RRFM 2009 Transactions - European Nuclear Society

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UPGRADED REACTOR SYSTEMS FOR ENHANCED SAFETY AT TRIGA-INR Marin Preda, Marin Ciocanescu, Emil Mugurel Ana Institute for <strong>Nuclear</strong> Research, Str. Campului no. 1, Pitesti, Romania Abstract After almost three decades of operation of stationary TRIGA 14MW with systems provided and installed at reactor first start-up, it appeared obvious that an extended modernization program is required, both for enhancing the nuclear safety and to expand the facility lifetime. A first step has been achieved through complete HEU to LEU core conversion, meaning also core refuelling possibility for the future. Systems that have been subjected to the upgrading program are: control rods, radiation monitoring, data acquisition and processing, ventilation, irradiation devices, and above all, the outstanding modernization of the I&C system, including a brand new reactor control desk. Taking into account own and research reactors community operation experience, IAEA guides and recommendations, the basic requirement for the Instrumentation and Control System is the separation between safety and operation components, in order to decrease human error consequences and avoid common cause failures. Modernization did not cover any sensor replacement, but preserve the present scram logic and conditions (as given and approved in the Safety Report and Licensed Limits and Conditions) The entire modernization program is performed according to QA system. Out of intrinsic nuclear safety enhancement, enhanced population and environment protection is a concern and an expected result of the program. Upgrading the overall performances of the reactor and extending its operational lifetime, the Reactor Department of Institute will be able to perform competitive irradiation tests for nuclear fuel and materials, and to continue to develop nuclear investigation techniques or isotope production. 1. Introduction TRIGA reactor from Institute for <strong>Nuclear</strong> Research Pitesti, Romania was commissioned in 1980 (first criticality has reached on November 17th 1979). In fact, as it could be seen in the Figure 1, the TRIGA reactor consists in two reactors: A Steady State Reactor, 14MW, initially loaded with HEU fuel (93% enrichment). An Annular Core Pulsing Reactor of 20.000MW. Figure1 TRIGA reactor pool arrangement DRY CAVITY ACPR REACTOR UNDERWATER NEUTRONOGRAPHY FUEL SSR 14MW REACTOR ROD CONTROL (8) EXPERIMENTAL LOCATIONS FOR LOOPS AND CAPSULES BERILIUM REFLECTOR PLUG LOADING TUBE THERMAL COLUMN STANDARD FOR NEUTRON FLUX CALIBRATION SILICON DOPING CAVITIES EXPERIMENTAL LOCATIONS IN REFLECTOR RADIAL CHANNEL REACTOR TANK TANGENTIAL CHANNEL REACTOR ENVELOPE DRY NEUTRONOGRAPHY PGNAA FACILITY TANGENTIAL CHANNEL 346 of 455 RADIAL CHANNEL NEUTRON DIFFRACTOMETER
During time the experimental facilities were used to perform the following test types: - Loop A*: o overpower type tests on fuel element, o power ramp type tests on fuel element, o corrosion and mechanical behavior studies on structural materials used in CANDU pressure tubes, o LOCA type tests and on line and off line water chemistry control: pH - 6 ÷ 10,5; conductivity O2 – 20 ÷ 100ppB; H adition; solid residues. - Capsule C1, C2*: o Fuel element dimensional measurement o Fission products pressure – on line o Power ramp o Short-time irradiation for residual deformation of the cladding determination o Central temperature measurement in the fuel element o Fission gases release effects on the measured temperature during irradiation o Fission gases composition for fuel element o Densification - fuel element - Capsule C5*: o Structural materials irradiation tests in inactive environment: Zircalloy-4, steel 403-M, Zr-2,5%Nb until 2,3X1022 nvt o Irradiation and tensile test of Chorpy standard minisamples – maximum 30 samples per irradiation campaign - Capsule C6*,**,***: o Thermomecanical behavior of CANDU type fuel element in fast power transients o Analysis of fuel elements clad failure limits and mechanisms for pellet clading interaction o Determination of energy level for fuel element failure depending on its geometry and microstructural characteristics o Studies on clad-fuel mechanical interactions o Database development regarding fuel element behavior in transient regimes - Capsule C9*,**: o Cycling tests on fuel elements that should confirm the fuel capacity to support a wide range of power cycling that occurs in normal operation of a CANDU reactor during power load following. * All irradiation devices are equipped with digital control system. ** The irradiation data are on line gathered and processed. *** C6 irradiation device is equipped with fast data acquisition system for fast transient processes. TRIGA reactor core conversion that was completed in June 2006 allowed us to restore the standard configuration with 29 fuel bundles. The performance tests illustrate the parameters the Table1 and 2: Table1 TRIGA SSR parameters comparison between 29 HEU and 29 LEU bundles core: Parameter HEU LEU Fuel bundles 29 29 Reactor Power 14 MW 14 MW Inlet Temperature 37°C 37°C Flow rate with one pump in primary cooling system 400 Kg/s 400 Kg/s Flow rate with two pumps in primary cooling system 700 Kg/s 700 Kg/s Safety limit for nuclear fuel 1150 °C 1150 °C Фmax (XC1 location) 2.9x10 14 (cm 2. s) -1 2.82 x10 14 (cm 2. s) -1 PPFmax 2.5 2.14 APFmax 1.30 1.313 CPFmax 3.25 3.01 347 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
Page 296 and 297: 5. Conclusions 1. Immersion of AA 1
Page 298 and 299: KEEPING AGING RESEARCH REACTORS IN
Page 300 and 301: inspected in presence of an expert
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Page 304 and 305: 1 A STRATEGY FOR MANAGING AGEING CO
Page 306 and 307: 3 2. With the aluminium-clad HEU fu
Page 308 and 309: 5 The pool and the support structur
Page 310 and 311: Fig.2. Reflector element of JRR-4.
Page 312 and 313: Dimensional change (%) 0.6 0.5 0.4
Page 314 and 315: SAFETY CONSIDERATIONS FOR CORE MANA
Page 316 and 317: 3 3. Core operation and monitoring
Page 318 and 319: 5 temperature, etc.). Effective sec
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Page 322 and 323: compared with that of the delayed n
Page 324 and 325: AD-HOC AND PREVENTATIVE MAINTENANCE
Page 326 and 327: Page 3 of 7 3. SAFARI-1OPERATIONAL
Page 328 and 329: Page 5 of 7 programs for instrument
Page 330 and 331: Page 7 of 7 DATE DESCRIPTION DATE D
Page 332 and 333: MICROSTRUCTURAL ANALYSIS OF MTR FUE
Page 334 and 335: In sample S3, similar zones could b
Page 336 and 337: increased temperature, the solid st
Page 338 and 339: Figure 8 BEI images covering Sample
Page 340 and 341: 660 °C, but since in the former me
Page 342 and 343: support. Both Governments have eval
Page 344 and 345: • Removal of sludge from the SNF
Page 348 and 349: Parameter HEU LEU a 6.0x10 -5 (°K)
Page 350 and 351: From 1992 until 2005 the TRIGA-SSR
Page 352 and 353: Figure9 - Dependency between reacto
Page 354 and 355: URANIUM CONTAMINATION IN PRIMARY CI
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Page 358 and 359: 6. Acknowledgement This work was pe
Page 360 and 361: 2. Steady State Thermal Hydraulic A
Page 362 and 363: in the scram, and the respective re
Page 364 and 365: CORROSION BEHAVIOR OF ALUMINIUM ALL
Page 366 and 367: 3.2. Pitting corrosion of 1050 and
Page 368 and 369: 4. Discussions This study shown tha
Page 370 and 371: The Steady State core was fully con
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Page 374 and 375: SYNTHESIS AND CHARACTERIZATION OF G
Page 376 and 377: 0 2 4 6 8 10 120 1,2 100 1,0 80 70
Page 378 and 379: Element, Wt %, At %, K‐Ratio, Z,
Page 380 and 381: U/cc, but the target uranium densit
Page 382 and 383: Fig. 5 Macroscopic cutting view (x
Page 384 and 385: SELF-ASSESSMENT OF APPLICATION OF T
Page 386 and 387: 2. The Code of Conduct on the Safet
Page 388 and 389: For emergency preparedness, conside
Page 390 and 391: inside of a compaction die by blowi
Page 392 and 393: Fig. 4. A compacted and sintered lu
Page 394 and 395: THE MANAGEMENT SYSTEM EVOLUTION OF
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etween CRPq specific process and IP
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aspect audits, mainly in relation t
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1 2 3 4 5 6 7 8 9 A B C D E F G H F
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MCNP5 is capable of calculating ter
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Similar to the case of stainless st
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LEU FUEL DISPOSITION AT WWR-M REACT
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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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