RRFM 2009 Transactions - European Nuclear Society

More documents

Recommendations

Info

ROLE OF RELAP/SCDAPSIM IN RESEARCH REACTOR SAFETY - RRFM 2009 ROLE OF RELAP/SCDAPSIM IN RESEARCH REACTOR SAFETY C. M. ALLISON, J. K. HOHORST, Innovative Systems Software, LLC Idaho Falls, Idaho 83404 USA iss@relap.com A. J. D’ARCY South African Nuclear Energy Corporation (NECSA) Pretoria, South Africa alan@necsa.co.za ABSTRACT The RELAP/SCDAPSIM code, designed to predict the behaviour of reactor systems during normal and accident conditions, is being developed as part of the international SCDAP Development and Training Program (SDTP). SDTP consists of more than 60 organizations in 28 countries supporting the development of technology, software, and training materials for the nuclear industry. The program members and licensed software users include universities, research organizations, regulatory organizations, vendors, and utilities located in Europe, Asia, Latin America, Africa, and the United States. Innovative Systems Software (ISS) is the administrator for the program. RELAP/SCDAPSIM is used by program members and licensed users to support a variety of activities. The paper provides a brief review of the applications of the code to the simulation and analysis of research reactors in the United States, Europe, Asia, and Africa. An example showing the application of the code to the SAFARI-1 research reactor in South Africa for licensing analysis and for use in an operator training simulator (SAFSIM) is included in the paper. 1.0 Introduction The RELAP/SCDAPSIM code, designed to predict the behaviour of reactor systems during normal and accident conditions, is being developed as part of the international SCDAP Development and Training Program (SDTP) [1,2]. SDTP consists of nearly 60 organizations in 28 countries supporting the development of technology, software, and training materials for the nuclear industry. The program members and licensed software users include universities, research organizations, regulatory organizations, vendors, and utilities located in Europe, Asia, Latin America, Africa, and the United States. Innovative Systems Software (ISS) is the administrator for the program. Three main versions of RELAP/SCDAPSIM, as described in Section 2, are currently used by program members and licensed users to support a variety of activities. RELAP/SCDAPSIM/MOD3.2 and MOD3.4 are production versions of the code and are used by licensed and program members for critical applications such as research reactor and nuclear power plant applications. The most advanced production version, MOD3.4, is also used for general user training and for the design and analysis of severe accident related experiments such as those performed in the Phebus and Quench facilities. In turn, these experiments are used to improve the detailed fuel behaviour and other severe accidentrelated models in MOD3.4 and MOD4.0. MOD4.0 is currently available only to program members and is used primarily to develop advanced modelling options and to support graduate research programs and training. Section 3 highlights various research reactor 1 204 of 455
ROLE OF RELAP/SCDAPSIM IN RESEARCH REACTOR SAFETY - RRFM 2009 applications. An example showing the application of the code to the SAFARI-1 research reactor in South Africa for licensing analysis and for use in an operator training simulator (SAFSIM) is included. 2.0 RELAP/SCDAPSIM RELAP/SCDAPSIM uses the publicly available RELAP/MOD3.3[3] and SCDAP/RELAP5/MOD3.2[4] models developed by the US Nuclear Regulatory Commission in combination with proprietary (a) advanced programming and numerical methods, (b) user options, and (c) models developed by ISS and other members of the SDTP. These enhancements allow the code to run faster and more reliably than the original US NRC codes. MOD3.4 and MOD4.0 can also run a much wider variety of transients including low pressure transients with the presence of non-condensable gases such as those occurring during mid-loop operations in LWRs, in pool type reactors, or in spent fuel storage facilities. The most advanced version of the code, RELAP/SCDAPSIM/MOD4.0[5], is the first version of RELAP or SCDAP/RELAP5 completely rewritten to FORTRAN 90/95/2000 standards. This is a significant benefit for the program members that are using the code for the development of advanced models and user options such as the coupling of the code to other analysis packages. Coupled 3D reactor kinetics and coupled RELAP/SCDAPSIM-SAMPSON [6] calculations are examples where MOD4.0 is used because of a significant reduction in the code development effort and expense to link the packages. MOD4.0 also includes advanced numerical options such as improved time advancement algorithms, improved water property tables, and improved model coding. As a result the code can reliably run complex multidimensional problems faster than real time on inexpensive personal computers. Plant simulation and integrated uncertainty analysis are among the most important applications benefiting from the improved speed and reliability of MOD4.0. MOD4.0 includes many enhanced user options to improve the accuracy of the code or to offer new options for the users. For example, the addition of an alternative material property library designed for Zr- Nb cladding materials is important for VVER and CANDU reactor designs, particularly for severe accident related transients. The addition of an advanced water property formulation is important for many transients, in particular those involving super critical water applications. 3.0 RESEARCH REACTOR APPLICATIONS A combination of RELAP/SCDAPSIM/MOD3.2 and MOD3.4 is being used to analyze research reactors. A brief summary of the early work by several countries was given in Reference [6]. The research reactors noted in this paper include (a) the LVR-15 reactor located at the Nuclear Research Institute in Rez, Czech Republic, (b) the CARR reactor being built in Beijing, China by the China Institute of Atomic Energy, and (c) TRIGA reactors located at the Atomic Energy Research Establishment in Dhaka, Bangladesh and National Nuclear Energy Agency in Bandung Indonesia. LVR-15 is a light-water moderated and cooled pool type reactor with a nominal thermal power of 15 MW. The pool operates at atmospheric pressure with an average coolant temperature in the core of 320 K. The reactor also has closed high pressure/temperature loops suitable for testing of materials under PWR and BWR conditions. The reactor core is composed of several fuel assemblies of Al-U alloy arranged in square concentric tubes. CARR is a tank-in-pool design, cooled and moderated by light water and reflected by heavy water. The rated power is 60 MW. The core consists of plate-type fuel assemblies of Al-U alloy. The Indonesian and Bangladesh TRIGA reactors are pool type reactors with 2 MW and 3 MW thermal power respectively. The reactor cores are composed of solid U-ZrH fuel rods arranged in a hexagonal array and are cooled by water in either forced or natural circulation, depending upon the conditions. More recently, the analysis of two additional reactor types have been reported in References 7-9. The first is for the SAFARI-1 research reactor located in South Africa [7,8]. The second 2 205 of 455
Page 2 and 3:
© 2009 European Nuclear Society Ru
Page 4 and 5:
RESEARCH REACTOR COALITIONS - SECON
Page 6 and 7:
- Eastern European Research Reactor
Page 8 and 9:
ut a comprehensive alignment of tec
Page 10 and 11:
EERRI will be launched. As noted ab
Page 12 and 13:
Nuclear material in the form of hig
Page 14 and 15:
The GTRI domestic radiological mate
Page 16 and 17:
2. State of the art For fuel plate
Page 18 and 19:
It is known since the early days of
Page 20 and 21:
End of 2008 first DU-8wt.%Mo (deple
Page 22 and 23:
References [1] D. AB. Robinson et a
Page 24 and 25:
• CNEA have been working in the f
Page 26 and 27:
O 75 Bignan.doc - DI - 1 / 10 20/02
Page 28 and 29:
O 75 Bignan.doc - DI - 3 / 10 20/02
Page 30 and 31:
O 75 Bignan.doc - DI - 5 / 10 20/02
Page 32 and 33:
O 75 Bignan.doc - DI - 7 / 10 20/02
Page 34 and 35:
O 75 Bignan.doc - DI - 9 / 10 20/02
Page 36 and 37:
THE EAST EUROPEAN RESEARCH REACTOR
Page 38 and 39:
namely Jozef Stefan Institute TRIGA
Page 40 and 41:
Information and organizational issu
Page 42 and 43:
Action Item Table 3. Description (t
Page 44 and 45:
Table 6. Materials/fuel test experi
Page 46 and 47:
In a recent development, it should
Page 48 and 49:
1. Introduction: a renewed context
Page 50 and 51:
interaction, close retention of fis
Page 52 and 53:
3.3 Reference concept and alternati
Page 54 and 55:
changes can be caused by void swell
Page 56 and 57:
For its manufacturing, an additiona
Page 58 and 59:
minor actinides (MA) considered as
Page 60 and 61:
In France, the decision to build a
Page 62 and 63:
eactors (HFR, OSIRIS and then JHR)
Page 64 and 65:
DECOMMISSIONING PROGRESS OF THE DOU
Page 66 and 67:
Page 68 and 69:
Page 70 and 71:
SECURITY OF SUPPLY FOR FISSION MEDI
Page 72 and 73:
But the worst crisis developed by e
Page 74 and 75:
IRE and COVIDEN are following close
Page 76 and 77:
Fig 1. Design of the core and refle
Page 78 and 79:
Table 3: Set of detailed fuel funct
Page 80 and 81:
For the thermal-mechanical analysis
Page 82 and 83:
Behaviour under conditions represen
Page 84 and 85:
6. References [1] - MC. Anselmet, G
Page 86 and 87:
2. Advanced nuclear fuel cycles The
Page 88 and 89:
3.3 India In addition to the ration
Page 90 and 91:
Session II Fuel Development 90 of 4
Page 92 and 93:
1. Introduction Since 1957, date of
Page 94 and 95:
A 3.1. Place a boron insert in a hi
Page 96 and 97:
Finally, a new tool was defined and
Page 98 and 99:
the end user (CEA), to define the f
Page 100 and 101:
Previous papers discussed character
Page 102 and 103:
(a) (b) (c) (d) (e) (f) (g) (h) (i)
Page 104 and 105:
The results of a third reported irr
Page 106 and 107:
In order to reproduce the heat tran
Page 108 and 109:
25 OXIDE THICKNESS 20 Kim et al pH=
Page 110 and 111:
Heavy ion irradiation of UMo7/Al fu
Page 112 and 113:
For the lowest flux values tested d
Page 114 and 115:
127 I beam 127 I beam 0 5 10 15 20
Page 116 and 117:
Weight fractions (%) U(α) UO 2 γU
Page 118 and 119:
activities to meet the need. The co
Page 120 and 121:
Heat capacity Information Thermal c
Page 122 and 123:
Decision point Activity Phase 2: Fu
Page 124 and 125:
IRIS PROGRAM: IRIS4 FIRST RESULTS F
Page 126 and 127:
The main characteristics of the IRI
Page 128 and 129:
4. In-pool plate thickness measurem
Page 130 and 131:
At that stage of IRIS4 irradiation
Page 132 and 133:
extrapolated thermal property value
Page 134 and 135:
optical microscope, acoustic emissi
Page 136 and 137:
CD WIRES AS BURNABLE POISON FOR THE
Page 138 and 139:
Figure 1. MCNP geometry model of fu
Page 140 and 141:
considered carefully: (i) maintain
Page 142 and 143:
discussed and validated together, k
Page 144 and 145:
The as-fabrication Si contents in t
Page 146 and 147:
uniform Si diffusion and consequent
Page 148 and 149:
Recently, fuel relocation induced b
Page 150 and 151:
characteristics of ILs obtained in
Page 152 and 153:
- Type 2 (Si content ≥ 5 wt%) : c
Page 154 and 155: 4. Discussion The main contribution
Page 156 and 157: METHODS OF INCREASING THE URANIUM C
Page 158 and 159: • strength properties of Zr-1% Nb
Page 160 and 161: 4. References 1. Birzhevoy G.A., Ka
Page 162 and 163: Dispersion fuel Monolithic fuel ATR
Page 164 and 165: − because of its Newtonian charac
Page 166 and 167: Session III Back-end 166 of 455
Page 168 and 169: Fig.1. The transfer hall (left in c
Page 170 and 171: The shipment included all the SNF f
Page 172 and 173: FRESH AND SPENT NUCLEAR FUEL REPATR
Page 174 and 175: modifications, spent fuel inspectio
Page 176 and 177: The shipment cleared Romanian Custo
Page 178 and 179: steel are chosen as effective gamma
Page 180 and 181: Fig 3. Long lived nuclides radioact
Page 182 and 183: Session IV Innovative Methods in Re
Page 184 and 185: Lower Gr-reflector Al-clad Upper Gr
Page 186 and 187: All described fuel elements were ca
Page 188 and 189: ANALYSIS OF NEW SAFETY CRITERIA FOR
Page 190 and 191: 3.2 Strain limit The value of 3.65%
Page 192 and 193: The comparison between clad tempera
Page 194 and 195: VALIDATION OF PLTEMP/ANL V3.6 CODE
Page 196 and 197: ΔT p 0.0234 2 ⎪⎧ q[W/cm ] ⎪
Page 198 and 199: Table II. Hot Channel Factors for B
Page 200 and 201: for the U-8Mo fuel meat [5]. The po
Page 202 and 203: correlation of Dittus-Boelter; the
Page 206 and 207: ROLE OF RELAP/SCDAPSIM IN RESEARCH
Page 212 and 213: SEVERE REACTIVITY INJECTION ACCIDEN
Page 214 and 215: expansion and steam formation were
Page 216 and 217: 3 Conclusion Since a few years, IRS
Page 218 and 219: analysis procedure (ENAA) except vi
Page 220 and 221: References Briesmeister, J.F., 2000
Page 222 and 223: Fig. 1 A geometric diagram of NIRR-
Page 224 and 225: Fuel plate temperatures during oper
Page 226 and 227: the primary circuit in the central
Page 228 and 229: Fig. 3: Surface temperature over th
Page 230 and 231: [4] “Test Irradiations of Full Si
Page 232 and 233: The core is made of 1488 stainless
Page 234 and 235: 4. Commissioning the facility for t
Page 236 and 237: ESTABLISHMENT OF A SINGLE-CRYSTAL F
Page 238 and 239: were provided to MU and INL for thi
Page 240 and 241: concrete covered with sheets of 1.2
Page 242 and 243: USE OF FISSION RADIATION IN LIFE SC
Page 244 and 245: Fig. 3. Scan of an etch track film
Page 246 and 247: Summary The unique fission neutron
Page 248 and 249: 2. Methods 2.1 Reactor produced rad
Page 250 and 251: 3.2 Viability studies The results o
Page 252 and 253: DESIGN OF A FLOWING-NAK EXPERIMENTA
Page 254 and 255:
4. Containment rig and sample-holde
Page 256 and 257:
7. Electrical heater The external h
Page 258 and 259:
Page 1 / 7 EVITA: a semi-open loop
Page 260 and 261:
Page 3 / 7 Fig.1: EVITA fuel Genera
Page 262 and 263:
Page 5 / 7 Challenges When operatin
Page 264 and 265:
Page 7 / 7 References: [1] M.C. Ans
Page 266 and 267:
Nuclear Research Institutes, and so
Page 268 and 269:
According to the latest IAEA inform
Page 270 and 271:
18. May Training of operating perso
Page 272 and 273:
2.1 Irradiation Facilities Brief te
Page 274 and 275:
3.1 Detection limits Although it is
Page 276 and 277:
PARTIAL DISMANTLING OF RESEARCH REA
Page 278 and 279:
• Core - liable to full scale rep
Page 280 and 281:
• Necessary individual dosimetry
Page 282 and 283:
SILICON DOPING AT FRM II H. GERSTEN
Page 284 and 285:
The resulting neutron flux density
Page 286 and 287:
4. References [1] Wagner F.M., Knes
Page 288 and 289:
The choice of fuel base and solutio
Page 290 and 291:
3.3 Increasing power beyond current
Page 292 and 293:
A COATING TO PROTECT SPENT ALUMINIU
Page 294 and 295:
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
Page 302 and 303:
After the inspection was finished,
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
Page 320 and 321:
1436 keV, Cs 138 1369 keV, Na 24 2.
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 346 and 347:
UPGRADED REACTOR SYSTEMS FOR ENHANC
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
Page 356 and 357:
1.00E+06 1.00E+05 Volume activity (
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
Page 372 and 373:
0.E+00 2.E+06 1.40E+07 2.E+06 1.20E
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
Page 396 and 397:
etween CRPq specific process and IP
Page 398 and 399:
aspect audits, mainly in relation t
Page 400 and 401:
1 2 3 4 5 6 7 8 9 A B C D E F G H F
Page 402 and 403:
MCNP5 is capable of calculating ter
Page 404 and 405:
Similar to the case of stainless st
Page 406 and 407:
LEU FUEL DISPOSITION AT WWR-M REACT
Page 408 and 409:
Tested fuel assembles are load in t
Page 410 and 411:
The PNPI experience of tests with W
Page 412 and 413:
CURRENT UTILIZATION AND LONG TERM S
Page 414 and 415:
Uusimaa hospital district decided t
Page 416 and 417:
FiR 1 has an important regional rol
Page 418 and 419:
STUDYING OF INFLUENCE OF A NEUTRON
Page 420 and 421:
Definition of crystal structure and
Page 422 and 423:
Microhardness measurement shows, th
Page 424 and 425:
THE SPENT FUEL STORAGE AT THE DALAT
Page 426 and 427:
2. Interim storage of spent fuel On
Page 428 and 429:
A rotating bridge crane of 3.6-ton
Page 430 and 431:
In order to investigate the perform
Page 432 and 433:
(a) (b) Fig. 2. (a) low and (b) hig
Page 434 and 435:
(a) (b) (c) (d) (e) (f) Fig. 5. TEM
Page 436 and 437:
Spot Al Si Mo U Zr G 90.6 7.8 0.6 0
Page 438 and 439:
volume in the material, which affec
Page 440 and 441:
THE STEREOMETRIC ANALYSIS OF GAS PO
Page 442 and 443:
of the crack of pillowing). It show
Page 444 and 445:
One gets the impression that during
Page 446 and 447:
switch from a fast to a thermal spe
Page 448 and 449:
He production to fission ratio vs t
Page 450 and 451:
INSPECTION EXPERIENCE WITH IEA-R1 S
Page 452 and 453:
camera system, received from IAEA i
Page 454:
core in the beginning 2007 (IEA-174
show all

RRFM 2009 Transactions - European Nuclear Society

Create successful ePaper yourself

Delete template?

Save as template?