Study of criticality safety and neutronic parameters of UO2 fuel in MTR research reactors using the MCNP4C code

Study of criticality safety and neutronic parameters of UO2 fuel in MTR research reactors using the MCNP4C code

Annals of Nuclear Energy 98 (2016) 144–156 Contents lists available at ScienceDirect Annals of Nuclear Energy journal homepage: www.elsevier.com/loc...

3MB Sizes 5 Downloads 158 Views

Annals of Nuclear Energy 98 (2016) 144–156

Contents lists available at ScienceDirect

Annals of Nuclear Energy journal homepage: www.elsevier.com/locate/anucene

Study of criticality safety and neutronic parameters of UO2 fuel in MTR research reactors using the MCNP4C code Ismail Shaaban ⇑, Mohamad Albarhoum Nuclear Engineering Dept., Atomic Energy Commission, P.O. Box 6091, Damascus, Syria

a r t i c l e

i n f o

Article history: Received 9 March 2016 Received in revised form 24 July 2016 Accepted 29 July 2016 Available online 7 August 2016 Keywords: MTR-22 MW reactor UO2 fuel Criticality safety Neutronic parameters MCNP4C code

a b s t r a c t The low enriched uranium UO2 fuel performance in MTR-22 MW reactor is investigated in this paper with One Neutronic Trap (ONT) and Three Neutronic Traps (TNTs) to increase the scientific utilization of the reactor. The MCNP4C code was used to estimate the criticality safety and the neutronic parameters of the MTR-22 MW before and after replacing the U3O8-Al original fuel by the UO2 one. The re-evaluated criticality safety parameters and the new neutronic characteristics of this reactor showed that the replacement of U3O8-Al Fuel Elements (FEs) by UO2 FEs would not significantly affect the neutronic characteristics of the MTR-22 MW as well as its criticality for both the ONT and the TNTs. This result leads to say that the replacement of U3O8-Al FEs by UO2 FEs will not affect the safety of the reactor. Ó 2016 Elsevier Ltd. All rights reserved.

1. Introduction It is well-known that the UO2 is the standard nuclear fuel used in modern, 2nd generation power reactors such as: the Light Water Reactors (LWRs) and the Advanced Gas-cooled Reactors (AGRs). This is due to the characteristics of the UO2 fuel such as: 1. A high melting point and a high-temperature stability, 2. a good chemical compatibility with the cladding and the coolant, and 3. its resistance to radiation. One assured disadvantage of the UO2 is a low thermal conductivity, which can cause large temperature gradients in UO2 fuel pellets. These temperature gradients produce thermal stresses that can lead to extensive fuel pellet cracking, an increased rate of fission gas release, more fission gas bubbles and fuel pellet swelling issues that limit UO2 fuel’s life within reactors (IAEA, 2005; Carbajo et al., 2001; Simnad, 1962; Yeo, 2013) on one hand. On the other hand, in research reactors many types of fuel were used such as: U-A1 alloy and dispersion-type including UA1x-A1, U308-A1, U3Si2-Al,U3Si-Al and U-ZrHx. Fuels in MNSRs (Miniature Neutron Source Reactors), MTRs (Material Testing Reactors) and TRIGA (Training, Research, Isotopes, General Atomics) reactors are enriched in 235U to about 90% in MNSRs and 20% in MTRs ⇑ Corresponding author. E-mail address: [email protected] (I. Shaaban). http://dx.doi.org/10.1016/j.anucene.2016.07.030 0306-4549/Ó 2016 Elsevier Ltd. All rights reserved.

and TRIGA reactors (IAEA, 1992; SAR, 1993; Bretscher and Matos, 1996). As mentioned above, the UO2 fuel has good characteristics to become one of the most promising new types of reduced enrichment fuels (less than 10%) for use in research reactors with very high power density. Therefore, many researches were performed in different countries to convert the reactor core from High Enrichment Uranium (HEU) to Low Enrichment Uranium (LEU) core using this fuel. In these researches, the UO2 fuel was proposed as fuel to convert MNSRs from HEU (90% 235U) to LEU with 235U enrichment less than 13.5% as in the French OSIRIS and the Shippingport reactors. In addition, ‘‘Benchmarks” calculations have been performed by different laboratories for two MTR-type reactors with power levels of 2 MW and 10 MW, respectively for conversion from HEU to LEU using UO2 as fuel (Albarhoum, 2011; Odoi et al., 2014; Hsieh et al., 1980; IAEA, 1980). The use of UO2 fuel in research reactors may contribute to: 1. The development of research reactors designs which adopt the UO2 fuel. 2. Reduce the economic cost to produce and re-process the nuclear fuel, since this fuel is produced in the fuel manufacturing facilities for power reactors. 3. Reduce the forms of the radioactive materials in the environment. Therefore, this paper will discuss the use of the UO2 fuel with Zircaloy-4 clad in a MTR-22 MW reactor with ONT and TNTs (the

145

I. Shaaban, M. Albarhoum / Annals of Nuclear Energy 98 (2016) 144–156

Egyptian Second Research Reactor (ETRR-2) as example) without changing the dimensions or the components or even the safe operational conditions of the reactor core. The MCNP4C code (Briesmeister, 2000) was used to track the effect of core modification on the criticality safety and on the neutronic parameters of the ETRR-2 reactor before and after the replacement of the U3O8-Al original fuel by the UO2 fuel.

2. Methodology 2.1. The ETRR-2 reactor The ETRR-2 is a Material Testing Reactor (MTR). It was commissioned in 1997. It is an open pool research reactor which uses low enriched MTR fuel elements (19.7% enrichment). It is cooled and moderated with light water and reflected by beryllium. The reactor power is 22 MW with high thermal neutron flux irradiation positions (>1014 n/cm2 s). The ETRR-2 core consists of 29 positions for fuel elements, where three distinct types of fuel elements are used in the ETRR-2 which are: 7 Standard Fuel Elements (SFE), 8 Fuel Element Type 1 (FE Type 1) and 14 Fuel Elements Type 2 (FE Type 2). The composition and general characteristics of the Fuel Material (FM), FE, Fuel Plate (FP), absorber material, active zone dimensions and water gap between plates of the ETRR-2 reactor for the U3O8Al fuel are given in Tables 1, 2 and 4. The reactor is controlled by 6 control plates made of Ag-In-Cd alloy (See Table 2) (Imami and Roushdy, 2002; Nagy et al., 2004; Khater et al., 2006; Gaheen, 2010; Hussein et al., 2011).

Table 1 Composition of the FE in a typical ETRR-2 reactor. Parameter

Wright percentage % SFE

FE Type 1

FE Type 2

235

12.377 50.450 25.91 11.263 4.802

6.598 26.894 60.504 6.004 3.299

8.398 34.230 49.730 7.642 3.655

U 238 U 27 Al 16 O Density (g/cm3)

2.2. The Monte Carlo model of the ETRR-2 The ETRR-2 reactor was simulated using the MCNP4C code (Shaaban and Albarhoum, 2015). The cross-section of the ETRR-2 core configuration is shown in Fig. 1. In this model, the ONT was taken and located near the center of the core. In the simulation the following specification of the ETRR-2 were used: 1. 2. 3. 4. 5. 6.

the fuel composition shown in Table 1, U3O8 as fuel material dispersed in an Al matrix, a plate type fuel element, an Al fuel clad, an Al frame for the fuel element with U3O8-Al original fuel, a de-mineralized light water as coolant.

3. Replacement of the U3O8-Al fuel plates by UO2 fuel plates in the ETRR-2 reactor The scope of this study is the replacement of U3O8-Al FPs by UO2 FPs, but with conserving similar operational capabilities, criticality safety conditions and neutronic parameters of the ETRR-2 reactor as the U3O8-Al original fuel and without changing the FE dimensions of the ETRR-2 core. To perform these analyses two methods were used: 1. The dimensions of the UO2 FM and FP in the FE, the active zone dimensions and the water gap between plates were taken as equal to the U3O8-Al original fuel dimensions used in the ETRR-2 reactor (See Fig. 2a, Tables 2 and 4). Where this procedure is applied in the calculations of the 2 MW and 10 MW reactors (IAEA, 1980). 2. The UO2 FM in the FP was divided into eight fuel pieces (See Fig. 2b, Tables 3 and 4) without changing the dimensions of the FP, the clad, the FE, the active zone and the water gap between plates. This procedure is applied in the Shippingport reactor (See Fig. 3) (Hsieh et al., 1980). 3. In both cases, the absorber material in the control plates material is changed from Ag-In-Cd to B4C, but without changing neither the dimensions nor the position (See Tables 2 and 3). 4. Zircaloy-4 also was used as a clad for both the UO2 fuel and as a frame for the FE.

Table 2 General characteristics of the fuel material, fuel element, absorber material of the ETRR-2 reactor fueled by the U3O8-Al original fuel and re-fueled by the UO2 with the ONT and TNTs for the fuel material consisting of the one piece. Parameter

U3O8-Al original fuel with ONT

UO2 fuel with ONT

UO2 fuel with TNTs

Fuel material Fuel meat Enrichment 235U (%) Total core loading 235U (g) Density of fuel meat (g/cm3) Cladding material

U3O8-Al 19.7 6944.5 See Table 1 Al-6061

UO2 4.44 7859.73 10.17 Zircaloy-4

UO2 5.4 8926.01 10.17 Zircaloy-4

Fuel element Number of the FEs Number of fuel plates in the FE Dimensions of the fuel piece (cm) (length  width  thickness)

29 19 80  6.4  0.07

29 19 80  6.4  0.07

27 19 80  6.4  0.07

Ag-In-Cd Wright percentage %

B4C Wright percentage %

B4C Wright percentage %

Ag In Cd

10

10

Absorber material Composition

3

Density (g/cm )

15 80 5 10.18

B B

48 32 20

11

C 1.80

B B

44 36 20

11

C 1.80

146

I. Shaaban, M. Albarhoum / Annals of Nuclear Energy 98 (2016) 144–156

Fig. 2. A cross section of the FP used in the ETRR-2 reactor using the MCNP4C code.

Fig. 1. A schematic representation of the 1/98 ETRR-2 core in the plane X–Y with ONT using the MCNP4C code.

Table 4 General characteristics of the active zone dimensions and the water gap between plates of the ETRR-2 core fueled by U3O8-Al original fuel and re-fueled by UO2 fuel with the ONT and TNTs.

4. Adding neutronic traps in the ETRR-2 core The ETRR-2 core consists of 29 positions for FEs and one central NT is located near the center of the core (See Figs. 1 and 4). The UO2 fuel was used for ONT and TNTs. For the TNTs, two irradiation boxes (two NTs) are placed inside the ETRR-2 core to increase the scientific applications of the ETRR-2 reactor and for the purpose of 99Mo production replacing two FEs to produce 99Tc and other radioisotopes, and the central NT is not changed at all as shown in Fig. 5. The frame of the three NTs is made of Al-6061 alloy with the same dimensions of the FE.

Parameter

One piece

Eight pieces

Active zone dimensions Active length (cm) Clad length (cm) External section of fuel element (cm2) Section in grid to house the fuel element (cm2) Plate thickness (cm) Meat thickness (cm) Meat width (cm) Side plate thickness (cm) Side plate width (cm) External distance between frames (cm) Internal distance between frames (cm)

80 80 88 8.1  8.1 0.15 0.07 6.40 0.50 8.00 8.00 7

79.4 80 88 8.1  8.1 0.15 0.07 6.20 0.50 8.00 8.00 7

Water gap between plates Of single fuel element (cm) Of different fuel element (cm)

0.27 0.39

0.27 0.39

Table 3 General characteristics of the fuel material, fuel element, absorber material of the ETRR-2 reactor fueled by the U3O8-Al original fuel and re-fueled by the UO2 with the ONT and the TNTs for the fuel material consisting of the eight pieces. Parameter

UO2 fuel with ONT

UO2 fuel with TNTs

Fuel material Fuel meat Enrichment235U (%) Total core loading 235U (g) Density of meat (g/cm3) Cladding material

UO2 4.6 7829.32 10.17 Zircaloy-4

UO2 5.65 8979.51 10.17 Zircaloy-4

Fuel element Number of the FE Number of fuel plates in the FE Number of fuel pieces in the FP Dimensions of the fuel piece (cm) (length  width  thickness)

29 19 8 19.85  3.1  0.07

27 19 8 19.85  3.1  0.07

B4C Wright percentage %

B4C Wright percentage %

Absorber material Composition

10

B B

64 16 20

11

C Density (g/cm3)

1.8

10

B B

44 36 20

11

C 1.8

147

I. Shaaban, M. Albarhoum / Annals of Nuclear Energy 98 (2016) 144–156

Fig. 3. A schematic representations of cladding and fuel components for compartmented plate-type oxide fuel element for the Shippingport reactor.

Fig. 5. A cross section of the ETRR-2 core in the plane X–Y with TNTs using the MCNP4C code.

5.1. Calculation of the criticality safety parameters The criticality safety parameters of the ETRR-2 reactor are: - the excess core reactivity q, - the Shutdown Margin (SM) of the control plates, - the SM of the control plates with Single plate Failure (SM with SF), - the Control Rod Worth (CRW), - the Reactivity Safety Factor (RSF) with RSF meaning total control rod worth/core excess reactivity q, - the effective delayed neutron fraction (beff).

Fig. 4. A cross section of the current ETRR-2 core in the plane X–Y with ONT using the MCNP4C code.

5. Calculation of the criticality safety and the neutronic parameters of the ETRR-2 using the MCNP4C code before and after replacement of the U3O8-Al fuel by the UO2 fuel To perform the criticality safety calculations using the MCNP4C code, the nuclear data for fissile and non-fissile materials such as: fuel and clad, coolant, moderator, control rod, clad, and reflector were taken from the ENDF/B-VI nuclear data libraries, and the thermal particle scattering S(a,b) was applied to treat the thermal scattering in beryllium reflector and in the hydrogen of the moderating water. In the MCNP4C model, three hundred million neutron histories (106 neutron particle source histories per cycle were made for 300 cycles (with an initial criticality keff guess of 1 and twenty non active cycles) are used to simulate the core and accumulate the reactor tallies to calculate the criticality safety and the neutronic parameters of the ETRR-2 reactor before and after replacement its U3O8-Al original fuel by the UO2 fuel with ONT and TNTs.

The criticality safety parameters were calculated by the KCODE criticality source card (Briesmeister, 2000) and using all fuel elements as fission source points, where the fission source is located in the middle of each FE. The criticality calculations were done for a clean fresh core (zero burn-up). The effective multiplication factor keff was estimated by running the input file of the ETRR-2 reactor by the MCNP4C code with All Control Plates with drawn Out (ACPO) from the reactor core. The excess core reactivity q was calculated using the following equation (Duderstadt and Hamilton, 1976).

Table 5 Measured and calculated values of the excess of reactivity, the SM and SM with SF of the 1/98 ETRR-2 core fueled by U3O8-Al original for the ONT. Core 1/98 of the ETRR-2 fueled by U3O8-Al original fuel with ONT (Fig. 1) Fuel material in the fuel plate consisting of the one piece Parameter

Measured

Excess of reactivity ($) SM($) SM with SF ($)

9.1 15.2 8.7

a

Calculated values using MCNP4C 8.964 ± 0.049 15.849 ± 0.029 8.929 ± 0.029

a References values of the core 1/98 were taken from the references (Nagy et al., 2004; Gaheen, 2010; Hussein et al., 2011; http://www.igorr.com/home/liblocal/docs/Proceeding/Meeting%208/ps2_villarino.pdf).

148

I. Shaaban, M. Albarhoum / Annals of Nuclear Energy 98 (2016) 144–156

Table 6 Calculated values of the excess of reactivity, the SM and SM with SF, CRW, RSF and the beff of the current ETRR-2 core fueled by U3O8-Al original fuel for the ONT. Current ETRR-2 core fueled by U3O8-Al original fuel with ONT (See Fig. 4) Fuel material in the fuel plate consisting of the one piece Parameter

Calculated values using MCNP4C

Excess of reactivity ($) SM($) SM with SF ($) CRW ($) RSF beff (pcm) beff (pcm)

9.376 ± 0.007 15.141 ± 0.036 8.429 ± 0.036 24.670 ± 0.043 2.631 ± 0.004 723 ± 7 740a

a Reference value was taken from the references (www.etrr2-aea.org.eg/Data Sheet about MTR-22MW.html and Main Core Data of MTR Reactor.html; http:// archive.is/Mnf73).

Table 7 Calculated values of the excess of reactivity, SM and SM with SF, CRW, RSF and the beff of the current ETRR-2 core re-fueled by UO2 fuelf or the fuel material in the fuel plate consisting of the one piece and eight pieces with the ONT. Current ETRR-2 core re-fueled by UO2 with ONT (See Fig. 4) Parameter

Calculated values using MCNP4C

Fuel material in the fuel plate consisting of the one piece Excess of reactivity ($) 9.145 ± 0.006 SM($) 15.645 ± 0.037 SM with SF ($) 8.448 ± 0.037 CRW ($) 24.791 ± 0.041 RSF 2.711 ± 0.004 beff (pcm) 767 ± 7

Table 8 Calculated values of the excess of reactivity, SM and SM with SF, CRW, RSF and the beff of the current ETRR-2 core re-fueled by UO2 fuel for the fuel material in the fuel plate consisting of the one piece and eight pieces with the TNTs. Current ETRR-2 core re-fueled by UO2 with TNTs (See Fig. 5)

Fuel material in the fuel plate consisting of the eight pieces Excess of reactivity ($) 9.352 ± 0.006 SM($) 14.867 ± 0.035 SM with SF ($) 8.654 ± 0.035 CRW ($) 24.219 ± 0.040 RSF 2.590 ± 0.004 beff (pcm) 726 ± 7

ð1Þ

The reactivity worth of the control rod was obtained using the relation:

CRW ¼ q þ SM;

ð3Þ

k1 ¼ kp þ kdeff

ð4Þ

where kp denotes the prompt neutron contribution to keff, kdeff the contribution of delayed neutrons. The obtained results of the beff for all the studied cases are given in Tables 6–8.

The reactor is designed to be used in a wide variety of fields including neutron physics, Neutron Activation Analysis (NAA), radioisotope production (e.g., 14C,32S, 51Cr,60Co, 89Sr, 153Sm,169Yb, 170 Tm, 192Ir), Neutron Transmutation Doping (NTD) of silicon, materials science, training and education for new operators, neutron radiography and boron capture therapy (SAR, 1997). Therefore, one of the duties of the calculations in this paper is to avoid degrading the capabilities of the reactor. On the contrary it is hoped to increase and develop new capabilities such as 99Mo production to produce 99Tc and other radioisotopes for scientific applications. These applications are somewhat related to the values of the Thermal Neutron Flux (TNF) in the NTs used for radioisotope production in the ETRR-2 reactor to evaluate the radioisotopes yields. The neutronic parameters in the ETRR-2 reactor are:

Calculated values using MCNP4C

Fuel material in the fuel plate consisting of the one piece Excess of reactivity ($) 9.088 ± 0.007 SM($) 15.105 ± 0.038 SM with SPF ($) 8.378 ± 0.038 CRW ($) 24.194 ± 0.042 RSF 2.662 ± 0.005 beff (pcm) 766 ± 7

q ¼ ðkeff  1Þ=keff

beff ¼ 1  ðkp =k1 Þ

5.2. Calculation of the neutronic parameters

Fuel material in the fuel plate consisting of the eight pieces Excess of reactivity ($) 9.218 ± 0.006 SM($) 15.199 ± 0.038 SM with SF($) 8.158 ± 0.038 CRW ($) 24.417 ± 0.040 RSF 2.649 ± 0.004 beff (pcm) 716 ± 7

Parameter

plate, respectively. The obtained results of the 1/98 ETRR-2 core (See Fig. 1) fueled by U3O8-Al FEs (the first digit is a correlative number and the last two digits are the year, then 1/98 is the first core in 1998 (Nagy et al., 2004) and the current ETRR-2 core fueled by U3O8-Al FEs (Fig. 4) are given in Tables 5 and 6. In addition, the MCNP4C results of the ETRR-2 core re-fueled by UO2 FEs for FM in the FP consisting of one UO2 piece and eight UO2 pieces with ONT and TNTs are presented in Tables 7 and 8. The effective delayed neutron fraction, beff, is given by the number of neutrons produced in reactions induced by delayed neutrons, divided by the total number of neutrons produced. To estimate the value of the effective delayed neutron fraction beff, the input file of the ETRR-2 reactor was run by the MCNP4C code using the ‘‘TOTNU” card with ‘‘NO” as the only entry turns off delayed neutrons in k-code mode, producing a final k-eigenvalue corresponding to kp. The value of the beff was calculated using the relation (Westren, 2007).

ð2Þ

where the SM and SM with SF is the negative reactivity, and they are calculated when the Control Plates are Fully Inserted (CPFI) and when the control plates are fully inserted but failing a single

1. the Average TNF (ATNF), Average Epi-Thermal Neutron Flux (AETNF) and Average Fast Neutron Flux (AFNF) in the central NT that is located in site 1 for the U3O8-Al original fuel and the UO2 fuel (current ETRR-2 core with ONT, See Fig. 4), 2. the ATNF, AETNF and AFNF in the TNTs which are located in site 1, site 2 and site 3 (current ETRR-2 core with TNTs, See Fig. 5) for the UO2 fuel, 3. the TNF, ETNF and FNF in the Be reflector for the current ETRR-2 core with the ONT and TNTs (See Figs. 4 and 5) for the U3O8-Al original fuel and the UO2 fuel, 4. the TNF, ETNF and FNF in the Central Irradiation Box (CIB) that is located inside the NT in site 1 (See Fig. 4) for the U3O8-Al original fuel and the UO2 fuel, and in the CIBs which are located inside the TNTs in site 1, site 2 and site 3 (See Fig. 5) for the UO2 fuel. Where the CIB is designed to produce 60Co for medical and scientific applications. To calculate these values, the F4 tally, the FS, the SD and the FM cards in the MCNP4C code were used in the input file of the ETRR-2 reactor, and then the input file was run by the MCNP4C code with the same conditions which were applied for the criticality safety calculations (See paragraph 5.1). The F4 tally, the FS, the SD and the FM cards are used as follows:

149

I. Shaaban, M. Albarhoum / Annals of Nuclear Energy 98 (2016) 144–156

The F4:n $ card: this card allows to estimate the track-length of the neutron flux in the desired cell. The FS $ card: this card allows to subdivide a cell or a surface into segments for tallying purposes. The SD $ card: this card allows to divide a volume or area into segments for tallying purposes. The E $ card: this is the Energy bins in MeV. The FMC $ card: this is a Tally multiplier (C – is the source strength of the ETRR-2 reactor (See manual of the MCNP4C code, Briesmeister, 2000)). The calculated values of the ATNF, AETNF and AFNF in the NTs, and the TNF, ETNF and FNF in the Be reflector, and the TNF, ETNF and FNF in the CIBs are tabulated in Tables 9–13 before and after replacement of the U3O8-Al original fuel of the current ETRR-2 core by the UO2 fuel for the ONT and the TNTs, and for the three following cases:

1. ACPO of the current ETRR-2 core fueled by the U3O8-Al original fuel for the ONT. In this case the keff = 1.07697 ± 0.00038 (See Table 9). 2. The current ETRR-2 core re-fueled by the UO2 fuel for ONT and TNTs (See Tables 10 and 12), and for the case the FM in the FP is consisting of one piece, and for the two following cases: - ACPO of the current ETRR-2 core. Where the keff = 1.07364 ± 0.00037andkeff = 1.073150 ± 0.00042 for the ONT and TNTs, respectively. - The criticality case. Where the keff = 1.00024 ± 0.00040 and keff = 1.00006 ± 0.00042, for the ONT and TNTs, respectively. 3. The current ETRR-2 core re-fueled by the UO2 fuel for the ONT and the TNTs (See Tables 11 and 13), and for the FM in the FP is consisting of eight pieces, and for the two following case: - ACPO of the current ETRR-2 core. Where the keff = 1.07427 ± 0.00038 and keff = 1.07543 ± 0.00044 for the ONT and TNTs, respectively.

Table 9 Reference and calculated values of the neutron flux of the current ETRR-2 core fueled by U3O8-Al original fuel for the ONT and the fuel material consisting of one piece. Neutron flux for the fuel material consisting of one piece Current ETRR-2 core fueled by U3O8-Al original fuel for the ONT (Fig. 4)

a b

Parameter

Reference value

Calculated value using MCNP4C keff = 1.07697 ± 0.00038

Calculated value using MCNP4C keff = 1.00038 ± 0.00039

TNF in the CIB in site 1 (n/cm2 s)  1014 ETNF in the CIB in site 1 (n/cm2 s)  1014 FNF in the CIB in site 1 (n/cm2 s)  1014 ATNF in the NT in site 1 (n/cm2 s)  1014 AETNF in the NT in site 1 (n/cm2 s)  1014 AFNF in the NT in site 1 (n/cm2 s)  1014 TNF in the Be reflector (n/cm2 s)  1014 ETNF in the Be in site 1 (n/cm2 s)  1014 FNF in the Be in site 1 (n/cm2 s)  1014

4.230a

4.215 ± 0.017 1.018 ± 0.031 1.533 ± 0.014 2.700 ± 0.007 0.867 ± 0.008 1.450 ± 0.007 ±0.014 0.227 ± 0.024 0.297 ± 0.019

4.026 ± 0.018 0.988 ± 0.032 1.388 ± 0.026 2.877 ± 0.007 0.948 ± 0.008 1.588 ± 0.007 0.998 ± 0.015 0.217 ± 0.027 0.019

2.700b

1.000b

Reference value was taken from reference Hussein et al. (2011). References values were taken from the references (www.etrr2-aea.org.eg/Data Sheet about MTR-22MW.html and Main Core Data of MTR Reactor.html).

Table 10 Calculated values of the neutron flux of the current ETRR-2 core re-fueled by UO2 for the ONT and the fuel material consisting of the one piece. Neutron flux for the fuel material consisting of the eight pieces Current ETRR-2core re-fueled by UO2 fuel and ONT (Fig. 4) Parameter

Calculated value for keff = 1.07364 ± 0.00037

Calculated value for keff = 1.00024 ± 0.00040

TNF in the CIB in site 1 (n/cm2 s)  1014 ETNF in the CIB in site 1 (n/cm2 s)  1014 FNF in the CIB in site 1 (n/cm2 s)  1014 ATNF in the NT in site 1 (n/cm2 s)  1014 AETNF in the NT in site 1 (n/cm2 s)  1014 AFNF in the NT in site 1 (n/cm2 s)  1014 TNF in the Be reflector (n/cm2 s)  1014 ETNF in the Be reflector (n/cm2 s)  1014 FNF in the Be reflector (n/cm2 s)  1014

4.424 ± 0.018 1.077 ± 0.030 1.591 ± 0.024 2.700 ± 0.007 0.881 ± 0.008 1.450 ± 0.007 0.998 ± 0.013 0.217 ± 0.026 0.277 ± 0.022

3.988 ± 0.018 0.979 ± 0.032 1.403 ± 0.026 2.933 ± 0.007 0.971 ± 0.008 1.625 ± 0.007 1.023 ± 0.012 0.202 ± 0.023 0.254 ± 0.022

Table 11 Calculated values of the neutron flux of the current ETRR-2 core re-fueled by UO2 for the ONT and the fuel material consisting of the eight pieces. Neutron flux for the fuel material consisting of the one piece Current ETRR-2core re-fueled by UO2 fuel and ONT (Fig. 4) Parameter

Calculated value for keff = 1.07427 ± 0.00038

Calculated value for keff = 1.00052 ± 0.00039

TNF in the CIB in site 1 (n/cm2 s)  1014 ETNF in the CIB in site 1 (n/cm2 s)  1014 FNF in the CIB in site 1 (n/cm2 s)  1014 ATNF in the NT in site 1 (n/cm2 s)  1014 AETNF in the NT in site 1 (n/cm2 s)  1014 AFNF in the NT in site 1 (n/cm2 s)  1014 TNF in the Be reflector (n/cm2 s)  1014 ETNF in the Be reflector (n/cm2 s)  1014 FNF in the Be reflector (n/cm2 s)  1014

4.431 ± 0.018 1.079 ± 0.030 1.602 ± 0.023 2.712 ± 0.012 0.878 ± 0.008 1.462 ± 0.007 1.000 ± 0.013 0.217 ± 0.027 0.278 ± 0.022

4.038 ± 0.018 1.016 ± 0.032 1.376 ± 0.026 2.925 ± 0.008 0.966 ± 0.008 1.612 ± 0.007 0.983 ± 0.014 0.202 ± 0.027 0.261 ± 0.022

150

I. Shaaban, M. Albarhoum / Annals of Nuclear Energy 98 (2016) 144–156

Table 12 Calculated values of the neutron flux of the current ETRR-2 core re-fueled by UO2 for the TNTs and the fuel material consisting of the one piece. Neutron flux for the fuel material consisting of the one piece Current ETRR-2core with UO2 fuel and TNTs (Fig. 5) Parameter

Calculated value for keff = 1.07315 ± 0.00042

Calculated value for keff = 1.00006 ± 0.00042

TNF in the CIB in site 1 (n/cm2 s)  1014 ETNF in the CIB in site 1 (n/cm2 s)  1014 FNF in the CIB in site 1 (n/cm2 s)  1014 TNF in the CIB in site 2 (n/cm2 s)  1014 ETNF in the CIB in site 2 (n/cm2 s)  1014 FNF in the CIB in site 2 (n/cm2 s)  1014 TNF in the CIB in site 3 (n/cm2 s)  1014 ETNF in the CIB in site 3 (n/cm2 s)  1014 FNF in the CIB in site 3 (n/cm2 s)  1014 ATNF in the NT in site 1 (n/cm2 s)  1014 AETNF in the NT in site 1 (n/cm2 s)  1014 AFNF in the NT in site 1 (n/cm2 s)  1014 ATNF in the NT in site 2 (n/cm2 s)  1014 AETNF in the NT in site 2 (n/cm2 s)  1014 AFNF in the NT in site 2 (n/cm2 s)  1014 ATNF in the NT in site 3 (n/cm2 s)  1014 AETNF in the NT in site 3 (n/cm2 s)  1014 AFNF in the NT in site 3 (n/cm2 s)  1014 TNF in the Be reflector (n/cm2 s)  1014 ETNF in the Be reflector (n/cm2 s)  1014 FNF in the Be reflector (n/cm2 s)  1014

4.185 ± 0.018 1.089 ± 0.031 1.465 ± 0.025 3.823 ± 0.019 0.841 ± 0.035 1.185 ± 0.028 3.822 ± 0.019 0.856 ± 0.034 1.212 ± 0.027 2.537 ± 0.008 0.830 ± 0.008 1.400 ± 0.007 2.387 ± 0.008 0.704 ± 0.009 1.164 ± 0.009 2.387 ± 0.008 0.701 ± 0.009 1.162 ± 0.008 1.025 ± 0.015 0.239 ± 0.025 0.301 ± 0.029

3.855 ± 0.019 0.916 ± 0.034 1.390 ± 0.026 2.778 ± 0.022 0.709 ± 0.037 1.069 ± 0.030 2.784 ± 0.022 0.678 ± 0.038 1.074 ± 0.030 2.762 ± 0.008 0.926 ± 0.008 1.652 ± 0.007 2.403 ± 0.009 0.740 ± 0.009 1.242 ± 0.008 2.400 ± 0.009 0.736 ± 0.009 1.241 ± 0.008 0.987 ± 0.015 0.225 ± 0.026 0.282 ± 0.020

Table 13 Calculated values of the neutron flux of the current ETRR-2 core re-fueled by UO2 for the TNTs and the fuel material consisting of the eight pieces. Neutron flux for the fuel material consisting of the eight pieces Current ETRR-2core with UO2 fuel and TNTs (Fig. 5) Parameter

Calculated value for keff = 1.07543 ± 0.00044

Calculated value for keff = 1.00089 ± 0.00041

TNF in the CIB in site 1 (n/cm2 s)  1014 ETNF in the CIB in site 1 (n/cm2 s)  1014 FNF in the CIB in site 1 (n/cm2 s)  1014 TNF in the CIB in site 2 (n/cm2 s)  1014 ETNF in the CIB in site 2 (n/cm2 s)  1014 FNF in the CIB in site 2 (n/cm2 s)  1014 TNF in the CIB in site 3 (n/cm2 s)  1014 ETNF in the CIB in site 3 (n/cm2 s)  1014 FNF in the CIB in site 3 (n/cm2 s)  1014 ATNF in the NT in site 1 (n/cm2 s)  1014 AETNF in the NT in site 1 (n/cm2 s)  1014 AFNF in the NT in site 1 (n/cm2 s)  1014 ATNF in the NT in site 2 (n/cm2 s)  1014 AETNF in the NT in site 2 (n/cm2 s)  1014 AFNF in the NT in site 2 (n/cm2 s)  1014 ATNF in the NT in site 3 (n/cm2 s)  1014 AETNF in the NT in site 3 (n/cm2 s)  1014 AFNF in the NT in site 3 (n/cm2 s)  1014 TNF in the Be reflector (n/cm2 s)  1014 ETNF in the Be reflector (n/cm2 s)  1014 FNF in the Be reflector (n/cm2 s)  1014

4.305 ± 0.018 0.956 ± 0.031 1.516 ± 0.026 3.813 ± 0.019 0.820 ± 0.034 1.265 ± 0.028 3.911 ± 0.019 0.867 ± 0.034 1.270 ± 0.028 2.575 ± 0.008 0.837 ± 0.008 1.412 ± 0.009 2.363 ± 0.008 0.706 ± 0.009 1.172 ± 0.007 2.375 ± 0.008 0.704 ± 0.008 1.163 ± 0.007 1.033 ± 0.014 0.231 ± 0.025 0.298 ± 0.029

3.827 ± 0.019 0.918 ± 0.032 1.370 ± 0.026 2.829 ± 0.022 0.732 ± 0.037 1.040 ± 0.030 2.776 ± 0.022 0.727 ± 0.036 1.087 ± 0.030 2.775 ± 0.008 0.916 ± 0.008 1.563 ± 0.007 2.425 ± 0.009 0.734 ± 0.009 1.233 ± 0.009 2.413 ± 0.008 0.738 ± 0.008 1.240 ± 0.008 1.003 ± 0.014 0.221 ± 0.026 0.277 ± 0.021

- The criticality case. Where the keff = 1.00052 ± 0.00039 and keff = 1.00089 ± 0.00041, for the ONT and TNTs, respectively. Figs. 6–8 show the variation of the TNF, ETNF and FNF along the height of the neutronic trap for the ETRR-2 core fueled by the U3O8-Al original fuel (one plate) and UO2 fuel (one plate and 8 plates) with ONT for the ACPO case, while Figs. 9–11 show the same quantities for the criticality case. The neutronics calculations were performed using three energy groups as: <0.625 eV for thermal neutrons, (0.625 eV to 5.53 keV) for epithermal neutrons and up to 20 MeV for fast neutrons. Calculations were also normalized to the steady-state power level of 22 MW. The validity of these calculations for the U3O8-Al original fuel is verified by comparing them with the reference values (See Table 9).

6. Burn-up of the fuel The cycle length of the ETRR-2 reactor fueled by the U3O8-Al original fuel is about 18 days (Khalil et al., 2005).Therefore, the MCNP4C and the ORIGEN-S codes (a depletion code (Hermann and Westfall, 1998)), and the PFCOM system (Shaaban and Albarhoum, 2015) were used to calculate the reactivity at the end of the fuel cycle for the ETRR-2 core fueled by the U3O8-Al original fuel and re-fueled by the UO2 fuel with ONT and TNTs. In the calculation the actinides 233U, 234U, 235U, 236U, 238U, 237Np, 238 Pu,239Pu,240Pu,241Pu,242Pu, 232Th,241Am, 242mAm and 243Am, major fission products 91Zr, 115In, 113Cd, 155Gd, 157Gd, 83Kr, 84Kr, 86 Kr, 95Mo, 99Tc, 109Ag, 101Ru, 103Ru, 103Rh, 105Rh, 131Xe, 135Xe, 133 Cs, 134Cs, 135Cs, 136Cs, 137Cs, 105Pd, 108Pd, 143Nd, 145Nd, 147Nd, 148 Nd, 147Pm, 148Pm, 149Pm, 147Sm, 149Sm, 150Sm, 151Sm, 152Sm,

I. Shaaban, M. Albarhoum / Annals of Nuclear Energy 98 (2016) 144–156

151

Fig. 6. TNF as along the height of the neutronic trap for the ETRR-2 core fueled by the U3O8-Al original fuel (one plate) and UO2 fuel (one plate and 8 plates) with ONT.

Fig. 7. ETNF as along the height of the neutronic trap for the ETRR-2 core fueled by the U3O8-Al original fuel (one plate) and UO2 fuel (one plate and 8 plates) with ONT.

151

Eu, 153Eu, 154Eu, 155Eu and light elements 1H, 6Li, 13C, 16O and Al were taken in the calculation. The calculated values of the reactivity at the end of the fuel cycle are given in Table 14 for all studies cases.

27

7. Results and discussion As a result it can be seen in Tables 5, 6 and 9 that there is a good agreement between the calculated values of the criticality parameters(excess core reactivity, SM and SM with SF) of the 1/98 ETRR-2 core (See Fig. 1), the calculated value of the beff, the neutron flux (TNF in the CIB, the ATNF in the NT and the TNF in the Be reflector) of the current ETRR-2 core (See Fig. 4) and the measured, and the reference values. In fact, the differences are less than 4.2%, 2.3% and 1.2% for the criticality parameters, the beff and the neutron flux,

respectively. This agreement will be used as the base to accredit reliability to the obtained results of the UO2 fuel in the ETRR-2 reactor. Table 6 shows that the calculated values of the criticality parameters (excess core reactivity, SM and SM with SF) of the current ETRR-2 core fueled by U3O8-Al original fuel (See Fig. 4) differ from the same parameters of the 1/98 ETRR-2 core by 4.39%, 4.47% and 5.6% for the excess core reactivity, SM and SM with SF, respectively. These differences should due to the change in the structure of the ETRR-2 core where the Be cubes were added around the current ETRR-2 core (See Figs. 1 and 4). Tables 7 and 8 show the calculated values of the criticality parameters (excess core reactivity, SM and SM with SF, CRW, RSF and beff) of the current ETRR-2 core (See Fig. 4) re-fueled

152

I. Shaaban, M. Albarhoum / Annals of Nuclear Energy 98 (2016) 144–156

Fig. 8. FNF as along the height of the neutronic trap for the ETRR-2 core fueled by the U3O8-Al original fuel (one plate) and UO2 fuel (one plate and 8 plates) with ONT.

Fig. 9. TNF for the criticality case as along the height of the neutronic trap for the ETRR-2 core fueled by the U3O8-Al original fuel (one plate) and UO2 fuel (one plate and 8 plates) with ONT.

by UO2 fuel consisting of one piece and eight pieces with ONT and TNTs in the ETRR-2 core (See Figs. 4 and 5). By comparing these values with the same values of the current ETRR-2 core fueled by the U3O8-Al original fuel, one can find that the maximum error between them does not exceed 3.32% for the UO2 fuel consisting of one piece and eight pieces with ONT and TNTs. As well as, the maximum error of the beff between the reference value and the calculated values in all studied cases of the ETRR-2 reactor with UO2 fuel does not exceed about 3.65%. This good agreement between the calculated values of the criticality parameters and the same parameters before and after replacing the U3O8-Al fuel by the UO2 fuel indicates that the UO2 fuel can be safely used to substitute the U3O8-Al fuel in the ETRR-2 reactor without any negative effect on the criticality parameters of the reactor. From Tables 10 and 11 it comes out that:

- The calculated values of the TNF in the CIB for the ACPO case of the current ETRR-2 core re-fueled by UO2 fuel are a bit higher than those of the reference values in the case of the U3O8-Al original fuel, where the difference between the calculated and the reference values are 7.11% and 4.59% for the UO2 FM consisting of one piece and eight pieces, respectively. This result is due to that the mass of 235U loaded in the ETRR-2 core is higher by about 13.21% and 12.77% for the UO2 FM consisting of one piece and eight pieces, respectively than the 235U mass for the U3O8-Al original fuel as can be seen from Tables 2 and 3. This reduces a bit the time required to produce the 60Co. - The calculated values for the ETNF and FNF in the CIB for the ACPO case of the current ETRR-2 core re-fueled by UO2 fuel are higher by 5.79% and 5.99%, and 3.78% and 4.50% than the calculated values of the ETRR-2 core fueled by the U3O8-Al original fuel (See above).

I. Shaaban, M. Albarhoum / Annals of Nuclear Energy 98 (2016) 144–156

153

Fig. 10. ETNF for the criticality case as along the height of the neutronic trap for the ETRR-2 core fueled by the U3O8-Al original fuel (one plate) and UO2 fuel (one plate and 8 plates) with ONT.

Fig. 11. FNF for the criticality case as along the height of the neutronic trap for the ETRR-2 core fueled by the U3O8-Al original fuel (one plate) and UO2 fuel (one plate and 8 plates) with ONT.

- The calculated values of the TNF in the Be reflector for the UO2 FM consisting of one piece and eight pieces with ONT are in good agreement with the reference values for the ACPO and the criticality case. - The maximum difference between the calculated values of the ATNF, AETNF and AFNF in the NT and the reference values does not exceeding 1.61% for the ACPO case. - The TNF in the CIB is reduced by 10.88% and 9.83% whereas the ATNF in the NT is increased by 7.85% and 8.63% for the criticality case and the UO2 FM consisting of one piece and eight pieces, respectively.

- The TNF in the CIB and the ATNF in the NT is higher than the 2.0  1014 n/cm2 s for the ACPO and the criticality cases. This value is sufficient to produce 99Mo and other radioisotopes such as: 14C,32S, 51Cr,60Co, 89Sr, 153Sm,169Yb, 170Tm, 192Ir (IAEA, 2003a, b; SAR, 1997; Shaat, 2010; NEA, 2010). - The AETNF and AFNF in the CIB is reduced by 9.01% and 11.82%, and 5.84% and 14.10% for the criticality case of the ETRR-2 core re-fueled by the UO2 fuel FM consisting of one piece and eight pieces, respectively.

154

I. Shaaban, M. Albarhoum / Annals of Nuclear Energy 98 (2016) 144–156

Table 14 Calculated values of the reactivity at the end of the fuel cycle of the all studied cases. Cycle length (day)

Reactivity (pcm)

U3O8-Al original fuel (one plate) with ONT 18

2412.37 (2456)a

UO2 fuel (one plate) with ONT 18

2429.50

UO2 fuel (8 plates) with ONT 18

2479.93

UO2 fuel (one plate) with TNTs 18

2471.37

UO2 fuel (8 plates) with TNTs 18

2548.36

a

Reference value (Khalil et al., 2005) of the reactivity at the end of the fuel cycle (17.61 day) of the ETRR-2 core fueled by the U3O8-Al original fuel.

- The AETNF and AFNF in the NT is increased by 10.21% and 12.07%, and 10.02% and 10.25% for the criticality case of the ETRR-2 core re-fueled by the UO2 fuel FM consisting of one piece and eight pieces, respectively. Figs. 6–11 showed that there was a good agreement between the calculated values of the TNF, ETNF and FNF along the height of the neutronic trap of the ETRR-2 core fueled by the U3O8-Al original fuel and the same values of the ETRR-2 core re-fueled by the UO2 fuel for the ACPO and criticality cases and with ONT, where the maximum difference between the average values does not exceed 4.3%. As a result the replacement of the U3O8-Al original fuel by the UO2 fuel does not have any negative effects on the neutronic parameters of the ETRR-2 reactor and its scientific applications which were available before the fuel replacement. Tables 12 and 13 show the following: - The difference between the calculated values of the TNF in the CIB and the ATNF in the NT in site 1 (See Fig. 5) for the ACPO case of the ETRR-2 core re-fueled by the UO2 fuel are far from

the reference values in the case of theU3O8-Al original fuel by about 1.06% and 6.03%, and 1.77% and 4.63%for the UO2 FM consisting of one piece and eight pieces, respectively. This result leads to conclude that the replacement of two fuel elements does not change the neutronic characteristics of this site when the UO2 fuel is used. Therefore, the reactor scientific applications which were available before fuel replacement will still be available now with the TNTs. - The TNF in the CIBs and the ATNF in the NTs in sites 1, 2 and 3 (Fig. 5) for the ACPO and the criticality cases are higher than 2  1014 n/cm2 s for the UO2 FM consisting of one piece and eight pieces. This value is sufficient to produce99Mo and other radioisotopes such as: 14C,32S, 51Cr,60Co, 89Sr, 153Sm,169Yb, 170 Tm, 192Ir (IAEA, 2003; SAR, 1997; Shaat, 2010; NEA, 2010). This result leads to say that the modified ETRR-2 core with UO2will allow to increase the scientific applications of the ETRR-2 reactor such as: NAA, NTD of silicon and materials science, and increase the amount of the 60Co and 99Mo produced in the ETRR-2 reactor if the reactor was used only to produce 60 Co and 99Mo. - The TNF, ETNF and FNF in the CIB in the site 1 decrease by about 9.49%, 9.92% and 7.37% whereas the calculated values of those parameters increase in the NT (site 1) by about8.32%, 10.50% and 14.34%, respectively. As well as, the calculated values of those parameters in the CIBs in the site 2 and site 3 decrease by about 27.32%, 15.83% and 13.28% whereas their values increase in the NTs (site 2 and site 3) by about1.35%, 4.72% and 6.28%, respectively for the criticality case and the ETRR-2 core re-fueled by the UO2 FM consisting of one piece and eight pieces. Generally speaking, as from Tables 12 and 13 it could be said that the calculated values of the ATNF in the NTs, the TNF in the CIBs in the sites 1, 2 and 3, and the ATNF in the Be reflector of the UO2 are in good agreement with the reference values of the same parameters for the U3O8-Al fuel with errors not exceeding 6.03% . This agreement indicates that the replacement of the U3O8-Al fuel by UO2 fuel and within the limits that are mentioned

Table 15 Calculated values of the neutron flux in the control plates of the ETRR-2 core re-fueled by the UO2 fuel for ONT and fuel material consisting from one plate. Parameter

UO2 fuel with Ag-In-Cd as a control plates

UO2 fuel with B4C as a control plates

Plate number 1 (See Fig. 12) Thermal neutron (n/cm2 s)  1012 Epithermal neutron (n/cm2 s)  1013 Fast neutron (n/cm2 s)  1013

6.651 ± 0.014 2.145 ± 0.014 4.603 ± 0.010

3.816 ± 0.016 1.026 ± 0.015 4.288 ± 0.010

Plate number 2 (See Fig. 12) Thermal neutron (n/cm2 s)  1012 Epithermal neutron (n/cm2 s)  1013 Fast neutron (n/cm2 s)  1013

7.738 ± 0.013 2.602 ± 0.013 5.700 ± 0.009

4.030 ± 0.016 1.215 ± 0.014 5.110 ± 0.009

Plate number 3 (See Fig. 12) Thermal neutron (n/cm2 s)  1012 Epithermal neutron (n/cm2 s)  1013 Fast neutron (n/cm2 s)  1013

5.555 ± 0.016 1.803 ± 0.015 3.925 ± 0.011

3.171 ± 0.018 0.844 ± 0.016 3.615 ± 0.009

Plate number 4 (See Fig. 12) Thermal neutron (n/cm2 s)  1012 Epithermal neutron (n/cm2 s)  1013 Fast neutron (n/cm2 s)  1013

6.758 ± 0.014 2.118 ± 0.015 4.602 ± 0.010

3.806 ± 0.016 1.036 ± 0.015 4.281 ± 0.010

Plate number 5 (See Fig. 12) Thermal neutron (n/cm2 s)  1012 Epithermal neutron (n/cm2 s)  1013 Fast neutron (n/cm2 s)  1013

7.570 ± 0.013 2.575 ± 0.013 5.683 ± 0.009

4.111 ± 0.016 1.232 ± 0.014 5.149 ± 0.009

Plate number 6 (See Fig. 12) Thermal neutron (n/cm2 s)  1012 Epithermal neutron (n/cm2 s)  1013 Fast neutron (n/cm2 s)  1013

5.662 ± 0.015 1.801 ± 0.016 3.986 ± 0.009

3.210 ± 0.018 0.876 ± 0.017 3.663 ± 0.010

I. Shaaban, M. Albarhoum / Annals of Nuclear Energy 98 (2016) 144–156

155

Table 17 Calculated values of the excess of reactivity, SM and SM with SF, CRW and the RSF of the current ETRR-2 core re-fueled by UO2 fuel and controlled by the Ag-In-Cd plates for the ONT and the TNTs.

Fig. 12. ETRR-2 core with control plates where NT- indicates to the Neutronic Trap (NT), Green color – is the water,

Parameter

Current ETRR-2 core re-fueled by the UO2 fuel for the ONT And the fuel material consisting of the one piece

SM ($) SM with SF ($) CRW ($) RSF

9.533 ± 0.022 4.223 ± 0.022 18.678 ± 0.030 2.042 ± 0.003

Parameter

Current ETRR-2 core re-fueled by the UO2 fuel for the ONT And the fuel material consisting of the eight pieces

SM ($) SM with SF ($) CRW ($) SRF

9.380 ± 0.021 4.230 ± 0.021 18.598 ± 0.029 2.017 ± 0.003

Parameter

Current ETRR-2 core re-fueled by the UO2 fuel for the TNTs And the fuel material consisting of the one piece

SM ($) SM with SPF ($) CRW ($) RSF

9.490 ± 0.023 4.970 ± 0.023 18.579 ± 0.031 2.044 ± 0.003

Parameter

Current ETRR-2 core re-fueled by the UO2 fuel for the TNTs And the fuel material consisting of the eight pieces

SM ($) SM with SPF ($) CRW ($) RSF

9.069 ± 0.023 4.687 ± 0.023 18.421 ± 0.031 1.970 ± 0.003

– is the fuel element, – is the control plate. (For interpretation of the references to color in this figure legend, the reader is referred to the web version of this article.)

Table 16 Material constants of the control plates. Type

Fast neutron

Ag-In-Cd plates D 1.23611E+00 Ra 1.05526E02

Epithermal neutron

Thermal neutron

5.68857E01 9.85794E02

3.67896E02 8.99976E+00

B4C plates for the UO2 fuel with the ONT and the fuel material consisting from one plate D 1.47186 + 00 3.17329E01 4.92927E03 Ra 3.78682E02 6.59878E01 6.72264E+01

in Tables 2 and 3 does not have negative effect on the neutronic parameters of the ETRR-2 reactor and its scientific applications. From Table 14 it results that the calculated values of the reactivity at the end of the fuel cycle of the ETRR-2 core fueled by the U3O8-Al original fuel and re-fueled by UO2 fuel are in good agreement with the reference value, where the maximum difference between these values and the reference value does not exceed 3.70%. This leads to say that there is no negative effect on the cycle length of the ETRR-2 reactor with the UO2 fuel. The absorber material of the control plates of the ETRR-2 core re-fueled by the UO2 was changed from Ag–In–Cd alloy to B4C material without any change in the dimensions and positions. The composition of B4C material is given in Tables 2 and 3 for all studied cases. The B4C absorber material showed more effectiveness than the Ag–In–Cd alloy for the ETRR-2 core re-fueled by the UO2 fuel with ONT and TNTs. The UO2 fuel reduces the effectiveness of the control plates (See Table 15) so they should be substituted by B4C plates. The calculations showed that in the case of the UO2 fuel the thermal neutron fluxes reduce in the control plates positions in average (See Fig. 12 and Table 15) by about 44.35% contributing to reduce the effectiveness of the control plates made of Ag-In-Cd alloy. The decreased effectiveness requires

more effective absorber to be used in the new control plates, which should substitute the old ones, such as B4C being this absorber more effective in the thermal neutron region (Table 16). The calculated values of the SM, the SM with SF, the CRW and SRF of the ETRR-2 core re-fueled by the UO2 fuel with Ag-In-Cd control plates differ from the same values for the ETRR-2 core fueled by the U3O8Al original fuel in average by about 38.13%, 46.28%, 24.85% and 23.29% as shown in Table 17. The cell constants of the absorber material (Ag-In-Cd alloy and B4C) for the control plates being used in this study are given in Table 16. These constants were calculated using the WIMS-D4 code. The dimensions of the cell were taken as: absorber material (0.18 cm thickness), stainless steel cladding (0.265 cm) and water channel (0.371 cm). In these calculations the neutron energies were selected as: - fast energy group: 10 MeV to 5.53 keV, - epithermal energy group: 5.53 keV to 0.625 eV, - thermal energy group: 0.625 eV to 0 eV.

8. Conclusion The UO2 fuel was proposed as fuel in the ETRR-2 reactor. The MCNP4C code was used to simulate and calculate the criticality and the neutronic parameters of the ETRR-2 reactor using a UO2 fuel. The obtained results for the criticality and the neutronic parameters showed a good agreement with the reference values. This work provides the evidence that UO2 fuel can be used in research reactors of type MTR with medium power.

Acknowledgment The authors thank professor I. Othman Director General of Atomic Energy Commission of Syria, for his encouragement and continued support.

156

I. Shaaban, M. Albarhoum / Annals of Nuclear Energy 98 (2016) 144–156

References Albarhoum, M., 2011. Performance of UO2 ceramic fuel in low-power research reactors. Prog. Nucl. Energy 53 (1), 73–75. Bretscher, M.M., Matos, J.E., 1996. Neutronic performance of high density in water moderated and water reflected research reactors. Argonne National Laboratory, 9700 South Cass Avenue, Argonne, Illinois 60439. Briesmeister, J.F., 2000. LA-7396-M, A general Monte Carlo N-particle transport code. Carbajo, J.J., Gradyon, L.Y., Sergey, G.P., Ivanov, V.K., 2001. A review of the thermophysical properties of MOX and UO2 fuels. J. Nucl. Mater. 299, 181–198. Duderstadt, J., Hamilton, L.J., 1976. Nuclear Reactor Physics. John Wiley and Sons Inc., Hoboken. Gaheen, M.A., 2010. Safety aspects of research reactor core modification for fission molybdenum-99 production. In: PERTR 2010 – 32nd International Meeting on Reduced Enrichment for Research and Test Reactors, SANA Lisbona Hotel, Lisbon, Portugal. Hermann, O.W., Westfall, R.M., 1998. ORIGEN-s: SCALE system module to calculate fuel depletion, actinide transmutation, fission product buildup and decay and associated radiation source terms. ORNL/NUREG/CSD-2/V2/R6. Hsieh, T.C., Jankus, V.Z., Rest, J., Billone, M.C., 1980. A study of UO2 wafer fuel for very high power research reactors. Argonne National Laboratory, 9700 South Cass Avenue, Argonne, Illinois 60439. Hussein, H.M., Amin, E.H., Sakr, A.M., 2011. Effect of core configurations on burn up calculations for MTR type reactors. In: Proceedings of the 8th Conference on Nuclear and Particle Physics, Hurghada, Egypt. International Atomic Energy Agency, 1980. Research reactor core conversion from the use of highly enriched uranium to the use of low enriched uranium fuels guidebook, volume 1: summary. TECDOC-233. IAEA, Vienna. International Atomic Energy Agency, 1992. Research reactor core conversion guidebook, vol. 1: summary. TECDOC-643. IAEA, Vienna. IAEA, 2003a. Status and advances in MOX fuel technology Technical Reports Series No.415. International Atomic Energy Agency, Vienna.

International Atomic Energy Agency, 2003b. Manual for reactor produced radioisotopes TECDOC-1340. IAEA, Vienna. International Atomic Energy Agency, 2005. Thorium fuel cycle – potential benefits and challenges, TECDOC-1450. IAEA, Vienna. Imami, M.M., Roushdy, H., 2002. Thermal neutron flux distribution in ETRR-2 reactor thermal column. Nucl. Technol. Radiat. Protect., 1–2 Khalil, M.Y., Amin, E., Belal, M.G., 2005. ETRR-2 in core management strategy. In: 9th International Topical Meeting on Research Reactor Fuel Management, Budapest, Hungary. Khater, H., Elmaty, T.A., Elmorshdy, E.E., 2006. Thermal-hydraulic modeling of reactivity accidents in MTR reactors. Nucl. Technol. Radiat. Protect. 2. Nagy, M.E., Elafify, M.M., Ashraf, M.R., Enany, A.M.R., 2004. Parametric study of reactivity changes in Egypt second research reactor (ETRR-2). Alexandria Eng. J. 43 (1), 11–19. NEA, 2010. The supply of medical radioisotopes: review of potential molybdenum 99/technetium-99m production technologies. Odoi, H.C., Akaho, E.H.K., Jonah, A.S., Abrefah, G.R., Ibrahim, Y.V., 2014. Study of criticality safety and neutronic performance for a 348 fuel-pin Ghana research reactor-1 LEU core using MCNP code. World J. Nucl. Sci. Technol. 4, 46–52. SAR, 1997. Safety analysis report of the ETRR-2, 0767–5325-3IBLI-001-IO. SAR, 1993. Safety Analysis Report for the Syrian Miniature Neutron Source Reactor. China Institute of Atomic Energy, China. Shaaban, I., Albarhoum, M., 2015. Minimizing MTR reactor uranium load with the use of MOX fuel by employing ORIGEN-S and MCNP4C codes. Ann. Nucl. Energy 83, 34–40. Shaat, M.K., 2010. Utilization of ETRR-2 and collaboration, IAEA-TM-38728. Simnad, M., 1962. Nuclear Reactor Materials and Fuels. University of California, San Diego, USA. Westren, D., 2007. Why faster is better – on minor actinide transmutation in hard neutron spectra (Doctoral Thesis), Stockholm, Sweden. Yeo, S., 2013. UO2-SIC composite reactor fuels with enhanced thermal and mechanical properties prepared by spark plasma sintering, Doctor of Philosophy. University of Florida.