- Email: [email protected]
MCNP5 modeling of the IPR-R1 TRIGA reactor for criticality calculation and reactivity determination
Nuclear Engineering and Design 241 (2011) 4989–4993
Contents lists available at SciVerse ScienceDirect
Nuclear Engineering and Design journal homepage: www.elsevier.com/locate/nucengdes
MCNP5 modeling of the IPR-R1 TRIGA reactor for criticality calculation and reactivity determination Clarysson A.M. da Silva a , Claubia Pereira a,∗ , Bruno T. Guerra a , Maria Auxiliadora F. Veloso a , Antonella L. Costa a , Hugo M. Dalle b , Maria Ângela de B.C. Menezes b a Departamento de Engenharia Nuclear – Escola de Engenharia, Universidade Federal de Minas Gerais, Av. Presidente Antônio Carlos, 6627, 31270-901 Campus Pampulha – Belo Horizonte, Brazil b Centro de Desenvolvimento da Tecnologia Nuclear, Comissão Nacional de Energia Nuclear, Campus da UFMG – Av. Presidente Antônio Carlos, 6627, 31270-901, P.O. Box: 941, Belo Horizonte, MG, Brazil
a r t i c l e
i n f o
Article history: Received 25 May 2011 Received in revised form 27 August 2011 Accepted 12 September 2011
a b s t r a c t The IPR-R1 TRIGA is a research nuclear reactor managed and located at the Nuclear Technology Development Center (CDTN) a research institute of the Brazilian Nuclear Energy Commission (CNEN). It is mainly used to radioisotopes production, scientific experiments, training of nuclear engineers for research and nuclear power plant reactor operation, experiments with materials and minerals and neutron activation analysis. In this work, criticality calculation and reactivity changes are presented and discussed using two modelings of the IPR-R1 TRIGA in the MCNP5 code. The first model (Model 1) analyzes the criticality over the reactor. On the other hand, the second model (Model 2) includes the possibility of radial and axial neutron flux evaluation with different operation conditions. The calculated results are compared with experimental data in different situations. For the two models, the standard deviation and relative error presented values of around 4.9 × 10−4 . Both models present good agreement with respect to the experimental data. The goal is to validate the models that could be used to determine the neutron flux profiles to optimize the irradiation conditions, as well as to study reactivity insertion experiments and also to evaluate the fuel composition. © 2011 Elsevier B.V. All rights reserved.
1. Introduction The IPR-R1 operates since November 1960 and it has been used for research in many fields such as radiochemistry, reactor physics, material science, environmental studies, medical and industrial radioisotopes production and staff training in the “Angra dos Reis” nuclear power plant (e.g. Oliveira et al., 2007; Mesquita et al., 2007; Zangirolami et al., 2009; Guimaraes, 1985; Mesquita and Rezende, 2006; Dalle et al., 2002; Dalle, 2005; Zangirolami, 2009; Reis et al., 2010). In addition to experimental works, some theoretical studies have been developed mainly for analyses of neutronic parameters, thermal hydraulic and sample irradiation. In these theoretical studies, some nuclear codes as WIMSD-5B, CITATION, MCNP4-B, MONTEBURNS, ORIGEN, SCALE, and RELAP5 have been employed (Dalle et al., 2002; Dalle, 2005; Zangirolami, 2009; Reis et al., 2010). Previous works (Dalle et al., 2002; Dalle, 2005), apply the MCNP4
(Monte Carlo N-Particle Transport) to IPR-R1 core modeling to analyze the reactor criticality, the neutron flux, the isotopic concentration, sample irradiation, etc. However, it would be interesting to know also the radial and axial neutron flux distributions at different operation conditions and at different geometric positions in the fuel and in the cooling channels. Moreover, to know the neutronic core conditions with control rods gradually inserted is an important data. In this way, using the MCNP5 code, two models will be evaluated. The first one, developed by Dalle (2005), referenced as Model 1, and the second model referenced as Model 2, that includes more geometry details: 36 axial nodes, superior and inferior grid, rotary rack, central thimble and pneumatic tubes. The goal is to compare the values calculated by the MCNP5 code with the experimental results to partial validation of both models. 2. Methodology 2.1. IPR-R1 general characteristics
∗ Corresponding author. Tel.: +55 31 34096686; fax: +55 31 34096660. E-mail addresses: clarysson [email protected] (C.A.M. da Silva), [email protected] (C. Pereira), [email protected] (B.T. Guerra), [email protected] (M.A.F. Veloso), [email protected] (A.L. Costa), [email protected] (H.M. Dalle), [email protected] (M.Â.d.B.C. Menezes). 0029-5493/$ – see front matter © 2011 Elsevier B.V. All rights reserved. doi:10.1016/j.nucengdes.2011.09.011
The IPR-R1 is a pool type reactor cooled by light water natural circulation and graphite-reflected having maximum power operation of 250 kW. It presents low power, low pressure, for application in research, training and radioisotopes production. The
4990
C.A.M. da Silva et al. / Nuclear Engineering and Design 241 (2011) 4989–4993
Fig. 1. IPR-R1 core representation.
fuel is homogeneous mixture of zirconium hydride and uranium 20% enriched in 235 U isotope. The core has a radial cylindrical configuration with six concentric rings (A, B, C, D, E and F) containing 91 positions able to host either fuel rods or other components like control rods, reflectors and irradiator channels. The core has 59 fuel elements cladding with aluminum, 4 fuel elements cladding with stainless steel, 23 radial reflectors elements, 3 control rods (shim, safety and regulating), 1 central thimble, and 1 neutron source. The fuel elements have three axial sections with upper and lower reflector (graphite), and the central portion filled with fuel (U–ZrH). The radial reflectors elements are covered with aluminum and filled with graphite having the same dimensions of the fuel elements. The control rods are composed by boron carbide and aluminum cladding. In the center of the reactor, there is an aluminum tube (central thimble) to irradiation of experimental samples. This tube is removable and when it is not in use, the reactor pool water fills its volume. Furthermore, the core has an annular graphite reflector with aluminum cladding. Such annular reflector has a radial groove where a rotary rack is assembled for insertion of the samples to irradiation. In such rotary rack is possible to place the samples in 40 different positions around the core. Moreover, tangent to annular reflector, there is a pneumatic tube where the samples also can be inserted to irradiation. Therefore, the IPR-R1 has three facilities for sample irradiation: the central tube, the rotary rack and the pneumatic tube. Fig. 1 shows the radial and axial core configuration and Fig. 2 illustrates the design of two fuel element types.
The core was configured in the MCNP using a cylinder that contains water, fuel elements, radial reflectors, central tube, control rods and neutron source. Each rod has a coordinate value. In the modeling, the rods were filled according to their individual characteristics. Around the core is the rotary rack with adequate groove to insert the samples to irradiation. The reactor was configured inside the pool where water surrounds the core and the rotary rack. Both models have the same dimensions and the same materials composition. However, the Model 2 presents more detailed geometry in comparison with the Model 1. The main improvements are: • in the Model 2 there are 36 nodes in the active length of fuel element which can be filled by different isotopic composition material; • in the Model 2, the control rods (shim, safety and regulating) can be gradually inserted in the core; • in the Model 2 the superior grid, inferior grid and pneumatic tube were configured; and • in the Model 2, the samples can be inserted in the rotary rack, or in the central thimble, or in the pneumatic tubes to irradiation. Using the Model 2, it will be possible to perform several studies such as: sample irradiations in pneumatic tubes or central thimble at different axial positions, neutronic parameters evaluations for gradual insertion of control rods, axial isotopic compositions analysis, and many others. In both models, the fuel composition was obtained from the reference (Dalle, 2005).
2.2. Modelings 2.3. Studied cases and evaluated neutronic parameters The reactor was modeled using the MCNP code. Figs. 3 and 4 illustrate the axial and radial view of the two simulated modelings, Model 1 and Model 2, respectively. These models were configured according with the geometry and the characteristics of IPR-R1 core described above.
As mentioned before, there are three-control rods types in the reactor core: regulating rod, safety rod and shim rod. Each control rod has a specific position in according with its function (see Fig. 1). The regulating rod provides fine reactivity control; the shim rod
C.A.M. da Silva et al. / Nuclear Engineering and Design 241 (2011) 4989–4993
4991
Fig. 2. Fuel elements design of the IPR-R1.
provides crude reactivity control; and the safety rod is used for reactor shutdown. In this way, the insertion of each control rod type generates different contributions to the core reactivity. Therefore, in this work, each control rod was individually inserted into the core and the reactivity effects were compared with experimental data obtained from reference (Souza et al., 2001). The following core conditions were evaluated:
code and and were calculated using, respectively, the following equations: =
keff − 1
= 2 − 1 =
(a) (b) (c) (d)
all control rods withdrawn; only the regulating rod totally inserted into the core; only the safety rod totally inserted into the core; and only the shim rod totally inserted into the core.
(1)
keff 1 1 − keff 1 keff 2
(2)
where the keff1 and the keff2 are the effective multiplication factors to control rods withdrawn and inserted into the reactor core, respectively.
3. Results The following three neutronic parameters were evaluated: effective multiplication factor (keff ), reactivity (), and reactivity worth of the control rods (). The keff was estimated by MCNP5
Table 1 shows the standard deviation ( ST ) and relative error R to the keff calculated by MCNP5 nuclear code. The standard deviation ( ST ) is estimated by the code that is printed in the output file with
Fig. 3. Axial configuration of the simulated models.
4992
C.A.M. da Silva et al. / Nuclear Engineering and Design 241 (2011) 4989–4993
Fig. 4. Radial configuration of the simulated models.
the respective keff . The relative error R was calculated using the following equation: R=
ST keff
material, this can explain the keff overestimated in the Model 1 presented in Table 2. Table 3 presents the keff and when the control rods are totally inserted in the core. It shows the values of these parameters when the regulating, or safety, or shim rod is inserted in the core individually. When a control rod is inserted in the core, there are obviously variations in the multiplication factor. Table 4 presents the keff variations (keff ) due to the insertion of control rods, calculated as:
(3)
According to the reference (Booth, 2003), the values are reliable when the relative error R < 0.05. Table 1 shows that the relative error is around 4.9 × 10−4 and therefore the calculated keff values are absolutely acceptable. Table 2 presents the keff and when all the three control rods are withdrawn. In addition, this table shows the differences between the calculated values and the experimental data. The difference was calculated as: Difference = |Calculated Value − Measured Value|
keff = keff 1 − keff 2
(5)
where keff1 is the effective multiplication factor to all control rods withdrawn and keff2 is the effective multiplication factor due only to one control rod (regulating, or safety, or shim) inserted into the core. Table 3 shows that the insertion of regulating rod generates the smallest keff while the insertion of shim rod induces the biggest keff as it was expected. In this way, the regulating rod has the smallest reactivity worth () while the shim rod has the biggest (see Tables 3 and 4). In the other hand, for the safety and shim rods, the keff and ( values are similar. The difference between the behavior of such rods is due to their positions in the core. The regulating rod is located in the periphery of the core (position F16)
(4)
Results show that the Model 2 presents smaller keff and difference than the Model 1 in relation with the experimental data. For Model 1 this difference is 0.35% while to Model 2 it is 0.03%. Although the models have the same dimensions, the Model 2 has more detailed geometry which presents the superior and inferior grid plates. In the Model 1, the volume of the top ad the bottom of the grid plates are filled with water. Therefore, the Model 1 presents water volume bigger than Model 2. As the water is a moderator Table 1 Standard deviation ( ST ) and relative error R to keff calculated by MCNP5 code. Inserted CR
No rods Regulating Safety Shim
Model 1 (M1)
Model 2 (M2) −4
keff
ST (×10
1.02215 1.01810 0.99724 0.99609
4.800 5.000 4.900 4.800
)
−4
R (×10 4.696 4.911 4.914 4.819
)
keff
ST (×10−4 )
R (×10−4 )
1.01891 1.01523 0.99832 0.99685
4.800 4.800 4.900 5.100
4.711 4.728 4.908 5.116
C.A.M. da Silva et al. / Nuclear Engineering and Design 241 (2011) 4989–4993
4993
Table 2 Effective multiplication factor (keff ) and reactivity () considering all control rods withdrawn. Parameter
Calculation Model 1
Calculation Model 2
Experimental data (Dalle, 2005)
Difference Model 1
Difference Model 2
keff (pcm)
1.02215 2.16700 × 103
1.01891 1.85590 × 103
1.01859 1.82500 × 103
3.560 × 10−3 3.420 × 102
3.200 × 10−4 3.090 × 101
Table 3 The keff and reactivity worth () of the control rods. Parameter
Inserted CR
keff
Regulating Safety Shim
(pcm)
Regulating Safety Shim
Calculation Model 1 1.01810 0.99724 0.99609 −3.892 × 102 −2.444 × 103 −2.560 × 103
Calculation Model 2 1.01523 0.99832 0.99685 −3.568 × 102 −2.024 × 103 −2.172 × 103
Experimental data (Booth, 2003) 1.01434 0.99615 0.99418 −4.110 × 102 −2.212 × 103 −2.410 × 103
Difference Model 1 −3
Difference Model 2
3.756 × 10 1.905 × 10−3 1.094 × 10−3
8.865 × 10−4 2.665 × 10−3 2.174 × 10−3
2.182 × 101 1.495 × 102 2.318 × 102
5.525 × 101 2.381 × 102 1.878 × 102
Table 4 Effective multiplication factor variations (keff ) and reactivity worth of the control rods (). Parameter
Inserted CR
keff1
keff2
keff (×10−2 )
Calculation Model 1
Regulating Safety Shim
1.02215 1.02215 1.02215
1.018100 0.997240 0.996090
0.40500 2.04910 2.60600
−0.389 −2.444 −2.560
Calculation Model 2
Regulating Safety Shim
1.01891 1.01891 1.01891
1.015230 0.998320 0.996850
0.36800 2.05900 2.20600
−0.357 −2.024 −2.172
Experimental data (Dalle, 2005)
Regulating Safety Shim
1.01859 1.01859 1.01859
1.014340 0.996150 0.994180
0.42500 2.24400 2.44100
−0.411 −2.212 −2.410
while the safety and shim rods are positioned near to the core center (positions C1 and C7) as it can be verified in Fig. 1. 4. Conclusions In this work, two models of the IPR-R1 TRIGA research reactor in the MCNP5, were used to obtain values of keff in some different situations of the reactor operation. The results were compared with experimental available data of keff . In comparison with the experimental data, the Model 2 shows smaller keff and difference than for the Model 1. Moreover, when regulating rod is inserted in the core, the difference of the Model 2 is smaller than Model 1. However, when safety or shim rod is inserted, the difference for the Model 1 is smaller than for the Model 2. Nevertheless, the differences between the two models are small and therefore acceptable. In this way, both models present simulation of the neutronic parameters with very good agreement between them and also with respect to experimental data. The Model 2 presents more detailed geometry in comparison with the Model 1.This fact allows Model 2 to perform more number of simulations with refined variations. Therefore, the Model 2 will be applied in future works. The models were validated to criticality calculation and reactivity changes. The next step is to validate the models using experimental results obtained with a instrumented fuel element. After the models validation, they could be used to determine the neutron flux profiles and consequently to optimize the irradiation conditions. Moreover, the models will be very useful to study reactivity insertion experiments and to evaluate the fuel composition.
(pcm) (×103 )
Acknowledgements The authors are grateful to CAPES (Brazil), FAPEMIG (MG/Brazil), CNPq (Brazil), and CDTN/CNEN (Brazil) for the support. References Booth, T.E., et al., 2003. MCNP-A General Monte Carlo N-Particle Transport Code (Version 5). Los Alamos National Laboratory, Report LA-UR-03-1987. Dalle, H.M., 2005. Simulac¸ão do Reator IPR-R1Utilizando Métodos de Transporte por Monte Carlo, Doctoral Thesis, Faculdade de Engenharia Química, Universidade Estadual de Campinas, Campinas, Brazil. Dalle, H.M., Pereira, C., Souza, R.G.P., 2002. Neutronic calculation to the TRIGA IPR-R1 Reactor using the WIMSD4 and CITATION codes. Annals of Nuclear Energy 29, 901–912. Guimarães, R.R.R., 1985. Levantamento das Distribuic¸ões dos Fluxos de Nêutrons Térmicos e Rápidos no Núcleo do Reator IPR-R1, Master Thesis, Escola de Engenharia, Universidade Federal de Minas Gerais, Belo Horizonte, Brazil. Mesquita, A.Z., Rezende, H.C., 2006. Experimental heat transfer analysis of the IPRR1TRIGA Reactor. In: 3rd World TRIGA Users Conference, Belo Horizonte, Brazil, August. Mesquita, A.Z., Rezende, H.C., Tambourgi, E.B., 2007. Power calibration of the TRIGA Mark I Nuclear Research Reactor. In: 17th International Congress of Mechanical Engineering, Sao Paulo, Brazil, July–September, vols. 19 n 3, pp. 240–245. Oliveira, P.F., et al., 2007. Measuring the dose rate at the core and tank of the CDTN IPR-R1 TRIGA Mark I Reactor. In: 2007 International Nuclear Atlantic Conference, Santos, Brazil, September–October. Reis, P.A.L., et al., 2010. Assessment of a RELAP5 Model for the IPR-R1 TRIGA Research Reactor. Annals of Nuclear Energy 37, 1341–1350. Souza, R.M.G.P., Resende, M.F.R., Mesquita, A.Z., Valente, E.S., Wakabayashi, T., 2001. Resultados dos Testes Iniciais para o Aumento de Potência do Reator TRIGA IPR–R1, NI-IT4-01/01. CDTN, Belo Horizonte. Zangirolami, D.M., 2009. Fluxo Neutrônico a 100 kW nos Terminais de Irradiac¸ão do Reator TRIGA IPR-R1, Master Thesis, Escola de Engenharia, Universidade Federal de Minas Gerais, Belo Horizonte, Brazil. Zangirolami, D.M., Ferreira, A.V., Oliveira, A.H., 2009. Specific induced activity profile at the rotary specimen rack of IPR-R1 TRIGA Reactor. Brazilian Journal of Physics 39, 260–263.
Contents lists available at SciVerse ScienceDirect
Nuclear Engineering and Design journal homepage: www.elsevier.com/locate/nucengdes
MCNP5 modeling of the IPR-R1 TRIGA reactor for criticality calculation and reactivity determination Clarysson A.M. da Silva a , Claubia Pereira a,∗ , Bruno T. Guerra a , Maria Auxiliadora F. Veloso a , Antonella L. Costa a , Hugo M. Dalle b , Maria Ângela de B.C. Menezes b a Departamento de Engenharia Nuclear – Escola de Engenharia, Universidade Federal de Minas Gerais, Av. Presidente Antônio Carlos, 6627, 31270-901 Campus Pampulha – Belo Horizonte, Brazil b Centro de Desenvolvimento da Tecnologia Nuclear, Comissão Nacional de Energia Nuclear, Campus da UFMG – Av. Presidente Antônio Carlos, 6627, 31270-901, P.O. Box: 941, Belo Horizonte, MG, Brazil
a r t i c l e
i n f o
Article history: Received 25 May 2011 Received in revised form 27 August 2011 Accepted 12 September 2011
a b s t r a c t The IPR-R1 TRIGA is a research nuclear reactor managed and located at the Nuclear Technology Development Center (CDTN) a research institute of the Brazilian Nuclear Energy Commission (CNEN). It is mainly used to radioisotopes production, scientific experiments, training of nuclear engineers for research and nuclear power plant reactor operation, experiments with materials and minerals and neutron activation analysis. In this work, criticality calculation and reactivity changes are presented and discussed using two modelings of the IPR-R1 TRIGA in the MCNP5 code. The first model (Model 1) analyzes the criticality over the reactor. On the other hand, the second model (Model 2) includes the possibility of radial and axial neutron flux evaluation with different operation conditions. The calculated results are compared with experimental data in different situations. For the two models, the standard deviation and relative error presented values of around 4.9 × 10−4 . Both models present good agreement with respect to the experimental data. The goal is to validate the models that could be used to determine the neutron flux profiles to optimize the irradiation conditions, as well as to study reactivity insertion experiments and also to evaluate the fuel composition. © 2011 Elsevier B.V. All rights reserved.
1. Introduction The IPR-R1 operates since November 1960 and it has been used for research in many fields such as radiochemistry, reactor physics, material science, environmental studies, medical and industrial radioisotopes production and staff training in the “Angra dos Reis” nuclear power plant (e.g. Oliveira et al., 2007; Mesquita et al., 2007; Zangirolami et al., 2009; Guimaraes, 1985; Mesquita and Rezende, 2006; Dalle et al., 2002; Dalle, 2005; Zangirolami, 2009; Reis et al., 2010). In addition to experimental works, some theoretical studies have been developed mainly for analyses of neutronic parameters, thermal hydraulic and sample irradiation. In these theoretical studies, some nuclear codes as WIMSD-5B, CITATION, MCNP4-B, MONTEBURNS, ORIGEN, SCALE, and RELAP5 have been employed (Dalle et al., 2002; Dalle, 2005; Zangirolami, 2009; Reis et al., 2010). Previous works (Dalle et al., 2002; Dalle, 2005), apply the MCNP4
(Monte Carlo N-Particle Transport) to IPR-R1 core modeling to analyze the reactor criticality, the neutron flux, the isotopic concentration, sample irradiation, etc. However, it would be interesting to know also the radial and axial neutron flux distributions at different operation conditions and at different geometric positions in the fuel and in the cooling channels. Moreover, to know the neutronic core conditions with control rods gradually inserted is an important data. In this way, using the MCNP5 code, two models will be evaluated. The first one, developed by Dalle (2005), referenced as Model 1, and the second model referenced as Model 2, that includes more geometry details: 36 axial nodes, superior and inferior grid, rotary rack, central thimble and pneumatic tubes. The goal is to compare the values calculated by the MCNP5 code with the experimental results to partial validation of both models. 2. Methodology 2.1. IPR-R1 general characteristics
∗ Corresponding author. Tel.: +55 31 34096686; fax: +55 31 34096660. E-mail addresses: clarysson [email protected] (C.A.M. da Silva), [email protected] (C. Pereira), [email protected] (B.T. Guerra), [email protected] (M.A.F. Veloso), [email protected] (A.L. Costa), [email protected] (H.M. Dalle), [email protected] (M.Â.d.B.C. Menezes). 0029-5493/$ – see front matter © 2011 Elsevier B.V. All rights reserved. doi:10.1016/j.nucengdes.2011.09.011
The IPR-R1 is a pool type reactor cooled by light water natural circulation and graphite-reflected having maximum power operation of 250 kW. It presents low power, low pressure, for application in research, training and radioisotopes production. The
4990
C.A.M. da Silva et al. / Nuclear Engineering and Design 241 (2011) 4989–4993
Fig. 1. IPR-R1 core representation.
fuel is homogeneous mixture of zirconium hydride and uranium 20% enriched in 235 U isotope. The core has a radial cylindrical configuration with six concentric rings (A, B, C, D, E and F) containing 91 positions able to host either fuel rods or other components like control rods, reflectors and irradiator channels. The core has 59 fuel elements cladding with aluminum, 4 fuel elements cladding with stainless steel, 23 radial reflectors elements, 3 control rods (shim, safety and regulating), 1 central thimble, and 1 neutron source. The fuel elements have three axial sections with upper and lower reflector (graphite), and the central portion filled with fuel (U–ZrH). The radial reflectors elements are covered with aluminum and filled with graphite having the same dimensions of the fuel elements. The control rods are composed by boron carbide and aluminum cladding. In the center of the reactor, there is an aluminum tube (central thimble) to irradiation of experimental samples. This tube is removable and when it is not in use, the reactor pool water fills its volume. Furthermore, the core has an annular graphite reflector with aluminum cladding. Such annular reflector has a radial groove where a rotary rack is assembled for insertion of the samples to irradiation. In such rotary rack is possible to place the samples in 40 different positions around the core. Moreover, tangent to annular reflector, there is a pneumatic tube where the samples also can be inserted to irradiation. Therefore, the IPR-R1 has three facilities for sample irradiation: the central tube, the rotary rack and the pneumatic tube. Fig. 1 shows the radial and axial core configuration and Fig. 2 illustrates the design of two fuel element types.
The core was configured in the MCNP using a cylinder that contains water, fuel elements, radial reflectors, central tube, control rods and neutron source. Each rod has a coordinate value. In the modeling, the rods were filled according to their individual characteristics. Around the core is the rotary rack with adequate groove to insert the samples to irradiation. The reactor was configured inside the pool where water surrounds the core and the rotary rack. Both models have the same dimensions and the same materials composition. However, the Model 2 presents more detailed geometry in comparison with the Model 1. The main improvements are: • in the Model 2 there are 36 nodes in the active length of fuel element which can be filled by different isotopic composition material; • in the Model 2, the control rods (shim, safety and regulating) can be gradually inserted in the core; • in the Model 2 the superior grid, inferior grid and pneumatic tube were configured; and • in the Model 2, the samples can be inserted in the rotary rack, or in the central thimble, or in the pneumatic tubes to irradiation. Using the Model 2, it will be possible to perform several studies such as: sample irradiations in pneumatic tubes or central thimble at different axial positions, neutronic parameters evaluations for gradual insertion of control rods, axial isotopic compositions analysis, and many others. In both models, the fuel composition was obtained from the reference (Dalle, 2005).
2.2. Modelings 2.3. Studied cases and evaluated neutronic parameters The reactor was modeled using the MCNP code. Figs. 3 and 4 illustrate the axial and radial view of the two simulated modelings, Model 1 and Model 2, respectively. These models were configured according with the geometry and the characteristics of IPR-R1 core described above.
As mentioned before, there are three-control rods types in the reactor core: regulating rod, safety rod and shim rod. Each control rod has a specific position in according with its function (see Fig. 1). The regulating rod provides fine reactivity control; the shim rod
C.A.M. da Silva et al. / Nuclear Engineering and Design 241 (2011) 4989–4993
4991
Fig. 2. Fuel elements design of the IPR-R1.
provides crude reactivity control; and the safety rod is used for reactor shutdown. In this way, the insertion of each control rod type generates different contributions to the core reactivity. Therefore, in this work, each control rod was individually inserted into the core and the reactivity effects were compared with experimental data obtained from reference (Souza et al., 2001). The following core conditions were evaluated:
code and and were calculated using, respectively, the following equations: =
keff − 1
= 2 − 1 =
(a) (b) (c) (d)
all control rods withdrawn; only the regulating rod totally inserted into the core; only the safety rod totally inserted into the core; and only the shim rod totally inserted into the core.
(1)
keff 1 1 − keff 1 keff 2
(2)
where the keff1 and the keff2 are the effective multiplication factors to control rods withdrawn and inserted into the reactor core, respectively.
3. Results The following three neutronic parameters were evaluated: effective multiplication factor (keff ), reactivity (), and reactivity worth of the control rods (). The keff was estimated by MCNP5
Table 1 shows the standard deviation ( ST ) and relative error R to the keff calculated by MCNP5 nuclear code. The standard deviation ( ST ) is estimated by the code that is printed in the output file with
Fig. 3. Axial configuration of the simulated models.
4992
C.A.M. da Silva et al. / Nuclear Engineering and Design 241 (2011) 4989–4993
Fig. 4. Radial configuration of the simulated models.
the respective keff . The relative error R was calculated using the following equation: R=
ST keff
material, this can explain the keff overestimated in the Model 1 presented in Table 2. Table 3 presents the keff and when the control rods are totally inserted in the core. It shows the values of these parameters when the regulating, or safety, or shim rod is inserted in the core individually. When a control rod is inserted in the core, there are obviously variations in the multiplication factor. Table 4 presents the keff variations (keff ) due to the insertion of control rods, calculated as:
(3)
According to the reference (Booth, 2003), the values are reliable when the relative error R < 0.05. Table 1 shows that the relative error is around 4.9 × 10−4 and therefore the calculated keff values are absolutely acceptable. Table 2 presents the keff and when all the three control rods are withdrawn. In addition, this table shows the differences between the calculated values and the experimental data. The difference was calculated as: Difference = |Calculated Value − Measured Value|
keff = keff 1 − keff 2
(5)
where keff1 is the effective multiplication factor to all control rods withdrawn and keff2 is the effective multiplication factor due only to one control rod (regulating, or safety, or shim) inserted into the core. Table 3 shows that the insertion of regulating rod generates the smallest keff while the insertion of shim rod induces the biggest keff as it was expected. In this way, the regulating rod has the smallest reactivity worth () while the shim rod has the biggest (see Tables 3 and 4). In the other hand, for the safety and shim rods, the keff and ( values are similar. The difference between the behavior of such rods is due to their positions in the core. The regulating rod is located in the periphery of the core (position F16)
(4)
Results show that the Model 2 presents smaller keff and difference than the Model 1 in relation with the experimental data. For Model 1 this difference is 0.35% while to Model 2 it is 0.03%. Although the models have the same dimensions, the Model 2 has more detailed geometry which presents the superior and inferior grid plates. In the Model 1, the volume of the top ad the bottom of the grid plates are filled with water. Therefore, the Model 1 presents water volume bigger than Model 2. As the water is a moderator Table 1 Standard deviation ( ST ) and relative error R to keff calculated by MCNP5 code. Inserted CR
No rods Regulating Safety Shim
Model 1 (M1)
Model 2 (M2) −4
keff
ST (×10
1.02215 1.01810 0.99724 0.99609
4.800 5.000 4.900 4.800
)
−4
R (×10 4.696 4.911 4.914 4.819
)
keff
ST (×10−4 )
R (×10−4 )
1.01891 1.01523 0.99832 0.99685
4.800 4.800 4.900 5.100
4.711 4.728 4.908 5.116
C.A.M. da Silva et al. / Nuclear Engineering and Design 241 (2011) 4989–4993
4993
Table 2 Effective multiplication factor (keff ) and reactivity () considering all control rods withdrawn. Parameter
Calculation Model 1
Calculation Model 2
Experimental data (Dalle, 2005)
Difference Model 1
Difference Model 2
keff (pcm)
1.02215 2.16700 × 103
1.01891 1.85590 × 103
1.01859 1.82500 × 103
3.560 × 10−3 3.420 × 102
3.200 × 10−4 3.090 × 101
Table 3 The keff and reactivity worth () of the control rods. Parameter
Inserted CR
keff
Regulating Safety Shim
(pcm)
Regulating Safety Shim
Calculation Model 1 1.01810 0.99724 0.99609 −3.892 × 102 −2.444 × 103 −2.560 × 103
Calculation Model 2 1.01523 0.99832 0.99685 −3.568 × 102 −2.024 × 103 −2.172 × 103
Experimental data (Booth, 2003) 1.01434 0.99615 0.99418 −4.110 × 102 −2.212 × 103 −2.410 × 103
Difference Model 1 −3
Difference Model 2
3.756 × 10 1.905 × 10−3 1.094 × 10−3
8.865 × 10−4 2.665 × 10−3 2.174 × 10−3
2.182 × 101 1.495 × 102 2.318 × 102
5.525 × 101 2.381 × 102 1.878 × 102
Table 4 Effective multiplication factor variations (keff ) and reactivity worth of the control rods (). Parameter
Inserted CR
keff1
keff2
keff (×10−2 )
Calculation Model 1
Regulating Safety Shim
1.02215 1.02215 1.02215
1.018100 0.997240 0.996090
0.40500 2.04910 2.60600
−0.389 −2.444 −2.560
Calculation Model 2
Regulating Safety Shim
1.01891 1.01891 1.01891
1.015230 0.998320 0.996850
0.36800 2.05900 2.20600
−0.357 −2.024 −2.172
Experimental data (Dalle, 2005)
Regulating Safety Shim
1.01859 1.01859 1.01859
1.014340 0.996150 0.994180
0.42500 2.24400 2.44100
−0.411 −2.212 −2.410
while the safety and shim rods are positioned near to the core center (positions C1 and C7) as it can be verified in Fig. 1. 4. Conclusions In this work, two models of the IPR-R1 TRIGA research reactor in the MCNP5, were used to obtain values of keff in some different situations of the reactor operation. The results were compared with experimental available data of keff . In comparison with the experimental data, the Model 2 shows smaller keff and difference than for the Model 1. Moreover, when regulating rod is inserted in the core, the difference of the Model 2 is smaller than Model 1. However, when safety or shim rod is inserted, the difference for the Model 1 is smaller than for the Model 2. Nevertheless, the differences between the two models are small and therefore acceptable. In this way, both models present simulation of the neutronic parameters with very good agreement between them and also with respect to experimental data. The Model 2 presents more detailed geometry in comparison with the Model 1.This fact allows Model 2 to perform more number of simulations with refined variations. Therefore, the Model 2 will be applied in future works. The models were validated to criticality calculation and reactivity changes. The next step is to validate the models using experimental results obtained with a instrumented fuel element. After the models validation, they could be used to determine the neutron flux profiles and consequently to optimize the irradiation conditions. Moreover, the models will be very useful to study reactivity insertion experiments and to evaluate the fuel composition.
(pcm) (×103 )
Acknowledgements The authors are grateful to CAPES (Brazil), FAPEMIG (MG/Brazil), CNPq (Brazil), and CDTN/CNEN (Brazil) for the support. References Booth, T.E., et al., 2003. MCNP-A General Monte Carlo N-Particle Transport Code (Version 5). Los Alamos National Laboratory, Report LA-UR-03-1987. Dalle, H.M., 2005. Simulac¸ão do Reator IPR-R1Utilizando Métodos de Transporte por Monte Carlo, Doctoral Thesis, Faculdade de Engenharia Química, Universidade Estadual de Campinas, Campinas, Brazil. Dalle, H.M., Pereira, C., Souza, R.G.P., 2002. Neutronic calculation to the TRIGA IPR-R1 Reactor using the WIMSD4 and CITATION codes. Annals of Nuclear Energy 29, 901–912. Guimarães, R.R.R., 1985. Levantamento das Distribuic¸ões dos Fluxos de Nêutrons Térmicos e Rápidos no Núcleo do Reator IPR-R1, Master Thesis, Escola de Engenharia, Universidade Federal de Minas Gerais, Belo Horizonte, Brazil. Mesquita, A.Z., Rezende, H.C., 2006. Experimental heat transfer analysis of the IPRR1TRIGA Reactor. In: 3rd World TRIGA Users Conference, Belo Horizonte, Brazil, August. Mesquita, A.Z., Rezende, H.C., Tambourgi, E.B., 2007. Power calibration of the TRIGA Mark I Nuclear Research Reactor. In: 17th International Congress of Mechanical Engineering, Sao Paulo, Brazil, July–September, vols. 19 n 3, pp. 240–245. Oliveira, P.F., et al., 2007. Measuring the dose rate at the core and tank of the CDTN IPR-R1 TRIGA Mark I Reactor. In: 2007 International Nuclear Atlantic Conference, Santos, Brazil, September–October. Reis, P.A.L., et al., 2010. Assessment of a RELAP5 Model for the IPR-R1 TRIGA Research Reactor. Annals of Nuclear Energy 37, 1341–1350. Souza, R.M.G.P., Resende, M.F.R., Mesquita, A.Z., Valente, E.S., Wakabayashi, T., 2001. Resultados dos Testes Iniciais para o Aumento de Potência do Reator TRIGA IPR–R1, NI-IT4-01/01. CDTN, Belo Horizonte. Zangirolami, D.M., 2009. Fluxo Neutrônico a 100 kW nos Terminais de Irradiac¸ão do Reator TRIGA IPR-R1, Master Thesis, Escola de Engenharia, Universidade Federal de Minas Gerais, Belo Horizonte, Brazil. Zangirolami, D.M., Ferreira, A.V., Oliveira, A.H., 2009. Specific induced activity profile at the rotary specimen rack of IPR-R1 TRIGA Reactor. Brazilian Journal of Physics 39, 260–263.