UNIVERSITÀ DI
UNIVERSITÀ DI PISA
GRUPPO DI RICERCA NUCLEARE GRUPPO DI
GRUPPO RICERCA NUCLEARE – SAN PIERO A GRADO (GRNSPG)
RICERCA NUCLEARE –
SAN PIERO A GRADO (GRNSPG)
JULES HOROWITZ REACTOR
Il reattore sperimentale per applicazioni scientifiche e industriali
10 Novembre 2011
ENEA - Via Giulio Romano 41, Roma
Competenze Italiane Utili per il Italiane Utili per il Progetto JHR il Progetto JHR e Progetto JHR e e
Competenze Competenze Italiane Utili per Opportunità di Ricerca per il Opportunità di Ricerca per il Mondo Accademico
Mondo Accademico
Adorni Dino Araneo,
Araneo Nikolaus Muellner,
Muellner Francesco D
D’Auria
Martina Adorni,
Auria
Any reproduction, alteration, transmission to any third party or publication in whole or in part of this document and/or its content is prohibited unless the University of
Pisa – San Piero a Grado Nuclear Research Group has provided its prior and written consent. This document and any information it contains shall not be used for any
other purpose than the one for which they were provided. Legal action may be taken against any infringer and/or any person breaching the aforementioned obligations.
JULES HOROWITZ REACTOR Grup
ppo Riceerca Nu
ucleare San Pieero a Grrado
The GRNSPG
The San Piero a Grado Nuclear Research Group
‰ Born in December 2003, aims at maintaining and improving the ,
g
p
g
Italian competences in the field of the nuclear technology –
with particular reference to Nuclear Power Plants.
‰ Performs R&D, engineering services, education and training activities, according to the tradition of the Department of Mechanics Nuclear and Production Engineering (DIMNP) and
Mechanics, Nuclear and Production Engineering (DIMNP) and of the University of Pisa (UNIPI).
‰ Led by Prof. Francesco D
Led by Prof. Francesco D’Auria,
Auria, involves more than 60 people involves more than 60 people
(mostly PhD level) among full‐ and part‐time staff and International Experts.
‰ Located in a high‐tech, comfortable and well‐connected site, shared with other research labs (e.g. BIOLAB, INFN).
JULES HOROWITZ REACTOR, 10 Novembre 2011, ENEA ‐ Via Giulio Romano 41, Roma
2/10
JULES HOROWITZ REACTOR FONESYS
Q
Quality y
Assurance
Education dT i i
and Training
Networks of E ll
Excellence
GRNSPG
Editorial ti iti
activities
Nuclear Reactor Technology: Safety &D i
& Design
Innovation
AVAILABLE CODES
Relap5/3D ©
ANSYS-CFX-10.0
3D NK - Nestle
XSEC derivation
Rel evant
ph
ph.
Criteria & Acceptabili ty
Thres holds
TransUranus
Phenomena
consideration
Most suited
(BE) code
Consideration
of qualification:
Nodalization
development
- Code
- Nodalization
- User
3
4
5
6
Select Accept.
Criteria
2
Assumptions
on BIC
Criter ia
Criteria
7
SA
Computational
Platform
Analysis
CA
Coupling with the
SYS TH code
RA
Purposes
8
9
Scenario
selection
Atucha II
Safety
Margins
Performing
the analysis
U
Quantification
?
1
Uncertainty and
sensitivity studies
Acceptability
criteria
CA
10
11
RA
FSAR
CIAU & BF
method
EM/Component
Stress An. Results
Diagram 15.0-1:
International Organizations
Transient core
power distribution
EM/Radiological
Consequences
CFD CFX
Boron distribution in
moderator tank
Linear Heat Rate
(kW/m)
LA
LD
300
LG
250
AL
200
150
AC
Y
100
AF
50
AK
15
18
21
24
30
BE
33
3
6
9
X
36
27
BB
39
12
0
42
TH RELAP5-3D©
Linear heat rate
Fast neutron flux
2
3
4
5
6
7
8
9
10
11
12 1 3 1 4 15
16
17
18
19
20
21 2 2 2 3 24
25
26
27
28
29
30 3 1 3 2 33
34
35
36
37
38
39 4 0 4 1 42
43
44
Analysis of AOO/DBA/SBDBA
Cladding temperature
Coolant pressure
45
1
277
265
257
52 8
5 28
BB 7
518
BA 8
512
BL 9
512
AK 10
5 21
517
511
503
AG 12
496
489
493
493
487
AC 16
488
AB 17 48 2
AA 18
527
51 6
L D 25
524
44 6
438
L C 26
436
428
42 1
42 4
4 29
421
416
4 08
295
294
40 8
40 5
4 08
401
404
4 10
395
3 98
7 BB
325
8 BA
3 25
332
331
330
1 0 AK
1 1 AH
3 42
3 41
340
347
354
9 BL
3 35
3 34
333
340
340
346
355
1 2 AG
3 43
3 49
1 5 AD
3 50
3 56
Fuel Performance/Behavior
1 7 AB
1 8 AA
1 9 AL
2 0 LK
363
372
378
380
900
1 6 AC
3 57
364
371
371
38 2
1 3 AF
1 4 AE
3 48
354
354
362
362
36 8
38 1
3 90
324
323
330
339
35 5
36 1
36 9
37 7
3 77
316
315
323
32 9
33 9
34 5
35 5
3 70
3 86
3 88
314
31 3
32 2
32 1
33 9
3 52
3 61
3 70
3 76
3 85
396
31 3
3 29
3 38
3 45
3 59
3 76
388
394
31 2
3 20
3 20
3 28
3 52
385
387
5 BD
6 BC
30 7
3 12
3 11
3 19
3 38
3 67
367
375
393
401
3 05
3 04
345
360
367
384
392
30 6
3 08
3 11
319
327
353
375
384
40 0
40 2
303
319
337
344
360
375
39 2
40 0
304
310
310
327
351
38 4
39 2
2 99
3 03
302
301
310
336
34 4
36 6
38 3
40 2
4 BE
297
296
301
30 1
31 8
35 8
39 1
40 0
4 09
416
415
414
28 7
30 0
32 6
34 4
37 4
4 09
4 09
412
29 4
29 4
29 3
30 9
3 65
3 99
4 07
411
298
289
289
28 7
28 7
2 93
3 17
3 91
3 99
407
416
422
28 8
2 80
2 50
3 91
406
411
413
3 BF
291
29 0
28 8
2 86
2 86
2 93
250
411
411
411
420
42 3
L A 28
2 80
278
250
250
250
418
419
419
2 79
279
271
250
250
417
425
418
419
29 2
28 4
28 1
2 80
272
271
266
250
250
432
425
42 7
4 31
271
271
251
456
432
43 2
42 6
42 7
28 5
2 82
2 81
274
267
259
251
504
48 3
44 8
42 6
4 30
L B 27
268
267
259
25 1
51 3
46 5
43 3
4 26
4 28
2 BG
2 83
275
275
273
268
26 0
52 2
49 7
43 9
4 40
4 33
4 33
437
26 2
26 0
4 83
4 57
4 40
434
435
277
276
275
269
268
25 2
5 23
5 05
4 75
4 49
449
440
441
442
442
2 60
5 14
4 90
483
466
449
443
443
2 52
5 23
514
499
476
458
458
450
444
26 9
26 2
2 52
523
506
492
484
466
458
450
44 5
253
525
499
477
459
45 9
45 1
269
2 63
2 61
514
506
506
491
484
467
46 8
45 2
4 47
526
515
509
498
47 7
46 0
45 1
4 52
4 53
270
2 64
261
254
526
515
50 0
49 1
48 4
47 8
4 61
4 62
454
455
L E 24
256
255
524
50 9
49 4
4 68
4 69
462
463
464
50 7
5 00
4 78
4 78
469
470
471
464
4 94
4 85
479
479
472
473
474
4 94
493
486
479
481
480
AL 19
L K 20
L H 21
L G 22
L F 23
51 6
5 08
5 01
501
AF 13 49 5
AE 14
AD 15
52 0
5 19
5 10
510
502
AH 11
258
52 9
3 73
2 1 LH
373
38 9
2 5 LD
3 97
2 6 LC
3
4
5
6
7
8
9
10
11
12 1 3 1 4 15
16
17
18
19
20
21 2 2 2 3 24
2 8 LA
25
26
27
28
29
30 3 1 3 2 33
34
35
36
37
38
39 4 0 4 1 42
43
44
45
Depletion PUMA
Burn
up
800
Oxidation at high
temperature
700
~ 1000°C
000 C
B ittl Failure
Brittle
F il
1000
TRANSURANUS
2 7 LB
403
414
29
2
1400
2 3 LF
2 4 LE
29
1
1200
2 2 LG
379
38 9
600
~ 500 kW/m
Embrittlement of
Cladding
800
500
400
600
- NUMBER OF FAILED RODS
300
400
200
Burnup
distribution
Linear Heat Rate
1
1
BG 2
BF 3
BE 4
BD 5
BC 6
Temperaturee
Experimental Activities
Simplified flowchart for
the proposed BEPU
approach for Atucha II
accident analysis
3D-NK NESTLE
45
Grup
ppo Riceerca Nu
ucleare San Pieero a Grrado
The GRNSPG world of competence
- FUEL SAFETY CRITERIA
100
0
0
1
2
3
4
5
6
7
8
9
Time [s]
# and id of failed rods
JULES HOROWITZ REACTOR, 10 Novembre 2011, ENEA ‐ Via Giulio Romano 41, Roma
200
Possible fracture
during quenching
Ballooning / Rupture
0
Fission Release and
Transport
Dose calculations
MELCOR
MAACS
3/10
10
JULES HOROWITZ REACTOR Grup
ppo Riceerca Nu
ucleare San Pieero a Grrado
Specific Competences Related to RRs
‰ International Projects:
¾
¾
¾
¾
IAEA expert missions
P ti i ti t C di t d R
Participation to Coordinated Research Projects
hP j t
Participation to Benchmarks
Preparation of Technical Documents
‰ OECD Activities:
¾ Working Group on Fuel Safety
Working Group on Fuel Safety
¾ Participation to Benchmarks
‰ Pubblications:
¾ More than 15 in International Journals
¾ More than 30 Conference Proceedings
More than 30 Conference Proceedings
¾ More than 20 Technical Reports
JULES HOROWITZ REACTOR, 10 Novembre 2011, ENEA ‐ Via Giulio Romano 41, Roma
4/10
JULES HOROWITZ REACTOR Grup
ppo Riceerca Nu
ucleare San Pieero a Grrado
NUTEMA
Nuclear Power Plant Technology Knowledge
Management System
or
Nuclear Technology Master
‰ Goal Æ
INTEGRATED SYSTEM CAPABLE OF MANAGING THE
OVERALL NPP/RR KNOWLEDGE AND EXPERTISE NEEDED FOR A
SAFETY USE OF NUCLEAR TECHNOLOGY THROUGH THE FULL LIFE
OF AN NPP/RR
‰ Main target Æ Utilities:
•
•
•
•
•
•
•
NPP KNOWLEDGE
CONSTRUCTION MANAGEMENT
MODIFICATION MANAGEMENT
OPERATION & MAINTENACE MANAGEMENT
LICENSING
PERSONNEL
SO
LEARNING
G
DESIGN AUTHORITY ROLE
‰ Principle Æ Defense
Defense-In-Depth
In Depth
JULES HOROWITZ REACTOR, 10 Novembre 2011, ENEA ‐ Via Giulio Romano 41, Roma
5/10
JULES HOROWITZ REACTOR Grup
ppo Riceerca Nu
ucleare San Pieero a Grrado
NUTEMA
NUTEMA NUTEMA
DATABASE DISCIPLINES NPP LIFETIME & Nuclear Power Plant Technology Knowledge Management System COMPUTATIONAL TOOLS
COMPUTATIONAL TOOLS
Or
WORKING MODES Nuclear Technology Master Nuclear Power Plant Technology Knowledge Management System BComponents, Materials
OBJECTIVE L Financing
S Structural
Mechanics
TO MANAGE AND TO PRESERVE THE KNOWLEDGE Or
/TARGETS
Structures
t
- DMS
ASSOCIATED WITH DESIGN, CONSTRUCTION AND OPERATION OF NPP COMPLEX SYSTEMS & St
T Neutron
Physics
M Design
Nuclear Technology Master ThermalDESIGN AUTHORITY U IRIDM
N Construction
‐
CRISIS CENTER
SCOPE
CENDF
or (IAEA INSAG‐19) /INSPIRATION hydraulics
(IAEA INSAG‐25) support support EMERGENCY PREPARADNESSS BASES
O Commissioning
DNJOY
V Radioprotection
E MCNP
P Operation
O
ti
NUCLEAR INDUSTRY ACTORS
END USERS NAMELY UTILITIES W Civil
Engineering
F TRANSURANUS
Q Safety
&
TO MANAGE AND TO PRESERVE THE KNOWLEDGE GANSYS
X Electronics
Licensing
DATABASE HRELAP
WORKING ASSOCIATED WITH DESIGN, CONSTRUCTION AND OPERATION OF NPP COMPLEX SYSTEMS DISCIPLINES NPP LIFETIME Y Informatics
& MODES R Decommissionin
eco
ss o
COMPUTATIONAL TOOLS I NESTLE
Z Chemistry
g
J MACCS –RODOS
BComponents,
Materials
AA
….
L Financing
Mechanics
K…
DESIGN AUTHORITY IRIDMS‐Structural
INTEGRATED RISK CRISIS CENTER
& Structures - DMS
INTEGRATED RISK INFORMED DECISION MAKING OBJECTIVE /TARGETS SCOPE
/INSPIRATION BASES END USERS M Design
T Neutron Physics
Q Safety
&
Licensing
R Decommissionin
g
W Civil Engineering
X Electronics
Y Informatics
Z Chemistry
AA
….
U Thermalor N ConstructionINFORMED DECISION MAKING (IAEA INSAG‐19) CENDF
hydraulics
O Commissioningg
DNJOY
(IAEA INSAG
25) support E MCNP EMERGENCY PREPARADNESSS
support P Operation (IAEA INSAG‐25) support
EMERGENCY PREPARADNESSS
V Radioprotection
INTERACTIVE TRAINING TRAINEE
F TRANSURANUS
GANSYS
HRELAP
I NESTLE
J MACCS –RODOS
K…
NUCLEAR INDUSTRY ACTORS
SUPERVISOR INSTRUCTOR
NAMELY UTILITIES
INTERACTIVE TRAINING TRAINEE
SUPERVISOR INSTRUCTOR
JULES HOROWITZ REACTOR, 10 Novembre 2011, ENEA ‐ Via Giulio Romano 41, Roma
6/10
JULES HOROWITZ REACTOR Identification of Interfaces between Computational Tools
Chain of Codes as used in CNA-2 FSAR Chapter 15
«Accident Analysis»
3D-NK NESTLE
Transient core
power distribution
p
CFD CFX
Boron distribution in
moderator tank
Linear Heat Rate
(kW/m)
LA
LD
300
LG
250
AL
200
150
AC
Y
100
AF
50
0
42
45
30
33
BE
36
X
39
18
21
24
27
3
6
9
12
AK
BB
15
TH RELAP5-3D©
Linear heat rate
F t neutron
Fast
t
flux
fl
2
3
4
5
6
7
8
9
Cladding temperature
Coolant pressure
10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 39 40 41 42 43 44 45
1
1
BG 2
277
BF 3
265
BE 4
257
BD 5
528
BC 6
528
BB 7
518
BA 8
512
BL 9
512
AK 10
521
517
511
503
AH 11
502
AG 12
501
AF 13 495
AE 14
496
489
AD 15
493
493
487
AC 16
488
AB 17 482
AA 18
486
481
480
AL 19
474
LH 21
LF 23
455
453
LE 24
452
447
LD 25
444
445
446
LC 26
442
442
438
466
434
435
436
LB 27
426
428
428
437
427
419
411
409
412
416
422
410
360
400
360
367
359
388
394
401
386
12 AG
11 AH
343
354
356
14 AE
350
16 AC
900
15 AD
357
F lP
Fuel
Performance/Behavior
f
/B h i
17 AB
18 AA
364
19 AL
20 LK
363
372
378
380
13 AF
349
348
354
371
371
381
342
362
362
382
390
398
347
355
361
9 BL
10 AK
341
340
354
368
377
377
388
395
396
355
369
370
376
385
340
346
335
334
333
340
355
361
370
376
385
393
404
345
7 BB
331
339
8 BA
325
332
330
339
352
367
367
387
401
405
345
6 BC
330
339
325
324
323
329
338
352
375
384
392
402
408
345
375
384
392
321
328
316
315
323
322
329
338
353
375
384
400
402
408
344
366
383
320
327
5 BD
314
313
312
320
319
337
351
392
409
408
416
415
414
374
400
409
327
344
391
399
407
319
307
313
312
311
311
319
336
358
391
399
411
416
344
391
406
318
326
365
407
413
421
420
423
250
411
411
419
421
424
250
411
418
418
309
310
306
308
305
304
304
310
310
299
303
303
302
301
301
4 BE
297
296
295
294
301
300
317
250
417
425
419
293
250
432
432
293
298
289
289
287
294
294
293
250
250
425
427
431
430
429
LA 28
448
426
426
250
250
432
433
433
286
278
250
456
439
440
433
251
483
457
449
271
291
290
288
287
287
286
3 BF
292
284
281
288
280
279
279
266
504
465
440
440
441
513
483
449
449
443
259
285
282
280
280
272
271
267
251
497
475
458
458
443
505
483
466
450
450
251
522
490
476
458
459
451
499
484
467
459
451
523
514
492
477
468
460
452
499
484
478
461
462
514
283
281
274
271
271
267
259
2 BG
277
275
275
273
268
268
260
252
276
275
269
268
262
260
252
523
506
491
477
468
469
454
498
484
478
462
463
506
269
262
260
252
523
514
506
491
478
469
470
464
494
485
479
471
464
509
500
494
479
472
473
LK 20
LG 22
494
493
479
507
269
263
261
253
525
526
515
270
264
261
254
524
526
515
509
500
256
255
524
527
516
516
508
501
258
529
520
519
510
510
373
373
21 LH
24 LE
25 LD
27 LB
403
28 LA
414
3
4
5
6
7
8
9
700
10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 39 40 41 42 43 44 45
Depletion PUMA
up
Burn
~ 1000°C
Brittle Failure
1000
TRANSURANUS
26 LC
29
2
800
Oxidation at high
temperature
23 LF
379
389
389
397
29
1
1400
1200
22 LG
600
~ 500 kW/m
Embrittlement of
Cladding
800
500
400
600
- NUMBER OF FAILED RODS
300
400
200
Burnup
distribution
Linear Heat Rate
1
y of AOO/DBA/SBDBA
Analysis
Temperature
Grup
ppo Riceerca Nu
ucleare San Pieero a Grrado
Features of the System
- FUEL SAFETY CRITERIA
200
Possible fracture
during quenching
Ballooning / Rupture
100
0
0
0
1
2
3
4
5
6
7
8
9
10
Time [s]
# and id of failed rods
Fission Release and
Transport
p
Dose calculations
MELCOR
MAACS
JULES HOROWITZ REACTOR, 10 Novembre 2011, ENEA ‐ Via Giulio Romano 41, Roma
7/10
JULES HOROWITZ REACTOR Grup
ppo Riceerca Nu
ucleare San Pieero a Grrado
The System
JULES HOROWITZ REACTOR, 10 Novembre 2011, ENEA ‐ Via Giulio Romano 41, Roma
8/10
JULES HOROWITZ REACTOR Grup
ppo Riceerca Nu
ucleare San Pieero a Grrado
The System
STATUS: The working modality as the database and/or management of computational tools have been proved at the present time, including a few thousands files (data base), a couple of dozen system codes and a few tens of input decks installed. The mode of p
y
p
operation “Safety” is possible.
JULES HOROWITZ REACTOR, 10 Novembre 2011, ENEA ‐ Via Giulio Romano 41, Roma
9/10
JULES HOROWITZ REACTOR Grup
ppo Riceerca Nu
ucleare San Pieero a Grrado
Conclusions
‰ NUTEMA constitutes an integrated system capable of managing the overall NPP/RR knowledge and expertise needed g g
g
p
for a safety use of nuclear technology. The exploitation of the planned capabilities of NUTEMA will require an additional effort. However in the existing configuration the system may already be used to support the Jules Horowitz Reactor analysis, including training and qualification for any level staff.
‰ UNIPI‐GRNSPG is interested in material testing, and can contribute with the design and analysis of results due to the importance for the development and the qualification of materials and nuclear fuel used in the nuclear industry. i l
d
l
f l
di h
l
i d
‰ UNIPI‐GRNSPG will contribute providing man‐power.
JULES HOROWITZ REACTOR, 10 Novembre 2011, ENEA ‐ Via Giulio Romano 41, Roma
10/10
Grup
ppo Riceerca Nu
ucleare San Pieero a Grrado
JULES HOROWITZ REACTOR THANKS FOR
YOUR
ATTENTION!
[email protected]
d i@i
i i it
JULES HOROWITZ REACTOR, 10 Novembre 2011, ENEA ‐ Via Giulio Romano 41, Roma
11/10