ML20154H837
| ML20154H837 | |
| Person / Time | |
|---|---|
| Issue date: | 10/14/1985 |
| From: | Asselstine J NRC COMMISSION (OCM) |
| To: | |
| Shared Package | |
| ML20154H353 | List: |
| References | |
| FOIA-85-745 NUDOCS 8603100384 | |
| Download: ML20154H837 (367) | |
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10/4/55 Remarks by James K. Asselstine Commissioner, U. S. Nuclear Regulatory Cermis!'cn at the 1985 Internaticra! Meeting on Reduced Enrichment fur Research and Test Reacturs Petten, The Netherlands October f4, ICES A TECHNICAL SOLUt!Cf TO A DIFFICULT PROBLEM--AN UPDATE It is a pleasure.to be invited again to speak before *.his distinguished group, i Last year at your meeting at the Argonne fl4tional Laboratory, I spoke to you frec a regulator's viewpoint un effci*ts to rcve away from fuels that are potential nuclear explosive materialf tcwards icw-enriched uranium (LEL) fucis. i Tcday. I would like to repcrt en scre further progress in the Uni ted States in actually rek'ng conversion happen. , While we believe that currently licensed U.S. non-peser reactnes using highly
- enriched uranium (HEU) fuel are operated without unoue risk to the public health and safety, the Nuclear' Regulatory Ccnrdssion is in the final stages of approving a rule that is inter.ded to further reduce the risk of theft ar.d diversion of HEU fuel used at such reacters. , The NRC staff has also been pursuing security improvements at ren-power reactor facilities over the last sestral years. While some impreverents have been made, the Commission's staff l
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l- 9603100304 960121 PDR FOIA 3 PDR d j AFTERQOGS-745 L.
currently is ceveloping a rule to improve ' security egeir.st theft at licensee facilities possessirg cre to five kilograms of hEl:. As long as high enrichea uranium exists at non-power reactors, some residual risk cf treirvc!ent use exists, btile we_ have no indication of a specific threat at a domestic facility, recent acts by terrcrists abroad have shown that a thrut can materia!* e without sufficient warning. Therefore, we believe that 1; is prudent to consider adoitiona; ceasures to increase protectice cf l'EU materials. One approach of course is the removal of such material. For the short-tcm we have ordered non-power reattcr licensees to return their " excess" fuel to the Department uf Energy. Excess fuel is that materia: tcycrd the amounts needed fcr the maintenance of normal cperaticns or 'or the replacement of depleted fuel and failed elements. In addition to rrecval of the~ excess fuel, the- staff 15 also developing a security upgrade to make it mcre difficult to steal the fuel. Cencideration is being giver te ta:rper-proofed and line-superviseo cetection systems, access berricrc GNer the reactor core, requiring *he presence of two persors durfrc acceJs to material, and imprcvirg connunications capability wi9 frw enforcement organizations. In the long-term we intend that ncn-pcwer reactor ifcensees cervert to icw-enrichec uranium. To effect conversion the Ccmission is currently developing a rule which directs licensees to use low enriched fue! under specific circumstances. he have carefully considered its impact on lit.ensees. For those seeking a ccnctruction permit for a reactor requiring bEb fuel, the rule requires that applicants cemcnstrate that the reactor has a " unique" purpose.
Oricuc purpuse is defined as a project cr prcgram which cannot reasonably be acccmplished without the use cf HEU fuel in the reactor. If the app!*ct.rt carret reet this uniqueness test, we will not issue a license. On the other hand, we believe there is suffictent flexibility in the rcle tc perrit legitimate research using HEU fuel. The other part of the ruic hat t treater impact on licensees. Ur.less the licensee can r..eet the unique purpnse test, the rule prchibits the licensee from acquiring additional PEU fuel if LEU fuel is available anc such fuel meets the Commission's health anc safety criteria. In addition, the rule rcquires each licensee te replace all fuel in his pcssessicn with the available LEU fuel in accurdance with a schedule basec Lpen the availability of that fuel and cent'deration of other factors. These other factors include the availability of shippirg casks, reactor usage, anc financial support. The final conversten schedule for each licensee woulc be approved by the Corn'st on. i The precest we t:nvision for 1ccomplishing conversicn o# existing reactors is that if the replacceent ui HEU fuel by LEU fuel does not cht.nge. the technical specifications frecrperated in the license or involve an unreviewed safety issue as definec in our regulations; the licentce vill be permitted to replace I:EU with LEU without amendment to his license. Information available to date suggests that the conversicn of r. cst licersed derestic non-power reactors from HEU to LEO is technically feasible. And, if the crats c' the RERTR program are successfully achieved. cerversion will be technically feasible for most of the remair4rg recctors.
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-4 There is considerable ccncern especially a'mong acacemic int-itutions, which operate most of the U. S. research reactors, abcut the ecsts of conversion.
The way w4 intend to hardle the tiraing and costs of conversion is that each cen-pcwer licensee using HEb will submit & picpesed schedule for conversien basec upcr. the availability of replacerent fuel and shipping casks. P.crecser, the licensee must include a certification that federal governcent funcing for conversion is avai*eble from the Department of Energy cr other appropriate t-federal agency. If such funding is not available, the licensee need only certify the non-availability of federal funding and resubmit Fit prcposal at 12 month intervals. Conversian will not be required by the Coinmission until federal funds are aveilable. On this point, we fully expect that funds wilt be available. The Department of Er,ergy has historically provided significent support for research and test L' reacters. In fact, fuel for university reactors is currently supplied anc paid for by the Depertment of Energy -- that is the U.S. te.wpayer -- and " loaned" to universities. The U.S. Serate has recently approved 30.9 million for LEU fuel assistance to universities. The Commission has been in it,uch with the appropriatt Cercressional committees urgir.g Ccetgress to provide tre necessary for. ding for conversion. Although the detcils of the funding program rust rtill be worked out by the Congress ar.d the Eepartment of Energy, we wculd anticipate , t federal funding for such costs as overall fuel replacement (including, but not limited to, transportation and reasonable inst fuel values for any licensees who cwned the HEU fuel beirg replaced), support for safety analyses, and rassunable ' lost revenues" caused sulely by the conversion requirement. Cests not likely tn be covered by the Federel Gcverrment would include facility B 4 9 .
security costs unrelated tc conversion and some potential licensing costs - associated with obtaining license amendeents. Another re'nt which arose during consideration o# the propnsed rule involved the licensing issues which could arise as a result of conversirr. A rumber of licensees were concerred that there could be significant litigation involved in obtaining license amenoments to convert to LEU fuel. Many licensees expressed the fear that such litigatfor cec 1d result in the shut down of reactors if
. conversion meant any significant ccercric impacts on the universities.
It is not our intent to close down research rer.cters. The preliminary juegment of the staff is that all avaf'tble information continues to indicate that conversion of icw pcwer research facilities is technically feasible and will not involve any significant hazards consideraticns. :n those instances in which license amendments may be needed to accomplish conversier, it appears that pctertial intervenors support, rather than oppose, conversion. Thus, exterded hearings on license amendments to accorplish conversion appear te be unlikely. However, to avoid the necessity of litigation in indhidrel proceedirge, the Commission is considering directing our staff to develop generic envelopes of safety limits for several types of rcr-prwer reactors. In parallel with the presert rule ordering conversion, we are alsc considering undertaking a rulemaking concerning tho.sc cereric safety envelopes. The
. purpose of the rulemakino is te rr.ke c finding that safe operations are not significantly affected by conversien.
., 3,-* i We believe cerve.rsion of research reactors can prove tc be of benefit to , licensees, particularly to universities. The use cf LEU fuel avoids the need for more extensive security upgrades that are likely to be ccstly and riay be incccpat16ble with an academic atmosphere. Ir. addition, conversion removes the ; potential for theft or diverster of HEU. Diversion attempts, even li unsuccessful, would likely jeeparci:e continued operation of most, if rict ail, university reactors.
. To irp!ement this change of policy in the Uniteo States has been a long and
,' arduous struggle. Many consicer this change insignificant as compared with cther steps that could be taken to protect public health ano safety. Hcwever, it if an eu.cple of an issue which can be scIved through research and the irplementation of that research. In implementing the Cone.issier's FFU rerversion rule, we believe that the public health and safety will be ,
well-served and that the berefits outweigh the costs. A goal of the Reduceo Enricisent Research and Test Reactor Prceram (RERTR) estct;litted by the U.S. Department of Energy is to cevelop the techichi n.eer.s !
, recde.-d to utilize LEU instead of hEU fuels in research and test reactors ard tc i do so without operational penalties and wl.thout affecting 56fet*/
characteristics. This goal has been pursued now for seven years in cooperatier. with many of you who are participants in this conference. The prcr;rar has made I significant propress in the development, demonstration and application of new LEU rescacch reactcr fuels. According to necent pre.iections, the percentaga cf
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HEU fuel in inteinattenal ccmerce which is technically replaceable with LEU fuels has been grcHer; steadily as research continues. There appears to be no
7 techict.1 barrier to eventual replacerer.t ut hEU tuel with LEU fuel in research erd test reactors. While there are mar.y politically complex and technicel'y difficult issues involved tr. r.or.prollteration, I conclude that the use of highly enriched weapcns grade uranium as a fuel in civilian research reactors , stands out as cce problem which has a straight forward tectrical solution. The r technical solutien is the effort you in the RERTR program are etr. barked on. Once that solutton l's in hand, I urge you to take the pelitical and legal initiatives to make convert cr. happen. By effecting conversion you will help i minimite the trace in highly enriched ure.rS;m and make the world a little safer 1 for all of us. I
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FINAL RESULTS OF QUALIFICATION TESTING OF TRIGA FUEL IN THE OAK RIDGE RESEARCH REACTOR INCLUDING POST-IRRADIATICN EXAMINATICN l
! Robert H. Chesworth and Gordon B. West GA Technologies Inc.
P. O. Box 85608, San Diego, California 92138, U.S.A. As part of the qualification testing of TRICA (uranium sirconium-hydride, rodded) low enriched uranium fuel, a 16-rod experimental cluster (Figure 1) was irradiated in the Oak Ridge Research Reactor (ORR) (Figure 2) between December 1979 and August 1984. A total of 19 fuel rods, with three different fuel ocapositions have been tested - Er-U-IrH containing 20, 30 and 45 weight percent uranium. Individual rod powers have been in the range of 25 to 53 W and the test cluster operated in-core for 901 full power days. Some of the fuel test parameters are shown in Figure 3 In May 1982, af ter 575 full power days, the rods containing 20 and 30 weight percent uranius were removed, having attained a burnup in the range of 50 percent of the contained urantua-235
! This was higher than the initial target burnup values of 35 and 40 percent, respectively. Destructive examination of one rod each of the 20, 30 and 45 weight percent fuel rods showed normal performance with no anomalies. Figure 4 shows the three rods destructively examined in the suaner of 1983 They were the highest burnup 20 and 30 weight percent rods and one of the high burnup 45 weight percent rods.
Irradiation of the 45 weight percent rods was continued to the target burnup value of 50% and extended into the burnup range of 55 to 65 percent of the contained uranium-235, up to a test cluster exposure of 901 full power days. At this point, one of the fuel rods exhibited a clad failure due to steam buildup during a return to reactor power. Extensive analysis and post-l irradiation examination led to the conclusion that the failure l originated in a cladding defect which became manifest af ter four 1 2-
GENERAL LAYOUT OF 16-ROD FUEL CLUSTER
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LOCATION OF TRIGA LEU FUEL TEST CLUSTER IN ORR
" WEST 89.845 8 REGION NUAB8ERS 44 CONTROL RODS 75.395 -2
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,,, VOL-% U (19.7% ENRICHED) 7 11 20 $ MAX CALC R00 POWER GENERATION (KW) -
m INITIAL CONFIGURATION - 36 35 FULL CLUSTER CONFIGURATION 41 43 48 45 WT-% ONLY CONFIGURATION - - 55 TIME AT POWER (FPD) INITIAL CONFIG (DEC 79-NOV 80) 0 278 278 FULL CLUSTER CONFIG (MAY 81-MAY 82) 295 295 295 45 WT-% DNLY CONFIG (JULY 82-NOV 83) D D 328 TARGET BURNUP OF U-235 (%) 35 40 50 FINAL BURNUP RANGE (%) 45-57 47-57 60-66 I[7ss
TRIGA LEU FUEL DESTRUCTIVELY EXAMINED IN 1983 AFTER 1RRADIAT10N IN ORR ' i
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6 years and about 5 x 1021 nyt. Non-destructive examinatien of a large sampling (over 100)of similar cladding - Incoley-800 - has led to the conclusion that this was a unique defect and not a generic characteristic of this type or batch of cladding asterial. The in-pile testing of the experimental cluster was reinitiated in July 1984 but was terminated shortly thereaf ter due to a cladding break caused by mechanical rubbing of the cluster spacers which resulted fros' removal of the thermocouple support member. The test was tarainated'at this point since all of the test rods had met or exceeded target burnups without difficulty. The burnup history of the 45 weight percent rods is shown in Figure 5. A comprehensive post-irradiation examination (PIE) campaign has been established for the 45 wt-5 fuel rods as the final step in the qualification of TRIGA Icw-enriched uranium (LEU) fuel. The PIE includes: Dissetral and bow measurements of all of the 45 wt-5 rods Hessureaant of pressure inside intact rod 1089 or 1090 ; Identification and quantitative analysis of gas in free ' space inside of rod 1089 or 1090 Identification and quantitative analysis of gas contained in fuel matrix of rod 1089 or 1090 Gamma scan of rod 1086 to determine peak-to-average burnup - Measurement of pressure and gss analysis of gasses in plenum of rod 1086 Performance of metallography on rod 1086 Performance of burnup analysis on section frca highest burnup region of rod 1086 Performance of physical testing of selected cladding segments of rod 1086 Tests completed to date include only the gamma scan of rod 1086, and some of the metallography on samples cut from red 1086. Bod 1086 is the rod with highest burnup. The gamma scan will
, help determine the peak-to-average power in the rod and was also used to help deterisine the best locations for metallography and burnup samples. The samma scan is shown in Figure SA.
l The metallography on rod 1086 indicates no major differences from the metallography which was done on the 45 weight percent
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fuel in 1983 (rod 1088) and in 1984 on rod 1093 The clad failure occurred on rod 1093 The burnup values for the three rods examined, 1088, 1093 and 1086 are 47, 53 and 66 percent, respectively. Typical metallography pictures of rod 1088 are shown in l Figures 6 through 9 They can be compared with typical metallography pictures for rod .1093, shown in Figures 10 through i l 12. Pictures for rod 1086 were not available at GA in time to i include in the paper. Figures 13 and 14 show a fuel-clad interface for rod 1093 There is no evidence of fuel-clad interaction. Metallographic l pictures of the clad break surface for rod 1093 are shown in
.] Figures 15 and 16. They show the characteristics of a typical
.; ductile failure, with no evidence of embrittlement, corrosive
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j A picture from the scanning electron microscope (SEM) is
-l shown in Figure 17 It also shows normal ductility. No shear
'; faces were detected.
! It is expected that the post-irradiation examinations and
*l their documentation will be completed in the first half of 1986.
f The LEU uranium-sirconium hydride fuel has completed all 3 essential development testing successfully. It is offered commercially in both 20 wt-5 and 45 wt-5 uranium concentration 8 for existing TRIGA reactors and as a conversion fuel for l plate-type reactors. , i Several reactor facilities have already begun conversion to i LEU uranium-zirconium hydride fuel. It is in use in a mixed j plate-and-rod fuel configuration in the 1 MW reactor at the
- National Tsing Hua University in Taipei, Taiwan. A complete core j of 20 wt-5 4-rod shrouded cluster fuel has been fabricated and
-l shipped for conversion and upgrading of the PRR-1 reactor in the l Philippines. The converted reactor will operate at 3 MW with ; forced cooling and have pulsing capability. Startup is scheduled l for early 1986. Also, a complete core of 20 wt-5 LEU fuel has 1
I been fabricated for the 3 MW forced-flow TRIGA Mark II reactor system with pulsing capability in Bangladesh. It is scheduled to i be completed in 1985. U-IrH LEU fuel has also been delivered to existing TRIGA reactor facilities in Malaysia, Thailand and _l Yugoslavia for their use as new fuel is needed to meet burnup requirements. I : i r F
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, *s . haf hs. W REFUELING THE RPI REACTOR FACILITY WITH LOW-ENRICHMENT FUEL
- l 0.R. Harris, F. Rodriguez-Vera, and F.E. Wicks Rensselaer Polytechnic Institute, Troy, NY 12180 ABSTRACT The RPI Critical Facility has operated since 1963 with a core of thin, highly enriched fuel plates in twenty-five fuel assembly boxes. A program is underway to refuel the reactor with 4.81 w/o enriched SPERT (F-1) fuel rods. Use of these fuel rods will upgrade the capabilities of the reactor and will eliminate a security risk. Adequate quantities of SPERT (F-1) feel
- rods are available, and their use will result in a great cost saving relative to manufacturing new low-enrichment fuel plates. The SPERT fuel rods are 19 inches longer than are the present fuel plates, so a modified core support structure is required. It is planned to support and position the SPERT fuel pins by upper and lower lattice plates, thus avoiding the
- considerable cost of new fuel assembly boxes. The lattice plates will be secured to the existing top and bottom plates. The design permits the fabrication and use of other lattice plates for critical experiment research programs in support of long-lived full development for power reactors.
- Work supported under U.S. Department of Energy Contract DE-FG02-85ER75205 1
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INTRODUCTION The critical f acility at the L. David Walthousen Laboratory at Rensselaer Polytechnic Institute (RPI) includes a water-moderated s reactor fueled by highly enriched uraniun (HEU) in the form of UO2 dispersed in stainless steel fuel plates. Since the beginning of its operation as an RPI facility in 1963, the reactor has operated in a highly satisfactory manner as a training and research facility. The reactor was reconfiguredl several years ago so as to reduce the U-235 inventory below Skg. The reactor operating license was renewed recently to the year 2003.2 ,3 The research carried out at the RPI f acility has included reactor noise and kinetic experiments as well as activation and reactivi ty worth measurenents. In these low-neutron-flux research operations, there is no need to use HEU fuel With its attendant
, security problems. In fact, low-enrichnent fuel (LEU) in the form of clad U02 rods would be advantageous because it would permit a variety of critical and other experiments in support of nuclear power prograns. The training function at the RPI facility similarly
. would be enhanced by the use of LEU oxide rod-type fuel because this fuel would be more closely related to fuel used in the nuclear power i ndu st ry .
A program is underway to refuel the RPI reactor with 4.81 w/o enriched SPERT (F-1) fuel rods clad in stainless steel 11.84 mm in outside dianeter. The stainless steel cladding will be compatible with the remainder of the RPI reactor, which contains stainless steel components; thus the water purification system need not be upg raded. The proposed fuel rod configuration and support structure will be such that critical experiments can be carried out in support i of light water power reactor prograns, particularly for high exposure fuel development. The enrichnent of the SPERT (F-1) fuel is somewhat higher than that currently used in light water power reactors, but it is typical of the higher enrichments which are required for high exposure fuel. The use of stainless steel clad rather ~than the Zi rcalcy clad commonly used in light water power reactors does not greatly affect such critical experiments, although it must be corrected for. Adequate quantities of SPERT (F-1) fuel rods are available, and their use will result in a great cost savings compared with use of LEU fuel plates manufactured for the purpo,se. Wi tn present fuel element technology, it is not possible to achieve the uraniun density required for simple replacement of HEU fuel plates by LEU fuel plates in the RPI reactor. It would be necessary instead to add numerous fuel plates and some fuel assenblies of low enrichment. In this case, there would be additional costs for core design and safety analysis although the basic core support structure and control system would renain unchanged. The new fuel plates would likely be uraniun silicide in aluminum and would represent,a
I
'a substantial cost. Care would be required to ensure chemical compatibility of the aluminum plates with the otherwise stainless steel system. The RPI Critical Facility will continue to operate at very low power with the SPERT (F-1) fuel in place so the SPERT (F-1) fuel rods can be reused later by the U.S. Government for other purposes. That is, the SPERT fuel will receive only very low exposure during the lifetime of the RPI reactor.
REFUELING The refueling program has a number of components. These include (a) nuclear core design; (b) mechanical redesign of support structure; (c) safety analyses to verify new core design for license modification; (d) fabrication of new core support structure components; (e) disassembly of old core structure and perinherals; (f) assembly of new core structure and peripherals; (g) redesign, reconstruction, and criticality analysis for the fuel storage vault; (h) shipment of old fuel to designated facility; (1) shipment of new fuel to RPI and storage in vault; (j) reanalyses of instrument response, shielded doses, and other reactor parameters for license modification; (k) alteration of Technical Specifications and other documents for 1icense modification; (1) loading of new core and
. experimental verification of reactor safety parameters; (m) reestablishment of instrument set points and other operational parameters; (n) development of new instructional and research techniques.
The principal ' technical problem is the design of the new core support structure and control system within reasonable cost limitations. Figure 1 shows the present core structure. The integrity of the core support structure is maintained by four vertical tie bolts in tension and by the massive top plate, middle (support) plate, and lower (carrier) plate. These units are retained in the new design which should be at least as strong as 1 that presently used. I The SPERT (F-1) fuel rods are 106.05 cm long,19 inches longer than are the present fuel plates. Nuclear analyses indicate that about 500 fuel rods will be required on a square lattice pitch of about 1.4 cm. The roughly cylindrical array of fuel pins will extend out to a radius (see Figure 2) which contacts the present control rod locations to an extent which is estimated to be adequate to yield suitable control rod worths within the present Technical Specifications. The total fuel weight will bd about one-half metric ton. The fuel rods will be supported and positioned on a new lower fuel pin support plate (see Figure 2) resting on and secured to the existing carrier plate. The lower fuel pin support plate will be drilled with 1/4" diameter holes in square array to accept the lower tips of the SPERT pins. A new upper fuel pin positioning plate (see Figure 2) resting on the top plate will be drilled through on the appropriate square pitch to position the upper ends of the fuel - i
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. .i pins. The lower fuel pin support plate and the upper fuel pin positioning plate will be secured by the present tie bolts. Such plates are used in many critical experiments and will be referred to here as upper and lower lattice plates. The use of lattice plates is much less expensive than new fuel assembly boxes and provides a
, regular fuel pin array unbroken by stainless steel box walls. The middle (support) pl' ate will be retained for structural reasons and
", to support a plastic spacer plate, but its center will be machined out. Four new tie rod spacers will be required.
The present control rods and control rod drives will be retained, but the HEU fuel followers will be removed. Additional absorber sections (on hand) will be inserted both for enhanced control rod worth and to maintain the control rod weight. The present control rod hydraulic shock absorbers (see Figure 1) will be retained as fastened below the lower (carrier) plate. The control cod motion will be laterally constrained by cutouts in the upper lattice plate and by new, longer positioning rods passing through holes already present in the shock absorbers. There is adequate space below the carrier plate for these positioning rods when the control rods are fully inserted. There is also adequate travel on i the control rod drive gears for the increased core height; however, minor modifications are required to the control rod height sensor mechani sms . , CONCLUSIONS
, The use of LEU SPERT (F-1) fuel in the RPI Critical Facility i reactor provides advantages in increased functional capabilities and i
! in lessened security risk over the present HEU fuel plates.
Moreover, the use of the existing SPERT (F-1) fuel is much less expensive than would be manufacture of new LEU fuel plates for the reactor. The lattice plate technique for positioning the fuel p, ins is considerably less expensive than would be fuel assembly boxes, and it permits the core to be a uniform array of fuel rods. Tne
, structural properties of the new core are expected to be similar to those for the present core. It is anticipated.that the control rod system will be used with little modification except for replacement of fuel followers by absorber material.
4 _ . - . ~ - _ - .- - - . _ - - . - . _ _ . _ _ _
1 5
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1 l l 1 REFERENCES P.R. Nelson and 0.R. Harris, " Reconfiguration of the Rensselaer 1. Polytechnic Institute Reactor Facility to Lower Critical Mass " Nucl . Tech. , 60,, 320 (1983).
- 2. 0.R. Harris, 0.C. Jones, F.E. Wi ck s, A.S. Harris, C.F. Chuang, and F. Rodriguez-Vera, " Design Basis Transient Analyses for Low Power Research Reactors," Proc. Int. Symposiun on Use and Development of Low and Mediun Flux Researctr Reactors, Cambridge, October 17-19, (1983), Cambridge, Ma ss. (1983) , Atomkernenergie Kerntechnik , M, 450 (1984).
- 3. 0.R . Ha rri s and F . Wi ck s, " Sa fety An aly si s Repo rt, Dock e t No.
,, 50-255, License No. CX-22," Januarf (1983) .
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Po p ths. W 1 , STATUS OF LEU PROGRAMS AT BABCOCK & WIILOX j KEN BOGACIK RESEARCH AND TEST REACIOR EUEL ELEMDir FACILITY LYNCHBURG, VIRGINIA, U.S.A. ABSTRACT Within the Iow Enriched Programs being conducted at Babcock & Wilcox the primary effort has been to establish, from past LEU developnent work and current production technology, an efficient production process that maintains product quality for both LEU UAlx and U3Si2 elements, h is effort has allowed the Babcock & Wilcox Cmpany to successfully complete a second IIU production contract for the 2-MW Ford Nuclear Reactor
, at the University of Michigan. Current U3Si2 contracts which include Standard snd Control Elements for the Oak Ridge Re. actor, SAPHIR Elements for the Swiss Federal Institate for Reactor Research, silicide (U3Si2) powder for the Danish Riso National Laboratory and Elements for Sweden's R2 Reactor at Studsvik are being , manufactured under the same guidelines of quality and efficiency improvements. We transition frm developncntal work to a production prncess for powder fabrication; capacting; plate and element fabrication along with inspection methods are highlighted within this report.
3 J I
2 INIRXXJCIION Research and developnent programs funded within the Naval Nuclear Fuel Division of Babcock and Wilcox have provided the technical knowledge to enter into "first time" Iow Enriched Silicide Production Contracts. Rese research and developnent programs are continuing in an effort to provide the additional information and data needed to make a more efficient element manufacturing process. Within this report an overview of the specific areas investigated during the past year is provided. FUEL POWER A significant effort has been devoted to the implementation of a production process for U3Si2 LEU fuel powder. Knowledge gained frm past developnent work allowed a transition frcrn small production lots to a " full" production process. Improvements in operating procedures, operating techniques and equipnent provided increased control of the final product quality. Rese improvenents also msulted in increased product yields. Controlling the silicon content of the final U3Si2 product was a primary concern to B&W. With pilot production quantities of a few kilograms, silicon content was easily controlled within the range of 7.3 to 7.4 weight percent. This range produced an alloy close to the desired U3Si2 stoichionetry. Subsequent studies conducted at Argonne National Iaboratory on B&W's developnent material identified small amoun'ts of free uranium (less than 1%) within the alloyed product. W erefore, silicon had to be added to the alloy to produce a silicon rich final product during ths ' manufacturing of production lots. Once production lot sizes exceeding 15 kilograms were initiated, B&W adjusted the target silicon content of the U3Si2 product to 7.7 weight percent. This additional silicon decreases the potential of having free uranitrn within the final product. Also, the additional silicon content produced an alloy which is more suitable for ccmninution. Figure 1 lists the certified silicon contents of B&W's current production run. As shown in the figure, a consistent silicon content was obtained during the production of the 13 fuel lots. e
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3 During process develegnent work, variations in the silicon content were first found to be caused by the spalling of the arc-melted alloy due to themal stresses during cooling. A "high" wall splatter shield was designed to contain spalled product within the arc-melt cavities. nis contaiment of spalled alloy provided a nminal silicon /u:anium alloy of 7.63 weight percent silicon with a variant of 0.14 weight percent as seen in the production lots listed in Figure 1. Increasing the silicon content improved the caminution characteristics which directly affected the processing efficiency and product yields during the crushing and screening operations. Bis silicon addition provided a size distribution which is similar to UAix as seen in Figure 2. Essentially no heavies (+ 149 micron) were generated. With B&W currently working to a 25% fine (- 44 micron) content in the final product blend of both the U3Si2 and UAlx fuel types the excess fines are recycled. men a calculated adjustment of the -44 micron fine content is made to the final blend which is identified in the Figure 2 Bar Chart. As seen in this figure, the -44 micron material is adjusted to the range of 20 to 25 weight percent. Success was obtained in the recycling of the excess of fines using two methods. De first method is recycle briquetting of the fines which is similar to the method used i.n UAlx production. Once a recycle briquetting pressure and die configuration were determined, a coherent U3Si2 briquette was produced that could be introduced into the melting process. We second method was a direct addition of fines into the melting process. This method is considered to be more desirable frm the standpfaint of efficiency. However, the addition of 50% fines to the product blend would essentially eliminate the need for recycle since the
-44 micron material is generally less than 50 wt. %. We affect of increased fine content on the volume loadings of fuel to alminm matrix material is being further evaluated.
A critical characteristic of U3Si2 is its affinity for oxygen and its reaction with altaninum matrix material in the presence of oxygen at elevated tenperatures. B&W conducted controlled experiments for determining the specific conditions required to inhibit the U3Si2 reaction. Rese experiments were run with snall quantities of fuel (2 grams) and cmpact sections (blended and capacted U3Si2 with alumintsn matrix) . Data generated fran these investigations provided the following information: (See next page).
4 U3Si2 EVEL REACTICN OIARACTERISTICS WIni OXYGEN
- 03Si2 oxidation will initiate at tenperatures of 350 F
, (177 C) producing an exothermic reaction.
- Once the initial reaction has subsided, a dull red burning of the fuel is sustained.
- Restricting oxygen fran contacting the surface will extinguish the reaction.
- Re-exposure of the reacted surface to oxygen will propagate the reaction until all fuel is constzned.
- Surface oxide developnent/ passivation data are provided in Figure 3.
COMPACT REACTICN CHARACIERISTICS , (preliminary data and results)
- U3Si2 will react with aluminum matrix powder in a vacuum at
, elevat.ed temperatures.
l No noticeable reaction of U3Si2/Al was detected at temperatures below 800 F (427 C). A minimal U3Si2/A1 reaction was detected at 900 F (482 C).
- Increased reactivity of U3Si2/Al was detected at temperatures exceeding 1000 F (538 C) .
We above characteristics were determined through an initial investigation directed at the identification of reaction tenperature ranges. All tests were run for approximately 400 minutes at tenperature. Further work is being conducted to better define temperature ranges within which the reaction is initiated. Also, the affect of time at tenperature is being evaluated. Wis ongoing work will involve both high oxygen content (2000 pga < oxygen < 7000 ppn) U3Si2 fuel and low oxygen (< 2000 ppn) 03Si2 fuel. PSW's fuel manufacturing process maintains an oxygen content below 1500 ppn as noted in Figure 4. 7 The increase of oxygen in lot 13 is a function of particle surface area since the majority of the lot consisted of recycle
, fines. Manufacturing U3Si2 fuel with a high oxygen content will require additional processing to add controlled amounts of oxygen to the fuel. t is increased oxygen content could be a factor in inhibiting the reaction of U3Si2 with the aluminum matrix.
5
. PIATE FABRICATICN An advantage of U3Si2 fuel ccmpared to UAlx or U308 fuel is the density of this fuel type. Provided below are theoretical fuel densities and theoretical maxinamn uranium densities within a fuel plate:
'IHEO. EUEL 'IEDORETICAL FUEL TYPE DENSITY MAX. PIKIE IDADINGS grams /cc grams U/cc UAlx 7.60 2.3 '
i I U308 8.40 3.2 U3Si2 12.20 4.8 We greater density of U3Si2 fuel is advantageous in obtaining higher uranim gram densities in the fuel plate. B is high density fuel has provided a likely choice for LEU conversions. Figure 5 shows the difference in U3Si2 versus UAlx densities in BtM production Mts. Figure 6 ccupares void voltzne data for specific uranium loadings of both LEU and HEU fuel plates. Bis void volume factor directly affects the volme loadings of fuel powder to
.l almine matrix powder when manufacturing high uranim gram density LEU fuel plates.
i As identified previously, U3Si2 has specific reaction characteristics which have to be considered in the fabrication of
! Cuel plates. Since this fual has a high affinity for oxygen; j storing, weighing, blending, and ccupacting require an oxygen free (< 4.0% oxygen) envirorynent. As a ccupact of U3Si2 fuel , dispersed with almine powder, the fuel is stable in ambient , conditions. his ccmpact can then be brought out into the ! at2nosphere and assembled with the frame and cover plates. During the assembly operation, care in handling the ccmpact is required to eliminate the generation of stray fuel on the frame / cover surfaces.
, U3Si2 oxidation will occur during the preheat . cycle prior to the first hot roll bonding pass when the assembled ccznpact/ frame / cover is heated to 900 degrees F. , Oxygen will enter through the exhaust port of the assembled billet during the preheating cycle. De resulting reaction product is black oxide. h is reaction is controlled by keeping oxygen away frcm the ccmpact during the preheat cycle.
,, _ _ _ _ , . _ , . - _ , , _ _ . _ _ _ _ _ y _,
6 Analysis has identified this product as uranium oxides, uranitzn-aluminides, silicon oxide, and aluminum oxide capounds. Mstructive inspections of a plate known to contain a reaction product revealed ultrasonic defects. ELEMENTS . B&W's LEU contracts / designs which are either cmpleted or are in production are listed below: NOMINAL N NINAL GRAM U235 GRAM U235 FUEL ELEMENT PIATE
! DESIGN TYPE IDADINGS IDADINGS STKIUS U. OF MICH. UALx 167 9.28 cm plete R2, Sweden U3Si2 325 18.06 cmplete U3Si2 490 27.22 in production j ANI/ORR U3Si2 340 17.90 in production U3Si2 200 13.33 in production SAPHIR U3Si2 412 17.90 in production
SUMMARY
A significant silicide information data base has been generated through past process developnent programs. However, continued research and developnent efforts will have to be directed at the improvement of silicide manufacturing until a caplete understanding of this "new" fuel type is obtained. Additionally, the periodic technical reviews of silicide elernent specifications by Argonne National Laboratory and the RERTR Program have aided in the generation of uniform product requirunents intended to assure the quality standards required by research reactor users and operators.
7 FIGURE 1 SUJCON CONTINT 10 9-8- g 7 u 3 on 6
., H 5' .
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.U3SI2 SURFACE OXIDE CEVEIGid.ENT DATA TEMPE3UCURE TLE PERCDTI CXYGEN F (C)
. HOURS PPM control -
1483 200 (93) '- 2 s 1584
- i 350 (177)* 2 4 5220 '
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j l 350 (177) 1 3750
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- Denotes cumulative times at tenperature.'
s Figure 3. Preliminary passivation data generated at B&W for 03Si2 fuel.
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'i FIGURE 4 CIYGEN CONTENT
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9 FIGURE 5 DENSITY 15.0 13.5 __ _ _ A 12.0.'[ b bl bI fl b b b bklb# k io.5./ ! II Ii !! l I i1 I II, I g9.0 / I I i 'l , I l l ll l l I
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1 2 3 4 [ 5 [T 6 7 [4[N[ 8 [ [ I ~[/ 9 10 11 12 13 LOT NUMBER 8 UAlx 0 u3st2
- Tr.ar AVERAGE CORE VOID VOIIIME OF B&W FUEL PIATES NCMINAL MEASURED B&W URANIUM PERCENT DESIGN PIATE FUEL ENRICHMCTI' IDADIfC CORE VOID MICH. 24 UALX LEU l.8 g/cc
- 5.9 -
MICH. 25 UAIX IEU 1.8 g/cc
- 6.5 MIT 26 UALX HEU l.6 g/cc 6.0 MURR 30 UALX HEU 1.5 g/cc 5.5 i R2 55 U3Si2 LEU 4.8 g/cc ** 9.1 ORR 56 U3Si2 LEU 4.8 g/cc 9.8
- Data based on a one plate measurenent.
** Value obtained frczn Argonne National Iaboratory.
Figure 6. Inmersion density measurenent data to detecnine plate core void volume.
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LICENSING CONSIDERATIONS IN CONVERTING NRC-LICENSED
, NON-POWER REACTORS FROM HIGH-ENRICHED TO LOW-ENRICHED URANIUM FUELS Robert E. Carter U.S. Nuclear Regulatory Commission ABSTRACT During the mid-1970s, there was increasing concern with the possibility that highly enriched uranium (HEU),
widely used in non power reactors around the world, might be diverted from its intended peaceful uses. In 1982 the U.S. Nuclear Regulatory Commission (NRC) issued a policy statement that was intended to conform witn the perceived international thinking, and that addressed the two relevant areas in which NRC has statutory responsibility, namely, export of special nuclear materials for non-USA non power reactors, and the licensing of USA-based non power reactors not owned by the Federal government. To further address the second area, NRC issued a proposed rule for public com-ment that would require all NRC-licensed non power reactors using HEU to convert to low enriched uranium (LEU) fuel, unless they could demonstrate a unique purpose. Currently the NRC staff is revising the proposed rule. An underlying principle guiding the staff is that as long as a change in enrichment does not lead to safety-related reactor modi-fications, and does not involve an unreviewed safety ques-tion, the licensee could convert the core without prior NRC approval. At the time of writing this paper, a regu-latory method of achieving this principle has not been finalized. 6 ,
- c. I
-s INTRODUCTION The principal purpose of this paper is to summarize current thoughts of the staff of the Nuclear Regulatory Commission (NRC) about licensing considerations in connection with the replacement of high enriched uranium (HEU) with low enriched uranium (LEU) fuel in non-power reactors in the U.S. When it was proposed several months ago to write this paper, the NRC staff was still working on the on-going rule-making procedure, and it was intended to present and ex-plain the final approved outcome. However, because this is a com-plex issue, some questions have not yet been resolved.
It must be made clear though, that the NRC Commissioners have established a policy that the agency will take steps to enccurage the owners of non power reactors subject to its jurisdiction to con-vert their reactors from the use of HEU fuel. This policy was first announced in 19821, was amplified in the proposed rule in 19842, and has been reaffirmed by Commissioner Asselstine at this conference, as well as at the 1984 Argonne meeting of the RERTR program. In the United States, as elsewhere, nuclear reactors may be grouped into one of two broad classes, power reactors, that are used for the commercial production of electricity, and non power reac-tors, that are useful for research, development, education, and testing. All of the first class are licensed and regulated by NRC, but only those non power reactors not owned and operated by the Department of Energy (DOE) or Department of Defense are licensed and regulated by NRC. Thus, the following remarks apply only to the NRC-licensed nonpower reactors in the United States. Before moving on to more detailed discussions, it is important to mention a broad admonition and constraint imposed by law in the US on the nuclear regulatory process. The Atomic Energy Act, under which both DOE and NRC function, encourages and supports nuclear research and development, and therefore asserts that reactors useful for these purposes are to be subject to the minimum regulation nec-essary to protect the health and safety of the public. NRC must comply with that law. BACKGROUND In order to help clarify the reactor licensing situation in the US, I will summarize the history and add a few words about the US regulations. The US Atomic Energy Act of 1954 provided the legal basis for the US Nuclear energy programs. A single agency, the Atomic Energy Commission (AEC), was established to carry out the law. In accordance with the law the AEC developed procedures and regulations for licensing reactors that it did not own, but airo built and operated many reactors that it did own but which were r.ot subjected to those regulations. The licensed reactors consisted of both commercial power reactors and non power reactors dedicated to 2
{- research, test, and education. In general, both classes of licensed reactors are subject to the same regulations, but because of the recognized larger potential radiological impact on the public, addi-tional more prescriptive requirements have evolved for power reac-tors. After about twenty years of experience, Congress changed the law, establishing the current Nuclear Regulatory Commission (NRC) whose function is to license and regulate nuclear facilities, and assigning the other functions of the AEC to a different agency that is now the Department of Energy (00E). This department continues to own and operate nuclear facilities that are not subject to NRC's regulations, and continues with a mission to help promote nuclear science and technology, including some financial and fuel's assis-tance to university reactors. DOE, of course, has supported the RERTR program. As we may remember, the first several nuclear reactors were operated with natural or low enriched uranium fuels. As high en-riched uranium became more readily available, it was used in some reactors to improve or optimize certain characteristics. Soon, even though probably not clearly required by performance considerations, HEU became the fuel of choice of both the US Government and reactor operators, and was used in most non power reactors except those of the TRIGA type. Even for those, however, an improved version of the zirconium hydride fuel was based on HEU. While research reactors were gaining wide international accep-tance and use, concern continued to grow that the increasing quanti-ties of HEU used for non power reactor fuel were increasing the potential risk of diversion to an undesirable level. NRC ROLE As noted earlier, NRC responded to the increasing concern about diversion of HEU fuel by issuing a policy statement in 19821 Among other things, this policy included an intent to encourage NRC li-censees to convert their reactors from the use of HEU fuel. In a second step, in 1984, NRC published a proposed rule for public com-ment that would require all licensees to cease using HEU fuel unless they could qualify for a defined unique purpose exemptions . A large fraction of the affected licensees expressed their concerns that conversion would require license amendments with a potential for public intervention that would involve unbearable litigation effort and costs, and even in the absence of litigation, conversion would require the expenditure of scarce money and manpower resources. The proposed rule acknowledged these two potential costs, and expressed concern about them, but did not include explicit provi-sions for coping with them. Both because a guiding principle of the reduced enrichment program was to avoid any significant negative impact on reactor performance, and because of the provision of the Atomic Energy Act noted above, the Commission directed the staff to 3
.f revise the proposed rule to make conversion dependent on assured funding by the federal government, and on the avoidance of public intervention in conversion licensing actions not accompanied by significant additional reactor changes. The staff has drafted a re-vised version of the rule that does require that federal funding for all conversion-related effort be assured before conversion is ini-tiated. The question of public participation has not been resolved
- at the time of writing this paper, but is a continuing concern. It is evident that NRC should avoid inadvertently causing excessive financial burdens on non power reactor licensees in its efforts to reddress the results of previous government policy.
OBJECTIVES AND APPROACHES In order to issue a reactor license, the NRC staff must pro-perly conclude, for each individual reactor, that there is reason-able assurance of no significant hazard to the public. However, many of the technical considerations can be generic. The staff is approaching conversion from HEU to LEU in that way, though not all steps are resolved at this time. The example discussed below of a generic environmental assessment currently in use suggests the fea-sibility of generic reviews for fuel conversion also. Several years ago, with reaffirmation about five years ago, the staff developed a generic environmental assessment for the licensing of non power reactor: that concluded that for such reactors licensed to operate at 2 MW and below, there was reasonable assurance of no significant impact on the environment or the public resulting from continued operation. This conclusion was based on a large number of detailed individual reactor safety reviews and on extensive operat-ing histories of such reactors. In recent years the staff has re-newed nearly 30 reactor licenses, and has relied on this generic environmental assessment, in conjunction with a review of an indi-vidual environmental report in each case to determine if there are any special features that render the generic assessment inappli-cable. No exceptions were found. Similarly, the staff's preliminary consideratier of the prob-able changes in reactor parameters and operating characteristics resulting from conversion from HEU to LEU has revealed no new haz-ards considerations. Therefore, we are considering grouping the reactors into categories, such as not more than 2 MW power level, and reviewing within that group the potential hazards to the en-vironment and the public due to the change of enrichment of the fuel. Table 1 gives a tentative grouping of all NRC-licensed non-power reactors using HEU. As in the example of the generic environmental assessment dis-cussed above, the staff would have to conclude for each reactor that there were no special features that rendered the generic evaluation inapplicable. The following are two ways in which that could be 4
e 9 Table 1 CATEGORIES OF HEU FUELED NON-POWER REACTORS LICENSED BY USNRC
- l. High Power Density 2 - University
- 11. Unique Fuel Configuration 3 - University 2 - Commercial Ill. High Pcwer (>5 MW) 1 - University 1 - Government i
1 - Commercial IV. Medium and Low Power ( < 2 MW) Plate Cores 13 - University V.TRIGA FLIP Fuel 4 - University 1 - Commercial
)
r
..j done: (1) each licensee would submit an application and an analysis for his proposed conversion in the usual way that other amendments are applied for, or (2) the licensees as a group would work with RERTR to assemble pertinent information, data, and analyses in a broad-scope document, and each licensee would submit a copy of that document, along with a comparison between his specific reactor para-meters and those embodied in the generic analyses. The NRC staff will continue working with the Argonne group and licensees to ex-plore flexible generic approaches to the review and evaluation of conversions.
DISCUSSION In the previous section, it is implied that a license amendment will be required. In most cases, that is probably true, unless the conversion would meet the current criteria of 10 CFR 50.59 of the NRC regulations, or some other mechanism can be found for relief from an amendment. Part 50,59(a), reproduced below, allows certain types of changei to a licensed reactor without prior NRC approval. If appro-val is required, in most cases it is accomplished by means of a li-cense amendment. S 50.59 Changes, tests and experiments. (a)(1) The holder of a license authorizing operation of a production or utilization facility may (i) make changes in the facility as described in the safety analysis report, (ii) make changes in the procedures as described in the safety analysis re-port, and (iii) conduct tests or experiments not described in the safety analysis report, without prior Commission approval, unless the proposed change, test or experiment involves a change in the technical specifications incorporated in the license or an unre-viewed safety question. (2) A proposed change, test, or experiment shall be deemed to involve an unreviewed safety question (i) if the probability of oc-currence or the consequences of an accident or malfunction of equip-ment important to safety previously evaluated in the safety analysis report may be increased; or (ii) if a possibility for an accident or malfunction of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced. In the previous section, nothing was said about out-liers, that is, those reactors that wouldn't qualify for inclusion in a category. Presently, the only alternative suggested is a case-by-case submit-tal by a licensee, review by NRC, and issuance of a license amend-ment by NRC. However, unless significant reactor modifications not required by the change in enrichment were made, the staff antici-pates that the licensee submittal need only address parameters and 6
l'., characteristics of the reactor altered by the fuel change. Figure 1 is a schematic flow diagram of the approach just discussed.
]
In the above approach, it has not been indicated that the path of minimal effort on the part of the licensee will lie with no change in the physical dimensions and hydraulics characteristics of the fuel elements upon conversion. Nor has the NRC staff suggested that the LEU fuel design must be optimized to achieve the maximum possible margins between safety limits and operating conditions in each case. Instead, NRC has a long-standing policy to standardize reactors when feasible. Therefore, we urge the licensees and DOE to propose standardized fuel parameters that will satisfy licensees' performance requirements and still provide bases for NRC to conclude logically that changing the fuel enrichment would cause no signifi-cant impact on the health safety of the public. CONCLUSION At the time of writing this paper, the NRC intends to issue a revised rule requiring all licensees to convert from HEU to LEU un-less they qualify for a unique purpose exemption. The revised rule is expected to have a provision making conversion dependent on as-sured federal funding. The revised rule also is expected to include provisions for avoiding public involvement on issues related closely to conversion from HEU to LEU. The schedule for issuance of the revised rule is currently uncertain. REFERENCES 4 1Use of High Enriched Uranium (HEU) in Research Reactors; Policy Statement, August 24, 1982, Federal Register (47 FR 37007). 2 Limiting the use of Highly Enriched Uranium in Domestic Research and Test Reactors, July 6, 1984, Federal Register (49 FR 27769). 7
V l l l REACTOR l Unique STET m j E " 9"' pproved "
- Perpose E Application \ Unique Changes Evaluated
\ Porpose
[ Densed l l conwees.on peos me,oe convers on Only mod.fwee.ons NOT dictated by convers.on Changes Evaluated Changes Evaluated i g% { :paus ciated
- CD SAR Changes SAR Changes SAR Changes 1
! No uneewewed un,e,.ewed E ' + e-- - , ._uestien sesee o. 2 setes, cue e.ca safety Question l Fuel Changes l hther Changes l E E License 10 CFR 50.50 License Amendment License and Amendment Amendment Apphcation Implement Application
. Application I I I j NRC Review ** NRC Review l NRC Review l
l i E u l l Amendment Amendment issued Amendment issued Issued )
^
1 - Implemeat emplement lf 4 4 1 e _ _____ _~ _ __
s .. L 9 6
)
f I l 8 9
+
T PROGMAYXE i I
- 6 9
@ J
w i Sunday, October 13 [ p.m. . Arrival of delegates I at Motel Akersloot
. Monday, October 14 i
1
; 08.00 Departure of coaches from Motel Akersicot
! to JRC Petten (FORUM) j 08.45 Arrival at Petten. Registration i
- h l 09.15-09.20 Welcome and Opening (Director of the Petten Establishment) 09.20-10.45 Session 1 i
- 1. J.K. Asselstine, Comissioner, U.S.-N.R.C.
l (invited speaker)
" Remarks on U.S. policy concerning LEU core ,
I conversion of research and test reactors" (preliminary title)
- 2. A. Trave 111, Argonne National Laboratory "The Status of the RERTR Program:
0verview, Progress and Plans"
- 3. I. Miyanaga, K. Kamei, JAERI, Tokai-Mura, and K. Kanda, T. Shibata, Kyoto University "Present Status of Reduce 1 Enrichment Program for Research and Test Reactor Fuels in Japan"
. 10.45-11.15 Break -
. ~ -
7ZC D DC - 3 0 4-fu bru ww q s MPmp ' t 11 lb-13.00 Sessicn 2 g)R R b a
, 'l b) 33 M 5 4 R. Muranaka, International Atomic trergy Agency, / EF /45')
- l '/1enna , Austria
!, "Research Reactor Core Conversion Programmes"
! 5. W. Krull, GXSS-Forsenungszentrum, L
Geestnacht, Germany
" Reduced enr1cnment activities at GKdS*
?
j 6. C. Bass, M. Barnier, J. Beylot, P. Martel
! . et 1. Merente, CEA (CEN Grencele et CEN Saclay)
,f " Rapport d'avancement sur la reduction de l'enrienissement du cembusttoie dans les reacteurs t 4 -
experimentaux du CEA" ! .l 1 l 7. J.L. bnelgrove, G.L. Hofman, Argonne National l Laboratory and G.L. Copeland, Oak Ridge National Laboratory { l
- 1. " Irradiation Performance of Reduced-Enrichment Fuels lested under the U.S. RERTR Program" i- , 13.00-14.30 Lunch i
14.30-15.45 Session 3
- 8. R.E. Carter, U.S. NRC
{
" Licensing Considerations in Converting NRC-Licensed i Non-power Reactors trem High-enr1cned to Low-r.
i enriened Uranium Fuels" l
- 9. 0 R. Harris, Rensselaer Polytechnic institute, J.E. Matos, Argonne National Laboratory ,.
H.H. Young, U.S. Department of Energy
. t
- " Conversion of University Reactors to .
Low-Enrichment ruel - tssues and Prospects" t i i
a, t.
.4-
. 10. M.D. Atfield, AEC Chalk R1,er, Canada
" Core Managerrent and Reactor Physics Aspects of tne j Conversion of the NRU Reactor to LEU" 15.45-16.15 Break i 16.15-17.55 Session 4 4
l
- 11. K. BUning, W. G13ser, J. Meier, G. Rau, A. Adhrmoser, L. Zhang, Physik-Oepartment E21 der
' Techn. UniversitHt MUnchen, Germany
," " Progress Report on the Design of a Compact Core f' or Upgrading the Munich Research Reactor"
- 12. J. Kollas, Greek Atomic Energy Commission,
. Agnia Paraskevi, Attiki, Greece U L u 1 R r tr
- 13. J.L. Snelgrove, M.M. Bretsener, R.J. Cornella, f
Argonne National Laboratory and R.W. Hobbs,
.j Oak Ridge National Laboratory ,
l "The Whole-Core LEU Fuel Demonstration in the ORR" l
- 14. A. Strcm1cn, Un. Siebertz, M. Wickert, Interatom GmoH, Bergisen Gladbach, Germany "The Conversion of the Dido-type Keactor PRJ-2, Studies and Conclusions" f
18.00-19.30 Cocktail ottered by the Director et tne Petten Establishment, Joint Research Centre, comission of , the European comunities
. : 19.30 Departure of e aches from Petten to Motel Alkmaar t
t
.. v.
-o.
Tuesday, Octooer 15 t i i 08.00 Departure of coaches from Motel Akersicot to
,- JRC Petten (FORUM) 09.00-10.30 Session 5
- 15. C. Costescu, C. Iorgulis, D. Simionovici, I
Institute for Nuclear Power Reactors, Pitesti, Rumania
! " Flux and Power Distribution in HEU, LEU and mi.xed I TRIGA SSR - 14 MW cores with control rods" l
l
- 16. T. Harami, M. Hirano, Y. Asahi, A. Kohsaka and
.l N. Ohnishi, JAERI, Tokai-Mura "A Safety Analysis of the Research Research
! Reactor JRR-3" i
i 17. K. Kanda, S. Shiroya, M. Mori, M. Hayashi and T. Shibata, Research Reactor Institute,
! Kyoto University, Japan
" Study on the Temperature Coefficient of MEU and HEU in the KUCA"
- 18. K. Bogacik, Babcock & Wilcox
" Status of LEU Programs at Babcock and Wilcox" 10.30-11.00 Break 11.00-12.45 Session 6 l l
- 19. R.H. Chesworth and G.B. West, GA Technologies Inc.,
l San Diego, USA
" Final Results of Qualification Testing of TRIGA (Uranium Zirconium-Hydride) Fuel in the Oak Ridge Reactor, including Post Irradiation Examinations" i
1 L.
- 20. Y. Fanjas and Ph. Dewez, CERCA, Romans, France "MTR fuel at CERCA - Status of Develocment 1985"
~
1 M O'/ fE21. M. Hrovat, H.-W. Hassel, NUXEM GmoH, Hanau, Germany i e., " Status of Cevelopment and 'crad1ation Performance db/ of Advanced Proliferation Resistant MTR Fuel at NUXEM" g/ce.(U) 22. J.C. Wood, O.F. Sears , L.C. Berthiaume, L.N. Hereert N - and J.D. Schaefer, Atomic Energy of Canada Ltd. , Chalk River, Canada "The Fabrication and Performance of Canadian Silicide
, utspersion Fuel for lest Reactors" 12.45-14.15 Lunch (+ poster set-up) 14.15-16.45 Poster Session (Break 15.30-16.00)
- 23. l.A. Collis-Smith, GEC Whetstone "M.T.R. 's - Experience with Modernisation and Upgrading" i
i 24. J. Deckers, TUV Rheinland, Germany
" Minimum cladding thickness of Material Test Reactor Fuel Plates"
- 25. J.R. Deen and J.L. Snelgrove, Argonne National Laboratory l " Reduced-Reactivity-Swing LEU Fuel Cycle Analyses for the HFR-Petten" ,
- 25. M. Caner and T. Aldemir, Ohio State University
" Preliminary Neutronics Calculations for the 05URR LEU Conversion / Upgrade Program"
., er e,
t - 7-
- 27. 0.R. Giorsetti, AAEC Buenos Aires -
" LEU Fuel Elements Manufacturing in Argentina"
{- '28. 0 R. Harris, F. Rodriguez-lera and F. Wicks, 3 Rensselaer Polytechnic Institute, Troy, New York
" Refuelling the RPI Reactor Facility with Lcw Enrichment Fuel"
- 29. K. Kanda, M. Hayashi, K. Mishima, Y. Nakageme and T. Shibata, Research Reactor Institute, t
Kyoto University, Japan t "KUR Core Conversion - Neutronics Calculation" 5 l 30. W. Krug, E. Groos and G. Thamm, KFA Julich i
" LEU-Plate Irradiations at FRJ-2 (DIDO) under the 4 AF-Programme"
- 31. R.L. Moss and P. Nay, JRC Petten
; "A Comparative and Predictive Study of the Annual l Fuel Cycle Costs for HEU and LEU Fuels in the High Flux Reactor, 1985-1993"
- 32. E. Nonbdl, Risd National Laboratory, Roskilde,
,l Denmark
" Development of a 3-dimensional calculation model
{ of the Danish research reactor 3R3 to analyse a
! proposal to a new HEU core design called ring-core" l
, 33. S. Shimakawa, Y. Nagaoka, K. Takeda, S. Koike,
- 8. Komukai and R. Oyamada, JAER!, Oarai-machi,
; Ibarakt-ken -
i " Critical Experiments of JMTRC Mtu Cores (II)" l-i l 4
.. ~
- 34. Y. Suco, M. Ikawa 'and M. Kaminaga , JAERI, Tokai-Mura
" Development of Heat Transfer Package for Core Thermal-
; Hydraulic Design and Analysis of Upgraced JRR-3" l.
i Jb. P. lott and A. Jensen, AILAS-0ANMARK A/S
" Differential tnermal analysis and metallograpnic
! examinations of U Si -powder anc V 51 Al (38 w/o)-
3 2 3 2 i m1ntplates" i 36. W.L. Woodruff, Argonne National Lacoratory
*The Bencnmark Proolem Revisited - Ine Accuracy of l
- Finite 01tterence Methods and the Potential for the Use 1
of Nodal Methods in the Analyses of Researen and Itst i
' Reactors" 17.15 Departure of coaches from Petten to Motel Akersloot l
I
}
Wednesday, October 16 1 l 08.15-09.15 Final session and closure of meeting In the Meeting room of Motel Akersloot j , " Highlights of the meeting, contributions / reviews of session chairmen" 09.30 Departure of coaches from the Motel for HFR visit (optional) and departure of a coach from the Hotel to Schiphol airport (optional) 10.30-11.30 Visit of HFR j 11.45-12.45 Lunch 12.50 Departure of coaches from Petten for Schiphol airoort and Alkmaar railway station
- 14.15 Arrival at Schiphol airport
t .. c;
. g.
'lenue i
j Sessions 1 through 6 will take place in the Aucitorium, e ground floor, FORUM building. l The Poster Session uses the Conference Recm,1st floor, FORUM building. 1 The Round Table on '4e'dnesday morning is held at the conference facility of Motel Akersloot, t f e e e i
- e e
e e
T' es c t i I 4 d
?
b 4 l i 0 3 6 e o I t i e 6 L.
1. Papor To. 2 9 THE STATUS OF TlfE RERTR PROGRM'. OVERVIE'J, PROGRESS AND PLMS A. Trave 111 (ANI,) , ABSTRACT 4 The status of the U.S. Reduced Enrichmene Research and Test Reactor (RERTR) Program is reviewed. The RERTR Program objectives, goals, and past accomplishments are briefly summarized to provide a background against which last year's progress can be measured. The main program events, activities, and accomplisbaents which occurred since the latest meeting are then reviewed in greater detail, placing special emphasis on their interrelationships and significance for the overall program. Finally, the plans and schedules for future program activities are presented with a detailed discussion of the i changes which have occurred during the past year, and of their causes and i implications. t e
l .. [ Papor No. 3 1 L l 1 Present S tatus o f Reduced Enrichmen Program for Research and Test Reactor ruels in Japan
'o Ichiro Miyanaga and Kazuo Kasei Japan Atomic Energy Research Institute Tokai-mura, Nakagun, Ibaraki 319-11, Japan
, and Keiji Kanda and Toshika u Shibata Research Reactor Institute, Kyoto University Kumat'o ri-cho , Sennan-gun, Osaka 590-04, Japan The Reduced Enrichment Program for JRR-2, JRR-4, JMTR of JAERI and KUR, KUHFR of KURRI are in progress under the Joint Study Programs with ANL.
In parallel with the Program, the revamped JRR-3, under construction, is pursuing its schedule of employ-f ing LEU fuels. Feasibility study for use of LEU fuels in JMTR and KUR is now being carried out. LEU silicide mini-plates have been irradiated successfuly in JMTR. MEU l critical experiments in JMTRC and KUCA paved the way to the smooth conversion. The irradiation tests of full-size MEU and LEU fuels are going on in JRR-2, JER-4 and JMTR. All these schedules are in process comprehensively with the related developments including the code system for analysis and design. o e l
m . . . __ __ . _ _ _ - - ._ - - _ _ _ _ _ _ . - _ _ . _. Paper No. 4 RESEARCH REACTOR CORE CONVERSION PROGRAMMES
. DEPARTMENT OF RESEARCH AND ISOTOPE 3 INTERNATIONAL ATOMIC, ENERGY AGENCY R.G. Muranaka, Physics Section This paper describes the Research Reactor Core Conversion Pro 8:ammes of the Department of Research and Isotopes. Primary responsibility for this area of research reactors rests with the Physics Section of the Department. Ongoins programmes and some future activities are briefly described along with past* activities to illustrate the types of programmes that are normally conducted.
Technical Committtee Meetings on the subject of Research Reactor Core Conversion to Low Enrichment Uranium Fuels have been held in Vienna since 1978. The results of these meetingt have been the
., following publications:
i TECDOC-233 - Research Reactor Core Conversion from the Use of Highly Enriched Uranium to the use of Low Enriched Uranium Fuels, Guidebook, 1980. TECDOC-304 - Research Reactor Core Conversion from the Use of Highly Enriched Uranium to the Use of Low Enriched Uranium Fuels, Cuidebook Addendum: Heavy Water Moderated Reactors, 1985. > Two additional publications are being planned: Guidebook on Safety and Licensing Issues Related to Core
, conversion - Licensin5 and Analysis - expected date, 1985.
Guide on Physical Properties and Irradiation Behavior of LEU l Fuels - expected date, 1986. Missicas Several missions of experts have visited counteles to discuss and help to plan Core Conversion programmes. Training Course on Conversion of Research Reactors to LEU Fuels was held in Caracas, Venezuela, in 1985. The Course was of four weeks duration. Technical committee Meeting on Core Instrumentation and Pre-Operational Procedures for Core Conversion was held in Vienna in 1983. TECDOC 304 on the subject was pub 11 shoo in 1984 l l'
- - - , - - - - - - , , - - + ,,e-,n -,w m-- - -- - - - - - - , - e,,w,-, m-,.v-.,- --m---,,.w- wm, , . , - - -
Paper No. 3 Reduced enrichment activities at GKSS t W. Krull GKSS - research centre Geesthacht GmbH The GKSS research centre is participating since 1980 in the national and international efforts on enrichment reduction to clarify the needs on safety and licensing issues, to decide on fuel tests, to perform fuel test and conversion calculations! These efforts are made within the IAEA working groups, the Geman AF program and CKSS R & D program. Especially within the fuel test GKSS has saccessfully tested 10 fuel elements, 23 places 45 %, UAlx , p = 1,41 g U/cc up to 64 % burnup. To-day casts are going on for 7 fuel elements, 21 places. 20 %,38 U 0 ' 0
- 3'I 8 UI""
4 fuel elements, 23 places, 20 %, 3U 2 S1 ' # " 3'I 8 UI"* l . 5 miniplaces, $ em x 18 cm, 20 %, U 3 SL 2 ' o = 4,75 g U/ce, fine , grain > 25 %. . Some of the fuel elements are instrumented with up to 8 thermocouples to perform loss of flow measurements. GKSS has done smae reactivity, microflux and temperature measurements. ' i Considerations are made for the conversion of our 5 MW-reactor FRG-1 and the 13 MW-reactor TRG-2 from 93 % to 20 % enrichment. l I
y Paper No. 6
' ' RAPTORT 4' AVANCEMENT SUR LA REDUCTION DE L' ENRICHISSEMENT 00 COM-
. BUSTIBLE DANS LES REACTEURS EXPER!MENTAUX DU CEA
PAR C. BAAS, M. BAANIER, J. BEYLOT, P. MARTEL ET F. MERCHIE 4 Resunt. s Le programme de qualification de combustibtu & bas enn.icM.ssmeM qui a dlbML dans te .ttacte.wt SILCE, il y a cinq ans, se pou,tsult conf ormtmeM aux. prLVisions.
! Apres L*LtradlMian de qua.cte plaques en aLLiage U3Si (S ,5 & 69/cm3) qui s'ut terninte en 1983, cette de quMrs plaques en a.llage USSi2 (2,0 &
S,49 /cm3)Lc d'un LitmeM complLc en aLLiage UI SL (6g/cm3) 1 dtbu.tl en l' 1984. Les Ltrndindons qu.i se totmineroM avant la fin de ce.tce annte, .
} se,tont suiviu de celle d'un litmen.c compte.C en alliage U3Si2 (S,2 9/cm3).
Ou muures sont regulLLtmeM ef f ectulu pendaM L,'LtondiMian (n.esu.tu
.d'apaisseur des plaques ec speettomLtt.it gammal.
Le rappor.c d'avancemeM precise les pn.incipaux. resulcMs obtenus jusqu'ici et les perspectives fu. cures. e 8 ^
r Papor No. 7 IRRADIATION PERFORMANCE OF REDUCED-ENRICHMENT FUELS TESTED UNDER THE U.S. RERTR PROGRAM J. L. Snelgrove, G. L. Hofman ( ANL), and G. L. Copeland (ORNL) ABSTRACT The fabrication and irradiation testing of high-density reduced-enrich-ment miniplaces has 3progressed during the past year. The fuels (uranium density range (Mg U/m ), enrichment) being tested are UA12 (3.0, M); Usi (3.8-3.9, L, M), U Si 3 U3 SiCa (5.3-7.0,2 3.9-5.2, M), U3 Si (4.8-7.2, L; 4.5-6.4, M), ' L;(4.8-5.7, 3.8-4.2, Lt M) , U 3Sij and U Fe (7.0-8.0, L). The medium-enriched uran,3 (5.1-6.0, lum (M) platesL;have 3.8-3.9, M), enrichments of6 40 or 45%. Postirradiation examination results for some of the higher-density l places with fission densities around 2.4 x 1027f ,3 are anticipated before , October. Channel gap and preliminary place-chickness sessurements have indi-i l cated the occurrence of breakaway swelling in 6.4 Mg U/z3 U 3 Si plates for fission densities of approximately 2.4 x 1027/md in both LEU and MEU places. There is also evidence of breakaway swelling in 7.8 to 8.0 Mg U/3 3 U6 Fe places at about 12x 1027 fissions /33 All U 3Si2 places are performing well to
- date.
l l'
- The irradiations of all 21 reduced-enrichment elements scheduled for the ORR have been completed.
Channel gap me,asurements indicate that all elements have performed well. Postirradiation examinations of ten of these elements have been completed to confirm the good performance. Several additional ele-ments are now being3 examined. It is anticipated that preliminary data on the six U3Si2 (4.8 Mg U/m ) elements will be available by October. Irradiations of U3Si 3 Sweden,2 and U Si elements are underway in SILOE in France and the R2 in and plans are being made for similar irradiations in the HFR-Peccan in The Netherlands and in the BR-! in Belgium. The paper will discuss t'io latest results and plans for the future. i l i l
4
- Paper No. 3 APSTRACT l
Licensing Considerations in Converting NRC-licensec Non-power Reacters 'crm Mich-enrichec to Lcw-enriched Uranium Fuels Robert E. Carter 1 j U.S. Nuclear Regulatory Ccmmission During the mid-1970s, there was increasing concern with the possibilit.y that hichly enriched uranien (MEU). widely used in non-power reactnes arrund the world, might be diverted frem its intended peaceful uses. In 11982 the U.S. Nuclear Regulatory Ccamission (NRC) issued a policy statement that was i intended to conform with the perceived international thinking, and that addressed the two relevant areas in which NRC has statutory responsibility, r namely, export of special nuclear materials 'cr ron-USA non-cower reactnes, l .and the licensing of USA-based nen-power reactors not owned by the Federal 3 government. To further address the second area, NRC issued a proposed rule' for public comment that would rer.uire all NRC-licensed non-power reactors using HEU to convert to low enriched uranium ILEU) fuel, unless they cculd demcnstrate a unique purpose. The preposed action-was not intended to require any licensee to ecnvert ferr one type of fuel to another (for examole, plate to red), to sustein significant decrecent in operating characteristics important in that reactor's use, to suffer significant interference with planned or in* ended reactor programs, or to be subject to burdenscre expenditures of resources that might lead to a decision to decommission the reactor. A sienificant ef#crt has been expended recently by the NRC staff in revising the orcpesed rule to be consistent with the above conditions, and yet to
;
- assure that HEU in most NRC-licensed reactors would be replaced by LEU shortly after acceptable fuel is available. An underlying principle guiding the
, staff is that as long as a change in enrichment does not lead to safety-related reactor modf #ications, and does not involve an unreviewed safety questten, the licensee could convert the core withnut prior NRC aceroval. At the time of writing this abstract, a regulatory merbnd of achieving this principle has not been finalized, but significant pregress is expected by fall 1985.
J a
- 1. Use of High Enriched Uranium (HEU) in Research Reactors; Policy i
Statement, August 24, 1982, Federal Register (47 FR 37007).
- 2. Limiting the use of Hiqhty Enriched Uranium in Ocmestic Research and Test Reactors, July 6,19Pa, Federe? Register (49 FR 27769).
4 i 4
i. L .. Paper No. 9
. Conversion of University Reactors to Low-Enrichment 'e Fuel--Issues and Prospects D. R. Harris * , J. E. Ma tos*, and H. H. Young **'
*Rensselaer Polytechnic Institute, Troy, NY 12181
+A gr onne National Laboratory
**U. S. Department of Enerzy i
. The highly-enriched uranium (HEU) fuel used in a score of United States
' university reactors can be viewed as contributing to security problems, and i
the U. S. Nuclear Regulatory Commission has issued a policy statement expressing its concern. The technical feasibility, functional impacts,
; licensing, and scheduling of conversion of these reactors to low enrichment fuel (LEU) have been assessed. The university reactors span a great range in form and function, from high power intense neutron sources where EEU fuel may
; be required, to low-power training and research facilities where EEU is
, unnecessary. Plans for conversion to LEU are facilitated by U. S. Department
, of Energy programs. Financial considerations are highly significant for LEU I conversion, but other problems are important as well.
c.- 5 t . e e 1 I! ~ l s 408 8
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r-Paper No. 10 Core Management and Reactor Physics Aspects of the Conversion of the NRU Reactor to LIU ABSTRACT The principal reactor physics effects of converting the NRU D2 0 research reactor from EEU to LEU are briefly examined, using the modellin6 methods established for that reactor. Changes in flux and power distributions, fuel rod usage, and reactivity values are considered, together with reactivity coefficients and some kinetics paramece'es. - In general the ef fects are ses11, less than 5%, although fuel usage is estimated to increase by 10%. No changes in reactor operating rules are required as a result of conversion and all reactor users can still be satisfied. Although there will be a slight loss of flexibility, no significant operational difficulties are identified. M.D.Atfield AEC Chalk River Canada i . e t
)
f
r: Paper No. 11 l Progress Recort on the Oesign of the Cccoact Core for Ucgrading the Munich Researcn Reactor X. Suning, W. G13ser, J. Meier, G. Rau, A. Rdhrmoser, t.. Zhang
.! Fakult1t fur Physik, Technische Universit&t MUnchen, 0-8046 Garching, West Germany 4
f Abstract 1 Our project is to modernize the Munich Research Reactor FRM using a f novel " compact core". This cylindrical fuel element is characterized by about 20 concentric fuel plate rings with an outer diameter of only about 20 cm an'd an active height of about 70 cm. It is cooled by light water, y ! which is not only technically more favourable than using heavy water but j also essential for the neutronic performance of this undermoderate small j core. The vertical tuce of the primary cooling circuit and the core are surrounded by a large heavy water moderator tank which will incorporate the e'xperimental f acilities, i .e. horizontal beam tubes and vertic'al irradia- -
; tion channels. The availability of the new high-density fuels is an impor-t j tant condition for the realizibility of this compact core, i
We report on a series of more detailed optimization calculations in which many of the relevant core parameters have been varied systematically. j These parameters include the overall core size, the dimensions of the fuel
! lattice, the enrichment and density of the fuel, the presence of a central ,
, control rod and of burnable poisen mixed with the fuel. The power density form factor can be decreased significantly by simply varying the fuel
!, enrichment or fuel density from ring to ring. The maximum value of the unperturoed thermal neutron flux in the 02 0 tank is predicted to be about 7 to 8 10 I# cm"2 s at only 20 MW reactor power.
I Paper No. 12 A COMPARISON OF THE RADIOLOGICAL CONSEQUENCES
, OF A HEU AND LEU FUELED RESEARCH REACTOR John G. Xollas
. Nuclear Technology Division
'NRC Demokritos
Greek Atomic Energy Commission i Aghia Paraskevi 153 10, Attiki, Greece . t 1 An analysis of the radiological consequences from the design basis accident (DBA)in the HEU and LEU fueled Greek research reactor (GRR) is pre-sented. GRR a 5 MW swimming pool type reactor is located within Athens the largest population centre of Greece concentrating 32% of its populat-
. ion. The' reactor is currently fueled with HEU and contains about 93% of U-235. From the ongoing safety analysis of GRR a loss-of-coolant accident accompagnied by a 20% partial core melting was chosen to represent its DBA.To estimate the source term 30 isotopes are taken into consideration and conserva-
! tive figures of fission product release are adopted. To estimate GRR's
. OBA consequences a CRAC2 consequence model version is used. Meteorological i data have been analyzed and a typical meteorological year of 8760 hourly .I observations has been selected. Population data were drawn from the latest census.
Doses and individual cancer risk frem exposure to the passing ra-dioactive cloud are estimated up to a distance of 20 km from the reactor site. Collective exposure and latent health effects due to initial expo-sure and ch'ronic exposure from inhalation of resuspended radionuclides and exposure to groundshine frem contaminated ground are estimated for the total Athens area of 3.081.000 inhabitants. The results indicate that the plutonittm isotopes buildup in the LEU fuel does not 1-ncrease appreciably
; the DBA consequences of the HEU fueled reactor. The plutonium impact concerns mainly bone effects and secondly lung and whole body effects.
The contribution to the limiting thyroid dose and the corresponding thyroid effects is insignificant, i b-I
f .. . Paper No. 13 I e i l L i i THE WHOLE-CORE LEU FUEL DEMONSTRATION IN THE ORR i t
- l'. 3 J. L. Snelgrove, M. M. Statscher, R. J. Corne11a ( ANL), and R. W. Hobbs (ORNL) l A3STRACT
, A whole-core demonstration of LEU fuel in the ORR is expected to begin l near the ciae of the Internacional Meeting at Pecten. Fuel elements vill
! 3
; contain U 3 Si 3.5 Mg U/a 3 .g at 4.8 Mg U/m and shia rod fuel followers will contain U Sig 3 at Shis rod assemblies (12) and/or fuel elements (100) are l currently being fabricated by B&W in the U.S.A., CERCA in France, and NUKEM in
- j. Germany, with deliveries scheduled to commence in August.
The primary objectives of P.he demonstration are to provide data for i . validation of LEU and mixed-core fuel cycle calculations and to provide a l large-scale demonstracion of the acceptable performance of produe ion-line l l U3 Si2 fuel elements. A detailed experiment plan is currently being I. ! prepared. It is planned.co approach the full LEU core through a series of i ; aixed corer where fresh LEU elements are introduced in place of REU f
- elements. Measurements to be made include flux (power) distribucion, l j reactivi:y swing, control rod worths, cycle length, fuel discharge burnup,
" gamma heating races, $,gg/1, and isothermal temperature coefficient. It is also planned to perfora seasurements on one or more critical configurations of fresh LIU fuel.
Preliminary safety approval has been received and the final saf ety assessment is being prepared. Key issues being addressed in the safety j assessment are fuel performance, zargin to burnout, and transient behavior.
- { The paper will provide details of the experiment plan and of the safety j s assessment.
1 i . I
- Paper No. la The Conversion'of the DIOC-Type Reactor FRJ-2 Studies and Conclusions A. Stremich, Ch. Siebert:, M. Wicker:
INTERATOM GmbH For the FRJ-2 (23 MW) of KFA-JQlich the conversion from HEU-to LEU-fuel was investigated. As this is the only heavy water mcderated research reactor in the Federal Republic of Germany our calculational methods had to be checked before starting the investigations. Therefore en one hand the IAEA-benchmark problem
'for heavy water reactors was calculated and on the other hand currently used FRJ-2 cores for HEU-operation were analysed.
In this way the qualification of the methods was achieved. There-after the U235-loading of the LEU-fuel element as well as the LEU core configuration could be optimized with regard to the cycle length, discharge burn-up, power density distribution and fuel handling precedure at end of cycle. Consequences on neu-tron flux values and distributions are shown. A transition phase from HEU- to LEU-core including mixed cores is investigated. The conversion procedure becomes very complicated and time consuming as the design for the tubular element must be modified when changing over to LEU in order to avoid high rejection rates during fuel element manufacture. i L
Paper No. 13 l i t l l Flux and Power Distribution in HEU, LEU and mixed TR:cA ssa - 14 MW cores with control rods AutorssC.Contesius C.:orgulis, D.Siminnovici. Abstract The Characterisa tion of LEU fuel in TR:CA SSR 14 HW core from the safecy and performance (in experimental loca Lons) point of view is strongly conected with accurace pin power distribution which in tur,n is strongly influenced by the unhomogenities ac che bundle edges (cwo dimensional spatial , effects) and by the control rods (threedimensional'afftLf$). A three dimensional neutronic comoucacional methods with control rods and bundle edge unhomogenities ,able to analyze HEUoLEU and mixed cores,has been assembled. For a measured HEU core configuration che calculated pin power distribution compares well with the experimental results on the whole core, and has o across each bundle and near the experimental locations, a di!!arent shape compared
, wLeh the Safety Report of the reactor.
Based on the above comparison with experimental results the method is applied to LEU and mixed cores. The computational machod is based on diffusion equation in chree spatial dimension, with cuo energy groups. The cross' sections for control rods are second stage homogenised in order to be good for accurate flux distribution in fuel. : The computational method togecher with the resulted
; ' gin power and flux distribucion for HEU, LEU and mixed cores are presenced in che paner.
1
e . Paper No. 16
~ .
Safety Analysis of the Upgraded Research Reactor JRR-3 T. Harami, M. Mirano, Y. Asahi, A. Kohsaka and N. Ohnishi Japan Atomic Energy Research Institute (JAIRI) 319-11, Tokai-Mura, Ibaraki-Ken, Japan JRR-3 is to be upgraded to a light water moderated and cooled, heavy water reflected 20 MWt pocl-type reactor with a 20% LIU plate-type fuel. The paper presents the analysis carried out for two of the design basis events, pipeline breaks at the heavy water loop in the pool and at the primary cooling loop. The first event can lead to skew of neutron flux due to dilution of heavy water. This initiates control rod withdrawal during the automatic centrol full power operation and inserts a reactivity in the core. The calculation showed that the transient safely comes to an and by monitoring thermal power instead of neutron flux. In the second event, LCCA, the core flow reversal from the forced circulation downficw to the natural circulation upflow is expected to occur just after the pool isolation frem the primary cooling loop. The analysis predicted tha: the peak fuel temperature and the minimum ON3R mee: the design basis criteria, althougn a sudden increase of the fuel temperature and a steep decrease of CNBR occur at the flow reversal. l l \ .. __ ______ . _ _ _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _
Paper No. 17 Studv on the Teneerature Coefficient of MEU and HEU C,res in the KUCA Keiji Kanda, Seiji Shiroya Masaaki Mori , Masatoshi Hayashi, and Toshikazu Shibata Research Reactor Institute, Kyoto University Ku=atori-cho, Sennan-gun, Osaka 590-0!.. Japan Depart =ent of Nuclear Engineering, Kyoto University Yoshida-honsachi, Sakyo-ku, Kyoto 606, Japan In accordance with the reduced enrichment for research and cast reactors (RERTR] program, the critical experiments using medium-enriched ! uranium (MEU] fuel in the Kyoto University Critical Asse=bly (KUCA] was launched in 1981. Thereafter, the KUCA experiments have been providing useful data vich regard to the international RERTR program. Recently, the measurements of the temperature effect on the reactivity was performed in succession to the study on the void reactivity, since the temperature coefficient as well as the void coefficient is a physical quantity closely related to the safety of liquid-moderated reactors. It is i=portant to investigate the effects of reducing fuel enric'hment on these quantities in advance to a core conversion from high-enriched uranium (HEU] to reduced enrichment uranium fuels. l The core employed in the present study was a light-water-moderated and heavy-vater-reflected one loaded with HEU or MEU fuel. The following effects on the temperature coefficient were investigated : (1) reducing the fuel enrichment, (2) leading pattern of the fuel places and (3) with and without boron burnable-poison in the MEU : ore. The investigated range of temperature was from 20'C through 70*C. The temperatures of light-water and heavy-vater were controlled to keep uniformity by the heaters and stirrers installed in the KUCA. The excess reactivity was measured by the positive period method at each temperature. The nessured data were analyzed using the code system SRAC to assess the computational technique for this quantity. l Through the experiments, no remarkable difference was observed between the temperature coef ficients of HEU and MEU cores, and the difference in the core configuration causes much more ef fect on this quantity than that caused by the difference in the enrichment. The calculated results approximately agreed with the experimental data. l
e - Paper No. L3 s STMTJS CF UT.J PPCGPA".S AT 2ABCDCK & .WILCOX KD3 BOCACIK RESEARCH AND TEST REACICR FTJEL ELE."DTr FACILITY triCHBUPG, VIPGIIIIA, U.S.A. ABSTRACT Within the Irv Enriched Progrra beira cenducted at Babcock & Wilecx the primary effcet has been to establish, fran past LEU develegnent wrk and current pecducticn technology, an efficient production prccess that maintains product quality for both LEU Calx and 03Si2 elements. This effort has allcwed the Babcock & Wilecx Carpany to successfully conplete a seccnd LEU producticn centract for the 2-W Ford Nuclear Reacter at the University of Michigan. Current U3Si2 centracts which include Standard and Centrol Elements for the Cak Ridge Peacter, SAPHIR Elements for the Swiss Federal Institute for Peacter Pesearch, silicide (U3Si2) powder for the Danish Riso !!aticnal Laboratcq and Elements for Sweden's R2 Peacter at Studsvik are being manufactured urder the same guidelines of quality ard efficiency improvernents. The transiticn fran develcpmental work to a pecduction process for pesder fabricaticn conpactingt plate and element fabricaticn along with inspection methcds are highlighted within this report. e
i Paper No. 19 Final Results of Qualification Testing of T3ICA (Uranius Zirconium-nydride) Fuel in the Cak Ridge Reactor Including Post-Irradiacien Examinatian by Robert H. Chesworth
. and -
Gordon 3. West GA Technologies Inc. P.0. Box 85608 San Diego, California 92138 U.S.A. AESTRAC* i As part of the qualification testing of TRICA (uranium zirconium-hydride, rodded) Icw enriched uranium fuel, a 16 rod experimental cluster was irradiated in the Oak Ridge Reactor (CRR) between December 1979 and August 1984 A total of 19 fuel rods, with three different fuel composit;Lons have been tested - Er-U-ZrH containing 20, 30, and 45 ' weight percent uranium. Individual rod pcwors have been in the range of 25 to 53 kW and the test cluster operated in-core for 901 full power days. At.575 full power days the rods containing 20 and 30 weight percent uranium
- were removed, having attained a target burnup in the range of 50 percent of the contained uranium-235. Destructive examination of one rod each of the 20 and ' 30 weight percent fuel rods showed normal perfersance with no anosolies.
Irradiation of the 45 weight percent rods was continued into the turnup
, range of 55 to 65 percent of the contained urantun-235, up to a test cluster exposure of 901 full power days. At this point, one of the fuel rods exhibited a clad failure due to steam build-up during a return to reactor power. Extensive analysis and post-irradiation examination led to the concluaica that the failure originated in a cladding defect which became manifest af ter fcur years and about 5 x 1021 nyt. Non destructive esamination of a large sampling of similar cladding - Incoloy-800 - has led '
to the conclusion that this was a unique defect and not a generio caaracteristic of this type of cladding material. The in-pile testing of the experimental cluster was reinitiated in July 1984 but was terminated shortly thereafter due to a cladding break caused by sechanical rubbing of the cluster spacers which.resulted from removal of the thermoccuple support member. The test was terminated at this point since all of the test rods had met or exceeded target turnups witnout difficulty.
l , . A comprehensive post-irradiation examinatica (PII) campaign has been established for the 45 w t-1 fuel rods as the final step in the qualification of TRIGA 1cw-enriched uranium (LIU) fuel. The PII includes: Dismetral and bow measurements of all of the 45 wt-1 r2ds Measurement of pressure inside intact rod 1089 or 1090 Identification and quantitative analysis of gas in free space inside of red 1089 or 1090 . Identification and quantitative analysis of gas contained in fuel matriz of red 1089 or 1090 Gamma scan of red 1086 to detemine peak-to-average turcup Nessurement of pressure and gas analysis of gasses in plenus of rod 1086 Perform metallography on red 1086 Perform burnup analysis on section from highest turnup region of rod 1086 Perform physical testics of selected cladding segments of rod 1086 The results of the post-irradiation examination and its imp 13 cations to the complete qualification of TRIGA LED fuel are discussed in this paper. f f o O
s Paper No. 20 MTR TUEL AT OERCA
$*ATUS OF 3EVELOPMENT 1935 Authors . Y. TANJAS - Ph. DEVE -
Since 1978 CERCA has been developing nigh densLty fuels allo-wing the use of LEU in replacement of HEU in research and test reac-tors. CERCA has de< eloped on its own funds the fabrication technology for three types of fuels . aluminide (U Alx), Oxide (U) Og) and sili. cides (U 3 51 2' U 3 31' U X Styl- foll wing densities can now ceen rea-ched :
. On industrial scale :
- With U Alx -
2,1 g U/cm3 e
" : 3,2 "
- U Og 3
- " U S1 "
3 2
- Ug Siy : 5.5
. On pilot scale :
-U St : 6,8 to 7 g U/cm3 3
for each of these types of fuels, an irradiation program has been set up. Thanks to international cooperation in the framework of the RER*R program, full size prototyp* fuel p 14 *.e s and fuel ele-ments have been irradiated in the USA (ORR, FNR), HOLLAND (HFR Petten), FRANCE (SIL0E) and JAPAN (JRR-2, JRR-4).
- So far, all the irradiation tests have been successful whatever the type of fuel, the uranium density, the uranium enrichment and the burn-up level :
- U Alx : satisfactorily tested from 1,7 to 2,3 g U/cm3
-U 3 Og : "
at 3,2 g U/cm3
- 4 -U SL : from 2.0 to 4,8 g U/cm3 3 2 I ,
-U 3 Si : at 5,5 and 6,0 g U/cm3 More testa are under preparation. In particular, tests eith a mixture of U 3 Si and U 3 S1 2 (Ux Sty) in the OSIRIS reactor.
The very good irradiation behaviour of silicides and the high densities they allow to reach have made them.the very fuels !. for reactor core ennversion. CERCA is presen*1y sanufacturing j 20 U 3 31 2 fuel elements loaded to 4,8 g U/tm3 for the full core l demonstration in the OAK RIDOE REACTOR,
a . Paper No. 21 i Status of Development and Irradiation Performance of Advanced Proliferation Resistant MTR Fuel at Nukem ., ABSTRACT I M. Hrovat H.-W. Hassel NUKIM GmbH, Postfach 110080 - D-6450 Hanau 11, FRG By the end of June 1985 the number of ordered full size fuel ele-ments to be fabricated by NUKZM for Material I,est and R,esearch ( MTR) Reactors with low enriched uranium (LEU) and m,edium enriched uranium (MEU) totalled 422. The fuel elements are being fabricated on production scale using dispersed CA1,, U0 38 and U 3 Si "*1** 2
; So far, 227 fuel elementes were delivered to the reactor opera-tors and inse,rted into various reactors. 87 fuel elements have reached their target burnup and were removed from the reactor for gost irradiation e,xamination (PIE). I The UA1 MEU fuel elements with a uranium density of 1.74 g U/cm2 meat reached a relatively high burnup of 73 4 235 0 in the '
Cak Ridge Reactor (ORR) . This burnup corresponds to a fission 1 density of 1.3 x 10 fissions /cm8 meat. The U 3g 0 and U3 Si 2 uel elements with uranium densities of 3.1 and 4.75 g U/cm8 meat respectively, reached with 82 4 235 U even higher burnup as the UA1, fuel elements. This burn-up corresponds to fission densities of 0.8 x 10-21 and 1.8 x 10 fissions /cm8 meat, respectively. In both cases LEU ses inserted.
I i i ! I a
- f The irradiation and pcs irradiation examinations on UA1, and UO 3g ORR experimental fuel plates (colled mini pisted were terminated in early 1993~. In both cases the burnups amount to l about 90 1 235 U. The investigations of the oxide layer for-mation during irradiation of there plates led to the results that j the 7.9 pm oxide layers of the unieradiated plates increased by ,
only 3.1 pm to 11 pm. The test plates with uranium silicide fuel of high density
, (U,Siy) are being irradiated. The irradiation of U,Si y test plates (majority U Si) with uranium densities up to 6.1 g l 3 l j U/cm3 meat in the FRJ-2 reactor at Julich is terminated.
! f Their burnup of 801 235 0 corresponds to a relatively high fis-sion density of 2.6 x 10 21 fission /cm8 meat. The U3 Si mini f j plates with so far highest uranium densities up to 6.7 g U/cm8 meat are being irradiated in the ORR. So far the reached burnup is about 45 % 3'U. 3
, All inserted fuel elements and experimental fuel plates performed
! well in endurance tests and exhibited an unobjectionable irra-
'diation behaviour, independent of fuel type.
r Paper No. 22 THE FABRICATION AND DERFORMANCE CF CANAO:AN S*LICIDE DISPERSION
. FUEL FOR TEST REACTORS -
J.C. Wood, O.F. Sears , L .C. B erth f aume, L .N. Kercert and J .D. Schaef er i
, AB STRACT Fuel fabrication effort is now concentrated on the comissioning of large scale precess equipment, defining product spec f f fcations, developing a quality assurance plan, and setting up a mini-comouter material accountancy system.
In the f rradiation testing program, seven full-size 12-element NRU assemol f es containing 20% enefched s11 fcf de disperston fuel have been irrad1ated successfully in the NRU reactor to burnups in the range 65-80 atomic percent.
. Irradiations have also been conducted on mini-elements naving 1.2 mm diameter holes in tnefr mid-sections, some deflied before irradf ation and others af ter frradiation to 22-83 atomic percent burnup. Uranium was lost to the coolant i in direct proportion to the surf ace area of exposed core materf al. Pre-fera-diation in the intact condition reduced in-reactor corrosion, probaoly because irradiation or in-reactor thermal effects had improved fuel-to-clad bonding cohesion.
l Fuel cores developed for the NRU reactor are dimensionally very stable, swelling by only 6-8% at the very high burnup of 93 atemic percent. Tw important factors contributing to this good performance are cylindrical clad restraint and coarse silicide particles. In wort afmed at defining specifi.
; cation limits we have irradf ated mini-elements containing very fine silicide
- particles dispersed in aluminum. The central regions of the cores of these I
sub-soecif fcation mini-elements transformed to aluminide and developed' enhanced porosity. With increasing burnup the pores (containing f f ssion gas) I accumulated in an annulus at the reaction front between the central aluminfde and the peripheral unreacted silicide dispersion material. radfated silfctde Thermal dispersion ramping testssegments fuels. Small were conducted in thereleased of fuel cores hot cells on gKr starting at about 793 K and pejking at about 853 K. After a holding period of 1 nour at
. 993 X a secondary c5Kr peak occurred during cooling (at about 600 K) i probably due to thermal contraction cracking. Whole mini-elements feradiated
! to 93 atomic percent burnup were also esmoed thermally. Encouragingly after
! about 0.25 h at 807 K the aluminum cladding developed very localized. small l blisters, some af th penetrating pin-hole cracks preventing gross of11owing or
- ballooning.
- Atomic Entry of Canada Limar.1
~
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, e . Paper No. 2] ; [ f
. ~
1
.n M.T.R. 's - E:C'EMENCI 'JITH P.CCERNISATION AND U7CRADING
.: i
~ '
licet existing reactors for research and irradiation were built in the 1950's. They were designed conservatively and have operated safety for more ch.c 20 years. Versatility and conservacias in the original designs has alloved the flexibility to extend both flux levels and experimental facilities. Pressures for =odernisation and renewal come from changing experimental ! needs, replacement of obsolete or vorn components and f ro= safety require- . l ments. In addition there are political influences affecting such factors ! l as fuel enrichment. : \ f l This paper highlights some of the significant aspects of a modernisation . I programme using illustrations drawn f rom ten years experience in this field. !
- l. ,
Examples will be taken from the Safety & Licensing area and the analytical work needed to establish the faulted canditions for a reactor. l Improvements in performance via enhanced control capability and the problems ' encountered in modifying or replacins sWsolete equipment will be described. I.A.Collis-Smith l GEC '4he ts tone U.K.
+
l l l L i t
l .. REB ~9AB P2per so. n ic:aw (ame:~n Minimum Claddine Thickness of Material Test Peactor Fuel Plates 4 J. Deckers For research reactoEs in the Federal Republic of Germany the average thickness of the cladding of fuel pistes is specified by German manufacturers at 0.35 mm or, as a function of the plate arrangement, at 0.36 to 0.41 mm. The minimum thickness may not fall below 0.25 mm. It is intended for the future to specify a minimum value of 0.30 mm for the average cladding thickness and a value of 0.2 mm for the minimum individual thickness at any location. This is in line with world-wide attempts of standard-1:ation for these fuel elements. This standardi:ation follows the successful use in the Oak Ridge HFIR research reactor of fuel elements with a specified minimum cladding thickness of 0.2 mm. The main objective of a study performed by TUV Rheinland was an assessment of the minimum cladding thickness from a safety point of view. The investigation covers the quality assurance measures puring fuel plate manufacture, the irradiation aad corrosion behaviour , of the fuel plates and fuel element damages. The presentation will give some detailed information regarding these aspects. The results of the investigation show that there are no objections against the reduction of the average cladding thickness to 0.30 mm and of the minimum cladding thickness to 0.20 mm for the fuels UA1 x and U 3g 0 as used in the irradiation experiments.
- - - - _ _ . - -_ ____._-_s. _ _ _ _ - - _ - . _ . - - - _. - - __ - - - - - - x . - . - -
.* O Paper No. 25 REDUCED-REACTIVITY-S'4ING LEU FUEL CYCLE ANALYSES FOR THE HFR-PETTEN J. R. Deen and J. L. Snelgrove Argonne National Laboratory Argonne, Illinois U.S.A. ABSTRACT The primary objective of these LEU fuel cycle analyses was to effect at least a 33: reduction in the reactivity sving now experienced in the REU cycle while minimizing increases in 235 U loading and power peaking. All LEU equilibrium fuel cycle calculations were performed using either a 19- or 20-place fuel element with 0.76-sm-chick mest and 0.5- or 0.6-=m-chick Cd wires as burnable absorbers and 16- or 17-p14te control red fuel followers with 0.76-sm-thick seat. Burnup-dependent microecopic cross sections were used for all heavy metala and fission products. A three-dimensional model was used to actaunt for ene effect of partially inserted control rods upon burnup profiles of fuel and of burnable absorbers and upon power peaking. Reactivity perf ormance equivalent to that of the reference HEU fuel c (420-g elempnes/290-g followers) was obtained in the LEU fuel cycles vt ch {cle 38U loadings of 500 g/340 g for the 19/16-place elements / followers and $25 g/365 g for the 20/17-place elements / followers. The equilibrium cycle reactivity sving (or, equivalently, control rod movement) was reduced by 50%. Average epithermal fluxes in the in-core irradiation positions were reduced by no more than 1% and thermal fluxes were reduced by 10 to 18%. Power density peaks occur primarily near the bottoms of low-burnup fuel elements or fresh fuel followers. The highest power density occurs in the f resh fuel follower upon reaching full power af ter startup when there are low Xe concentrations and maximum rod insertion. The magnitude of this peak is reduced by 11 when Xe equilibrium has been reached. Peak power densities at 50C in the LIU cores are ~3% higher than those in the HEU core for both the 19/16-place and 20/17-plate cosoinations. The equilibrium cycle length can be increased from 26 to 32 days by a 15: increase in the 23sU loading without significantly changing the reactivity swing.
o
- preliminary N:utronics Calculations npu ::o. 25 for the CSURR LEU Conversion /Uograce Program M. Caner and T. Aldemir Nuclear Engi.1eering Program The Ohio State University ABSTRACT The OSU Nuclear Reactor Lacoratory has emoarked on a facility modification and performance upgrade program, with the assistance of the Deoartment of Energy and ANL/RERTR. TL. goal is to have the reactor fueled to operate at a power level between 251 and 500 kW by 1988.
The OSU Research Reactor is of the swiming pool, natural convection type; it has a power level of 10 kW, and it is fueled at'oresent by MTR-type straight plate fuel elements with 93 enrichment. The present preliminary neutronics calculation is perfomed as part of the above mentioned program. The following two cores are considered:
- 1) A reference HEU core for the OSU Research Reactor, wnich serves the purpose of benchmarking the models used against available experimental data. (HEU fuel: UA1 alloy, 93 enrichment, 14.0 g U-235 per plate. ) -
- 2) A core s'imilar to the reference core, using the University of Michigan FNR type LEU plates (UA1x,19.5: enrichment, 9.3 g U-235 per plate), is assumed as a first LEU case.
The following parameters are being calculated: excess reactivity, ' flux and power distributions, and temoerature coefficients of reactivity. The University of Michigan code package is used: LECPARD for the cell calculations and 208 for the core global diffusion calculations (2-0, 4 energy grouos).
. . ?apor :fo. 27 r
j- , i LEU FUEL ELEMENT MANUFACTURING IN J.RGENTINA. THE STATUS OF LEU FUEL ELEMENT FANUFACTURING IN ARGENT!NA IS l REVIEWED. AFTER A BRIEF REVIEW OF THE ACLvr.LISHMENTS WHICH THE LEU PROGRAM HAS ACHIEVED UNTIL NOW, EM4 " ? IS PLACED ON THE PRESENT FUEL ELEMENT MANUFACTURING F, AND ON CURRENT PLANS l AND SCHEDULES. PROGRESS IN FUEL ELENCN. iCTURING HAS NOT BEEN i AS EFFECTIVE AS PLANNED BUT IS STILL CC' 4G IN SP!TE OF THE SEVERE ECONOMIC RESTRICTIONS WHICH THE C04. .nY IS UNDERGO!NG. l NEVERTHELESS THE UF6 70 U308 CONVERT!0N L4nE IS BE!NG SUCESS FULLY l COMPLETED, THIS BEING THE LA57 PHASE NECESSARY BEFORE THE PRODUCTION l OF THE PLATE-TYPE FUEL ELEMENT 9 CAN BEGIN. WE ARE THE POS! TION TO
! ANN 00NCD THAT, EVEN WHITH THIS SLOW BUT CONTINUQUS PROGRESS, l ARGENTINA WILL BE SELF-SUFFICIENT IN THE SUPPLY OF FUEL FOR ITS RA-3 REACTOR AND THE CORE OF THE RP-0 PERUVIAN REACTOR SEFORE M10 l l M!D - 1987. HAVING ACCOMPLISHED THESE OBJECTIVES, ARGENTINA IS IN THE l POSITION TO SE CONS!DERED A RELIABLE SUPLIER OF PLATE TYPE FUEL
! ELEMENTS. D.R. G!0RSETT! , l l 1 h e l l l f
O *
? aper No. 23 Refueling the RPI Reactoc pacility with Low Enrichment Fuel D. R. Harris, T. Rodriguaz-Vers. and F. Wicks Rensselaer Polytechnic Institute, Troy, New York The RPI reactor facility has operated for sany years with a core c' thin, highly-enriched stainless steel car-et fuel plates in twenty-five fuel assembly boxes supported in a large tank of light water. The reactor has been employed prisarily for teaching and research in support of the nuclear power industry. While the plate-type core is hignty flexible, it is not as suitable for its function as would be sore typical power reactor fuel. The RPI reactor is being converted to use 4.81 w/o enriched SPERT (7-1) C0P fuel rods. These rods are clad in stainless steel, which does not significantly impair their use in experiments in support of high-exposure fuel for the nuclear power industry, yet is compatible with the rest of the stainless steel system. The core structure requires substantial codification to support the longer SPERT fuel rods. The fuel assembly boxes are being replaced by a lattice plate structure in order to reduce costs and enhance lattice flexibility.
Structural stability is ensured by modification of the existing nassive place-and post support system. This design permits continuing use of the control rod system with the present fuel followers replaced by absorber. The redesign facilitates experiments related to power reactor startup as well as non proprietary critical experiuents for qualification of power reactor analysis setnods. e 6 l l l \
Papor No. 29 KUR Core Conversion - Neucr:ni:s Cal:ula::en Keiji Kanda, Masatoshi ifayashi, Katchiro Mishima, Yoshihiro Ns.tscce, and Tosnik.t:u Shiba:a Research Reactor Insti:ute,'$yoto Universi:7 . Ku=a:ori-cho, Sennan-gun, Osaka $90-04, Japan As one of possible fu:ure progrs=s for :he Kyoto Universi:y Re-search Reac:or (KOR), the Research Reactor Insti:ute of Kyoto Universi:7 (KURRI) has a plan for core conversion to the use of low-enriched uranium (LIU) fuels (see Table L for centative schedule), A feasibility study for this star:ed in November, 1983 as part of :he j oin: study between KURRI and Argonne Nacional Laboratory (MfL) . In L984, ther=al-hydraulic analysis was performed on the use of LIU fuels in the KUR. The results indicated no significant difficulties in core conversion in view of core thermal-hydraulics. During 1985, neutronics feasibility of the KOR core conversion was studied using the neutronics code system at A!(L . This paper presents some of the results obtained from the calculations. Table 1. Tentative schedule for the KUR core conversion.
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?2por No. 3C Abstract
- LEU-Plate Irradiations at FRJ-2 (OIDO) under AF-Programme W. Krug, E. Groos, G. Thamm In the frame of German AF-Programme treadiations of small test plates with LEU high density fuel are being performed at FRG-2 Gheesthacht and FRJ-2 Julich, i
The first treadiation campaign with ten test plates (3 plates containing silicide fuel, 2 ref erence plates with U33 0 ) has been finished October last yea'r at a maximum burn up of 2,41 x 1027m-3 PIE are under way in the hot cells of KFA. A d.iscription of the function and operation of the test fe;111ty as well as treadiation and recent PIE results will be presented. l 9 t
e ' Papar No. M A Comparative and Predictive Study of the Annual Fuel Cycle Costs for HEU and LEU Fuels in ene Hign Flux Peactor, Petten, 1985-1993 by R .L. Moss and P.May Ab s tr a ct The internationally agreed constraint on a va ilao tlity of supply of HEU fuels to Fesearch and Test rea cter s has necessitated that a coat analysis te carried cut to determine the financial effect cf converting the core of the HFP f rom
. H EU to LEU f uels . .
A computer p rogr am , written at Petten and cased on information ex tracted fecm studies in Europe and the USA, identifies the major ccst variables te te manufacturing, uranium, reprccess ing and transpor t cos ts . Ccepar iscn be tween HEU and ITU co*res have been carried out and includes the effects of inflation and exchange rate fluctuations. Co n v e r.- sion of the HFP core to IEU fuels is shown to be financially d isadvantag eo us. i 4 l
t .
?sper No. 32 DEVE* 0P.v2NT OF A 3-D .v. ENS:0NAL OALCULATION MODEL OF THE DAN!SH P23EAECH REACTOR OR3 00 ANALYSE A PFCPCSAL TO A NEW HEU CORE DESIGN CALLED RING-COPI. ,
E. NONB0L Riso National Lacoratory, OK-4000 Roskilde, Oenmark A project was started two years ago with the purpose of developing a 3-d imensional calculation model of the Danish research reactor DR3. Demands of a more effective utilization of the reactor and its f acilities had required a more detailed calculation tool than applied so far. The capacilities of the model should be to calculate reactivity enanges, burnup di-stributions and flux distributions in 3 dimensions. As for the flux it should also be possible to calculate this in the dif-ferent irradiation rigs in both the heavy water reflector and the graphite reflector. The model is based on diffusion theory with 3 energy groups. The equations are solved in 3 dimensions using finite-difference technique. The discretication is done according to the ususal 7 point scheme and successive overrelaxation is
' applied together with different accelerating devices (1,2 ] .
With a mesh size of 7 cm the total amount of meshes for the whole reactor comprises about 27,000. The coarse control arms of the reactor are treated separ-ately in the model. First the intersections between the CCA-outlines and the mesh grid are calculated as function of the
, CCA-angle. Applying this it is poss ible to determine the adjacent area occupied by the control arms for each mesh in the core. This area is then substituted into an expression for the neutron absorbelon cross section of the coarse control arms to be added to the fuel element absorti'on cross sections.
In Fig. I is shown a comparison between calculated and measured thermal flux in axial direction in a fuel element. The . results seem to agree very well. In Fig. 2 is shown the thermal flux in a horizontal plane in the center of the core. The pic-ture reflects a proposal to a new core design called ring-core-design where 4 central fuel elements are replaced by 4 dummy elements with heavy water. The purpose was to obtain a greater thermal flux in the center of tne core and the intenston seems to have been fulfilled quite well. Although the verification of the model still ts in progress, it seems as if it is possible to obtain reasonable results with
- the model. Re activ ity calculations on cri'tical configurations
! show agreement within 14. The intension ofausing tne model for l burnup calculations, flux distributions and reactivity calcu-lations thus appears to be possible.
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- I FIG.1. Comparison between measured and calculated thermal flux in axial direction in a fuel element. i l
THERMAL TLUX LN 71(E CCRE or DR3 l l RING cCRE DE3tCN ' t z(ccP)-o , j cca-se t t i : l : . s
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l % ./ ' D i I L .j FIG.2. Thermal flux in a horizontal plane in the center of the j j core. i REFERENCES l ' (1} KRISTIANSEN, G.K., CC4 to-day, RF-memo-183(1973). ' (2) LINDSTROM JENSEN, K.E., Development and verification of l Nuclear Calculation Methods for Light-water reactors. l Rise Report No. 235. i l [
d s CRITICAL EXPERIMENTS OF JMTRC MEU CORES ( II ) p,7,, go, 3: S. Shimakawa, Y. Nagaoka. K. Takeda. S. Koike. B. Komukai and R. Oyamada
,' Japan Atomic Energy Research Institute Oarai-macht Ibaraki-ken Japan l
i' . Ui ABSTRACT m. The critical experiments have been completed on the MEU
, (45%-enriched 235-U) fueled core in the JuTRC. the critical facility of the Japan Materials Testing Reactor (JMTR).
The analysis results were presented some parts of the critical experiments i .e. , cri tical mass, excess react ivi ty, control rod worths and flux distribution, at the international meeting of RETER. October, 1983. Tokal. Japan. Since then, the following experiments have'been carried out on the MEU fueled core, and their results are presented here. comparing with.the data on the HEU (90%-enriched 235-U) core. (1) Shut-down margin ( by Rod drop method and Pulsed neutron technique )
- (2) Void coefficient (3) Kinetics parameters ( A/2 )
( by Pulsed neutron technique ) In addition, all results of-the experiments are compared with those of neutronics calculations using SRAC code system. i
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- Paper No. 34 Development of Hea: Transfer ?ackage for Core Ther=al-Hydraulic
. Design and Analysis of Upgraded JR2-3 o
.a 4
.] - Yukio SUDO, Masahiro IKAWA and Masanori KAMINAGA Department of Research Reactor Operation, Tokai Research Establishment; Japan Atomic Research Institute (JAIRI) r
, Abstrac:
l A'hea: transfer package was developed for the core thermal-hydraulic design and analysis of the Japan Research Rese:or-3 (JRR-3) wh1ch is :o be remodeled :o a 20 MW: pool-type, ligh: water-cooled-reac:or with 20 : low enriched uranium (LEU) plate-type fuel. -This paper presen:s the constitu:1on of :he developed I heat transfer -package and the applicability of the hea transfer correlations adopted in 1:, based on the hea: :ransfer experimen:s in which :hormal-hydraulic features of :he new JRR-3 core were properly reflected. Major concerns in the development of the hea: l Oransfer package are :he differences in the hea: transf er character ' istics between upflow and downflow, both of which are adopted as the natural circulation mode and the forced circulation mode, respactively, for :he core cooling.
*e i
-i 'e Paper No. 35 Differential ther al analysis and metallogra bic examinations of U,Si.,-oowder and U.Si.,/A1 (35 o o)-miniolates
- a. ;.
Palle Toft Arne Jensen Atlas-Danmark A/S *) Advanced Engineering Division 3) A 11 S T R A C T The paper describes the thermodynamic and metallurgical behavior of U.Si,-powder a - l and miniplates at increasing temperatures. The min 1 plate core is U,$1.,/A1 (35 w/0) a - and the' cladding 6061 A1. Specimens of U 3 Si,-powder and specimens of the miniplates are heated and cooled at the same rate (200C/ min.) in a differential thermal analysis (DTA) apparatus. , The specimens are beated in air and helium to 14000C for the U Si -p wder and 3 2 to 7000C, 900*C,11000C and 14000C for the miniplates. Metallographic examinations and EDAX-analysis where carried out on miniplate specimens after DTA. The results { I from DTA of the U 3Si 2-Powder show a thermodynamic stable powder when heated in
, helium. When heated in air there was a heavy oxidation starting at ~ 1200C.
l The results from the DTA and the metallographic examinations of the miniplate specimens show when heated in air an exothermic peak at 4400C which probably is due to oxidation in the U 3Si 2/A1 - core. When heated in helium the miniplates j were thermodynamic stable. The metallographic examinations of the miniplate spe-cimens heated to 7000C, 9000C and 11000C show a formation of a ternary interme-tallic compound (U x Si A1.) at 640-7000C. All the U 3Sigin the core is trans-7 formed to the. ternary compound at this temperatur. The sececheometric composi-tion once formed, is fixed at increasing temperatures. The ternary crystalls show a rapid growth in the temperature range 900-1100*C. In the same range an eutectic composition between the ternary compound (U Si7 Al,) and microsegrega-
.j
- ted Al occured. All this were confirmed by the EDAX-analysis.
; The study concludes, that fuel plates will not remain stable under accident i.
condition. A fuel element will most likely melt down before a' stable structur is established. i
*) vork perforr'd under contract with Riso .ttianal L,boratory L
e ae Paper No. 36 ASSTRACT The 3enchmark Problem Revisited - The Accuracy of Finite Diff,erence
.. Methods and the Potential for the Use of Nocal Methods in ene Analyses of Research and Test Reactors W. L. Woodruf f Argonne National Laboratory Nodal methods have proven to be very useful in t.WR and LMF3R applications for reducing the computing costs by allowing a rather coarse node structure while stili maintaining an accurate solution.
The neutronics codes uset at ANL are capable of providing not only finite difference solutions but also aporoximate nodal dif fusion theory solutions in both hexagonal and Cartesian geometries and nodal transport theory solutions in Cartesian geometry. The IAEA benchmark reactor has been used as a base for this assessment. The reactor is modelled both with side plates represented explicitly and with the side plates homogenized into the fuel element. The nodal =ethod available prevides and assortment of levels of approximation for the flux and the leakage in addition to a choice of node structure for the problem. An extrapolated finite difference solution is taken as the standard for comparison. The results of this study are not without some surprises. The higher level flux approximations give improve =ents up to cubic, but the quartic flux approximation gives a poorer estimate for the eigenvalue than the cubic. Two polynomial approximations are provided for the leakage of either constant (zero order) or quadratic (second order). The quadratic approximation for leakage transverse to the axial direct-
- ion gives an improved estimate for the eigenvalue, while this choice in the axial direction exhibits convergence problems.
The nodal method was found to be very ef fective in reduciag computation costs, and the accuracy of the solution is often higher than that attained with finite dif f erence. In fact in order to have a finite difference solution of comparable accuracy the costs can be prohibitive. The usual choice of mesh structure may produce signifi-cant inaccuracies in the solution, and transport effects in many
. cases may be quite large. The use of transport nodal methods for some applications may be justified and far less expensive than Monte Carlo or Sn transport methods.
t l . i'
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ci . 0 SY $10 t PRESENT STATUS OF REDUCED ENRICHMENT PROGRAM FOR RESEARCH AND TEST REACTOR FUELS IN JAPAN Ichiro Miyanaga and Kazuo Kamei Japan Atomic Energy Research Institute 2-2 Uchisaiwai-cho, Chiyoda-ku, Tokyo 100, Japan and Keiji Kanda and Toshikage Shibata Research Reactor Institute, Kyoto University Ku=atori-cho, Sennan-gun, Osaka 590-04, Japan ABSTRACT The reduced enrichment program for the JRR-2, JRR-4 and JMIR of Japan Atomic Energy Research Institute (JAERI), and the KUR and KUHFR of Kyoto University Research Reactor Institute (KURRI) are in progress under the Joint Study Programs with Argonne National Laboratory (ANL), U.S. A. In parallel with the program, employing LEU fuel is pursued in upgrading JRR-3. Feasibility study for use of LEU fuel in the JMTR and KUR is now being carried out. LEU silicide min 1 plates have been irradiated succes,sfully in the JMIR. MEU critical experiments in the JMTRC and KUCA paved the way to the smooth conversion. The irradiation tests of full size MIU and LEU fuel elements are going on in the JRR-2, JRR-4 and JMTR. All these plans are in progress comprehensively together with the development of the code systems for analysis and design. INTRODUCTION Among 18 research reactors and critical assemblies in Japan (Tables 1 and 2), those which are relevant to the RERTR program are the JRR-2, JRR-3, JRR-4 and JKIR of JAERI and the KUR and KUHFR of KURRI (Table 3). The RERTR program in Japan has been pursued to the extent in which various circumstances can permit under the direction by the Five Agency Cocsittee on HEU, which has played the remarkable role in deciding policy related to the program and in close cooperation with w_ M
+
2 ANL through Joint Studies which have been very worthwhile (Table 4 Table 5). In JAERI the program for the first step, in which the JRR-2 and JMTR are to be converted to use of 45% EU UAlx dispersion fuel and the JRR-3 and JRR-4 to use of 20% EU UAlx dispersion fuel, is steadily in progress, hence it 1s approaching the final point. Further core conversion of the JNTR into use of 20% EU silicide fuel element with burnable poison has been studied since 1984 in accordance with Phase C of Joint Study with ANL. Along with this direction, silicide miniplates were irradiated in the JMTR and full size silicide fuel elements are to be irradiated in 1987. On the other hand, in KURRI the sa=e ef forts as in JAERI to reduce the enrichment of the KUR fuel is in progress. JAERI
~
JRR-2 The irradiation of two full size test fuel elements loaded with 45% EU UA1x (%1.6 gU/cc) has been under way since October, 1984. By early August, 1985, they are approximately 30% depleted. The irradiation lasts until December, 1985 to reach goal burnup of 40%. Full core conversion is scheduled in mid-1987. This schedule has been somewhat delayed by safety review. JRR-3 Safeyy review of the upgraded JRR-3 using 20% EU UAlx (%2.2gU/cc) was approved by the Government December,1984. The safety analyses are presented in this meeting. The removal of the old reactor body was permitted August, 1985. The upgraded core will reach criticality in 1989. JRR-4 The results of the neutronics measure =ents for full size test fuel ele =ents loaded with 20% EU UA1x (2.2 gU/cc) showed good agreement with the predicted values. The irradiation of these elements was started October,1984 and is continued in both of the JRR-2 and JRR-4 by loading one element in each reactor. The irradiation in the JRR-2 is to acceler-ate burnup (reached 22% by early August, 1985) and will last until April, 1986 to attain goal burnup of 35%, whereas the irradiation in the JRR-4 - will last until 1988. Full core conversion, though it was previously scheduled in 1987, is delayed at least one year because of the same reason as in the JRR-2. o m
f 3 "3 JMTR MEU fuel (UAlx, 1.7gU/cc) -
- 1. The critical experiments in the JMTRC have been successfully completed. The result is presented in this meeting.
- 2. The irradiation of two test elements (1.5 cycles, 4.5 weeks) was completed by March, 1985. PIES will be completed' by October,1986.
- 3. The inspection of fuel plates was finished July, 1985 at NUKEM F.R.G. They are now being composed into fuel elements. The full core demonstration is scheduled in mid-1986.
LEU fuel
- 1. Miniplate test Irradiation was finished June,1985 (2 cycles, 7 weeks, 25 s 34%
depleted), and FIEs are in progress. Fuel types, uranium density and number of miniplate are as follows: Type Uranium density Number A U3 Si2 4.8 gU/cc 4 B U3 Si2 5.3 gU/cc 3
,C U 3 Si 5.3 gU/cc 4 Size of miniplates: 50 cm x 150 mm ceat thickness 0.51 =m plate thickness 1.27 mm Under visual inspection and sipping test, they are pvoved to be sound.
The PIES of these plates,. are scheduled from Dece=ber, 1985 through May,1986 (Table 6) .
- 2. Ducmy fuel elements Hydraulic test by 2 dummy fuel ele =ents is planned to be performed in 1986, in order to confirm enough margin of mechanical strength of fuel elements against exaggerated coolant flow.
- 3. Full size fuel elements The irradiation test of two full size fuel elements is planned in
- 1987. The application for safety review of the irradiation test is prepared.
The elements are to be loaded with LEU of 4.8 gU/cc (mainly . U3Si2 ) . Insertion of cadmium wires into side plates as burnable poison is considered.
t 4 KURRI KURRI has had a joint study program with ANL since May,1978. The program is now in Phase C which is designed to study the conversion of KURRI reactors as shown in Table 7. The KUR has been successful in operation with 93% HEU, and will be converted to use LEU fuel when
'the new KUHFR is established. Even if the construction of KUHFR is delayed, the KUR will be converted to use LEU af ter the ANL-KURRI Joint Study Program Phase C is completed.
Since the 1984 RERTR Meeting in ANL, the following works have been performed, and some results are presented at this meeting. (1) Void coefficient =easurement and analysis for MEU and HEU cores. (2) Temperature coefficient measurement and analysis for MEU and HEU cores. (3) Neutronic calculation for use of LEU in the KUR. REFERENCES
- 1. K. Kanda, " Reducing Enrichment Program for Research Reactors in Japan",
I in Proceedings of the International Meeting on Research and Test Reactor Core Convension from HEU to LEU Fuels, Argonne, USA. I November 8-10, 1982, ANL/ RERTR/ TM-4, Conf. 821155, pp.24-32 (September 1983).
- 2. K. Sato, " Opening Statement to the International Meeting on Reduced Enrichment for Research and Test Reactors", in Proceedings of the i International Meeting on Reduced Enrichment for Research and Test Reactors, Tokai-mura, Japan, October 24-27, 1983, JAEFt-M 84-073, pp.8-10 (May 1984) .
- 3. K. Kanda et al., "Sta,tus of Reduced Enrichment Program for Research Reactor Fuels in Japan", in, Proceedings of the International Mee~ ting on Reduced Enrichment for Research and Test Reactors, ANL, USA,' October 15-18, 1984, ANL/RERTR/TM-6 Conf-8410173, pp. 11-20 (July 1985).
- m _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
, .i
~
Table 1. Japanese Research Reactors in Operation Name Owner Site Type and enrichment Max power Start-up date i JRR-2 JAERI Tokai D2 0 (CP-5) U-Al 93% 10 MW- 1960.10 UTR-KINKI Kinki Univ. Higashiosaka 11 0 (UTR) U-Al 90% 1W 1961.11 2 TRICA-II R1kkyo Univ. Yokosuka fly0 (TRICA) U-Zrli 20% 100 kW 1961.12 l TTR Toshiba Kawasaki 1I 0 (pool) U-Al 20% 100 KW 1962. 3 7 JRR-3 JAERI Tokai D2 0 (tank) NU 10 HW 1962. 9 UOy 1.5% v' TRICA-Il Musashi Kawasaki 110 7 (TRICA) U-Zril 20% 100 kW 1963. 3 MUSASIII Inst. Tech. KUR KURRI Kumatori 11 0 (tank) U-Al 93% 5 HW 1964. 6 2 J RR-4 JAERI Tokai 110 (pool) U-Al 93% 7 3.5 HW 1965. I JMTR JAERI Oarai II20 (MTR) U-Al 93% 50 MW 1968. 3 I l l YAYOI Univ. of Tokai fast U 93% 2 kW 1971. 4 Tokyo (horizontally movable) NSRR JAERI Tokai 110 (TRICA) U-ZrH 20% 300 kW 1975. 6 7
7, e'.,
. y l
Table 2. Japanese Critical Assemblies in Operation -l Name owner Site Type and enrichment Max. power Start-up date i SHE JAERI Tokai Craphite U 20% -10 W 1961. .l. horizontally split TCA JAERI Tokai H2 O (tank) UO 2.6% 200 W 1962. 8 l UO -Puo y 2.6% 2 NCA NAIC Kawasaki H2 O (tank) UO 2 1 % 4.9 200 W 1963, 12 e JMTRC JAERI Oarai H2 O (pool) U-Al 90% 100 W 1965. 10 FCA JAERI Tokai fast U 93% 2 kW 1967. 4 l U 20% horizontally split DCA NPC Oarai D2 0 (tank) UO 20.22% 1 kW 1969, 12 UOy -Puoy 1.5% KUCA KURRI Kumatori various U-Al 93% 100 W 1974. 8 U-Al 45% 1 kW l mult1-core '(short time) a
g_ - . - 7 Table 3. Research Reacters Relevant to RERTR in Japan
#8 "* #"*"#E "
h Name Power Critical Enrichment
- i l
KUR-1 (KURRI) 5 MW 1964 HEU + LEU 1989 I j KUHFR (KURRI) 30 MW (?) HEU + MEU (?) i JRR-2 (JAERI) 10 MW 1962 HEU + MEU 1987 JRR-3 (JAERI) 10 + 20 MW 1962 NU + LEU 1989 E JRR-4 (JAERI) 3.5 MW 1965 HEU + LEU 1988 JMTR (JAERI) 50 MW 1968 HEU + MEU 1986 Related Critical Assembly KUCA (KURRI) 100 W' 1974 HEU + MEU 1981 JMTRC (JAERI) 100 W 1965 HEU + MEU ' 1983
* + means the re'ducing enrichment in progress.
O e e ti 's
r
. 8 Table 4. History of Reduced Enrichment Program for Research Reactors in Japan 1977. 11 Japanese Committee on INFCE WG 8: started 1977. 11 Proposal of joint study by.Dr. R. Lewis (ANL) at the time of the application of export license of HEU for KUHFR. j J
1978. 5 ANL-KURRI Joint Study Phase A: started. 1978. 6 Five Agency Committee on Highly Enriched Uranium *: organized. 1978. 10 Five Agency Committee tentatively agreed to reduce KUHFR fuel enrichment from 93% to 45% as recommended by INFCE WG 8 in Japan. 1979. 2 ANL-KURRI Phase A: completed. l l 1979. 5 Project team for RERTR formed in JAERI. 1979. 7 ANL-KURRI Phase B: started. 1980. 1 ANL-JAERI Joint Study Phase A: started. I 1980. 8 ANL-JAERI Phase A: completed. 1980. 9 ANL-JAERI Phase B: started. f 1981. 5 MEU full core experiment in KUCA: started. I 1983. 3 ANL-KURRI Phase B: completed. 1983. 8 MEU full core experiment in JMTRC: started. 1983. 11 ANL-KURRI Phase C: started. i 1984 3 ANL-JAERI Phase B: completed. 1984 4 ANL- JAERI Phase C: started. 1984 4 MEU-HEU mixed core experiment in KUCA: started. 1984. 10 Irradiation test of MEU and LEU UAlx full size elements started in the JRR-2 and JRR-4, 1985. 3 Irradiation test of HEU UAlx full size elements and LEU UxSiy miniplates started in the JMTR.
- Until now, the committee has met 33 times. The member organizations are: 1. Science and Technology Agency
- 2. Ministry of Foreign Affairs
- 3. Ministry of Education. Science and Culture
- 4. Japan Atomic Energy Research Institute
- 5. Kyoto University, Research Reactor Institute (Observer: Naclear Fuel Industries).
9 Table 5. ANL-KURRI Joint Study Phase A (May 1978 - February 1979) (1) Feasibility Study of MEU (45%) Fuel. (2) Planning for Implementation of MEU (45%) Fuel: Critical Experiment Burnup Experiment Legal Procedure. Phase B (July 1979 - March 1983) (1) Detailed Calculations of the KUHFR with MEU (45%) Fuel. (2) Technical and Economical Evaluations of MEU (45%) Fuel, and Commercial Considerations. (3) Detailed Planning for Critical Experiments in the KUCA(C) with MEU (45%) Fuel. , (4) Application of Safety Reviews to Japanese Government for MEU (45%) Fuel to Be Used in the KUCA(C). (5) Fabrication of MEU (45%) Fuel for the KUCA(C). (6) Detailed Planning and Arrangements for Burnup Tests in the ORR and Post Irradiation Examinations at ORNL of Fuel Elements Using MEU (45%) Fuel Fabricated in the FRG. France and USA. (7) Fabrication of Fuel Elements for Burnup Tests in the ORR. (8) Performance and Analysis of Critical Experiments with MEU (45%) in the KUCA(C). (9) Performance and Analysis of Burnup Tests with MEU (45%) in the ORR. (10) Feasibility Calculations for Use of High-Uranium-Density , Fuels with LEU (<20%) in the KUHFR. l l (continued) l l l
10 (Table 5 continued) Phase C ( Nove=ber 1983 - March 1986) (1) Detailed Calculations of the KUHFR and KUR with LEU (<20%) Fuel. (2) Information Transfer for FNR with LEU Fuel. (3) Technical and Economic Evaluations of LEU (<20%) Fuel, and Co==ercial Considerations for KUHFR and KUR. (4) Detailed Planning for Critical Experiments in the KUCA(C) with LEU (<20%) Fuel. (5)* Application for Safety Review by Japanese Government for LEU (<20%) Fuel t'o Be Used in the KUCA(C). (6)* Fabrication of LEU (<20%) Fuel for the KUCA(C). (7) Detailed Planning and Arrangements for Burnup Tests in the ORR and Post Irradiation Examination at ORNL of Fuel Elements Using LEU (<20%) Fuel Fabricated in the FRG, France and USA. (8) Fabrication of Fuel Elements for Burnup Tests in the ORR. (9)* Performance and Analysis of Critical Experiments with L. (<20%) in th KUCA(C). (10) Perfor ance and Analysis of Burnup Tests with LEU (<20%) in the ORR. (11)* Application for Safety Review by Japanese Government for MEU (45%) and/or LEU (<20%) Fuel in the KUHFR. (12) Application for Safety Review by Japanese Government for LEU (<20%) Fuel in the KUR.
- If Japanese budget allows.
4 _. _ _ _ _ _ _ _ _ _ _ . _
11 Table 6. PIE Items for Irradiated Miniplates
- 1. Gamma Scanning ( for relative burnup measurements)
- 2. Absolute burnup measurement by chemical analysis
- 3. Measurement of swelling vs. burnup
- 4. Oxide layer thickness of cladding
- 5. Tensile strength measurements of plates i
- 6. X-ray radiography on plates for inspecting cladding defects
- 7. XMA (X-ray Micro-Analysis)
, 8. Measurements of blister temperature thresholds l
1 l 1 9
.a 4
e Table 7. Tentative schedule for the KUR core conversion Fiscal . Yea r 1984 1985 1986 1987 1988 '1989 1990 in use new (for 2 years) extra (for 1 year)? 93%EU-fuel = c __________ Burnup 25% ~ 35% Safety - Safety Thermal- Neutronics Fuel Tests Review Examination Ilydraulic Analysis (U3S17 -AO = = for LEU Analysis 8 20%EU-fuel = 2 1: Fab ricat ion Operation e
'fM
i h 1 RAPPORT D'AVANCEMENT SUR LA REDUCTION DE L'ENRICHISSEMENT DU CopBUSTIBLE DANS LES REACTEURS EXPERIfENTAUX DU CEA C. BAAS CEN/Grenoble M. BARNIER CEN/Saclay . J.P. BEYLOT CEN/Saclay P. MARTEL CEN/Grenoble F. MERCHIE CEN/Grenoble Le rapport prdsente l'dtat d'avancement du programme de qualification dans le rdacteur SILOE des combustibles & bas enrichissement. Ce programme comporte principalement des irradiations de plaques et d'didments complets U3Si et U3Sf 2 chargjs respectivement 6 g/cm3 et 5,2 g/cm . Des examens pdriodiques jusqu'& sont effectuds en cours d 'i rradia tion , en particulier des mesures d'dpaisseur de plaques ; Enfin, on prdsente les caractdristiques d'un didment mixte U3 51-U 3 S1 2 dont l 'i rradia tion est prdvue dans OSIRIS.
- 1. GENERALITES Ce rapport prdsente les rdsultats obtenus en 1985 dans le ddroulement du programme de qualification des combustibles & bas enrichi ssement dans les rdacteurs expdrimentaux du CEA.
Ce programme a deja dtd prdsentd dans son ensemble
& plusieurs reprises. Nous n'y reviendrons donc pas.
Depuis octobre 84, les principaux dv6nements & signaler sont les suivants :
- Poursuite de l' irradiation dans SILOE des 4 plaques U3Si2 . Cette irradiation ddma rrde en juillet 84 devrait se terminer & la fin d'octobre 85.
l
~ . - _ _ . _ _ . _ __- _ .-
2
- Ddbutde en octobre 84, l 'i rradia tion dans SILOE d'un didment standard U3 Si doit dgalement se terminer en t octobre 85. !
- Prdparation de l' irradiation dans SIL0E d'un didment '
standa rd U Sig qui prendra le relais de l 'i rradia tion de l'didment U Si ci-dessus.
- Ddcision de preparer l' irradiation dans OSIRIS d'un didment standard constitud d'un melange U3 Si-U3 Si2 -
Jusqu'& prdsent, toutes ces irradiations se ddroulent normalement et les resultats obtenus sont satisfaisants. Nous ddveloppons ci-apr6s chacun de ces points.
- 2. IRRADIATION DE 4 PLAQUES U 512 3 EN VRAIE GRANDEUR Cette irradiation a ddjh dtd ddcrite lors du prdeddent meeeting & ARGONNE en octobre 1984. Rappelons cependant que ces plaques fabriqudes par la CERCA ont des dimensions identiques & celles qui sont utilf sdes pour la fabrication des didments du rdacteur SILOE. L'alliage utilisd est un silicure de formule U Si2 3 chargd respectivement pour chacune des plaques & 2 ; 3,7 ; 5,2 et 5,4 g/cm3 en Uranium Total (cf. Tableau 1 ci-joint).
Ces plaques sont irradides dans un bottier spdcial ' qui est placd dans le rdflecteur du rdacteur. Elles sont retirdes apr6s chaque cycle de fonctionnement de SIL0E, c 'e st-n-di re apr6s 3 semaines d 'i rradia tion . Elles sont alors examindes visuellemen t sous eau et passdes ensuite sur un banc de mdtrologie immergd afin de ydri fier le
- gonflement du combustible sous irradiation par mesure de J
leur dpaisseur. Le rdsultat de ces mesures apparaf t sur la figure 1 pour la plaque la plus chargde (5,4 g/cm3). On remarquera ; que ce gonflement apr6s 11 cycles d' irradiation (courbe ' supdrieure) est tr6s faible puisqu'il est de l'ordre de 0,04 mm pour un taux de combustion moyen calculd dgal & , 59 %. Prdef sons qu'a ce taux de combustion moyen correspond un taux de combustion maximal d' environ 80 %. Les controles effectuds comprennent dgalement une l ddtection de rupture de gaine effectude avant chaque nouveau ' cycle de fonctionnement du rdacteur. o , i i
3 Enfin, il doit stre mentiennd qu'une campagne de controle par spectrometrie gamma a 4td effectude & mi-irradiation afin de recaler les taux de combustion calculds des 4 plaques. Cette campagne sera h nouveau effectude sur l' ensemble des plaques & la fin de l 'i rradiation . Les rdsultats obtenus jusqu'ici confi ment donc l' excellent comportement de l'alliage U Si 3 2 pour des charges importantes et pour des taux de combustion dievds.
- 3. IRRADIATION D'UN ELEMENT STANDARD U3 Si Commencde en octobre 84, cette i rradia tion doit se terminer en octobre 85.
Elle constitue le prolongement logique de l' irradiation de quatre plaques U3 Si effectude en 1982-1983 dans des conditions analogues A celles des 4 plaques U3Si 2 ddcrites plus haut. L'didment, fabriqud dgalement par la CERCA, est absolument identique geometriquement aux didments combustibles utilisds normalement dans le rdacteur SIL0E. En ddbut d' irradiation, la masse d'U235 contenue dtait dgale & 507 g, l'alliage U 3Si utilisd dtant chargd & 6 g/cm3 en Uranium Total. Deux des 23 plaques ne sont pas serties et sont donc amovibles, ce qui perme. & chaque arrat de cycle de les retirer sdpardment et de mesurer leur dpaisseur, c'est-h-dire le gonflement du combustible sous irradiation. Les rdsultats obtenus appa raissent sur la figure 2. Aprbs huit cycles de fonctionnement, le taux de combustion moyen est supdrieur h 50 % tandis que le gonflement observd est de l'ordre de 0,08 m. Cette valeur est tout-A-fait compa rable & celles observdes sur les 4 plaques U3Si irradides prdeddemment. Enfin, il doit etre mentiennd que cet didment, comme tous les combustibles qui sont testds dans SILOE, fait l'objet d'un controle de rupture de gaine avant chaque cycle, d'un examen visuel aprbs chaque cycle et d'un controle par spectromdtrie gamma en cours d' irradiation et en fin d' irradiation pour recalage du taux de combustion calculd. _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ~
I 4 i 4. IRPADIATIONS PROGUMEES 4.1. Irradiations prograsundes dans le rdacteur SIL0E l L' irradiation de I'didment U3 Si ddcrite prdeddemment dtant arrivde & son terme, celle d'un nouvel didment fabriqud l
& partir d'un alliage U Si2 3 chargd & 5,2 g/cm3 d' Uranium Total, soit une charge totale d' Uranium 235 dgale & 434 g va prendre immddiatement le relais. Cet didment comporte l dgalement 2 plaoues amovibles.
Le programme d' irradiation de cet didment U3 Si 2 sera identique & celui de I'didment U3 St, & savoir : l~ - & chaque intercycle : 1
. examen visuel
. contrale de rupture ;e gaine
. mesure de gonflement du combustible par mesure de l'dpaisseur des 2 plaques amovibles.
- & mi-irradiation et en fin d' irradiation :
. contr61e du taux de combustion par spectromdtrie
\ ga ma. I En plus de la mesure de la largeur des deux plaques amovibles, nous prdvoyons d'effectuer rdgulibrement le , contr61e de la largeur des canaux d' eau (canaux interplaques). L'appareillage qui sera utilisd pour ces mesures a 4td I testd de fagon satisfaisante sur I'didment U 3 Si et pourra permettre une ydrification d6s le ddbut de 1' irradiation. Cette irradiation sera probablement suivie de l 'irradia tion d'un deuxieme didment U3 51 2 dont la charge
' sera directement reprdsentative de celle des didments qui composeront le futur coeur & bas enrichissement de SILOE.
4.2. Irradiation prograsunde dans le r1!acteur OSIRIS Ce rdacteur utilise depuis cinq anndes le combustible CARAMEL & bas enrichissement. Cependant, 11 a paru intdressant d'y effectuer dgalement une i rradia tion d'un didment combustible fabriqud & partir d'un alliage silicium afin de recueillir le maximum de donndes sur le comportement des combustibles & base de silicium.
p . l Cet didment est actuellement en cours de fabrication
& la CERCA et son irradiation ddbuterait au cours du premier semestre 1986.
Ses principales ca ractdristiques apparaissent sur le tableau 2 ci-apr6s. L' irradiation de cet didment durera environ une annde. TABLEAU 2 Principales caractdristiques de I'dideent & f rradier dans le rdacteur OSIRIS
- DESCRIpTIF DE L' ELEMENT RETENU Pas OSIRIS : 87,4 mm Alvdole carrd : 84,4 mm Eldment standard 1 21 plaques planes toutes identiques Charge totale en 235U = 500 g/didment complet Epaisseur ame fissile = 0,6 mm Gainage AG 3 = 0.38 mm Epaisseur totale plaque = 1,36 mm Canal intdrieur = 2,55 mm Plaquettes absorbantes en partie supdrieure MATERIAU FISSILE Melange U3512-U351 d = 4,66 g U/cm3 Enrichissement 235U = 19,76 %
Pourcentage Al en poids = 20 %. l
6 CONCLUSI0ft Tous les essais de qualification actuellement en cours ou programmds dans nos rdacteurs concernent des combustibles
& base de silicium. Les rdsultats obtenus jusqu'& prdsent sont satisfaisants et permettent d 'ef fectuer des dtudes plus ddtailldes sur les ' performances optimisdes que l'on peut attendre de coeurs complets constituds avec ce type de combustible.
4 i f 9 'u
m . M
. m._ .
PLAQUE U3SI, (5,4 g/cc) CYCLES : 00-02-0"-.09 'I 00 80 .. @ e C 60 _~ a-a 40 . 34 FIGURE 1 7 a
- 20 *
,00 !SO 60 14 0 120 60 0 l - i:80 t%- 3 - 3
- c0 l
g l:40 t ':20 l TRACE : 5 11 0 .
. ..-_.2 ____.____m m m .
m 0 100 200 300 400 500 000 COTE (T1f1)
IRRADIATION DANS SILOE ," DE COM80STIBLE BAS ENRICHISSEMENT i Etat d'avancement SEPTEMBRE 1985 charge Entree Sortie Nombre Mombre Variations gg Type initia}e gut /an en pile de pile de cycles de cycles prevus " d'epaisseur Observations realises mm 5,5 10/93 46 % (M) 0,05 o plaques 5,5 g 11/s3 53 % (M) 0,06 Bon comporttment U
- Si 6 11/93 54 % (M) 0,105 general 6 10/93 43 % (M) 0,095 2 64 % (E) 0,030 Bon comportement O plaques 3,7 59 % (E) 0,040 general 5,2 OW hh6 ND H U,Si, 85 * "
5,4 59 % (E) 0,040 realise fin 6ene cycle Elenent complet 3eme 45 % (E) 0,060 Bon comportement U, Si 6 10/94 trimestre N9 8 general (2 plaques amo. ) 85 53 % (E) 0,000 Fnman y scanning realise fin 4eme cycle U s 8A 5,2 hre N9 Irradiation en a preparation (2 plaques amo.) as TABLEAU 1
( Plaque de l'515 ment U351 (6,0 g/cc) , l ! CYCL.ES : 00-05-06-0'l-08 10
- E 10 g C
' ;o - 5 na
.o y e.,
'm FIGURE 2
)0 to o-20 -
v co F " % *e .?o em g 2 r No f
- o -
,:0 , TRACE : 5 30 e . c0TE inn)
j,. Papertiu. T 4 IRRADIATION PERFORMANCE OF REDUCED-ENRICHMENT FUELS TESTED UNDER THE U.S. RERTR PROGRAM ! J. L. Snelgrove,* G. L. Hofman,* and G. L. Copelandi
*Argonne National Laboratory, Argonne. IL 60439 U.S.A. l t0ak Ridge National Laboratory, Oak Ridge, TN 37839 U.S.A. '
l i ABSTRACT Considerable progress in the irradiation testing of high-density, reduced-enrichment fuels has been ande during the past year. Minipistes containing UA1, U3 Sig, U3 Si g 3, U 3Si, U SiCu, 3 and U6 Fe have been irradiated. footirradiation examina-tions have revealed that breakavey swelling has occurred in 6.4-Mg U/m3 U3 31 plates at ~2.8 m 1027 , fissions /m3 . U3 12 plates are continuing to show satisfactory per ormance. The testing of full-sized fuel elements in the ORR and the silos reactor have continued with good results. Poetirradiation examinations are confirming the satisfactory performance of these elements. INTRODUCTION Significant progrees has been ande during the past year under the U.S. Reduced Enrichment Research and Test Reactor (RERTR) Program in assessing the irradiation performance of high-density, reduced-enrichment fuels. The irradiation of a second series of etniature fuel plates (minipistes) has continued in the Oak Ridge Research Reactor (ORR) and poetirradiation examinatione (PII) have begun at Arganne National Laboratory (ANL). Evidence of fuel per-formance limits is beginning to emerge. Irradiation of full-staed
. .g
~
[ fuel elements has been completed in the ORR and continues in the SILOE reactor in France. All elements appear to have performed well. A summary of the significant results obtained during the past year is presented in this paper.
, MINIPLATE TESTING As reported last year,I a second series of miniplate
- irradiations is in progress to establish the performance limits of U3 Si 2 and U3 Si fuels, to determine whether a small copper addition dll improve the performance of U3Si, to show that highly loaded UAl2 dispersion fuel performs well at medium enrichment, and to e investigate the basic behavior of U6 Fe dispersion fuel. The test matrix, shown in Table 1, has been expanded during the past year to include USi dispersion fuel since, according to the uranium-silicon phase diagram, the depletion of uranium (through burnup) would tend to transform U3Si2 into USi. Therefore, knowledge of the irradiation behavior of USi might help in understanding the behavior of U3 Si2 at high burnups. Also, a few highly loaded U 3 Si 3 and U3 Si miniplates containing natural B C 4 as a burnable poison in the fuel meat will be irradiated to determine if the helium gas formed has any ef fect on the fuel meat swelling. The test matrix also contains six U Si 3 miniplates fabricated by NUKEM and one U 3 Si and five U3 Si2 miniplates f abricated by the Argentine a CNEA. It is expected that the last of the miniplates to be fabricated will begin soon. The status as of September 19, 1985, of the miniplate irrad-1ations is shown in Table 2. The first four modules to be irrad- - lated have been removed from the reactor. Some of the miniplates in these modules have been selected for PIE and others have been . placed in new modules for continued irradiation. The burnups and fission densities have been estimated on the basis of flux -- measurements and represent an average for each module. Channel gap measurements have been made at the end of each reactor cycle, 8 and the results have proven very useful in determining the onset - of accelerated fuel meat swelling. The PIE of twenty-five miniplates began at ANL during /.ugust 1985. To date, visual inspections, gauma scans, and plate thick-ness measurements have been made. The results of the thickness measurements are shown in Table 3. Note that t he thickness changes include the oxide layer buildup on the cladding surf ace. , The gamma scans indicate that substantial axial burnup gradients (peak / averages of 1.15 to 1.25) exist owing to the axial flux profile in the top half of the ORR core, where all of their irradiation occurred. This leads to a large range in thickness , change for plates in which burnup limits are being reached. The t U3 Si miniplates loaded to ~6.3 Mg U/m have experienced breakaway 3 swelling (pillowing) in regions where the fission density is
~ , e .
b 3
.Toble 1 Test wateln for the Second Series .
of Minlplete Irradiations (September 1985) i Densi , Number of irrediated in edule ru.i ryp. Ma u/ rarier.,ent* pi.tes nnnnanna 29 UAl 2 3.0 M 2 2 23 U$l 3.7 - 3.8 0 3 3 3.8 - 3.9 L 4 4 3.8 - 3.9 M 3 3 U 51 3 2 4.8 0 2 2 8 4.8 - 5.7 L 21' 3 8 2 1 3.9 - 5.2 4 4 1 3 39 - 1.7 H 2 1 1 19 U3SI 6.1 - 6.2 0 2 2 d 4.8 - 7.2 L 19" 2 4 6 I 9 gg 4.S - 6.4 M 5 5 2.0 H 2 2 U3Sicu 5.3 - 7.0 L 9 2* 4 3.8 - 4.2 M 3 3 U3Sit .5 5.1 - 6.0 L 5 3 2 39 3.8 - 3.9 M 2 2 U6Fe 6.9 - 7.1 0 2 2 7.0 - 8.0 L 4 2 2 8
- 0:0.21, L 19.4-19.81, W:40.0-44.81 H:92.6-93.01.
D Four conteln 8 C dispersed in meet. 4
" Originally Irradf ated in meule 17 '
4 0riginally irradiated in Modulee 17 and 22.
' Fuel particles partially coated eith ouillte (3Al23 9 '2$102 3 '" **"#*# **d
' pestle.
Condif fons of Note e opply for one plate. 9 0riginally irradlated in mdule 23. bree contain 8 C4 dispersed in emet. l h
. e, 9
4
/
. Table 2. Status of Second Series of Miniplate lerediations (End of Cycle 173-B, September 19, 1985)
Exposure. Es t. Full-Power Burnup Fuel Est. LoadedAve. Fiss. Plate at Dens. Enr. of2'**3 10 '/m
%dulo Loestion Dev a i Type . M H 8
17 PIE. 25 174.65 41 U 51 3 2 0.9 (11/15/84) U3SI 1.3 2.3 1.6 18 PIE 254.74 80 U SI 2.1 3 2 (3/14/85) U SI 3 2.4 22 PIE. 25* 126.74 34 U SI 1.2 1.2 3 2 (1/20/85) U 51 3 1.1 U6 I* I*2 23 Pool, 29' 9 1.46 23 UAl 2 0.7 . (4/3/85) U Si 3 2 1.0 0.8 U3 SIy ,3 0.6 U3Si 0.7 U6Fe 0.8 24 Pos. 4 228.48 c U3512 U 51 3 U6 E' 8 25 Pos. 5 1M.37(25) U 51 3 2 I'7 309.02(17) 72 U3SI 2.4 261.11(22) 68 27 Pos. 3 62.63 44 USI 0.8 U 51 I'I 3 2 U 51 3 1.5 1.0 -- U3 Sicu 1.1 28 Pos. 2 62.63 31 USI 1.0 1.1 I*I U3Sil .5 U 51Cu 1.1 3 29' Pos. I 24.87(29) UAl 2 0.8 116.33(23) 26 U3Sl2 1.2 0.9 U351 g ,3 0.7 i U SI 3 0.8 e Module containing selected plates for continued Irradletion. D 0 ate of last Irradiation.
"All fueled plates contain depleted uranium, d
Contains plates previously irradiated in %dules 17(3 pl.) and 22(8 pl.).
'Contains plates previously irradiated in mdule 23(10 pl.).
5 Table 3. Thickness Changes of Miniplates Irradiated in Modules 17, 18, and 22 Est. Fiss. Thicknegs No. of U Dens., Dens., 1027/m3 Change, um Fuel Enr.a py,ge, ggf,3 Ave. Max. Ave./ Plate Range U3 Si2 L 1 5.2 0.9 1.1 25 23-36 L 8 5.1-5.7 2.0 2.3 18-48, 0-109 M 1 4.0 1.2 1.4 15 15-18 H 1 1.7 1.2 1.4 51 51-51 03 Si L 1 7.1 1.3 1.6 20 10-30 L 4 6.2-6.3 2.4 2.8 206-328 107-681 M 2 4.5 1.6 2.0 61-81 43-102 M 3 6.2-6.4 2.3 2.9 58-218 25-635 H 2 2.0 1.6 2.0 53-64 48-94 U6Fe L- 2 7.8-8.0 1.2 1.4 254-373 51-889 at:19.8%, M:40.0%, H:92.6-93.0% b Includes oxide layer of 13 to 25 um thickness. estimated to be ~2.8 x 10 27 f ,3, both for 19.8% and 40.0% enrich-ments. It is also seen that similar swelling occurred in the U6 Fe ., miniplates, but at one-half the fission density. Optical metallo-graphy has confirmed that the growth of fission gas bubbles in the fuel particles is in a state of breakaway swelling. In a manner similar to that previously observed in U3 SiA1, large bubbles have linked across particles, forming a pillow. The U 3Si2 miniplates continue to show good behavior, although there is evidence that the swelling of the highest loaded plates is beginning to accelerate at fission densities approaching 2.3 x 1027/m3 , FULL-SIZED ELEMEttr TESTING The irradiation of all of the full-sized elements scheduled for the ORR has been completed, as shown in Table 4. Channel gap measurements during irradiation indicated satisf actory perfor-mance. PIE has begun on six of these elements during the past year, including five U3 Si2 elements. Except for blister testing of selected plates, the PIE of NSI202, CSI201, and CLE204 are essenti-ally complete. Plate thickness measurements in the area of peak burnup indicated that swelling, corrected for the oxide buildup
Table 4 Status of Full-Stred Fuel Elements Irradiated in the ORR (October 1985) Element Fuel No. of 2350 Depletion, Enrictment, Present initlet U Inittel irred. PIE No.* Type Plates 5 tocetton Dens.. Mg/m I 2330 , g 5 Compt. In Prog. (bapt. T291X U038 19 45.0 Pool-Rd 1.7 280 56 10/13/88 X T2924 Up, i9 43.0 iRtEt* i.7 280 72(70>* 4/i6/82 x T293X U3 0, M 45.0 Rep m . 1.7 ISO 55 10/1 W 2 e T294X 00, 3 19 45.0 Reproc. l.7 28 0 58 6/22/82 e NLE451 UAI, 19 44.9 tRLEt 1.7 284 73(75)* 10/11/82 X NtE452 UAl 19 44.9 Pool-R I.7 284 59 6/11/82 X CLE451 U4l 19 44.9 Pool-R I.7 282 76(71)* 11/18/82 X CLE452 I;Al, 19 44.9 Pool-R I.7 28 2 56 4/16/82 X CLE453 UAI, 19 44.9 DRLEt 1.7 284 75 9/95/83 X l NtE208 U0 13 19.6 Pool 2.* 340 77 t/20/85 38 m NLE202 U03g 13 19.6 M 2.3 340 58 5/30 M l CLE201 UAI, 83 19.8 Pool 2.5 312 56 5/30/e5 CLE202 048, 13 19.8 Pool 2.3 336 78 I/20/B5 CLE203 00, 3 18 19.7 Pool 3.2 326 74 4/22/84 CLE204 003g 18 19.7 FRtEt 3.2 326 57(54)* 9/29/83 X NSl201 U3Sly 19 19.7 DetEt 4.8 340 49(41)* t/14/83 X NSB202 U 35ty 19 19.7 DELEt 4.8 340 82(82)* 8/l4/84 X CS1201 U3Sly 19 19.8 tRtEt 4.8 339 56(54)* 10/13/83 X C$l202 U3Sly 19 19.8 tetR 4.8 339 82 8/14/84 X l 851201 U 59 19 19.8 tRLEt 4.8 339 55 4/22/84 X 3 7 l 851202 U3Sly 19 19.8 DRtEL 4.8 339 82 12/19/84 X
*Ftrst letter in element no. designetes fabricator Babcock & Wilcow (8), CERCA (C), NUKEM (N). or Teres Instruments (T).
b l High Radletion Level Enemination Laboratory,4PIEl.
- Depletion in perentheses based upon prellalnary evoluetions of measurements, d
Avelting shipment for reprocessing.
*>h Plo planned.
1
w 7 averaged 76 um (range of 20 to 127 um) for NSI202, 51 um (range of 8 to 76 unj35for CSI201, and 61 um (range of 41 to 89 um) for CLE204. The U burnup in these areas were 963,' 67%, and 70% for NSI202, CSI201, and CLE204, respectively. Figure 1 shows the microstructure of the fuel meat in the high-burnup areas of fuel plates from NSI202 and CSI201. The large bubbles have been shown, by comparison of unirradiated and irradi-ated samples in a scanning electron microscope, to be formed in particles which initially were predominantly U 3 Si. However, no areas were observed with any significant linking which could lead to breakaway swelling. Current specifications and production tech-niques for U3 Si2 fuel alloy should substantially reduce the amount ! of U3 Si in fuel being produced today. Some fabrication porosity ! remains in the low burnup end of NSI202. The structure of the meat in CSI201 is similar, but shows less incidence of particles contain-ing gas bubbles, consistent with a lower initial concentration of I U3 Si particles. The fabrication porosity has been consumed in the high-burnup region of CSI201, and less remains in the low-burnup region than in NSI202. This is consistent with the low value of porosity (4%) sessured in an unieradiated plate from the CS1201- l CSI202 production. All evidence indicates that these full-sized l U3 Si2 elements have performed in a perfectly satisfactory manner. l The U 03 8 element (CLE204) showed structure typical of that ob-
; served in the highly loaded U 30g miniplates and the sedium-enriched full-sized elements,2,3 There was extensive reaction with the matrix aluminium; however, significant matrix aluminum remained. No I fabrication porosity remained. The fuel particles showed a large central void with a network of smaller bubbles around it. There was no significant linkup of these large central void to indicate that breakaway swelling was imminent. The fuel showed entirely satisfactory performance at this burnup level.
Irradiation of RERTR elements in Eurofean reactors under co-
! . operative arrangements has continued, as shown in Table 5. All , results have been satisfactory thus far, and some more detailed I results are reported elsewhere." Discussions concerning similar .
irradiations of uranium-silicide fuels in the HTR-Petten and the BR-2 are underway.
SUMMARY
Substantial progress in the characterization of the irradiation behavior of high-density, reduced-enrichment fuels has been made during the past year. It has been found that burnup 11mits exist for low-enriched U3 Si fuel at high loadings and that Ugre does not perform well under irradiation. The irradiation and PIE of full-sized elements containing U 3Si 2' U Si,and 3 U 03 8 at somewhat lower uranium densities have confirmed the satisfactory performance of these fuels.
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1 1 Tetle S. Status of Full-Sized fuel Elements Scheduled for Irradiation In the European Reectors (September, 1964)
'3hl Element Fuel Enrichment, Present laittel U I Itlet Depletion, Re.ctor me, p.ncle. tor
_Trt*_ *'.s Lec. tion 0.n.. . nom 3 235.. U .t.s R -erbs WR LC-Ol e ERCA UAI, 19.8 Storage 2.1 330 73 Irred. completed 4/11/83 WR LC-02 GRCA uAI, 19.8 Storage 2.8 329 48 Irred, completed 10/4/02 WR LHl4 NUKEm U3 0, 19.6 Storage 2.1 328 74 trred. completed 2/14/83 eWR t>02 NuKEM U3 0, 19.6 Storage 2.1 328 45 Irred. completed 6/7/s2 SILOE SAHUO0l ERCA UAI, 44.8 e 2.2 420 50 Irred completed II/81 SILOE b CERCA U 58 19.8 Pool c c >SS trred. completed 11/03 3 SILOE SDJZ001 GACA U 58 19.4 Core 4.0 507 50 Irred, began 10/84 3 SILDE b GRCA U351g 19.0 Core 4 4 60 trred. began 7/84 SILOE SSJA001 GRCA U 51 19.8 e 5.2 4 30 - 3 3 R2 85004 Mw U 51 3 2 19.8 Pool 4.8 444 0 In core 6/84 for fluu mees. R2 RW-005 MW U3512 19.0 t 4.8 90 - R2 i CERCA U351 , 19.4 i 4.8 M90 -
- R2 t IsuuEM U 51 19.8 i . 4.8 4 90 -
3 7 OSIRIS g CERCA U3 58, 19.8 g 4.7 @ -
- PIE cogleted at G55eclays nom at CE*Grenoble emelting shlpeent for reprocesslag.
*four plates la special Irradletion element.
" Plate loadings are S.S and 4.0 Hg U/e3 (20.2 and 22.8 g 2350 ).
Plate Boedings are 2.0. 3.6. 5.2, and S.3 Its u/m3 (7.S, 13.2, 10.7, 19.1 g 23'ul.
'At SILOE; Irroeletion scheduled to begIn late In 1905
'One elements tetricatIon scheduled during 1985 9
0ne element; tabrication scheduled during late 196S and early 1906.
F_.
\D .-
10 REFERENCES
- 1. J. L. Snelgrove, "RERTR Program Fuel Testing and Demonstration
- An Update," Proceedings of the 1984 International Meeting on Reduced Enrichment for Research and Test Reactors, Argonne, Illinois, October 15-18, 1984, Argonne National Laboratory Report ANL/RERTR/TM-6 (July 1985).
- 2. G. L. Copeland and J. L. Snelgrove, " Examination of Irradiated High-U-Loaded 38 U 0 -Al Fuel Plates," Proceedings of the International Meeting on Research and Test Reactor Core Con-versions from HEU to LEU Fuels, Argonne, Illinois, November 8-10, 1982, Argonne National Laboratory Report ANL RERTR/TM-4 (September 1983).
- 3. G. L. Copeland, G. L. Hofman, and J. L. Snelgrove, " Irradiation Performance of Low-Enriched Uranium Fuel Elements," Proceedings of the 1984 International Meeting on Reduced Enrichment for Research and Test Reactors, Argonne, Illinois, October 15-18, 1984, ANL/RERTR/TM-6 (July 1985).
- 4. F. Merchie, et al., " Irradiation of Low Enrichment Fuels in the SILOE Reactor," these proceedings.
so 1 e a O
Parv A4. I' - LICENSING CONSIDERATIONS IN CONVERTING NRC-LICENSED NON-POWER REACTORS FROM HIGH-ENRICHED TO LOW-ENRICHED URANIUM FUELS Robert E. Carter U.S. Nuclear Regulatory Commission . A8STRACT Ouring the mid-1970s, there was increasing concern with the possibility that highly enriched uranium (HEU), widely used in non power reactors around the world, might be diverted from its intended peaceful uses. In 1982 the U.S. Nuclear Regulatory Commission (NRC) issued a policy statement that was intended to conform with the perceived international thinking, and that addressed the two relevant areas in which NRC has statutory responsibility, namely, export of special nuclear materials for non-USA non-power reactors, and the licensing of USA-based non power reactors not owned by the Federal government. To further address the second area, NRC issued a proposed rule for public com-ment that would require all NRC-licensed non power reactors using HEU to convert to low enriched uranium (LEU) fuel,
' unless they could demonstrate a unique purpose. Currently the NRC staff is revising the proposed rule. An underlying principle guiding the staff is that as long as a change in enrichment does not lead to safety-related reactor modi-fications, and does not involve an unreviewed safety ques-tion, the licensee could convert the core without prior NRC approval. At the time of writing this paper, a requ-latory method of achieving this principle has not been finalized.
, = . .,
INTRODUCTION The principal purpose of this paper is to summarize current i thoughts of the staff of the Nuclear Regulatory Commission (NRC) about licensing considerations in connection with the replacement of ' high enriched uranium (HEU) with low enriched uranium (LEU) fuel in non-power reactors in the U.S. When it was proposed several months ago to write this paper, the NRC staff was still working on the on-going rule-making procedure, and it was intended to present and ex-plain the final approved outcome. However, because this is a com-plex issue, some questions have not yet been resolved. I It must be made clear though, that the NRC Commissioners have established a policy that the agency will take steps to encourage the owners of non-power reactors subject to its jurisdiction to con-
! vert their reactors from the use of HEU fuel. This policy was first ;
i announced in 19821, was amplified in the proposed rule in 19842, and 4 has been reaffirmed by Commissioner Asselstine at this conference,. j as well as at the 1984 Argonne meeting of the RERTR program. In the United States, as elsewhere, nuclear reactors may be grouped into one of two broad classes, power reactors, that are used for the commercial production of electricity, and non power reac-tors, that are useful for research, development, education, and i
' testing. All of the first class are licensed and regulated by NRC, but only those non-power reactors not owned and operated by the Department of Energy (DOE) or Department of Defense are licensed and regulated by NRC. Thus, the following remarks apply only to the ,
NRC-licensed nonpower reactors in the United States. 8efore moving on to more detailed discussions, it is important ' to mention a broad admonition and constraint imposed by law in the j US on the nuclear regulatory process. The Atomic Energy Act, under 4 which both DOE and NRC function, encourages and supports nuclear
! research and development, and therefore asserts that reactors useful
- for these purposes are to be subject to the minimum regulation nec-
, essary to protect the health and safety of the public. NRC must comply with that law.
i , BACKGROUND i In order to help clarify the reactor licensing situation in the i U$ I will summarize the history and add a few words about the U$ 4 regulations. The US Atomic Energy Act of 1954 provided the legal l 4 basis for the US Nuclear energy programs. A single agency, the j Atomic Energy Commission (AEC), was established to carry out the i law. In accordance with the law the AEC developed procedures and regulations for licensing reactors that it did not own, but also built and operated many reactors that it did own but which were not
-{ subjected to those regulations. The licensed reactors consisted of
! both commercial power reactors and non power reactors dedicated to 2 ?
research, test, and education. In general, both classes of licensed reactors are subject to the same regulations, but because of the recognized larger potential radiological impact on the public, addf-tional more prescriptive requirements have evolved for power reac-tors. After about twenty years of experience, Congress changed the law, establishing the current Nuclear Regulatory Commission (NRC) whose function is to lic'ense and regulate nuclear facilities, and assigning *.he other functions of the AEC to a different agency that is now the Department of Energy (00E). 'his department continues to own and operate nuclear facilities that are not subject to NRC's regulations, and continues with a mission to help promote nuclear , science and technology, including some financial and fuel's assis-tance to university reactors. 00E, of course, has supported the RERTR program. As we may remember, the first several nuclear reactors were operated with natural or low enriched uranium fuels. As high en-
, riched uranium became more readily available, it was used in some reactors to improve or optimize certain characteristics. Soon, even though probably not clearly required by performance considerations, HEU became the fuel of choice of both the US Government and reactor operators, and was used in most non power reactors except those of the TRIGA type. Even for those, however, an improved version of the zirconium hydride fuel was based on HEU.
While research reactors were gaining wide international accep-tance and use, concern continued to grow that the increasing quanti-ties of HEU used for non power reactor fuel were increasing the potential risk of diversion to an undesirable level. NRC ROLE As noted earlier, NRC responded to the increasing concern about diversion of HEU fuel by issuing a policy statement in 19821 Among other things, this policy included an intent to encourage NRC 11-consees to convert their reactors from the use of HEU fuel. In a second step, in 1984, NRC published a proposed rule for public com-ment that would require all licensees to cease using HEU fuel unless they could qualify for a defined unique purpose exemptions . A large fraction of the affected licensees expressed their concerns that conversion would require license amendments with a potential for public intervention that would involve unbearable litigation effort and costs, and even in the absence of litigation, conversion would require the expenditure of scarce money and manpower resources. , The proposed rule acknowledged these two potential costs, and expressed concern about them, but did not include explicit provi-sions for coping with them. Both because a guiding principle of the reduced enrichment program was to avoid any significant negative impact on reactor performance, and because of the provision of the j Atomic Energy Act noted above, the Commission directed the staff to 1
- 3 l
r we .
l . I' I i i revise the proposed rule to make conversion dependent on assured funding by the federal government, and on the avoidance of public l intervention in conversion licensing actions not accompanied by significant additional reactor changes. The staff has drafted a re-l vised version of the rule that does require that federal funding for l all conversion-related effort be assured before conversion is ini-tiated. The question of public participation has not been resolved at the time of writing this paper, but is a continuing concern. It is evident that NRC should avoid inadvertently causing excessive financial burdens on non power reactor licensees in its efforts to reddress the results of previous government policy. OBJECTIVES AND APPROACHES In order to issue a reactor license, the NRC staff must pro- '. perly conclude, for each individual reactor, that there is reason- ! able assurance of no significant hazard to the public. However, - many of the technical considerations can be generic. The staff is approaching conversion from HEU to LEU in that way, though not all steps are resolved at this time. The example discussest below of a ! generic environmental assessment currently in use suggests the fea-sibility of generic reviews for fuel conversion also. Several years ago, with reaffirmation about.five years ago, the staff developed a generic environmental assessment for the licensing of non power reactors that concluded that for such reactors licensed f to operate at 2 MW and below, there was reasonable assurance of no significant impact on the environment or the public resulting from continued operation. This conclusion was based on a large number of detailed individual reactor safety reviews and on extensive operat-ing histories of such reactors. In recent years the staff has re-l i newed nearly 30 reactor licenses, and has relied on this generic environmental assessment, in conjunction with a review of an indi-vidual environmental report in each case to determine if there are l any special features that render the generic assessment inappli- ! cable. No exceptions were found. l ! r ! Similarly, the staff's preliminary consideration of the prob-able changes in reactor parameters and operating characteristics i- resulting from conversion from HEU to LEU has revealed no new haz-l ards considerations. Therefore, we are considering grouping the , 1 i ' reactors into categories, such as not more than 2 MW power level, and reviewing within that group the potential hazards to the en-vironment and the public due to the change of enrichment of the fuel. Table 1 gives a tentativt grouping of all NRC-licensed non-power reactors using HEU. t l As in the example of the generic environmental assessment dis-l cussed above, the staff would have to conclude for each reactor that
- there were no special featurst that rendered the generic evaluation
! inapplicable. The following are two ways in which that could be ! l l 4 l l
1 .- . Table 1 i CATEGORIES OF HEU FUELED ) NON-POWER REACTORS ! LICENSED BY USNRC , i
- l. High Power Density i 2 - University l
- 11. Unique Fuel Configuration ;
3 - University 2 - Commercial ! 111. High Power (>5 MW) ] 1 - University
- 1 - Government 1 - Commercial I '
IV. Medium and Low Power ( 5.2 MW) Plate Cores 13 - University i
~
V.TRIGA FLIP Fuel ! 4 - University 1 - Commercial i I _ . _ _ _ _ . _ . _ . _ _ _ .__ . ._ ___ _ _ ___ __. _
done: (1) each licensee would submit an application and an analysis for his proposed conversion in the usual way that other amendments are applied for, or (2) the licensees as a group would work with RERTR to assemble pertinent information, data, and analyses in a broad-scope document, and each licensee would submit a copy of that document, along with a comparison between his specific reactor para-meters and those embodied in the generic analyses. The NRC staff will continue working with the Argonne group and licensees to ex-plore fis:xible);eneric approaches to the review and evaluation of conversions. DISCUSSICN In the previous section, it is implied that a license amendment will be required. In most cases, that is probably true, unless the conversion would meet the current criteria of 10 CFR 50.59 of the NRC regulations, or some other mechanism can be found for relief from an amendment. Part 50.59(a), reproduced below, allows certain types of changes to a licensed reactor without prior NRC approval. If appro-val is required, in most cases it is accomplished by means of a 11-cense amendment.
$ 50.59 Chango , tests and experiments. ;
(a)(1) The holder of a license authorizing operation of a production or utilization facility may (i) make changes in the facility as described in the safety analysis report, (ii) make changes in the procedures as described in the safety analysis re-port, and (iii) conduct tests or experiments not described in the safety analysis report, without prior Commission approval, unless the proposed change, test or experiment involves a change in the technical specifications incorporated in the license or an unre-viewed safety question. (2) A proposed change, test, or experiment shall be deemed to involve an unreviewed safety question (i) if the probability of oc-currence or the consequences of an accident or malfunction of equip-ment important to safety previously evaluated in the safety analysis report may be increased; or (11) if a possibility for an accident or mm1 function of a different type than any evaluated previously in the safety analysis report may be created; or (iii) if the margin of safety as defined in the basis for any technical specification is reduced. In the previous section, nothing was said about out 11ers, that is, those reactors that wouldn't qualify for inclusion in a category. Presently, the only alternative suggested is a case-by-case submit ' tal by a Itcensee, rev4ew by NRC, and issuance of a license amend-ment by NRC. However, unless significant reactor modifications not required by the change in enrichment were made, the staff anticia pates that the licensee submittal need only address parameters and 6
i '.P. I I l : l ! characteristics of the reactor altered by the fuel change. Figure 1 ! is a schematic flow diagram of the approach just discussed. - l l i In the above approach, it has not been indicated that the path of minimal effort on the part of the licensee will lie with no change in the physical dimensions and hydraulics characteristics of the fuel elements upon conversion. Nor has the NRC staff suggested that the LEU fuel design must be optimized to achieve the maximum possible margins between safety limits and operating conditions in , each case. Instead, NRC has a long-standing policy to standardize . reactors when feasible. Therefore, we urge the Itcensees and 00E to j propose standardized fuel parameters that will satisfy licensees' ( performance requirements and still provide bases for NRC to conclude : l logically that changing the fuel enrichment would cause no signift-1 l cant impact on the health safety of the public I CONCLU510N L At the time of writing this paper, the NRC intends to issue 4 i revised rule requiring all Itcensees to convert from HEU to LEU un- ! less they qualify for a unique purpose exemption. The revised rule is espected to have a provision making conversion dependent on as-1 sured federal funding. The revised rule also is espected to include l provisions for avoiding pubite involvement on issues related closely L to conversion from HEU to LEU. The schedule for issuance of the revised rule is currently uncertain. , l t REFERENCES i r
, t I
i tune of High Enriched Uranium (HEU) in Research Reactors Policy i Statement, August 24, 1982, Federal Register (47 FR 37007). ! 8 Limiting the use of Highly Enriched Uranium in Domestic Research ! and Test Reactors, July 6,1984. Federal Register (49 FR 27769). ( l
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e CORE MANAGEMENT AND REACTOR PHYSICS ASPECTS OF THE CONVERSICN OF THE NRU REACTOR TO LEU i M.D. Atfield Atomic Energy of Canada Limited Chalk River Nuclear Laboratories Chalk River, Ontario CANADA K0J IJO A3STRACT Results of work done to assess the effects of converting the NRU reactor to LEU are presented. The effects are small, and the ( operational rules and safety analysis, appropriate to the EEU core, will still apply. INTRODUCTION The NRU reactor at Chalk River was commissioned in 1937 as a large sultipurpose facility demonstrating heavy water technology , and on-line refuelling. Tor six years it ran on natural uranium metal fuel at 220 MW. To11owing conversion to EEU in 1964, it started operation in its present role as a fully-fledged research reactor, used for developing CANDU fuels and saterials technology, the production of isotopes, and fundamental physics investi-gations. In 1973 the aluminum calandria was replaced. The most recent major development for NRU comes from the US initiative tewards the use of LEU, for antiproliferation reasons. The reactor itself consists of a vertical cylindrical tank of D2 0, 3.5 m in diameter and J.7 m high, into which a maximum of 227 rod-like assemblies may be suspended in a hexagonal lattice of pitch about 20 ca. Of these 227 sites, about 95 at present are EEU driver rods, which are taken to about 40% burnup. The reactor is very heavily utilised, with only a few vacant rod sites around the periphery, so the concern in going to LEU was largely about the 9 5 9 II
I l 2 effect on core space and associated research art development pro-l grams. Fortunately, this effect has turned out to be reasonably j small. 1 The current KEU (93 enriched) driver fuel contains 1.8 g l U-235/cm in 12 sheathed pencil elements of uranius-aluminum l alloy. Since AECL has extensive experience in the production, handling, and analysis of these rods, the basic premise adopted I ' to keep the conversion to LEU simple was that the driver rod geo-setry should stay the same, and only the fuel composition would i change. Considerable research and testing by yuel Materials Branch at CRNL has resulted in the selection of U3 Si as the l mest suitable fuel material (l) for the operating conditions. l , The major change from the neutronics standpoint, then, is the l addition of a considerable quantity of U-238 to the driver l elements, which operate in a well-thermalized spectrum. METHOD l l The change in driver rod composition does not lead to any changes in the operating philosophy of the reactor. The reactor will still be operated to provide required flux levels for loops, l isotope production and neutron physics experiments. The loops, ;
- in particular, need a nominal flux of 3 x 101' n/cm2/s axial '
l peak in the moderator. The LEU core flux radial shape should 1 therefore be the same, as far as possible, as the EEU. l. i In practice, it is not possible to achieve this outcome exactly. The central core area in NRU is full, and central non-driver sites cannot reasonably be converted to driver sites. Thus any increase in iriver sites, to compensate for the extra U-238 load, must occur on the outside of the core. Even there, few sites are available. Nor can the average fuel content of central driver sites be increased, since individual central ' driver powers sometimes approach their limits, and a margin of flexibility must be maintained to ensure ef ficient operation.
' The addition of extra fuel to the core periphery will therefore be necessary, and will flatten the flux shape somewhat.
The method used in this assessment was to choose a typical REU core loading, replace REU driver rods with LEU rods, and then to add peripheral rods and adjust burnup in order to achieve i close to the same k,gg and flux shape as before. The re- . sulting LEU loading was used to assess other core parameters. ' i The comparison of REU and LEU core parameters is thus based i on one core loading only. l Nevertheless, these results give a l good indication of the sajor effects of going to LEU driver fuel. t Such things as the sensitivity of everage core properties to i normal load and flux anape changes (normal regional flux changes Ray be
- 13I) were not addressed.
i l 1 i L i !
_ ~ _ - - _. _ . - . _- - - - - . _ _ - - _ _ _ _ _ .
.' ~
3 l RESULTS Core Management The sajor effect of substituting LEU driver fuel for RE" is the addition of about kg of neutron-absorbing U-238 to every driver site. This extra load results in a reactivity loss af about 13.3 ak. Extra fissionable fuel (U-235) La theref ore required to support this load, and also extra power to provide neutrons for absorption in the U-238 and still leave the soderator flux at loop sites unchanged. These requirementa could be set by staply reducing driver rod burnup at every site, while leaving the core size the same. Since there are a few empty i , peripheral sites available, however, it is more ef ficient in I cerus of total rod usage to use some of these, so that exit
- buruup need not be reduced so much.
i The fuel managtment practice in NRU is to put fresh fuel near the outside of the core, move fuel in towards the centre, and l finally back to the outside. Each rod occupies an average of 5 core positions during its lifetime. This scheme will not be l affected by the change to LIU. 3 The initial HEU core loading is given in Fig.1 with an i indication of the extra sites used for the LIU loading. Five l extra peripheral rods were added to form the LEU core (3 high, 2 tow burnup), and burnups were altered on 16 other driver rods to reduce the exit buruup to 300 mwd (f rom 313 MVd), and to make the flux shape and k,gt similar to the REU case. Under normal operating conditions with an REU core, there can be considerable variation in the global flux shape. In comparison, the dif f erence in calculated flux shapes between the REU and LEU cores was l negligible. This also means that there will be insignificant flux change at the control and safety ion chambers located on she core boundary. i The calculated power " distributions were equivalent within the limits of variability seen under normal operating conditions and so normal fuel rod power limits will still provide adequate flexibility. If reactor power is derived by equating total loop operating powers, the LEU reactor power is about 4% higher. This is operationally insignificent during normal operation (except in , I economic terms), but does restrict flexibility in seating requests for extreme conditions, e.g. some loop power boosts. , One potential restriction follows from the fact that VRU ! approaches its 133 MW total power limit when providing nominal l flux conditions for four fuelled loops in the present eore loading, with HEU driver rods taken ts 313 MVd. This condition I l I
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(- has not occurred often over the past several years, and can be sanaged briefly without resorting to removing old fuel rods (to reduce total power), thus raducing exit bu rnup . With the , increased reactor power resulting f rom LEU 3 river rods, however. this . limiting condition would be approached sore of ten. It is not clear at present whether this ef fect would require a further reduction in LZU exit burnup over that already given. since it will depend on loop operation, and loop arrangement will be changing in the near future. No allowance was nade, the re f o re . for this effect. Some general' fuel management inf ormation is given in Table
- 1. The choice of an exit burnup fir the 12U drivet rods is some-what arbitrary. It could be increased or decreased slightly by adding or removing very old rods to or from the core periphery, but the 300 mwd value is a best guess based on the current core '
loading and flexibility requirements. , Reactivity Values
, There are 18 control rods in NRU. C?s 9-18 are actually used for operational control, while 1-4, 3-8 are designated as safety banks, used for shut-down reactivity control.
Control rod reactivity worths are given in Table 2. A slight overall reduction in values in the LEU core is observed, as expected, but this poses no probles with regard to reactivity limits. Safety banks are worth greater than tre design value of 30 sk. TMM 1 . COHPARISON OF HEU and LZU FUELLZD CORES LEU LEU % CHANGE Reactor Power (same loop flux) 132 MV 137.3 +4 . 0 Driver rode 90 15 +5.6 Exit burnup 313 MVd 300 -4.8 Driver power 114.3 MW 119.7 +4.7 Average driver power 1.27 MW 1.26 -0.8 Max. driver power (limit 2.4 MW) 2.21 MW 2.0$ --- Residence Time (70% efficiency) 354 4 340 4 -4.0 Av. core positions / rod 3 5 0 Driver rods /yr (70% efficiency) 92.7 102.0 +10.0
6 TABLE 2 REACTIVITY WORTH COMPARISON FOR CONTROL AND ADJUSTER RODS (calibracion done in sequence) , CRf HEU ak LEU CRf HEU ak LEU 1 28.9 29.0 10 6.1 5.9 2 13.0 10.9 11 8.3 9.4 3 5.0 4.2 li 10.1 8.8 4 6.2 6.7 13
- 9.2 9.1 3 14.8 10.4 14 4.9 4.0 6 11.4 12.5 15 4.9 5.1 7 14.2 16.4 16 7.9 7.8 8 6.3 5.1 17 3.9 3.7 9 15.7 13.1 18 12.8 11.2 TOTAL 183.6 173.3 arf arf 1 6.7 61 3 13.4 11.2 2 9.8 10.7 4 5.9 5.9 (adjuster rod worths calculated with CRs 1 to 16 up)
Reactivity worths of reactor assemblies are compared in Table
- 3. La expected once again, the values in the LEU core are gene. rally slightly smaller; insignificant exceptions occur due to the 1.nperfect reproduction of the HEU flax shape by the LEU loading. The general similarity of reactivity values, however, shows that no changes in operating philosophy will be required in converting to LEU.
. Equilibrium reactor poison worth will hardly change due to the 4% increase in reactor power, but will be reduced slightly due to the estra U-238, in approximately the saar proportion as control rod worths. The reactivity scale for NRU is thus not af fected.
The shatdown peak poison worth will increase in step with the reactoe power relative to the control rod s, thus decreasing time-to poison,slightly (f rom 65 min to 60 min), and increasing poison-out-time by about an hour (f rom 33 h to 34 h). Reactivity Rates The reactivity values for fuel and absorber rods in the LEU core are similar to current values to the HEU core, and so the same reactivity limits will apply, as will the established analysis of the eff ects of reactivity rates achievable under certain postulated accident conditions. l l
7 TABLE 3 REACTIVMT WORTH COMPARISON Site Description HEU ak LEU R18 Driver (2 mwd) to 0,0 -3.35 -3.91 M09 Driver (114 mwd) to*D,0 -7.15 -5.85 L10 Driver (2C7 mwd) to D;0 -5.43 -4.42 F13 Driver (296 mwd) to D*,0 -2.47 -2.29 H13 Driver (195 mwd)toDjo -6.29 -5.38 H13 D,0 to driver (fresh) +12.69 +11.00 H17+ Dliver (220 mwd) to D 0 -4.21 -4.86 H17+ D2 0 to driver (fresh)2 +12.08 +11.13 H17+ corresponding total rod power 3.30 MW 3.11 MW 013 Mk IV Fast Neutron Rod to D,0
+1.96 +1.92 K21 C-14 Rod to D,0 +8.08 +8.61 D09 C-14 Rod to D;0 +8.88 +4.28 H23 Co-60 Rod to 0,0 +6.31 +6.34 Kil Mo-99 Rod to D;0 +0.31 +0.26 H 1' Mo-99 Rod to D'O +0.20 +0.20 E2) Loop fuel to D,0 +3.02 +2.85 E23 Loop fuel void 3d +2.59 +2.39 017 Loop fuel to D 20 +7.85 +8.23 067 Loop fuel voided +3.51 +3.57 E20,017 Loop fuels (nat.U) voided together +5.80 +5.67
+ highest core flux position.
Reactivity Coefficients In order to properly determine the ef fect of LEU driver fuel on core moderator temperature and power coefficients, it is , necessary to rerun the whole core model under dif ferent sets of : temperature conditions. Due to the large number of assemblies in NRU, this procedure would be very time-consuming. Theref ore a simplified core was set up, which should nevertheless enable good estimates of the reactivity coef ficients to be made. This simpli-fled core contained driver rods, using a current real core burnup distribution, strong absorbers (cobalt carrier rods) and D 20 sites. Other core assemblies were replaced by these three assemblies in accordance with their effect on the core in terms of neutron production or absorption, and in order to bring k,gg l close to 1.0. l l i d
! 8 TABLE 4 L ( Temperature Reactivity Coefficients (Simplified core) f Temperatures ('C) KEU LEU Fuel Coolant Moderator v v v = 1/k,ff 41 41 41 0.97569 0.98803 81 81 41 0.97866 0.99093 201 81 41 0.97900 0.99133 201 41 41 0.97615 0.98858 , 81 81 81 0.98706 0.99896 i i Fuel temp. coefficione ' HEUs (0.97569-0.97615)/160 = -0.0028 sk/'C
- l LEU (0.98803-0.98858)/160 = -0.0034 Coolant temp. coefficient: EEUs (0.97615-0.97900)/40 = -0.071 "
LEU (0.98858-0.99133)/40 = -0.069 l Moderator temp. coefficient: EEUs (0.97569-0.98706)/40 = -0.28 " LEU (0.98803-0.99896)/40 = -0.27 l l l Table 4 shows the range of temperatures used in the models of I these assemblies, together with the eigenvalues derived f rom their use in a 3D diffusion code simulation. A fine convergence cri-terion (10-5 on eigenvalue) was used since the reactivity i I dif ferences were small. ' l The moderator temperature coefficient is defined as the 1 change in core reactivity per unit change in the heavy water inlet '. temperature. This was derived by raising the simplified core from .' a uniform temperature of 41'C to 81*C. The EEU core coef ficient was -0.284 sk/*C, while the LEU core coefficient was -0.273 sk/*C. This chante in coefficient will have negligible operational' l significance. l The power coefficient is defined as the change in core react- , ivity per unit reactor power increase, while holding the heavy water inlet temperature constant. This would be very dif ficult to calculate, but since most of the reactivity ef fect would come f rom changes in fuel and coolant temperatures, the coef ficient may be examined by changing these temperatures only, as shown in Table 4. ; in this way, the calculated fuel temperature coef ficients were KIU, .0029 sk/'C; LEU, .0034 sk/'Cl while the calculated coolant ; temperature coefficients were: HEU, .071 sk/*C; LEU, .069 sk/ *C. . I It can be seen that these coef ficients are small, so that the total power' coefficient will be saali, and that the REU and LEU values are similar, so the change will assin have negligible operational significance.
9 The moderator purity coefficient was derived by reproducing cell parameters for the three assemblies at a Dg0 purity of 99.75%, instead of 99.85%. Over this range, at 41 C, the HEU coef ficient was 8.08 sk/0.1% D 2 0 purity, while for LEU it was 8.00 mk/0.1%, D2 0 purity. This change is again not significant. (Previously ; calculated values (1964) are: KEU: 10 mk/0.1%, and natural uranium cores 2.4 mk/0.1%). I Kinetics Parameters t '- Proept Neutron Lifetime. The slowing-down tip in NRU is close to that in the D20 moderator i.e., 4.6 x 10"Js. The average dif fusion time of thermal neutrons in the core may be l roughly estimated from the neutron balance by calculating equivalent homogeneous cell parameters:- .
, 1, = =pf- (7 = 2.2 x 103 cm/s) a and including leakage, g, 1 ett V(* a+D8 ) VOY f
y - also total core neutron production rate = V*g 1 2 Using the present simulations of an KEU loading and its equivalent rebalanced LEU loading, we derive 1(HEU) = 1.78 x 10-3s, l
. . and 1(LEU) = 1.69 x 10-3 s, giving total prompt neutron lifetimes of 1, (EEU) = 1.83 x 10"3 s, t, (LEU) = 1.74 x 10*3 s.
i This small change will not produce any significant ef fects l on reactor response characteristics. > l Delayed Neutrons. Table $ shows that, at the LEU driver rod l exit burnup, at least 91 8% of fissio's n still come from U-233, while 7.0 are due to Pu-239, and 0.9% to Pu-241. Since the average driver burnup in the core is less than 60% of the exit i burnup, the average importance of the Pu fissions will be reduced. , ! Examination of the table indicates that an average relative Pu fission rate of 3.0 vill be about right f or NRU. This es411 l l percentage will not change the dynamic response of the reactor very much, and previous dynamic analysis of NRU will still be applicable. I i i 4 6
5 l ., ..o., f l e ( ' i i 10 l> TA3.E 5 f I Fission Rates per Material for 1,IU Driver Rods ( 13 11 12 12 I x 10 x 10 x 10 x 10 t n/kb %'d /es U-235 U-238 Pu-239 Pu-241 I
* ('
0 O 37.99 7.03 0 0 0 0.2
- 1. >
(99 8 0 0)! } O'.71 .L9 6 4.3 3.2 0.27 l l , 1.0 (97 1 0.2 26 0 1)! l 2.4 1.190* 7.93 2.27 6 10 0.80 ' i 2.4 (91 4 0.3 7.0 0.9): approximate value at centre of 300 mwd (exit burnup) rod
+ estimated values for core-average burnup.
]
i , Delayed Photoneutrons. Since the Pu fissions are only about ! f 33 on aversee, the yields of fission products producing sammes ! I t above the photoneutron threshold of 2.226 MeV in D2 0 will be l similar in the MEU and 1.EU cases, and may be taken se coming fros [ l U-235 for most purposes. Thus, as with delayed neutrens, i l previously used data for delayed photoneutrons in NRU may still be ! l applied. : f : l ACCOUlff1NG l l The present NRU fuel inventory accounting procedure does not I track fuel isotopes other than U-235 and U-236 on a daily beats. l The plutonius build-up in 1.EU fuel will require the inclusion of !' other isotopes, for esse of reprocessins/ burial accountancy. This ( , modest expansion can be readily achieved using data f rom current cell codes (VIMS). [ CONCLUSIONS l The effects of replacing the present NRU driver fuel (MEU, 93 l wt! enriched uranius, U/A1 alloy) by 12U fuel (20 wt3 enriched l l uranius, s111 cide dispersion fuel) were assessed f rom the reactor ; l . physics viewpoin6 Overall the effects are small, with perhape l the largest be Ang an increase of about 101 in fuel consumption. ' No significae.t operational dif ficulties were identified. l R2yERENCES
- 1. The Fabrication and Performance of Canadian Silick~de Dispersion Fuels for Test 11eactors, D.F. Sears, et al., Paper
- to be presented at RIRTR Conference, Petten,1983.
l l i
0 MW d. I 1 J THE CCNVERSICN OF THE DIDO-TYPE PEACTCP FPJ-2 STUDIES AfiD CCNCLU$!ONS
,A. Strcmich, Ch. Sieterts. M. Wickert Interstem CmcH Bergisen Gladbach, Cernsny ADSTPACT For tne FPJ-2 (23 ffd) of the KTA-Julich the con-version frem HEU. to LLU-fuel was investigated. De-fore starting tne conversion calculations our metnods were qualified for the application to heavy water me.
derated researen reactors. A combination of LEU-ele-ments with two differ ent U-235 loadings of 100 g and 225 g was found as suitaele for conversion. Witn these LEU-elements a worwing core and a transition phase was calculated. The change of the mechanical fuel element design was tagen into account. INTPCDUCTICf! This contribution deals with the investigations for the conversion fecm HEU to LEU-fuel of the 23 MW-research reactor FPJ-2 operated at tne KTA JU1Leh. There are two items wnien in-fluenced the studies essentially, a methodical one and a techni-cal one
- As the DICO-type reactor FRJ-2 is the only heavy water modera.
ted research reactor in Germany our calculational methods nad to be qualified for this purpose.
- When changing over fecm HEU to LEU the mechanical design of the fuel element has to te modified. As a censeq'annce of this tn9re arise additional steps within the converaton procelure.
. t I
2 , BENCHMARK CALOULATIONS l j Before starting special conversion calculations we had to adapt and to qualify our standard metnods /1/ for tne application to heavy water moderated research reactors. One way of cotaining this qualification wa. the calculation of the IAEA.Denchmark pro-l blem for D2 0 research reactors /2/. A very good agreement with the , results of ANL was acnieved and is displayed exemplarily for the infinite multiplication factor in fig. 1 and for flux distribu. l - tiens in fig. 2. Ine results strongly support the reliability of the methods in performing conversion studies on D 0 researen reactors. Iney l are described nere in a snort sammary
. Tne fuel element group constants in dependence en burn-up and several other parameters are calculated by the pregram MONSTRA.
l It was modified for the application to D,0-reactors witn respect to n.2n-reactions in D o calculation of' decay products and increase of the energy group nu.gete,r. The reflector group constants are ge-nerated by IAN5N, an Interatom-version of the transport code ANION. With these group constants as input data the xy. core calculations are performed in four energy groups Dy tne program IAMADY based I en the diffusion tneory. r
l MIIIM - Fig. 1 K. cf benchmark calculations se 7 ,
. e . 4%
= = ttff84f0M 10 '
11 1
\'
%gg ,
N y q, . ,,s u, ,, .. IM W> $ AM* l gi o e L
i o '.
- 3
, g. n i. is % @WNU
~ ~'-~ y \
'2
}\ s, y,,, .. . as
- m rrearm
( \
' \
l rwen. g s x N f*f"U"-
\s ccer \
oas \i- M am ?t-s
. '. m -
..,.< , & .. ~ e a ,. ., -
- F.isj. 2 Flux distribution of tencrearx calculaticns SHORT DESCRIPTICN OF THE FRJ-2 The FRJ-2 is a DIOC-type neavy water modersted closed tanx l reactor operating at a nominal power of 23 MW. Its core is cenposed
- of 25 fuel elements. A large numter of norizontal team tuees and vertical facilities extend into tne heavy water reflector. Six l
" signal arm" absorters centrol the resctor. The arrsegement of cere, reflector and experimental facilities is shewn in fig. 3a.
The tubular fuel elements (cf. fig. 5,CD) consist of five cqncen-tric tuees, four inner ones of wnien contain the fuel, the outer-most beren as a burnacle poisen. The core is composed of fuel e'.ements with two different U-235 loadings of 150 g and 170 g witn an enrichment of 80 %. CALCULATICN CF THE ACTUAL CORE STATUS After having checked our metnods at the idealized core cen. figuration and boundary conditions of a tenermark problem we applied them to the actual HEU-core of the FRJ-2. This was done in order . 4 i - to obtain a furtner qualification under real conditions for a real core and reflector configuestion and
1 f 4 mJ e
. Mr_ Lran .
jy V .h..j , .
, ..;.'a v y !. T, - 1 g. .. ;,. .
.x s.: ,* .
r
,...m. ...
/... .
, , s .l: L .
1 r, . .\ .[ . , , n..y /../.;.; 3....:;:f: N:' 7 @*/g o 4.r. i , ,....rn
,- e?' J..:p.:.g
- s. ._
e
.a . :
r t ;A y ..g s.
^
[;'= ,
;g ,, , ,
a
>p.e ~ ,
. , t s '. . ..
s, e . ,o ' N. ;{.y ... ,/ 7 . t.q.g{k ' ' 's ,'k'. ' .
- : -a
, Ul . NIL.l.if i i s
u . E g .
- 1 p+++=,w .
. . d ,
[
+. . . . .
] .i+4.. H.. .d . [ ' , b
.et4 7
, 1- -
. P e***?** .'s ".
\
h
)
I ! Fig. 3 a Crcss sectior. of FEJ-2 core and reflector b Calculational Ocdel l 4
- Y 5
i l i i I _ b i 80C ,.,, .,,, ,, ,, __ cyg, EOC ,,, t ,, ,, ,, 1 06 1 26 80 .86 -- W a l . I 10 1 27 39 91
; .90 1.11 1 23 1 06 82 83 to 1.16 -123 t ot 79 .m
.31 1.11 1 29 99 79 to .93 1.12 1.26 94 .70 .82 1 15 1 Of f 31 1 02 .90 1.16 106 1 31 .90 .87 1
1.21 1.01 1.35 1 07 95 1.3 .90 1. M t02 .96
- i. .93 .46 1 19 1.16 .3 72 .M If 1. 2 1.16 .73 .3 f ,
.4 .M 1.20 1.29 .76 .72 .79 76 1 25 1.24 .74 .77 LOS 109 .94 .75 1.11 f.13 .96 .75 t.07 1.te .9e .n 1.s tm .99 .77 i
t i Fig. 4 Power peaking factors for HEU-core 2/83 *
'i l.
- to assure that the used methods and models treat the special core and core-related components correctly.
4 The calculational model is given in fig. 3b. Calculations taking into account the absorbers were performed with the transport
- j. code DOT 4.2.
Seven states of currently used HEU-cores (four different
.! cycles) were treated. The calculated excess reactivities for
! these states agree with the values from measurements within i AS 51 1 %. The power peaking factors from calculations and mea-surements agree in general better than within 10 % as shown in
. fig. 4 DESIGN OF THE LEU FUEL ELEMENT For the conversion to LEU originally a fuel elegent with 200 g U-235, U,0,-Al fuel at a density of 2.9 g U/cm was chosen by the KFA-JulIc5 as a reasonable solution. Six elements of this 5'
type were fabricated by Nukem for irradiation tests. One of them e was irradiated and discharged at a burn up of 55 98 %. Experience during L.W. element fabrication has shown that the mechanical de-
' sign must be modified in order to avoid high rejection rates.
The new element design will in future provide for three bent 1 fuel plates being connected to form the fuel tube by rolling into 4 three aluminium webs provided with grooves. At present the three prebent plates are completed to form the fuel tube by electron beam welding. In fig. 5 the old design (CD) and the new design j (ND) are shown for comparison. It is not intended to change other i
c - 6 3 M M\ // g
/ /
I w
- - - . -}- - - j j j y , &lil; l
C: l 00 NO I 1 I
+ -
l Fig. 5 Sketch of old design (OD) and new design (ND) fuel [ element i i parameters, e.g. meat thickness and cladding thickness. Concerning the fuel material it was a clear outcome of the RERTR-meeting 1984 f at ANL that for technical and economic reasons the use of silicide
; fuel for LEU-operation is preferable.
Consequently the fuel element for conversion to LEU will be a ND-element with U Si -Al-material. 3 2 4 CONVERSION PROCEDURE In agreement with the licensing authorities the conversion is envisaged to be perfor=ed in two phases, af ter forerunning tests with 2 to 3 ND-elements with HEU-material (170 g U-235) as well as with LEU-U S1 -*"t*#1*1* 3 2
- Changing of the complete core from CD-HEU-elements to ND-HEU-
. elements.
- Conversion of the HEU-core with ND-elements to a LEU-core via a transition phase.
In this way the mechanical design change of tne fuel elements
, will be taken into account sufficiently.
HEU-CORE WITH NEW DESIGNED ELEMENTS With respect to tne first phase of conversion the BOC- and EOC states of the HEU-core 2/83 were recalculated with tne only change to ND-HEU-elements (same U-235 loading). This enange effec-
"ted a reduction of, excess reactivity of about A S = 1 % and only small deviations in flux and power distributions. Hence there are no serious problems tc be expected from the neutrcn pnysical point of view. The result of a thermal hydraulic investigation was that
e 7 [ctiu61FJ-1 68-C tM g I W% CD 50g 100% N0 i 158-NO mgIN%l NO 200g #3% j h0 2g t3%
\
l
. , a a .
W. SURwuP 1 g 2MyI Fig. 6 K ,3 for different fuel elements there are also no objections against the change frem CD- to ND-fuel elements with respect to flow instability. LEU-CORE WITH UNIFORM U-235 WADING OF FUEL ELEMINTS ror a LEU-fuel ele =ent of new design 'gith the originally en-visaged U-235 loading of 200 g (3.1 g U/cm ) cores with two dif-ferent burn up distributions were investigated (fig. 7). Charac-teristic features are
- (A) 15 % burn up step; about 27 fpd cycle lengths charging of 8 (9) fresh elements ! - (B) 12 % burn up step; about 22 fpd cycle lengths charging of 6 (7) fresh elements l
The calculated reactivity values are within the assumed
, limits which we derived from calculated HEU-core states
+
, - upper limit from BCC of core 2/83 (absorber angle 10.5 *)
- Icwer limit frcm ECC of core 4/84 (absorcer angle 40*,
completely withdrawn) The corresponding HEU- and LEU-reactivity values are simi-lar although there is a reducticn of 5 % to 10 % for the infinite f l t i
E f
@ H:it/M - @hM-e 15 15 0 a 36 24 12 , 0 a 12 12 24 0 0 60 3 29 9 16 0 00 8 o 15 m x m 0 0 15 0 15 E 26 0 24 % ?e 24 0 00 17 0 0.0 18 9 33 0 2e 6 15 30 m 30 15 0 30 15 0 30 12 M M M 12 0.0 35.2 13 4 00 3: 4 0 30 30 30 15 0 12 15 E O 15 h' o 25 w n x 0 15 0 %.6 31 4 0.0 45.5 3, 5 3 ,
0 15 15 0 26 36 0, 12 , O tacsi2cw 0 24 12 12 31 1 0.0 23.3 13.3 O 225, s 2c v. Fig. 7 Burn up distribution Fig. 6 Burn up distributica for 200 g-element with two U-235 leadings multiplication factor frc= HEU to LEU as snown in fig. 6. The co=pensation can be explained by the shift of tne flux spectrum and the resulting increase of moderation cutside the core in the LEU case. According to the distributler, with increasing burn up free outside to inside the core the power distribution is relatively flat. But it i= plies a snuffling of fuel ele =ents at BOC. The calculations proved the ?;D-LEU-ele =ent with 200 g U-235 loading as suitable for the conversion of tne FRJ-2. LEU-CORE WITH TWO DIFFERENT LOADINGS OF FUEL E.' EMENTS With the preceding results as a basis test calculations fer the ccmbinations 190 g/215 g and 180 g/225 g U-235 respectively were performed with re!pect te power distributiens. Tne aim was tc avoid the element snuffling at B00. Evaluating the results the (2.8cperateg) g U/cm determined tne combina;ien of the U-235 leadings 180 g and 225 g (3.5 g U/c=~) for further investigatiot:s. A core configuration witn an idealized burn up distribution is shown in fig. 6 with the characteristic features 15 %/12 % burn up step for 160 g/225 g
- 45 %/48 % discharge burn up accut 23 fdp cycle length
- charging cf 4 (5)/3 fresn elements
c .
. ..e.... :
t l 9 l
,ny ,o . ,, , @"*"*
i m Liu r
- J t
~ l
~ ~ l l l l
]
_______o___._.________=___a_.._,l l 3
, ,. > > . , i
,o .
l l . l i i , ' l 1 i
.. I l I I i
l 1 l l 1 I l l I o i I l l'
- ' 2-
; I I
i l -l--i l I i I [ l r [; n z tv v I vi l vis ! e l l , f vs , isoo eia moo uno l Fig. 9 Reactivity and power peaking factor (FR) in dependance on time As there is no fuel element shuffling this idealized burn up { pattern is repeated af ter 13 cycles. Such a chain was calculated
; with reasonable reactivity values within the assumed limits des-
; cribed in the preceding chapter.
4 i Starting frem ECC of HEU-core 2/83 (however with ND-elements) a transition phase from HEU to LEU was calculated exemplarily. ( Fig. 9 shows the reactivity as a function of time during the tran-sition and for some cycles after ecmplete conversion. Moreover the number of charged fresh LEU-elements is indicated. One can see that the conversion is completed at the beginning of the 4th cycl.e, i.e. after about 68 fpd. Up to the 7th cycle the complete chain was calculated without fuel element shuffling. At the be-ginning of the 8th cycle the elements were rearranged according
! to fig. 8,B obtaining a burn up pattern very similar to the ices-
{ lized one (fig. 8,A). Fig. 9 shows additionnally that the radial (element' averaged) power peaking factors (indicated by arrows) are in an acceptable range. The subcriticality for the shut down core during this chain of cycles was calculated to more than A3er-5.5 %, whereas a minimum value of a3 = -5 % is allowed. A (L> comparison of the thermal flux (E < 0.625 eV) of the HEU-core 2/83
' at BOC and the 8th core (LEU) of Ehe calculated chain shows in the LEU-case a reduction of about 15 % inside the core and about
{g 10 % in the D 2 0-reflector (fig.10). The decrease of the fast group
^
~.*..
10 ess FITNG2B - 6 C EE 0.0-
~ [ /%
t/
\
. .j S4'
.s .
.. ss-cam .
ogo coat ogo cum - ls-f a n.
.s ,
] 's e ies n.azr5f ta :3c Ine
.} Fig. 10 Distribution of thermal flux (E < 0.625 eV) u n fluxes is about 5 to 10 % in the core, i s-'
CONCLUSIONS For the conversion of the FRJ-2 a LEU-element combination f) with two different loadings of 180 g and 225 g U-235 wgs evaluated 1 as suitable. The corresponding densities of 2.8.g U/cm and 3.5 g
.1 U/cm# are in a well qualified density region of U Si -fuel. With
~ .) thiselementcombinationaLEU-corecanbeoperat$dbithoutele-
., 'l
! ment shuffling at BOC. For the transition phase from HEU to LEU core there are no serious problems to be expected. The first LEU-
'i.
- core will be reached af ter about 70 fpd. The thermal flux will de-crease about 15 % in the core and about 10 % in the reflector. The
.g f change of the mechanical design will effect additional steps in
' I the conversion procedure but no other serious nuclear or thermal hydraulic problems.
,' .I i
REFERENCES
~ 7, j _.
,{' 1'. Research Reactor Core Conversion from tne Use of Hignly En-
,I
? riched Uranium to the Use of Low Enriened Uranium Fuels, P, ,.
Guidebook, IAEA-TECDOC-233, 1960
- 2. Guidebook Addendum: Heavy Water Moderated Reactors, IAEA-TECDOC-326, 1985 1 --
t __m_W.
yH Q Y
. Study on Temperature Coefficients of MEU and EU Cores in the KUCA Keiji Kanda, Seiji Shiroya, Masa-aki Mort, Masatoshi Hayashi, and Toshika:u Shibata Research Reactor Institute, Kyoto University Kumatort-cho, Sennan-gun, Osaka 390-04, Japan ABS *2ACT 3 Recently, the measurements of the temperature 4 reactivity coefficients were performed in the IUCA in d succession to the study on the void reactivity effects.
t, ne objective cores of study were light-water-moderated and heavy-water-reflected ones loaded with EU or MEU fuel. De following effects on the temperature coeffi-
, cients were in7estigated for the range from 20*C through i 70*C: (1) the reduction in fuel enrichment, (2) the fuel
'; loading pattern, and (3) the existence of boron burnable poison. ne measured data were analyzed using the $2AC system to assess the computational technique for ths
, temperature effects on reactivity, nrough the prasent i study, no remarable difference was obserred between the temperature effects in EEU and MEU cores. It was found
! that the difference in the core configuration causes a
'; much more effect on this quanticy than the other diff er-
, ences including the fuel enrichment. De calculated results approximately agreed with the experimental data.
1 DITRODUCTICN i l In accordance with the international Reduced Enrichment for l Research and Test Reactors (RERTR] program, the critical expert-ments using nedium-enriched-uranium (MEU] fuel was launched in l
- l l-.
75 1
[ ., . 1981 using the Kyoto University Critical Asse=bly (KUCA] . There-after, the KUCA experi=ents have been providing useful data with regard to the RERTR program.1-15 Recently, the sessurements of the temperature effects on reactivity were perfor=ed in the KUCA in succession to the study on the void reactivity effects 2 .4 6.t]s11 1*eL5, since the te=per-ature coefficient as well as the void coefficient is a physical quantity closely related to the safety of liquid-moderated reac-cors. It is i=portant to investigate the effects of reducing fuel enrichment on these quantities in advance to a core conversion from high-enriched-uranium (EEU] to reduced-enrich =ent-uranium fuels. With use of light-water-coderated and heavy-water-reflected annular cores constructed in the EUCA, the following effects on the temperature reactivity coefficients were investigated for the temperature range from 20*C through 70*C: (1) the reduction in fuel enrichment, (2) the fuel leading pattern, and (3) the exis-tence of boren burnable poison (3P]. The measured data were analyzed using the SRAC system 16 go assess the computational technique for the te=perature coeffi-cients. For the calculation of this quantity, 3 physical process-p es were taken into consideration; namely, the (1) Doppler, (2) J thermal expansion, and (3) thermal neutron spectral shift effects.
. 51 EIPERLL57. TAL A schematic cross-section of the light-water-moderated and heavy-water-reflected core constructed in the KUCA is shown in Fig. 1. The core can be divided into following 7 concentric regions: the (A) central light-water, (3) inner fuel, (C) control red. (D) outer fuel, (E) outer vessel, (F) heavy-water reflector, and (C) outside light-water regions. The outer fuel region consists of 12 fuel elements, and the inner fuel region 6 fuel e'l aments.
Six fuel loading patterns were employed in the present study; 4 patterns of MEU cores (see Fig. 2) and 2 patterns of HEU cores (see Fig. 3). These cores can be classified into 2 types; namely, "1" and "II". In the type "I" core, all outer fuel elements were fully loaded to its capacity with 17 fuel plates and the criti-cality was essentially adjusted by the number of fuel places inserted into the inner fuel ele =ents from the outside toward the inside in order. In the type "II" core , all inner fuel elements were fully loaded with 15 fuel places and the criticality was adjusted by the number of fuel places inserted into the outer fuel elements from the inside toward the outside. Therefore, the
thicknesses of the central light-water region and the outer vessel region of light-water vere different for the type "I" and "II" . In Figs. 2 and 3 "no 3P" means that thers are no side-plates containing SP; " Cuter 3P" and " Inner 3P" sean that all outer and inner side-places containing 3P, respectively. A heater and a stirrer were installed in the heavy-vater reflector in addition to 3 heaters and a stirrer installed in the dump cank from which light-water is pu= ped up and fed to the core tank in every operation of the KUCA. Seven thermocouples and 2 quart:-type thermometers were also installed to =enitor the uniformity of the temperature as shown in Fig. 4 At several temperatures in the range from 20*C through 70*C, the excess reactivities or the suberiticalities were seasured by the positive period method or the source multiplication method. Then, the measured data of temperature dependent excess reactivities were fitted to a quadratic equation using the =echod of least squares as; c(T) = aT2 + bT
- c , (1) where, p(T) is the excess reactivity at the temperature T ['C],
and a, b, and c are the constants. Thus, the ta=perature coeffi-cients s(T) was determined as; s. a(T) = 2aT + b . (2) M CAI.CUI.ATIONS The temperature effects on reactivity were calculated by the procedure shown in Fig. 5 using the SRAC systemM. The effective multiplication factors were calculated at 3 temperatures (namely, 27'C, $2*C, and 77'C) for which the scattering kernels are pre-pared in the neutron cross section library in SRAC. In the calculation, following 3 physical processes were taken into account; the (1) Doppler, (2) ther=al expansion,
- and (3) ther=al neutron spectral shift effects.
From the public library of 107 energy groups in SRAC based on the ENDF/3-IV file 17, the user library of 50 energy groups was generated. f41th assuming a fixed source problem, the primary cell calculations were performed by the collision probability routine in SRAC. In this step, the cell-everaged 19-group constants f or the actually fueled region was obtained with approximating a curved geometry. by a slab oned. Using the T'40TRAN code in SRAC, the secondary cell calculations were perfor=ed in order to take into account the neutron flux distributions in the azimuthal direction for obtaining 19-group constants of the fuel region. In this step, a curved geometry was approximated by a rectangular one and a special attention was paid to preserve the volu=es of the F' E actually fueled region and the 3P layer. Therefore, with dividing a fuel element into several regions, the plural calculations were performed for the fuel region as shown in Fig. 6. With use of the 19-group constants obtained through the above procedures, the core calculations were performed using the CITATION code in SRAC. A one-dimensional [1-D] cylindrical model was employed in this eigenvalue calculation (see Fig. 7) and the 10-group constants were generated in this step. In this step, the experimental data of reflector savings 3'S were used for the vertical transverse buckling. With use of the 10-group constants, two-dimensional (2-D} R-Z calculations were performed using the CITATION code in SRAC (see Fig. 8). In order to check the differ-ence between the 1-D and 2-D calculations, 1-D calculations using the 10-group constants were also carried out, i . l RESULTS AND DISCUSSION
, Figure 9 shows the comparisons between the calculated and seasured temperature effects on excess reactivity. Note that the calculated values are normalized to the experimental ones at 27'C.
The . calculations gave slightly larger effective multiplication factors than the experimental data, however, these differences were less than 3%. The 2-D calculations simulate fairly well the f a tendencies in temperature effect, whereas the 1-D calculations underestimate chose tendencies. This can be attributed to the d neglect of the positive temperature effects caused by the light-3- water reflectors above and below a core in the 1-D model. Figure 10 shows the comparisons between the calculated and
, measured temperature coefficients of reactivity. The differences between the cesperature coefficients in MEU and HEU cores are not so significant and it is considered that they depends strongly on the fuel loading patterns (see Figs. 2 and 3) as mentioned below.
4 The calculated results approximately agree with the experimental
, ones, however, there exists some discrepancies in the gradients of
'i temperature coefficients between them. The agreements are better j in the type "I" core than in the type "II" core. This tendency 4
was previously found in the analyses of the BP effect measure-ments8 '13 This may be attributed to the difficulty in the
! generation of group constants for the inner fuel region where the
, neutron importance is highest and the curved geometry is most severe.
Figure 11 show's an example of the dependences of temperature effects on 3 physical processes calculated by the 1-D model. In the MEU core, the Doppler effect causes a slightly negative reactivity effect, whereas that in the HEU core is close to zero. The thermal expansion effect causes a negative reactivity effect, 4
}
- whereas the thermal neutron spectral shif t effect causes a large positive effect which overweigh the other effects.
Figure 12 shows the region dependent temperature effects on reactivity in the HEU cores calculated by the 1-D model. It is found that the temperature effects in the fuel regions causes negative reactivity eff ects and those in the heavy-water reflector are approximately tero, uhereas those in the light-vater regions causes positive ef f ects. The positive temperature effects can be attributed mainly to the ef fects caused by the central light-water and outer vessel regions which depend strongly on the thicknesses of light-water layers. In view of the above, the temperature eff ects on reactivity depend strongly on the core configurations, rather than on the fuel enrichment and the existence of SP. It is considered that
' the negative temperature effects in the fuel regions are attribut-ed mainly to the decrease in =acroscopic neutron scattering cross section with the increase in temperature, and the positive temper-ature effects in the light-water regions are attributed to the decrease in macroscopic neutron absorption cross section.
ACKNOWLEDGEMENTS The authors wish to express their thanks to all staff s of the KUCA including Mr. Keiji Kobayashi for their generous assistance l in the experiments. The calculations were performed at the Data Processing Center of Kyoto University. This study was financially supported by the Ministry of Education. Science and Culture. REFER.rNCES
- 1. K. Kanda, et al., "KUCA Critical Experiments Using MEU Tuel",
IAEA-SR-77/30 (1981).
- 2. K. Kanda, et al., "KUCA Critical Experiments Using Medium En-riched Uranium Tuel". Annu. Rep. Res. Reactor Inst., Kyoto Univ., 15,,
5 1 (1982).
- 3. S. Shiroya, et al., " Measurements of Neutron Flux Distribu-tions in a Medium Enriched Uranium Core", ibid. , 141 (1982).
4 K. Kanda, et al. , "KUCA Critical Experiments Using MEU Fuel (II)", ANL/RERTR/~M-4 CCNT-821155, 426 (1983).
- 5. S. Shiroya, et al., " Analysis of the KUCA MEU Experi=ents Using the ANL Code System", ibid. , 449 (1983).
- 6. H. Tukui, et al., " Experimental Study on the Void Reactivity Coefficient in the KUCA", Annu. Rep. Res. Reactor Inst.,
Kyoto Univ., J_6,, 1 (1983). l
- 7. S. Shiroya, et al., " Analysis of the KUCA MEU Experiments Using the ANL Code Systa=", ibid., 17 (1983).
- 3. S. Shiroya, et al., " Analysis on the KUCA MEU Experinents (II), 3cron Burnable-Poison Effect", JAERI-M S4-073, 369
-(1984). ,., 9. M. Hayashi, et al., "Calculatien of the MEU-EEU Coupled Core in the KUCA", ibid. , 377 (1984) .
- 10. K. Kanda, et Ill., " Experimental Study on the Void Reactivity Coefficient in the KUCA", ibid. , 388 (1984) .
- 11. Y. Senda, et al., " Analysis on the Void Reactivity Measure =ent in the KUCA", ibid. , 399 (1984) .
- 12. K. Kanda, et al. , "HEU-MEU Mixed-Core Experi=ents in the KUCA", International Meeting on RERTR,15-18 Oct. , Argonne.
USA (1984). j 13. S. Shiroya, et al., " Analysis of Critical Experiments Using i Mediu=-Enriched-Uranius Fuel in Kyoto University Critical l Asse=bly (KUCA)", J. Nucl. Sci. Technol., 22, 507 (1985).
- 14. H. Fukui, et al., "Effect of Reducing Fuel Enrich =ent on the Void Reactivity, Part I; Experimental Study" Nucl. Technol.,
70,, 301 (1985).
; 15. T. Senda, et al., "Effect of Reducing Fuel Enrichment on the
$ Void Reactivity, Part II; Analytical Study", ibid. , 31S 3', (1985).
',' 16. K. Tsuchihashi, et al., SRAC: JAERI Ther=al Reactor Standard Code Svste= for Reactor Design and Analysis, JAERI 1285 (1983).
- 17. D. Garber, ENDF/3 Su==arv Documentation. BNL-NCS-17541 (CIDF 201), 2nd. Edition (1975).
l
, m c.ecr.1 u. -e..,t t.e
/ .C. E) (3) tamer hat tagtse (C) Caecret ted testae I
c1 secar h A tasase j
' (1) 3 star hose 1 tagtse (1.Asas-=ecar restas secween gg3 tan encar fuel sleemmes and cae.neewser reflestart (F) leavy=.ieter tatiester lagtse (G) Lisas desar lagtse heside tas Eneve-water tailesser Fig.1 Schematic cross-section of the KUta : ore.
r s s,- ir leir s ' s.s to t m aus eut m BE M ir % .s /b4 / W- w ir/ (as *5 ' s- M _ r.
. it - .n fir y, ,r t je g ., ,s gs s
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- 10 -
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SDCUS OF DEVECPMIINr AND IRPADIATICN PERfCPMANCE CF ADVANCED l PICLIFEIUC'ICN RESISTANT MTR EUEL AT NCIGM l l M. Hrovat, R.-W. Hassel and E. WehrAr l NtEEM Qatst i
, EHl450 Hanau 11, Postfach 110000, FIG i
mis paper describes the current status of dev=1m and irradiation performance of fuel elements for storial Test and assearch (MPR) Reactors with :nedium enriched uraniun (REU, 4 451 l -
- i. 735-U) and .l.ow e.nridied .urantun (IJU7 4 20 4 735-C) .
I ?
- 1. FUUrSIZE W Atc LEU FUEL IIJMNFS .
l By the end of Septanter this year the nunber of ordered MEU and ' LEU fuel elenants to be fabricated by NUIGM totalled 425: ! 4 i . 251 tal. f.aol elements ; , j e 118 Uso s afuel elements l l
! . 56 U3 Si t M elements.
J !-l 1.1 tale Full-Size feu and LEU.Puel Elenants Se 251 orders for tais fuel elements with reduced uraniun enrichment at the end of Sgtember consisted of 212 MEU and 39 LEU f-elements. So far 156 talx 1.wl elemer.ts have been delivered and '
+
inserted into various reactors. 78 MEU f.aol elenents reached their j target burnup and wre removed from tim :eactor. Se highest burnup 4 of 73 4 235-0 was adtieved by the fuel elenents with an uraniun den-sity of 1.74 g U/cos meat,in the ORR. B is burnup corresponds to a
, fission density of 1.3 x 10*' fissims/m a meat. Se 39 IJU fuel
] elements are still under irradiatica (s. Fig. 2) . ,
. 1.2 U3 0g Full-Size Fuel Elenants '
; Fer fabrication of the U 08 3 fuel powier a new procedure ws j developed at 2 ERIN. An essential step is t!w preparation of an uni- ,
; fermly dense U 3 05 powier with an appropriate grain size distribution !
t by crystal grensth treatment. Bus a fuel powder with a very narrow
- I 1
i
- o. .
' grain size spectr ra ard high sL est cecretical density can be ob-tained. Sis fabricatien step has preved itself verf wil in pec>-
duction scale. Cut of 2e total of 113 fuel ele ents cedered, 25 have s1 ready been delivered to reacter eptors. 2e remainder of 93 fuel ele-ments is in preptien (s. Fig. 3) . Seventeen of de delivered fuel elernts reached their target turnup ard wre removed frcm the reacters. Fig. 3 sh::ws the prcperties of the rencved U 303 :.EU fuel elements. Be uraniun density in the meat of these fuel ele-ments varied bee.een 2.08 (FSR) and 3.1 (FT,-2) g U/cn2 meat,
- depending en plata geanetry ard reacter requirements. Be meat thickness varied witin a relatively vide range fr:m 0.51 mm (standard) up to 1.5 m, latter e.g. in case of CPR fuel elements.
Pest irradiation examiraticn of de FJR elenents has been ccm-r pleted. At a .iaxt:t:m burnup of 73 % 235-U, de thickness gr wth of
' all investigated plates is belcw 40 pn. In ceder to test the mechanical integrity de plates wre subjected to a blister test.
Here, by successive heating of the plates to higher and higher tatperatures, de te perature is detected at which blisters are created at the plate surface. me results are very enceuraging. 2e measured blister ter:perature for all investigated plates was above 500*C. 1.3 U3Sit Pall-Size I.EU Nel F.lements Uraniun silicides are distinguished by very high densities I which, in the case of U> Sit with 12.2 g/cn2, exceed the value for U C, (8.4 g/en2) by a facter of ateut 1.5. Be first U 3Sit S full size fuel eierients were fabricated by NCKDi in 1982 and inserted into CPR. Since then a total of 56 fuel elanents have been crdered and 6 of them already inserted in the CPR and FPG-2 reacter. Pig. 4 shcws the properties of de inserted fael elements. me CPR fuel elenents have been renoved fran the reacter. one of the eas elenents reached a high burnup of 82 % 235-U,
' corresponding to a fissicn density of ateut 1,79 x 10
- fissicns/cn2 meat.
me pest irradiatien examiratien of these tas elenents is corpleted. Fig. 5 shcus a typical phetanicrograph of a fael plato
- with a burnup of 82 % 235-U. me USSit fael grains '4th this high i
burnup are presented in Fig. 6 by SD4 photcgraph. Se Figures 3 and 6 reveal a uniform fael structure wiccut arr/ cracks. In additien plate thickness measuremnts wre perfecned. Se j icw thickness gr wth is in a good agreenent with the gr:wh of te stardard UAlx fael plates. L_
l
, 2. TES:' PIATES Test plates (can ed experimental mini plates) have been irra-diated in the CPR, FR7-2 and FPG-2. In the frame wrk of mini plate 1.%tien tests de fonowing fuels wrs used:
-% (2.2 g U/en2 meat)
-U03 3 (3.2 g U/cn2 meat)
- U 3Sit (5.0 g U/en2 meat, majerity U3 Sis)
- U 3 Si (6.9 g U/cn2 meat, majcrity U3 Si) i de densities in brackets represent the maxinun values. Fig. 7 t
shows a survey of test plate investigations. t e 1. % tien tests with UAls and U3 Og in the ORR are,ter-minated as wu as the correspcniing post irradiaticn examina-
*MJ . .
Fig. 8 represents the microstructure of the U3Og test plates l before and after L%tien. In spite of high volune loadLg and
' high burnup a crackfree micrcstructure is evident. m e good metal-
; lurgical bcrsiing between cWing and meat is cbvious.
l Se tests with the uraniun silicides are corpleted to sone extend. met advanced is the capsule irradiaticn test in the FR7-2, i the am=' lated burnup is about 81 % 235-U. Ibr the U Si plates l with an uraniun density of 6.05 g/an a meat this bupup3 corresperds to the so far highest fission density of 2.41 x 10" fissices/en2 l meat. In general it can be stated: All inserted fuel elenents and experimental mini fuel plates 1 ! j fabricated at :D2M performed wu in endurance tests and l > exhibited an unobjecticeable irradiaticm behaviour, independent of j fuel type. ' j . I l 3. CINC:L'SICN AND C(MICK
% fuel has proved very won in endurance irradiation tests.
! }bwever, the relatively icw attainable uraniun density in the meat l t makes this fuel applicable with IEJ for low power reactors cnly. !. ' U3 0, fuel up to a 0-density of 3.2 g U/cn2 meat is a reliable well proven I.EU fuel with following advantages:
, - relavitely favourable fabricaticn costs
- good 1.%ticn perfernance and
- alacet similar uraniun chemistry as it is required for the fabricaticn of the oxide fuel for powr reacters.
I l . k
= - - . _ _ . - - - _ _
i l t U3 Si fuel is distinguished by a high attainable unniun density in the meat, up to 7 g U/en2 meat. C:nsequently this fuel l is able to meet the requirements placed on I.m fuel elments even i fer high pcwer MTR . Because of the by far greatest fuel demand of high pcwer reactors this fuel M11 play the .est i::pertant role in 9.e future. It is 'crth mentiening, that about 56 silicide fuel ele- l ments have been ordered by StJCM since .4y,1984. Carrently the process parameters are being established to erect the fabricaticn ; j line with a capacity of 15 000 silicide fuer plates correspcniing to about 750 standard full size elements. Be irndiatien perfer-j mance of this fuel is unobjectionable.
' % investigate the irradiatien perfonnance of high loaded i
silicide plates with 3.C as burnable poisen a second irrv11stien experiment in the FRJ-2 is b progress. Be test plate details
; are corpiled in Fig. 9.
l , he volune loading of all plates is equally adjusted to about 45 "/o f'ael in the meat. S e silicon ecntent is varied bee m t 4 and 5.9 /o. Cbnsequently the attainable density varies betwen 6.0 and 7.0 g U/cn* meat. To meet the required areal 235-U loading the meat thicknesa L is within the range of 0.48 and 0.56 ::in. Irradiaticn start::p is l projected for the erzi of this year. We waulcal fabricaticn of IRJ MTR fuel elements necessi-tates the intensivaticr. of the Aalve betwen reactor cperator and fuel fabricator with the aim to standardized fuel plate speci- '
' fications. Without taking into account any penalty concerning the safe irradiation perfa m preferabely such properties have to i be reviewed, which effect either excessive rejection rates or un-necessary high quality inspectica efforts. Currently the folicwing itars appear to be reevaluated:
Size distributicn of the fuel powder (firms fraction)
- Cladding thickness
, - White spots
; - n:e1 hcmogeneity in the meat
- Surfacs defacts. i 5
5 6
Total Nurnber 425 ' e UAlx FE 251
+ U0 38 FE 11 8 e
U3Si2 FE 56 Status Oktocer 1985 . FULL SIZE MEU & LEU FUEL ELEMENTS ORDERED by NUKEM rin,1 Total Number 251 (212 MEU, 39 LEU) e in Preparation 95
- Inserted 156 (117 MEU, 391.EU ) ,
- Removed 73 MEU
- Max. Values of Removed ORR FE:
- U Density 1.74 g U/cm3 meat
- Plate / Meat Thickness 1.27/ 0.51 mm
- Burnup 73" ass U.13 10 8' f/cm3 meat Status Oktet:er 1985 UAix FULL SIZE MEU & LEU FUEL ELEMENTS Fipm 2
, I Total Number 118 l - In Preparation 93
- Delivered 25 (17 Removed )
i Reactor
, ASTRA OR 3 FRG-2 HFR ORR FE Details Number of FE l 3 3 7 2 2 I U Dendly , 2,9 ' 2.65 l 3.1 2.08 2.18 ; M/cm3 meat )
Maat Th kness 0.84 0.65 0.51 1 32 1.50 (m ) t' , Burnup % nsu l 50 1 65 ' 60 , 73 80 Fiss n s (tc me t))Q , 0.73 l 0.86 : 0.93 j 0.75 j 0.88 Status Oktocer 1985 U30s FULL SIZE LEU FUEL ELEMENTS Figure 3 Total Number 56
- In Preparation 50
, - Delivered 6 ( 2 Removed )
Reactor I l ORR FRG-2 l
, FE Details i ,
! Number of FE 2 4 l .
I U Densit i I(g U/cm5 meat ) I 4.72 3. 7 i Meat Thickness 0.51 1 0.51 ' (mm) l l l Burnup (M issU) l 82 30 l
; Fission Density !, 1.79 0.49 4(f/cm3 meat ) a 1021 {
Status Chtecer 1985 U3Si 2 FULL SIZE LEU FUEL ELEMENTS F17xe 4
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( w%g o,... A ., s : : p* 1 100 pm , l i Photomicrograph of U3 Si 2Fuel Elemen l i Irradiated in ORR,Burnup 82% 23sU i Figure 5 i i
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; SEM Photographs of U 3 Si 2 Fuel Element, i irradiated in ORR, Burnup 82% 2ssU :
i rig re 6 1 I l I I
l Reactor , l ORR FRJ-2 FRG-2 ' Plate Detsits Fuel Type UAixU081 3 U3 Si U30 3.U3Si2 U3 Si U3Si2-Number of 4 8 8 1 5 3 5 Test Plates D'" ' gU/c 2.2 3.1 6. 9 ' 2.4 5.0 6.1 4.8 meat ) , , Burnup (% ssU) 90 90 65 82 82 81 25 ff c Jm at) 21023; 1.6 M3 2.0$ G1,1.90 2.0 M2'
- Status Okt.1985 i PIE ccmp. ltrad.PfE in progress .irrad.. I l 1 i !
i 1 TEST PLATE INVESTIGATIONS Figure 7 i before irradiation OCoum.
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l Fuel U3Si 10 0 70 70 50 50 Fraction (mole M) U3Si2 30 30 50 50 Si Content (%) 4 4.8 4.8 5. 9 5. 9 U Density 7.0 6.5 6. 5 6.0 6.0
.(gU/cm3 meat)
Meat
, ness
- 0.48 0.52 0.52 0.56 0.56 osson B.C Number of 2 2 2 2 2 iTest Plates 84C Content . 14 5 mg /em3 me.it TEST PLATE DETAILS for SECOND IRRADIATION EXPERIMENT in FRJ -2 Figure 9 S
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n e THE FA3RICATION AND PERFORMANCE OF CMADIM SILICIDE DISPERSION FUEL FOR TEST REACTORS D.F. SEARS, J.C. WOOD, L.C. BERTHIAUME, L.N. HER3ERT AND J.D. SCHAEFER Accaic Energy of Canada Limited - Research Company Chalk River Nuclear Laboratories
, Chalk River, Ontario K0J IJO .
o ABSTRACT Fuel fabrication effort is now concentrated on the consissioning of large-scale process equipment, defining product specifications, developing a quality assurance plan, and setting up a mini-computer zacerial accountancy system. In the irradiation testing program, full-size NRU assemblies containing 20% enriched silicide dispersion fuel have been irradiated successfully to burnups in the range 65-80 aconic percent. Irradiations have also been conducted on mini-elements having 1.2 na diameter holes in their mid sections, some drilled before irradiation and others after irradiation to 22-83 aconic percent burnup. Uranica was lost to the coolant in direct proportion to the surface area of exposed core aster 141. Pre-irradiation in the intact condition appeared to reduce in-reactor corrosion. Fuel cores developed for the NRU reactor are
'; dimensionally very stable, swelling by only 6-8% at the very high burnup of 93 atomic percent. Two important factors contributing to this good performance are cylindrical clad restraint and coarse silicide particles.
Thermal ramping tests were conducted on irradiated silicideg2ispersion released fuels. Small segments of fuel cores Kr starting at about 520*C and peaking at about 680*g3 Af ter a holding period of 1 hour at 720*C a secondary Kr peak occurred during cooling (at about 330*C) probably due to thermal contraction cracking. -Whole mini-elements irradiated to 93 aconic percent burnup were also raged thermally, w1th encouraging results. After about 0.25 h at 530*C the aluminum cladding developed very ' localized small blisters, some with penetrating pin-hole cracks preventing gross pillowing or ballooning. l
y . 4
. r f
THE FA3RICATION AND PERFORMANCE OF CANADIAN SILICIDE DISPERSION
, FUEL FOR TEST REACTORS
'd D.F. SEARS, J.C. WOOD, L.C. BERTHIAUME, L.N. HEREERT AND J.D. SCRAEFER Atomic Energy of Canada Limited - Research Company Chalk River Nuclear Laboratories Chalk River, Ontario K0J IJO ABSTRACT Fuel fabricacion effort is now concentrated on the commissioning of large-scale process equipment, defining product specifications, developing a quality assurance plan, and setring up a mini-computer material accountancy system.
In the irradiacion testing program, fulf-size NRU assemblies containing 20% enriched silicide dispersion fuel have been irradiated successfully to burnups in the range 65-80 atomic percent. Irradiacions have also been conducted on mini elements having 1.2 na diasecer holes in their mid-sections, some drilled before irradiacion and others af ter irradiation to 22-83 atomic percent burnup. Uranium was lost to che coolant in direct proporcion to che surface area of exposed core sacerial. Pre-irradiation in the intact condicion appeared to reduce in-reactor corrosion. Fuel cores developed for the NRU reactor are . dimensionally very stable, swelling by only 6-8: at the very high burnup of 93 atomic percent. Two important factors contributing to this good performance are cylindrical clad restraint and coarse silicide particles. Thermal ramping tests were conducted on irradiated spersion fuels. Sas11 segments of fuel cores silicide releasedggKr starting at about 520*C and peaking at
,about 680*g3 After a holding period of I hour ac 720*C a second.s ry Kr peak occurred during cooling (at about 330*C) probably due to charmal contraccion cracking. Whole mini elements irradiaced to 93 atomic percent burnup were also raged chermally, vich encouraging results. After
' abouc 0.25 h at 530*C che aluminum cladding developed very localized small blisters, some with penetracing pin-hole cracks preventing gross pillowing or ballooning.
ke
p T l I
.: INTRODUCTION At Chalk River Nuclear Laboratories we have successfully develo
- j. ) 235U) silicide ped anddispersion irradiatedfuels low enriched uranium with which (LEU, we plan i.e. 20 to replace
, the 93% enriched Al-U alloy fuel currently used in Canadian .'
research reactors. Seven full-size 12 element NRU reactor assen-
.e* ' blies containing silicide dispersion fuel are currently under i irradiation, and to data have achieved burnups in the range 65 to l 80 atomic percent. In other fuel experiments in-reactor corrosion l~ rates in deliberately- defected mini-elements were found to be satisfactorily low and post-irradiation thermal camping tests have
. shown that silicide dispersion elements maintained considerable ! structural integrity after heating to high temperatures. Our fabrication development effort has been concentrated on commissioning large scale process equipment for manufacturing j full-size assemblies. This work also includes defining product specifications, developing a quality. assurance plan, and setting up a mini-computer asterial accountancy system. l FUEL FA3RICATION DEVELOPMENT l l The LEU fuels developed for use in Canadian research reactors l contain particles of uranium silicide dispersed in an aluminum matrix. Fuel elements are clad in aluminum. The dispersed l I silicides studied in the development program were U3 Si l { (U-3.9%we%51), USiA1 (U-3.5we%Si,1.5 wt% A1), and Usi*Al l i (U-3.2ve%S1, 3.0 we% A1), the uranium being enriched to 20% , 235 U The silicide loadings were such that the fuels had the same 235 U densities and geometries used in existing NRU rods.
- ;' The procedures developed for laboratory-scale fabrication of i
dispersions containing about 62 wt% silicide (equivalent to the 23,5U loading needed for NRU) and about 73 wt% silicide (the i I equivalent for the NRI reactor) have been described elsewhere (2,3 ) . Large scale production equipment has now been installed and is undergoing commissioning. 9 , , Melting, Casting and Heat Treatment The zelting/ casting furnace, powered by a 30 kW high-frequency , induction unit, is capab,le of casting 4 kg silicide billets. The zirconium oxide crucibles which hold the uranium and silicide charge have been redesigned with tapered bottoms to enhance the
; nelt flow and reduce slag sticking to the bottom. Heating times have been typically 0.25 h for a 4 kg charge and the scrap race has been reduced significantly to levels as low as 2%, compared to l the 5% previously attainable. The cast billets have been vacuum I
i r
I heat-treated as before (2) with "deltizing" having occurred af ter 96 h at 300*C. Chip Machining, Washing and Haamer-Milling Lathe machining using a drill with replaceable, indexable carbide inserts has been selected for silicide chip production because the production rate was high and chip thickness (approx 1-mately 150 um) could be closely controlled by adjusting the cutting speed,. feed race and coolant pumping rate. A large rigid lathe was needed to reduce the vibrations from the cutting opera-tion and enhance the carbide cutting tool life. A centrifuge proved effective for washing the large amounts of chips required for manufacturing full-length fuel elements. Batches of up to 4 kg of chips have been washed and 99% of the cutting fluids and oils on the chips were removed using the centrifuge. The remaining traces of oil were removed from the r chips by vacuum baking in an oven. A large rotary hammer a111 has been installed for comminucing the silicide chips into powders. This impact aill has a built-in rotary classifier system which allows the finely powdered product to leave the grinding chamber, but the oversized particles from
- the initial grinding stage are continuously returned to the mill for further grinding. The ground material is passed through a cyclone which discharges the product classified at both ends of the specified range. New glove boxes have been designed to house ,
the ba===r-n111 and the centrifuges in an inert atmosphere, and r these are presently under construction. l Silicide Dispersion Core Extrusion The fuel cores are fabricated by direce extrusion of blended , powders of aluminua and uranium silicide which have been heated in the hot die ecvity of an extrusion press (2,3). The die block and the die inserts were redesigned to allow quick tooling changes to be made while the die block is hot. As previously reported (1), ! our high production rate enables us to annufacture the annual l-requirement of NRU cores in a 12-week extrusion campaign. . Material Accountancy System + l The production line for manufacturing fuel assemblies is
- divided into a number of zones or asterial balance areas (MBA's).
Each MBA has a critical limit of fissile asterial which can be l resident in that area, and as fuel moves through the proddecion i l sw'
Line an account has to be made of material coming into and leaving each K3A. As a result of the large quantity of material that will be required for manufacturing the innual requirement of fuel rods for the test reactors, the several forms in which the uranium exists while in process, and the numerous transfers that will have to be made to and from the EBA's as the material moves through the manufacturing process, we are setting up a sini-computer sacerial accountancy system. This will greatly reduce the time an operator needs to transfer material from one process to the next, and since all areas within the system will be monitored simultaneously, closer inventory control can be effected for better process control, criticality control and accountability. IRRADIATION EXPERIMENTS The test vehicle for irradiating silicide dispersion fuels in most tests so far has been the mini-element, but we have also fabricated and successfully irradiated full-size NRU dispersion fuel elements. The mini element fuel core diameter (5.5 mm) and clad wall chickness (0.76 mm) are the same as in full size NRU elements. Hini-elements are however only 184 am long compared with 2.9 m for NRU elements. The mini elements also resemble NRU elements in that they have six cooling fins at 60* intervals around the cladding, the fin width being 0.76 as and fin height 0.96 an. Irradiation Program Status The irradiation program is outlined in Table I which presents the scope, objectives and current status of the tests. The most important achievements during the past year were: * (1) the completion of the FZZ-911 and FZZ-915 defected fuel irradiations, post-irradiation zetallography, and the
! determination of the amounts of fuel lost from the purposely drill-defected mini elements, , (ii) the irradiation of seven full-size LZU driver rods in the NRU reactor (Experiment FZZ-913).
(iii) completion of post-irradiation thermal ramping tests with l.
~
sessurement of fission product activity released from silicide dispersion fuel. Previous work in the irradiation program have been described elsewhere (1-3). These earlier results showed that the swelling behaviour of the Al-USiA1 silicide dispersion fuel with 20% enrichment was similar to that of the Al-U alloy fuels with O
equivalent U-235 loadings. The swelling was about 4.5 volume percent af ter a final burnup of 56.4 aconic percent (Table 1 - Experiment F22-905). The FZ:-909 experiment showed that mini-elements containing the NRU compositions (Al-61.5 vt% USLA1 and Al-62.4 wt% USi*Al and the NRX compositions (Al-72.4 ve% USiA1 and Al-73.4 wt% USi*A1) behaved similarly. The NRU composition sini-elements had swollen by slightly less chan 1 volume percent per 10 atomic percent burnup up to a terminal burnup of 93 atomic percent, and the NRX composition mini elements had swollen by slightly more than 1 volume percent per 10 atonic percent burnup up to a terminal burnup of 82 atomic percent. These are very i=portant results
, because swelling was approx 1:ately linear right up to 93 (NRU) and
, 82 (NRX) accmic percent bu rnup , so the silicide dispersion fuels l reached their terminal burnup without exceeding the threshold of breakaway swelling. Fission gas bubbles ranging in siae up to 5 m in diameter were contained in the kernels of the fuel i particles.
4 In the FZ:-910 experiment deliberately sub standard minielements containing a high loading of' fine silicide powders showed greater amounts of swelling af ter 60 atomic percent burnup compared with similar tests on fuels with coarser particles'. The lesson learned from the performance of the FZZ-910 sini elements is thac the amount of fine s111 cide powder particles (<44 um) in the dispersions must be controlled, particularly at high silicide loadings. We have therefore designed an additional experiment (FZZ-918) to assist us in establishing the limits on the particle
, sizes to be used in manufacturing the dispersion fuels.
A brief review of the important features of the work in the past i year follows.
] FZZ-911 Experiment. In the first part of this experiment, four mini-elements had L.2 mm diameter holes drilled in the cladding sid-section and were irradiated in the linear power range i
60-87 kW/m in NRU. The first two mini-elements were removed from reactor after reaching 21 and 39 atomic percent burnup respec-tively. The third and fourth nini-elements were removed after reaching 57 atomic percent bu rnup. Post-irradiation metallography and neutron radiography revealed that a small roughly henispherical cavity of 1 mm radius i had developed in the first fuel element after 29 days at power in-reactor. The second aini-element developed a roughly ellip-soidal cavity 3.7 =m major axis and 1.2 mm ainor axes af ter 62 days at power. The third and fourth mini elements were both removed after 98 days at power in-reactor. The third nini element
had developed an ellipsoidal cavity with 4.4 zm major axis and 3.2 mm sinor axes while the fourth sini element's cavity measured 10.5 mm =ajor axis and 3.2 mm ninor axes. The elongated ellipsoid shape with the major axis parallel to the fuel element axis was somewhat unexpected. Rather than the shallow ellipsoid, we had expected the hotter material from the centre of the fuel core to be removed preferencially. These cavities correspond to 1.1 mg, 3.2 mg, 9.3 mg and 48.0 mg of 235U lost to the coolant respec-tively (Figure 1). To a first approximation fuel loss is directly proportional to the exposed surface area. The corrosion rate of the purposely defected fuel elements appears acceptably low. The two nini elements that were removed af ter 98 days irradiation lost amounts of 235 U differing by about a factor of five (48 mg vs 9.3 mg). Meta 11ography revealed that the fourth nini element (#61.3) which lost the greatest amount of fuel had developed a crevice between the cladding and the fuel core, and thus exposed a larger surface of the core to the hot coolant than the third mini element (#61.3). The original objective of the second part of this experiment was to evaluate the performance of intact nini-elements having slight as-fabricated or deliberately introduced imperfections in the core surface as part of a feasibility study into the potential for relaxing core surface quality control standards. However, the mini-elements contained very fine silicide particles of the size that led to enhanced swelling in experiment FZZ-910. The enhanced swelling occurred in FZ:-910 mini elements with a high silicide loading (72.4 we%) compared with the more dilute dispersion - 61.5 wt% in the FZ -911 irradiation, but as a precautionary measure we have decided to perform periodic swelling measurements. The core volume of a sample nini-element (N-6) had only increased by approximately 4.9 vo1% af ter 60 atonic percent burnup , compared with 7.0 to 17.5% swelling in the FZ -910 cores at about the same burnup. The swelling of the N-6 mini element is in the same range as FZZ-905 ind FZ:-909B mini elements at burnups of approximatley 60 atonic percent, indicating that the swelling performance is clearly acceptable at the lower silicide loading. Visual inspections indicated that the naturally-occurring core surface imperfections had no detrimental effect on in-reactor performance. FZZ-913 Experiment. This experiment is the first irradiation of full-length 12-element NRU fuel rods. Three rods (36 elements) containing Al-62.4 ve% USiA1 and four rods (48 elements) contain-ing U3 Si (20% U-235) are currently being irradiated in the URU reactor, and to date have achieved burnups in the range 65-80 atomic percent. All rods have been performing reassuringly well.
F 2-915 Exoeriment. This experiment was similar to the FZZ-911 experiment, the =ajor difference being that the mini-elements were pre-irradiated to a variety of burnups in the range 22.3 to 82.6 atomic peretat before the 1.2 mm diameter holes were drilled in the clad mid-sections. The mini-elements were further irradiated is the NRU reactor to evaluate the performance of the " defected" fuel elements, and during the additional 38 days the mini-elements were in-reactor, no increase in activity in the coolant above the normal background was detected. Neutron radiography and metallographic examinations revealed that the cavities were typically 0.7 =m deep by 1.3 am across, i.e. only . marginally larger than the original cavity made by the drill tip. These results seem very encouraging, indicating that the corrosion 4 resistance of the LIU fuel is possibly increased by previous burnup, probably because irradiation or in-reactor thermal ef fects had improved fuel to clad bonding cohesion. However, conclusions as to the effect of prior irradiation drawn from a comparison of the FZZ-911 and FZZ-915 experiments are compromised somewhat by the fact that the silicide particles in the FZZ-911 dispersions were finer and may therefore have influenced the corroston race of the dispersion. POST-IRRADIATION HEATING TESTS OF SILICIDE DISPERSION FUELS Thermal ramping tests were conducted in the Hot-Cells to determine the effects of temperature excursions on the dimensional stability and fission product activity release from previously-irradiated silicide dispersion fuel. Whole mini-elements and short segments of mini elements with the fuel meat exposed were chosen, having fuel burnups of either 23 or 93 atomic percent. Ralf the samples contained Al-61.3 we% USLA1 and half contained Al-62.4 we% USi*A1. The equipment used in the tests is sketched in Figure 2. Each spec. en was placed in a stainless steel boat which was inserted into a replaceable stainless steel containment tube passing through a tubular furnace. A stream of argon was passed over the specimens at a flow race of 1.66 mL/s. The gas flowed into a delay chamber for gamma counting by a germanium detector coupled to a gamma spectrometer. A sodium iodide crystal was also used for gross gamma measurements. The temperature was controlled by four thermocouples, one of which was placed adjacent to the specimens in the furnace boat. . The samples were heated at races between 0.2 and 0.4 'C/s to prescribed temperatures in the range 530 to 720*C. The tempera-ture was then held constant (14*C) for one hour, at which time the power was turned of f and the specimens cooled to ambient
temperature. Af ter the tests the samples were examined visually and photographed. The samples were then sectioned, mounted in resin, metallographically polished and then etched in Murakami's reagent to make identification of phases with an optical
- aicroscope easier.
Preliminary data are disclosed in Table 2 indicating the tempera-cures at which fission product activities (85Kr and 137Cs) were observed and also post-test dimensional and visual observa-
. tions. These data are incomplete pending chemical dissolution of l the remnants of the samples to determine the quantities of retained fission products, and hence fractional releases. Figure 34 shows, erature, the distribution of recorded asafunctionoftimeandtang7 gamma activity for 85Kr and 1 Cs released from specimen KiB1, a 25.7 na segment of minielement that had been irradiated to 23 atomic percent burnup. The small segment released 85Kr starting at about 520*C and peaked (5.96 x 107 3q) at about 680*C. Af ter a holding period of I hour at 720*C a secondary peak occurred during cooling (at about 330*C) probably due to thermal contraction cracking, the secondary peak reaching 3.46 x 106 sq before dropping to background at 150-200*C. The 137Cs gave a much weaker signal, two or three orders of magnitude less than 85 Kr and the onset of release was delayed 0.1-0.2 h relative to 85Kr release. A major problem with the 137Cs was that it placed out in the gas line and counting chamber, giving erroneously high apparent release data as the test proceeded, culminating in apparently high releases even when the specimens had cooled (dashed lines in Figure 3a).
Whole mini-elements irradiated to 93 atomic percent burnup were also ramped thermally. The release behaviour of specimen K556 which was ramped to 640*C is shown in Figure 3b. 65Kr was released at 580*C during' the temperature rise, peaked (2.93 x
- 108 Bq) just before reaching the 640*C plateau, and peaked again i
(7.58 x 106 Sq) during cooling. There is a striking resemblance between the krypton release patterns obtained from the short fuel specimen of burnup 23 aconic percent (Figure Ja) and the whole l aint element of 93 atoale percent burnup (Figure 3b) notwith-standing test temperature differences. Mini element K6B7 which had been heated to $30*C and had not begun releasing 85 Kr until 0.25 h af ter reaching 530*C had not l pillowed or ballooned but rather had developed several very l localised blisters. Some of the blisters had developed pinhole , ! cracks and released the fission products from the core. This behaviour is interpreted to mean that coolant channels would not become blocked, even if the fuel was subjected to some hypothet-ical abnormal event with the potential to cause overheating of the i core e.g. 530*C compared with the normal operating aaximum of l 200*C. - l l 4e* * -+ - -
On completion of the thermal ramp tests, when the specimens were examined, all the fuel cores were observed to have swollen. Known swelling mechanisms include (a) gas bubble coalescence and (b) thermal reaction between the s111 cide particles and the aluminum matrix which pecduces the aluminide UA13 and voids (condensed vacancies) at the particle-matrix interface.' Metallography confirmed that these mechanisms had indeed contributed to the observed core swelling. Macallography of the nini-element segments thermally ramped to 720*C revealed that the silicide dispersion fuel core had completely reacted with the aluminum matrix material and the ' cladding aluminum had melted or been partially consumed by reaction with the peripheral fuel core material. The core had transformed into an essentially homogeneous two phase alloy probably containing UA13 and UA14 phases with Si dissolved substitutionally throughout the aluminide lactice (Figure 4). The stucture of the transformed core was similar in both mini-element segments ramped to 720*C except the specimen previously irradiated to 23 atomic percent burnup contained smaller fission gas bubbles than the specimen irradiated to 93 atomic percent bu rnup. CONCLUSIONS
- 1. Four additional full-size NRU 12-element assemblies containing Al-61.0 wt% U 3Si have been f abricated and installed in the NRU reactor. This brings the total to seven full-sized assemblies successfully irradiated in the NRU reactor. Burnups are in the range 65-80 atomic percent.
- 2. The in-reactor corrosion resistance of purposely drill-defected mini elements appears very good, the uranium lost to the coolant being in direct proportion to the surface area of exposed core material. Pre-irradiation in the intact condition appeared to reduce in-reactor corrosion, possibly because irradiation or in-reactor thermal ef f acts improved fuel-to-clad bonding cohesion.
- 3. In thermal ramp casts, a whole mini-element irradiated to 93 atomic percent burnup developed small localized blisters, some with pinhole cracks releasing fission products (83Kr and 137Cs) after 0.25 h at $30*C. This behaviour prevented gross pillowing or ballooning.
4 A mini-element irradiated to 93 atomic percent burnup and ramped to 640*C developed radial cracks which tear the cladding and release fission products from the core. Even at this high temperature the element maintained considerable structural integrity.
l . l j 5. The new induction-heated bottom pouring furnace has cast 4 kg U3 Si billets and achieved less than 2% scrap race.
- 6. Lache-machining using a drill with replaceable, indexable carbide inserts produces uranium silicide chips and the chip
- sizes can be closely controlled. Hammer milling these chips produces suitable silicide powders.
l . REFERENCES (1) J.C. Wood, M.T. Foo, L.C. Berthiause, L.N. Herbert, J.D. Schaefer and D. Hawley, " Reduced Enrichment Fuels for Canadian Research Reactors", Proc. International Meeting on Reduced Enrichment for Research and Test Reactors, ANL, l Argonne, Illinois, 1984 October. (2) J.C. Jood, M.T. Foo and L.C. Berthiause, "The Development and i Testing of Reduced Enrichment Tuels for Canadian Research l Reactors", Proc. International Meeting on Research and Test i Reactor Core Conversions, Argonne, Illinois,1982 November. (3 ) J.C. Wood, M.T. Foo, L.C. Berthiause, L.N. Herbert and J.D. Schaefer, " Advances in,the Manufacturing and Irradiation of Reduced Enrichment Fuels for Canadian Research Reactors *, Proc. International Meeting on Reduced Enrichment for l Research and Test Reactors, Tokai, Japan, JAKRI-M-84-073, l 1983 October. i r l l
TAutE la lleWGIATKlel listatal Ape} stAllE . U-235 Fimi ( Euleriammt b da:r of Gwe hterial Entide- Test Qa sesit knap [ El.sms.ts (a) ansa 2 Otabetives Statam (1) (4:1) ; i l l 62Z-9h itt A141.5 LUtAl 20 Omp.nre Disga:rsicens Irrallatican l 41 Al-21 U 9) witta Alloys Gaylete %.4 41 Al-37 U 45 (h0El Gagamititais) P.I.E. Gasplete l 67Z 'MM til Al-72.4 IE1A1 20 Test tell Irrallatite Gmg.lete 82 tel Al-73.4 IE!*Al 20 Gagamig liams P.I.E. Gasplete ! 122-9hti A141.5IE!Al 20 IUtda bnnap itt last te elesasats diacimargpl - til A142.4 IE!*Al 20 Gaifismet tima at 93Z laina,. 91 e P.I.E. Gmailete l 67Z-910 191 Al-72.4 IEIAI 20 T at finer sill- Escensive ana:lling at a Att Al-73.4 LE1*Al c1Je garticles, laanap of 42 atZ. P.I.E. ompletaz!. Irrallatlost u) ; terminutest et a lammy of 60 at%. EE-911 41 A141.5 LEIAl 20 Fiseer SilicJde pert- Irrellaticia naplete at 21 1cles. Drill.at 39 asal 57 at1. S7
.h:lects in clattisu .
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}
} MTRs - EXPERIENCE WITH MODERNISATION AND UPCRADING 1
J. A. Collis-Smith GEC Energy Systems Limited Leicester LE8 3LH England ABSTRACT Most reactors for research, training and irradiation were built in the 1950s. They were designed :onservatively and have operated safely for more than 20 ye s.r s . Versatility and conservatism in the original designs has allowed the flexibility to extend both flux levels and experimental facilities. Pressures for modernisation and renewal come from changing
, experimental needs, replacement of obsolete or vorn components and from safety requirements. In addition, there is currently i political pressure on operators to reduce the enrichment of their j fuel to less than 20% and, because there is as yet no plug-in I replacement fuel available for most reactors, this change in itself
! requires a re-appraisal of the physics, thermal hydraulic j performance and safety of the reactors, in addition to the l development of new fuel materials.
l This paper highlights some of the more significant aspects of a renewal or modernisation programme, using examples drawn from ten 3 years' experience in this field. l . t f INTRODUCTION , The United States representatives at the first meeting of the IAEA Working Group on Materials Testing Reactor (MTR) renevel in 1973 (1) formally introduced the proposed Reduced Enrichment for Research and Test Reactors (RERTR) programme to I
i l - ) i ) ! 2 operators in Europe. Since that time, a great deal of effort and
, ingenuity has been expended in two principal areas In improving techniques for calculation of core physics and thermal hydraulic performance, and in the development and proving of higher fuel meat l
densities. An idea of the scale of expenditure on the RERTR programme so l far was given recently in the USA (2). It is 40 million dollars. l The IAEA lists contain around 320 research reactors, distributed in - l l 1 53 countries worldwide. While many of these reactors originated in i designs from the USA, the original designers and vendors are no longer in the training, research and test reactor business. Training, research and test reactors provide a wide range of
' services to society and these services evolve as training and research progranmes develop. Concerns have been expressed (3) that I
future needs for neutron research with reactors will cuestrip l . performance in these areas. These needs can be met, given l . appropriate funding, by a range of pulsed beam and reactor neutron l sources. The facilities may be provided by new installations, by upgraded and improved existing reactors and beam halls, and by a supporting range of lower power facilities combining the training and supporting research roles. The costs of new high-performance I facilities are considerable and these are likely only to proceed as a part of a national or international collaboration programme. There is, therefore, a great incentive to capitalise on the large investment an existing facility represents, and to modernise and upgrade existing equipment in order to keep pace with new experimental requirements, new safety standards or economic i pressures. I SAFETY AND LICENSING ' i The national regulatory and licensing authorities were formed early in the power reactor programmes to oversee and regulate the design, construction and operation of nuclear power plants. The early TRTRs were designed and built before the evolution of recognised safety rules and codes. This generally means that when ' improvements are being considered, either the logic devised to cover power reactors is invoked, or specific changes are argued out on a case-by-case basis. In fact, power reactors and research reactors are fundamentally different, both in purpose and in the mode of operation. A power reactor has a single use, runs to a . pre-set well-proven routine and is staf fed accordingly. An HTR is intended to provide a flexible irradiation source and operates within certain limiting guidelines (eg excess reactivity), but will often require frequent changes both in core configuration and in the type and distribution of irradiation loops. The early
r 3 4 sarety reports for MTRs have usually been supplemented, expanded or redraf ted as changes in power level or experimental programmes have been made. These changes have been initiated by the owner / operators and approved by the locai Satety Committee and the appropriate licensing body. Petten Studsvik and Delft The scale of the technical and licensing task can be put into perspective by the HTR Petten vessel replacement programme. The new vessel has been treated as a "repiacement component". Nevertheless, in the course of completing the design, manuf acture
! and installation tasks, five informal and nineteen tormal licensing meetings were held. Arising from these meetings, more than twenty-six additional tasks were identitied. These tasks ranged from the l estabiishment of the upset and faulted conditions appropriate to the new vessel design (pressure spikes, reactor transients, loss of
! vessel cooling etc) to the establishment of the material properties of the aluminium alloy chosen for the fabrication of the vessel. In addition, both the Swedish (Studsvik HFR) and the Dutch (Petten HFR) licensing authorities decided that although their i replacement vessels were designed and fabricated to the intent at the ASME code, Section III, an external restraint structure should be provided, so that (in the event of a hypothetical, catastrophic failure of the primary vessel) both shutdown capability and coolable geometry could be assured. In the Petten example the external restraint structure requirement (evolved af ter the main design stage during the early fabrication phase of the new vessel) conflicted with the access
' requirements to the two poolside facilities (PSFs) and the new begg tubes. It was eventually integrated into the provisions for N control outside the core box. This provided a composite water-cooiing system for the PSF walls, beam tube ends and activated poolwater control, and a structure capable or limiting relative movement between the top and bottom of the core assembly to less than 2 mm even with spontar.aous disappearance of the core box itself.
The instrumentation refit completed at Delft (IR1, Holland) followed a rather easier course. This project started from a , preliminary functional specification prepared by the customer and agreed with the Licensing Authority. During a period of negotiation, the conceptual design was evolved and frozen prior to
- the start of detail design and manufacture, allowing project completion essentially to the original programme and buJget.
l l l
4 For HFR Studsvik, the benefits of experience gained at Petten are clearly appareat. The vessel failure logic was accepted and agreed before design commenced. The replacement vessel project started with a short feasibility study which established six main activities -
- 1. Improve the irradittion facilities.
- 2. Review the existing vessel for changes in construction required to meet the code intent for fabrication and inspection.
- 3. 1.ook for vessel options to ease assembly or machining problems.
- 4. Provide material specification to meet design requirements.
- 5. Justify code class selection for the new vessel.
- 6. Provide budget and schedule for vessel design or supply.
PERFORMANCE IMPROVEMENTS The High Flux Reactors at both Petten and Studsvik were adapted from an Oak Ridge reactor in the (ISA. Each of these reactors evolved from the original design to suit the particular experimental needs of the operators at the time. In addition. there were significant variations in both the licensing procedures and the requirements in the two cases. In the recent refits, now
! completed at both sites, the designs have diverged further.
+
In both cases it has proved practicable to remove the i obsolete therral column equipment and fit new twin-beamtube 1 assemblies. Petten is capable of irradiating long vertical rigs close to the core box on either side of the core in the poolside facilities, while the remaining faces accommodate new beameube assemblies. Studsvik has two poolside faces available, similar in scale to the original single-face arrangement, while two integral heavy water tanks provide thermal neutron facilities on the other two - sides, either via beameubes or vertical irradiation tubes accessible from the pool top. . At both Studsvik and Petten the vessel designs and their cooling systems have provided a significant design margin in reactor power and flux over the current power rating. This margin l l i i
5 will allow around 20% increase in flux for operational flexibility in the future. IRI Delft All the control and instrumentation equipment for the 2W Delft University reactor has been completely replaced to modern safety standards. The general concent is to orotect the reactor with conventional instrumentation and hardware alarms, but in aidition to provide the flexibility and safeguards of fered by a
'Ebs'pilter system.
, Performance has been improved in a number of ways. The automatic computer controller can maintain any demanded power level between 1 W and 2 W by varying the position of any one of six black control rods, using a three-term control function. The new control rod mechanism and magnet can be positioned to an accuracy of about 0.05 mm. and rod position is known to 1 part in 10,000 using an optical encoder driven from the output shaft of the control rod drive gearbox. These improvements have made significant reductions in unwanted flux oscillations possible.
A new, additional powered-shutdown facility has been provided via a dedicated microcomputer which carries out surveillance of a number of reactor parameters. If these parameters vary outside pre-set limits, the reactor in shutdown by a controlled drive-in of
, the rods.
The nucleonic equipment includes the followingt-
- 1. Four auto-reset neutron tiux fully failsate shutdown channels.
- 2. Four pooi gamma flux failsafe channels.
- 3. One stack off-gas activity channel.
I 4 One pool fission product channel to monitor delayed neutrons. S. Three area gamma switched range monitors.
- 6. Wide range log / linear channel covering startup to tuli power -
using pulse / Campbell mode with automatic changeover.
' Other improvements include the storage, retention ~and presentation of 64 channels of analogue data and 320 digital status signais. The cceputer calculates reactivity using a single-point algorithm based on data received trom the wide-range logarithmic channel. Reactor power is calculated from neutron flux and I
C i l
6 corrections ge made to compensate for poisoning by reference to a calibrated N gamma channel at the cooling water outlet. , l
! The process instrumentation uses commercial conditioning equipment where appropriate and failsafe components where a prime safety function is involved.
The computer installation consists a of PDP 11/34 minicomputer, input keyboard, printer / plotter, colour VDU, dual hard disks, magnetic tape unit and video hard copier. All plant interfacing is done via the international standard CAMAC system using a dual-crate interface, which, besides carrying out all input / output interchanges, also has a LSI-11 microcomputer resident in one module which carries out a secondary safety function. This installation scans all plant data every 100 ms, provides the reactor auto-control function, updates plant data files on disk and provides the power shutdown via the LSI-11, even in the event of PDP 11/34 failure. Rod drop timing is achieved automatically and simultaneously on all control rods using a scaler in the CAMAC crate. Plant status loss of all parameters are provided on the printer / plotter on a scheduled basis or by operator demand. Sof tware alarms are carried out on all plant parameters and these are printed out and/or passed to alarm annunciators as appropriate. The memory stored on disk covers three ranges:- Scan interval Elapsed time stored 100 as 2 mins 10 s 30 mins j 30 mins 7 days This data is available to be accessed by the graphics display and/or the printout, as requested by the operator. In addition, in the event of a reactor trip, two further files are updated. These files are termed " saved files" and include the 100 as and 10 e data. These saved files are transferred on-line to magnetic tape for further post-incident analysis by another computer or for simple archiving. They can also be accessed by the PDP 11/34 for i dis 6ay on the colour VDU if required, providing the operator with iradiate visual display as a function of time. The graphics display consists of a Ramtek controller ' and colour monitor which provides the main operator / machine interface and allows mimics and trend displays to be easily generated.
F ! 7 l Computer checka are maintained by two watchdog systems. The first is contained in the LSI-11 microcomputer which monitors the PDP 11/34. If the PDP 11/34 fails, the watchdog switches the automatic power control to manual and informs the operator. The reactor continues to operate normally with the automatic rundown system and its conventional instrumentation fully operational. The second watchdog is a CAMAC hardware one and this checks the operation cf the LSI-11. If this fails, the automatic shutdown facility is vetoed and the operator informed via an alarm annunciator. The sof tware watchdog is also initiated, putting the reactor on manual. The reactor continues to operate using its conventional instrumentation. CONCLUSIONS Nearly ten years' work in the upgrading and modernisation of reactors has proved that renewal programmes can be successfully and economically undertaken. With a planned approach, with reasonable time allowed for the necessary scheme optimisation and safety logic factors to crystal 11se, quite extensive changes can be completed and meet pre-determined programmes and budgets. The man-Rem burdens in dismantling and removing obsolete equipment can be well within accepted international limits, and reconstruction and commissioning has also been shown to be safe. The costs of major refits involving removal and replacement of the entire reactor block can be economically justified against a new greenfield installation. Examples on this scale are under consideration in Germany, Japan, the United Kingdom and the United States. Previous experience can be invaluable. This can be illustrated by comparison between the programmes at Petten and Studsvik. Petten was planned for 41s years, after a pre-design phase of some 4 years (4): the actual programme was 6 years. Studsvik was planned for 3 years and took 31/4 years. Technical experience in safety and licensing, reactor performance, metallurgy and non-destructive testing are all - essential ingredients towards completdon of a successful project. Some or all of these aspects Iow-enrichment-uranium convarsinns a tr a n whmen na sinnificant vill arisa in =nme
) .
nges in experimental facilities are envisaged. 3
,o .
L 8 REFERENCES
- 1. Advisory C roup Meeting on Reactor Renewal and Upgrading Programs, IAEA, Vienna, May 23-26, 1918.
- 2. 12th Biennial Conference on Reactor Operating Experience, American Nuclear Society. Williamsburg. 4-7 August 1985.
(Transaction Supplement 2. Volume 49)
- 3. Major Facilities for Materials Research and Related Disciplines, Committee on Physical Sciences, Mathematics and Resources. National Research Council. Washington DC.
4 M. R. Cundy and P von der Hardt. "HFR Vessel Replacement at Petten", Nuclear Europe 5/1985 p. 39-42. 1 f l
\
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f L . l I l , t l I I ! i i l l , I I A JAytTT ' ANALY3:3 0F TME RESEARCH REACTOR 1RR-3 !
! I l
*aikan 1ARAM1. Maassai H:RANO. Yosaire AJAH:. !
l Atsuo XCH3AXA and Nocua41 *KN:3H Japan Atomic Energy Researta Institute (JAZRI) . Tokai-sura, Ibarski-ken, Japan 319-11 I t i i Austeset The upgrading program of tae researca reactor JRR-3 is Detag carried forward to a 1133: water sodersted and j cooled, nesty water reflected 20 NW pool-type reactor
, with a 205 LIU plate-type fuel. This paper provides tae c analysis done for- two of tae design basis events, a pipe-line treak at :Re heavy water icop in the pool and a pipeline treak at tse primary coolant loop. In 13e I former event. tae calculation snowed tast 13e transient safely :an come to an end ty sonitoring taersal power ;
tastead of neutron flux. In :se latter event. tae analy- : sia predicted taat tae peak fuel resperature and tae sinimus ONS ratio can seet the design tasts criteria, ; although a sudden incrosse of tae fuel temperature and a steep decrease of :ne CNB ratio would occur at tae flow reversal. [
! INTRCDUCTICN -
[ i The researca reactor IRA-) at IAIRT is a sort taan 20 yeare 1 old, 10 MW Sanx-type reactor. The upgrading program at tae *RR ) { is now being carried terward to tacrease neu:ron tiux and .'moreve i utility capanill tes. The new reactor is 1esigned is te a ligat I water sedersted and :coled reactor vita a 201 '.ZU plate-ty;e fuel. l The DgG tank surrounds tse ore Jylindrically (Fig.11. The reac. ! or power is increased :o 20 MW. The primary :colant f13ws in tae [
. i I
h ! l
I l l 1 3 1 \ core downwards in the forced :irculation and upwards in the natural circulation trough a natural circulation valve installed between
- the lower plenum and the reactor pool. The primary cooling loop is equipped with the sypnon Dreak valves. in order to prevent all the pool water from seing discharged through the erstk. When the
;ool water level goes down to the syphon treak level the reactor pool can be isolated from the primary cooling loop.
The neutronic and thermal-hydraulic design of the reactor and the safety design at thenormalstaagystateoe{ationweredis= cussed in detail in the previous RERTR meeting 'I2I. In the meantime, the safety analyses of the design Basis events at an abnormal operational transient and accident were done in the licensinE procedure (3)*(d) . Major design Basis criteria for an abnormal operational transient are that the sinimum CNB ratto shall not be less than 1 5 and the maaimum fuel temperature shall be less than 400 C. This paper presents two of the design basis events, i.e. a pipeline Dreak at the heavy water icop in the pool and a pipeline break at the primary coolant loop. The analyses were done using the computational codes, EUREKA-2 and TXY E-P, developed at JAER:. The large pipe Dreak won't to expected to occur in a research reactor which is operated under the conditions of low pressure and low temperature. We made an assumption in the analyses of postu-
. lated piping f ailures that the maximum break area is Ot/4 The assumption is accepted for moderate energy fluid systems in a light 4 waterr7actorbythereviewin'StandardReviewPlan*ty U.S.NRC Si, l
]
II A BRIEF CCCE DESCRIPT CN USED IN THE ANALYSE 3 1 (1) EUREKA-2 The EURIKA-2 code can predict the course and consequence of a ! reactivity addition accident in a nuclear power plant ( 6), ;3gg code provides a coupled thermal-hydrodynaste and point kinetics capability. The feedDack reactivities are evaluated by a Lapor- ' tance weighted sum of the contribution in each region of the core. (2) THYDE-P The TRYDE-P code can analyte a thermal-hydraulic transient and l accident in a nuclear power plantIU. This code applies a one- I dimensional node and junction method. EAtensive code verification studies have LOFT facility ) . gencarriedout for the experimental results of the l l (3) Modification of the Codee The code zodification was made for the pacxage of the heat transfer correlations and the CNB heat flux correlations, which are applicaele to noth downflow and upflow in a plate-type fuel of the JRA-3reactorundertheconditionsoflowpressureandlowtemgera-ture. Tacle 1 shows the correlation set used in the analyses ( 3 ) l l l l l l
i ( 1 3 i
- AN ANALYTICAL Rg3 ULT AND :3CU33:0N i
; 1. A Pipeline Break at the Heavy Water Loop in the Pool 1
We'll make a focus on the accident that the pipeline areax of the heavy water cooling loop would occur in the pool (Fig. 11 This can lead to skew of neutron flux aue to dilution of the heavy water, which is shown in Fig. 2. The figure shows the thange of the thermal neutron flux in relation to the D 0 :oncentration in the heavy water tank. 11 dilution of the heavy water manes the neutron flux decrease relatively by ADout 201 at the neutron 1 counter which is placed just outside the tank. When the reactor is in the automatic control full power operation, this skew of neutron flux can initiate control rod withdrawal and insert a reactivity in the core. The other reactivity addition accidents e.g. uncon- ; trolled control rod withdrawal can be suppressed $y the reactor scraa signal from the neutron counter. In this event, however. the neutron counter will not work to indicate the correct reactor ; power. The reactor scraa signal from an alternative parameter to neutron flux is necessary to cepe with the transient. 1 Fig. 1 shows the transients of the reactor power. the inserted, total reactivity and the D:0 concentration of the tank. ' The major contribution to a feedbacx reactivity, which is the dit-forence of the inserted and total reactivity, is the moderator
, ;; temperature effect. t. however, doesn't have a significant
- 3 effect on the transient. The reactor is suppressed by the control
}
rod insertion in the core. It takes longer time to shut down the reactor in response to the monitoring of temperature compared with that of neutron flux. The scras delay time is important in order to evaluate whether the safety criteria can te set. Fig. 4 shows the sensitivity calculation on scram delay time, sven tt the telay i time of 16 sec, the sinLaus CNB ratto is calculated to be ADout 1.d ~ and the sazimum fuel temperature is 119 C. 16 seconds of the scram , delay time is acceptable for the hardware design of the monitoring > of temperature. The analytical result shows that the scram signal i from the monitoring of thermal power can control the transient
; safely.
i l i 2. A Pipeline Break at the Primary Coolant Loop i We'll show the analytical result of the transient under the pessimistic sesumption that the break would have the sasimum area I and occur at the suction line of the main pump. the main pump ; vould therefore fail to work due to pump degradation secause of the suction of air through the Dresx into the primary coolant. This assumption would sake the transient just after the event initiation severe. t Fig. $ shows the changes of the pool water level and the core i i flow after the event segins until the pool isolation. While the l core flow decreases fue to pump degradation. the primary coolant i i i h e l
i i l l i 4 l l l discharges through the treak. The pool water level starts to go down. The reactor scram taxes place when the primary coolant ft:w i decreases colow a specified value (55 of the normal flow ratei. l After the :castdown of the main coolant pumps is completed. the forced cirtulation core cooling Oy the auxiliary pump continues ! until the water pool isolation. (h t i the natural circulation valve is opened, a part of the primary scolant flow bypasses the : ore througa the natural circulation valve and the core flow secreases l slightly. After the peol water level decreases to the level of the
' syphon treax line, it is kept constant and the decay heat is removed by the natural circulation cooling.
i The event sequence has two important thermal-hydraulic ! Denaviors in Fig. $ from the safety point of view. (1) Just after the event initiation I Fig. 6 shows the transient behaviors of the power. the core I flow and the fuel surface heat flua at the hot spot until e seconds
- after the transient begins. Fig. 7 shows the fuel temperature at the not spot and the DNB ratio. The transient is dominated Dy the competitive process of the decrease in the core flow against the decrease of the decay power after the reactor scram. The minimum ON8 ratio. 1.7. and the highest fuel temperature.120 C. of the whole transient were calculated here at the beginning of the
) control rod insertion. ;
t 1 l [ (2) After the reactor pool isolation
- Fig. 8 shows the core flow the fuel surface heat flua at the b hot spot at the flow reversal. Fig. ? shows the DNB ratio and the fuel temperature at the not spot. Before the pool isolation occurs the fuel temperature reaches a certain velue cecause all the decay heat is removed Dy the core flow, the fuel temperature starts to increase just after the reactor pool isolation tocause i
of the stored energy in the fuel and has the peax value. 30 C. After the core flow reversal, the fuel temperature reaches another i steady value in the natural circulation. The pHU heat flua i decreaseswiththeflowcoastdowntothevaluecalculatedoyq$ND.)
, in sq. (2) in Table 1 and stays at the value turing the short flow reversal time. The surface heat flus suddenly tecreases with the heat transfer coefficient decreasing due to the flow coastdown.
' The heat transfer coefficient reaches the lowest level calculated By Nu = 4 and stays at the value during the flow reversal. The l sinimum CNB ratio was calculated to to 3.1. This shows that the '
decay power level is sufficiently low at the flow reversal.
. i IV CONCLUDING AgMARKJ ,
This paper presented the analyses :arried out for two of the l tesign masis events of the JRR-3 reactor. The analyses showed the i i l i
\ . i i l 5 l fol'. awing results. (1) The transient initiated by skew of neutron flux sue to dilution of heavy water :an ne suppressed safely my the [ reactor scram signal from the sonitor of thermal power instead of neutron fluz. (2) Although a sudden increase of the fuel temperature and a steep decrease of the ON8 ratio would secur at the flow reversal. the peak fuel temperature and tne minimum ON8 ratto :an teet
. the design Dasis criteria.
References (1) H. Tsuruta. H. Ichikawa and J. Iwasakt. *Neutronics Design of l Upgrated JRR-] Research Reactor
- in Proceedings of the l , International Meeting on Development. Patrication and Application of Reduced Enrichment Fuels for Research and Test Reactors. Argonne. Illinois USA, October 15-18. 1984 i (2) Y. Sudo. H. Ando. H. Ikawa and N. Chnishi. ' Core Therso- -
l hydraulic Cesign with LEU Fuels for Upgrade' Research Reactor. ! JRR-]" in Proceedings of the Ir.ternational Meeting on Development. Fabrication and Application of Reduced Enrichment Fuels for Research and Test Reactors. Argonne. Illinois. USA. Octocer 15-1d. (1984).
. (3) 7. Haraat. M. Uesura and N. Ohnishi. " Reactivity Initiated
] Accidert Analyses for the Safety Assessent of Upgraded IRR-)*.
(in Japanese) JAERI-M 94-142. (1984).
. (4) M. Mirano and Y. Sudo.
- Analytical Study on thermal-Hydrauli:
4 Senavior of the Transient fres Forced Circulation to Natural Circulation in the Japan Research Reactor-]", sutaitted to Nucl. Sci. Technol. (5) KURIC-0800. U3NRC STANDARD REVIEW PLAN, July 1981. (6) N. Onnishi. T. Haram1. H. Hirose and M. Veaura. 'A Ccaputer Code for the Reactivity Accident Analysis in,a Water Cooled Reactor *, (in Japanese). JAIRI-M $4-074 (1984). (7) Y. Asahi.
- Description of THYDE-P (Pet 11minary Report of Methods and Models". JAERI-M 7751. (1979).
(8) M. Mirano. " Analysis of LOFT L)-6/L8-1 with THYOE-P (C3NI International Standard Problem No.11 and THYDE-P 3 ample Calculation Run 60)*, JAERI-M 92-026. (1982). (9) Y. Sudo. H. !kawa. H. Mirano and M. Chnishi.
- Development of Heat Transfer Package for JRR-] Therschydrodynamic Analysis".
(in Japanese). JAIR:-1 S4-066. (1784). D
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i j MINIMUM CLADDING THICKNESS OF MATKRIAL TEST j REACTOR FUEL PLATES i j t Jurgen Deckers I
~
Technischer Uberwachungs-Verein Rheinland e.V. , J K51n, F.R. Germany I For research reactors in the Federal Republic of Germany the average thickness of the cladding of i fuel plates is specified by German manufacturers at 0.35'mm or, as a function of the plate' arrange-ment, at 0.36 to 0.41 mm. The minimum thickness may not fall below 0.25 mm. It is intended for the future to specify a minimum value of 0.30 mm for j the average cladding thickness and a value of j 0.2 mm for the minimum individual thickness at any
; location. This is in line with world-wide attempts j of standardization for these fuel elements. This standardization follows the successful use in the Oak Ridge HFIR research reactor of fuel elements
! with a specified minimum cladding thickness of 0.2 mm. The main objective of a study performed by TUV Rheinland was an assessment of the minimum cladding thickness from a safety point of view.
The investigation covers the quality assurance measures during fuel plate manufacture, the irra-diation anc corrosion behaviour of the fusi plates and fuel eletent damages. The presentation will give some detailed infor-mation regard.ing these aspects. The results of the investigation show that there are no objections against the reduction of the average cladding thickness to 0.30 =m and of the - , minimum cladding thickness to 0.20 mm for the fuels UAl and U 3g 0 as used in the irradiation experimen s
> 7,l .=
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- 1. IUTRODUCTION l This presenta, tion reports about an assessment of ,
the minimum cladding thicP. ness from a safety point of view. An important requirement for safe and undisturbed reactor operation and experimentation is to prevent
; a release of fission products from the fuel element and, as a consequence, an activation of the reactor cooling circuit. It is, therefore, necessary that ,
the integrity of the fuel plate cladding remains intact for as long as the fuel element stays in the reactor. As most of the fission products developing . during operation are retained in the matrix of the dispersion fuel, the main function of the cladding
,. ; under normal operating conditions is to prevent a
' contamination of the reactor coolant with uranium.
j This is a requirement which.must be fulfilled also , under high irradiation rates. ' j In addition, the cladding must be in a position to i prevent any larger release of fission products into
- the cooling circuit in the case of incidents with
; extreme temperatures which, at the fuel plate sur- [
face,-can reach to the solidus line of the cladding material. In the course of the last fifteen years it became
- evident, that the treatment of the fuel plate
.j surface is of major importance for the performance j of the fuel elements under normal operating and '
accidental conditions. This aspect will be briefly discussed in connection with fuel element damages.
- 1. FUEL PLATE MANUFACTURE l The fuel plates are manufactured using the frame technique. Figure 1 shows the subsequent steps to r produce the fuel " pictures", i e., the unciadded .
! rectangular fuel plates. These " pictures" are put '
1 into aluminium frames and covered on both sides by aluminium sheets. The sheets are secured to the frame by spot welds and the entire assembly is then rolled in a number of steps at temperatures around 400 'C. During the rolling process a per-
- fact intermetallic bond between the meat and the cladding is obtained. The adjustment to final fuel zone and cladding geometry is done by cold rolling.
The rolled fuel plates are then aligned, X-rayed, cut to final dimensions according to the fuel zone, f
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and surface treated.
. The following examinations are of impor,tance for the quality of the cladding layer with a view to further manufacturing and to the use in the reactor:
Verification of the specified fuel powder grain size and size distribution, Verification of homogeneous uranium distri-bution in the fuel plate, Verification of cladding and meat thickness Visual examination of the fuel plate surface, and Blister testing. . This test is of particular importance, as it assures a perfect intermetallic bond between the fuel matrix and the cladding under extreme conditions as might occur during an accident. The blister test is also always part of post irradiation examinations.
- 3. IRRADIATION EXPERIMENTS Irradiation experiments were performed in national programs in order to test the new fuels with high .
uranium density in reactor service. In table 1 the irradiation experiments are listed for NUKEM fuel j elements and fuel plates. The post irradiation, examinations performed cover the following: 4 burn-up rate obtained during irradiation, v.isual inspection of the fuel plates and elements, changes of tuel plate dimensions during irradiation, fuel and cladding layer microstructure, corrosion performance of the cladding, oxide formation during irradiation,
- blister test, and release of volatile fissile products.
In the following discussion only NUKEM manufactured fuel-plates and elements are considered and among these only those, for which the post irradiation experiments have been entirely completed.
Twenty NUKIM fuel elements with UAl, and U 0 fuel enriched from 20 % U-235 to 39,70 U'3235 we$e 3 irradi-ated in the Oak Ridge reactor and in the High Flux Reactor in Petten between May 1981 and February 1904. The results can be summarized as follows: The burn-up rates obtained for the plates irrgdia-
-ted in Oak Ridge lie between 1.2 and 1.9 x 10
fisssions per em) corresponding to a U-235 burn-up
.of more than 90 %. The average U-235 burn-up rates obtained for the plates irradi,ated in Petten lie between 45 % and 74 %. .
All test plates showed a slight oxide layer across the entire cladding and longitudinal streaks caused by the coolant flow. The contours of the meat could clearly be identified. No defects could be detected at the plate surfaces except for a few scratches and slight deflections caused by dismantling of the modules. Plate thickness was measured for all plates. Measurement were taken either at 3, 4 and 12 locations respectively or across the entire plate length and along three tracks 20 mm apart from each other. Comparison with the pre-irradiation measurements led to the following results: The plate volume had increased in all cases. The amounts of increase were low and lay
- for the lengths between 0.05 and 0.17 mm,
- for the widths between 0 and 0.07 mm, and
- for the thicknesses between 0.023 and 0.15 mm, always with reference to the pre-irradiation values.
Immersion density was determined after the oxide layer had almost entirely been removed. The results corr.clate well with the dimensional measurements. Some of the test plates were metallographically examined, two of them after blister testing. The results of these examinations can be summarized as follows: All sections show an internetallic bond bet-ween fuel matrix and cladding.
> _ _ _ _ -w
No indications of rearrangements or demixing could be detected in the fuel matrix. Pore size and frequency are normally distri-buted over the fuel matrix. No defect in the cladding layer could be detected in any of the plates. l'
.The visual inspection before and after removal of the oxide layer revealed no indication of corrosion on the test plate surface. Oxide layer was removed from irradiated and, additionally, frem unirradiated test plates. The oxide layer thickness was determined and an average increase of the oxide layer thickness by irradiation of 0.03 =m was indicated.
Blister test was performed on 10 fuel plates. Starting temperature was 350 *C. The temperature steps were 50 *C up to 450 'C and 20 'C thereafter up to the temperature at which blister formation could be detected visually or by measurement. l The blister temperatures lie between 470 *C and 550 *C. Metallographic investig'ations after blister testing show that for the plate with the_UA1 fuel the matrix surface was locally lifted up wit 5 the blister, whereas this phenomenon did not occur in the case of the U 38 0 fuel. The activity of all plates was measured using a Ge (Li) detector. The results showed, that during and after irradiation no release of fission products was detected.
- 4. CCOLING WATER QUALITY The corrosion behaviour of aluminium is among others a function of the pH value, the temperature condi-tions, and the cooling water flow. Investigations
, showed that at pH values between 5 and 7 the corrosion rate was rather low, irrespective of cooling water flow or temperature. For pH values below 5 or above 7, the corrosion rate increased with temperature and decreased with increasing cooling water flow rate.
Of stronger influence than pH value, temperature, and cooling water flow rate is a prolonged exposure to impurities in the cooling water, in particular to chlorides and heavy metals. Concentrations around 0.1 ppm are already sufficient to cause pitting corrosion in the fuel plate cladding.
- 5. FUEL ELEMENT DAMAGES Little informations is available in the literature on damage cases regarding material test reactor fuel elements. The reason for this is that a serious fuel element damage can only occur in connection with an accident and that no such accidents have become known for research reactors.
One type of fuel element damage had occurred in 1973 in several research reactors in the Federal Republic of Germany. Pitting had partly extended down to the fuel :ene and was independent of cladding thickness. Minute traces of chloride that had remained on the aluminium cladding after hot water sealing of the finished fuel plates had led to the pitting described after short pericds of use of the ruel
, elements concerned. As a consequence, the cladding material was changed (to A1 99,5 and AIMg alloys) , with the result that this damage has never occurred again.
Another type of fuel element damage occurred in a German research reactor in 1973. Investigations revealed that anomalies of the mechanical structure of the fuel' elements or of the quality of the aluminium material used could be excluded. The only deviation frem normal reactor operation concerned a considerably increased conductivity of the heavy water to above 14 uS/cm. As a consequence of the increased conductivity in the primary circuit the radiolytic decomposition of the heavy water increased. A change in the chemical and physical properties of the oxided protection layer can at wall temperatures around 100 *C not be excluded. Investigations show, that layer thickness deter-minations resulted in between 40 and 60 um for the Boehmit layer and between 360 and 390 um for the cladding material. The minimum cladding thickness reduction to be assumed is about 15 to 20 %. The maximum cladding temperatures occurring during a serious reactivity accident have been determined from experimental investigations and subsequent calculations using appropriate computer codes. The results show that below 550 *C only neglectable release occurs. Below 500 *C the fuel plates did not show any change whereas at temperatures from about 525 to 550 *C deformations occur with all plates.
-y-
- 6. CONCLUSICNS Concerning quality assurance during fuel plate manufacture a lot of characteristics have to be specified. Among others these are grain si:e distribution and uranium distribution. A further important point in the fuel plate u nanufacturing is the fuel fillment of the cladding thickness requirements. A review of the values realized is given in table 2 for NUKEM prototyp plates with depleted uranium.
Concerning irradiation behaviour of the fuel plates no peculiarities were found at the fuel plate surfaces. The amounts of the dimensional changes found for the fuel plates have no influence on the fuel element designs as been used. The dimensional changes are small. A mechanical strength of the cladding layer can also be neglected. ' The low increase of oxide layer thickness during the entire time, for which the plates stay in the reactor, plus the time until the post irradiation examinations, shows that significant effects on the performance of minimum cladding thicknesses, for which a value of 0.2 =m is intended to be specified, are not to be expected. This is supported by corresponding operating experience in HFIR
, research reactor.
The irradiation experiments and the post irradiation examinations have shown that the test plates and the fuel element plates essentially retain all fission products under normal reactor operating conditions. The investigations show also that the cladding thickness neither influences the mechanical performance of the fuel plates nor the release of fission products. Based on above mentioned results, there are no objections against the reduction of the average cladding thickness to 0.30 mm and of the minimum cladding thickness to 0.20 mm for the fuels UAl and U 0 as used in the irradiation experiments 7 38
I t I Figure 1: Fuel " Picture" Manufacture i
; Fuel powder
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m-t e Table 1: AF-Progra==e Irradiation Experiments on NUKIM Fuel Plates and Fuel Elements e e. t 3 . g - . . . . . . . - . . . . . .
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- j. ; Table 2: MPR 30 Fuel Plate Thicknesses as Determined on u ; Metallographic Specimens
, t Specimen No. Specimen Location Claddine-Thickness (nm) Meat Thickness Top Side botton Side (mm) 30-7-31 rront end, longitudinal I ='0,35 I = 0,35 I = 0,63 s = 0,0234 s = 0,0208 s = 0,0237 l
. Bacx end, longitudinal I = 0,34 I = 0,33 I = 0,65 j s = 0,0298 s = 0,0255 s = 0,0419 .
- f. Center, transverse I = 0,35 I = 0,35 I = 0,63 s = 0,0213 s = 0,0217 s = 0,0302 4
30-7-49' front end, longitudinal I = 0,35 I = 0,34 I = 0',65 l., s = 0,0287 s = 0,0204 s = 0,0393 sack end, longitudinal I = 0,34 i = 0,35 ,I = 0,64 s = 0,0255 s = 0,178 s = 0,0313
, Center, transverse I = 0,34 I = 0,36 I = 0,64
, s = 0,0221 s = 0,0204 e =*0,0333 I
! { 30-7-13 Front end, longitudinal I = 0,34 I = 0,35 I = 0,62
- r. (
f a = 0,0195 s = 0,0191 s = 0,0298 l l Back end, longitudinal I = 0,33 I = 0,34 I = 0.65 4 x = 0,0276 s = 0,0244 s = 0,0376 i Center, transverse x = 0,35 x = 0,34 x . 0,63 l ; s = 0,0186 s = 0,0181 s = 0,03c8 I 3 C 46 rront end, longitudinal I = 0,34 I = 0,34 x = 0,64 3 e = 0,0396 s = 0,0239 s = 0,0447 3ack end, longitudinal I = 0,32 I = 0,35 I = 0,65 s = 0.0431 s = 0,0'209 s = 0,0487 l ~f center, transverse i = 0,34 I = 0,34 I = 0,63 l 1 f ,s = 0,0203 s = 0,0139 s = 0,0276 l l s
- P i
i i l i t_
l P w REDUCED-REACTIVITY-SWING LIU FUEL CYCLE ANALYSES FOR EFR PETTEN , s J. R. Deen and J. L. Snelgrove { Argonne National Laboratory
, Argonne, Illinois 60439 U.S.A.
ABSTRACT The primary objective of these low enriched uranium (LEU) fuel cycle analyses was to effect at least a 33% reduction in the reactivity swing now experienced in the high enriched uranius (HEU) cycle while minimizing increases in 235U loading and power peaking. All LEU equilibrium fuel cycle esiculations were performed using either a 19- or 20-place fuel element with 0.76-en-chick seat and 0.5- or 0.6-am-thick Cd wires as burnable absorbers and 16- or 17-plate control rod fuel followere with 0.76-am-thick meet. Burnup-dependent microscopic cross sections were used for all heavy metals and fission products. A three-dimensional model was used to account for the effect of partially inserted control rode upon burnup profiles of fuel and of burnable absorbers and upon power pe king. The equilibrium cycle reactivity swing (or, equivalently control rod movement) was reduced by 50% using LEU fuel with U meat densities (4.8 Mg/m3 . INTRODUCTION The Reduced Enrichment Research and Test Reactor (RERTR) Program, the Joint Research Centre (Petten Establishment), and the Netherlands Energy Research Foundation have been engaged in a continuing joint study to determine the most suitable LEU fuel L__
2 element design for the High Flux Reactor at Petten. Additional fuel cycle optimization calculations were needed to investigate LEU fuel cycle performance in anticipation of future upgraded operation strategies and to reduce the reactivity swing by 1/3 compared to that of the reference HEU equilibrium fuel cycle with 235U loadings of 420/290 g (standard / control). Previous LEU fuel cycle analyses explored the reactivity and power peaking trends of using various LEU fuel elements in the reference fuel cycle.1 A number of XY LEU equilibrium cy-:le calculations were performed using a 19-plate element with 0.76-am-thick meat and Cd wires as burnable absorbers for the standard element and a 16-
! plate control fuel follower. The ratio of the standard to control element 235U loadings was set equal to the HEU ratio of 420
- 290
, in order to minimize power peaking in the control elements. A l description of the elements is given in Table 1. The XY fuel cycle calculations were made using " dummy" experiments for all in-core i
- radiation positions.
Also included in this paper are the results of XYZ calcula-tions of fuel cycles utilizing one 19/16-plate (standard / control) fuel element, two 20/17 plate LEU fuel elements, and the reference HEU fuel element. A coarse-mesh XYZ REBUS-3 model was used to obtain equilibrium cycle burnup distributions for use in finer-mesh beginning of equilibrium cycle (BOEC) and end of equilibrium cycle (E0EC) calculations to determine power distributions and peaking, reactivity swings, and fluxes in irradiation positions. The control element fuel management was changed from the reference
! HEU cycle pattern of loading on the average 1.5 fresh control elements per cycle to loading one fresh control element per LEU ~
cycle.
; Table 1. Description of Standard LEU Element and Control Fuel Followers ,I Number of Plates, (Std/ Cont) 19/16 20/17 Fuel Meat Composition U3Si2 -Al U3S12 -Al ' Fuel Meat Thickness, nun 0.76 0.76 Water Channel Thickness (Std/ cont), un 2.656/2.656 2 45/2.35 Ratio of LEU Standard Element l Moderator Volume to HEU Moderator Volume 1.007 0.977 Control Fuel Clad Thickness, em 0.38 0 38 Standard Inner Plate Clad Thickness, un 0.38 0.38 l Standard Outer Plate Clad Thickness, um 0.57 0.57 Fuel Meat Width, em 6.315/5.958 6.315/5.99 60.0 Fuel Meat Length, em 60.0 Ratio of 235U Loadings (Std/ Cont) 0.69 0.69 t
3 i l EQUILIBRIUM FUEL CYCLE MODELS All REBUS-32 equilibritus fuel cycle calculations were made 4 ' with burnup-dependent microscopic cross sections for all heavy metals and fission products. Mid-life cross sections were used for all other asterials. Cross sections were calculated using
. identical models in EPRI-CELL.3 Very-high-burnup microscopic cross sections were required for the longer-residence-time control fuel followers. .
The fuel management strategy for simulating the current i i operating cycles has not been changed for LEU cycles. The XY ' diffusion-theory mesh for all fuel cycle calculations was 92 x
- 81. Each in-core position was modelled by a 4 x 4 mesh grid. The [
inner (fuel) sone of each standard fuel element was assigned a 2 x 4 mesh and each side plate zone a 1 x 4 mesh. The fuel ~one of i the control follower was assigned a 2 x 2 mesh. The XY mesh p structure was 61 x 55 x 24. The reduction in the planar mesh
. occurred primarily in the ex-core regions by increasing the mesh intervals and by more material homogenization. Acceptable agreement was obtained between the detailed-XY-model (92 x 81) planar in-core flux solution and the coarse-mesh (61 x 55) planar flux solution from the XYZ aodel. Each standard and control follower fuel element was divided into eight axial burnup sones.
Control rods were all banked at the estimated average position during the entire cycle. Beam tubes were not modelled owing to convergence difficulties encountered. The core consisted of 528 separate burnable absorber zones, 264 standard fuel zones, and 96 control fuel follower zones. A black boundary condition (j/$ = i , 0.4678) was imposed at the surface of the control rod absorber material for the thermal group (gp. 5). ln The XY model reactivity swing is based upon REBUS-3 k,gg results for BOEC and E0EC. The BOEC calculation in REBUS-3 assumes equilibrium Xe and Se only in all previously irradiated fuel. All XYZ model reactivity swing results are based upon a 9 separate calculation of excess reactivities at BOEC and E0EC with equilibrius Xe and sa concentration in all fuel assemblies. I Therefore the reactivity swing (for XYZ calculations) is defined 1 to be 4 swing = (kBOEC - kE0EC) 6 k BO all rods fully withdrawn and'EC*kE0EC, equilibriuswhere Xe andk,gg's Sa inare allfor i fuel. BENCHMARK COMPARISONS OF DIFFUSION MODEL WITH VIM-MONTE CARLO Previous LEU standard fuel element designs using 0.4-am-0D Cd < wires and lower 235 U loadings have been benchmarked for an '
- infinite lattice.1 With higher 235 U loadings and larger cd wire ; '
l 1 j .
e 9 4 diameters, it was necessary to make additional benchmark compari-sons to determine the adequacy of the RE8US-3 model to predict ab-sorption rates and reactivity for a fresh element in an infinite lattice as well as near partially inserted control rods and experiments. The VIM 4 model used for the benchmark comparison was a detailed model of each individual fuel seat, clad, moderator, side-plate, and Cd wire. Each element was run for 100,000 neutron histories in order to obtain accurate (<*2%) power shapes and Cd absorption rates in these small volumes. The first comparison was made for a 470-g 235g g9.plag, element with a Cd wire of 0.5-am OD located at each edge of each
. fuel plate. The de 235U was 5.2%
higher in the DIF3Dgletion rate inof the Cd VIM relative to for model than model the beginning of life (BOL) condition in an infinite lattice environment. Good agreement in the reactivity of the lattice was also achieved. The DIF3D model k, was within two standard deviations of the VIM k . The same conclusions were also reached for the higher-loaded 600-g 235 U element case with 38 Cd wires of 0.6-am OD. The depletion of Cd relative to 235 U was 3.2% higher in the DIF3D model than in VIM. However, the DIF3D ( was lower than the VIM ( by 0.0066 A(. In order to validate the depletion rate of 10 3 in the top of the core adjacent to control rods in the XYZ REBUS nodel for the NEU core, a four-element XYZ representation of the core center was
', modelled in DIF3D and compared with VIM-Monte Carlo. One quadrant had a control element positioned at 51.7 en while two other ad-jicent quadrants were loaded with fresh HEU elements containing 1g 10 B and 420 g 235U. The remaining quadrant contained a
" dummy" experiment. The same mesh and homogenization of materials were used in the DIF3D model as were used in the REBUS-3 model. A black boundary condition in group 5 was used to sisuiste the absorption in the Cd control rod. Zero currant boundary condi-tions were used at the X and Y boundaries 'of the four in-core positions, and no-return-current conditions were used at the axial reflector boundaries. The result was that the total depletion rate for 10 B vos 3 4% higher in DIF3D than in VIM. At axial nodes near the top of the core, the depletion rate was 5 to 10% higher adjacent to the control rod and ~5% lower in the quadrant adjacent 1 to the experiment. The k, for the entire problem was 1.3027 in DIF3D and 1 3002
- 0.0031 in VIM. Therefore, the 10 B depletion rate is being calculated quite well by diffusion theory, and it does not appear to be necessary in this model to generate different burnable poison cross sections for different neighboring elements or experiments.
A LEU comparison for a similar configuration was also made using the 20/17-plate fuel element loaded with 525/365 g 235gf element and 20 Cd wires of 0.5 mm-0D per sideplate. The control ,
- 4. %
W-a w -w
5 rod position was changed to 58.15 cm, and the experiment was loaded with a smeared concentration of 10 5 (N(10B ) = 4.0 x 104 atoms / barn-es) to simulate the presence of an " average" experi-ment. The results indicated an underprediction of Cd wire depletion rate by 1 to 2% in the upper 15 cm of the core control-led by the cd rods. Conversely, a 8 to 10% overprediction in the Cd depletion rate was observed in the portion of the core opposite the control fuel followers. The total Cd wire depletion rate was 9% more than the VIM depletion rate. RESULTS OF XY FUEL CYCLE CALCULATIONS Several different fuel cycle options were examined with RZBUS-3 using XY geometry. In these no attempt was made to change the fuel loading schese for the reference 6/15 (standard element / control element) reloads per cycle or the core positions of partially depleted fuel. The first two cases presented in Table 2 are for the reference HEU fuel cycle and for a reduced control element batch size. The reference HEU cycle is Case #1. It can be compared with the proposed reduction in control element loading to one per cycle (Case #2). The reactivity penalty would be ~940 pcm without any changes to the fuel management strategy. The discharge burnup of the fuel follower would increase signi-ficantly. The burnup values for the fuel follower are probably higher in the IT REBUS-3 model than would be predicted by a full XYZ calculation with control movements modelled during the cycle. However, the predicted percentage increases in fuel follower burnup should be representative of all types of cycles having similar rod movements. In order to achieve the same reactivity swing as with the reference HEU fuel cycle. 475/328-g 2350 (standard / control) elements would be needed with 46 Cd wires of 0.4-am 00 loaded into each fresh standard element (Case #3). More than 38 wires are needed to previde the additional control poison at BOEC without increasing significantly the amount at E0EC. However, more than 46 wires would be needed (perhaps eight more) to reduce the reactivity swing the additional ~500 pcs needed to achieve the design objective of a 1/3 reduction in reactivity swing from that of the reference cycle. As an alternative to using more 0.4-se OD wires, 38 Cd wires of 0.5-m OD vere used with a 500/345-g element loading (Case #4). This case does achieve the desired reactivity swing and EOEC k with only a 1% higher peak power at E0EC than with the use of ,gg 0.4-am OD wires (Case #3). The peak power density increases 10% during the cycle because the power and flux shift as the Cd depletes. This power peak increase might be minimized by altera-tion of the fuel shuf fling scheme when using larger-diameter Cd wires. However, such changes might result in loss of reactivity and/or fast fluz in the irradiation positions.
6 thble 2. Owartem of fuls hterance nel cycle mW IHlace im mal enlam mi-1-ad wth om Ir w.1 anferwu:n Chas #1 72 f3 A #5 M mv wo tm tm tm tm c3cle 1muth. dare M & A 3 & A
. ht weycla 6/1.5 6/1 6/1.5 6/1J 5/t.5 5/1 2s % 1and g/e1. 430/290 4J0/290 475/3 3 300/D5 530/366 2 0/366
# places /elammt 3 /19 D/19 19/M 19/ 5 19/16 19/16 Oimnrul gap, as 0 218 0.218 0.36 0.3b6 0.36 0.36 lest p,. g/ce' 1 4 16/0.895 141&/0.895 4.40/3 JO 4.&3/4m 4.91/4.3 4.91/4.3 mal anac ttk., um 0.31 0.31 0.76 0.M 0.76 0.76 34/e'.ames Ig 18afelm 46 Chr 38ofW.--
(O h @) 0.3r= GC h 1.Onst least 14773 1m97 t. Cast im62 N 1.0518 1.0619 1.0693 1.0676 1.0536 1.0649 N - 14390 1.0386 1.0h06 14383 1215 h 1.0352 1.052 1.03 3 1266 1.03E 1.0369
. s m (,m) es 15m un ion isso is70 2350 leal REC, 4 11.857 11.380 14.001 15.019 15.427 15.383 EEC, 4 10.363 10.057 12.448 U.4Gt 14.223 13.983 haimb;e Assetur 185 185 11801 31801 33801 II808 BE,s 11.52 11.35 24.81 3542 30.36 30J9 EEC, 3 5.40 5.42 2.96 7J7 6J2 6.47 DLP.furge kenes STD. I M.2 30.9 41.8 39.9 44.4 45.0 Cat, 3 52J 72.1 44J 424 41.1 57.1 Punk Paw Deimity EE, Wasl 906 ( 5 ) 964 (tB) 989 ( 3 ) 938 (G) 1005 (E) 1014 (E)
BE,Wd 963 (G) 972 (G) 1018 (G) 1010 (G) gr% (en) EDs (36) J Osma ." 4 M fle #tt #12 1m tm tm tm tm tm Qele lagth,dsys 32 ~ 32 32 32 & B n=f==da/cytta 4/1J 6/1 6/1J 6/t 5/1J S/1 1s% 3and 3/e1. 3M/M6 530/M6 575/397 60nfou 575/397 6mf4M f platamfelm 19/14 19 /16 19/M 19/16 19/16 19/14 Gamal sup 0.36 0.36 0.36 0.36 0.34 0.3h6 lese p., s/d, as 4.91/4.36 4.9 t/4.3 5J2/4.40 5.55/4.s0 5.32/4Jo 5.3nr4.a0 Runt emme tit., m 0.76 0.76 0.4 0.M 0.76 0.M 34/elamat 30 01ut.-_ (0.Se e) (0.5 e e) (0.6= e) (0.6 = m) (0.6en m) (0.6= GD 1.0062 1.0e62 1m69 1Ager 14a69 Im07 1.05La 1.062) 1.0693 1.0tes 1.0D0 1.053B 143E3 1.0198 1.0614 tatt 1810 1812 1.0107 1.0222 1 0385 tales 1.0181 1.0M0 3dsg (pan) 1839 ;358 904 932 M38 ID2 28SU load REC, 4 15.301 15.001 16.992 17.439 17J21 527 EEC, 4 13 .574 13.385 15 257 15Jp5 15.9 7 16.385 temble amorteur 13808 11801 11801 !!8 0s 1380d 1880s BEC, s $1.96 31.75 33.07 53Jo 51.45 52. 2 RE,s 3.71 3.53 12.06 13.07 17.33 18.38 Diacturge names f!D. I 45J 46.3 42.3 41.2 elJ 40 4 03rr, I 49 4 66J 45J 40 4 38 4 51.1 Punk Paar thsutty EEC, Waal 1029 (G) 1051 (G) 960 (G) 952 (G) 988 (G) 135 (G) REC, Wasp 1059 (G) 1075 (G) 1057 (G) 104 (G) 1005 (36) 1GB (BL) 4 9 I
7 l
)
Also presented in Table 2 are the fuel cycle results using a 530/366-g element combination with 38 Cd wires of 0.5-am OD for the desired upgraded fuel cycle performance of increased cycle length or reduced reload batch size operation. For both types of e upgraded fuel cycle operation, the reactivity swing was too high and E0EC reactivity was too low. Clearly more Cd and 235U are needed to achieve the desired performance. The reactivity swing was reduced slightly by loading fewer control elements per cycle. The peak power increases ~2% for the reduced control rod reload strategy for 32 day cycle operation and increases 3 to 4% for the increased cycle length from 26 to 32 days. The same fuel cycle operation strategy shown in Cases #5-8 was used again for a 600/414-g 235U alement combination for the reduced control element reload batch size (Cases #10 and #12) and a 575/397-g 2350 element combination for the reference control element reload batch size (Cases #9 and fil). The standard fuel elements contained 38 Cd wires of 0.6-am OD in order to reduce the reactivity swing. This goal was achieved for the 32-day cycle Owing (Cases #9 and to the greater 11#10)Cd inventory reduction, during the ay 32while cycle, maintaining an a the power peaking increased 12% from BOEC to E0EC. Comparable E0EC k,gg's are noted for cases #11 and #12 but with higher reactivity swings relative to the equivalent fuel elements used in a 32-day cycle. This is caused by more unburned Cd at E0EC in the 26-day cycle than in the 32-day cycle. Therefore, longer cycles are more advantageous when larger Cd wire diameters are needed to suppress the reactivity swing. The advantage of increased Cd at E0EC was a reduction in power peaking during the cycle by 5 to 7%. RESULTS OF THREE-DIMENSIONAL REU FUEL CYCLE CALCULATIONS Comi.arisons of three-dimensional REBUS-3 equilibrium HEU fuel cycle results with those using previous two-dimensional (2D) models have shown only minor differences for the case with all rods fully withdrawn. The 2D model (Case #1, Table 2) gave re-activities which were 1.11,1.3%, and 1.2% lower than those of the 3D model at BOL, BOEC, and E0EC, respectively (Case #1, Table 3). This suggests an overestimate of the axial leakage in the 2D model. The element powers in the 2D model were overpredicted by
~3 to 4% in the fresher elements and underpredicted in the higher-
. burnup elements in the core peripheral column H by 5 to 10%. The peak power density in the 3D model remained nearly constant during the cycle at 860 W/cm 3. The peak power density was somewhat higher in the 2D model with a value of 904 W/cm3 in location 54 at 50EC, increasing to 943 W/cm3 at E0EC. The reactivity swing was 128 pcm higher in the 3D model than in the 2D model.
_.- - + -
r- , 8 Another 3D REBUS-3 HEU equilibrium fuel cycle calculation was run with all control rods banked at position 51.7 cm (the average for the entire cycle) using "dusay" experiments with Al plugs in Al filler elements. Some notable dif ferences were observed be-tween these two cases (with and without control rods inserted). The 105 burnable absorber-depletion rate decreased significantly in the top of the core near the inserted rods, which resulted in more unburned 10B remaining in-core at E0EC. The peak power density is higher because the flux is pushed toward the bottom of the core by the partially inserted control rods. The peaking values listed in Table 3 are for the rods at their critical position at the RIBUS BOEC and E0EC conditions. Since the rods would be even louer in the core for the start-up initial condition (no Xe), the actual peak power during the cycle will be somewhat higher than indicated. The radial power distribution was only affected by 5 to 10% in most locations, however. The k,gg was
. 1.0157 at BOEC and 0.9893 at E0EC.
In order to better simulate the reactivity effect of the 17 in-core experiments, a final HEU 3D equilibrium fuel cycle calcu-lation was made. ECN calculations 6 indicate that the actual control rod position at E0EC should be 61 cm with the 420/290 g fuel. The estimated 80EC control rod position is 55.3 cm af ter one day of full-power operation. Since modelling each experiment in the detail required would not be computationally feasible in a 3D REBUS-3 calculation, s' small concentration of non-depleting 103 was added to each of the 17 in-core dummy experiments to represent the poisoning effect of the actual experiments on total core performance. The asount of los used was 4.0 x 10 % atoas/bn-cm, which resulted in k,gg : 1.00 at BOEC with rods at 55.3 cm and at E0EC with rods at 61 ca. The REBUS calculation was run with the control rods at the average position of 58.15 ca. The effect of the experiments on the core performance result-ed in a radial shift of power into the A and B columns. The average increase in element power in the A and B columns was 7 to 10% with respect to the case without the experiment absorption modelling. All in-core experiments are located in columns C . through H, which accounts for the power shift away from that por-tion of the core into columns A and B. The other change from the previous case was a reduction in the peak power density from 1016 W/cm3 to 992 W/cm3 in the B4 fuel follower. This change was
, caused primarily by withdrawal of the control rods to positions
- higher in the core. Calculations with a more detailed mesh and with the rods at their actual critical positions are needed to determine the exact peaking for each control rod position at B0Er and E0EC.
0 , :.J., X ,,
- a 2
T .
* ~
9 RESULTS OF LEU FUEL CYCLE CALCULATIONS USING XYZ MODEL The three LEU fuel cycle cases that were calculated using the XYZ model are presented in Table 3. LEU Case #1 used a 19/16 plate (standard / control) element loaded with 500/339 g 235 g/ element with the same fuel management strategy as the reference HEU core. All LEU cases presented in Table 3 were calculated with a non-depleting 10 B content in the 17 in-core experiment locations of N(10 5)= 4.0 x 10-6 atoes/bn-cm and with all six control rods positioned at 58.15 cm during the entire cycle. The effects of the control rods upon the fuel cycle charac-teristics can be noted by comparing Case #4 in Table 2 with LEU Case #1 in Table 3. The main dif ferences occur in the burnable absorber depletion rates and peaking factor shif ts during the cycle. The results indicate that the LEU Case #1 fuel cycle would provide essentially the same excess reactivity at BOEC and slightly more at E0EC when compared to the reference HEU core characteristics in Table 3, HEU Case #3. The overall Cd wire depletion rate has been reduced by the introduction of control rods into the top of the core. This contributed to the 44% lower reactivity swing from BOEC to E0EC. The peaking factor actually decreases instead of increasing as the XY model indicated by with-drawal of the control rods to their E0EC position. The second LEU fuel cycle in Table 3 uses a 20/17-plate element loaded with 550/385 g 235U / element. The fuel cycle was altered slightly compared to the 19/16 plate LEU Case #1 by reducing the number of control elements loaded per cycle to 1 from 1.5. From previous XY nodel results presented in Table 2, the is about 806 pcm for 19 plate LEU fuels. The reactivity choice of penalt{5 550 g 2 /Ustandard element corresponds to a uranium meat density of 4.8 Mg/ud. The k,gg,, at BOEC and E0EC were slightly larger for this second LEU case compared to the reference HEU cycle even af ter the 33% reduction in control element loading per cycle. The average discharge burnup of the control fuel follower has increased by 21%. The 113Cd burnable absorber inventory has increased slight 1{35owing tothan 0 more the harder apectrum the 19-plate 20-element plate element loaded with 50 g of LEU Case fl. The increased 235U loading and faster depleting Cd burnable absorbers have contributed to the 69% reduction in reactivity swing conpared to the REU reference cycle. The peaking trends are similar to those of HEU Case #1 on the same Table 3. The third LEU case presented in Table 3 has equivalent reactivity performance to the reference HEU cycle at 50EC and slightly more at E0EC. Therefore, the control rods required withdrawal of 2.2 cm during the 26-day cycle instead of 5.7 cm for the reference EEU cycle. The reduced movement during the cycle is reflected in the 63% reduction in the reactivity swing. With
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B
e -- 11 reduction in 235U loading for this 20/17-plate fuel element of 235U / element, the spectrum becomes sof ter, which increases 25/15 the 11 gCd depletion rate and, consequently, the equilibrium cycle reactivity. t NEUTRON FLUX CHANCES IN IN-CORE EXPERIMENT POSITIONS i A summary of the average group-dependent neutron fluxes for the three I.EU fuel cycles modelled in KYZ geometry relative to the reference HEU fluxes is presented in Table 4. The inner irradia-tion positions refer to the nine non-peripheral locations C-3, E-3, G-3, C-5. E-5, G-5, C-7, E-7 and G-7. These inner loca-tions are modelled in an identical fashion to the eight outer (peripheral) locations D-2, F-2, H-2, H-9, H-6, H-8, D-8, and F-8, except that a stainless steel liner is placed around the central Al plug for all inner positions. All irradiation positions are primarily a mixture of Al and H2 O with a small concentration of non-depleting 10 B to simulate the total effect of resetivity reduction on the core performance. Table 4. QWson of JWernge Neutron Flux Ratios in Nine Imer ad Eight 02ter In-One Irradiation Positions in IRJ Relative to HEU Equilibrius CF.le (bres (bntrol 235U P4mber Ibd Amrage Group That Ratios (IRJ
- IED 46.;; tamiirs Plates / Position .
L-eim (g/el) element krrup (cm) +1 +2 +3 $4 +5 Imer 500/339 19/16 BOEC 55.3 0.967 1.004 0.986 1.2(D 0 .9 15 0 car 1417 1.030 1.006 1.152 0.950 Inner . Euc 61.0 0.965 1.002 0.978 1.187 0.880 Oscar 1.0 14 1.027 1.001 1.143 0.924 Inner 550/380 20/17 BOEC 55.3 0.967 1.011 0.964 1.030 0.832 Outer 1.047 1.065 1429 1.041 0.905 Inner EDEC 61.0 0.962 1.006 0.974 0.996 0.793 Qatar 1.(D3 1451 1.0 15 1.023 0.870 Inner 525/365 20/17 BOEC 55.3 0.995 1.017 0.992 1.023 0.857 Oscar 1.048 1.067 1.034 1.047 0.922 Inner Euc 61.0 0.990 1.0 13 0.983 1.008 0.821 aster 1.006 1.054 1421 1.031 0.889 Imer lac 57.5 0.992 1.016 0.986 1.010 0.820 O2 car 1.047 1.066 1.031 1.041 0.897 I
., ~ .
12 i The neutron flux ratios (LEU
- HEU, Ref.) for ueutrons with
. energies >1.85 eV range between 0.985 for the inner experiments to 1.065 for the outer experiments. The primary flux redistribution for the 19/16-plate case occurs from group 5 (En < 0.625 ev) into group 4. Fluxes are 5 to 12% lower in group 5 and 15 to 20%
higher in group 4. This spectral hardening is caused primarily by a the increased 235 U loading in the 19-plate LEU element relative to the 23-plate HEU reference element since the water channel flow area is nearly equivalent for these two designs. The group-5 flux ratios at E0EC are ~3% lower than at BOEC since the sof tening of the spectrum during the cycle is less with the LEU core relative to the HEU core. The effect of a change in 235U loading and control rod position upon the average fluxes in the irradiation positions can be observed by comparing the two 20/17-place cases. The harder spectrum 20/17-plate cases have caused slightly larger flux ratios in groups 1-3 and reductions in groups 4 and 5 compared to the 19-plate case. The change in 235 U loading and control rod position
. have only a very minor effect upon average neutron fluxes.
However, movement of control rods do cause substantial flux changes in exial zones near the bottom of the Cd control sacerial. For the inner experiment locations group fluxes are within st2% for En > 0.625 eV and are reduced by 14 to 18% En< 0.625 eV for the 525/365 g case." For outer or peripheral in-core positions fast flux increases of 3 to 7% and reductions in only the lowest thermal group of 8 to 10% can be expected. CONCLUSIONS i From the neutronics point of view either the 19- or 20-plate standard element would be a suitable replacement for the current HEU design and fuel management strategy without significant increases in power peaking or losses in neutron flux in the irradiation positions. Reductions in the annual control element loading requirement by 33% and and reduction in the movement of control rods by 60% during the cycle have been achieved for all proposed LEU fuel cycles. These added features were possible using uranium seat densities of <4.8 Mg/m3 .
._n ._.a. .-.
4 k 4$Wp 4 $ g
,- o-
r 13 REFERENCES
- 1. J. R. Deen and J. L. Snelgrove, "A Neutronics Study of LEU Fuel Options for the HFR-Petten," in the Proceedings for the 1984 International Meeting on Reduced Enrichment for Research and Test Reactors, Argonne, Illinois, ANL/RERTR/TM-6, CONF-8410173 (October 1984), pp. 323-333.
- 2. B. J. Toppel, "A User's Guide for the REBUS-3 Fuel Cycle Analysis Capability,"' ANL-83-2 (March 1983).
- 3. B. A. Zolocar, et. al., "EPRI-CELL Description," Advanced Recycle Methodology Program System Documentation, Part II, Chapter 5, Electric Power Research Institute (September
- 1977). EPRI-CELL code supplied to Argonne National Laboratory by Electric Power Research Institute, Palo Alto, California (1977).
- 4. R. E. Prael and J. J. Milton, "A User's Manual for the Monte Carlo Code VIM," FRA-TM-84 (February 20, 1976).
- 5. K. L. Derstine, "DIF3D: A Code to Solve One, Two, and Three Dimensional Finite Difference Theory Problems," ANL-82-64 (April 1984). l
. 6. G. 01thof and A. Tas, " Methods to Overcome Stortage of
, Reactivity by Means of Increasing the 235U Mass and the Effect of a Modified Control Rod Loading Pattern," ECN Petten Memo No. 84-15 (May 1984).
e I I ? -- .. .. . . t ,
;L.?7 KUR CORE CONVERSION - NEUTRONICS CALCULATIONS Keiji Kanda, Masatoshi Hayashi, Kalchiro Mishima, Yoshihiro Nakagone and Toshikazu Shibata Research Reactor Institute, Kyoto University Kumatori-cho, Sennan-gun, Osaka 590-04, Japan
! ABSTRACT d
As one of possible future programs for the
' Kyoto University Research Reactor (KUR), the Research Reactor Institute of Kyoto University (KURRI) has a ; plan for core conversion to the use of low-enriched uranium (Lau) fuel. A feasibility study for this started in November, 1983, as part of the joint study between KURRI and Argonne National Laboratory (ANL).
Thermal-hydraulic analysis on the use of LEU fuels in the KUR was Performed in 1984. The results indicated that core thermel-hydraulics posed no
. significant difficulties for the core conversion.
i l During 1985, neutronics feasibility of the KUR core conversion was studied using the neutronics code j (3PRI-CELL) system at ANL. This paper presents some i of the results obtained from the calculations, including the comparison between the current highly 0.586 U/cm3 )
, enricheduraniumHEU(93%,UAl-allog),
and LEU (19.75%, U3 31 2 -A1, 3.23 U/cm as follows:
- 1. Critical mass
- 2. Neutron flux distribution
, 3. Fluz peaking factor t
- 24
INTRODUCTION The KUR (Kyoto University Research Reactor) is a light-
! water-moderated tank-type reactor of SNW. It has been successfully operated for more than twenty years, during which period it has served as one of the most useful inter-university research reactor facilities in Japan. It is believed that the KUR deserves upgrading during core conversion for longer use.
As a possible future program, a feasibility study has been performed on the core conversion to the use of LEU (low enriched uranium) fuel. This study is a part of the Joint Study Programme between the Research Reactor Institute of Kyoto University (KURRI) and Argonne National Laboratory (ANL) in Phase C. At the last RERTE Neeting at ANL, in October 1984, we presented a paper on the thermal-hydraulic analyses for the KUR. It showed from a thermal hydraulic viewpoint that the current HEU (highly enriched uranium) fuel, 93% U-235 EU, UAL alloy, 0.5855 U/cm 3 , could be converted to LEU fuel, 19.75%, U 3s12 -A1, 3.23 U/cm3 without major design changes, if the peaking factor in the converted core would be less than 3.2. The present paper shows the neutronics calculations for both HEU and LEU fuels, which are very satisfactory. So far, we have a plan of the KUR core conversion as shown in Table 1. DESCRIFTION OF THE KUR l ; Figure 1 shows the horizontal cross section of the KUR. ( Figure 2 shows a typical core loading pattern. Currently, we l . use a typical curved NTR-type fuel element shown in Fig. 3, which will be changed to a straight NTR-type one shown in Fig. 4 at the occasion of the core conversion. CALCULATION 4 In the present calculation, the well-known ANL/RERTR version of the RPRI-CELL code was used, whose flow diagram is shown in Fig. 5. l l
. 1
. +-
Criticality
' The eigenvalues for the current core and for possible converted cores, whose specifications are shown in Table 2, were calculated as shown in Table 3. The agreement between j calculation and experiment for the initial core is excellent,
; which confirms that the ANL code system is useful for such reactors as the KUR.
Flux Distributions and Peaking Factors Figure 6 shows the fast, epithermal and thermal neutron flux distributions for the current KUR with 93% EU and the converted one with 20% EU. Figure 7 shows the thermal neutron flux distributions in HEU and LEU cores in both I- and Y-axis. From these figures we understand that the fluz change caused by the reduced enrichment is small. The peaking factors of various cores are tabulated in Table 3. i CONCLUSION
- 1. The ANL code system is suitable to KUR conversion calculations.
- 2. The flux peaking in the LEU core is less than 3.2, which satisfies the limit of thermal hydraulic condition.
- 3. The change of flux distribution in the converted core is small.
ACKNOWLEDGMENT This work was supported by the Ml'istry of Education Science and Culture. Dr. James E. Natos, Dr. William L.
, Woodruff and Dr. Armando Trave 111 of ANL gave useful suggestions and support.
REFERENCES
- 1) K. Mishima, K. Kanda and T. Shibata, Thermal-Hydraulic j Analysis for Core Conversion to the Use of Low-Enriched Uranium Fuels If the KUR, KURRI-tr-258. (Dec. 1984)
I I
, 1
i Table i Tentative schedule for the KUR core conversion Fiscal Year 1984 1985 1986 1987 1988 1989 1990
, in use new (for 2 yeare) extra (for 1 year)? , 93%EU-fuel : 2 Burnup 25% - 35%
Safety Safety Thermal- Neutronice Fuel Tests Review Examination Hydraulic Analysis for LEU (U3517-A1) = = Analysis 20!EU-fuel - Fabrication Operation O 6 W
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no a a F F F F F a en nr a a a a a a a a en F: STAM0Aa0 FUEL CLD'ETT C: SPECIAL Futt ELIMENT FCs CONTROL n00 M: MALF-LOAct0 FUEL ELIMINT a: GRAPMITE atFLECTCn NY : NYORAULIC ISRA0!ATION Pont PN : PhttMATIC Tuel Poaf PL : PLUG FOR LONG Tisre laAADIAT!0m Fig. 2. Typical core loading pattern of the KUR t_________-._____________--______-.______________-. - _________________-_____-------_;
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, Fig. 4 The modified KUR fuel element (see element in Table 2)
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[ CINDER l Constees eefeth T" he output Fig. 5. Flow diagram for the EPRI-CELL code Table 2. Fuel Element Geometries for LEU Fuels Plate Inseber Flate Channel Heat Flow Area Hydraulic Element Shape of Flate** Thick.* Thick.* Thick.* (cm2 /Eles.) Dia. (mm) A Ostved * *
- 18/9 1.52 2.81 0 50 35.337 5.341
, 5 Flat 18/10 1.52 2.81 0.56 35.337 3.390 C Flat 19/10 1.27 2.81 0.31 37.761 5.392 D Flat 21/12 1.52 2.20 0.50 33.771 4.259 j E Flat 22/12 1.27 2.20 0 51 35.232 4.259 l . " Unit in mm.
** Number of Flate Standard Elseent/ Control Element.
*** Curvature Radive s 139.7 as.
**** Effective Core Heights 384 me. Width of Heat: 63 mm.
Notes LEU:19.751EU. U3512-A1. 3.23U/cm HEU:93.01EU, UAl-alloy 0.585gU/cm 1hble 3. Eigenvalues and Peaking Ihetors for Critical Core Loading Pattern of the KIR EdnENT ENn CHMENT K EXP PEAKING NAnE FACT 0m
.l A 93.191 1.0045 1.0055 - 2.46-B 19.751 1.0053 2.52 C 19.751 1.0291 2.58 D 19.751 1.0176 2.54
'E 19.75% 1.0429 2.61
- DIF3D 3-n MENSION, 5-GROUP.
e KUR Flux for Groups 1,3 and 5 at 93%E PIva
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fO 1 i i 6 Y LEU-PLATE IRRADIATION AT FRJ-2 t (DID0) UNDER THE GERMAN AF-PROGRAMME l
- l E. Groos, W. Krug, J. Seferiadis, G. Thamm ,
Kernforschungsanlage Julich GmbH l Julich, Nordrhein-Westfalen, BRD l ABSTRACT l r 10 LEU fuel plates (8 with uranium ! silicides max. U-density 6.1 g/cm8) have l been irradiated at FRJ-2 (0100) of KFA-1 1 Julich till cnd of October 1984 during 321
! fuly power days up to max. burnup of 2.41 x
. 10 fissions /m8 without major interrup-l tions and troubles, PIE began recently in 4
KFA hot cells. Visual inspections revealed -
! no damage or greater deformation for the majority of the plates, but red / brown coloured layers (partially peeled off) on the cladding over the fuel. Aluminium (oxide) is the chief constituent of the layer with smaller portions of Ni and Fe the latter causing the red / brown colour.
[ The major part of the layer (~50 pm) i most probably has been formed during 20 h immediately after experiment start-up under ( abnormal conditions of the coolant water. l Gamma scanning has been completed. 01 men-sional measurements are under way confirm- ! ing first observations of severe swelling i i (pillowing) of 1 plate. Density and i blister testing as well as metallography and burnup analysis remain to be accompli-shed end of 1985/beginning of 1986. l l L
'l l
1 E I f 1. Introduction l The main objective of the irradiation tests under
; the German AF-Programme is to demonstrate that reduced l -
enrichment uranium (REU) fuels fulfill the same
' safety criteria during reactor operation as do the current high - enriched uranium (HEU) fuels.
Four types (stages) have been identified for the 1 1rradiation tests, all of which an entirely new developed fuel has to undergo /1/. These stages, which have been agreed upon by the German licensing autho-j rities and their experts, are: d 1. Irradiation of small plates (5-10) in l' well instrumented irradiation devices wi'th subsequent extensive post irradiation 4 examinations (PIE) to determine the fundamen-tal irradiation behaviour of new fuels. .
- 2. Statistical irradiation performance tests of
{ a " greater" number of fuel plates (20-30) I with partially extensive PIE prior to { complete element testing.
- 3. Irradiation of complete test elements (3-5, possibly in several reactors) with
.I minor PIE (primarily in the case of failure)
.! to achieve standardized REV fuel element j manuf acturing specifications. 4 Irradiation with prototype fuel elements
, (optimized with a view to the future LEU fuel cycle according to fuel element design, type j of fuel and U-235 content) immediately prior to the HEU-LEU conversion of a reactor.
15 small REU fuel plates have been irradiated
; respectively are being irradiated in 2 different
- irradiation devices at FRJ-2, KFA-Julich, and FRG-2,
; GKSS Geesthacht. The first irradiation campaign with 1
10 test plates (8 plates with uranium silicide fuels,
, 2 with U 10,) has been finished October 1984 at FRJ-2 i ( 010 0 ) . 7 N are under way in hot cells of KFA. A I
I description of the test facility (AF-loop) as well as irradiation data and recent PIE results (preliminary) j will be given. 4 i t
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- 2. Description and data of the AF-loop The irradiation device in the FRJ-2, called AF-Loop-Julich, consists of the inpile section, the loop system and the electrical instrumentation and control system (fig. 2.1). It is designed for a maximum power removal of 250 KW.
The inpile section contains 10 testplates within the core region, partially instrumentated, which are l arranged in two groups of 5 plates each, one above the I other, as shown schematically in fig. 2.2 f fE ? ??? ! @ E
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6 - o,.- . s .. - , , " , , .ni _. r i. ~, ., ,~,, , ,-,,,, e cladeng motmot. Al Mg 2 US *loodmg : 6700 mg i plot, l i Fig. 2.2: Arrangement of the test plates Detailed informations of testplates and. fuel are given in the following table 2.1. m m'o**** imas % us. er w ,,,,,,. passermes, Ig/M (%) t islast u eco u, o, x 2 tm / asi Ars 17s u, s, 3 i n / 448 104 itrs u, s,
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e 5 The loop system, which operates as an individual closed circuit is connected to the inpile section. It has to cool the test plates and to control the test plate temperatures. Deionat is used as cooling medium with a controlled conductivity of less than 3 pS/cm at an constant overpressure of 3 bar. Massflow, t emp e ra - tures, pressure and activity are controlled continuous-ly. In case of a failure, the reactor will be automa-tically scramed. To avoid an overheating of the testplates caused by a failure of the main pumps (1 in operation, 1 for reserve), the auxiliary cooling system, which is independently working, will be automatically put in operation in case of a scram. The main cooling system and the auxiliary cooling system are only connected to each other in the top of the inpile section. Also the corresponding magnet valves for closing and opening the systems are directly connected to the top. So a simultaneous failure of both couling systems by a pipe break can be avoided. All safety data will be indicated, registrated and controlled by the electrical instrumentation and control system. All experimental data are registrated by a data acquisition system.
- 3. Irradiation Data The main irradiation data and results are given in the following table 3.1.
Irredsson cemoni.n:1107.83 28 f(L M Pueldsomer. power: Pus poner days: 320.8 . se Toeni energy: T25083IM.i - en.rt of irressa.on: of irreisanon' .2.1 kW127 8 kW US est auf 1 E0 f.TF .10" t,1 10" F43 31J f 38 120 El 3 .1. 11 10" 1 . 10 " Ftl 31. . . . 3 t.1 f 10 " 10. 10" FSA 34 . . - 4 EF 223 10" 1 1. to" Fia M. . . = 8 53 121 10" 14 . to" FSJ 38 f 28 1 82 ta.
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Table 3.1: Main irradiation data of the AF. Loop l 1 I [ l L
6
! Some characteristic temperature profiles of the testplates for the reactor periods 4, 11, 13 and 15 (final period) are shown in fig. 3.1.
5 130 -
. c 120-N . period 4 11 0 - . .
* . period 13 100-90-L
*N, . period 15 M i . , ,
'! T26 T25 T24 TE-crrangement T23 Fig. 3.1: Temperature profiles of the test plates
.i During the whole irradiation campaign no problems arised with the operation of the experiment. The conductivity of the cooling water never exceeded 1,5 pS/cm and no activity increase has been measured.
This is valid with the exception of the first day of
! operation, 1. e. during start up, which will be l discussed later (see S.).
1
.i '! 4. POST IRRADIATION EXAMINATIONS 1
i Visual Inspection i All 10 fuel plates were covered with a thin, i reddish brown layer (Fig. 4.1). The colour appeared darker in fuel areas of higher temperature. The layer partly peeled off in small chips of about mm-size, leaving in some cases marks in the metal surface like scratches with a knife. Along the line circumscribing
] the fuel meat, the layer had formed small blisters, i some of them had peeled off, exhibiting the clean and ,j undeformed metal underneath. Very distinguished was the ! colour as well as the blister formation at the post-i tions of the thermocouples (Fig. 4.1, Plate No 1 and j 6). A small amount of peeled off chips were collected for further analysis.
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No.9 No. 9 FIGURE 4.1: PsoToGRAPHS OF 3 FUEL PLATES (FRONT, REVERSE) ' i
8 One plate (No. 1) with U 0o-Al fuel had developped during irradiation a flat bl3 i s t e r of about 20 mm i diameter about twice as thick as the o.ri gi nal plate i symmetrically on both sides of the plate. However no damage of the metal surface could be observed (Fig. 4.1, Plate No 1). 4
' Two other plates with U Si and one with V Si fuel showed very slight indicat1ons 3 of blister for$ation
- with about the same diameter but much flatter (Fig.
, 4.1, Plate No 9).
t It seems noteworthy that all 4 blisters had formed
] in similar positions of the plate, i.e. in a corner i
upstream of the cooling water. Layer Analysis The collected chips were analysed by three methods: l
- optical emission spectroscopy (quantitative),
- micro probe analysis (qualitative) and
{ - gamma spectrometry (qualitative). i i The results are summarized in 3 tables 4.1-4.3, I indicating that the layer is AI i ties of which the Fe causes the ,01 Tendish with common colour. impuri-I AL 3P % MG 0,4 % CR, CA, CU, M#, ZN NI 5% SI 0,4 % LESS THAN 0,3 %
.{
FC 1% NA 0,3 % TAR. 4.18 EMISSION SpCCTROSCOPICAL RESULTS MAIN CoNSTJTUENTS: LESSER CONSTTTUENTS: TRACES: AL , o p FC, CR,'MG SI, S TAR. 4.P MICR0 PROBE RESULTS (PRELIMIP8ARY) l , t ZH b5 3,83 SC 4b 0,2 CR 51 0,02 l Co 40 2,0 FE 5'l 0,12
! MM 54 08 50 124 0,1%
TAB. 4 3: GAMMASPECTROMETRICAL RESULTS {10 b 00) L.
9
. 9 The radioactivity is very low, only about 6 x 10 0 Bq ( ~150ect ) per gr with Zn65, Co 60 and Mu 54 as main constituents. I t
l Gamma Scanning l Each plate was scanned along its length with a slit 0.2 mm wide and 60 mm high (scanning along the , width is under progress). Every 2 mm along the plate a l spectrum was measured and evaluated for total gamma ' activity and various nuclides. There was no indication of migration of fission t
- ' products, slight peaks at both ends of the profiles are i probable caused by the rolling of the plates (dog-i boning) during fabrication. Gross activity and Cs'134 profiles for 4 typical plates are shown in fig. 4.2 and 4.3. The figures are arranged as if the cooling ,
water was flowing form left to right. ~
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{ Dimensional Measurements Because of the alumina layer covering a conside-rable part of the plate surface, measurement of plate
; thickness is problematic. Up to now, no attempt has been made to remove the layer in order not to damage the metal surface.
i Tentative thickness measurements were made with l l micrometers and calipers with semispherical tips, how-l ever, the results are very uncertain and have so far { only shown that the blister in plate No. I has a thickness of about twice that of the plate and that the alumina layer reaches a max. thickness of 0.1 mm in some spots. l r 1 l 1 I 1
( [. 1.1 It is uncertain if with an alumina layer of such a ' dimension a thickness measurement can be of any significance after a not negligable part of the metal has been consumed in forming the layer. At present possible alternative methods to establish the dimen-stonal behaviour of the fuel are being discussed. Future Examinations Apart from the thickness measurements the alumina layer will also influence the density measurement. However, these nondestructive examinations have to be completed - or cancelled - before any further examinations can be performed. Intended are metallo-graphic examinations, blister tests and burn up determinations as destructive tests. The dimensional behaviour of the fuel can, if all other methods fail, be determined from metallographical sections.
- 5. OXIDE LAYER FORMATION AND WATER QUALITY Immediately after the irradiation test started (July 1983) the acitivity level of the coolant was reaching relatively high values due to abnormal water conditions mainly coming from impurities (Al, Na, NH4, NO3, SO4) with concentrations much greater than during later operation after a complete exchange of the primary coolant water (see table 5.1).
t abnormal conditions normal conditions I mg/l . mg/l At 0.22 0.01 Mg 0.023 0.04 Na 2.16 < 0.01 Ca 0.011 0 07 NH 4 2.0 06 CI < 0.1 < 0.1 NO3 0.2 < 0.01 SO4 1.6 0.77 ph: 9.1 ph: 5.4-6.0 ei. conductmty: 1.9 mS/m el. conductmt/: 0.10413 mS/m 1 Table 5.1: Water quality during irradiation - p__
l 12 The major part of the Al-oxide layer is assumed to be formed during start up of the experiment within about 20 h of operation under abnormal water quality conditions. That increased oxide formation is due to abnormal conditions also has been found in former US-LEU tests where typical oxide film thicknesses of less than 10 pm only had been exceeded (up to 25-50 m) e.g. in the case of high temperatures the plates encountered during PIE in hot cells. I i
- 6. RESULTS AND CONCLUSIONS I Although the major part of the PIE still remains to be done the results to date and the conclusions which can be derived preliminarily will be summarized
- as follows:
S. 10 LEU test plates have been irradiated troublefree l up to very high bu rn u p2(a t e s (more than 80 %, cor-
; responding to 2.4 x 10 fissions /m8) which was de-7 monstrated by the continuous recording of essential operational data (e.g. very low and constant activi-ty level of the coolant).
Visual inspections during PIE indicated that one < plate was severely swelled (pillowed). The max. ) thickness increase determined by dimensional ! measurements which still are under way has been j found to be about 250 % (ref. original fuel meat i j thickness) for the Ug 0a-Al plate (Ng 1) achieving a burnup ratM of 1.9 x 10 f/ms. . Irradiation under conditions of the coolant that , i deviate from normal values (ph in the range of 5.5 - ! 6.0, el. conductivity 0.15 mS/m) leads to increased I production of Al-oxide on the plates. If operation under abnormal conditions is limited to ! t about 1 d serious damage of the plates with conse- )
; quent fission product release into the coolant can 1 be excluded. '
i Operation experiences at HTR-research reactors (e.g. i at FRJ-1/FRJ-2 KFA-Julich) show that longer periods 1 (several days) of " abnormal" operation results in a j s i gn i ficantly greater corrosion attack on the fuel plates followed by the release of fission products. i I 1 I m
, 13 Acknowledgement:
1 i Analytical work was performed by Dr. H. Heckner, KFA/ZCH and Dr. H. Grubmeter KFA/IRW. References
/ 1/ G. Thamm, "Five Years German AF-Program Review and Outlook", Proceedings of the 1984 International Meeting on Reduced Errichment for Research and Test Reactors, ANL, Argonne, Illinois, USA, October 15-18, 1984, ANL / RERTR / TM-6, CONF-8410173
/ 2/ W. Krug et al., "Bestrahlungserprobung von neu
- entwickelten, hochdichten niedrigangereicherten Brgnstoffen im Rahmen des AF-Programms",
l 26 Plenary Meeting of the Irradiation Devices Working Group, 8.-10. October 1980, Geestacht I /3/ W. Krug et al., "Bestrahlungstests mit niedrigange-reicherten Brennstoffen tyd Brennelementen fur For-i schungsreaktoren", 27 Plenary Meeting of the 3 Irradiation Devices Working Group, October 1982, Grenoble
/ 4/ W. Krug et. al., "Sicherheitsbericht LV 50",
j AF-Loop-Julich im FRJ 2, April 1983 i f { i a l l O 9
f~. _ - .. .. c . t Paper .No.31 s F 1 L , A CCMPARATIVE AND PREDICTIVE,jTUDY OF THE ANNUAL FUEL
~
CYCLE COSTS FOR HEU" AND LEU TUE LS IN THE HIGH FLUX REACTOR, PETTEN, 1985-1993 5 R.L. Moss and P.May
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HFR Division, Joint Research Centre, , Commission of the European Communities, Petten, .N L Aostract t- . Tne internationally agreed constraint on availability [ of supply of 'HEU f uels to Research and Test Reactors has ' necessitated that a cost- analysis be carried out to . determine .the financial eff ect of converting the core of the HFR'from HEU to LEU f uels. A computer program, written at Petten and based on information extracted f rom studies in Europe and the USA, identifies the major cost variables to be manufacturing, uranium, reprocessing a nd transport cos ts. Comparison between HEU and LEU cores have been carried out and , includes the effects of inflation and exchange rate . fluctuations. Conversion of the HFR core to LEU f uels is i
;. shown to. be financially disadvantageous. *
- HEU - High Enriched Uranium
*
- LE U -
Low Enriched Uranium t HFR - High Flux Feactor m 4
- m. -
F s .
- 1. INTPCCUCTION Due to internationally agreed constraints on the availability of supply of HEU fuel tc Research and Te s t Reactors throughout the western w o r ld , it has ceen necessary to execute cost analyses on the effect of converting the HFP Petten core fecm dLU to LEU f ue ls . In j the HFR at Petten which at present requires for operation approxima tely 35 kg . U- 235 per annum of HEU fuel, tne change tc LEU fuel necessitates its own detailed anal-l ysis, which should include projective analyses for estim-ating cos ts in f uture years.
l Es tlier similar studies, /1,2,3/, have already b ee n carried out in Europe and the USA. Whilst these studies i primarily concentrated on fuel design and f so r ica tion l optimisation, it could be concluded, /3/, that if in the i HEU fuel case, the fuel losoing and burn-up of a fuel ele ment is low, then there will ce seme economic advan-tage in the LEU case. On the other hand, the converse applies, that is, tha t if high HEU burn-up, then conver- ! sion to LEU is disadvantageous. l Frcm Matos and Freese 's work, /1,2/, the major cost l items were identified as manufacturing ( -201 ), uranium, i including enrichment ( - 271 ), rep roce ssing ( -20% ) a nd l shipping ( -15% ) . Manufacturing was evidently identified l as the major variable. However, it was concluded and l pred icted that possible eventual standardisation of the LEU manuf acturing process would reduce the overall costs quite considerably. From these conclusions, it transpires that individual cost analyses for a particular reactor is , necessary. hence, the analysis for the HFR case. The work l here includes aspects f rom each of the studies above and l utilises the most recent cost information. Projection to l future years has also been carried out and is based on educated guesses of expected inflation thecughout the countries involved in an analysis and on the corres-ponding exchange ra te fluctua tions. [ The analysis has been performed with the help of a computer program specially written at Potten. The pecgrae
- allows for easy input of parameter variation and was
! aimed to present its results clearly in both taoular or g raphical form. The program is described in more detail i in section 4. l l l I l I l l I
i L The results of the prosent analysis are discussed in section 5. Various conclusions are put torward in sec-tien 6.
- 2. HIGH FLUX FEACTOR, PETTEN The HFR at Potten is a 4'S '% Materials Te s ting Reactor, cooled and moderated oy light water. A general lay-out of the reactor is shown in Fig.1. Ine core lattico, see Fig.2, is a 6 x 9 array, containing 33 fuel "
assemblies of the MTP type, 6 control members, 17 ! experimental positions and 16 Beryllium reflector elem-ents. Each fuel assemoly contains 23 vertically arranged f uel plates. The uranium is 93 n enriched in U-2 35 and , as , of recently, each fresh fuel assemoly contains 4209 U-235. The control of the reactor is achieved oy tne o control rods, comprising a f uel section surmounted by a cadmium sec tion . Drive mechanisms below the reactor move the control rods upwards, displacing the cacmium with fuel. The f uel content in a f resh control rod is at p resent 290g. Calculations on the comparative thermohydra ulic and nuclear as pec ts of the HFR in a standard HEU core and a typical LEU core have been previously studied and were presented at the last REPTR at Argonne in 1984, /4/. t
- 3. FUE L CYCIE COST FACTCRS The model and assumotions to compute the annual f uel cycle costs are based on information and CCE formulae ex tracted from Refs. 1, 2 & 3. Fixed costs for the manu-f acture and transport of the fuel elements and control rods at the European manufacturers are obtained directly from the companys involved.
The calculational procedure is shown s chema tically step-by-step in Fig.3. The variables are as f ollows: l 3.1 HFR Requirements , The amount of U-235 required to feed a years cycle in l the HFP is determined by the formula r 1 I t l r
7 , r . . . j amount of U-235, US say, =3p x Mr + N. x M. (1) where, N, = no. of f uel elements p. a. , M, = U-235 per element,
= no. of control rods p. a. ,
N. and M. = U-235 per control rod. Cue to expected processing losses of 2.5%, the amount of enriched uranium required is given py: 4 Ug = 1. 0 25 x U, / d, , a t E.1 e n r ichm en t . (2)
, The equivalent amount of U nat required is given by:
U ,gg = (&.-t s)/ (a,- EJ x Ug (3) 4 where a s = assay of tails ( = 0.2% ) and Et = assay of feed ( = 0. 7111 ). The amount of U nat which actually needs to ce purchasec for processing is reduced by the amount of
, credit uranium in the spent fuel which is ul tima te ly i returned at the end of the cycle to the reprocessing plant. The amount of credit uranium is determined as folicws:
if B is the percentage burn-up at the end of the cycle , then the amcunt of uranium given in ogn(1) l
. is reduced by (1-B) a nd its enrichment is given by the DOE formula: ,
Sa = (1-B)2. /(1-0.84 &,B) (4) o where dg = reduced enrichment after burn-up, an. Hence the amount of U returned is (1-B) U, AEg and the equivalent U nat is given by: S U nat = (4 -A n j/(15 -A g) x (1-B)U,,/Ag, (5)
F L . . i -S-which, due to reprocessing and conversion losses, must be reduced by 2.3%. Hence, th e ne t requirement of U nat to be purchased is given by; e s Net U nat = U nat . - U nat (kg) (6) The cost of U nat per kg. was, at tne tima of publication, 40-45 S/kg. The expectations are enat this t w ill remain con s ta nt until 1990, whereaf ter it may be i ' assumed to rise with inflation. 3.2 CONVEFSION
- Conversion of U nat to Urs is simply given oy a fixed price pe'r kg U nat. Present prices are $6.5/kg.U nat.
i Thereaf ter , it is expected to rise at a similar rate to '
- i. ; reprocessing cost increases.
- j. .
- ! 3.3 ENRICHMENT i ,
The cos ts of enrichment of the UE6 is determined in quantities of Separative Work Units (SWU) . These are again de te rmined f rom DCE formulae and are given per kg l U. The SWU formula is I ! t SWU. = V (1. ) -V(4. ) + ( f. - 4 ) / (4, - 4 ) .V ( Eg ) -V ( i p ). (7) l where V(&L) = (1-21;) in ( (1-ic ) /& L ) , for i= o,f,t,
. the amount of kg U being given by egn(2) .
. Again, credit is taken account of for the SWy, due to the uranium re turned to the reprocessing plant. The credit SWU is determined by; no. of StiU re turned =
0.977 x SHUg x U-returned (s)
= ShU , say g
J where SWU3 is given by egn (7), out witn th substituted for 1 a 9 f e 8
f Hence, Net SWU = S W U, - SUU g (9) The cost per SWU was recently S135 and was expected to rise by approximately 10 % p.a. Hcweve r , f rom recent COE information, it transpires that for s e para ti ve work on HEU with assays >10% U-235, the new price f or 1966, under the existing Euratom contract is S125.44 per 3hU. 3.4 MANUFACTU RE OF FUE L E LEMENTS Afl0 CCNTFO L RCOS For the HFP Petten, f uel elements are purchased from both NUKEM in W. Germany and CEFCA in France. Tne control rods are purchased from NUKE!!. bhilst de tailed analyses could be carried out cased on prices per kg U per element, a final design of LEU f uel elements is as yet still to be f ixed. Initial estimates suggest 1.5 to 3.0 x the price of HEU elements. However, as pointed cut in reference (1), eventual standardisation of tne LEU elem-ents manuf acturing process wculd probably red uce the resultant price. .Ne ve r thele s s , sensitivity analyses are here presented comparing HEU ILEU manuf acturing cost rat-ios. ( Note tnat element refers to both f uel element and control cod ). The 198 5 prices f rom NUKEM are CM 20,790 per fuel e leme nt and CM 35,690 per control rod. Fuel elements purchased from CEPCA cost FF 60,350 per element. 3.5 TPANSPORT AND SHIPPING Transport and shipping are included as additional cost items for the movement of -
- 1) shipping of tt.e enr iched UF6 to Europe (fixed p.a. cr $300/kg U)
- 11) transport of the elements f rom manufacturer to Pe tte n (fixed per transport) 111) shipping of spent fuel to USA f or reprocessing (fixed per shipment or 01500-S3500/ element) iv) transport of re-conver ted UFg to re-enrichment plant
- (523/kg U in U ff )
The fixed prices per transport or shipment are again
. _ __ -_ _ - _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ . _ _ _ . _ _ - m _ _ . _ _ _
i '. l . i ( I lt
+:
based on previous costs and are conse rvative e s tima tas . l Prices per kg U per element are also quoted. Both types, where ap p rop ia te , are considered in the sensitivity ' analysis. The transpor t of enriched Urs to Euroce and of the manufactured elements to Petten is taken to be tne same. For shipment of spent f uel to the US A, a price per l trip ( average _2.25 per year ) or costs per element are
, quoted. Sotn types are considered.
i l 3.6 RE-PROCESSING , ! l The re-processing of the spent fuel returned to the USA is charged in $/ag (Al+U) . The amount of A1+U is de te rmined from burn-up and weight of material returned: kg(Al+U) =U6x (1-B)/43 + 4.2(Np + N ),. (10) 1 4 where 4.2 is the numoer of kg of Al per element. The i
. , price per kg/(A1+U) is at present $1100. The ex tra costs I
for LEU fuel p roces sing due to the use of uranium i silicides instead of UAl., have been quoted at +10%, /2/, '
; or up to 2x HEU costs, /3/. Ag a i n , the effect of the l
increased price on overall costs is considered. i i '
}
3.7 RE-CONVERSION AND RE-PEOCESSING ! \ - l The resultant UNH from the re-processing of A1+U is ' i
,' re-converted to UF6. The amount of kg U in the UhH is )
i i given as above, that is: - l' s 6 U x (1-B)/is, and reduced by 2.3% due to U losses f rom the re-con-version and re-enrichment processes. The charge for S/kg U are presently $175/kg U. Transport of the material to
; the enrichnent plant is currently $23/kg U. Price incre-ases in subsequent years are assumec to follow US inflation expectations. .
l
- 4. THE COMPUTER PROGR AM '4 CAST' !
, A computer program written on a HP9845B at Petten and named 4 CAST calculates the annual f uel cycle costs f or HEU a nd EU fuels. The program has been written in a user i
r l
.1 ,
( i ! l !
\.
_g - friendly manner. The input of data and the output of i results are thus in a clear and precise format. The input data is presented in tabular form. Tne
- input items required are of fuel element and control rod data, inflation data and exenange rate data. The form of the taole is snown in Tables I to III. By means or a running cursor, data can be changed or updated directly on the Vtu screen. Unit priceg, indicating yearly infla-1 tion ef fects up to 199 3, are shown in Table IV. Calcul-ations can be carried out for a particular year or a
. number of years, and for any type of fuel. The results are p re sented in taoular or graphical form. Examples of 1 taoular output are given in Tables V to VIII. The cost i items, as described in section 3, are indicated accord-ingly and calculated in the national currency, then
'; ex changed to ECU's, the unit of currency of tne European
- Community. The table shows the total costs in current and fixed ECU, which indicate the change in costs as compared I to 1985 prices (fixed ECU) .
.) Graphical presentation of results are also presented and give an immedia te
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indication and overview of cost j ' burdens in the near future. Typical plots are show n in Figs. 4-14. The results compare LEU to HEU costs against
. years, relative processing costs, relative manufacturing I cos ts, quantity of U-235 required, inflation and exchange
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rate fluctuations. To ennance the results, the 3-dimen-sional graphs have oeen w r itte n with perspective and
]j hidden line routines.
l l S. DISCUSSION CF RESULTS 1 The results of tne HFR fuel cycle cost exercise has been aimed towards comparing the costs of a HEU to LEU
]i loaded core, the increase in costs with respect to future
'i years and the sensitivity of total costs as a result of l changing various parameters, such as, amongs t others,
' l manuf acturing and transport costs, uranium price incre-ases, exchange ra te fl uc tua tions and US S inflation
- rs effects. The basic data used in the current exercise is shown in Table I. The EEU column gives the data p resently n used in the HFR. The 3 columns under tne IEU heading E 'j t indicate the range of variation considered due to uncert-
, ainties in the final design of an LEU f uelled core . In i Tables II and III, the Inflation and Exchange Fate values E 2 are educated guesses based on present day information'and e x pe c tation s .
I i
, _ - - - . _ , - , . . _ , . , . _ - _ - - , . . - _ _ . . . - , _ . . ~ _ , , , . - - - - _ . ~ . . -- - - . _ , _.m . - -
9-
' The various cost itecs, as senematically indicated in Fig.3, are calculated accordingly and presented in tao-ular form' as shown in Isoles V to VIII . The creakocwn of each item in its base cur re ncy and excnange to CCU is clearly shown. The costs in Tables V correspond to tno input items in Tacles I ano III for the HEU data and are shown xor years 19o5, 193 9 and 199 3. Tables VI-VIII show
- the co r re spond ing results for the m in imum , mean and maximum LEU data.
The growth in HEU costs over the coming years is snown in Fig.4. The dip in costs from 1964-1985 is due to the relatively large decrease in uranium and SUU costs during this period. A similar effect is coserved in tne pur suing figures. For the HEU f uel, the annual costs in
. 1985 are 2,932 kECU ( S2.1 million ), show ing tot the following years an average increase of app roxima tely 9%
per annum. The corresponding figures for LEU f uel, based i on the mean data, are 3,303 kECU (S2.4 million), with again a 9% per annum increase for the following years. In Fig.5, the increase in EEU costs for the maximum, mean
, and min imum d a ta are shown. Maximum expectations would not be expected to exceed an extra 10% of the mean costs, whilst minimum costs would be up to 25% lower.
, In comparison, a LEU (mean) fuelled core wculd be
- approximately 12% more costly than the existing HEU f uelled core. A LEU (minimum data) fuelled core, however, indicates similar or even lower costs. The la t te r core
; however is unlikely to oe acnieved as a combination of i all the minimum data would be unrealistic.
l
, A b reakd own of the total cos ts reveals, as in prev iou s studies, that the major f actors are uranium i costs, manuf ac tur ing cos ts , transport costs and repro-cessing cos ts . The HEU analysis givas proportions rela-tive to the total costs of 364 for ur anium costs, (ie.
items 1+ 2 ) , 31% for manufacturing costs, (i tems G +7 +6 ) , 20% for reprocessing costs (item 11) and 11% for tran-sport costs (items 5+9+10+13 ) . The corresponding values i for LEU fuel are 37, 25, 24 and 11%. Any op tim isa tion or 'minimisation of total costs should the re fore concentrate on these items. Hcweve r in
! the case of uranium costs, these are marke t dependent and are more likely to increase than decrease. A 50 % increase in uranium costs, which is not unrealistic, would infer a 18% increase in total costs. In tne case of manuf acturing costs, in ref (1), it was inf erred that a standarcisation i
i
l
; l l
i of the LEU fuel manuf acturing crocess could reduce tce total cos ts. On the ctner hand, until standardisa ticn is achieved, LEC manu f ac tur ing costs are expe c te f to be h ig he r than the present HEU costs. The effect of tne relative total co s ts to the
, rela tive manuf a ctu r ing costs in, for example, 1965 for IEU and HEU fuels is shown in Fig.6. An increase of d=50%
in the manufacturing costs wculd correspendingly irply
. *12% in total costs. The eff ect of yearly inflation is
, j shown in Fig.7, where a 3-D plot of the relative i manufacturing cos ts against years is stown. As the years j p rog res s , the effect of increased costs in LEU is seen to i become much more marked. A similar analysis has ccan carried out for reprocessing costs. Ecwever, this latte r process wculd not be as easy to min imiz e or cha nge in j te rms of costs, as the former analysis. However, new CCE
- announcements on reprocessing price changes are excected si '
soon. Until then, the ef tects of varying the costs are shown in Figs. d and 9.
. It appears therefore tnat LEU costs will remain hig her than HEU costs and may only become less once s tandardised manuf acturing procedures are attained.
I Other items which have cee n analysed include the uncertainties in US inflation and US S fluctuaticn rates. The cost items credited in dollars account for 57% and
, 65% of the total costs in HEU and LEU respectively. For the LEU analysis, graphs are produced showing the e ff ec t
, of 20% variation in US S fluctuation, see Fig.10, and US 3
inflation predictions of 2, 6 and 10 % , see Fig. ll. It can l be seen that dollar increases of 20 % wculd increase total costs of the annual LEU costs by a roughly similar level,
; whilst a 20% drop in the dollar would reduce overall costs oy only 11%. Underestimating or overestimating US l inflation by Wh4%, as indicated in Fig.ll, only becomes O significant as the years progress. In te rm s of 4 year cost e s timate programs, as in Euratom pol ic ies , the effect is not too cr itical.
In Figs .12-14, three 3-D analyses were carried cut to observe possible unforeseen effects from multi-item anal-yses. In Fig.12, the variation in the tctal costs for 1985 with respect to the combined changes in the rela tive number of fuel elements and the relative manu f ac tur ing costs. The overall e f fect is slightly non-linear, indic-
. ating for example that an increase of 20% in required IEU fuel elements increases the rela tive costs by approx-i
imately 15 6. The result of a similar analysis on reprocessing costs is shown in Fig.13. The overall results are ve ry similar. In Fig.14, the amcunt of U-235 per fuel e lem e nt and control rod arc varied and their combinea ef fect en the overall LEU : HEU costs are shown. The increase of U-235 per fuel element is tne more critical 'ut c this is almost entirely due to the 4 times more fuel e leme nt s than control rods used in any year. The increase in U-235 per control rod has little ef fect.
- 6. CCSC LUSICNS Cue to international agreement, the HFB core must consider e ventual conversion to a LEU f uelled core. The effect on the overall annual fuel cycle costs has, therefore oeen assessed. As a result, the cost analysis which was carried out by means of a HFF de velop ed computer program and as reported here , concluces tha t ;
i) in comparison with the present HEU core, tne costs of an exp ec ted LEU core configuration would be greater by approximately 12% and wculd incresso under present inflation and exchange rates, ey approximately 9% per annum,
- 11) the major cost items identified in a LEU core are uranium costs (37%), manufacturing costs (25%),
reprocessing costs (24%) and tr anspo,r t costs (114), iii) possible increases in uranium costs by up to 50 %, would increase total LEU costs by only 0.5% in comparison with HEU costs increases, iv) recent SWU price changes have negligible ef fect on overall LEU costs, v) due to standardisation of the manuf acturing process for LEU f uels, manuf acturing is the only item were overall costs would favourably be reduced. Howe ve r higher manuf acturing costs for LEU fuels will ce main ta ined prior to s ta ndardisa tion and , as a result, LEU total costs in the immed iate future would increase by up to 12% over the present LEU projections presented in Table s II-IV ,
c vi) a 20 % fluctuation in the US dollar with respect to the ECU would reduce or increase LEU total costs oy
*2% with respect to HEU total cos ts, vii) variation in US inflation expectations oy no more than em4% aosolute, would give less enan 14 vari-ation in total costs oetween LOU and hEU cores.
- 7. EEFERENCES
- 1) Ha tos , J . E . , and Freese, K.E. ' Fuel Cycle Cost Comparisons with -Oxide and Silicide Fuels', Inter-national Mee ting on Fesearch and Tes t Feactor Core Conversions from HEU to LEU Fucis, Arg onne ,198 2.
- 2) Matos , J . E . , and Freese, K.E. ' A . Fuel Cycle Cos t
; S tudy with HEU and LEU Fuels',Inte rna tional Mee ting on -
Red uced Enrichment for Research and Test Reactors', Argonne, 1984.
- 3) Krull, W. ' Remarks on the Influence of Enrichment Reduction on Fuel Cycle Costs', Idem
- 4) Pruimboom, H. a nd Tas , A . 'Compariscr. of Thermo-hydraulic and Nuclear Aspects in a Standar1 HEU Core and a Typical LEU Core for the HFR Petten', Idem O
6 4 9 l l l J
- 13 l
Table I
+++ ' ITEM' TABLE +++
T-~~~~~ T~~~~~~~' LEU ITEM HEU Min : Mean : Max No. of FE's p.a. 70 55 66 70 U235 per FE (kg) 42 45 .55 .55 No. of CR's p.a. 17 11 14 17 U235 per CR (kg) .29 .305 .39 .38 Enrichment (*) 4 93 19.75 19.75 19.75 Burn.up (::) 46.7 40 45 50 No. of HUKEM FE's 20 16 19 20 Ho. of CERCA FE's 50 39 47 50 Table II
+++ 'INFLATICH' TABLE +++
.... __.-________....______ y...___~T~~~~~~~T~~~~~~~'
CURRENCY UNITS: YEAR USA : SWU : France : Repro 1984 6 10 10 5 1985 6 10 le 5 1986 6 10 10 5 1987 6 10 10 5 1988 6 le to 5 1989 6 to 10 5 1990 6 10 to 5 a 1991 6 10 le 5 1992 6 te le 5 1993 6 to 10 5 Table III _ ___ __ -_- --___ __ _ ___ ..___ ___. .... ._... ... ..... __ __ __.
+++ ' EXCHANGE RATES' TABLE +++ -
____.~T~~~~~~~T~~~~~~~T-~~~~~~~ CURREHCY UNITS: 1 ECU = ... YEAR DM : FF : Hft : . USs 1984 2.34 6.86 2.52 .73
, ,1985 2.23 6.82 2.52 .72 1986 2.22 6.87 2.52 .72 1987 2.2 6.88 2.51 .71 1988 2.19 6.89 2.5 .71 1989 2.18 6.9 2.49 .7 1990 2.17 6.91 2.48 .7 1991 2.16 6.92 2.47 .69 1992 2.15 6.93 2.46 .69 1993 2.14 6.94 2.45 .68
8 j* 8 3 792 3 3 6 5 6 3 559 7 4 45. 58 6 3 8 6 8 227 3 9 28 6 5 946 9 7, 8, 8, 3, 7, 4,6,2 8 4 3 3 6 9 262 5 3 5' 2 3 5 1 1 0 8 5 2 437 56 9 9 2 7 e 5 5 2 283 5 5369 6 6 9 7 4 46 3 4 946 9 27
. 9 3, 6, 2, 6, 6, 6, 4, 5, 2 8
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ssssssss5 ss5 ss5 s s s s s s s 3.s s s s s s s No ! TEM U$8 Dn f FF f6ECV l 3381 kg. Un at (Hetto) 152,184 ! 1 -- -- 211 2 4451 Shu (Hetto) 600,999 -- -- 835 . 3 Conversion of 3341 >g. Un at 21,932 -- -- 31 4 Cylinder EenteFill Costs. (Sus 14,d28 -- -- 20 5 Tr an s por t of UF4 to Ewrope. -- 33.?eo -- 15
- 6 28 Fuel Elements erom HUKEM.
~
415.30@ -- tid i ? 17" Control Socs from NUAEM. -- 6@6,733 -- 2*2 a 50 Fuel Elements from CE2CA. -- -- 3,017,500 4e2 9 Teansport to H.F.R. Retten. -- 33,?00 -- 15 14 Return used Elements to USA. -- 74?.350 -- 317 11 Reprocessing of 3?9 kg. (U+Al) 417.494 -- -- 530 12 Re-conversion of 22 kg. U. 4.009 -- -- ( 13 Tr an s por t UF4 for Enetchment. 52? -- -- g 14 Tot al per Year, Current ECU's. 2, 732 15 Fised ECU's. 2.469 H.F.R. Petten i Forecast of H.E.U. Fuel Cycle Costs for 1999 sssssssssss,sssssss,ssss,,,,ss Ho ! TEM U$s DM FF fhECU 1 3388 kg. Unas (Netto) 'J2,104 -- -- 217 2 4451 $Wu (Netto) 879,923 -- -- 1,257 3 Converston of 3301 kg. Unat 21,982 -- -- 31 4 Cyltnder RenteF611 Costs. (5x) 18,464 -- -- 26 5 T r ans port of ,UF6 to Europe. -- 42,545 -- 24 6 20 Fuel Elements from NUKEM. -- 524,939 -- 241 1 ? 17 Control Rods from HUKEM. -- 765,947 -- 354 4 50 Fuel Elements from CERCA. -- -- 4,417,922 640 9 Tr ansport te h.F.R. Petten. -- 42,545 -- 20 14 Return used Elements to USA. -- 993,444 -- 414
!! Reprocessing of 379 kg. (U+Al) 507,953 -- --
726 12 Re-conversion of 22 kg. U. 5,061 -- -- T 13 Transport UF6 for Enrichment. 665 -- -- 1 i 14 Tot al per Ye ar, Current ECU's. 3,947 15 Flued ECU's. 3,?$7 s l N.F.R. Petten Forec ast of H.E.U. Fuel Cycle Costs (or 1993 ssssssssssssssssassssssassasss Ho ITEM U5s CN FF mECU f 1 3381 kg. Unas (Netto) 192.129 -- -- 213 2 4451 $wu (Hetto) 1,298.295 -- -- 1,495 3 Conversion of 3381 kg. Un at 27,752 -- -- 41 4 Cylinder Rent,F611 osts. (Sm> 23,315 -- -- 34 5 Transport of UF6 to Europe. -- 53,?t3 -- 25 6 29 Fuel Elements from HUKEM. -- d62,?22 -- 314
? 17 Control Rods from HUKEM. -- 967,441 -- 452 8 50 Fuel Elements from CERCA. -- -- d,464.279 132 3 Tesnsport to H.F.R. Petten. -- 53,713 --
25
. 10 Return used Elements to USA. -- 1,128,285 --
527
!! Reprocessing of 379 kJ. (U+AI) 6t?,420 -- --
946 12 Re-conversion of 22 kg. U. 6.390 -- -- 9 13 Tesnsport UF4 for Enetenment. d49 -- -- t . 14 Tot al per Ye ar , Current ECU's. 5,442
, 15 Femed ECU's. 5,121 Table 5
F~
- 16 e
H.F.R. Petten Forec ast of L.E.U. (Min.) Fuel Cvele Costs foe 1945 ssssssssssssssssssssssssssssss No ITEM U$s DN , FF kECU l L 2447 kg. Unas rHesto) 144,320 -- -- l 150 2 3193 Swu (Hetto) 430,450 -- --
. 594 3 Conversion of 2407 kg. Un at 15,646 -- --
22 4 Cylinder Rent < Fill Costs. (5x) 14,629 -- -- 20 5 Tr an s por t of UF6 to Europe. -- 33,704 -- 15 6 16 Fuel Elements (com HUAEM. -- 332,640 -- 149 7 11 Control Rods from NUxEM. -- 392,592 -- t?6 9 39 Fuel Elements from CERCA. -- -- 2,353,650 345 9 Teknsport to H.F.R. Petten. -- 33,700 -- 15 10 Return used Elements to USA. -- 707,950 -- 317 11 Reprocessing of 400 kg.-(U+AI) 440,692 -- -- 612 12 Re-converston of 129 kg. U. 22,716 -- -- 32 13 Tr an spor t UF6 (or Enetchment. 2.986 -- -- 4 14 Total pee Year. Current ECV's. 2,456 15 Femed ECU's. 2.443 H.F.R. Petten Forec ast of L.E.U. (Min.) Fuel Cycle Costs or 198) 5 sssssssssssssssssssssssssssss No ITEM U$s l DM FF lmE;U 1 2407 k g. Un at (Netto) !@S,329 -- -- 155 2 3184 SWU (Netto) 638,222 -- -- 900 3 Conversion of 2447 kg. Unat 15,646 -- -- 22 4 Cylindee Rent / Fill Costs. (5x) 18,464 -- -- 26 5 Tr an s por t of UF6 to Europe. -- 42,545 -- 20 6 16 Fuel Elements from NUKEM. -- 411,950 -- 193 7 11 Control Rods from HUKEM. -- 495.639 -- 227 8 39 Fuel Elements from CERCA. -- -- 3.445,979 499
, 9 Transport to N.F.R. Petten. -- 42,545 -- 20 10 Return used Elements to USA. --
493.644 -- 414
!! Reprocessing of 400 mg. (U+AI) 535,664 -- -- ?65 12 Re-conwersion of 129 kg. U. 28,678 -- -- 41 13 Tr ans p or t UF6 for Enricnment. 3,769 -- --
5 14 Tot al per Ye ar, Current ECU's. 3,244 15 Fixed ECU's. 3.149 s h H.F.R. petten F seec ast of L.E.U. (Min.s Fuel Cycle Costs for 1993 sssssasssssssssssssssssssasss s No ITEM U$s DM FF kECU l l 1 2407 k g. Unas (Netto) 136,752 == -- 201 2 3108 $wu (Het t o) 922,707 -- -- 1,35? 3 Conweeston of 2407 kg. Un at 19,753 -- -- 29 4 Cyltnoer s ent< Fill Costs. (Sm> 23,315 -- -- 34 5 Transport of UF6 to Europe. -- 53.713 -- 25 6 16 Fuel Element s from HuxEM. -- 534,174 -- 14% 7 in Control Rods from HuxEM. -- 625,732 -- 292
, 9 39 Fuel Elements from CERCA. -- -- 5,445,254 ?2?
) Tr an s por t to H.F.R. Petten. --
53.713 -- 2* 10 Return useo Elements to USA. -- 1.128,205 -- 527 11 Reprocessing of 400 h 4 (U+AI) 651.103 -- -- 953 12 Re-conversion or 129 hg. U. 36,206 -- -- 53 13 Tr an s por t UF6 (or Enetcnment. 4,758 ,
-- -- 7
~~
14 Total per Year, Current ECU's. 4.444 15 F6=ed ECU's. 4.216 Table VI
. - 17 -
H.F.R. Retten Forecast of L.E.U. (Mean) Fuel Cycle Costs for 4945 sssssssssss sssssssssssssssssss No ! TEM U$s 3M FF .ECU 1 3975 kg. Unat (HestoJ 178.333 -- -- 244 2 5264 $WU (Hetto) 780.708 -- -- 16? 3 Conversion of 39?5 ag. Unat 25.839 -- -- 3i 4 Cylinder Rent. Fill Costs. (5m) 14.d29 -- -- 10 5 Tr an s por t of UF6 to Europe. -- 33.706 -- 15 6 19 Fuel Elements from NUKEM. -- 395.014 -- 177 7 to Control Roos from HUKEM. -- 479.dd3 -- 22'4 6 47 Fuel Elements (com CERCA. -- -- 2.43d 450 4td 9 Tr an spor t to M.F.R. Petten. -- 33.749 -- 15 10 Return useo Elements to USA. -- 747.450 -- 317 11 Reprocessing of 519 kg. (U+A14 571.373 --
-- 794 12 Re-conweeston of 191 kg. U. 33.453 -- --
46 13 Transport UFd for Enrichment. 4.397 -- . -- d 14 Tot al per Y e me . Current ECU's. 3.303 15 - Ftmed ECU's. 3.23d N.F.R. Petten Forecast of L.E.U. (Mean) Fuel Cycle Costs for 1939 i A
, sssssssasssssssssssssassssssss a
i No ITEM U$s On FF kECU 1 3975 kg. Unat $HettoJ 178.843 -- -- ;56 2 5264 $WU (Hetto) 1.040.548 -- -- 1.44d 3 Conversson of 3975 kg. Un at 25,439 -- a 37 4 Cyltnder Rent < Fell Costs. (5x) 14.468 -- -- 2d 5 Transport of UF4 to Europe. -- 42.545 -- 24 6 19 Fuel Elements from NUKEM. ** 499.691 -- 229 7 14 Control Rods from NU6EM. -- 634.813 == 299 j 8 47 Fuel Elements from CERCA. -- -- 4.152.34d d42
, 9 Transport to M.F.R. Petten. --
42.545 -- 2e
! 16 Return used Elements to U$A. --
893.d44 -- dia 11 Reprocessing of 519 kg. (U+RI) 694.547 -- -- 192 12 Re-conversion of 191 kg. U. 42.233 -- -- da 13 Transport UF6 for Enetenment. 5.551 -- -- 3 14 Total per Year. Current ECU's. 4.435
, 15 Flmed ECU's. 4.254 t
H.F.R. P e t t e'n Forecsat of L.E.U. (M e an) Fuel Cycle Costs for 1993 ( sssassssssssssssssssssssssssss No ITEM U$s DN FF lkECU
," 1 3975 kg. Unat (Hetto) 225.835 -- --
332 2 5264 SWU (Hetto) 1.523.466 -- -- 2.244 3 Converston of 3975 kg. Unat '32.d21 -- -- 48 4 Cyltader Rent < Fill Costs. (S=) 23.315 -- -- 34
> 5 Tr an s por t of UFd to Europe. --
53.713 -- 25
- f. d 19 Fuel Elements (com NUKEM. --
d29.536 -- 294 7 14 Control Rods from HUKEM. -- 796.387 -- 372 0 47 Fuel Elements from CEsCA. -- -- 6,404.!J2 47d 9 Transport to M.F.R. Petten. -- 53.713 -- 25
. 10 Return used Elements to U$A. --
1.124.245 -- 527 11 Reprocesetng of 519 kg. (U+Ali S44.179 -- -- 1.248 12 Re-conversion o( 191 kg. U. 53.318 -- --
?$
l 13 Tr an s por t UFd (or Enescnment. 7,409 -- -- 10 14 Tot al per Y e ar . Corrent ECU's. d.405 15 Fixed ECU's. 5.736 Table VII
I
, H.F.R. Petten s.:r e c na t of L.E.g. (M aa. ) Fuel Cycle Costs for 1)s5 5 5 sss ssssss sssssssssss ssss ssss No l TEM U$s DM l FF fmECU l 1 4706 kg. Unat (Hettos 211.399 -- --
234 2 6226 $ku (Hetto) 844,582 -- -- 1.167 3 Conversion of 4704 k3 Un as 30,648 -- -- 43 4 Cylinder Rent / Fill Costs. 55m) 14,628 -- -- 20 5 Tr an s por t of UF6 to Europe. -- 33.748 -- 15 6 28 Fuel Elements tech MunEM. ** 415,300 -- t$$ 7 17 Control Rods (Poh HuAEM. *- 606,733 -- 272 3 50 Fuel Elements from CERCA. -- -- 3,417,500 44; 9 Tr an s por t to H.F.R. Petten. -- 33,700 -- 15 la Return used Elements to USA. -- 707.450 -- 317 11 Reprocessing of 568 kg. (U+AI) 6t? 052 -- -- 457 12 Re-conveeston of 203 kg. u. 35,693 -- -- 50 13 Transport UF6 roe Enrichment. 4,691 -- -- 7 14 Total pee Y e ar . Current ECU's. 3,646
, 15 Fixed Ecu's. 3,612 H.F.R. Petten Forec ast of L.E.U. (Maa.) Fuel Cycle Costs for 1969 8 s.s s s s 8 sssssssssassssessssssas No ITEM U$s 2M FF l=Ecu 1 4700 kg. Unat tHetto) 211,399 -- == 19?
2 6226 $wu (Hetto) t,238,696 -- -- 1,756 3 Conversion of 4700 kg. Unas 34,684 -- -- 44 4 Cvitneer Rent < Fill Costs. (5m) 10,468 -- -- 26 5 Transport of UF6 to Europe. -- 42,545 -- 20 6 20 Fuel Elements from HuKEM. -- 524.930 -- 241 7 17 Control Roos from HukEM. -- 765,997 -- 251 0 50 Fuel Elements from CERCA. -- -- 4,417,922 640 l 9 Transport to H.F.R. Petten. -- 42,545 -- 29 l 10 Return used Elements to USA. -- 493.644 -- 480 11 Reprocess 6ng of 560 k g. (U+AI) 758,031 -- -- t,a?! 12 Re-conversion of 263 kg. U. 45,962 -- -- 64 13 Tr an s por t UF6 for Enricneens. 5,922 -- -- 4 14 Tot al pee Ye ar, Current ECU's. , 4 )5? 15 Flued ECU's. 4,755 H.F.R. Petten Forec ast of L.E.U. (Mae.) Fuel Cycle costs for 1993 ssssssssssssssasaeasssas ssssss
.4 No ITEM U$s LM FF =ECu i 1 4708 k g. unat (Netto) 267,517 -- --
373 2 4226 SWU (Hello) 1,891,$62 ** ** 2,65@ 3 Conversion or 4700 kg. Unas 38,641 -- -- 57 4 Cyltader Rent < Fell Costs. (5m) 23,315 -- -- 34
'l 5 Transport of UF6 to Europe. -- 53,713 --
35 6 20 Fuel Elements from HuKEM. ** 662,722 -- 310 7 17 Control Rods from HuKEM. -- 967,@41 ** 452 3 50 Fuel Elements from CERCA. -- -- 6,444,2?9 932 1 Tennsport to H.F.R. Petten. -- 53,713 -- 25 t@ 8etuen used Elements to USA. -- 1,128,285 -- 527 11 Reprocessing of 564 m g. (U+AI) 111,667 -- -- 1.341 12 Re-conversten or 203 kg. u. 56,889 --
-- 94 13 Transport UF6 for Enricnnent. 7,477 -- -* 11 14 Tot al pe r' ve ar, Current Ecu s. 6.441 15 Fined ECU's. 6,424 Table VIII
. 99 l
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. -c. - Fuel Cycle Cost Factors HFR 1) kg Unat required per annum requirement (+2.5% process losses) Spot 2) puri:hase Unat ($/kg Unat) market l U rat
. Conversion 3) conversion costs ($/kg Unat) olent 4) cylinder filling & renting costs , l UF 6 Enrichment 5) enrichment costs (SWU/kg U) ; olant enriched UF 6 6) shipping costs (fi-d p.a.)
Manufacture 7) manufacturing costs : (Euroce) i) NUKEM (DM/ Fuel Element) FE's and CR's ii) NUKEM (DM/ Control Rod) gg lii) CERCA (FF/ Fuel Element)
- 8) transport costs (fixed p.a.)
spent fuel 9) shipping costs (fixed per trip) Reprocessing clant (USA) 10) reprocessing costs ($/kg(U+Al) ret.) UNH Re-conversion
- 11) conversion costs ($/kg U in UNH) l UF 6 Re-enrichment 12) transport costs ($/kg U)
A plant I . credit U & SWU - Fig.3 Flow Diagram for the Calculation of the HFR Annual Fuel Cycle Costs
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i kECU's 8000 . 1 . . . 1 . . . 7BBB - 1 LEU max. GBBB -.---.';--'...............,......:.'...... . . LEU mean 9000 : -i--- - - r s LEU min. w- ;- - -- AB00 -l-- 3000 ----:---.--.
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.5 1 1.5 2 2.5 3 FIG 6. LEU /HEU Total Costs vs. LEU /HEU Manufacturing Costs, for 1985 e
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4fu $r. DEVEIDPMENT OF A 3-DIMENSICNAL CAICUIATION MODEL OF 'IHE DANISH RESEAROI REAC'IOR DR3 'IO ANALYSE A PIOPOSAL 'IO A NEN CDRE DESIGN CALLED RING-CORE Erik Nonbel* Rise National Laboratory 1 DK-4000 lbakilde, Dennark ABSTRACT A 3-dimensional calculation model of the Danish research reactor DR3 has been develeped. Demands of a more effective utilization of the reactor and its facilities has required a more detailed calculation tool than applied so far. A great deal of attention has been devoted to the treatment of the coarse control arms. 'me model has been tested against mea-surements with satisfying results. Furthermre the model has been used to analyse a proposal to a new core design called ring-core where 4 central fuel elements are replaced by 4 dunoy elements to increase the thermal flux in the center of the reactor. INTBODUCTICN A project was started two years ago with the purpose of devel-oping a 3-dimensional calculation model of the Dansih research reactor DR3. Demands of more effective utilization of the reactor and its facilities had required a more detailed calculation tool . than applied so far. It conprises calculation of burnup reactivity, flux distribution in different irradiation tubes and study of new fuel element designs. Finally the DR3-staff wanted to reduce the ntaber of their rather troublescue flux measurements. 1
=;
a ,- 2 DESCRIPTION OF T!iE REACIOR DR3 he DR3 reactor is a 10 M heavy water cooled and moderated research reactor with highly enriched U-Al fuel elements. Be reactor is of design similar to the British " Pluto" type and its operation began in,1950 (1).- Se DR3 design is illustrated in Fig. 1 showing a horizontal cross section through the reactor. Se core consists of 26 fuel elenents, each1.one entains four mncentric aluminim clad fuel tubes, which are arranged to provide a 5 on centre hole for exper-iments.
- Se reactor core and heavy water are mntained in an alminium tarA of 200 cm diameter. Vertical and horizontal test holes of 10 and 17.5 on diameter are located in the radial D20 reflector. Out-side the alminim tank is a 30 cn graphite reflector with 10 on vertical test holes.
.me reactor is controlled by means of cadmim absorbers. Mere are seven coarse control arms and one fine control rod. Be coarse control auns move like signal acns between the rows of fuel el-enents. me fine control rod and two safety rods are situated in the corners of the core.
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7V) n=9 7V2 FCR 1 -ll
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'If_b ,~2 V E~, w . NId2
, s Fig.1 A horizcntal cress section through the reactor
_ , - ,_ - _ _ , _ _._ ,..__...m.- --.. _ _ - _ . . _ . _ . . . _ _ , . , , , . , - . _ - , .
"* = 3 3 DESCRIPTIN OF THE LODEL DR3/SIM 2e model is based on diffusion theory with 3 energy groups.
- 7 * (p g(r)Veg(r)) + E gg(r)$g(r) =
3 3 . I E gg (r)$g,(r) + Ax 9E,(vEg)g (r)$g (r) s e e 9*9' 9 g = 1,2,3 where 4 eg(r) is the neutron flux Ag(r) is the diffusion coefficient Egg (r) is the removal cross section Egg' is the scattering cross section Eg(r) is the fission cross section xg is fission spectrun 4 v is neber of neutrons produced per fission A is the reciprocal effective multiplication factor 1 - 2e equations are solved in 3 dimensions using finite-difference technique. Se discretization is done according to the usual 7 point scheme and successive overrelaxation is applied together with different accelerating devices (2, 3). When using finite-difference technique the mesh size has to be anall in order to obtained accurate flux values. Sus with a mesh size of acout 7 an the total amount of meshes for the whole reactor ocuprices about 27.000. Se mesh size is variable to facilitate the positions of the different rig tubes. Furthermore the cubic material meshes may be divided into two prisms by mean of diagonal planes to describe the circular circisaference of the reflector boundaries. ,a em- -
4 Representation of the Cross Sections he cross sections of the fuel elements are given as function of fuel type and burnup. Bey are calculated frtan UKNDLr-73 library with 76 energy groups as basis applying a collision probability code called CCCMO (4) for the detailed burnup distribution in the four concentric fuel regions. Finally the fuel cross sections are homogenized and condensed to three energy groups with the following , energy limits: l l Group 1:. 15.0 MeV - 5.53 kev
> Group 2: 5.53 kev - 0.625 eV Group 3: 0.625 eV - 0.1 meV 2e DR3 staff measures the thermal flux by means of Co-wires
- and the fast flux by means of Ni-wires. Berefore the microscopic l cross sections of these two materials are calculated in the center i
of the fuel elements taking disadvantage factors into account. 2ese cross sections are also given as functions of fuel type and burnup. The progran CCOO is also used to calculate the cross sections of the different rigs and their irradiation materials. l' Aepresentation of the Coarse Control Arms CCA me seven coarse control arms made of ca&nitun that move like signal arms between the rows of fuel elenents, have caused a great trouble to represent in the model. Soy move from an angle of 300 with a vertical plane at the start of a reactor cycle to an angle
! of 600 at the end of a reactor cycle.
I A model has been chosen where the cross sections of the control arms are added to the cross sections of the adjacent meshes which varies with the. coarse control angle CCA. 2-dimensional calcu-lations of the following type have been done. hgenW CCA-absorber fuel 1 / - - - . .
\ ! ! ! i ! s 2 s 2 s l 2
- : - s
! k h
. gCCA 1CCA 1%CCA 2CCA 6
5 2e homogenized 3 group cross sections of a fuel elenent with a 1/2, 1, 1 1/2 and 2 control arms as neighbours are detemined. Rese calculations are conpared with the corresponding calculation without control ams. Because the threshold value of Cd is within the third energy group only this group is considered. A. subtraction gives the third group cross section of the control arms for differ-ent ntaber of control arms. S e results show the following variation of the CCA-cross section: {f=K-(0.4*x2 + 0.5 x) (3.1) where K is a mnstant and x is the ntaber of adjacent control arms , of the mnsidered mesh. Sua the resulting third group cross section
! of a mesh with control arms is represented as
[f'"[3 +[ and the corresponding diffusion mefficient as Dps , 3jg jjp1 + 3.jp) In the model the intersections between the CCA-cutlines and the mesh grid are calclulated as function of the CCA-angle. Applying this it is possible to determine the adjacent area occupied by CCA-arms for each :nesh in the z-y plane. Bis area relative to the total z-y area of the mesh concerned is substituted for x in ex-pression (3.1) and thereby determines the absorbtion cross section of the coarse control arm to be added to the mesh absorbtion. VERIFICATICN OF 'IHE MX)EL 2e results frca calculations with DR 3/SIM have been ccupared with flux measurenents made by the DR 3 staff (5). Below is shown sane of these conparisons. Fig. 2 shows the thermal flux in axiel direction in a fuel ele-
, ment and fig. 3 the corresponding fast flux.
Se results seem to agree very well with the measurements except for the fast flux where there .is a discrepancy of about 10%. Still it is within the limits of the error frce the measurement method.
1 j.
. , \
4 6 I
! THmm. rwx cwen a.io i4) OR3, FL.EL ELEMENT C3 3.00
!' e--. ==de w I ! anowe num
>- +
. . . am .
.au =
a.00 i :
, i
.i .
x
* 'f c
s.00 w F :
-t , i
. l i.
s J
- -s0 -30 0 3o so FI.g.2 01stmx rum ccsE czrera pure (m)
I
, i rasr rum twen a.iO is) OR3 FUEL ELEPENT C3 4.00
' . e--e **d==
?
- causwa rum I : ..e.
3.00 ! 6 : 3
.i :
. 2.00 l
~
; / i a
l g,oo e ,
.i, :
e
-e i
i 0.00
+ -so -30 . 0 3o $o Fig.3 0:stsect ram coac cmr% cvec cen wm o*** .
7 CORE DESIGN CAIGLATICNS Se model has been used to study a proposal to a new core de-sign where the 4 central fuel elements B3, B4, C3 og C4 are re-placed by 4 dunny elements without fuel (6,7). We U235 content of the 22 remaining elements is increased to give the see total
- amount of U235 in the reactor. 'Ibe purpose of the new core design was to obtain a greater thermal flux in the center of the core.
In fig. 4 is shown a crmparison of the axial thermal flux in fuel position C3 for 3 different cases:
- 1) Old design with 93% enriched fuel
- 2) Old design with 20% enriched fuel
- 3) New design with 20% enriched fuel It is' clear how the thermal flux is reduced from case 1 to case 2 (15%) and increased from case 2 to case 3 (50%).
Fig. 5 shows the thermal flux in a horizontal plane in the cen-ter of the core for this new design called ring core design. By conparing with fig. 6, showing the flux for the normal design, it is quite clear how the thermal flux in the center of the core is increased. Finally fig. 7 and fig. 8 show the corresponding fast fluxes. It appears clearly how the renoval of the 4 central fuel elements causes a " hole" in the fast flux. n(Rwo Flux (4/Cu 2*t0**le) CR3 FUEL ELEMENT C3 3.00 CLD CCSos (132) CLO CC50s (2cs) l Reso CORC CC50s (202) l Pow 0l=1Cuw. CCA=tt l 2.00 -' V
,,co /
. ./ -
-50 *30 0 30 50 CtSTANCC FRou CCPC CtNTRAL PWC (Cu)
Fig.4 A ccmparison between old and new core design. wm-- - ,
8 THERMAL FLUX IN THE CORE OF DR3 RING CORE DESIGN Z(CCP)=0 CCA- 18 n O'
~
m h[ c a.
, \
l- . 2*
*;.. s e -
hA a? 1
. Fig.5 THERMAL FLUX IN THE CORE OF DR3 PERIODE NO 303 Z(CCP)=0 CCA- 18 m
O
-j
- q M
g 'o s e,: - 1, - s N O- . o O
,e th.
".. .7 i
~% ,, .-
Fig.6
g ._. __ _ _ _ -. . _
- o. .
9 FAST FLUX IN THE CORE OF DR3 RING CORE DESIGN , Z(CCP)=0 i CCA- 18
;[ 0- -
l 3 .
's u Go' , .
kU -
/
eg .
/
1 / P u ."-
. ~a i
% %* e
-,7 r
Fig.7
.i
-l FAST FLUX IN THE CORE OF DR3
'l PERIODE NO 303 Z(CCP)=0 CCA- 18 1
f I
; u
! w -
, O
=~
4
]w -
.j EN .
9, w . -,.
= #
l
~
*=0*
%. ., ..s i
Fig.8
,i..,,, 10 CONCWSION Although the verification of the nodel still is in progress, it e seems as if it is possible to obtain reasonable results with DR j 3/SIM. Reactivity calculations on critical mnfigurations show agreement within 1%. Se intention of using the model for burnup
; calculations flux distribution calculations and reactivity calcu-i lations thus appear to be possible.
However the model of the coarse control arms needs to be veri-fled more carefully before DR 3/SIM can be used as a reactor simu-lator. i i REFERDiCES
- 1. Reactor Irradiation Services, AEC Rise.
- 2. G.K. Kristiansen, DC4 to-day, RF-memo-183 (1973) .
t
- 3. K.E. Lindstran Jensen, Developnent and Verification of Nuclear Calculation Methods for Light-water reactors, Rise Report No.
235 (1970). l l 4. C.F. Harjerup, he Cluster Burn up Program CCC and a (bmparison
! of its Results with NPD Experiments, Rise-M-1898 (1976).
1
- 5. K. Falcon Nielsen, Reactor Run no. 300, Rise-I-172 (1984).
, 6. K. Haack, Means of Preserving Irradiation Qualities after Con-version to Iow Enriched Uraniun Fuel, IAEA-SR-119/26 (1985) . ! 6. E. Nonbel, Preliminary Calculations of Ring-core-design, RP-06-85 (1985).
i i i
.i r, o,.n.. se.e. cem., a.
C. CRf?fCAL EXPERIMENTS OF JMTEC VEU CCPES (Hi S. Shimakawa. Y. Nagaoka. S. Kotke. K. Takeda. B. Komukat and R. Oyamada Japan Atomic Energy Research Institute Carat Research Estabitshment Oaral-mas.1 tbarakt-ken Japan A8STRACT Critical experiments in the Japan Materials Testing Reactor Critical factitty * ( JMTRC) with medium-enrichment-urantum (MEU. 45%) fuel elements have been carried out. . The purposes of the expert-ments are to obtain nuclear characteristics and to validate neutrontes calculation performed by SRAC code system used for analyzing the JuTR MEU core. This paper describes the results of experiments, such as reactor kinettes parameters, shut-down margin and vold coefficient following the previous paper pres ented at RERTR meeting . 1983. The calculated results are in satisfactory agreement with the measured results. It is indicated that the changes
'of nuclear characteristics due to the core converston from the HEU to the MEU core gt,ve no serious problem from the viewpoint of reactor safety.
INTRODUCTION The Japan Materials Testing Reactor Critical factitty (JMTRC). a 100 W swimming pool type critical factitty and moderated and cooled by light water. has been operated as neutronics mock-up for JMTR (Japan Materials Testing Reactor.50MW tank type)." Critical experiments have been carried out in the JMTRC with medium-enrichment-urantum (MEU. 45%) fuel element.1. The purposes m
E O 9 2 of the experiments are to obtain nuclear characteristics and to validate neutrontes calculations performed by SRAC code system 2) us ed for analyzing the JMTR MEU core. The some results of the experiments. such as Crit 1 Cal mass excess reactivity. Control rod worths and flux distributton were presented at the international l meeting of RERTR. 24-27 October.1983. Tokat-mura. 2)
! This paper describes the results of the experimen*.s such as l reactor kinetics parameters A eM/ 2 ( AeM : effective delayed-neutron fraction and 2: ; prompt-neutron life time ). shut-down margin and vold coefficient following the previous paper. In order to compare nuclear characteristics of the MEU and previous high-enrichment-uranium (HEU.90%) core. the experiments we.~e also i
carried out in the HEU core prior to the expertments in the MEU core with the same configuration. The validity of the neutronics calculations is confirmed by the experiments. in both the MEU and HEU core. t 9 t CORE CONFIGURATION The core configuration is a dupiteate of the JuTR core. i.e.. number and arrangement of fuel elements and control rods. the beryllium matrix etc. as shown in Figure l. .The fuel region consists of 7 by 5 lattice spaces. each 7.72 centimeter square. These spaces contain the 22 standard fuel.41ements. 5 control rods with follower fuel sections. and 8 aluminum experiment holes. The horizontal cross sections of MEU fuel elements are illustrated in l Figure 2. Surrounding the fuel reston is a reflector reston I containing a number of beryllium and aluminum experiment holes. The core has 5 mock-up loops including two mock-ups of hydraulic
~
rabbits. The control rods are made from borated stainless steel containing 1.6 wt-% national boron. For the JuTRC experiment. 31 MEU fuel elements were fabricated without any significant change in dimensions and shapes from those
' of the HEU fuel elements. The uranium density of fuel elements and U-235 per element are summartzed in Table 2. The JWTRC fuel l
elements were fabricated. such as three kinds of standard fuel elements (A. B. C). in order to simulate the equilibrium core of the JMTR. PEACTOR KINETICS PARAVETERS (Bedd / f t ) j The kinettes parameters ( AeH/ft) were measured by the pulsed neutron technique, as a ratto between effective delayed-neutron fraction ( AeM) and prompt-neutron life time (as). 1.e.. Prompt-neutron decay constant at critical t ac).
*e 3
The outline of the experiment system using the pulsed neutron technique is shown in Figure 3. An instantaneous pulse of neu-trons, which is generated at an accelerator assembly. Is injected i into a subcritical core and ensuing flux of neutons is measured by BF3 counter. And the decay constant at a subcritical (a) is defined as the time constant of a fundamental prompt-neutron mode. as shown in Figure 4. The a was measured at vartous control' rod pos i tions . And a: was obtained by extrapolating the data of a to that at tne control rod posttton at critical. The accelerator assembly was located at 4 lattice spaces (G-3. G-4. H-3 and H-4) in the core shown in Figure 2. The BF3 counter was set up at K-12. and the measurements were also carried out when the counter was set up at F-12. in order to check whether the data depended on counter positions in the core. I Figure 4 shows the a versus control red post tton from critical i pos i t ic,q. The reactor kinetics parapeters ( Aeff/22) measured are til sec in the MEU core and 103 see in the HEU core. The value in , the MEU core is about 7 % larger than that in the HEU core. SHUT-00wN MARGtN e . The shut-down margio, was measured b,y. two.dt f f erent methods : .y . .. the pulsed neutron technique cutilned above and the rod drop method.. . The__measucements. by.--_the pulsed neutron technique were .. . _. carried out at three different locations of the BF3 counter. t.e.. F-12. K-12 and C-14. In the core shown in Figure 1. And the shut-down margin is determined by following equation." 0 = (1 - a/ac) / (1/4eff - a /ac ) where o is reactivity and A sff 13 effective delayed-neutron fraction. On the other hand. in the* measurements by the rod drop method. SF3 counters (R-1. R-2) were located at B-14 and Q-1. In
, the core shown in Figure 1.
The data by two methods are shown in Table 2. In the measure-ments of the shut-down margin. since the pulsed neutron technique is less dependent on the BF3 counter positions than the rod drop method. the pulsed neutron technique is superior than the rod drop method. Therefore the data by pulsed neutron technique have been used as evaluated data. The shut-down margins obtained are 14.0 l %4k/k in the MEU core and 16.4 %dk/k in the HEU core. The margin in the MEU core is about 2 %4k/k smaller than in the HEU core. l but is still enough as the criterton is 10 %dk/k in the JuTRC and the JMTR. l e o l
1 4 VOfD COEFFICIENT in the measurement of vold coefficient. the void was simulated by inserting aluminum plates into the core. It has been confirmed by the neutronics calculations that a discrepancy between aluminum and void effects is small enough to be ignored. In this experiment ten aluminum plates ( 2-mm-thick. 60-mm- width, and 840-mm-height )
. were inserted vertically into water-gaps of the standard fuel l
! element.
The results'of void coefficient measured at various positions
, are shown in Table 3. As shown in Table 3. the locations of the i
fuel element with aluminum plates were symmetric with respect to the column "I" in the core shown in Figure 1. There is almost no difference in the vold coeff!ctent between the MEU and HEU core. ' l VAlfDITY OF NEUTRONICS CALCULATIONS l The neutronics calculations were performed using the SRAC code I system. which was developed in Jcpan Atomic Energy Research
} Institute t JAER! 1. In this study, the lattice calculation to generate group constants was performed by colliston probabil.ity
- method, and the cort catrulation was-- perf ormed- by 3-D diffusion -
code. The kinetic parameters. Eeff and As. were calculated by means of the perturbation theory. The few group energy' structures were the 4 groups tupper energy boundarles : 10 MeV. 1.0 MeV. 1.83 kev. 0.6823 eV). f i Table 4 shows the results of the measured and calculated f excess reactivity, the control rod worth 3. the shut-down margin and i the void coefficient. As shown in Table 4. the calculated excess
', reactivities are 0.3 %4k/k and 0.6 %Ak/k higher than the measured ones in the MEU and HEU core. respectively. The calculated control ! rod worths agree with the measured ones within 0. 4 14 k/k per control rod. The calculated shut-down margins are 1.3 %dk/k and l 1.8 %4k/k higher than the measured ones in the MEU and HEU core.
respectively. The void coef ficient agrees by 0.001 %dk/k/ nut % bet-
! ween the measured and the calculated values.
3 Table 5 shows the results of measured and calculated reactor i kinetics parameters. As shown in Table 5. the calculated 4eff/2:
! are 13-15% larger than the measured ones. And the changes of the . . parameters due to the core conversion from HEU to MEU fuel are about 6% decrease in the 2 and no significant change in the Seff.
- l. The reason why the prompt-neutron life time is smaller in the MEU
; core than that in the HEU core is mainly because of increased e urantum-235 loading in former core.
l Table 6 shows the neutron flux changes due to the core ! i l
- l
z - f 5 converston. The calculated thermal neutron flux (<0.68eV) is in satisfactory agreement with measured one. and the decrease of thermal neutron flux agrees with the measured ones within 2-5 %. , And the changes of calculated fast neutron flux (>t.0MeV) due to the core conversion are almost nc difference in any reston. CONCLUSION The MEU core with the larger urantum loading, is validated by the JMTRC experiments resulting in the feature as follows. The reactor kinetics parameters 4eff/2s is 112 sec"' in the MEU core, which is about 7% higher than in the HEU core. h - ( The shut-down margin is 14.0 %Jk/k in the MEU core, which is } 2.0-3.0%dk/k smaller than that in the HEU core. l ! The void coefftetent is -0.01.2 %Jk/k/wid-% in the MEU core. l There is no signtitcant difference in the void coefficient between the HEU and the MEU core. i It is indicated that the changes of nuclear characteristics l due to- the core converston- from the HEU to MEU core .stve no -- serious problem from th,e viewpoint of reactor saf ety. l Concerning the va11dity of the neutrontes calculations, it is confirmed that the calculated results are in satisfactory agree-ment with the measured results. t.e.. the differences are: 0.3-0.6
%Jk/k in excess reactivity. 0.4 %4k/k in control rod worths.1.3-1.8 %dk/k in shut-down margin. 0.001 %dk/k/etd-% in vold co-l efficient. 13 - 15 % in kinet1Cs Parameters and 2 - 5 % in flux distribution.
l l l ! REFERENCES
- 1. Diviston of JWTR project. "JWTR trradiation Handbook.* JAERI-M .
83-053. Carat Research Estabitshment. JAERI(19831
- 2. K. Tsuchthashi. H. Takano, K.Horikant. Y. Ishtguro. K. Kaneko and T.Hara. "SRAC: JAERI Thermal Reactor Standard Code System for Reactor Design and Analysts." JAERt-1285(19831
- 3. Y. Nagaoka. K. Takeda. S. Shimakawa. S. Kotke and R. Oyamada.
" Critical Experiments of JMTRC MEU Cores.* RERTR Proceeding (1983) l l 4. B.E.Stamon and J.S. King.:Nucl. Sct. Eng. 3, P.595 (19581 l
l i ___M
6 IO ij i i I i I I I I I I I i l l lm ""
, I I I i i ! l ;
a I I I - l I l i i I l 1 I l g
. I i I . . i. c I i i a i M 101 101 101 IOi i I I i 1 9 I I l o t 1.101 1,5 i,:Oi 1. i !c! I I i
. 'l i 61 J l l 1.bH I.N 1.1 f.I l l l l i
' 'i i l iOI l.51.0 f B I,1 iO J l I '
'i I i J t.l 1. 8 l.t H f.I i O I ! I i g W f ici r 101 1.8 1.101 1.f J o J l ii i i I JOi f0f !Of I Ed J f d I i i i !O! JOI 10! !fJ l
.u t t l l l l l l l l l ( l l l ,
'a @l I i ! I l l ! l ! I l l li e
Q . a., e us.nn
, , , _ o .
- w i. si O ==
- sa s. mi o...
a s.i
,,%. c O ==~
- i. =>
- i. .
,u=-..
l Fig. I JMTRC Standard Core J i I yg_ gee oc g1 a 20.io 38 '
- 4. 0a kI! ,
l
.- 0 a 882 8
e"4 'L8
**;)
4 i ,ee L jQ3 ) lUCO3
'50 *p 2# o e : y- :
a a e o i ,n a g m We
" i N i
li - a -
~
m :i _
>g o l$1 i,~.i., i.,~,. a , :
3 1g follower fuel section standard fuel element E na *, Fig. 2 Horizontal Cross Section of JMTRC MEU Fuel Elements L
C e 7 i l l i F as , j 4 ..s6. 1 F .a f e.e.. .. . . de h )
! l e r N -
c -... ; i c.. g e s ., ! i ***.a* . . g <
= 2_w. _ %.
i
- 1* :: %.v 1 i
- y. .~.
i.._.,. ....... 3 : Hu. e . . .:. r
..I . . , , , , .
., u.s r. -
i
- 11 a=== A ll
.j,,,,,,,,, "
a u s .a e-a, . .. .... e ie as se t 7'" , ti .r... t. . .. 4 s . p. l twric Fig. 4 Counter Responce versus Time
. . Fig. 3 Outline of Pulsed Neutron after injected of Neutron Pulse
- Technique System into Subcritical Core I i i i ISO -
g'- g# n0 -
",s- -
i00 - - # . ,, _ o
- f. i:0 -
.e - - -
o* o e llo p#e
*---* M E U 100
'J-*-o HEU
$0 - ..
ss ., I l l I e la 10.0 15.0 200 Centrol Red Insettien Lenetn (n) Fig. 5 Decay Constant (a) versus Contro!
. Rod Position from Critical Position
l l 8 ; l l Table 1. JMTRC Fuel Element 1.oading l l Kind of Plates per Uranium U Element Element Density. 9/cm 3 Content. 9 M l
, Standard fuel l element A 19 1.6 310 B 19 1.4 280 l C 19 1.3 250 l
- Fuel follower
. element 16 1.6 205 HEU Standard fuel I element A 19 0.7 279 B 19 0.6 237 C 19 0.5 195 Fuel follower element ,
16 0.7 195 l 1 f l Table 2. Shut-Down' Margin l l Method & MEU HEU 40 %dk/k BF3 Counter Position %dk/k %dk/k IMEU-HEUI Rod drop method B - 14 11.0 12.5 -l.5 Q-1 13.0 15.5 -2.5 l Pulsed neutron technique K - 12 13.7 16.8 -3.1 l
' F - 12 13.4 16.1 -2.7 l C - 14 14.7 16.4 -1.7 ! , Average 14.0 16.4 -2.4 l
l l l l 1 l l
9 i Table 3. Void Coeffletent f Location of Fuel MEU HEU + l Elements wIth
%Jk/k /wid % %4k/k /voi4 %
Aluminum Plates t 2 2 H-8&J-8 -0.97 x 10 2
-1.02 x 10 2 G-9&K-9 -1.39 x 10 2
-1.55 x 10 -
F-8&L-8 -1.59 x 10 -1.80 x 10,2, 2 F -10 & L -10 -0.72 x to ~0.68 x to * ' Avera9ed value -l.15 x 10 -1.22 x 10' l ., 1 I l Table 4. Calculated Exess Reactivity. Control Rod Worths. Shut-Down Wargin and Vold Coef ficient. Compatring with Measured Ones l l WEU HEU l Weasured Calculated Jo Weasured Calculated -Ja tot e s int-wi p . I Exess reactivity 11.2 11.5 +0.3 10.0 10.6 +0.6
%dk/k Contrl rod worths EJk/k SH-l & SH-2 11.3 11.7 +0.6 11.7 12.5 +0.8 ,.
! SA-l 3.1 2.9 =0.2 3.2 3.1 , *0.8 > SA-2 S.9 6.0 +0.1 6.3 6.4 +0.1 SA-3 3.4 3.2 -0.2 3.4 3.J 0.1 Shut down margin 14.0 15.3, +1.3 16.4 14.2 +1.8 1 1
%Jk/k Vold coefficient =0.012 =0.013 +0.001 0.012 0.013 -0.001
, 1Jk/k / ee % l l 4 i i l l i ; l , I
Q .',',. 10 Table 5. Calculated Kinetics Parameters. Ef f ective Delayed-Neutron Fract ion Deff and Prompt-Neutron Lif e Time Is. Compatring With Measured Ones MEU HEU
. Measured Calculated C/M Measured Calculated C/M 4eff/fe, sec lit 125 1.13 103 118 1.15 D eff -
0.00766 - - 0.00766 - - As. usec - 61.1 - - 64.8 - Table 6. Calculated Thermal Flux Changes by Core Conversion from HEU to MEU Fuel. Compatring with Measured Ones Measured Calculated Thermal neutron flux Fuel region -8 ~ - 12 % -8 ~ - 13 % Be reflector region -l ~ -3 % 0 ~ -3 % Fast neutron flux Fuel and Be reflector - +2 ~ -2 % region l
i - y,
\f :
. O p o r IV3, )y 1
i f
.p
~i
.i DEVELOPMENT OF HEAT TRANSPER PACKAGE FOR CORE THERMAL-iiYDRAULIC
, DESIGN AND ANAt.YSIS OP UPGRACED JRR-3
; Yukio SUDO, Hiremasa IKAWA and Masanori XA.tNACA f
Japan A:cmic Energy Research Institute (JAERI)
! Tokat-sura, Ibsgaki-kert,. Japan .319-11 ,
. Abstract A neat transfer package was developed for the core
. thermalanycraulic design and analysis of the Japan
$ Research Reac:ce-3 (IRR-3) whica is to be remodeled :o jt a 20 .We pool-type, ligat water-cooled reactor with 20 %
low enriened uranius (LIU) plate-type fuel. This paper
-i presents :ne constitution of 33e developed acat transfer package and the applicability of :ne neat transfer correlations adopted in it, based on the neat transfer
, experiments in wnica thersal-nydraulic features of :ne new JRR-3 core were properly reflected. ,
I INTRODUCTION At 23e Japan Accaic Energy Researca Institute, the J2 pan Research Reactor-3 (JPR-J) is now under remodelling into a 20 .Wt 7 light water sederated and cooled, teryllius and heavy water
,- reflected, pool-type rescior with 20 $ low enriened uranium plate-type fuel. Por a reference, Pigure 1 shows a schematic diagram - ! :ne resc ct including a core, a light water primary cooling sys*. a and a agavy water ecoling system. *he major features of tae JRR-3 for the taersal-nydrau11: sesign and safety analysis are as ts11cws.
(1) *wo modes are adopted for core ecoling under normal operation, one is a natural-convec ton cooling vita uptiow in :ne : ore for icw j : ore power up to 200 K'4 and ne other is a forced convec:Lon : col-i ing for aign core power up to 20 W. Flow airec-ton in tae : ore
; for :ne forced-convection ecoling sede is townward fres :ne e
2 viewpoint of attenuation of l'N. With downflow in the core at the normal operation. a core flow reversal should occur to remove the decay heat af ter scris with upflow in some of operational transi-ents and accidents assu=ed in the safety design. (2) Two major design criteria have been set up for the core thermal-hydraulics so that fuel plates may have enough safety margin for the condition of normal operation. One is to avoid nucleate boiling of coolant any'<here in the core and the other is to keep the minimum ONBR not less than 1.5. This paper introduces the heat transfer experiments in which key thermal-hydraulic features of the new JRR-3 core were properly reflected and presents the constitution of the heat transfer pack-age for the core thermal-hydraulic design and safety analysis. II HEAT TRANSFER IXPERIMENTS FOR DEVELOPMENT OF HEAT TRANSFER PACKAGE
- 1. Experimental Rig Pigure 2 shows a schematic diagram of the test loop which is used to develope the heat transfer package for the new JRR-3 Any of upflow and downflow can De selected in the test.
The test section is composed of a flow channel, a lower plenum and an upper plenum. The configuration of the flow channel which is composed of adjacent two heating plates is rectangular with 50
; mm in width. 2.25 =m in water gap and 750 mm in length. The heat-ing plates are made of Inconel 600 with 1.0 nm in thickness.
- 2. ONB Conditions The ONB temperature has been determined by the following two simultaneous equations in the thermal-hydraulic design under normal operation with forced-convection.
2.33 3
; q = 1. 7 6 x 10-s , pt.tss[ ,(7,g , 7s If ' P (1) t
! q=0.023Rj'8 Pa.. , , {(7 7) g , (7s*IIf' b III Fquation (1) was proposed by Bergies and Rohsenow III for water, and Eq. (2) was proposed by Dittus and BoelterI2) for the forced con-tion single-phase flow. Por the evaluation of the margin of fuel surface temperature against the ONB temperature. the precision of i Eqs. (1) and (2) should have been made clear. The precis *.on of Eq.
(1) was not always clear for the application to the subchannel of the JRR-3 because the amount of the available data was small for the forced convection downflow though some data for upflow have t been reported.I3'I)
; Pigure 3 shows the comparisen of the experimental results ob-t tained for both upflow and downflow in this experiment vita Eq.
; (1), with respect to the relationship of 1 'te 47, at the onset of l <
nucleate boiling, le can te pointed out in this figure that no l 4 significant difference between upflow and downflow is observed with [ 1 l
3 respect to the relationship of q vs a?, at the onset of nucleate
, boiling.
Figure 4 shows the ecmparison of the ex erimental results including the available existing dataI1'3*"* I with the predictions by Eq. (1) in order to evaluate the error of Eq. (1). This figure clearly indicates that the error of Eq. (1) is about -1 K against the lower limit of the measured superheats at the onset of nucleate boiling.
. 3 Forced-Convection Single-Phase Flow
- The problem is the dif-forences in the single-phase forced-convection heat transfer char-acteristics between uptiow and downflow for a vertical retangular channel simulating a JRR-3 subchannel. It is considered that there are no differences in the single-phase forced-convection heat transfer characteristics between apflow and downflow for a high velocity at the normal operation condition of the JRR-3 The ef-
, facts of buoyant force should, however, enhance the heat transfer for upflow and on the contrary, decrease the heat transfer for j downflow at low velocities, which should occur during the opera-1 tional transients and accidents.
Figure 5 shows the comparison of the experimental results for Re larger than 1 Dittus-Boelter(2}000 , with the existing Sieder-Tate (6) andcorrelations proposed Colburn(7) for by turbulent forced-convection heat transfer. The characteristics of the turbulent forced-convection heat transfer for narrow vertical flow channel are su=marized for both j upflow and downflow as follows. (1) Sy use of equivalent hydraulic diameter for rectangular channel, any existing heat transfer cor-relation of Dittus-Boelter, 31eder-Tate and Colburn is available for both upflow and downflow even in the channel whose gap is as i narrow as 2.25 =m. (2) No significant differences were coserved i between upflow and downflow in turbulent forced-convection heat transfer characteristics ard therefore, it is considered that there
; are no effects of buoyant force for the turbulent flow.
] Por Re less than 2,000 at which the flow is laminar Pig. 6
, shows the comparison of experimental results on Nu for upflow with
! . those for downflow. I5 is obviously recognized that Nu at downflow are smaller than those at upflow for Re less than about 700. Upper l limits of Nu correspond to Nu at the inlet of channel (x = 0 m) and lower limits to Nu at the end of heating plates (x = 0.75 m). For almost the same Re, Nu decrease with the increase of distance x
, from the inlet of channel for both upflow and downflow.
In order to make clear the effects of buoyant force, Pig. 7 is
; presented by adopting Grashof number Gr for the abscissa and Nu/Nuo
! for the ordinate. Here, Nuo is the Nusselt number obtained for the rectan alar channel heated from both sides fica the analysis by
, Hwang l ) et al. The effects of buoyant force are evident in the
'I
? downflow for Re less than about 700 and Or larger than aoout 1.000.
3 ONB Heat Plux The major features vf CNB heat flux to be
, predicted for the JRR-3 subchannel are; (1) The flow channel is rectangular with a narrow gap of 2.26 :s and is long with a large
a i length-to-hydraulic diameter ratio 1/Ce of 170, (ii) :he coolan: velocity range of interest is fairly wide ranging from downflow to upflow, and (iii) the coolant conditions are at low pressure and low temperature. A=ong a small amount of experiments with rec-tangular channel, the, study conducted by Mish1=al3.10) is most Un-portant in the viewpoint of application to the JRR-3 subchannel. But, the errors of :he correlations proposed 5y Mishima and the conditions under which the correlations can be applied have not
, always been made clear primarily because the quantity of experi-mental data is very small. .
Figure 8 shows all the available CNB heat flux data obtained for u flow both in the rectangular channels and in other chan-
, nels. 11) The bold line gives a correlation of qCNB vs G* shown below.
q$NB,1=0.005C'8*"I (3)
. Equation (3) is correspondent to 1.5 ti=es of the lower limi:
of the data, and is recommended as a CNB heat flux correlation for upflow at not very small G* not only in the rectangular channels but also in other channels. Figure 9 shows all the available CNB heat flux data obtained for downflow both in the rectangular channels and in other channels. From this figure it is clear that qCNB for downflow both in the rectangular channels and in other channels are lower than those for
, uptiow, at low G of about 2 % 400. On the other hand, it appe ars t
.{ that qCNB is almost the same between uptiow and downflow at high C, i- larger than 400 and at very low G including zero.
Figure 10 presents the scheme of CNB heat flux correlations proposed in this study for upflow and downflow. The scheme pro-posed in this study are composed of three correlations. Eqs. (3), (4) and ( 5 ) . Eqs. (4) and ( 5 ) are expressed as follows. A 1h. .
# ""' U" (
4CNB,2 * "[{ ((
- QCNB,3 = 0.7(A/AH)(W/Af / f1'I Y /g Y1) (5) j for upflow and downflow The errors of these correlations were evaluated in the viewpoint of safety design and analysis. Equation ( 3) is a new correlation for predicting the CNB heat flux for u flow, and Eqs. (4) and (5) are correlations proposed by Mishima.( p0) Por downflow, there tre three regions, Regions I, II and III. In region I, the CNB heat flux is predicted by Eq. (5), which is the same as the correlation predicting the minimus CNB heat flux for upflow. In Region II, .
l the CNB heat flux is predicted by Eg g (4). wnich gives much lower heat flux than Eq. (3) at the same G . In Region III, the CNB heat flux is predicted by Eq. (3), which is the same as the correlation for upflow. l
5 III CONSTITUTION OF MEAT TRANSFER PACKAGE - (1) Single-phase forced-convection flow: For a downward flow. the following were acopted depending on Reynolds numeer Re. Nu = 4.0 for laminar flow (Re < 2000) , (6) Nu = 0. 02 3 Re * * 'Pr ' *
- for turbulent flow (Re g 2500), (7) and Nu is evaluated by interpolation with Eqs. (6) and (7) for transition region (2000 g Re < 2500).
For an upward flow in a natural-convection cooling mode, the 4 following were provided taking into account the effect of buoyant force. Nu = max (Eq. (6), Collier correlation] for laminar flow (3) (Re < 2000 ) ,
; where Collier correlation is given as fo110ws.(12) i
; Nu = 0.17 Ref
- 3 8Pr[**3{Prf/Pr)L/*{gSD(7,-Tl#V y
3 b }h'l (9) i' The scheme of heat transfer correlations for single-phase forced-convection flow described above is applied for both liquid flew and steam flow. (2) ONB Temperature: Eq. (1) is adopted for the correlation pre-dicting the onset of nucleate boiling (ONB) temperature both for upflow and downflow. k (3) Nucleate Boiling Heat Transfert The following correlation
) which was developed by ChenI13) are adopted for the heat flux e under the subcooled and saturated nucleate boiling.
q = 0. 02 3 { R,g( 1-2 ) f ' Pr' + * (TW~I) g
- P i
. o.rs s.es s.es .
! I ' E 3 0.00122 (Tv-Ts )
Pw -P,)#8
,0'*8 u{**'h[g**o*2'g ,
(10) l
*he factors P and 3 are determined depending on flow conditions (subcooled boiling or saturated boiling, and so on). It has been confirmed that Eq. (10) can be applied also for downflow, from comparisonwiththeexistingavailabletestdatafordownflow.I{he #3 (4) ONB Heat Plax Scheme: A scheme of CNB heat flux correlation for upflow and downflow is composed of Eqs. (3), (4) and (5).
(5) Transition Boiling Heat Transfer and MSPB Temperature: In the heat transfer package. the following correlation vnich was proposed by Bjonard and Oriffith(15) is adopted for the neat flux in the transition boiling regiont . 4 = dq0NB
- II*$I9MSFB' *h8f' $"( ) - (!1)
,ONB"'MSFB f
r i I
6 The minimum stacle film toiling (MSFS). temperature is definec by the following correlation proposed by Henryll5) , T3373 = Tgy * (TEN-T ) (ykCp );/(ykC p)w . wnere 733 = 324*C (12) (6) Film Soiling Heat Transfer: The following modified 3remley correlation is adopted for the film boiling heat transfer!143; kfy(yg-y)hrg'(1-2) g g I!* q=C( ) (T.,-Ts I ' ug le(Tv -I S I (13) Gr (I<*Ts ) $ c )1/2 where hgg,= hfg{1+05 t. Ae .= 27 ( yg,yg 3,E and C = 0.62. Table i su==arizes the constitution of heat transfer package developed for the core thermal-nydraulic design and safety analysis of the new JRR-3 IV CONCI.UDING REMARKS
,. A heat transfer package was developed for the core thermal-
; hydraulic design and safety analysis of the new JRR-3 With the heat transfer package thus developed, the core thermal-hydraulic design has successfully been carried out.IIII (Nomenclature]
As Plew area, Ag Heated area, C: Constant, C: p Specific heat
! Dte Equivalent hydraulic diameter. C: Mass flow rate, G- Di-mentionless mass flow rate a G//Ayg tyg-yg)g, g: Acceleration of gravity, hgg , Latent heat of evaporation, dht : Inlet suDcooled enthalpy, kt Thermal conductivity, Nu: Nusselt number, P; Pressure, Pp: Prandt1 number. 1: Heat flua, q'. Dimen-sionless heat flua = q/hrg/Ayg lyg-yg)g, R,: Reynolds numeer, 7: Temperature, AT,1 Superheat, x: Cuality, W: Width of channel, 3: Void fraction, St Volumetric expansion factor, y: Specific weight, or Density, V: Kinematic viscosity, ut Dynamic viscosity, A: Critical wave length for CNB heat flua
= (c/(yg-y )]'/8 Ar g e Critical wave length for film boiling g
8, c Surface tension *
= 27(c/(yg-)y (Suescript )]8 5 bulk, CNB: departure from nucleate boiling, f: film, g steam, RN: homogeneous nucleacion, 1: water, MSPB: mini-sum stable film coiling, CNB: onset of nucleate boting, 3: saturated, W: vall
r,* . 7 i I References i
. (1) Sergies, A.E., et al.: Trans. ASME Ser. C, 86-3, 365 (196a).
(2) Dittus. P.W. , et al.: Univ . Calif . Pubs . Eng. , 2, 443 (1930). l (3) Clark, J.A., et al.: Trsns. ASME. 76-2, 553 (195.4). (4) Hada, M.: Technical Memoranda. ISNSE, ANL, Oct. (1958). (5) Sato, T., et al.: Trans. JSME, 29(204), 1367(1963), (in Japanese). (6) Sieder E.N. , et al. a . Inc. Eng. Chem. , 28( 12 ) , 1429 (1933). (7) Colburn, A.P.: AICHE J,. 29,' 174 (1933). (8) Hwang, C-L., et al.: Appl. Sci. Res., Section A. vol. 13.
. 401 (1964).
(9) Mishima, K. , et al. : Int. Mtg. on RERTR, Tokai, Japan, Oct. (1983). !^ (10) Mishima, K. : Soiling burnout at low flow rate and low pressure conditions, Dissertation Thesis, Kyoto Univ.
- -j (11) Sudo Y. , et al.
- J. Nucl. Sci. Technol., 22(3). 202 (1985).
1
; (12) Collier, J.G.: " Convective Boiling and Condensation",
j McGraw Hill (1972), New York. (13) Chen, J.C.: ASME paper NO. 63-HT-34, (1963). ! (14) Sudo, Y., et al.: JAERI-M 84-066 (in Japanese). l (15) Sjonard, T.A., et al.: Symposium on Thermal and Hydraulic l- Aspects of Nuclear Reactor Safety, vol. 1, light water
- - reactors, pp. 17-14 ASME. . .New York (1977 ) .
l ,j (16) Henry R.E.: AIcht Symposium Series, 138, vol. 70, pp. 81-90 (1974).
,5, (17) Sudo. Y., et al.: J. Nucl. Sci. Technol., 22(7), $51 (1985).
Li f . l l i L 1 . l \
- I k
3 j - t
- --,u,..,
' *-- == = u 'i ;*;a;;;' -, e.4.w cA, T-
,=
- '. \ :=, u ., . - - - ,
, ;y --m _N_1 - . w
'- EE;,
~
l =%. \ Iy f
.- -f -
"N 9=- N I 'l lT N
'...s~,.. h'$ (l g: I 7# .. ] . 3! I g @i, n is g.',. _ . / i l
- ,s ,.
.. a yp- r-
. , ), , i
.I
~-
- ii /
Ijfp M,- s . .
\=.n.,W .. ] !t } Pta
, .x / p w .
I .- 7 e .-..
. 5
.,e r zw
.CD 4.
. dam.
i Pig. 1 Senesatic diagram of Pig. 2 Seneestic slagram of the upgraded JRR-) for experimental rig used core cooling for development of neat a transfer sacrage 4 xm a . . .., u . . . ,
$ n lnO n es l4__ ,,
3 i @ m, ,p *-a ' 50 ,
- f. ,,t N l5l*" @ -
_. 20 . p.i . e - se.m - J !! 4'l 8
**f*
; I0 s . - % ,s Ni .e,. .,
- I
~,
- t -
4 K y s - e.
*C i
'e e., er a 8':
s
~
~
i e w ei e I 'I ' e s . 2 -
* =
.. ws e,
' i
.
- I 2 S 10 20 50 l C y'ucrc :a= (K3 to. I i 8 4 30 SC Pig. e comparison netween the sensured c g . ;.. i. m ana greatetes superneste at tne CNR wita given neat flus Pig. ) Heat fluz vs superment et the CNR in downflow and upflew ,
o
rm 2
'C8 - to.rf -
'p. ( =a ) t:) ..
l ~ n- . Cze .
. $.M ze .
fdl'['i ,
'x,nr 1,.. -
.s,c. .1 i
'*I'#-i_$ g 1 's , A%, gJWPf 2
l -- iC 3 tC* I 8 3 2 45 90 10' to3 3 5 to* a.i-i a, i_, ,,, i , Pig. $ comparison of tae esperimental resulta vita :ne estating correlations. (a) Dittue-Boelt er's. ( t ) Steder-Tate's and (c) Colturn's for tursulent fisw on single-onase forces-convection neat transfer enaracteristice ( Ae 3 1000)
.1 .
)
l t -
,.on _ *
~ * "'
WM? io' - 1.p , * ,. W #, j._ .9r Op->g,g3l ei e' n
,,,y
. Ls ' " \,'Q* Q '
N c.s }. %.ca . % j $g ' (2$04Ae<40Ci 1 - ' O2 g4 gga
- 10"
& t-1 10g 10 1 2 5 gol 2 Pig. f gf rette of Grsenof Nuncer on
$4 ( ~) Nueselt MusDer in single-passe forced-convection lastnar.
Fig. 6 Cooperison of Nu vs Pe sownwars flow Setween upflow and dewnflow few Ae less taan 2000.
e . ,
, e '
c* . s 10
-t,- .
g
'J'*Ww
(' *j ,l:
,,,[; ~
/. ' ,
i i Ocm#ww 34+133
/
/
.4 t
, k, 14. ( 13
- f
- 16[ --
. ) . [
- s f 1 10
- 4 4
.) ,eA - -
- ,, .qgg,1 -
i- ,6'
.b
/ 4g3d [ ,
g ;
' I I !
'0 -
0 :S T -
- 8. .
to- ic- 'O 0 15 so' tc
- io. if G*4-1 g*ii Fig. 8 hpertaental results af DNS Pig. ) h pertsental,results of ONS neat flu for upflow neat flu for townflow i
Y
/
- l 0 >
$ $ Pig, r0 1 9' g e *
*/
*/ s Illustration of the proposed 8 0 ',
ao
, p <
seneee of DNa neat fl u correlations for upflow and to . ( S {g e dowerlow 3
- % .n og l~
** ***'"' J t 0*'0 t so ,oa i g, . ,C'
- 0 * (~*)
Taele 1 Constitution of tae neat transfer package proposed for the design act analysis of sne new'JRP ) gaat transf er sees e Gewa rl sw i verlo, Sang &c. Laannar ta.Iet ta !8) Passe 'rama t t i an !*tereelatteg
# tad 'urmulant 'a t (Stttam-teeltere 3me t - *ature i 'a f>! tiarelas.aaassaows eve&asse i hasselae ' i.a ' iqs toestftaa cwas teiltaa I laturatae t ia i10' '0%eae
.e , eat n.a ,.,,<o ,,n ..a,o ..s. oam. .a, < o . n u .,,, .
l
";;=;- l se . . . . ...., e .4, , , n . .
}
*.a a a *~ se.,i, ......,
1 i
,,e ta...
, o . ,. 1, a- o ,,,, ....,f .. ,,,,e..
..-..a..i ,, . . . 6. .. . e . . . . . . . . e . n ..
masaa fisw ! p.m- ,
I dE Differential thermal analysis and metallographic examinations of Unsippowder and Unsip/Al(38w/o)miniplates Palle Toft Arne Jensen Atlas-Danmark A/S Advanced Engineering Division I ABSTRACF The paper describes the thermodynamic and metallurgical behavior of U3Si2 Powder and miniplates with regard to the following aspects:
- increasing temperatures op to 1400 C
, - tuel plate stability
- formation of tornery intermetallic compounds
, - growht of the formed ternery intermetallic crystals at increasing temperatures.
The miniplate core is U 381 2/A1(38w/o) and the cladding is A1 6061. The study concludes, that fuel plates will not remain stable under accident conditions. A fuel element will most likely melt down before a stable structure is established.
! The work is performed under contract with RIG 4 National
; Laboratories, Denmark.
I e i 1 t
'P
INTRODUCTION The aim of the work is to examine the thermodynamic and metallurgical behavior of U3 Si2 Powder and U 3 Si2 /Al (38w/o) miniplates at increasing temperatures. 1 The powder and miniplates were manufactured at Argonne National Laboratories, Chicago, USA. EXPERIMENTS l DTA examinations (Differential thermal analysis) Samples of U3 Si2 Powder and of U 3 Si2 /Al (38w/o) mini-plages are heated and cooled at the same rate
, (20 C/ min) in a gTA apparatus. The maximum heating 3 temperature is 1400 C. The samples are heated in helium i or in air. Solidification / liquidation temperatures, oxidation and phasetransformations are registered on a
, plotter as endothermic and exothermic reactions.
i Metallographic examinations
,i Metallographic examination is carried out on U3Si2/Al a
(38w/o) miniplates before and after heating in DTA. The composition of the intermetallic compounds are analysed by EDAX technique. Preparation of samples j Two samples of U3Si2 powder weighing 700 mg each are O mounted in Al203 crucibels. Samples of U3Si2/Al (38w/o)
,! miniplates are cut to the following dimensions: 7 mm x j
- 2. mm x 1.5 mm. The cutting surface of the core is during the DTA runs exposed to the ambient atmosphere.
j The samples are mounted in A1203 crucibels. As referen-
; ce is used Al2O3 Powder.
Treatment of samples ( sample no.)
- 1. DTA on U 381 2 Powder heated to 1400 C in helium.
~
] 2. DTA on U 3 812 Powder heated to 1400 C in air.
4 l 3. Metallographic examinanation of the U 3 Si 2 /Al-l (38w/o) miniplate core before heated in DTA. EDAX analysis of phase compositions. j - t i 1 t
m 4 i
- 4. DTA gn U 3 Si 2 /Al(38w/o) miniplate heated to 1400 C in helium. Metallographic examination.
EDAX analysis of phase compositions.
- 5. DTA gn U3 Si 2 /Al(38w/o) miniplate heated to
, 1400 C in air
, 6. DTA on cladding material A1 6061 heated to 1400 C in air.
7,8,9 DTA on U 3 Si2 /Al(38w/o) miniglate heated to
, 700 C, 900 C and 1100 C in helium.
Metallographic examination. EDAX analysis of phase composition. EVALUATION OF RESULTS U Mis powder samples SAMPLE 1: OTA ON U,51, POWOER HEATED TO 1400'C IN HELIUM r exaramw No EN00/ EXOTHERMIC REACTIONS i
"o m >N ..__ _____
- b. . x%,s. m, anoornaw i
I Sample 1: The U 381 2 Powdgr is thermodynamic stable when f heated in helium to 1400 C s __ , w --a 'W^" "'
. 1 1
4-SAMPLE 2: DTA ON U,514 PCWCER MEATED TO 1400*C IN A!R J
- 1. CIIDATIO's CF U 351 2 POWDER Exortnu l
1 l i. Ed* C 95b'C N I f f f f I N f f f I t t t I t
; 4 6 6 4 4 IN 4 1 4 l l g g g l
I i
; acorunu Sample 2: A powerfull exotgermic reaction occurs in the temperature range 150-950 C caused by oxidation of the U3Si2 Powder. The oxidation occurs mainly in to steps, at approx. 300 C and at approx. 630 C, both silicon and uranium oxides are probably formed.
U Mip/A1(38 w/o) miniplate samples igt, -ro.g.
% ,.. ~ 1q.v -- g 3 ;;.<;.y3,,_.j.g;~g~c . 4 g :.; ;g.s -
.' ~'
3:.;C,* '*.G;*,W,'QL
~f *. . e. , "s*f ;, . f,' _ , .- r . .' .*Q='O $$*be '( * . ..* ' *
. y l*-f ~_alj.* 'l Sk
.,. 0*d,?~ $? ' T, "' '
;?'
. . 1. ' . '
l ~ l l 2 l I ! p' :;*.. \;1. s.- l
- y. ;, , ,y- *,.; . ' . s, .* r .. 2 .' .-
,'^..
'.f. .: .'K.? 'h* h:g; . (M :p!Qt..';.'g',',#Ql:
'. .e
- ?,;;J . :. *- ...- -
- . . .. . x,.
- g. .% s :st .*, ; p f_Jr. ..', y,,i ';* ?-Q' g.?,a
~'st....s' L. ,r a , . s t,g. ,,
- e. .
n e
- Micrograph 1 (x100)
[ l
~.
.O Sample 3: The metallographic examination (micrograph 1) of the miniplate core before heating in DTA shows that the core material consists of U3 Si 2 (1) Particles embedded in a pure aluminium matrix (2). The EDAX analysis gives the following composition: U Si2 Par-ticles: U 90.3 w/o (52.2 a/o), Si 7.3 w/o (335.6 a/o).
Matrix material: 99% A1. f SAMPLE 4: DTA 04 Ui$13 /A1(38w/.) MIN! PLATE MtAtt0 TO 1400*C tm NElluM i trotram 2. 610*C 1. ALUMINIUM 5 MELTlhG POINT
- 2. ALUMINIUMS SOLIDIFICATION Polmi l
1 t t t t t t t t t t t I t t li t 1, i i i i ( i i 6 6 6 6 1 i ls 1
- s. .
a
- 1. 640*C i
GN00Tinu l 2 i f Sample 4: The endothermic (1) reaction is due to i melting of both the cladding material and the matrix material in the core. The exothermic (2) reaction shows
; the solidification of the micro segreated aluminium.
The metallographic examinination of the sample shows that there is an reaction between U3Si2 and Al at or immediately above aluminiums melting point. The ternery intermetallic compound which is formed has
! according to EDAX analysis the following composition:
U 72.1 w/o (22.9 a/o), Sir 6.6 w/o (17.6 a/o), Alt 21.3 w/o (59.5 a/o).
I .k ^' N E a ._
'f%,& f- f.
M-- 7 a .
&. ke& s . %
.:. .q g .r ,s, .
3 .
}?g. ~ -
3' t k' '
' ~
^
AfiM@f?W.2~u.
'. , . y
?
\.l'l'w,;
E . ' . ' . ') *? -h .
,,,.g!. p . _ s nJ';
- 17.c.
n
. ;, :. 4 _ 3., .,1.s c , ....--
- 1
.hhk; r . . \3.e.d.
y; . . . ,(1N[tf - a/ h_
- x'.1)*$$r% kA N!Ns li Micrograph 2, 1400 C (x100)
Micrograph 2 shows the ternery intermetallic compound as a big regular crystal (1). The micrograph shows also an eutectic phase (2) (fine and coarse laminated pha-se). This ternery phase consist probably of the above described ternery phase and micro segreated aluminium (3). SAMPLE 5: DTA ON U351,/A1(38v/) MIN! PLATE HEATED TO 1400*C IN AIR
- 1. 440*C 1. 0110AT10N OF U3512 IN TME CORE ExonAAf 2. ALUMINIUM 5 M(LT!NG P0!NT
- 3. ALUMINIUM 5 5OLIDIFICATION POINT f
" 2. 640*C t
I f g- - ,ld 1 Nt t l e I t t t t e t t 6 6 4 3 S 6 i i 4 & s 1 # i
\ 3. 610*C sucorsano
i Sample 5: The exothermic reaction (1) at 440 C is due to oxidation of the core surface (rafering to the preparation of samples). The oxidation products is probably silicon oxides and uranium oxides. SAMPLE 6: OTA DN CLA00!NG MATERIAL A1 6061 NEATED TO 1400'C [N AIR
,,,,, 2. 610*C 1. ALUMINIUM 5 MELf!NG POINT
- 2. ALUM!N!UM5 50L:0!FICAf!04 POINT
(_v _N - _ "_
.i i j t
i [ fi A _Yt i C 1?
.I. 1_. ,.
I 6 l iN 4 6 6 i i 6 3 I il s N k
- 1. 640*C w oorsau Sample 6: The oxidation of U3Si2 in the core is confir-med by the DTA examination of the cladding material Al 6061 who does not show gny exothermic reactions before the melting point at 640 C.
~< a m e;7.- p -~ w *~+ .' ;-
- wif,.. s .
, yo- j;i/.C-yhM W 8??F,
't
. n. .x <s.; , = ,,1 ' , ., s,.. %g:: 'g,
, ..y.- .e ..v-'
dMIq;45 f 9' sn ~g -,
.y
%. sigh;rk:j@M.R?; 2 ..
r.= 4. 4 a.a .a. w.eg *
%mv C. MRV %n 7 g&-
apmq A 8 m Q% htl,f
,M - -
r Ex w.= ,
.k, ; ' .. a? .?.. .
1
,3 g c g g ,' ,
vt c ~. 6 .
.c >
6.* py;;g 9 *;' - s . l v.4 E6.9 om Je a
.2
J:...fi t
'y, .. . dt. ../
.6 s . . . :.s x 100 x 100 x 100 Micrograph 3, 7000C 4, 900 C 5, 1100 C
4 Sample 7,8,9: In order to examine the formed ternery intermetallic crystals coalecens (growth to larger and fewer crystals)C at increasing temperagure, samples is heated to 700 C, 900 C and 1100 C in helium. The micrographs 3 and 4 shows a slow growht of the terngry crystals in the temperature range 640 to approx. 900 C. Mi 5 9 raph 5 shows that there is a rapid ggowht aftgr 900 C. Soraewhere in the temp. range 900 C to 1100 C formes an eutectic phase consisting of the ternery intermetallic compound and micro segregated aluminium. This indicates that there is dissolved as mutch of the ternery compound in the melted aluminium that an eutec-tic composition is obtained. During cooling the eutec-tic phase is formed as a laminated phase. CONCLUSION
, Considering the metallographic examination and the results from the DTA runs, the following conclusion can be drawn.
UnSip powder The U 3Si2 Powder is thermodynamic stable to 1400 C in
, helium. When heating in air there gs a heavy oxidation in the temperature range 150-950 C. The oxidation products is probably silicon and uranium oxidation.
UnSio/Al (38 w/o) miniplates b When heating in helium there is no exo/ endothermic reactions below aluminiums melting point which could indicate any changes in the core and the cladding mate-rial. In or immediately above aluminiums melting point there is a reaction between U3Si2 and A1. The ternery intermetallic compound which is formed has according to EDAX analysis the following composition: U: 72.1 w/o 22.9 a/o) Si: 6.6 w/o (17.6 a/o), Als 21.3 w/o (59.5 - a/o). All the U Si in the core is transformed to the ternery (approx. intergetallic compound at this temperature 640 C). The stoecheometric composition once formed is fixed at increasing temperature. The ternegy intermetallic gphase is in the temperature range 640 C to approx. 900 C homogenious distributed in the form of crystals embedded in an aluminium matrix.
~,. s'
_9_ l This concludes that the miniplate will not remain stable after aluminiums melting point. After 640 C there is a coalecens (growht to larger and fewer crystals) of the ternery crystals at increasing temperatures. During heating and during the growth process the crystals have a tendency to move towards the surface of the sample. This will result in an in-homogenious distribution of ternery crystals in the aluminium matrix. At temperatures above 900 C an eutectic phase is formed in the aluminium matrix. In order exactly to measure the -ternery intermetallic composition,and when and how the eutectic phase is formed, further examinations is required. When heating in air there is a non reversible exothermic reaction at approx. 440 0C. This reaction is due to oxidation of the UiSi2 core surface. ACKNOWLEDGMENTS The authors are indebted to RIS$ Metallurgical Gepart-ment for the acces to and the use of the departments equipment. Special thanks to Ole Olsen for preparing and photographing the metallographic specimens and to Kjeld Larsen for assistance with the differential ther-mal analysis. The assistance of Jan Borring in prepa-
, ring the paper is gratefully acknowledge.
l l
.]
-- ~
, As .
NEUTRONIC AND THERMAL-HYDRAULIC CALCULATIONS
, FOR THE CONVERSION OF THE 5 MW "LA REINA" REACTOR USING MEU FUEL J. Klein, O. Mutis, J. Medel,
{ 0. Villegas and J. Zambrano Chilean Nuclear Energy Commission Santiago, Chile ABSTRACT This paper presents the neutronic and thermal-hydraulic calculations for the conversion of the 5 MW La Reina reactor i from 80% to 45% enriched uranium fuel elements. Mixed-enrich-i ment cores are also studied where MEU fresh fuel elements subs l titute gradually the HEU depleted fuel elemen:s in the equili-l- . brium core. Thermal-hydraulic calculations have been carried
. out to determin changes in the characteristics of the con-
, verted reactor during steady conditions and transient response to a coolant flow loss. -
1
- 1. INTRODUCTION The Chilean Nuclear Energy Commission has been performing studies for the conversion of the La Reina MTR-type reactor. It was establishedl that, in order to match the cycle length of the current 80% enriched uranium (HEU) fuel l design with 20% enriched fuel, an uranium density of about 3.7 g/cc is required.
't The La Reina reactor will be converted to use 45% enriched uranium (MEU) fuel elements. The MEU fuel elements have only minor changes in the de-
, i sign of the fuel plates and no changes in the design of the fuel element geome l .' try. With the same element geometry, the thermal-hydraulic characteristics of the core will be virtually identical with both HEU and MEU fuels. l i The neutronics calculations have been performed using well verified j computer codes available at the NEA Data Bank. A thermal-hydraulic subroutine l package has been developed to analize the behaviour of the reactor in steady conditions at nominal power and during primary pumps failure. i
- 2. REACTOR AND FUEL ELEMENTS DESIGN
! The La Reina reactor is a light water-moderated, water-cooled and be ryllium-reflected reactor. It emoloys a flat plate MIR-cype fuel, Fig. 1, wiEh l 1 highly-enriched aluminium-uranium alloy. L' The configuration of the core can be changed to suit experimental re
. quirements and all core elements are similar in overall shape. Six blade-type
- control absorbers pass through the core in two groups of three. Slots are for med between the elements and in the lattice place to allow passage of the ab-sorbers.
I .
. i The lattice place is a 12 cm thick plate of aluminium and a rectan-gular array of 10 x 3 holes are suitable to accom=odate the core elements.
Tables 1 and 2 su=marize the reactor and fuel element design descril tions with the HEU fuel and MEU fuel.
- 3. NEUTRONICS CALCULATIONS 3.1 Cross Section Generation The cross sections were prepared for different regions in the core using the WIMS-D cell codeZ . This code will accept rod or plate fuel geome-
- tries in either regular arrays or in clusters and the basic library has been compiled with 14 fast groups, 13 resonance groups and 42 thermal groups. Tem-perature dependent thermal scattering matrices for a variety of scattering laws are included in the library for the principal moderators. The creat=ent of resonances is based on the use of equivalence theorems with a library of accurately evaluated resonance integrals for equivalent homogeneous systems at a variety of te=peratures.
Different cell models were needed to generate appropriate cross sec-tions for the various reactor regions3 '" in the standard five-group structures connonly used at Argonne National Laboratory for MTR plate-type reactor. The broad group energy boundaries are listed in Table 3. Once generated, these cross section. sets were used in diffusion or transport calculations. Most of the results of this study are based on KY multigroup diffu-7 sion calculations using ERE3US* and CITATION codes with bucklings i= posed for the axial di=ension. Two-dimensional transport theory calculations were performed running Tk'OTRAN a code. 3.2 MEU versus REU Comoarison 3.2.1 Critical Configuration The results of the critical experiment performed during October 1974 will be used to confirm the calculation methods com,aring the calculated va-lue of the effective multiplication factor with the one experimentally obser-ved. During the critical experiment, with all control blades withdrawn, it was found that the system was slightly suberitical with the first 19 fuel ele ments (Fig. 2 without the fuel element in position E9) and supercritical when element 20 was added. For the 19 and 20 fuel element configurations kegg was calculated using an KY model and 6.65 cm extrapolation distance *. The diffu-
~
sien theory calculations gave an effective multiplication factors of 0.99673 for the 19 fuel element case and 1.00799 for the 20 element case. Thus, the calculated results allow predicting kegg quite well for the cold and clean critical core using HEU fuel. Critical calculations were performed for the MEU fuel using the sa=e extrapolation distance as the EEU fuel case. For the 19 fuel configuration the system was supercritical and gave an effective multiplication factor of 1.00526. However, 13 fuel elements in the core were not critical but slightly
suberitical. 3.2.2 Reference Core Fig. 3 shows the 32 fuel element reference core, consisting of two regions of 16 (2 by 8 array) fuel elem2nts separated by a file of solid alu-ninium elements. The core is beryllium-reflected in the east and west sides. The experimental value of k,g, estimated for the reference EEU core using fresh fuel was of 1.056 e 0.004 . The ERE3US, CITATION and TWOTRAN codes were used to calculate the keff for both the EEU and MEU reference core with 32 fresh fuels using the same axial buckliog", 3:2 . t,936 x 10-3 cm-2 The k gg results are shown in Table 4 and for the EEU case the codes predict well the experimentally deter-nined kegg value. The flux distributiens in X and Y directions for groups 1, 3 and 5 at the nominal power of 5 MW are shown in Figs. 5, 6, 7 and 8. 3.2.3 Equilibrium core In the equilibrium cycle studies using ERESUS code, the fuel cycle
! length for the core configuration, as shown in Fig. 4, was computed for beth EEU and MEU fuels using the same fuel shuffling pattern. This fuel shuffling pattern is shown in Fig. 9, and represents an X-Y model of the core with 1/2 core symmetry. This fuel shuffling pattern is intended to be illustrative for current calculation purposes, and does not necessarily represent the fuel shuffling pattern that could be used in the operation of the reactor.
In this shuffling pattern one fresh fuel element is inserted into position 1 outside the core, and the remaining fuel elements are rotated se-i quantia117 after each cycle. With 1/2 core symmetry, this pattern is equiva-lent to loading into the core two fresh fuel elements per cycle. The elements are discharged from position 18 after 18 operational cycles. Table 5 shows the results of the burn up studies. In each case, the fuel cycle length that would provide an end of equilibrium cycle kegg of 1.005 was computed. Fig. 10 shows a comparison between the thermal fluxes at the end of equilibrium cycle. Fast, epithermal and thermal flux ratios between the MEU and HEU cores along the east-west direction, also through the center of the core, are shown in Fig.11. 3.2.4 Safety Related Parameters In this section, results of calculations are presented on the safety related neutronic parameters needed for transient analysis of the 5 MW La R91 na reactor with HEU and MEU fuels. 3.2.4.1 Kinetic parameters.- The reactor kinetics parameters have 'seen cal-culated for the reference core and at the end of equilibrium cycle 3. The number of energy groups was increased from five groups to ten groups. see Table 3. The WIMS code was run using the ten-groups structure to generate
.= 1
. i l
cross section data for each region of the core. I i The neutron generation time ( A), the prompt neutron lifetime (1) and I
, the effective delayed neutron fractions (Seff), using the two dimensional di-ffusion theory perturbation capability of the CITATION code, are given in Tables 6 and 7. Table 8 gives the delayed neutron fractions in six energy groups for both reference and equilibrium cores.
3.2.4.2 Isothermm1 Tameerature and void Coefficients of Remetivitv.- The iso
- . thermal temperature and void coefficients of reactivity were computed separately as functions of temperature or void fraction for the EEU and MEU cores. To determine chose coefficients, the reactivity changes were cair.ula-ted as functions of temperature due to the following effects: (1) the harde-ning of the neutron spectrum caused by increasing the temperature of the water only, (2) the iacrease in the neutron leakage when the water density is decrea sed, and (3) the increase in absorption in the U-238 epithermal resonances due to the increase of the fuel seat temperature (Doppler effect).
f 1 The calculations were made for the reference fresh fuel cores and also for the 36 fuel element configurations at 30C both EEU and MEU fuels. These results are shown in Tables 9 and 10. The initial slopes of each of the tempe '
] rature coefficients are shown in Tables 11 and 12.
3.2.4.3 Eautithrium Xe-135 Worths.- The effect of Xe-poisoning is calculated I by setting the Xe-135 concentration zero for each considered burn-up and recalculating the cross sections with the WIMS code. The reactivity worths of equilibrium concentrations of Xe-135 were j evaluated at the beginning of equilibrium cycle for EEU and MEU fuels and the t results are given in Table 13, 3.2.5 Control Rod Worths
- g The control worth calculations of the six fully inserted Cadmium-1 stainless steel control blades for the reference and equilibrium at 20C cores, l both HEU and MEU fuels, are based on a diffusion calculation using blackness theory. ,
i
! The six absorber blades are manually controlled in pairs and the lo-cation of the control rods in the core configuration are shown in Fig. 4. Each of the six control blades is 0.757 en thick and 63.5 cm long. The central :
,j blade pair is 19.37 en wide and the two outer pairs are 26.67 cm wide. The Cadmium section has a width of 0.254 cm.
1
~! The cross sections were generated using the WIMS code with the slab geometry option in six broad groups. The cross sections edit over the portion
. of Cadmium was utilized to produce the blackness coefficients 2 and 310 The
~ . ! group averaged values of 2 and 3 are used to obtain equivalent diffusion -
, theory parameters (D and I,) for the Cadmium slab.
A j-To calculate the control worth of the Cadmium-stainless steel control
- . ; blades, using the CITATION code, the reactivity of the reactor for the blade i in (k i) and blade out (ko) configurations are to be calculated.
}'
3 s t 9 l
_ The_ blackness coef ficients 2 and 3, and the diffusion theory para-meters D and Ia are shown in rable la and 15, and the reactivity corresponding to the control blade out and blade in3 are shown in Tables 16 and 17 for both i reference and equilibrium cores using HEU and MEU fuels. 3.3 Mixed Cores Mixed-enrich =ent cores might be considered as an option for the con-version of the La Reina reactor where MEU fresh fuel elements substitute gra-dually the HEU fuel in the equilibrium core. This situation is expected to occur during the core conversion and in the planning of such conversion it has to be able to predict the accurate behaviour of the mixed cores. j For the mixed core calculations it will be used the same shuffling paccern that was utilized for the equilibrium core calculations, Fig. 9. At {
, EOC the fuel element in position 18 is removed from the core and the remaining
, EEU fuel elements are rotated sequentially and then, a fresh MEU fuel is in-l serted into position 1. Due to the 1/2 core sy= metry, this pattern is equiva-lent to loading two fresh MEU fuels per cycle and, consequently, the core is being gradually converted from the use of EEU to MEU fuel.
. Calculations were performed at 30C for the first mixed-enrichment
. core, with Xe-135, using the ERE3US code. It was found a k egg of 1.02271. In section 3.2.3 it was obtained a k,gg of 1.02228 for the REU equi-librium core and 1.02318 for the MEU equilibrium core, both at 30C with Xe-135. Thus, it has to be concluded that the kegg will gradually change from that co-rresponding to the first mixed core to that completely converted MEU core; i.e. from a kegg of 1.02271 to 1.02318. I
! It is also important to evaluate for mixed-enrichment core the chan-ges in the power distribution in the particular case where lower-enriched fuel elements with higher U-235 concent are placed in a higher-enriched core. Thus ,
it was calculated the form-factors, defined as the ratio of the average power of a fuel element and the average power of the core. Fig. 12 shows the form-
. factors for each of the fuel elements of the REU equilibrium core and the first I
mixed-enrichment core both at 30C. It can be observed that in position 1, where there is a fresh MEU fuel element, the form-factor is increased in 3.4% for the case of mixed core with respect to the HEU core. This increase has not a signi ficant effect over the thermal-hydraulic safety margins. Figure 13 shows the form-factors of each fuel element for the EEU and MEU equilibrium cores at BOC. I 4 THERMAL-HYDRAULIC CALCULATIONS The conversion of the La Reina reactor from EEU fuel to MEU fuel was considered without changes in the fuel element geometry. Only minor modifica-
, tions in fuel plates were made where the width of the meat was changed from its actual value of 6.0147 cm to 6.0833 cm in the MEU case. Thus, the thermal-hydraulic behaviour of the converted core would be virtually identical to the HEU core.
=, - - -- . .
l . . The thermal-hydraulic calculations have been carried out using a thermal-hydraulic subroutine package ll (THSP) in order to determine changes in the characteristics of the converted reactor during steady conditions and i l
- t transient response to a loss of coolant flow for both EEU and MEU reference core.
In these studies the reference core was considered for the thermal-
- . hydraulic calculations in order that, for this case, the heat flux reached ;
the maximum value of 29.73 v/cm2 La the fuel elements located in positions 55,
~
~. 56, F5 and F6, Fig. 3. For the EEU equilibrium core the maximum heat flux is 20.76 w/cm2 in positions D2 and D9. Thus, if it is found that the reactor is
! safety operating the reference core at 5 MW, the equilibrium core configura-tion will be safer. The calculations for the reference core, both EEU and MEU
. fuels, confirm that there are minor changes in the thermal-hydraulic behaviour, Table 1, and for this reason the results for the fresh fuel MEU reference core only will be shown. The technological hot channel factors were also utilized-1
.A based on a statistic treatment 21,
, i 5 Figure 14 shows the form-factors for the EEU and MIU reference cores at the beginning of life.
_: 4.1 Reactor in Steadv State Conditions
~
j Figures 15 and 16 show, along the hottest channel, the coolant tem-
; perature and the clad surface temperature, respectively, at the nominal power j of 5 MW. >
' The onset of nucleate boiling (ONB) conditions was evaluated using the Bergies and Rohsenow 12 correlation. This correlacion is widely used and applicable down to the low pressures characteristic of research reactors.
The security margin of the reactor will be defined as the difference between the ONB and maximum cladding temperatures in steady conditions. In i che case of La Reina reference core the Bergies and Rohsenow correlacion pre-diets that ONE will occur when the channel temperature may become higher than 126.6 'C. However, in steady conditions the maximum clad surface temperature
,d is 71.5 *C, giving for the reactor a security margin of 54.9 *C.
': The operation power of the La Reina reactor under natural convection cooling using the Vernier 13 and Westwaterl4 correlations and considering the same security margin for operation at 5 MW (54.9 'C) is 200 KW with a coolant velocity of 5.18 cm/s and an exic coolant temperature of 54 *C.
The heat output from water-cooled reacter fuel channels is limited by, among other factors, the phenomenon of departure from nucleate boiling ' (DNB) commonly referred to as burn-out. The DNB conditions were determined using the Mirshak15, Bernach st and Labuntsow 17 correlations applicable to low
~, pressure plate-type fuel reactors. In the case of natural convection the Rohsenow and Griffith18 correlation will be used. However, the maximum power output from reactor fuel channels, cooled by sub-cooled water, may be limited by the occurence of excursive flow instability. Prediction of the conditions under which excursive instability will occur depends on the knowledge of the
.1 i
_ ~ - - _ , . - . . . . . , . _ _ _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ . , _ _ _ . _ _ , _ _ . _ _ _ _ , _ _ - . _ _ .
pressure drop-flow characteristics of the coolant channels. The channel pre-ssure drop-flow curve depends on the channel geometry, inlet and exit resis-tances, flow direction, subcooled vapor void fraction and heat flux distribu-tion along the channel. The onset of flow instability (0FI) condition will be evaluated by the Forgan and Whittle ll relation and the Waters 2a correlation valid for a nonuniform heating distribution. The margin to DNB and 0FI, for fresh fuel reference core, are given in Table 1. 4.2 Thermal-Hvdraulie Transients 4.2.1 Loss of Coolant Flow The transients due to the less of coolant flow, when one or more
, pumps fails, were calculated starting from full power operation at 5 MW with a coolant flow rate of 654 m /hr 3 and a coolant inlet temperature of 35'C. As the flow decreases, two flap valves are opened below the core so that the re-maining reactor power can be really dissipated to the bulk water in the reac-ter by natural convection circulation. The direction of flow through the core must reverse so that natural convection may occur. Reactor scram is program-med to starr when the flow is decreased below 490 J/hr and the time to reach this minimum flow is shown in Table 18.
Figures 17,18 and 19 show the evolution of the flow calculated by THS? code when one, two, or three pu=ps have failed, respectively. Figure 19 shows clearly when the flap valves are opened and the three hydrodinamics regimes are also shown: forced, mixed, and natural convection. Figure 20 shows the cladding temperature when three pumps have si-multaneously failed. The minimum clad surface ta=perature of 41.l'C occurs at the end of the scram; the maxi =um clad surface te=perature of 89.0*C is reached at the end of the mixed regime due to the disadvantage of the heat transfer. The mixed convection starts with the opening of the flaps at 3 9.13 s and for a flew of 110 m /hr. Theflagsneed0.5sectobecompletely open and the flov for each flap toes to 40 m /hr. Table 19 shows the secuen ce of the transient response due to the simultaneous failure of three pumps. The THPS code predicts a reactor power of 200 kw at the beginning of the opening of the flaps. However, to reach the onset of flow instability with a flow of 110 m 3/hr it is necessary a reactor power of 2.2 MW. 4.2.2 Electrical Energv Sueolv Cucoff Figure 21 shows the evolution of the flow when the electrical ener gy supply is interrupted. In this case, the mixed convection ti=e is longer than in the latest case of simultaneous failure of three pumps due to a de-crease in the ascension force. Notice that the inversion of the flev starts at 14.3 s. ,
The evolution of the cladding temperature is shown in Fig. 22. The maximum and minimum cladding temperatures are 33.0*C and 37.0*C. respecti-vely. The temperatures are loss than in the case of failure of three pu=ps because there is no delay in the scram. The case of an interruption of electrical energy supply is studied simultaneously with one of the two flap valves shut. In this case, the na-cural convection starts at 15.4 see and the flow reaches 55 m /hr. The maxi 3
~
zum clad surface temperature is 37.3*C, This accidental event causes smal1 changes in the thermal-hydraulic behaviour of the system and the TESP code predicts that 44% of the total flow passes through the flap valves when both are open and 30% when only one is open. 4.2.3 Pump Blockage The evolution of the flow and the cladding temperature for one bio ekaded pump are shown in Figs. 23 and 24 The flow rapidly falls down up to 420 m 3 /hr and the perturbation of the flow reaches the core 0.5 s later than the blockage of the pump. The blockage of a pump is one of the most severe accidents because the nominal power of the reactor (5 MW) continues during the hydrodinamic '
! transient of 0.9 s with risk of reaching the onset of flow instability con-dicions.
The maximum cladding temperature of 102*C is reached in 0.6 s mean while the reactor continues at 5 MW. The scram of the reactor starts at 0.8 s and the clad temperature rapidly descends to the asymptotic value of 35'C. In this transient, OFI conditiens would be produced if the reactor operation power were 7.1 MW. For a reactor power of 5 MW che onset of flow instability three conditions pumps. would However, be reached this accident haswith the simultaneous an occurrence blockage probability of ,,/ of 2x10 year. CONCLUSIONS In every aspect can the 5 MW La Reina reactor be converted to use 45% enriched uranium without modifications in the design of the fuel element geometry where the uranium density changed from its actual value of 0.630 g/cc to 1.258 g/ce. The most important neutronics effects in the equilibrium core perfor mance as a result of the conversion, from 80% to 45% enriched uranium fuel elements, is the reduction of 4% in the thermal flux in the irradiation posi- , cious, 6% in the control blade worth and 7.7% in the neutron generation time. ( An increase of about 10% and 5.3% was obtained for the cycle length and the
, discharge burn-up, respectively, i
l l l i l
[~ ~ . Mixed-enrichment cores might be considered as an option for the con-version of the reactor where MEU fuel elements substitute gradually the EEU i fuels in the equilibrium core. The thermal-hydraulic behaviour for the reference core was identical
-to both EEU and MEU fuels. The effects of flow transients over the thermal-hydraulic characteristics of the reactor for the MEU reference core at begin-ning-of-life have demonstrated that the reactor can be opera'ted at 5 MW.
l- .
- For the equilibrium core the flow through the fuel element channels
( , has a reduction estimated in 10% in relation with the reference core; con-sequently, the maximal head flux is reduced in about 30% comparing the same configurations. It could be, thus, deduced that the reactor using the equi-librium core is safer than the reference cores configuration. However, this conclusion must be confirmed in future calculations.
} .
t REFERENCES I'
- 1. U.Schutt, et al., "IAEA Technical Mission to Chile," Oct. 29 to Nov. 2 1979.
- 2. J.R. Askew, T.J. Fayers and P.B. Kenshell, "A General Description of the Lattice Code WIMS," Journal of the British Nuclear Energy Society, Octo-ber 1966.
- 3. J. Klein and O. Mutis, " Reactor Physics Studies for the Conversion of the Lo Aguirre Reactor from HEU Fuel to LEU Fuel," Work performed at the ANL, Argonne, Illinois, May 1981.
! 4 O. Mutis, "Escudios de Conversi6n del Reactor RECH-1. C11culos Esti-
, cicos," Comisi5n Chilena de Energia Nuclear, Diciembre 1983. ,
$. Guidebook on Research Reactor Core Convection from the Use of Highly Enriched Uranium to the Use of Low Enriched Uranium Fuels, Internacio-nal Atomic Energy Agency, Report IAEA-TECDOC-233, August 1980.
- 6. M. Console A. Danari and E. Salina. "EREBUS, A Multigroup Dif fusion Depletion Program in Two Dimensions for the IBM-360," FN-E-38,1967. *
- 7. T.B. Fowler, D.R. Vondy and G.W. Cunningham, " Nuclear Reactor Core Analysis Code: CITATION, " ORNL-TM-2496, Rev. 2, July 1971.
- 8. K.D. Lathrop "TWOTRAN - A Fortran Program for Two Dimensional Trans-port," GA-8747, 1968.
- 9. J. Medel, "Determinacion de la Importancia de las Barras de Control y Cniculos de los Parn=etros Cinnticos para el Reactor RECH-1," Comisi6n Chilena de Energia Nuclear. (to be published).
- 10. M. Bratscher, " Control Worth Calculations for Slab Geometry with Appli cation to the Rhode Island Nuclear Science Center Reactor," Argonne National Laboratory.
- 11. O. Villegas, "THSP: Thermal-Hydraulic Subroutines Package," Comisi6n Chilena de Energia Nuclear, (to be published) .
- 12. A.E. Bergies and W.M. Rohsenow, "The Deter =ination of Forced-Convection Surface-Boiling Heat Transfer," Transaction of the ASME 86 (Series C -
Journal of Heat Transfer), August 1964
- 13. Ph. Vernier, " Convection Natura11e dans un Canal vertical de Section Rectangulaire Chauffi Uniformiment," CEA-CENC-Rapport CEA N' 2197, 1962.
14 Y.Y. Hsu and J.W. Westwater, " Approximate Theory for Film Boiling on Ver tical Surf ace," Chem. Eng. Prog. Symp. , 56, 30, 1962.
- 15. S. Mirshak, W.D. Duran and R.H. Towell, Heat Flux at Burnout ," Du Pond,.
DP-355, February 1959.
- 16. L. A. Bernach, "A Theory of Local Boiling Burnout," Heat Trans. Symp. ,
A.I.Ch.E. National Meeting. Louisville, Kentucky, 1955.
- 17. D.A. Labuntsow, " Critical Thermal Loads in Forced Motion of Water which is Heated to a Temperature Bellow the Saturation Temperature," Soviet Journal of Atomic Energy, (English Translation) 10, November 1961.
- 18. W. Rohsenov and P. Griffith, " Correlations of Maximum Heat Transfer Data for Boiling of Saturated Liquids," Chem. Eng. Prog. Symp. 52, 1956.
- 19. R. Forgan and R.H. Whittle, " Pressure-Drop Characteristics for the Flow of Subcooled Water at Atmospheric Pressure in Narrow Heated Channels,"
Part 2. AERE-M 1739, 1966.
- 20. E.D. Waters, " Heat Transfer Experiments for Advanced Test Reactor,"
BWNL-216, 1966.
- 21. S. Fabrega, "Le Calcul Thermique des Reacteurs de Recherche Refroidis par Eau," CEA-R-4114, 1971.
Table 1. Reactor Design D'escription Parameter HEU MEU Reactor Type Pool Type MTR
- Steady-State Power Level, MW $
Nanber of Reference Core Configuration 32 Control Blade Material Cadmium Number of Control Blades 6 Grid Place 10x8 Lattice Picch, mm - 76.149x76.149 Primary Coolant System Flow, m /hr 654
- Core Pressure Drop, kg/cm 0.15 Primary Coolant System Pressure, kg/cm 1.92 Core Entrance Water Temperature, 'C 35.0
. Core Exit Water Temperature, 'C 41.4 Average Heat Flux, W/cm 13.97 13.81 Peak Heat Flux, W/cm 29.73 29.84 Maximum Cladding Temperature, 'C 70.6 71.5 Maximum Meat Temperature, *C 71.9 72.8 Radial Factor 1.60 1.60 Axial Factor 1.33 1.35 ONB Ratio 1.85 1.85 DNB Ratio 4.25 4.25 0FT Ratio 2.76 2.77 l Coolant, Moderator Water Reflector Beryllium h
E. _
Table 2. Fuel Assembly Design Description Parameter HEU MEU Uranium Enrichment, 80 45
~
Type of Place MTR, Straight Fuel Element Dimensions, m 74. 740x74. 549x581.05 Place Thickness, m 1.53 Water Channel Thickness, m 3.177 Places / Fuel Element 16(14 internal + 2 external) Fuel Meat Composition U-Al Alloy Meat Dimensions, m 0.61x60.147x581.05 0.61x60.833x581.05 . Clad Thickness (A1), m 0.46 U-235 Density in Fuel Meat: External Place, g/'em 0.258 0.283 Internal Place, g/cm 0.516 0.566 U-235/ Place: External Place, 3 5.50 $ 6.11 Internal Place, g 11.00 12.21 f Uranium Density in Fuel Meat:
+ External Place, g/cm 0.322 0.630 Internal Place, g/cm 0.645 1.258 Uranium / Place:
External Place, 3 6.785 13.578 Internal Place, g 13.750 27.133 l' U-235/ Fuel Element, g 165.00 183.16
+
k
y-Table 3. 3 road Group Energy Structure Used in the Diffusion or Transport Calculations Upper Energy Upper Energy Upper Energy Group (5-Group Set) *(6-Group Set) (10-Group Set) 1 10 MeV 10 MeV 10 MeV 2 821 kev 821 kev 821 kev 3 5.53 kev 5.53 kev 9.118 kev 4 2.10 eV 2.10 eV 5.53 kev
! 5 0.625 eV 0.625 eV 2.10 eV
. 6 0.140 eV 1.15 eV I
7 0.625 eV e 8 0.4 eV 9 0.14 eV 10 0.058 eV Table 4 Effective Multiplication Factor Calculated
. for Reference Configuration with HEU and
.MEU Fuels EREBUS CITATION T'40TRAN T 7p*
k %4k/k k 2ak/k k %Ak/k HEU 1.05666 5.36 1.05652 5.35 1.05805 5.49 MIU 1.06496 6.10 1.06305 5.93 1.06417 6.03 i n n
i Table 5. Equilibrium Cycle Performance Characteristics Parameter HEU MEU Cycle Length", Days 36.8 40.6 BOC k,gg 1.02228 1.02318 BOC k,gg(without Xe) 1.05284 1.05517 EOC k,gg 1.00486 1.00505 U-235/ Element ,g 165.00 183.16 U-235 Burned", 2 54.6 57.5
" Based on a power level of 5 MW. .
b U-235 content is for fresh fuel element. ,
"U' 235 burned in discharged fuel element.
t Table 6. Kinetics Parameters for the HEU and MEU Reference Cores Using Fresh Fuel Fuel k,gg A, us 1, us S,ff Type HEU 1.05652 70.81 74.81 7.432-03 MEU 1.06305 66.00 70.16 7.423-03 Table 7. Kinetics Parameters for the Equilibrium HEU and MEU Cores at BOC Fuel k,gg A, us 1, us 8,gg 67P* HEU 1.02228 71.07 72.65 7.315-03 MEU 1.02318 65.62 67.14 7.279-03' u
Table 8. Delayed Neutron Fractions for the Reference Fresh F 1 Cores and at the Beginning of Equilibr.um Cycle Reference Core Beginning of Eq. Cycle Group HEU MEU HEU MEU 1 2.80356-04 2.79541-04 2.75944-04 2.74181-04 , 2 1.58293-03 1.57949-03 1.55799-03 1.54911-03 4 3 1.39767-03 1.39540-03 1.37564-03 1.36850-03 4 3.02618-03 3.02196-03 2.97849-03 2.96365-03 1 5 9.50362-04 9.50949-04 9.35355-04 9.32442-04 6 1.94721-04 1.95379-04 1.91638-04 1.91532-04
'l
\
- i i
1
.t 4
i t 1 3 l'
. ' l.
k L
3 -t Table 9. Temperature Coefficients of Reactivity
- i' Due to Change of Fuel Temperature, Wa-ter Temperature, and Water. Density for-the Reference HEU and MEU Fresh Fuel
., Cores
'1
- Chante of Fuel Temoersture T , 'C HEU, -4c+10 3 MEU, -40 10 3 7
~
20 - - 100 0.08 0.81
. 200 0.17 1.71 400 0.32 3.20 650 0.45 4.64
( Chante of Water Temperature
} T,, 'C HEU, -40 10 3 MEU, -4c+10 8 20 - -
- - 45 2.26 2.02 1 70 4.48 3.99 120 9.07 8.01 Chante of Water Density w' H2 0, g/en' HEU, -60 10 8 MEU, -40 10 8
, 20 0.9982 - -
.j 50 0.9881 5.70 3.22 0.9749 8.14 5.74 o 75 l 100 0.95 13.99 11.39 i 100 0.90 20.59 19,58
, 100 0.80 36.50 34.78 n:
L I i f 4
- ..i i
i w"'
* ~
t . Table 10. Temperature Coefficients of Reactivity Due to Change of Fuel Temperature, Wa-ter Temperature, and Water Density for l the Equilibrium HEU and MIU Cores at j 30C Change of Fuel Temperature
; T. C HEU, -a0+10 3 MEU, -60+10 3 F
1 20 - - t 100 0.09 0.94 ' l Il 200 0.19 1.99
- l. ] 400 0.35 3.71
, 650 0.49 5.38 t
; Change of Water Temperature I
i
] T,, *C HEU, -40 10 3 MEU, -40 10 3 20 - - i 45 4.47 4.12 70 8.72 8.01 1
120 16.93 15.46 I Chante of Water Density , , f. f' j T,, *C cH2 0, g/cm 3 HEU, -60 10 3 MEU, -ac 10 8 l l > 20 0.9982 - - l 50 0.9881 4.31 4.69 I 4 75 0.9749 9.80 9.01 l 100 0.95 16.84 15.02 100 0.90 25.77 24.02 i 100 0.80 46.63 49.15 t 6 2 i :! i ! l, I
.l
[
- 1 I
1 l I l l l I L
n . L I l Table 11. Initial Slopes of the Temperature Coefficients of
; Reactivity for the Reference HEU and M.EU Fresh
- Fuel Cores r-
[ HEU MEU Effect ( -ao/AT 10"*/'C -do/aT*10
~
'/'C' L .; Doppler (20*C - 100 *C) 0.0010 0.0101
+ ; Water Temperature (20'C - 70*C) 0.0896 0.0798
'>'! ~ Water Density (20'C - 75'C) 0.1480- 0.1044 i 4 0.2386 0.1943
- l'
(:.' d f E 7; Table 12. Initial Slopes of the Temperature Coefficients of
'i Reactivity for the Equilibrium HEU and MEU Cores -
at BOC l j *
' .i
- 1. HEU MEU l ,, Effect l '-
-ao/AT+10~'/*C -ac/aT+10~3/*C
.i Doppler (20*C - 100*C) 0.0011 0.0118
! Water Temperature (20*C - 70*C) 0.1744 0.1602 l Water Density (20'C - 75'C) 0.1782 0.1638 l-
- 0.3537 0.3358 h ;
i -1 ! -j L
, j Table 13. Xe-135 Reactivity Worths at the BOC i
r ! '=' u.ff u.ff l Type with Xe without Xe ak, % do,
! HEU 1.02228
- 1.05284 3.06 2.84 .
t MEU 1.02318 1.05517 3.20 2.96 e I' f p.- l t s t--w o
i , . -i
+
,_ . -- .. - - . - . . . - . . . - - . - - - . . . ~ - . _ - ...f.
- ?
J
~
Table 14. Blackness Coef ficients aind Ef fective Dif fusion Parameters for tise Cadmita Control Blade for the llEU and HEU Reference Cores (.
- HEU HEU.
Group _ a E,, cm ,
, 8 D, cm E,, cm g a S D , cm a
1 - - 1.88353 3.83186-03 - - 1.87860 3.83322-03 a 2 1.26746 4.17298-03 1.26265 4.17298-03 a 3 - - 9.65129-01 7.58797-03 - - 9.68765-01 7.55387-03 l 4 7.48236-02 3.87175 4.88274-01 5.92860-01 7.47708-02 3.87471 4.88652-01 5.92424-01 5 4.98319-01 5.01687-01 1.53556-02 1.11021+01 4.98246-01 5.01761-01 1.55100-02 1.10446401 6 5.00000-01 5.00000-01 7.00330-03 1.49920601 5.00000-01 5.00000-01 7.00330-03. 1.49920101 6 = D, E =E. a a
+,
Y ^ r ( _1<- l
~
~
s . e '* -e * - - -. c. , i i . f i i L Table 15. Blackness Coef ficients and Ef fective Dif fusion Parameters for the Cadmium Control Blade for the Equilibrium HEU and HEU Cores at BOC HEU HEU I Croup G B 5 cm f,. co" a S' 6, em f , cm-I A I - - 1.88129 3.82503-03 - - 1.87796 3.82670-03 A 2 - 1.30 % 9 4.17298-03 1.30255 4.17298-03 e 1 3 - - 9.56948-01 7.62226-03 - - 9.61195-01 7.58973-03 l 4 7.49020-02 3.86736 4.87742-01 5.93478-01 7.46403-02 3.87081 4.88153-01 5.92981-01 l 5 4.98501-01 5.01504-01 1.49510-02 1.1255t+01 4.98420-01 5.01586-01 1.51358-02 1.11848101 { 6 5.00000-01 5.00000-01 7.00329-03 1.49920+01 5.00000-01 5.00000-01 7.00329-03 1.49920601 J 4 ! - 6 = D, f a =E. a
i I Table 16. Control Worth of the Cadmium Control Blades for the HEU and MEU Reference Cores Fuel k k Control Type "E " Worth HEU 1.05652 0.89740 lo.78
; MEU 1.06305 0.90974 15.85 i
l. l l' ~ Table 17. Control Worth of the Cadmium
'[ Control Blades for the Equi-librium HEU and MEU Cores at BOC Fuel k k Control
"" i" 4 Type Worth HEU 1.02228 0.92213 10.62 s
MEU 1.02318 0.92832 9.99 4 i i l T
l Table 18. Time to Reach the Flow Trip Point i Number of Time, s Delay, ms Failed Pumps 1 2.00 70.0 2 0.74 70.0 3 0.42 70.0 t j Table 19. LOTA Secuence Action Signal Trip Point, s Delay, ms Failure o'f Pumps (3) Flow 0.0 0 Scram Flow 0.42 70 Beginning of Control Blade Insertion Flow 0.49 0 End of Control Blade , Insertion Flow 0.391 0 Beginning of Opening of
; the Flap Valves Pressure 9.13 0 End of Ocening of the Flap Valves Pressure 9.43 0 i End of Mixed Flow Flow L3.1 0 Beginning of the Inversion -
of the Flow Flow 14.0 400 All Control Blades: 72% Out. F
----4 --
l o. l l 76.149 i 72.135 __ ! I I 7.. _ i t.r_e_
,J,
'I
_y,_,1 I ! 1 , I
;i
- i i ! I l
I I I '. l i I ~
,1
_ , 3.17 7
- i. i ,
l I l 1 i I t 4 l, I i l r- I I' ! ll J e' , l I
, ! I e
- I ,
; j v o l'l i l t I
+
W
- s@ ii 3 i l
- l ,
1 { gil i l i r- i i J _I1.53 1 ; l l ; i! l , i ' l l l I i 1 i l i l l li l l I l l l
--[.. . . . .d .dd . . .d ;
f 74.549 i HEU Fuel Plata t
, 10.46 1.53 lI -
l0.65
T
, I, - 60.147 -l 71.63 __
1
'i MEU Fuel Place h
6 I0.46 153 l0.61I T l 60,833 l 5 i 71.63 Fig. 1. La Reina Reactor Fuel Element.
! 1
1 E G F I I I I E I 3 E U U U U O U U U U C U U U U [ l B U U U U A U U U U H 1 2 3 4 5 6 7 8 9 10 W Fig. 2. HEU Core Critical. Configuration. 6 l l
l I E G B B B B B B B B B B F Sk U U U U U U U U Sk I I I I I I E Sk U U U U U U U U Sk D Be Al Al Al Al Al Al Al Al Al N S C Sk U U U U U U U U Sk E I I I I I B At U U U U U U U U Sk A @ Be Be Be Be Be Be Be Be Al H Be Be @ Be Bk Sk Be P P Be 1 2 3 4 5 6 7 8 9 10 W U: Fuel Element Be: Beryllium Reflector Element A1: Solid Aluminium Element Bk: Blanking Element R: End of Rabbit System P: Lead Block Fig. 3. XY Reference Core Configuration. v h
E - G Sk Al Be Be Be Be Be Be At S( F Al Be U U U U U U Be Bk I I I I I I E Be U U U U U U U U Be N o Be u u u u u u u u Be s C Be U U U U U U U U Be E I I I I I B Al Be U U U U U U Be Bk A @ Sk Be Be Be Be Be Be Sk Al H Sk Bk @ Be Bk Bk Be P P Bk i 1 2 3 4 5 6 7 8 9 10 W U: Fuel Element Be: Beryllium Reflector Element Als Solid Aluminium Element Bk: Blanking Element R: End of Rabbit System a P: Lead Block Fig. 4. Equilibrium Core Configuration.
r-- , i l l r l' o o
- f. 80 %
N t A O CROUP
- 1 o
! S , a CROUP s 3 o-E + GROUP s 5 8 i
~
. 5-
{ Q** , e * ! i . I I o a l o !
- l. f. !
e
- . e
- I i l i . x l r
go J" ! ! . u o-o e
~
e
, m I ; a p
! 8 ' M ' 4. 7 I I N - 2 : : a 3 1 a
-i g d-e
~
- . n .
I i - g i ] I< s -- - -- g ! !! !!! i i o
.. e
[ _
; e C
- . . . 6 i . . 6
, I 6 1 0.00 16.00 32.00 48.00 6+.00 90.00 96.00 EAST-WEST DIRECTION
.l b
l' Fig. 5. Kid-Plane Flux in East-West Direction for Groups 1, .3 and 5 at 5 MW. Refe- -
. rence HEU Core.
i______
i o o 6 ce 45 ?. O GROUP s t 8
- GROUP s 3
) , . .
- f. .
+
GROUP s 5 t 8
^ .i i
I d. t Q "@
.4 .
.' l 1 l
o l.
.. g o 1 4 d. ,
t . . 4 ? {
! g ) l I,
2 - t 1 ao I l W* f, e' : g- , 6
,t ,
I
, /
o I + o b- ' 2 '
...e- .- l
./ . , l ; Q c- I i I
,1 f..
~
3 g i . 1 l 1 a r l o
-* f. I
.l .
~
l
. f . t f
. a !
' E g
i
. ,s i 3
- i Ei
- o.
e i o
, o
- o a -
. . . 6 i
- l. *o.co 16.00 32.00 48.00 64 00 80.00 96.00
.} 1 EAST-WEST DIRECTION
.).
Fig. 6. Mid-Plane Flux in East-West Direction for Groups 1, 3 and 5 at 5 MW. Reference MEU Core. i
+
i C t e a
- f. 80 7.- ;
e -,
" O GROUP e t g I
- cRour 3 b-N
" + CROUP e 5 .
j = 7
." d- ;
i o2 i
-i r
n , 1 e i i l i 1 6 l
'. x2
- D i
i
.J $ "
wa 4 ) ,) ! 7
~
4 i-k ,
?
1 2 d- !- I, ~.
! 1 .
i . I 8 {! = t a 3' i
- 8. .
I
- e ,I
- i i
- 8. . .i $
. .I I. I.
. I. .I l S 5 I
-h -
i
' 8 I i g -
} 0 00 t6.00 32.00 48.00 64.00 80 00 96 00 Ca -
; NORTH-SOUTH DIRECTION , ROW 3 j
Fig. 7. Mid-Plane Flux in North-South Direction-for Groups 1, 3 and 5 at 5 MW. Reference
- HZU Core.
8 d-a C GROUP e 1 E a GROUPe 3 N'
* \
+ CROUP e s S
- d- l
.l.* og l .
l t o-
\
M E t 3 . . s a wg d- f ;
, ? f
. 1 8
. \
! 3 i
d-
=
!s !
y :
. I 1. -
. d .
- i' I S. -
-l 3 l
= g i ! ! ! ! ! ! !
- i. [ S l
I a l 1 i 8 1
- g. '
, O.00 16.00 32.'00 44 00 64.00 40.00 96.00 g.
NORTH-SOUTH DIRECil0N , Row 8 - L 1 t Fig. 8. Mid-Plane Flux in North-South Direction for Groups 1, 3 and 5 at 5 W. Reference MEU Core, l i i
o WATER Se Be Be Al Sk 11 6 3 Se Sk
?
15 14 9 4 Be . 18 17 to 2 Se m 16 13 8 5 Se 5
! 12 7 1 Be Sk
?
j . Se Be Be Bk Al I Be : Beryllium Reflector Element. I Sk : Blanking Element. A1 : SolM Aluminium WATER Element. Fig. 9. Fuel Shuffling Paccern Showing Half of the Core. t t 6
e o G-N p. o o o-
= .
o o. o-e in oo
-9
.o- , . .
,.. l l i l HEU
. d-eo ., l
. f ! !. .M. .E.U. . . . .
wo .
. c.
-o e- .. . . ,. ., : -
** .(
l l, ,. vo wa
. l
/ .
a g. : l l : i e wm
~ . . : :-
= - ; ; : l c o , , . .
=o . .
-g. : .
. l l :
A.
,~ :
i - . . : ; g o .i
- o. ,
- e. ,
o i I ' 1
- l i : :
ruct ,l rutt ;scaittiun: WATCR o .g. . . C. l i. ' . l o
- l l l
= .
. g o .
o. C 0 00 8.d0 16.00 26.00 32.00 40 00 48 00 56 00 Eist-west Direetson CM. Tig. 10. Thermal Fluxes at the End of Equilibrium Cycle for HEU and MEU Fuels.
9 g 2 . 4 o, o, 1 3 5 _ . p P. P.
, u.
u.. u a o . o s t r.
- r. r i
.. G. C. C n N u C u_
n_ 0
.. ~ .
_ 0 0 3
.. g
_ ... n o
. t l
a
. l f
u . C L O L E
.. I t
R E a
.., S s
- o
...' *
- i t
.. m. a 0 R
. 0 s e
_ . .. 0
.. 2 x u
. l .
_ L . F n W o
. ei
. I nt
. ac l e
. P r
. i dD i
Mt
. s caeso>o $o h 1h 0
8 -
/t o, - Z s
_ .. 0 5 s
. 0 4E
=.. .. ,
0 3
. I
- I a . .
g
..... L i
F
- E
... ... U
. ~ F
. . ~ .
. . 0
... 0
- 0 me.o am.o om s
ee o
*N.Om44 N N*e,LA
I 9 . l l I I 1.046 0.945 f0.982
, .979 1.042 0.942 It e .
1.012 0.982 1.002 0.940 1.009 0.979 1.000 0.938 ,
, ts is s =
0.993 0.969 1.044 1.0$$ I O.991 0.967 1.043 1.054
- I to 17 to 2 0.994 1.024 1.0$0 0.920 j .994 1.023 1.047 0.918 i
! ' le 13 e I l
'O.972 1.037 1.033 .
l , 0.972 1.032 1.068 11 1 1 A A: Only HEU Fuel. B B: One Fresh MEU Fuel in Post-cion 1. Fig. 12. Form-Tactors for the HEU Equilibrium Core and the First Mixed Core at BOC. 9
O I l 0.982 1.046 0.945 0.987 1.053 0.952 I ti s : 1 1.012 0.982 1.002 0.940 1.000 0.975 1.006 0.949
, is t. s .
0.993 0.969 1.044 1.055 0.968 0.951 1.045 1.061
' to 17 to 2 0.994 1.024 1.050 0.920 0.980 .020 1.055 0.932 ts ts e s e 0.972 1.037 1.033 0.976 .047 .044 12 7 l
l A A: Only HEU Fuel. B B Only MEU Fuel. Fig. 13. Form-Factors for the *dEU and MEU Equilibrium Cores at BOC.
. 1
' f
, 1.299 1.183 - 0.961 0.738 1.297 1.192 0.960 0.741 i .
1.139 1.049 0.884 0.747 a I .137 .048 0.884 0.751 e I . . . . . . . . . A As only HEU Fuel. B B only NEU Fuel. Fig. 14. Form-Factors for the HEU and MEU Reference Cores at the Beginning of Life. I h . - i
F , t 5 W LA MIM MKTot Ca 45 % Inni FMSH MFCROCE CCM 0 A' 0 c 44 A M T 44 T f JR M P
- JE T
P
! N 1
se
- It N N N N 60 MlA 0!STMG FAtW W W KTM LMIN . W Fig. 15. Coolane Temperature Distribution.
1 t 4 g 5 til LA WIM MACTOF 40 % ht!ti FMSN MCAENCC CCM
)77 0
69 N /
# /
[ g S. f*
., j
/ \ ~
n j 4 N ll 0 is a k is W ,, MlK DISTMCC FMth W W KT!VC LAN6TH , C1 Fig. 16. Cladding Temperature Distribucion along the Active Length. t
. e Q $ W La %:!M nTACTCR , MM 45 % tfu (RESH kTfGCQ CCFZ 4M
}
Got i 53% i s,
; ME %
\
N 4N; - e '2 '6 'G 'q 't o 'g2 'g 4 '16 fitti , 5%C Fig. 17. Evolution of the Coolant Flow Due to the F.tilure of One Pump. h gQ S W LA Rtt.M RCACTCR 45 g tCI TRCSH R000C[ CCR[ sui 196 \ no; edd;
#J#j
\
'N s K4; JJ#;
294; e 't 't 't 'e I0 I2 14 It TLYC , frC fig. 18. Evolution of the Coolant Tiow Due to the Failure of Two Pumps.
r-- ., ,
,t . *- l
( l nh
~
$ nW LA MIM MAC100 i
45 % MU fMSV MfEMME CQGC i sad
$N. \
. Sdd;
- l. 4M: s-
- u. \,-
J7e; %
- i 74.; :
2N. \g% l
- i 2fd.
190.
'N t I
i > idd. , 30. j dd; 1 l 0 '2 ~4 't 't, 't I t'2 14 1C 12 M o SEC
.l' I' Fig. 19. Evolution of the Coolant Flow Due to the Failure of Three Pumps.
l a h 1 :
)
[ 5 W LA MIM MKIDt l l I 45 % MV /MSV ACrgggy cogg '
,' c #2 ,
! IM
.d t
' t
, O N.
s 0 14
\ '
', N G pg) rg '
t
, n u,
, 'E ,
- L *C
. 't . Ef: ,
1 45i ., L_,
't
' , ', , . s ,
' 0 4 6 9 10 12 14 it it 20 Tvit , $fC
- t. 1 Tig. 20. Maximus Cladding Temperature Due to i ;
- . , the Failure of Three Pumps.
r 0 l
e s, s
& 5 5 V MM MACTCR ;
l
, '_n g
+1 % frty FMSN MFEMMC C3C 69t SM ,
54$ 49e-444 \. ! M \.
\
I 3% ' 29e . 144 tse , N i .. , 1 s. a 1. ",,
' TIM , SCC
! Fig. 21. Evolution of the Coolant Flow Due to l the Electrical Energy Supply Cutoff.
i
'i .
J $ M M MM MAC709
) y +5 x Mu insk wtunct car c
1 W ,/^ ,
# 18.
D i t 17
'I N ,
I f~ ' / i 7 65 {
- } {K ,
l
~ ? s7 j '
i l 'C g
- 95 t l
; 9 '2 '$ 'C '4 50 52 I4 EG it it
.) Tid o $EC
, 1
; Tig. 22. Maximum Cladding Temperature Due to '
. the Electrical Energy Supply Cucoff.
t +m=h-44 t
.o e .
^
.o* ,.
e Q 4 . f.' t. i. f ' ?I g[y $ M LA Mim MAClot n x Mu FM5W MfEPEXZ CCAC i
< SSe,
.t ,
.l f-g ,
i f- ' 544
- s%
o .: s l t 450 l- ; u 3 16 12 14 - 0 2 4 4 8 it h T M o SEC
.c j Fig. 23. Evolution of the Coolant Flow Due to
- l. 4 One Blockaded Pump.
- l. .
l
.I
_l 1 M LA MIM MACiot i: n x av msk Anremet ccM
.. IN y gM r ,
[ 1 i LM Ay f l ' '. j se;
! M; M
g y< , 1 M l T 40-a g
. i M N ,
,3 P $0 o ' 46-
- N: q
- i. ~' * 'A 4 'l 't E0 12 14 IG 10 N TM , KC l -:
l Fig. 24. Maximus Cladding Temperature Due to l One Blockaded Pump. e
?
i .1 i ! 1 L _-__-_____.____)}}