 1 Fig 6 5 1 Measurement Points for CP Deposits Deposition Density SN9410 89 089 107 Deposition Density Ci cm 107 HOT LEG COLD LEG COLD LEG R ZV IHX PUMP R V 107 0 20 40 60 80 Distance m 6 5 2 Deposition of Corrosion Products MK 2 gt O Mn 54 C Hot 16th Cy Mn 54 C Cold 1 Mn 541iCl Cold 2 MK 2 Mn 54 E Hot 12th Cy Mn 54IEIColdul 2 Mn 54 2 7th Cy Co 60 C Hot MK 2 60 Cold 1 12th Cy Co 60 2 MK 2 60 Hot 16th Co 60 E Cold 1 9 Co 60 E Cold2 2 7th MK 2 2nd Cy Deposition Density Ci cm 0 20000 40000 60000 80000 100000 120000 7 0 20000 40000 60000 80000 100000 120000 Cumulative Reactor Power MWd Cumulative Reactor Power MWd Fig 6 5 3 Comparison of Measured vs Calculated CP Buildup in Primary Main Circuit Piping A 5 9410 89 089 1982 7 2 1983 7 26 1984 1 730 1985 9 3 1987 4 20 1988 10 11 Dose Rate mR h Distance m Fig 6 5 4 Surface Dose Rate Distribution LOOP 1982 7 9 1984 2 6 1985 9 9 a 1987 4 27
60. 150 MW Table 11 3 1 4 IC 63 385 5 9410 89 089 Se loc 71113 Main Intermediate Heat Exchanger CA Loop Core Support Section Main Intermediate Heat Exchanger B Loop Upper Plenum Section Fig 11 3 1 Section of Evaluation Object 158 691 Analysis Case Section of Evaluation Object 1 Reactor Vessel Cone Section Leakjacket Joint 2 Reactor Vessel Core Support Section Lower Flange 3 Main Intermediate Heat Exchanger Loop Upper Plenum 4 Main Intermediate Heat Exchanger A Loop Lower Plenum 5 Main Intermediate Heat Exchanger B Loop Upper Plenum Table 11 3 1 Tresca Stress kg Loss of Power Neutron Flux High
61. Table 6 3 1 63 338 5 9410 89 089 Table 6 3 1 Result of Neutron Fluence Measured Irradiation Rig 1 01 essen _ 1619 seris 157 sens 025 04 sisse co reso cio saris enis J Standard Deviation in Percent Irradiation Rig Il 01 eem summ sime een me _ mafon aan iana oran aafia Gas esan Car Ora ua 18 me eres as cien nme an 23 Standard Deviation in Percent TOTAL gt 1 0 MeV n 20 1 MeV dpa dpa 559410 89 089 6 4 1 2 55 Gr 1 RAO PERSE OD I BN Lr HER Ic k EFE DEUS Cr UR OO ER ERRARE 2 1 9 2 17 3 HE 1
62. 2H 100MW 17 1 9 4 3H 31 70 AT v 7G fe 100 MW 18 1 3 HO 2 1 1 2 1 2 2 1 3 2 1 8 17 2 1 1 1 fl R amp Table 2 1 1 Monthly Operating Summary January 1989 through March 1989 LI m sare e EZ amma nom mm 7 P MU 7 saasom el e e sw 91 m 680 68 017686 ONd 2 1 2 J X Table 2 As 2 Accnmulated Operating Summary through March 31 1989 5634 MK II 563
63. 28 18 224 tg BF 25 0 25 95 NaBF NaBP NaF BF 100 ppm 100 ppm NaBF 1 10 W 3 2 32mm 63 438
64. Gr 1 B 2 CG 1 CG Kr CG 3x 108 Gi cad OO iC 6 5 X10 IAE IB CG B G 6 x10 92 3 x10 3 CG 10 P133 126 2 4 28 126 63 399 108 5 9410 89 089 7 2 1
65. E E 12 3 eee 14 3 1 MK I 17 miss RETR IT k Ge 6323545 15 3 2 MK 17 meet 6923642 16 3 3 MK 17 63 367 370 393 415 428 220 34 MK 17 e 63 311 419 24 35 63 376 383 26 3 6 MK 17 717 MAGI 63 377 28 3 7 63 433 29 38 MK 18 ee 63 450 31 bu ha an d b RM MM ADI UN 32 4 1 1AEA Stage 2 EX 2 MM 63 395 33 42 IAEA Stage 2
66. 1 LI 6080 68 0196 5 2 1 6 Table 2 1 6 Operating History Data of JOYO January 1989 MS Se BEE BEE e Te S Te T8 s E UNM aras T I 15 85 KW 1 91 ELA CLL E vw 125584555428 war 3 1 Na fit Lu 2 BUR 246 SHMIR 1 UPR 1 Na 8 HE uiu eei Age 1 X Na t fd que 3 Fy EEE es RP ZU 5 SMIR 4 TH TT E exin Elo ie amam 1 12022106 680 68 0196 5 ONd x 217 Table 2 1 7 Operating History Data of JOYO February 1989 IT r mE GL 6100 2203 SS 310 100 85 HLLLLL SHAG Na Bie 472 HT TET E ZH cR ekek 5 HEV CURRERE LE TT LL T gt NNUS
67. 63 399 108 2 2 eM 63 405 109 13 2 FFD CG ere 63 434 110 SN9410 89 089 7 4 2 enn 63 455 111 8 45 FH Qy 7 060 OREA SE 1 15 81 37 e 63 374 116 9 P 125 91 Nal Na 63 336 126 92 Nai eene 8 63 347 128 93 T 83 578 130 94 FC Na EEE iG tat 63 392 132 95
68. 12m Zh 63 396 Heat Transfer Coefficient Over Flow 12ni h 10 500 kcal nf h C Over Flow 0m h 2 0 kcal nf h Over Flow 0n 12nr h 2 0 kcal nf h C 10 500 kcal nf h 100 0 kcal nf h C 3 5 kcal nf h Gap 0 01 mm 3 NOR 7 1 PORSAS TA BOE Sa coder 2217 UU L U EE ce whe ar adt L4 HIT ur PRS Sour ee parsers SaaS I I k e Heat Transfer Coefficient No HE i i eg x x Y Fig 5 6 1 Condition for Heat Conduction Analysis of the Safefy Vessel Penetrating Part of Return Pipe 680 68 0196 5 Fastening gt 3 A 2 1 1 4 1 27 1 on 177 1 ap X 7 Section Y Direction Fixed End Y Z Section X Direction Fixed End Fig 5 6 2 Condition for Thermal Stress Anal
69. 2 413668595 28 63 451 148 5 9410 89 089 1988 2 29 0 0 1989 3 31 0 0 JOYO MK II CO26 0 00 12 50 25 00 3750 50 00 62 50 75 00 87 50 100 00 11280 12500 5 800 MW PNC JOYO JOYDAS PLOT Fig 10 3 1 History of Vane Opening Signal IA 149 SN9410 89 089 10 4 GEYSER DUMP Gr DB S 4 DES MK DASD GEYSER DB DBS GEYSER DUMP GEYSER DUMP
70. 1 0 min 30 60 HH 2 Na ERED SOADUO HBARL LCHEMIC Cold leg 7 81 x10 75 Na 350 2 13 m 10 200 h Hot leg 1 48 x10 4Ci S Na 460 2 TTT MAME 10 H H 63 375 398 5 9410 89 089 6 7 SMIR 4
71. 2 0 ER 3 2 2 58 E E NSK DOES H k 7 3 f 1 Fig 5 6 1 MEA 10 500 kcal mi h C 2 0 kcal Z m sh C ASRR 2 0kcal nf h 280 sec COR 2 0kcal rf h 10 500 kcal h C 2 Fig 5 6 2
72. 205 1 5 3x 10 GEXA 3x10 7x10 CH 3 1 9 2 2 177
73. 3 ICRP Pub 30 Supplement 4 5 63 412 420 448 176 539410 89 089 2 51 9 28 1 5x 107 mSv 16 10 3 14 4 1 9 63
74. A 74798 A B FP 1507 A 990 77774 28 RE 3 m 1 13 2 180 C A B 5 80 C B 3 C 3 600 4 P D 5 A 20 C B UR 6 N HAMA 7 63 391 171 5 9410 89 089 12 3
75. Na 650 C 2 7 14 Na Fig 9 1 2 In D 3 80 2 30 10 xt D t Na C 4 Na WE 500 C 800 C Nal 63 336 126 561 9 1 1 Concentration of Sodium Jodide Sodium Heating at 500 C Heating at 650 C 5 34 5 58 5 63 4 2 7 14 17 14 7 20 28 5 76 5 53 5 66 10 28 Fig 9 1 1 Test Capsule Heating at 800 C Heating Time conc Nal Heating Time conc Heating Time conc Nal Day 45 Day Day 5 Day Heating Time SUS 316 TUBE 19 0 mm LD 164 mm 500 600 700 800 Na Temp C Fig 9 1 2 The Heating Time Curve for Preparation of Saturated Solution 680 68 0196 5 5 9410 89 089 92
76. 12 REFA 12 Subassembly as sd Subassembly 2 Cycle Address Cycle BORD Number cw Number 5 PPD129 PD129 9 13 000 000 450 4 2072 20728 PFD217 9 19 252 282 468 8 1326 132 6 PFD 135 359 7 10 14 2E1 350 4 209 8 PFD 145 472 d 1096 PFD221 10 14 202 362 4 352 5 156 347 4 171 3 PFD 223 10 14 369 4 174 3 PFD 208 356 d 2047 PFD 224 10713 282 261 4 199 9 PFD 212 367 p p 2D2 _ PFD215 9 14 3F1 343 4 1494 149 4 value by FPGS code 63 340 5 9410 89 089 65 CP Gr 1 CP 2 CP 7 MIKK Db 16447 VATE ic
77. 2 Co Fig 6 5 3 6 3 35 IHX PUMP A 259mR h 344mR h 345mR h 562mR h B 1709mR Zh 1620mR h 500mR h 146 mR h Fig 6 5 4 Fig 6 5 5 amp 63 361 4 OCP Na Na CP CP CP
78. 3 GHA BHAA 30 4 1 3 1 30 3 63 394 21202555 2 m 63 2 854 1 1 30 a d RE 9 13 K DHF eRe Eo dE ORG ICES 30 42228 1
79. 1 10 5 JOYDAS 10 17 Fig 5 5 1 10 16 17 63 384 Outlet Temperature 500 490 480 470 6A2 Fig 5 5 1 aS n Qi O ES x 10thCYCLE O IZthCYCLE X 13thCYCLE 14thCYCLE 15thCYCLE 16thCYCLE 6B2 6C2 6D2 6E2 6FZ Position Number Subassembly Outlet Temperature of Inner Reflecter 10 1 17th CYCLE 11thCYCLE gt 680 68 0176 5 559410 89 089 56 2 7 8 Gr 1 63 127 O 2
80. eee 81 61 ATR MM 63 330 331 89 6 2 389 15 093 03 soe 83 6 3 63 338 85 64 1 2 eee 63 340 87 65 EJ CP 002639617 88 66 63 374 398 92 6 7 SMIR 4 Opi u S aan 63 387 93 6 8 SMTR 2 tt 63 388 96 6 9 He KO SMIR 16 He 63 418 426 432 99 6 10 872311 k 2 BBEAGUSEIERBREEE 63 423 100 6 11 Neutron Fluence Evaluation of AMIR 4 Dosimeter Sets 63 427 102 6 12 FPGS 3 5 ICE eee 63 442 103 6 13 Measurement and Evaluation of Burn up Distribution on the JOYO Spent Fuels 63 449 105 7 MAELEEBESESESRER a fx wh DBE gs 107 7 1
81. 50 1 40 s 30 4 5 20 6 10 1 8 0 28 32 36 40 44 48 52 56 min CYCLE 17 89 01 30 Fig 10 2 2 Data at Stability test 141 SN9410 89 089 10 3 JOYDAS Gr X JOYDAS 1 2 Fig 10 3 1 100 MW 98 100 MW 0 C _ y KE C ta 3 09 n 0 JOYDAS 44 3
82. 1 EX 1B 2 1 F BEACON 2 MM 20 3 20 Fig 4 3 1 4 se 2 62 3 31 EX 1A table 4 3 1 Fig 4 3 2 LRP URP 20 1 2 Ny OSESUES O2 TH 32 EX table 4 3 2 Fig 4 3 3 D
83. 63 353 SN9410 89 089 5 4 MK 17 M AMAL Gr 1 M f K Na 2 M 1 26 1 30 M 1A 2A 1B 2B 16
84. JOYDAS PC 9801 C AMAC 2 dn ma n oe m gt Ci Ci m 2 2 2 11 x EET roa sn Arcs rcs
85. INTA 8 18 6 MK C amp SG FFD AR F 3B HTTR 1CRP Pub 26 R amp D He SN9410 89 089 FAKS RHCHL
86. 1 table 4 2 1 1 EX 2 table 4 2 1 2 2 1 0 table 4 2 2 3 EX 1A B EX 2 2 Fig 42 1 4 30 EX 1A table 4 2 3 EX 1B B k table 4 2 4 2 S A 0 59 S A Na 1 table 5 3 1 PNC SN 9450 88 003 IAEA IWGFR Coordinated Research Programme Intercomparison of LMFBR Core Mechanics Codes Japanese ex reactor data for the validation exercise Revised Edition 2 BEACON d 683 411 SN9410 89 089 mm 2 mm 2500 2440 2410 1 50 0 2370 2310 2170 2070 1970 1870 1840 1770 1670 1570 1500 1470 31 61 1370 1340 1270 1170 1070 1040
87. OMW 250 2 665 AkZk OMW 250 MAGI 2 920 Ak Zk 2 17 EOC 4 1 0 OMW 250C 0 930 OMW 250 MAGI 1 292 3 MAGI 0MW 250 0 255 0 362 Calculated Value Measured Value Excess Reactivity at 0 930 Operation time SN9410 89 089 x107 45 4K K 8 0 Power Reactivity Coefficient 0 10 20 30 40 50 60 70 80 90 100 10 Thermal Power MW Fig 3 3 1 Power Dependency of Power Reactivity Coefficient Excess Reactivity 4K Z K Operation Time days Fig 3 3 2 Change of Measured Excess Reactivity Depend on Burn up SN9410 89 089 2710 5 4k Zk Z flow Y 2 5029D 04 X 2 6995D 00 2 685 e x 2 660 0 10 20 30 40 50 60 70 80 90 Primary Coolant Flow Rate flow Fig 3 3 3 Flow Dependency of Excess Reactivity 0 960 55 Ak ZkZ flow Y 1 2477D 04 X 9 49180 01 42 E 0 945 Q 3 0 930 0 10 220 30 40 50 60 70 80 90 Primary Coolant Flow Rate flow Fig 3 3 4 Flow Dependency of Excess Reactivity 100 100 110 110
88. 63 465 559410 89 089 Table 2 41 Event Condition Uncontrolled Control Rod Withdrawal Accident at Normol at Reactor Operation Start Up Scram Doppler Effect 4 K ZK maximum 0 37 x 1073 Coolant Density 4K K C maximum 5 7 107 Wrappler Tube maximum 0 39 x 1076 ft Pellet Expansion m Clad Expansion 7 Reacti vity Coefficient maximum 0 63 x 107 Core Support Structure Expansion 10085 Flow x Loop Secondary p 100 5 Flow ump B Loop A Loop Blower 1 1 2 Paramater Table 7 4 2 Peliet Density Condition of Each Paramater Case 10 melt Channel Clad Fuel Conductance Linear Power Flow maximum maximum Wem Wem Ckg sec Temperature C Temperature C 0 570116 545 056 1 8483 2854 12 553 184 1 8791 2862 98 599 021 2 0544 2868 93 1 05489 592 883 2 0307 2868 18 625 939 2 1588 2872 04 468 419 1 5632 2839 54 496 740 1 6616 2841 78 0 390610 493 035 1 6539 2841 66 513 010 1 7280 2842 79 2500 00 680 68 0196 5 SN9410 89 089 16 940 Fuel Maximum Temperature 14 860 1 Melted Fuel Fraction 12 780 5 s gt 2 S 10 5 700 az 8 5 2 8 a 8 8 620 E Clad Maximum Temperature 2 6 8 540 2 5 8 4 46
89. 0 3 d 0 1 3 2 0 Measured Value Reference P 1 1 X 1 2 Decay Heat W 150 200 250 300 350 Cooling Time day Fig 6 12 1 Comparison between Calculated Value and Measured Value on Decay Heat 104 5 9410 89 089 6 13 MEASUREMENT AND EVALUATION OF BURN UP DISTRIBUTION THE JOYO SPENT FUELS Sigit Asmara Santa Gr 1 OUTLINE Experimental theoritical investigation were carried out on fuel irradiated JOYO FBR order to study correlations between radioactive fission products which were determined by means gamma spectroscopy and calculated burn up by MAGI code The gamma ray distribution of spent fuel assemblies were measered means of gamma spectrometry system to obtain and identify radioactive fission products and to survey the radial and axial burn up distribution The special aim of these experiments was to investigate the relation of fission poducts activity and calculated burn by MAGI code used to develope MAGI code and make easy and practical manner relative burn determination 2 RESULT OF EVALUATION Th
90. MK I III S63 MK MK I II B 5 38 Fal JU K at E at S 57 11 22 5 52 4 24 5 57 11 22 CS 52 4 24 8 57 11 22 6 53 4 24 H 1 1 31 H 1 2 28 H 1 3 31 1 s e _ 12 12 few maemo cm ae sos ml mm wel wp 5 1 1 C T rananow s C T examzew 1 T 1 1 2 680 68 01 6 6 ONd x 2 1 8 Table 2 1 3 Chronology of Principal Activities in January 1989 4 20 7 RARE dioi sree TCT TT FL m amp W 2 Dwowwemmoue 4 gt m x gt gt INTA SHEL _ 1 RERE E H I EO HERE yeah b CERE SEO M DE EB K FERE R BHEN BRF FEMM S ARE URE m
91. 20 63 395 8 20 fig 4 41 3 4 s 3 m 3 1 table 44 1 KO f ig 4 4 2 0 9 SA 2 3 7 5 0 LRP URP 9 5 1 X SA 5 32
92. 0 14 mNa 63 335 SIoXO 11 uomem3uguoo 9100 175794 022262022 2 2222 020202229 22 MOD D D D D Do OO Solo oco e 0008022 5589410 89 089 5 3 Gr 7 MK 17 28 j s ee 77 e Car a 2 28 11 07 08 2A TE312 2A 2B 10 05 05 0 000 7 ROSE 12 A TE312 24 3 00 oo oo C eo oo oo D UT 15 4 oo oo oo DAC eee ee De 79 79 o RC TR
93. MK LT 1 45 JOYFL MK IT 4 07 kg 5 0 kg cm it 0 59 kg cm 0 9 kg cm 0 1 17 95 kg 5 14 44 kg MK II 14 6 kg 63 425 160 5 9410 89 089 ee lt 161 Fig 11 4 1 Calculation Model of Hydraulic Hold Down Force 559410 89 089 1 5 MK 1 MK K 6 5 MK 32 cy 35c AR
94. i x Xi Rod Stroke Excess Reactivity COEF 0 3 78070165 D 00 COEF 1 1 784267670 02 COEF 2 6 34711667D 05 COEF 3 1 61715157 07 COEF 4 2 04795987 D 10 COEF 5 9 60610050D 14 400 450 500 550 600 Rod Stroke mm Fig 3 2 4 Stroke Curve 4 650 680 68 OTP6NS ONd 261 Excess Reactivity 2 4 Fitting equation Y 1 xX X Rod Stroke Fitting equation Y XiCoef i x X X Rod Stroke Y Excess Reactivity COEF 0 5 33614659D 00 COEFil 3 399441920 02 COEF 2 1 27494283D 04 COEF 3 2 84630979 07 COEF 4 3 20675374D 10 COEF 5 1 39464062 13 Excess Reactivity Ak Zk Reactivity 2 99504217D 01 COEF 1j 1 81447614D 02 2 8 68799294D 05 COEF 3 1 524880270 07 4 1 21108422D 10 COEF 5 3 753160440 14 350 400 450 500 550 600 650 350 400 450 500 550 600 650 Rod Stroke Rod Stroke mm Fig 3 2 5 Stroke Curve 5 Fig 3 2 6 Stroke Curve 6 680 68 0196 5 ONd 559410 89 089 3 3 MK HH 17 Gr 1 MK HH 17
95. rw 3 7 8 84 54 UNIT 1 CL07 LP03 04ML 95 2 1 90 52 89 60 81 64 UNIT 2 CL02 1 01 02 ML UNIT 2 CL07 1 03 04 ML UNIT 1 CL02 LP01 02 MH 1 5 UNIT2 CL02 LP01 02 MH 1 CL07 LP03 04 UNIT2 CL07 LP03 04 MH 680 68 OIP6NS SN9410 89 089 5 2 1704 72490 Gr MK 17 9 6 63 181 L BRER WmesmisHISHOO 9 30 12 00 522 2 2 1 505 ae 6D6 2 Fig 5 2 1 3 Av 7 48 64mNa B 48 09mNa A BE FE 8 UU XE RH O A x B 4 8 H A B 20 06 INTAS A B
96. gt Setting 1 2F1 545 3 3F1 540 4 3F2 534 5 4 507 6 4 4 502 8 4 2 526 9 423 505 606 EREN 13 ni DUE 16 573 506 17 5 4 509 18 19 2 680 68 OIF6NS 5 9410 89 089 35 Gr 1 jJ SA 6 10 17 Table 3 5 1 2 6 63 376 383 SN9410 89 089 History of SA out temperature 6 line 1267 i4cy 16CY i7cy AVE 17 AV 1 10 12 11 13 12 4415 3 5 1 CO Nf lt lt Il m lt N i gt ET 2 1 480 418 483 479 492 487 478 491 47 68 4 11 4 01 41 484 471 488 497 496 492 496 496
97. 1 Gr 1 AEA IWGFR STAGE 2 3 _ 2 D BEACON 2 20 fig 4 5 1 3 6 iD CASE 2 ees 6 3 1 CASE 1 table 451 fig 4 5 2 3 BEENEN13 16 2 2 table 4 5 2 fig 4 5 4 5 D
98. 5 9410 89 089 34 MK 17 1 8 17 94 7 D ESPERA COMAMEE 63 371 419 44 Setting Address Panel 14 000 573 1E1 556 565 8 241 561 2 5 107281 556 54 4A 1 4 2 17 4 4 517 17 4 4 517 517 GE 483 456 sai 476 5 642 47 sas 480 saa 485 Cm eas 476 1 Table 3 4 1 Setting Tempe Panel Address rature m s 555 4 382 527 5 scr 9s Fe ses sm e 15 e 484 si 40 9 Im 58 5 2 465 TRAI 1 2 5B4 516 E i m 564 498 480 194 480 196 9 ecs so 01 Predict of Subassembly Outlet Temperature 2 2 550 2E1 542 2E2 557 554 543 520 546 503 510 3 2 1 2 4 4 401 542 517 5 4E2 so 6 scs 45 He sp2 4ee _ soo 5D4 5D5 5E 1 2 6 6 602 603 co 65 52 6
99. INaOH IKCZ 4 00 0 40 2 00 0 20 0 00 O 0 00 0 20 40 60 80 100 Na O wt 100 80 60 40 20 0 NaOH wt Plus Internal Standard 20 wt Fig 9 3 2 Plot of Calibration Data 131 SN9410 89 089 9 4 FC Na Gr 1 FC Na SN 9520 88 015 Na Fig 9 4 2218 3 HE RBS 25 RBS 25 Fs RBS 25 HzSO 2 SO 310 CEIA IC KOU 2
100. 1 TYPE1 i TYPE 2 TYPES 5 3 DOT 8 5 6 11 6 1 1 1 164 SN9410 89 089 11 61 Se s 7 88E 06 T51E 06 1 01E 07 1 3 10 16 2 92 16 3 901 17 6 19 02 6 27 E 02 1 67 02 2 6 48 14 598 14 6 96 14 TYPE 3 919B 13 8 0 13 103 14 Neutrons sec Total Photons sec Total E PNC SN9410 89 089 1 7 1 MK
101. 1988 10 28 Dose Rate mR Z h 41 c Distance m Fig 6 5 5 Surface Dose Rate Distribution B LOOP SN9410 89 089 66 201 Gr 1 MK 16 2 Na K CH 2 677 6 mm 1 2mm 1000 BS 100 mm Na Cold leg Hot leg 100 CCHS 2 2 230
102. 3 1 4 1 5 1 5 1 6 Cr p i 6 Cl p i 7 Cr p i 7 SN9410 89 089 4 1 3 TEMF DATA for IAEA MARK TEST 2 Number of Assembly 1 Length of SA 2440 mm Number of NODE 20 Length of NODE 122 mm Wall No R7 gt R1 LI gt L7 Input No 1 m 8 gt 4 Room Temperature 20 Temp Deference between surface of wrapper wall along the axes INFUT DATA for BEACON code Temp data One point par wapper tube 122 00 24400 366 00 488 00 610 00 732 00 854 00 976 00 1098 00 1220 00 1342 00 1464 00 1586 00 1708 00 1830 00 1952 00 2074 00 2196 00 2318 00 2440 00 0O 3 ON 0 oc SN9410 89 089 42 IAEA Stage 2 EX 1 EX OFF pk BEACON Gr 1 IAEA Stage 2 EX 1 2 BEACON 2 1
103. 650mm 2 6 MK 17 D 3 k k Ci R R XX mm COEF 0 4 1920333D 00 COEF 0 1 6905478D 00 0 2 6928953D 00 COEF 1 1 82412220 02 COEF D 4 63475020 03 COEF D 6 4612224 03 COEF 2 5 4809690 05 COEF 2 2 94462500 05 COEF 2 1 6927572 05 COEF 3 1 27101090 07 COEF 3 2 8155428D 08 COEF 3 6 6158812D 08 COEF 4 1 5474936 D 10 COEF 4 1 20969190 11 COEF 4 1 0637497 D 10 5 7 07661010 14 COEF 5 1 85613400 14 COEF 5 5 5716502D 14 COEF 0 3 78070162 00 0 5 33614660 00 COEF 0 2 9950422D 01 COEF 1 1 78426775 02 1 3 39944190 02 1 1 8144761D 02 2 6 3471167D 05 COEF 2 1 2749428D 04 COEF 2 8 6879929 05 COEF 3 1 61715160 07 COEF 3 2 84630980 07 COEF 3 1 5248803D 07 COEF 4 2 0479599D 10 COEF 4 3 2067537 0 10 COEF 4 1 21108420 10 COEF 5 9 60610050 14 CO
104. E H 1 0 Node 10 318 20 0 11 Node 10 071 10 055 20 37 8 1075 Node 2 T hH 9 14 Node 10 0 Node 20 415 Node 20 057 Node 10 415 i Node 20 2 09x 1075 Node 2 5 68 _ i Node 10 0 Node 20 611 Node 20 043 Node 10 611 Node 20 9 95 107 Node 2 i 22 1 Node 10 0 Node 20 9 ps 50 17 7 x 1079 i Node 2 4 8 4801 20 011 10 480 20 214 x10 Node 2 865 10 0 Node 20 3 510 Node 20 009 10 510 Node 20 357 10 2 113 8 1 Node 10 140 1 Node 20 5 369 Node 20 082 I 10 369 Node 20 345 10 Node 2 2023 Node 10 0 1 Node 20 336 1 20 062 10 3 36 I Node 20 507 10 Node 2 253 9 Node 10 142 i Node 20 196 1 Node 20 096 Node 10 1 96 1 Node 20 719 107 Node 2 0 Node 10 1797 Node 20 84 9 055 20 0 25 10 0 55 20 216 107 Pivot Point mm Node Node 2 10 2044 Piyot Point KD 244 1220 2440mm
105. 10 08 009 10 1 1 12 13 14 300 X X M A Bredig 195 1 5 C G Allan 1973 Sodium anelysis group presumption curve 1 6 1 7 1 8 Na Temp 1 10 7 Fig 9 2 1 Solubility curve of sodium iodide in sodium 129 SN9410 89 089 93 Gr 1 Na Na Na Na 2 1 Na HS Fig 9 3 1 2 Na
106. 1013 1077 1012 1078 800 600 400 200 0 200 400 600 Distance from the Core Center mm Fig 6 8 1 Axial Distribution of Neutron Flux 1022 102 1029 101 800 600 400 209 0 200 400 600 Distance from the Core Center mm Fig 6 8 2 Axial Distribution of Neutron Fluence 102 10 109 107 TOTAL A gt 1 0 MeV gt 0 1 MeV DPA DPA dpa sec 100 MWt O TOTAL gt 1 0 MeV gt 0 1 MeV DPA DPA dpa 5 9410 89 089 6 9 He SM IR 16 He WHE Gr He SMIR 16 He 1 He n Q He He
107. NM MERE Flux Tilting pore 29 C3M 8 1 3 2 3 1 3 2 5 9 10 12 13 SHMIR F no 680 68 0196 5 X 2 14 2 1 4 Chronology of Principal Activities in February 1989 9191918 IAEA H z 45465 A2D B6 054 INTA S UPR 1 8 2 AMIR 3 1 3 2 SMIR 9 10 12 13 SHMIR 1 m ASS Ar 737 M 5 EX m bp F Il ES 8 SUR EPR 680 68 OIL6NS Xx 215 Table 2 1 5 Chronology of Principal Activities in JOYO March 1989 D 2 s a a aeo a LLLLLLLLLELLELLET EEE EEE EEE FFARR TAR 2D C3M C4F E INTA S 1 3 CPR 2 3 1 3 2 SMIR 9 10 12 13 SHMIR 1 mH FFD NE INE x pz PMSA i 1
108. een 18 6 Result of Stress Analysis Tresca Stress Loss of Power 5 32 7 Neutron Flux High 33 9 Loss of Power 32 Neutron Flux High 33 6 680 68 OIL6NS 5 9410 89 089 1 4 MK FOU y9 EFD 1 ADS C EFICZIAMB amp MK BE Fig 11 4 1 MK 2
109. 970 870 800 770 670 1000 570 470 370 200 170 70 0 18418 1 51 72 I o 18 03 530 Fig 4 2 1 1 2 Subassembly Model 5 9410 89 089 Table 4 2 1 1 Wrapper tube flat to flat temperature difference for 1 B Axial Level BX 1 Table 4 2 1 2 Wrapper tube flat to flat temperature difference for 2 Axial 2 80 ize 206 252 283 25 7 309 252 309 FS 26 me Lee pe mee epe m s ase o direction No 0 9 SN9410 89 089 Table 4 2 2 Temperature at point on circumference of wrapper tube vall Axial leyel 770 mm Cr temp 1 Cr p l Cr temp 1 1 Cr temp 1 1 Cr temp 1 Cr P 1 Cr temp 1 Cr p 1 Cr temp 1 1 Cr temp 1 Cl temp 1 p C1 Ci temp 1 1 1 temp 1 1 1 temp 1 1 1 C1 temp 1 Cl p 1 temp 1 Cl p i Cl temp 1 111111 l H H H Cr temp 2 7 0 000 Cr temp 2 6 016 Cr temp 2 031 Cr temp 2 094 Cr temp 2 Cr temp
110. Row 2 Row 0 1 Form of Cauitations in each Flow Zone Fig 11 2 1 SN9410 89 089 11 3 MK Gr 1 E MK 100 MW 2 Fig 11 3 1 A B 3 1 2 38m 3 Sm 59 4 38m 33 6 kg mm B 44 6 kg mm 38m 31 2 kg mm
111. a 1 1 RH Na P fit 2 A 2 62 1 1 3 1 4 1 2 1 1 680 68 OIV6NS X 218 Table 2 1 8 Operating History Data of 4 March 1989 y Tw e e y Te 2 5 4 8 8v T5 8 9 57 PAK EE sn 1 Na di RF ALI Na 370 C _ Tr TT us 1 amp 3 Na it Tit Na it 100 9 eter 038 Na 472 C HHHHH ga T 1 Ol 7a head See ca EEBE L T anke S mum 2 Na S P zm ika I 2 Na i HE 2 8 Le 1 Ar S P 86 62 14 3 1 4 2 1 1 680 68 0196 5 559410 89 089 i 0 50920 2108 a OZE R SQ R 80 T 13 R 188 add R 178 15 R 160 2 1 1 100 MW 17 Fig 2 1 1 Core Configuration for 17th Cycle pd ee ee ee EL odo 1331 S ik Jj 298 A
112. mo Calculation 60 40 20 0 20 40 60 mm 4 4 2 5 Comparison between experiment and calculation result 680 68 OIP6NS CASH mm 2 440 20 1 830 15 1 220 10 8 T 6 670 5 4 3 2 Fig 4 4 2 6 ee HEH Experiment Calculation 60 40 20 0 20 40 60 mm Comparison between experiment and calculation result mm 2 440 1 830 1 220 20 19 18 oO G A Q Experiment Calculation 60 40 20 0 20 40 60 mm Fig 4 4 2 7 Comparison between experiment and calculation result 680 68 0196 5 66 No 8 RE mm Node mm Node 2440 L 20 o 2 440 20 19 1 e 18 18 17 17 9 16 He 16 ge 1 830 15 1 830 15 14 14 13 8 13 a 12 e 12 2 3 1 220 10 1 220 10 9 9 8 8 7 7 6 6 670 5 610 5 9 4 4 3 HOH Experiment 9 Experiment 2 Calculation 2 Calculation 1 1 0 0 60 40 20 0 20 40 60 mm 60 40 20 0 20 40 60 mm Fig 4 4 2 8 Comparison between experiment and Fig 4 4 2 9 Comparison between experiment and calculation result calculation result 680 68 OIL6NS ONd 5 9410 89 089 45 lt 5 2 3
113. 001 1 241E 20 1 7358 20 7 155R 01 5 384 19 5 082E 19 1 059E 00 OVER 0 1 MEV e gion 20 8577E 20 6 901E 01 1 554 21 2 478E 21 6 270E 01 6 776E 20 1 054 21 6 426E 01 Table 6 7 4 Distribtion of Neutron Energy Ratio of Neutron Fluence to Total MAGI MAGI NEUPAC NEUPAC 1 000 00 1 000 00 1 000 00 1 000E t 00 1 000 00 1 000E 00 OVER 1 0 MEV 2108E 02 1 266E 02 2 933 02 2 729E 02 2 4868 02 1 685E 02 OVER 0 1 MEV 2 796 01 2 925E 01 3 673E 01 3 900E 01 3 128E 01 3 495E 01 Table 6 7 5 Axial Distribtion of Neutron Fluence Standardized by the value at Core Center MAGI NEUPAG MAGI NEUPAC MAGI NEUPAC 5 004 01 4 614E 01 1 000E 00 1 000E 00 5 119 01 4 7478 01 OVER 1 0 3 596E 01 2 141E 01 1 000 00 1 000 00 4 338E 01 2 930E 01 OVER 0 1 MEV 3 809 01 3461 601 1 0008 00 1 000 00 4 360 01 4 255 01 5 9410 89 089 1018 1073 TOTAL lt gt 1 0 MeV gt 0 1 MeV O DPA u t 3 gt lt gt gt 9 10 5 8 gt Mu E lt A gt 1012 1077 2 1012 1078 600 400 200 0 200 400 Distance from the Core Center mm Fig 6 7 1 Axial Distribution of Neutron Flux 10 102 TOTAL gt 1 0 MeV gt 0 1 MeV D
114. 3C3 3D3j 3E3 5A2 5D2 23 24 Fig 11 1 1 Table 11 1 1 1 0 1 2 2386250 B 3 2 11 4 6 3 E 5 1 D 3 2 4 2 B RE 0 1 MeV 2 3 0450 Total 1 6 ZATE 5 CMIR SMIR 1 3 63 339 350 152 SN9410 89 089 11 11 Neutron Flux of Irradiation Rigs Neutron Flux n cm sec 2 418 10 2 440 10 2 2 022 10 2 035x10 3 009x10 3 028x10 2 243 10 2 295x10 1 883 10 1 920 10 2 866 10 2 913 10 2 899 10 2
115. EX 1 EX 2 n nnn 63 411 37 43 1AEA Stage 2 EX 1 A B 1 63 413 4 4 44 IAEA 430 49 Stage 2 EX 2 1 63 SN9410 89 089 45 IAEA Stage 2 EX 3 Oen 6 9193 62 60 5 mH 65 51 563 334 65 52 17 47 5 f 68 335 70 53 63 958 72 54 MK I 17 M ee 63 380 ie 55 17 verven 63 384 76 5 6 3 nM 62 396 18 6
116. Table 6 8 3 Neutron Fluence 05022 Ds023 CMAGI NEUPAC MAGI CMAGD NBOPAC MAGI NEUPAO OTAL 1 893E 21 1 792E 21 1 056 00 8 263E 21 8 165 21 1 012 00 2 074 21 1 817 21 1 141 00 OVER 10 2 492E 19 3 269E 19 7 624 01 7 331E 20 1 292 21 5 675 01 4 545 19 6 108 19 7 441 01 OVER 0 1 5 0088 20 5 697E 20 8 791 01 4 6438 21 5 302 421 8 758 01 7 552 20 7 469 20 1 011 4 00 Table 6 8 4 Distribution of Neutron Energy Ratio of Fluence to Total 05021 05022 5023 NEUPAC MAGI NEUPAC 1 000 00 1 000 00 1 000E 00 1 000 00 1 000 00 1 000 00 OVER 10 1316 02 1 824 02 8 872 02 1 582 01 2 191 02 3 361 02 OVER 0 1MEV 2 646E 01 3 178 01 5 619 01 6 493 01 3 641 01 4110 01 Table 6 8 5 Axial Distribtion of Neutron Fluence Standardized by the value at Core Center DS 023 MAGI NEUPAC MAGI MAGI NEUPAC 2291 01 2 195E 01 1 000 00 1 000 00 2 510E 01 2 226 01 OVER 1 0 3 399 02 2 530 02 1 000 00 1 000 00 6 200E 02 4728 02 OVER 0 1 MEV 1 079E 01 1 075 01 1 000 00 1 000 00 1 627 01 1 409 01 5 9410 89 089 Total Neutron Flux n sec 100 MWt Total Neutron Fluence nZ cm 10 1073 1015 1075 1 0 32 1 07
117. 141 SN9410 89 089 5 2 B JOYO 0 4 ppm Na BEF NaBF NaBF 4 142 SN9410 89 089 x10 2 E Ole AE 2 59 x10 9 u m et amp E 4 x10 10 0 10 20 Standing Time day Fig 9 8 1 Hydrolysis of Sodium Fluoborate at About 20 C 30 559410 89 089 10
118. 3 360 00 11 TD 2B 356 00 Outlet Sodium Temperature 2B Outlet Sodium Tempereture 34600 Outiet Sodium Temperature 2 85 Outlet Sodium Temperature 1A F E 8 x or Inlet Sodium Temperature B Inlet Sodium Temperature lt 8 mm E 2 8 37 8 Outlet Sodium Temperature 1 x a RJ 5 U u lt 8 S ug bi Intet Sodium Temperature as 88 Inlet Sodium Temperature 8 8 E 8 Subassenbly Outlet Temperatura 000 8 Eg 2 a Neutron Flux 6 2 5 84 00 6 00 68 00 14200 216 00 29000 364 00 438 00 51200 58600 66000 734 00 868 00 892 00 56 00 103000 110400 89 1 30 TIME SEC 10 15 12 58 PLANT STABILITY TEST DHX 95 MK2 1 Fig 5 4 3 Plant Stability Test 17 Cycle 5 9410 89 089 55 17 Gr 1 MK 17 BOC 17 16 17
119. 5 ke 239 241 8 4 f eo Co waa 844 866 690 794 956 730 875 95 5 98 e 680 68 0196 5 T 3 2 3 10 11 12 1 2 3 4 5 6 7 8 9 10 11 12 1 2 3 1011121 2 3 18 19 20 20 MK IHE A C FFD R 680 68 OTVENS ONd 5 9410 89 089 12 4 ICRP Pub 26 Ae Gr 1 8 2 1 2 6 1
120. 6 2 1 Table 6 2 2 63 337 SN9410 89 089 Table 6 2 1 Result of Neutron Flux Measured of CA 02 Sets Table 6 2 2 Results of Neutron Fluence Measured of 02 Sets gt 1 0 MeV 2 25 10 cem gt 0 1 4 58 1018 SN9410 89 089 6 3 Gr 1 MK 2 SBI MK 1 01 M 1 I 01 R 2 27 887 75 MW 3 2 X 10 sec Ni Cu NEUPAC JLOG 3 5 NEUPAC ncrrf sec BS dpa sec OMAK 4
121. 892x10 2 411 10 2 403 10 3 619 10 3 620x10 1121x107 1 1118 10 SMIR 10 9 464 10 9 439 10 505 1 697 105 1 701x107 1 Axial Postion from Core Center gt 0 1 MeV 2 Core Average E gt 0 1 MeV 3 Core Average Total 153 559410 89 089 Fig 11 1 1 Neutron Flux at Core midplane 154 Flux x 105 em see Address 23 cycle 24 cycle 24 cycle 23 cycla 5 9410 89 089 1 2 MK 1 Hm XE MK 1 5 K
122. Fig 6 7 2 5 BF 1 0 MeV 0 1 MeV MAGI Table 6 7 3 Table 6 7 5 MAGINEUP 0 62 1 20 fk 63 387 5 9410 89 089 Table 6 7 1 Total Neutron Flux TOTAL 4739 14 982 1 027 15 983 4 875E 14 9 89 OVER 1 0 6 001 E 12 18 02 2 803E 13 17 74 8 213 12 17 96 n Zem2 sec Z100 MWt OVER 01 MEV 13866 14 14 93 4 005E 14 14 23 1704 14 14 69 5 745B 08 12 83 L610E 07 11 79 6 7948 08 12 37 dpa 7100 MWt Table 6 7 2 Total Neutron Fluence 0841 Z 452 0842 2 5 0843 2 358 TOTAL 2 933 21 9 82 6 355 21 9 83 3 017 21 9 89 OVER1 0 MEV 3 7148 19 18 02 1 735E 20 17 74 5 082 19 17 96 OVER0 1 8 577 20 14 93 2 478E 21 14 23 1 054 21 14 69 3 555 01 12 83 9 963E 01 11 79 4204 0101237 dpa Standard Deviation in Percent Z Distance from the Core Center mm Table 6 7 3 Neutron Fluence DS 41 DS 42 DS 43 MAGI MAGI NEUPAC MAGI NEUPAC ngop NEUPAC TOT 2 172 21 2933E 21 7219E 01 4231E 21 6 355E 21 6 657E 01 2 166E 21 30178421 T 180E 01 OVER 1 0 MEV 4 463R 19 3 714 19 1 202
123. IHX Pri Main Pump OFC A B TLD CP Ge Ge BOOOERL Ge LC 7 2044 CP Fig 6 5 1 K 20 3 1 CP Co 3Co Fig 6 5 2 fc Mnh Cold leg 3 8 Co Hot leg 0 31 4Ci cm Cold leg 2 0 18 WACi Ek 8Co Hot leg 0 05 cm Co Na
124. No D 7 8 4 Room Temperature 20 Temp Defernce between surface of wrapper wall along the axes INPUT DATA for BEACON code Temp data gt One point par wrapper tube face 70 00 170 00 200 00 370 00 470 00 570 00 670 00 770 00 800 00 870 00 970 00 1040 00 1070 00 1170 00 1270 00 1340 00 1370 00 1470 00 1570 00 1670 00 1770 00 1880 00 1880 00 1990 00 2000 00 2110 00 2330 00 2330 00 2440 00 2440 00 Do FP N 5 9410 89 089 4 25 TEMP DATA for IAEA TEST EX 2 Number of Assembly i Number of NODE 30 Wall No R7 gt R1 11 L7 Input No 1 7 8 WRAPPER TUBE NO Room Temperature 20 Temp Deference between surface of wrapper wall along the axes INPUT DATA for BEACON code Temp data gt One point par wrapper tube face 70 00 170 00 200 00 370 00 470 00 570 00 670 00 770 00 800 00 870 00 970 00 1040 00 1070 00 1170 00 1270 00 1340 00 1370 00 1470 00 1570 00 1670 00 1770 00 1840 00 1870 00 1970 00 2070 00 2170 00 2310 00 2370 00 2410 00 2440 00 SN9410 89 089 43 Stage 2 EX 1A 1B 1 Gr 1 IAEA IWGFR STAGE 2
125. between experiment and calculation result CAI mm Node 2 440 20 19 18 17 16 1 830 15 14 13 12 11 1 220 10 9 8 7 6 610 5 4 3 2 1 0 Fig 4 4 2 1 o 9 1H e TI HEH Experiment Calculation 60 40 20 0 20 40 60 Comparison between experiment and calculation result 680 68 0196 5 CASH SIA mm Node 2 440 20 19 18 17 16 1 830 15 14 13 12 11 1 220 10 9 8 7 6 610 5 4 3 2 1 0 Fig 4 4 2 2 o o cO 9 o o EH Experiment Calculation 60 40 20 0 20 40 60 mm Comparison between experiment and calculation result mm Node 2 440 20 19 18 17 16 1 830 15 14 13 12 11 1 220 10 670 Fig 4 4 2 3 o HEH Experiment Calculation 60 40 20 0 20 40 60 mm Comparison between experiment and calculation result 680 68 OIL6NS mo C a mm Node 2 440 20 19 2 18 17 16 8 1 830 15 14 13 12 11 d 1 220 10 9 1 8 as 7 d 6 670 5 4 3 KOH Experiment Calculation 1 0 60 40 20 0 20 40 60 mm Fig 4 4 2 4 Comparison between experiment and calculation result ASH STA mm Node 2 440 1 830 1 220 670 Experiment
126. 0 2 380 0 300 TIME SEC Fig 7 4 1 Uncontrolled Control Rod Withdrawal Accident at Normal Operation 16 940 14 860 12 77 S 10 g 3 S 8 5 620 s 2 Fuel Maximum Temperature S 6 5 sob E Clad Maximum Temperature 4 5 460 Coolant Maximum Temperature Wa 2 380 Melted Fuel Fraction 0 300 0 4 8 12 16 20 24 28 36 40 TIME SEC Fig 7 4 2 Uncontrolled Contrd Rod Withdrawal Accident at Reactor Start Up 5 9410 89 089 8 115 5 9410 89 089 81 3 Gr 1 5 3 63 374 116 SN9410 89 089 D METHUSELAH METHUSEBILLAH
127. 00 40 00 40 00 90 0 CONTROL SIGNAL AN 4 00 5 38 6 76 8 14 9 52 10 90 12 28 13 66 15 04 10 42 17 80 19 18 VANE SIGNAL 4 00 15 52 DAMP SIGNAL MA Fig 5 1 2 2 Vane damper characteristic 1988 12 7 SN9410 89 089 3 5 s 53 OPENING f 40 0 CONTROL SIG NAL 00 5 41 5 81 8 22 9 53 11 03 12 44 13 85 15 25 16 66 18 07 1947 VANE SIGNAL 4 00 15 52 DAMP SIGNAL MA fig 5 1 3 18 Vane damper characteristic 1988 12 7 28 VANE x DAMPER E X T LAN mim THENE eRe s ENS miss minuna psp nep app BHESZSEHLWESEHTENEREBNEZIBEE Sea HH OPENING E e e EN 0 Vee AA 21 22 11 42532 4 00 13 03 13 85 VANE SIGNAL Rig 5 1 4 2B Vane damper characteristic 1988 12 7 4 00 15 52 DAMP SIGNAL 5 1 1 set values of Vane damper examination date 7 th December 1988 setting date 10th December 1988 calculating value re setting value setting setting opening setting point remarks value value Bam me a fee
128. 1 110 16 15 9 14 13 8 12 11 6 1220 10 9 5 8 7 4 6 5 3 4 2 3 2 0 Upper core Structure Sleeve Deflection 530 Point Temperature Experiment Analysis Model Fig 4 3 1 Subassembly Model mm Node Cose BX1A mm Node Case No 2 440 20 2 440 20 19 19 18 4 18 17 17 16 16 1 830 15 1 830 15 14 14 13 13 12 12 11 1220 L 10 1 220 10 8 8 7 1 610 5 610 54 4 46 l O4 Experiment P 2 T Experiment Calculation 1 16 Calculation 0 20 10 0 10 20 mm 0 10 20 mm Fig 4 3 2 Comparison between Experiment and Calculation Fig 4 3 3 Comparison between Experiment and Calculation result for EX1A result for EX 680 68 0196 5 5 9410 89 089 44 7 2 STAGE 2 EX 2 1 1 BH IAEA ZIWGFR STAGE 2 EX 2 EHE 10 2 80 10 0
129. 1 I 179 9507 1745 10 9 8 0 0 59 3836 e Analysis model Experiment Fig 4 5 1 Subassembly Model 5 9410 89 089 measurement 1000 analysis 800 600 400 200 0 10 20 30 displacement of subassembly Fig 4 5 2 CASE 1 Relation between top displacement of subassembly and load measurement analysis 300 200 100 0 10 20 30 em Top displacement of subassembly Fig 4 5 3 5 Relation between top displaccment of subassembly and strain SN9410 89 089 Load O measurement 1000 analysis 800 600 400 200 0 10 20 30 displacement of subassembly Fig 4 5 4 2 Relation between top displacement of subassembly and Load Strain measurement analysis 300 200 100 0 10 20 30 cm Top displacement of subassembly Fig 4 5 5 CASE 2 Relation between top displacement of subassembly and Load SN9410 89 089 5 BOEXERPO 77 v SN9410 89 089 51 MK 17 MBE
130. 2 Cr temp 2 temp 2 C1 temp 2 temp 2 temp 2 temp 2 Cl temp 2 temp 2 e 11111111 11 temp 3 Cr temp 3 Cr temp 3 Cr temp 3 Cr temp 3 Cr temp 3 Cr temp 3 Cl temp 8 Cl temp 3 Cl temp 3 Cl temp 3 Cl temp 8 Cl temp 3 Cl temp 3 Cl n i LD Cr p i 1 Cr p i 2 Cr p i 2 Cl p i Cr p i 7 SN9410 89 089 Table 4 2 3 TEMP DATA for IAEA BENCH MARK TEST EX 1A Number of NODE 30 Wal R7 gt R1 11 gt L7 Input No 0 8 gt 14 LLL LLA TUBE NO Room Temperature 20 Temp Deference between surface of wrapper wall along the axes INPUT DATA for BEACON code Temp data One point par wrapper tube face Temp 6 10 00 170 00 200 00 370 00 470 00 570 00 670 00 770 00 800 00 870 00 970 00 1040 00 1070 06 1170 00 1270 00 1340 00 1370 00 1470 00 1570 00 1670 00 1770 00 1840 00 1870 00 1970 00 2070 00 2170 00 2310 00 2370 00 2410 00 2440 00 omn 035 Qo WN 5 16 MYM Ww HY 9 C2 DN e a Cc o t0 OO e e 559410 89 089 Table 4 24 TEMP DATA for IAEA MARK TEST EK 1B Number of 30 Wall No R7 gt L7 Input
131. 2 33 34 35 36 Operation Cycle Calculated Excess Reactivity between 31th Cycle and 35th Cycle 680 68 OIL6NS SN9410 89 089 1 6 MK ERE T R Gr 1 MK Il R 2 ORIGEN 79 3 Neutrons sec ii Photons sec 4 J 2 0 2 3 MW SA 75000 MWd t 2 60 85 1 445 90000 MW d t 85 Case 1 15 1 45 60 oCase 2 15 1 70 85 H 3 90 000 MW d t 85 H 3
132. 494 490 47 488 404 4 9 1 4 4 6B5 487 496 494 489 494 495 494 401 496 487 493 2 2 5 5 6 6 491 498 495 491 499 500 496 499 500 491 496 7 3 4 sl 479 489 487 486 484 491 482 486 491 482 487 480 418 481 484 487 474 471 475 476 473 487 m 0 c lt lt lt u t wm 0 r 100 o gt itr lt 5 lt lt p 1 lt lt 1 lt SO mi OO lt h 42 O 4 cO e OOo wo sc 6C2 539410 89 089 36 17 MAG Bis Gr l H 17 BOC K 1 5 64 C 0 144 Ak MAGI MAGI 17
133. EF 5 1 3946406D 13 COEF 5 3 7531604 14 2 Total Worth Ak k tk 63 364 Excess Reactivity Ak Zk 350 400 450 Fitting equation Y XiCoef i x X X Rod Stroke Excess Reactivity COEF 0 4 19203329D 00 COEF I 1 824122150 02 COEF 2 5 48096903D 05 COEF 3 1 27101089D 07 COEF 4 1 54749358D 10 5 lt 7 07661008D 14 500 550 600 Rod Stroke mm Fig 3 2 1 Stroke Curve 1 650 Excess Reactivity 4k k 400 450 Fitting equation Y DCoef i XX Stroke Excess Reactivity COEF 0 1 69054778 0 00 4 63475015 D 03 2 2 94462500D 05 COEF 3 2 81554285D 08 COEF 4 1 209691870 11 COEF 5 1 85613400D 14 500 550 600 Rod Stroke mm Fig 3 2 2 Stroke Curve 2 650 680 68 0196 5 Excess Reactivity 8 Ak Zk 400 Fig 3 2 8 Fitting equation Y XOoefli x x Rod Stroke Y Excess Reactivity COEF 0 2 69289534D 00 COEF 1 6 46122238D 03 2 1 69275721D 05 COEF 3 6 61588116D 08 COEF 4 1 06374971D 10 COEF 5 5 57165020D 14 450 500 550 600 650 Rod Stroke mm Stroke Curve 3 Fitting equation Y
134. Gr 1 MK 17 2 T2 D 0 5 65419 1 5 3 1 16
135. PA p 1021 10 5 5 8 amp 2 10 5 2 1019 1077 600 400 200 9 200 400 Distance from the Core Center mm Fig 6 7 2 Axial Distribution of Neutron Fluence 5 9410 89 089 6 8 SMIR 2 HEE Gr 1 MK H SMIR 2 C SMIR 2 1 0 MeV 0 1 MeV 2 1 SMIR 2 MK 8 1984 4 19 1984 6 10 4437 MWd 2 SMIR 2 503 3 Fe Cu Co Al 1 NU EU V 3 SMIR 2 NF END B V DOT 3 5 NEU
136. PAC Cn cm sec dpa sec m Table 6 8 1 Table 6 8 2 24124139 Fig 6 8 1 Fig 6 8 2 1 0 MeV 0 1 MeV MAGI Table 6 8 3 Table 6 8 5 MAGI NEUP 0 56 1 14 H 63 388 Table 6 8 1 Total Neutron Flux 05021 2 612 05022 2 452 D8023 2 457 4 675 14 9 54 2 130 15 6 41 4 740 14 9 52 OVER 1 0 MEV 8 527E 12 17 78 3 370 14 8 46 1 593E 13 16 75 n em7 sec LOOMWt OVER 0 1 MEV 1 486 14 14 71 1 383E 15 9 98 1 948E 14 13 93 5 950 08 12 29 6 505E 07 6 72 7 9318E 0811 15 dpa ZsecZ 100 MWt Table 6 8 2 Neutron Fluence TOTAL 1 792E 21 9 54 8 165 21 6 41 1 817 21 9 52 OVER 1 0 3 269 19 17 78 1 292 21 8 46 6 108E 19 16 75 OVER 0 1 5 697 20 14 71 5 302 21 9 98 7 469E 20 13 93 2 281E 01 12 29 2 4946 00 6 72 3 036E 01 11 15 aa Standard Deviation in Percent 2 Distance from the Core Center mm
137. S S 5 24 Gr l 1 7 BIC LAAIERREROR FPGS 3 5 2 PFD156 41000 MWd t 4D1 0 8 0 9 134 350 a b c d 3 55 cm E FPGS 3 5 DATA MAGI a MAGI
138. TOR BUBBLE DECONTAMINATION SOLUTION pH 4 5 120 F 0 200 400 600 800 1000 1200 1400 TEST RELEASE PRESSURE psi g Fig 8 1 2 DECONTAMINATION OF IODINE IN FUEL HANDLING ACCIDENT 124 5 9410 89 089 9 125 5 9410 89 089 91 Na I FM Gr 1 H X Na Nal 9 Sid Na 2 Fig 9 1 1 2 Nal 500 C 650 C 800 C 3 Na Na 1 3 R Table 9 1 1 Na 500 14 650 14 800 C 7
139. c o MK 1 FOR PUBLICATION PNCT N9410 89 089 May 1989 Preliminary Report on Experiments Analyses and Evaluations Performed in Reactor Technology Section Experimental Reactor Division Quarterly Report Vol 5 No 4 Toshihiro Odo Yoshio Arii Takayoshi Kobayashi Makoto Sawada Yuuichi Shimada Toru Sone Toshio Funada Yoshioki Yamashita Abstract This report summarizes results on experiments analyses evalua tions performed by Reactor Technology S
140. e experiments performed on spent fuel assembly of JOYO have been shown that 7 spectrometry of the Rh 14Cs Cg and 144Pr be used as burn up monitor routine basis analysis Other fission products such as Zr Nb and also were detected in spent fuel assemblies but have no potential interest for the development of correlation since their half lives are too short The correlations between measured activity and calculated burn up MAGI were made on spent fuel assemblies with different irradiation history and irradiation address have been shown that Ce Pr and Ru Rh have best chance to be utilized for the determination of relative burn up value in speut fuel The Cs and 4Cs also can be used but with low sensitivity correlation ones as shown in Fig 6 13 1 63 449 105 901 Relative Gamma ray Activity Cps Fig 6 13 1 106 Ru 06 Rh 1050 3684 10000 20000 30000 40000 50000 Burn up MWd t Relationship between Measured Activities and Burn up Calculated by MAGI 60000 680 68 0196 5 ONd SN9410 89 089 7 07 SN9410 89 089 7 1
141. ection Experimental Reactor Division during January through March 1989 Each result described in this report was reported as the internal memoranda of Reactor Technology Section for further analyses evaluations and or discussions This report contains the following items 9 Results of measurements analyses and evaluations for nuclear Characteristics of JOYO 9 Results of analyses and evaluations for core mechanics 9 Results of measurements and analyses for the plant characteristics of JOYO Results of measurements and analyses of neutron flux gamma ray decay heat of JOYO Preparation of Constr ction Permit Amendment for the operational reliability tests of FBR fuel assembly Preparation of Construction Permit Amendment for spent fuel strage facility Reactor Technology Section Experimental Reactor Division Engineering Center PNCTBN9410 89 089 Results of study and development for analytical technic Production and or arrangements of analyses codes and their manual Analyses and evaluations Core Miscellaneous results The final report will be published for each program after further discussions analyses and evaluations 5 9410 89 089 E II X 1 2 eee eH 2 bd ed ect ee ee er ee ene ere 2 22
142. ed in JOYO 1 core to investigate the irradiation performance of neutron absorber materials Reaction rate of three dosimeter sets which had been loaded in this subassembly were evaluated Utilizing these measured reaction rates the neutron spectra at dosimeter irradiation positions were unfolded with Jl type unfolding code The total neutron flux and fluence fluence above 0 1 MeV fluence above 1 0 MeV and displacements per 2 Irradiation Condition Irradiation Time Accumulative Power Rated Power Effective Irradiation Time Irradiation Core Address Dosimeter Materials 3 Result of Evaluation Position of Dosimeter Set from Core Cente 27 Midplane mm Total gt 1 0MeV gt 0 1MeV PS atom were calculated from the unfolded spectrum MK 8 12 duty cycle December 1 1985 December 6 1986 P 22 322 MWd 100 MWt 1 93 10 seconds 6C6 Fe Ni Cu Al Al Fluence n cm Displacements 7 15E 20 10 6 3 38E 18 18 6 1 558 20 16 1 6 76E 02 14 4 1 662 22 10 5 1 42B 21 16 3 8 73 21 13 4 3 85 00 10 7 177 21 10 0 L15E4 19 18 03 5168 20 15 7 1 52 01 13 8 Note Read error 10 Ratios of MAGI to calculational results were within the range between 0 76 and 1 37 at the level of 44 mm fromthe core midplane 63 427 27141225 5 9410 89 089 6 12 FPG
143. on No 0 9 252 283 257 309 252 Table 4 1 2 Tamperature at the point on the circumference of wrapper tube wall Axial level 770 mm Normal Temp Cr temp C1 7 0 000 Cr temp 1 6 014 Cr temp 1 5 072 Cr temp 1 4 140 Cr temp 1 3 330 Cr temp 1 2 590 Cr temp 1 1 1 000 Axial level Cl temp 1 Cl temp 1 temp 1 temp 1 C1 temp 1 Cl temp 1 C1 temp 1 Cr temp 2 Cr temp 2 Cr temp 2 Cr temp 2 Cr temp 2 Cr temp 2 Cr temp 2 temp 2 Cl temp 2 Cl temp 2 Cl temp 2 Cl temp 2 Ci temp 2 C1 temp 2 1 1 000 2 820 3 380 4 190 5 096 6 014 7 0 000 7 0 000 6 016 5 031 4 094 3 410 2 660 1 1 000 1 1 000 2 750 3 350 4 160 5 0 000 6 120 7 0 000 SN9410 89 089 Axial 2300 mm Cr temp 3 7 0 000 temp 3 6 083 temp 3 5 210 Cr temp 3 380 Cr temp 3 630 Cr temp 3 920 Cr temp 3 1 1 000 Cl temp 3 1 1 000 Cl temp 3 2 760 temp 3 3 480 Cl temp 3 4 200 Cl temp 3 5 040 Cl temp 3 6 0 000 Cl temp 3 7 0 000 Cl p i 1 Cr p i 1 Cr p i 2 Cr p i 2 Cr p 3
144. um Surface GL 9490 Top of Fuel Assembly GL 9200 M U A Inside of Equipment i 1 1 LI MI Cable 1 Data Logger Inside of Equipment lt Connector Ar Gas Sodium Electical Boiler Outline of Experimental Equipment for Acoustic Boiling Detection 680 68 0196 5 5 9410 89 089 2122 CGCS Gr 1 H X CGCS 8 PP 7 FFPD R 407 HNH 63 8 22 9 19 2 ik Ar WA FP Xe Kr CGCS 2
145. x 10 cps CG CPS 63 434 5 9410 89 089 74 MIMIR N2 z KICKS P TM Gr 1 MIMIR N2 PTM 2 1 2 2 D Table 7 4 1 2 PTM Table 7 4 2 3 1 15 C 11 Fig 7 4 1 2 1236 ko Ci Fig 7 4 2
146. ysis of the Safety Vessel Penetrating Part of Return Pipe 680 68 0196 5 539410 89 089 6 amp Ami SN9410 89 089 61 ATR Gr ATR R amp D Fe Cu Nb 3 Fe Cu 7 7 BOB75 Nb 3 Ge Li K XMO ATR Measured Specific Activity at the end of Irradiation Specific Activity decay Z sec Z g EE aor io
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