Impact of extended burnup

Transcription

1 IAEA-TECDOC-699 Impact of extended burnup on the nuclear fuel cycle Proceedings of an Advisory Group Meeting held

2 The IAEA does not normally maintain stocks of reports in this series. However, microfiche copies

3 IMPACT OF EXTENDED BURNUP ON THE NUCLEAR FUEL CYCLE IAEA, VIENNA, 1993 IAEA-TECDOC-699 ISSN Printed by the IAEA in Austria April 1993

4 FOREWORD Currently, burnup extension is one of the most advanced and intensively studied areas in fuel technology and management worldwide. The average design discharge burnups that

5 EDITORIAL NOTE In preparing this material for the press, staff of the International Atomic Energy Agency have mounted and paginated the original manuscripts as submitted by the authors and given some attention to the presentation. The views expressed in the papers, the statements made and the general style adopted are the responsibility of the named authors. The views do not necessarily reflect those of the governments of the Member States or organizations under whose auspices the manuscripts were produced. The

6 SUMMARY CONTENTS

7 PAPERS PRESENTED

8 SUMMARY

9 2.2. Impact

10 It

11 With respect to burnable absorbers the higher loadings needed with extended burnups gives an increased incentive to use integral fuel absorber designs. There are many integral fuel absorber designs available

12 used

13 3.3. Mechanical design Structural design impacts As mentioned previously, there

14 discharge burnups, because the dominant cost component is the fabrication cost, which is essentially independent of burnup. Another factor for MOX is that the isotopic composition of recovered Pu will change depending on the discharge burnups of the fuel from which the Pu is recovered. This has a minor impact on the nuclear design of MOX cores. 4. IMPACT OF EXTENDED BURNUP ON THE BACK END OF THE FUEL CYCLE 4.1. General issues Criticality

15 4.3. Transport Transport casks

16 3. The programme to assess and benchmark computer codes for fuel thermal-mechanical performance at high burnups should be continued. 4. Measures to decrease uncertainties in predictions of isotopic composition and decay heat generation

17 PAPERS PRESENTED AT THE ADVISORY GROUP MEETING

18 DEVELOPMENTS

19 TABLE 4. ENRICHMENT PLANNED STEPS OF HOMOGENEOUS SEU ATUCHA-1 CORE

20 TABLE

21 W N) Other consequences of core enrichment (see Table 6), are: - extension of fuel residence time, - reduction of fuel assemblies consumption and lower frecuency

22 7. SEU PROGRAMS FOR OTHER POWER REACTORS Argentina

23 3.8m

24 Fuel channels Inside oooooooooooooooooooooo oooooooooooooooooooooo oooooooooooooooooooooo oooooooooooooooooooo oooooooooooooooooo oooooooooooooooooo. oooooooooooooooo ooooooooooooo oooooooooooo

25 ro CD 250 >, 200 -i D 0 a PWR D 3.25% enrichment o 4.25% CANDU, 13 To r> c ISO- Natural Uranium B 0 A» CO D) 'I

26 Figure 7: Comparison of CANDU and PWR Fuelling Costs This then yields the following value for the price of RU: CRU 1/Fj

27 ro

28 Channel K-11 (5543 kw) Ä 500. o> 1 o. to c m 5 Axial Bündle Position n 3 Ù- 0 Bündle 1 g 0~?"0 IL T j U -D 600! S, 500- <D Ï 300 t 7 à 200

29 CO o Strategies

30

31 CO (VI 4.2

32

33 GJ -t». The Qinshan project

34 3. ECONOMIC EVALUATION FOR FUEL CYCLE COSTS There are some analysis models and codes for fuel cycle econonic have been developed

35 G3 O> The basic research includes:

36 PRESENT STATUS

37

38 Alternative Fuel Cycles Comparison CSFR Fuel Fabrication Alternative Fuel Cycles Comparison CSFR Cumulative Fuel Fabrication Alternative Fuel Cycles Comparison CSFR Spent Fuel Disposal Alternative Fuel Cycles Comparison CSFR Cumulative Spent Fuel Disposal Fabrication. ttu Spent Fuel.

39 -p.. o SO 40

40 Alternative Fuel Cycles Comparison CSFR RA Waste Storage Alternative Fuel Cycles Comparison CSFR Funds Available Spent f net. ktu

41 Similar 3.3. BACKEND

42 k$/kgü 1000 S/kg

43 5 COHCLUSIONS At present the average fuel burnup in Czechoslovak WER reactors is 30 MWd/kgU with

44 2 The fuel cycles 2.1 TVO I and TVO II (twin ABB Atom BWRs) Front end (TVO) Öl

45 O)

46 3.2.2 Loviisa

47 -pk 03 tion, in the international Battelle High Burnup Effects Program

48 regards reactor power densities and linear powers or discharge burnups have been avoided.

49 Öl o In this respect, c/kwh

50 2 - ENRICHMENT AND FUEL MANUFACTURING

51 Öl ro 8 H G F E D C B A H G F E D C B A Batch? Batch

52 6 - STORAGE OF SPENT FUEL ELEMENTS Spent fuel storage

53 Oi

54 DATA Burn-up Cooling time U235 initial Fission products : Including solubles insolubles gazeous FP activity Activation products. hull. end fittings Actinides :. Uranium 232. Uranium 235. Uranium 236 Total Plutonium including : TABLE 1 HBU FUELS - COMPARED CHARACTERISTICS ESTIMATED VALUES Data refers to 1-metric ton of initial uranium (MTU)

55 Oi

56 Total Reprocessing waste volume EXTENSION OF BURNUP IN INDIAN PHWRs

57 CJ1 00 either plutonium or U.In the third stage, Th-U reactors would be the source

58 chosen. The core is divided into two or three radial zones to achieve radial power flattening in equilibria» condition and in each zone a discharge burn up Öl CO

59 CD O TABLE-2 Microscopic Fission and capture Cross Sections( barns) at 2200 m/s neutron speed for Uranium and Plutonium Isotopes Isotope U-238 Pu-239 Pu-240 Pu-241 Vu-242 Fission Capture Beta energies

60 with significant difference in irradiation or of natural uranium bundles and MOX bundles would lead to power peaking Reactivity Coefficients

61 The licit on the power in each channel according to the coolant flow in that channel, fixed

62 5.1 Equilibrium Reactor For the MOX-7 reactor, a central region with all natural uraniun fuel

63 TABLE - 4 Burnup (GWD/T) O.O(NoXe) 0.0(with

64 0.2 Target Burnup Extended burnup o> UI

65 O) o> New

66 upper Tie-Plate Large Center later Channel Ferrule Type Spacer Ferrule Type Spacer Lower Tie-Plate Loier Tie-Plate TypeA TypeB

67 CO Though higher burnup fuel TJ C R V a l u S 3 p g o O 01 o - O- XJ \ O O o o 3 T3 t» 0 E) O CO o co H W T) CJ -I \ «-* ö c *-t 3 CD 3 en O Q. \ r-r in W ^d a CD CO CD

68 OJ CD 0. 5 o a: 0

69 -4 O fabrication service o M- m R t i n w 01 ~r~ o T~

70 z S c H l D O (N n \ s Harden > Neutron Spectrum f N \ Increase Void Coefficient * Increase AMCPR during Transient

71 The main effects of the burnup extension in the nuclear property

72 1.6 Parameter : Enrichment, H/U 12H/0 H/U H/0 H/U-

73 because the point will be far from peak point and K-infinity increase will be very small.

74 defference between fuel bundles.

75 -J O)

76 reactor coolant water chemistry on zirconium alloys and to accelerate

77 --J (scrap uranium iuo., Powder] Solution* (Tank Diameter) PH Control* (Tank Diameter) Powder Arrival Storage* (Can Spacing) Precipitation i Mixing* Mixing Machine> (

78 -vl (O As for ventilation, drain and solid waste control, the impact

79 CD O 4

80 l

81 GO ro *o

82 HIGH BURNUPS FOR WATER REACTOR FUELS - A UK PERSPECTIVE K HESKETH British Nuclear Fuels pic, Salwick, Preston, United Kingdom Abstract CD W

83 03 -t* Wolf Creek

84 Table 1 Impact of High Burnup Fuel Cycles on Core Characteristics Parameter Radial peaking factor

85 CD O) restrictive

86 EFFECTS OF EXTENDED BURNUP ON THE NUCLEAR FUEL CYCLE P.M. LANG United States Department

87 00 00 unchanged cycle length. For reactors operating at the historical burnups in the neighborhood of 30 MWd/kgll, approximately half of this uranium saving can be realized if burnup and cycle length are simultaneously increased, to the higher values

88 CD CO To establish utility confidence and regulatory acceptability for extended burnup fuel design

89 CD O pools.

90 IMPROVEMENT OF WER FUEL BURNUP AND

91 (O N> 2

92

93 CD 4. FUEL HANDLING IN THE CORE AND MONITORING OF WER REACTOR REACTIVITY BEHAVIOUR IN THE PROCESS OF FUEL BURNUP INCREASING 4 1 WER-440 In the beginning o± a tuel cycle the VVER-440 and WER-IOOO core possesses a maximum reactivity margin intended tor bringing the reactor to an operating condition. Fuel burnup during 7000 hr (292 eff.days)

94 ,1.0 0, Temperature,

95 0> reactor emergency protection (e g with 97 members ot control and protection system a repeated criticality at the end of a burnup cycle

96 <o application ot integrated burnable neutron poisons (uraniumgadolinium fuel

97 (O 00 ' r Fig.5 Decay heat unâer the conditions of long-tern storage of WEE-440 spent \ I0 2 9 Ö I0 1

98 4 Astakhov S A Proselkov VN et al Optimization ot WER cores to improve tuel utilization and reactor safetv. VANT. ser Fizika yadernykh reaktorov. No , Onutnev V D, Proselkov V N.Simonov V D Safetv assurance and economy improvement in water-cooled Soviet reactors with uranium-gadolinium fuel

99 o Safe transportation of NPP spent fuel is generally provided with

100 The TK-11 cask

101 o M packages. For example, with the TK-13 cask fully loaded and WWER-1000-type fuel being cooled during up to 3 years, its burnup

102 fuel and, consequently, its long-term storage in quantities exceeding capacities available

103 IMPACT ON THE BACK END OF THE NUCLEAR FUEL CYCLE

104 flssio«pbooucis

105 O a> 3. Spent fuel storage and final disposal. For a 36 ^ increased discharge bumup, the residual heat of spent fuel is increased by 26 \ after a cooling time of 1 yr and by 49 * after 5 yra ( 2 ). Since the decay heat is one limiting feature for spent fuel transportation, the storage period In the spent fusl pools at the reactor site might have to be Increased considerably. For Instance, taking the decay heat generation versus cooling time per assembly from the presentation

106 GWd/tU 33 PU SgB EOL Pu 236 Pu 236 Pu 233 Pu 240 Pu 241 Pu 242 Am 241 Reference 12X10" , ( 5 )

107 o o> ACTIVITY FACTOR ALPHA HEAT GENERATION Ü GAMMA NEUTRON -<200 KeV MeV > 1 MeV FIG.

108 based

109

110 LIST

111

112 based

113 ( 2 ) - Compatibility of Extended Bumup with French Fuel Cycle Installations.

114 ARGENTINA LIST OF PARTICIPANTS

115