RRFM 2009 Transactions - European Nuclear Society

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6. Acknowledgement This work was performed within the scope of research intent MSM 267 224 4501 of the Ministry of Education, Youth and Sports. 7. References [1] NRI Rez, 2008. Research Reactor LVR-15, [2] Viererbl, L., Lahodová, Z., Horák A., Marek M., Voljanskij A., 2008. Radionuclide Activity Measurement in Primary Circuit Water of the LVR-15 Research Reactor, Conference proceedings, XXX days of radiation protection, Liptovský Ján, Slovak Rep. [3] Viererbl, L., Lahodová, Z., Klupák, V., Marek, M., Voljanskij, A., 2007. Source of Radionuclides in Primary Circuit Water of LVR-15 Reactor, RRFM-IGORR conference, Lyon, France. [4] Viererbl, L., Lahodová, Z., Horák, A., Marek, M., Voljanskij, A., 2005. Analysis of Radionuclide Generation in Primary Circuit Water of LVR-15 Reactor, Conference proceedings, XXVII days of radiation protection, Liptovský Ján, Slovak Rep. [5] Viererbl, L., Marek, M., Voljanskij, A., Brož, V., Burian, J., 2004. Detection of Damaged Fuel Assembly in LVR-15 Reactor with Spectrometric Water Activity Measurement, International Conference on Topical Issues in Nuclear Installation Safety, Beijing, China. - 5 - 358 of 455
ANALYSES OF IRT, SOFIA LEU CORE: STEADY-STATE THERMAL-HYDRAULIC AND ACCIDENT ANALYSES S.I. BELOUSOV, T.G. APOSTOLOV Institute for Nuclear Research and Nuclear Energy of Bulgarian Academy of Science Tsarigradsko 72, 1784 Sofia, Bulgaria N.A. HANAN, J.E. MATOS RERTR Program, Argonne National Laboratory Argonne, IL 60439-4815 USA ABSTRACT The initial LEU (IRT-4M fuel assemblies, 19.75% 235 U) core of the new IRT, Sofia research reactor of the Institute for Nuclear Research and Nuclear Energy (INRNE) of the Bulgarian Academy of Science, Sofia, Bulgaria is jointly analyzed with the RERTR Program at Argonne National Laboratory (ANL) to evaluate its characteristics important to safety. Analyses are carried out for the fuel properties presented in the actual fuel Catalogue Description specified for IRT, Sofia research reactor. The initial configuration using 16 fuel assemblies (four 8-tube and twelve 6-tube fuel assemblies) is analyzed at a critical core state preferable for BNCT tube operation. Results of detailed steady-state thermal-hydraulic calculations and transients for this initial core configuration are presented in this paper. These results show that one pump in the primary circuit is sufficient for safe operation at 200, 500 and 1000 kW reactor power levels and that safety is maintained for all transients. 1. Introduction A joint study concerning IRT, Sofia research reactor (RR) between INRNE and the RERTR Program at ANL was initiated in 2002. The initial steps studies [1-4] were mainly focused on neutronics properties significant for reactor application and safety analyses. The results of further neutronics thermal hydraulic and accident analyses significant for safety assessment of the LEU core for the critical state preferable for Boron Neutron Capture Therapy (BNCT) beam tube operation were presented too [5, 6]. The BNCT activity is considered nowadays as one of the most important for future IRT, Sofia application. However these results were based on the preliminary data for the LEU fuel assemblies (FA) [7, 8] that differed from the actual fuel data [9] which became available later. Presented here are results of the steadystate and transient thermal hydraulic calculations using the actual LEU fuel assemblies’ data. The actual fuel data that are most significant for the safety assessment analysis and differ from the previously used include: the FA maximum inlet coolant temperature; the maximum admissible temperature on the fuel element (FE) surface; and onset of nucleate boiling ratio (ONBR). Moreover the hot channel factors (HCF) impact is not considered in the actual fuel passport. The actual fuel data include the temperature of ONB (118ºC) that was not presented in the previously used data. The actual catalogue description also defines the fraction of the reactor power generated in the FE equal to 95% whereas the value of 94% was used in our previous calculations. In the safety assessment it is important to remember that increasing this fraction leads to corresponding increase of the peak power density in the core; this needs to be included in the thermal-hydraulics analyses (steady-state and transients). 359 of 455
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© 2009 European Nuclear Society Ru
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RESEARCH REACTOR COALITIONS - SECON
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- Eastern European Research Reactor
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ut a comprehensive alignment of tec
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EERRI will be launched. As noted ab
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Nuclear material in the form of hig
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The GTRI domestic radiological mate
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2. State of the art For fuel plate
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It is known since the early days of
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End of 2008 first DU-8wt.%Mo (deple
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References [1] D. AB. Robinson et a
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• CNEA have been working in the f
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O 75 Bignan.doc - DI - 1 / 10 20/02
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THE EAST EUROPEAN RESEARCH REACTOR
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namely Jozef Stefan Institute TRIGA
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Information and organizational issu
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Action Item Table 3. Description (t
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Table 6. Materials/fuel test experi
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In a recent development, it should
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1. Introduction: a renewed context
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interaction, close retention of fis
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3.3 Reference concept and alternati
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changes can be caused by void swell
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For its manufacturing, an additiona
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minor actinides (MA) considered as
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In France, the decision to build a
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eactors (HFR, OSIRIS and then JHR)
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DECOMMISSIONING PROGRESS OF THE DOU
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SECURITY OF SUPPLY FOR FISSION MEDI
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But the worst crisis developed by e
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IRE and COVIDEN are following close
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Fig 1. Design of the core and refle
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Table 3: Set of detailed fuel funct
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For the thermal-mechanical analysis
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Behaviour under conditions represen
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6. References [1] - MC. Anselmet, G
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2. Advanced nuclear fuel cycles The
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3.3 India In addition to the ration
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Session II Fuel Development 90 of 4
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1. Introduction Since 1957, date of
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A 3.1. Place a boron insert in a hi
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Finally, a new tool was defined and
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the end user (CEA), to define the f
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Previous papers discussed character
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(a) (b) (c) (d) (e) (f) (g) (h) (i)
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The results of a third reported irr
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In order to reproduce the heat tran
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25 OXIDE THICKNESS 20 Kim et al pH=
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Heavy ion irradiation of UMo7/Al fu
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For the lowest flux values tested d
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127 I beam 127 I beam 0 5 10 15 20
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Weight fractions (%) U(α) UO 2 γU
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activities to meet the need. The co
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Heat capacity Information Thermal c
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Decision point Activity Phase 2: Fu
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IRIS PROGRAM: IRIS4 FIRST RESULTS F
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The main characteristics of the IRI
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4. In-pool plate thickness measurem
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At that stage of IRIS4 irradiation
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extrapolated thermal property value
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optical microscope, acoustic emissi
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CD WIRES AS BURNABLE POISON FOR THE
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Figure 1. MCNP geometry model of fu
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considered carefully: (i) maintain
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discussed and validated together, k
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The as-fabrication Si contents in t
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uniform Si diffusion and consequent
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Recently, fuel relocation induced b
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characteristics of ILs obtained in
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- Type 2 (Si content ≥ 5 wt%) : c
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4. Discussion The main contribution
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METHODS OF INCREASING THE URANIUM C
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• strength properties of Zr-1% Nb
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4. References 1. Birzhevoy G.A., Ka
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Dispersion fuel Monolithic fuel ATR
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− because of its Newtonian charac
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Session III Back-end 166 of 455
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Fig.1. The transfer hall (left in c
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The shipment included all the SNF f
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FRESH AND SPENT NUCLEAR FUEL REPATR
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modifications, spent fuel inspectio
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The shipment cleared Romanian Custo
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steel are chosen as effective gamma
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Fig 3. Long lived nuclides radioact
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Session IV Innovative Methods in Re
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Lower Gr-reflector Al-clad Upper Gr
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All described fuel elements were ca
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ANALYSIS OF NEW SAFETY CRITERIA FOR
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3.2 Strain limit The value of 3.65%
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The comparison between clad tempera
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VALIDATION OF PLTEMP/ANL V3.6 CODE
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ΔT p 0.0234 2 ⎪⎧ q[W/cm ] ⎪
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Table II. Hot Channel Factors for B
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for the U-8Mo fuel meat [5]. The po
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correlation of Dittus-Boelter; the
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ROLE OF RELAP/SCDAPSIM IN RESEARCH
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SEVERE REACTIVITY INJECTION ACCIDEN
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expansion and steam formation were
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3 Conclusion Since a few years, IRS
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analysis procedure (ENAA) except vi
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References Briesmeister, J.F., 2000
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Fig. 1 A geometric diagram of NIRR-
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Fuel plate temperatures during oper
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the primary circuit in the central
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Fig. 3: Surface temperature over th
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[4] “Test Irradiations of Full Si
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The core is made of 1488 stainless
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4. Commissioning the facility for t
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ESTABLISHMENT OF A SINGLE-CRYSTAL F
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were provided to MU and INL for thi
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concrete covered with sheets of 1.2
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USE OF FISSION RADIATION IN LIFE SC
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Fig. 3. Scan of an etch track film
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Summary The unique fission neutron
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2. Methods 2.1 Reactor produced rad
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3.2 Viability studies The results o
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DESIGN OF A FLOWING-NAK EXPERIMENTA
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4. Containment rig and sample-holde
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7. Electrical heater The external h
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Page 1 / 7 EVITA: a semi-open loop
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Page 3 / 7 Fig.1: EVITA fuel Genera
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Page 5 / 7 Challenges When operatin
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Page 7 / 7 References: [1] M.C. Ans
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Nuclear Research Institutes, and so
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According to the latest IAEA inform
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18. May Training of operating perso
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2.1 Irradiation Facilities Brief te
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3.1 Detection limits Although it is
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PARTIAL DISMANTLING OF RESEARCH REA
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• Core - liable to full scale rep
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• Necessary individual dosimetry
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SILICON DOPING AT FRM II H. GERSTEN
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The resulting neutron flux density
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4. References [1] Wagner F.M., Knes
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The choice of fuel base and solutio
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3.3 Increasing power beyond current
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A COATING TO PROTECT SPENT ALUMINIU
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The treatment consisted of simple i
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5. Conclusions 1. Immersion of AA 1
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KEEPING AGING RESEARCH REACTORS IN
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inspected in presence of an expert
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After the inspection was finished,
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1 A STRATEGY FOR MANAGING AGEING CO
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3 2. With the aluminium-clad HEU fu
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Page 332 and 333: MICROSTRUCTURAL ANALYSIS OF MTR FUE
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Page 374 and 375: SYNTHESIS AND CHARACTERIZATION OF G
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Page 378 and 379: Element, Wt %, At %, K‐Ratio, Z,
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Tested fuel assembles are load in t
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The PNPI experience of tests with W
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CURRENT UTILIZATION AND LONG TERM S
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Uusimaa hospital district decided t
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FiR 1 has an important regional rol
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STUDYING OF INFLUENCE OF A NEUTRON
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Definition of crystal structure and
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Microhardness measurement shows, th
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THE SPENT FUEL STORAGE AT THE DALAT
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2. Interim storage of spent fuel On
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A rotating bridge crane of 3.6-ton
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In order to investigate the perform
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(a) (b) Fig. 2. (a) low and (b) hig
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(a) (b) (c) (d) (e) (f) Fig. 5. TEM
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Spot Al Si Mo U Zr G 90.6 7.8 0.6 0
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volume in the material, which affec
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THE STEREOMETRIC ANALYSIS OF GAS PO
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of the crack of pillowing). It show
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One gets the impression that during
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switch from a fast to a thermal spe
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He production to fission ratio vs t
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INSPECTION EXPERIENCE WITH IEA-R1 S
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camera system, received from IAEA i
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core in the beginning 2007 (IEA-174
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