Journal of Nuclear Science, Engineering and Technology (JONSAT)
In cooperation with the Iranian Nuclear Society

Control rods effect analyses on thermo-hydraulic safety parameters in TRR

Document Type : Research Paper

Authors

A. Lashkari
F. Khoshahval
M. Amin Mozafari

Reactor and Nuclear Safety Research School, Nuclear Science and Technology Research Institute, AEOI, P.O.Box: 14155-1339, Tehran – Iran

https://doi.org/10.24200/nst.2020.1114
Abstract
In this study, the effect of control rods on TRR safety parameters was investigated in both cases of static and transient behavior. In this paper, the effect of the control rods movements (70%) on the power peaking was used in the form of the normalized function instead of the COS function in the TERMIC cod. The results of the new function were investigated on thermo-hydraulic parameters in the steady-state and transient conditions. In the steady-state with 5 MW thermal power, the results showed that the safety margin in new condition (70% insertion) decreased respect to the state that all control rods where out. Also, in the transient section, the axial function of the power distribution was applied in the input of the PARET cod. The direct influence of control rods on the power peaking was the production of nucleate boiling earlier than the COS shape function. This effect, reduced the power and temperatures due to the void coefficient feedback. In the last section, the effect of the control rods insertion on the reactivity feedback coefficient was considered in the transient scenario. Increment of the reactivity feedback coefficient due to control rods insertion along with the direct effect of control rods caused a significant decrease in all thermo-hydraulic safety parameters.

Highlights

1.             AEOI, Safety Analysis Report for the Tehran Research Reactor, Tehran-Iran, (2018).

 

2.             A. Lashkari, Experimental and numerical analyses of control rods effect on Tehran Research Reactor power peaking factor, Iranian Physic Conference, Mashhad university, (1394) (in Persian).

 

3.             A. Lashkari et al. Neutronic analysis for Tehran research reactor mixed-core, Pnucene, 60, 31–37 (2012).

 

4.             M. Torabi, et al. Neutronic analysis of control rod effect on safety parameter s in Tehran Research Reactor, Nuclear Engineering and Technology, 50, (2018), doi.org/10.1016/j.net.2018.05.008.

 

5.             INVAP, MTR_PC V. 3.0. Neutronic, Thermal-Hydraulic and Shielding Calculations on Personal Computers. Nuclear Engineering Division, INVAP, Bariloche. (2006)..

 

6.             C.J. Taubman, J.H.L., WIMS-D4 Code, A Genera description of the Lattice Code WIMS, United Kingdom Atomic Energy Authorit.

 

7.             E. Villarino, CITVAP v3. 1 Improved version of CITATION II. MTR_PC v2, (2001).

 

8.             INVAP, P.A. Termic V4.1, A Program for the thermal-Hydraulic Analysis of A MTR Core, In Forced Convection. (2003).

 

9.             E., P.A.-I.S., CAUVAP v3.60, A computer program for flow distribution and pressure drop calculation in a MTR type core. (2003).

 

10.          A.E. Bergles, W.M. Rohsenow, Forced-Convection Surface Boiling Heat Transfer and Burnout in Tubes of Small Diameter, DSR Report No. 8767-21, Department of Mechanical Engineering, Massachusetts Institute of Technology, Cambridge, Massachusetts, (1962).

 

11.          IAEA-TECDOC-233, Research Reactor Core Conversion from the use of Highly Enriched Uranium to the use of Low Enriched Uranium, Appendix A 'U.S. Generic Enrichment Reduction calculations for plate type and rod type reactors', ANL (USA) (1980).

 

12.          W.L Woodruff, A kinetics and thermal-hydraulics capability for the analysis of research reactors, Nuclear Technology, 64, 196–206, (1984).

 

13.          IAEA, Safety analysis for research reactors, Safety Reports Series No. 55, IAEA, Vienna (2008).

Keywords

Research reactors
Thermo-hydraulic
Transient behaviors
Safety parameters
1.             AEOI, Safety Analysis Report for the Tehran Research Reactor, Tehran-Iran, (2018).
 
2.             A. Lashkari, Experimental and numerical analyses of control rods effect on Tehran Research Reactor power peaking factor, Iranian Physic Conference, Mashhad university, (1394) (in Persian).
 
3.             A. Lashkari et al. Neutronic analysis for Tehran research reactor mixed-core, Pnucene, 60, 31–37 (2012).
 
4.             M. Torabi, et al. Neutronic analysis of control rod effect on safety parameter s in Tehran Research Reactor, Nuclear Engineering and Technology, 50, (2018), doi.org/10.1016/j.net.2018.05.008.
 
5.             INVAP, MTR_PC V. 3.0. Neutronic, Thermal-Hydraulic and Shielding Calculations on Personal Computers. Nuclear Engineering Division, INVAP, Bariloche. (2006)..
 
6.             C.J. Taubman, J.H.L., WIMS-D4 Code, A Genera description of the Lattice Code WIMS, United Kingdom Atomic Energy Authorit.
 
7.             E. Villarino, CITVAP v3. 1 Improved version of CITATION II. MTR_PC v2, (2001).
 
8.             INVAP, P.A. Termic V4.1, A Program for the thermal-Hydraulic Analysis of A MTR Core, In Forced Convection. (2003).
 
9.             E., P.A.-I.S., CAUVAP v3.60, A computer program for flow distribution and pressure drop calculation in a MTR type core. (2003).
 
10.          A.E. Bergles, W.M. Rohsenow, Forced-Convection Surface Boiling Heat Transfer and Burnout in Tubes of Small Diameter, DSR Report No. 8767-21, Department of Mechanical Engineering, Massachusetts Institute of Technology, Cambridge, Massachusetts, (1962).
 
11.          IAEA-TECDOC-233, Research Reactor Core Conversion from the use of Highly Enriched Uranium to the use of Low Enriched Uranium, Appendix A 'U.S. Generic Enrichment Reduction calculations for plate type and rod type reactors', ANL (USA) (1980).
 
12.          W.L Woodruff, A kinetics and thermal-hydraulics capability for the analysis of research reactors, Nuclear Technology, 64, 196–206, (1984).
 
13.          IAEA, Safety analysis for research reactors, Safety Reports Series No. 55, IAEA, Vienna (2008).

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