Towards Standard Methodology in the Safety
Analysis of Research Reactors
Dr. A. Hainoun,
Head, Reactor Safety Division and Energy System Analysis
AECS, Nuclear Engineering Department
E-mail: [email protected]
International Conference on Research Reactors:
Safe Management and Effective Utilization
14-18 November 2011, Rabat, Morocco
Content
Background;
Safety Analysis Scheme of RR;
Qualification of Safety Analysis Codes;
Examples: Application on Selected RR;
Role of IAEA in Adopting a Safety Analysis
Codes;
Conclusion.
Classification of Research Reactors
Low to medium flux reactors:
High flux reactors:
Highest flux reactors:
scm 21410
scmscm 214214 10510
scm 214105
There is a wide spectrum of RR types:
Various RR and FE Types
Neutron flux vs. reactor power for various research reactor types
Selected RR
.
Compact core
design
Specific Features of RR
Wide spectrum of RR types
Compact core
design
Short FE cycle
length
Small
fission product
inventory
Fast transient behavior
Small
fission product
inventory
Short FE cycle
length
Small
fission product
inventory
High power density
High heat flux
Low fuel melting point
Low system
pressure
Low hazard
potential
Various FE Types
Involutes Fuel Plates
Scope of RR Safety Analysis
Analysis of event sequences and evaluation
of PIEs consequences,
and comparison of the achieved analysis
results with design limits and radiological
acceptance criteria.
General Aspects of RR
Despite the difference between RR and
power reactors, the safety objective is the
same;
Variety of RR has limited the effort to
establish detailed standard approach and the
development of comprehensive Safety
Analysis Codes for RR.
Safety Aspects of RR
Safety Limits:
Sub-cooled boiling (ONB, OSV);
Thermal hydraulic instability (OFI),
Parallel channel instability,
DNB (saturated or subcooled).
Safety Aspects of RR (Some PIE)
RIA, LOFA, LOCA;
Loss of elect. power
Internal and external events,
Human errors, etc..
Specification of core, FE
and loops geometry
Neutronic Analysis:
Criticality and burn-up
(MCNP, CITATION, …)
TH Analysis:
SS, Transients
(RELAP, ATHLET, MERSAT,
CATHARE, PARET, CATENA)
Key dynamic
parameters:
F, , b, a, L
Core
specification
Distribution of
T, v, p, e
Safety limits
and margins:
DNBR, OFI
Design&
Safety Analysis Safety analysis of DBA: RIA, LOFA, LOCA, ATWS
(TH Codes +3D-Kinetic)
Approach for Comprehensive Safety Analysis
Verification and Validation Procedure
Sensitivity analysis
Specifying the Variables Ranges
Setting of Validation Criteria
Validation Experiments
Validated Code
Separate Effect Experiments
Problem Selection
Computer Code
Code Modification
Additional Experiments or Code Modifications Experiment Analysis and
Comparison with Code Results
Ranking of Sensitivity
Validation Matrix (single &integral effect test)
Physical Phenomena
ONB, OSV
OFI & PCI
DNB & Transition Boiling
Fuel Melting
Flow Reversal
Natural Circulation
Reactivity Feedback & 3D
Effects
Axial Void
Distribution A P
Static
Instability Experiments
A P
Parallel
Channel
Instability
P A P P
RIA P A
LOFA P A P P P P
LOCA A A P
Loss of Heat
Sink P A P
Ex
per
imen
ts
Two Phase
Heat Transfer A A P
Covering range of the experiment regarding the physical phenomenon:
A: Appropriate for code validation, P: Partially appropriate for code validation
Selected Examples:
MNSR, IAEA-10 MW, FRJ-2, FRM-2
ETRR2, RSG-GAS, IEA-R1, MNR, SPERT-IV
قسم الهندسة النىوية هيئة الطبقة الذرية السىرية
MSNR Models (MERSAT, ATHLET, RELAP)
MCNP 3-D Model of MNSR
TH Experiments and Validation Results
0 50 100 150 200 250 300 350 400 450 500 550
20
25
30
35
40
45
Tout (o
C)
Time (s)
0.000
0.002
0.004
0.006
0.008
0.010
0.012
0.014
Vav (
m/s
)
20.000
20.500
21.000
21.500
22.000
22.500
Tin-ATHLET
Tout-ATHLET
Tin-Experiment
Vav
Tout-Experiment
Tin (
oC
)
0 2000 4000 6000 8000 10000
20
22
24
26
28
30
32
34
36
38
ATHLET
Experiment
Inle
t T
em
pera
ture
Tin(c
)
Time (sec)
0 100 200 300 400 500 600
0.0
0.5
1.0
1.5
2.0
2.5
3.0
3.5
Relative Average Core Temperature
Relative Reactor Power
Experiment
ATHLET
Experiment
ATHLET
Re
lative
Un
its
Time [s]
LOFA and RIA Analysis of IAEA-RR (10 MW)
RELAP& MERSAT
Nodalization Model
LOFA Analysis of IAEA-RR
30
40
50
60
70
80
90
100
110
120
50 55 60 65 70
Time (sec)
Tem
per
atu
re (
C)
Fuel Clad Coolant
Onset of flow reversal
HC Temperatures during LOFA.
RIA Analysis of IAEA-RR
0
50
100
150
200
250
50 50.5 51 51.5 52
Time (sec)
Tem
pera
ture
(C
)
Fuel Clad Coolant
PEAK POWER TIME
HC Temperatures during Fast RIA (1.5 $/0.5 sec)
40
42
44
46
48
50
580 600 620 640 660 680 700 720 740 760 780 800 820 840 860 880 900
Time (Sec)
Te
mp
era
ture
(c)
Tinlet ( UP)
Toutlet(LP)
Core inlet and outlet
coolant temperatures
during the LOFA
LOFA Simulation of IEA-R1
40
45
50
55
60
65
70
75
80
85
90
600 650 700 750 800 850 900
Time (sec)
Tem
pera
ture
(C)
Tclad-10
Tfuel-10
Flow Reversal (ATHLT Application)
ETRR2 Reactor
MERSAT Nodalization Model
On going IAEA’s-CRP1496 on:
“Innovative Methods in Research Reactor Analysis”
Scope of the Project:
Assessment and qualification of selected
computational codes for the application in
neutronic, thermal hydraulic and safety
analysis of research reactors
Development Phase:
Establishing a technical working group to initiate the
development program (adopting modular structure),
Utilizing the experience in restructuring the advanced codes
like RELAP, ATHLET, CATHARE,…with emphasis on 3DN
Conclusion
The performance of standard safety analysis for RR
require the establishment of qualified deterministic
safety analysis code (TH-system code& 3-D Neutronic)
IAEA could start an initiative to set up such SAC !
Proposal for possible working program
Conclusion
Testing and Verification:
IAEA’s WG and selected teams in MS,
Validation:
Establishing a robust validation matrix using RR data
base,
Distribution to MS for initial test to receive feedbacks
and recommendations,
Establishing and freezing the first version:
accumulating user recommendations for periodical
updating,
Improve the performance and utilization of RR
especially in developing countries,
Enhance the safety culture in MS by exchange of
experiences in RR safety analysis,
Open possibility to simulate combined event
sequences (lesson learned from F-D accident).
Conclusion
This effort could support the standardization of SA
of RR resulting in:
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