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: ahainoun@aec.org.sy

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: