Performance of the MTR core with MOX fuel using the MCNP4C2 code

Performance of the MTR core with MOX fuel using the MCNP4C2 code

Applied Radiation and Isotopes 114 (2016) 92–103 Contents lists available at ScienceDirect Applied Radiation and Isotopes journal homepage: www.else...

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Applied Radiation and Isotopes 114 (2016) 92–103

Contents lists available at ScienceDirect

Applied Radiation and Isotopes journal homepage: www.elsevier.com/locate/apradiso

Performance of the MTR core with MOX fuel using the MCNP4C2 code Ismail Shaaban n, Mohamad Albarhoum Nuclear Engineering Dept., Atomic Energy Commission, P.O.Box 6091, Damascus, Syria

H I G H L I G H T S

 Re-cycling of the ETRR-2 reactor by MOX fuel.  Increase the number of the neutronic traps from one neutronic trap to three neutronic trap.  Calculation of the criticality safety and neutronic parameters of the ETRR-2 reactor for the U3O8-Al original fuel and the MOX fuel.

art ic l e i nf o

a b s t r a c t

Article history: Received 23 November 2015 Received in revised form 10 May 2016 Accepted 10 May 2016 Available online 11 May 2016

The MCNP4C2 code was used to simulate the MTR-22 MW research reactor and perform the neutronic analysis for a new fuel namely: a MOX (U3O8&PuO2) fuel dispersed in an Al matrix for One Neutronic Trap (ONT) and Three Neutronic Traps (TNTs) in its core. Its new characteristics were compared to its original characteristics based on the U3O8-Al fuel. Experimental data for the neutronic parameters including criticality relative to the MTR-22 MW reactor for the original U3O8-Al fuel at nominal power were used to validate the calculated values and were found acceptable. The achieved results seem to confirm that the use of MOX fuel in the MTR-22 MW will not degrade the safe operational conditions of the reactor. In addition, the use of MOX fuel in the MTR-22 MW core leads to reduce the uranium fuel enrichment with 235U and the amount of loaded 235U in the core by about 34.84% and 15.21% for the ONT and TNTs cases, respectively. & 2016 Elsevier Ltd. All rights reserved.

Keywords: MTR-22 MW reactor MOX fuel Criticality safety Neutronic parameters MCNP4C2 code

1. Introduction In the present days, research reactors are used for many purposes such as nuclear science research, technology development, reactor services, neutron radiography, radioisotope production, Neutron Transmutation Doping (NTD) of silicon, gem coloring or silicon doping and boron neutron capture therapy. In addition, research reactors provide important training to scientists/engineers and help in generating a pool of skilled human resource for future requirements (IAEA, 2003; Kaushal, 2005; Liu et al., 2004; Shaaban and Albarhoum, 2015a,b). The fuel elements that have mainly been used in research reactors include; uranium metal, uranium alloys, dispersions in aluminum (UAlx,U3O8, U3Si2), dispersions in graphite and stainless steel (UO2) and UZr-hydride (IAEA, 1992; SAR, 1993; Bretscher and Matos, 1996). Nowadays, most of the researches which are carried out in the world concentrates on theuse of MOX fuels (UO2&PuO2) in the advanced Light Water Reactors (LWRs) and the use of transuranic nuclides (Neptunium, Americium and Curium) in their fuels. This n

Corresponding author. E-mail address: pscientifi[email protected] (I. Shaaban).

http://dx.doi.org/10.1016/j.apradiso.2016.05.009 0969-8043/& 2016 Elsevier Ltd. All rights reserved.

techniques help to: reduce the volume of the spent fuel for storage and for non-proliferation issues, reduce the risk of these elements in the environment, reduce the enrichment in 235U of the fuel material and increase the proliferation resistance 238Pu/Pu ratio (Rahn et al., 1984; IAEA, 2003). The MOX fuel is widely used in LWRs but until now using MOX fuel in Research Reactors (RRs) does not seem to have been investigated. Therefore, the data base about using MOX fuel in RRs are very small or even absent. In addition to the aforementioned reasons for investigating the use of MOX fuels in RRs there may be other reasons of which may be mentioned: 1. To increase plutonium isotopes utilization contributing to extend the spreading of RRs which used only MOX fuel and MOX with minor actinides in future, 2. To reduce the volume and the risk of the spent fuel operation up to the repositories, 3. To help designing new RRs using only MOX fuels. Therefore, the objective of this paper is to design a MOX-Al ((U3O8&PuO2)-Al core of the MTR-22 MW (the Egyptian Second Research Reactor (ETRR-2) as example) reactor with ONT and TNTs,

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with similar operational capabilities as the original U3O8-Al core and without changing the dimensions or the components or even the safe operational conditions of the reactor core. The MCNP4C2code (Briesmeister, 2000) was used to model the ETRR-2 and calculate all the neutronic parameters of the ETRR-2 reactor for both the U3O8-Al original and the MOX fuels, whereas, the PECOM system (Shaaban and Albarhoum, 2015a), an interface program between MCNP4C2 and ORIGEN-S was used to calculate the reactivity of the reactor as a function of time.

2. Methodology 2.1. The ETRR-2 reactor The ETRR-2 is a Material Testing Reactor (MTR). It is an open pool research reactor which uses low enriched MTR fuel elements (19.70% enrichment). It was commissioned in 1997. It is cooled and moderated with light water and reflected by beryllium(36 blocks). The reactor power is 22.0 MW with high thermal neutron flux irradiation positions (41014 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, FE, Fuel Plate (FP), absorber material, active zone dimensions and water gap between plates of the ETRR-2 reactor for the U3O8-Al fuel are given in Table 1, Table 2, Table 3 and Table 4 (Imami and Roushdy, 2002; Nagy et al., 2004; Khater et al., 2006; Gaheen, 2010). The reactor is controlled by two control systems; the First Shut-down System (FSS) which consists of 6 control rods 4 of which for reactor control and 2 for the safety of the reactor, while the Second Shut-down System (SSS) is formed of a solution of gadolinium nitrate to be injected when required to shut-down the reactor in emergency conditions as shown in Fig. 1 (Hussein et al., 2011). 2.2. The Monte Carlo model of the ETRR-2reactor with ONT and TNTs The ETRR-2 reactor was simulated using the MCNP4C2 code (Shaaban and Albarhoum, 2015a,b). The cross-section of the ETRR2 core configuration is shown in Fig. 2. As shown in Fig. 2 the ETRR-2 core consists of 29 positions for FEs and one central NT is located near the center of the core. The central NT in the ETRR-2 reactor is used to perform Neutron Activation Analysis (NAA), for radioisotope production (e. g., 131I, 125I, 32P, 151Cr, 192Ir,60Co) and for other scientific applications in Materials Science. In the center of the NT the Central Irradiation Box (CIB) is located and used to irradiate various samples. In this study, the ETRR-2 core was modified to allow for additional scientific applications of the reactor. In this case two NTs were placed inside the ETRR-2 core for the purpose of 99Mo production replacing two FEs to produce 99Tc and other radioisotopes, and the position of the central NT is not changed at all as shown in Table 1 Composition of the fuel in a typical ETRR-2 reactor. Parameter

235

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

Weight percentage % SFE

FE Type 1

FE Type 2

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

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Fig. 3. The new ETRR-2 core consists of 27 FEs and TNTs located in the reactor core. The dimensions of the new NTs are the same as the dimensions of the replaced FEs. The ETRR-2 reactor was resimulated with ONT and TNTs, and with MOX (U3O8&PuO2) fuel without changing dimensions of the ETRR-2 core or any component (See Table 2). The nuclear data for the fissile and the nonfissile materials such as: the fuel, clad, coolant, moderator, control rod, and the reflector were taken from the ENDF/B-VI nuclear data libraries, and the thermal particle scattering S(α,β) was applied to treat the thermal scattering in both beryllium and hydrogen of the moderated water. In the simulation the following specifications of the ETRR-2 core were used: 1. The fuel composition shown in Table 2, 2. A plutonium isotopes composition taken from Table 5 (Breza et al., 2008), 3. A U3O8and PuO2as fuel material dispersed in an Al matrix, 4. A fuel element as shown in Fig. 4, 5. A fuel plate as shown in Fig. 5, 6. An Al fuel clad (Al-alloy 6061), 7. An Al frame (Al-alloy 6061) for the fuel element, 8. A de-mineralized light water as coolant.

3. Replacement of the U3O8-Al fuel by the MOX fuel in the ETRR-2 reactor The U3O8-Al Fuel Plates (FPs) in the FEs were replaced by MOX fuel plates (a kind of fuel dispersed in an Al matrix without technological miscibility problems (Hagenauer et al., 2003) without changing the dimensions of the FPs whether in the FE, or in the control plates. The general characteristics of the fuel material, the fuel element, the control rod material, the active zone dimensions and the water gap between plates of the ETRR-2 reactor for the U3O8-Al original fuel and the MOX fuel for the ONT and the TNTs are given in Table 2, Table 3 and Table 4.In the modified core only the absorber material in the control plates was changed from Ag-In-Cd material to the B4C material without changing neither the dimensions nor the positions (See Table 2). The replacement of the fuel material in the ETRR-2 reactor shall, hopefully, preserve both the operational capabilities and the safe operational conditions (criticality safety conditions and neutronic parameters) of the ETRR-2 reactor as the U3O8-Al original fuel. This will be discussed in the next paragraphs.

4. Calculation of the criticality safety and power distribution plus the neutronic parameters of the ETRR-2 reactor before and after replacement of the U3O8-Al original fuel by the MOX fuel 4.1. The criticality safety calculations The KCODE criticality source card (Briesmeister, 2000) was used in the input file of the ETRR-2 reactor to be executed by the MCNP4C2 code to calculate: 1. The effective multiplication factor (keff) of the 1/98 core (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) (See Fig. 2) and the current core (See Fig. 6) and then the excess core reactivity using the equation (Duderstadt and Hamilton, 1976).

ρ = ( k eff − 1)/k eff

(1)

2. The Shutdown Margin (SM) of the FSS, the Shut Down Margin of the FSS with Single Failure (SM of the FSS with SF), the

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Table 2 General characteristics of the fuel material and the fuel element of the ETRR-2 reactor for the U3O8-Al original fuel and the MOX fuel for the ONT and the TNTs. Parameter

Fuel material

Fuel meat Weight percentage (%) Enrichment 235U(%) Total core loading 235U (g) Saved mass of 235U (g) Density of fuel (g/cm3) Fuel element Number of the FE Number of fuel plates in the FE Dimensions of the fuel plate (cm) (length  width  thickness)

U3O8-Al original fuel with ONT

MOX fuel with ONT

MOX fuel with TNTs

U3O8-Al 100 19.70 6944.5 See Table 1

U3O8-Al 97.20 7.10 4523.19 2421.31 4.90

U3O8-Al 96.20 9.70 5886.29 1058.21 4.90

29 19 80  6.40  0.07

29 19 80  6.40  0.07

PuO2 2.80

PuO2 3.80

27 19 80  6.40  0.07

Table 3 General characteristics of the absorber material of the ETRR-2 reactor for the U3O8-Al original fuel and the MOX fuel for the ONT and the TNTs. Absorber material Composition

Ag-In-Cd for the original fuel Weight percentage (%) Ag 15 In 80 Cd 5 Density (g/cm3) 10.18

B4C for the MOX fuel with ONT Weight percentage (%) 10 B 38.4 11 B 41.6 C 20 2.52

B4C for the MOX fuel with TNTs Weight Percentage (%) 10 B 40 11 B 40 C 20 2.52

Table 4 General characteristics of the active zone dimensions and the water gap between plates of the ETRR-2 reactor for the U3O8-Al original fuel and the MOX fuel for the ONT and the TNTs. Fig. 1. A cross section of the current ETRR-2 core in the plane X–Y with the FSS and the SSS using theMCNP4C2 code.

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) Cladding material

U3O8-Al original fuel with ONT 80 80 88

MOX fuel with ONT 80 80 88

MOX fuel with TNTs 80 80 88

8.1  8.1

8.1  8.1

8.1  8.1

0.15 0.07 6.40 0.50 8.00 8.00

0.15 0.07 6.40 0.50 8.00 8.00

0.15 0.07 6.40 0.50 8.00 8.00

7

7

7

Al- 6061

Al- 6061

Al- 6061

0.27 0.39

0.27 0.39

Water gap between plates of single fuel element (cm) 0.27 of different fuel element 0.39 (cm)

Reactivity Safety Factor (RSF) with RSF meaning control rod worth/core excess reactivity, and the Control Rod Worth (CRW) of both the 1/98 core and the current one. Table 6 and Table 7 shows the measured and calculated values of the criticality safety parameters of the 1/98 ETRR-2 core and the current ETRR-2 core for the U3O8-Al original fuel and ONT, the current ETRR-2 core for the MOX fuel and ONT and TNTs. The values of the SM and the SM of the FSS with SF are calculated when the FSS is fully inserted and when the FSS is fully inserted

but failing a single plate, respectively. Whereas, the calculated value of the CRW was obtained using the relation:

CRW = ρ + SM

(2)

4.2. Calculation of the beta effective βeff, for the current ETRR-2 core fueled by both the U3O8-Al and the MOX fuels The effective delayed neutron fraction, βeff, is given by the number of neutrons produced in the reactions induced by delayed neutrons, divided by the total number of neutrons produced. To estimate the value of the effective delayed neutron fraction βeff using the MCNP4C2 code the MCNP4C2 code was used to run the input file of the ETRR-2 reactor with the "TOTNU" card with "NO" as the only entry which turns off delayed neutrons in the k-code mode, producing a final k-eigenvalue corresponding to kp. The value of the βeff was calculated using the relation (Westren, 2007).

βeff = 1– ( kp/k1 )

(3)

k1 = kp + k deff

(4)

Where kp denotes the prompt neutron contribution to keff, kdeff the contribution of delayed neutrons. The MCNP4C2 code results of the βeff for all the studied cases are given in Table 7. 4.3. Calculation of the neutronic parameters The reactor, as aforementioned, is designed to be used for NAA,

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Fig. 2. A cross section of the 1/98 ETRR-2 core in the plane X–Y using the MCNP4C2 code.

Fig. 3. A cross section of the ETRR-2 core in the plane X–Y with the TNTs using theMCNP4C2 code.

radioisotope production (e.g., 131I, 125I, 32P, 60Co, 151Cr, 192Ir,…etc.), materials science, training and education for new operators, neutron radiography and boron capture therapy (SAR,1997). These applications depend on the values of the thermal neutron flux in the NTs. Therefore, the values of the Thermal Neutron Flux (TNF) in the modified ETRR-2 core with MOX fuel for the ONT and the TNTs should have the same order of values of the thermal neutron flux for the U3O8-Al original fuel. The aforementioned neutronic parameters include the following parameters: a. The Average TNF (ATNF) in the central NT which is located in site 1 as shown in Fig. 6, and in the CIB which is located in the center of the central NT using the U3O8-Al original fuel and the MOX fuel for ONT. b. The TNF in the CIB located in the central NTs in sites 1, 2 and 3 (See Fig. 3) using the MOX fuel for the TNTs, c. The Average TNF (ATNF) in the NTs and in the Be reflector using the U3O8-Al original fuel and the MOX fuel for the ONT and the TNTs. To calculate these neutronic parameters the input file of the ETRR-2 reactor was run by the MCNP4C2 code using the F4 tally, the FS, the SD and the FM cards in the MCNP4C2 code (Briesmeister, 2000). In the MCNP4C2 code, tallies are normalized per source particle except in criticality calculations. The flux tally will then be in units of neutrons/(cm2 source particle), and this will give the correct spectral shape of the neutron scalar flux but not the correct magnitude of the flux. The normalized flux can be calculated using the average number of neutrons produced per fission ν ̃, the reactor operating power (P in MW) and the MCNP4C2 flux tally

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Table 5 Composition isotopes.

of

the

plutonium

Parameter Weight percentage % 238

Pu Pu Pu 241 Pu 242 Pu 239 240

2.78 55.46 23.20 12.16 6.4

normalized per source neutron:

⎞ ⎛ neutrons ⎞ ⎛ 1MeV ⎟× ⎜ φ=φMCNP × P × ν ̃ ⎜ ⎟× ⎝ seconds ⎠ ⎝ 1. 6022 × 10−13Joules ⎠ ⎛ fission ⎞ ⎟ ⎜ ⎝ 200MeV ⎠

(5)

The power in Eq. (5) is in the units of Joules/seconds, to give the normalized flux in the correct unit of neutrons/cm2 s (Makgopa, 2009).Therefore, to produce P watts of power, one needs 3.1203  1010P fissions per second. This produces 3.1203  1010x P x ν̃ neutrons/s, which is the source strength for this power level. The source strength(normalization) should be written in the tally on the FM card to calculate the TNF in the desired cell (Briesmeister, 2000). The MCNP4C2 code results for the ATNF in the NTs, and in the Be reflector, and the TNF in the CIB are listed in Table 8 and Table 9 of the current ETRR-2 core fueled by the U3O8-Al original fuel and re-fueled by the MOX fuel for the ONT and the TNTs, and for both cases:

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Table 7 MCNP4C2 code results of the safety criticality parameters current ETRR-2 core fueled by the U3O8-Al original fuel and the MOX fuel for the ONT and the TNTs for the all control plates are full withdrawn out from the ETRR-2 core. Current ETRR-2 core fueled by U3O8-Al original fuel for the ONT keff ¼ 1.07697 70.00038 (See Fig. 6) Parameter Excess of Reactivity ($) SM of the FSS ($) SM of the FSS with SF ($) CRW ($) βeff (pcm)

Calculated (Using MCNP4C2 code) 9.3767 0.007 15.1417 0.036 8.429 7 0.036 24.6707 0.043 7237 7

Current ETRR-2 core re-fueled by the MOX Parameter For the ONT (See Fig. 6) keff ¼ 1.07230 70.00042 Excess of Reactivity ($) 8.990 7 0.007 SM of the FSS ($) 15.5197 0.040 SM of the FSS with SF ($) 8.329 7 0.040 CRW ($) 24.2097 0.043 βeff (pcm) 7627 6 Parameter For the TNTs (See Fig. 3) keff ¼ 1.0742707 0.00044 Excess of Reactivity ($) 9.218 7 0.007 SM of the FSS ($) 14.405 7 0.038 SM of the FSS with SF ($) 8.456 7 0.038 CRW ($) 23.622 7 0.043 βeff (pcm) 7797 6 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). Fig. 4. A schematic horizontal cross section of the fuel element used in the ETRR-2 reactor using the MCNP4C2 code.

Fig. 5. Dimensions of the fuel meat, cladding and water channel in the fuel plate using the MCNP4C2 code.

Table 6 Experimental and the MCNP4C2 code results of the safety criticality parameters of the 1/98 ETRR-2 core fueled by the U3O8-Al original fuel for the ONT. Type

Core 1/98 of the ETRR-2 fueled by U3O8-Al original fuel for the ONT (See Fig. 2) Measureda

Excess of Re9.1 activity ($) SM of the FSS ($) 15.2 SM of the FSS 8.7 with SF($)

Calculated (Using MCNP4C2 code)

Calculated/ Measured

8.9647 0.049

0.985

15.8497 0.029 8.929 7 0.029

1.042 1.026

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).

1. All Control Plates are fully withdrawn Out (ACRO) of the current ETRR-2 core fueled by theU3O8-Al original fuel for the ONT. In this case the keff ¼ 1.07697 70.00038 (See Table 6), 2. The current ETRR-2 core re-fueled by the MOX fuel for the ONT and the TNTs (See Table 9), and for two cases:

Fig. 6. A cross section of the current ETRR-2 core in the plane X–Y with the ONT using the MCNP4C2 code.

- ACRO of the ETRR-2 core. Where the keff ¼ 1.07230 70.00042 and the keff ¼1.0742707 0.00044for the ONT and TNTs, respectively,

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Table 8 Reference and the MCNP4C2 code results of the TNF, βeff and the Apf for the current ETRR-2 core fueled by the U3O8-Al original fuel for the ONT for the all control plates are full withdrawn out from the ETRR-2 core. Current ETRR-2 core fueled by the U3O8-Al original fuel for the ONT (See Fig. 6) and the keff ¼ 1.076977 0.00038 Parameter 2

14

TNF in the CIB in site 1 (n/cm s)  10 ATNF in the NT in site 1 (n/cm2 s)  1014 ATNF in the Be reflector (n/cm2 s)  1014 βeff (pcm) Apf

Reference values with all control plates are out 4.230 a 2.700 b 1.000b 740b 1.355c

Calculated (Using MCNP4C2 code)with all control plates are out 4.215 70.017 2.7007 0.007 0.988 70.014 7237 7 1.428

Calculated/Reference 0.996 1.000 0.988 0.977 1.054

a

Reference value was taken from Ref. 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; http://archive. is/Mnf73). c Reference value was taken from reference (Mohamed, 2011. Analysis of Reactivity Accidents in MTR for Various Protection System Parameters and Core Conditions, Thesis, Master of science. Faculty of Engineering, Alexandria University). b

Table 9 MCNP4C2code results of the TNF for the current ETRR-2 core re-fueled by the MOX fuel for the ONT and the TNTs, and for the both cases: The all control plates are full withdrawn out from the ETRR-2 core, where the keff ¼1.07230 7 0.00042 and the keff ¼ 1.074270 7 0.00044, and the criticality case, where the keff ¼ 1.000757 0.00044and the keff ¼1.00092 7 0.00044. Current ETRR-2core re-fueled by the MOX fuel for the ONT (See Fig. 6)

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

Calculated value (Using MCNP4C2 code ) keff ¼1.00075 keff ¼ 1.07230 4.1827 0.019 3.940 7 0.020 2.6787 0.010

2.898 7 0.008

0.983 7 0.013

1.038 7 0.017

Current ETRR-2core re-fueled by the MOX fuel for the TNTs (See Fig. 3) Parameter Calculated value (Using MCNP4C2 code) keff ¼ 1.07427 keff ¼1.00092 TNF in the CIB in site 1 3.612 70.020 3.35370.020 2 14 (n/cm s)  10 3.3317 0.021 2.5067 0.023 TNF in the CIB in site 2 (n/cm2 s)  1014 3.223 7 0.021 2.542 7 0.024 TNF in the CIB in site 3 (n/cm2 s)  1014 2.6337 0.007 2.450 7 0.008 ATNF in the NT in site 1 (n/cm2 s)  1014 2.6147 0.008 1.845 7 0.009 ATNF in the NT in site 2 (n/cm2 s)  1014 2.592 7 0.008 1.8707 0.009 ATNF in the NT in site 3 (n/cm2 s)  1014 0.986 7 0.011 1.0147 0.013 ATNF in the Be reflector (n/cm2 s)  1014

- The criticality case. Where the keff ¼ 1.00038 70.00039 and the keff ¼1.00092 70.00044, for the ONT and TNTs, respectively.

The calculated values of the TNF for the current ETRR-2 core fueled by the U3O8-Al original fuel and re-fueled by the MOX fuel for the ONT and the ACRO case are also given in the Table 10. The neutronics calculations were performed using three energy groups as: o0.625 eV for thermal neutrons, (0.625 eV to 5.53 keV) for epithermal neutrons and up to 20 MeV for fast neutrons. The validity of these calculations is verified by comparing them with the reference values as shown in Table 8 too. 4.4. Calculation of the power distribution in the FEs The total power production in the FEs depends on the neutron flux and the properties of all nuclides in the fuel material. The Power Distribution (PD) in the FEs is studied to determine whether the replacement of the U3O8-Al original fuel by a MOX fuel causes any changes in the value of the power peaking factors (IAEA, 1980).To calculate the values of the power distribution in the FEs of the ETRR-2 reactor. The following equation (Lewis, 2008) was used:

Power in the FEs=

ρ . σf . V . φ h

(6)

Where ρ – is the atom density of the fuel material, V- is the volume of the fuel material in the FE, σf - is the total fission cross section of the fuel material, h¼ 3.1203  1010 - is the number of fissions generating 1 watt per second, φ - is the actual total neutron flux in the FE its value being calculated from Eq. (5). Now, the F4tally, the FS, the SD and the FM cards (Briesmeister, 2000) were used in the input file of the ETRR-2 reactor as: F4: n $ This card allows to estimate the track-length of the neutron flux in the desired cell. FS $ This card allows to subdivide a cell or a surface into segments for tallying purposes. SD $ This card allows to divide a volume or area into segments

Table 10 MCNP4C2code results of the TNF for the current ETRR-2 core fueled by theU3O8-Al original fuel and re-fueled by the MOX fuel for the ONT and the all control plates are full withdrawn out from the ETRR-2 core, where the keff ¼1.07230 7 0.00042 and the keff ¼ 1.0742707 0.00044. Current ETRR-2core fueled by the U3O8-Al original fuel and the MOX fuel for the ONT (See Fig. 6) Calculated (Using MCNP4C2 code)with all control plates are out Parameter

U3O8-Al original fuel keff ¼ 1.07697

MOX fuel keff ¼ 1.07230

Original fuel /MOX fuel

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

4.2157 0.017 2.7007 0.007 0.988 7 0.014

4.1827 0.019 2.6787 0.010 0.983 7 0.013

1.008 1.008 1.005

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Black color - Number of the FE in the ETRR-2 core, where the number 16 denotes to the NT. Red color - Calculated value of the power distribution for U3O8-Al original fuel and ACRO case. Blue color - Calculated value of the power distribution for U3O8-Al original fuel and criticality case. - Control plate Fig. 7. Calculated values of the power distribution in the fuel elements (MW) of the current ETRR-2 core with ONT for the U3O8-Al original fuel and the ACRO and criticality cases.

Black color - Number of the FE in the ETRR-2 core, where the number 16 denotes to the NT. Red color - Calculated value of the power distribution for MOX fuel and ACRO case. Blue color - Calculated value of the power distribution for MOX fuel and criticality case. Fig. 8. Calculated values of the power distribution in the fuel elements (MW) of the current ETRR-2 core with ONT for the MOX fuel and the ACRO and criticality cases.

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for tallying purposes. E $ energy bins in MeV. FM4 C m -6. Where:

C=

A. ρ . V h

m – is the number of the fuel material in the input file. -6 – denotes to the total fission cross section in units of barns in the MCNP4C2 code. TheMCNP4C2 code calculations of the PD (MW) in the FEs of the current ETRR-2 core (See Fig. 6) for the U3O8-Al original fuel and the MOX fuel are tabulated in Figs. 7 and 8 for the ONT, respectively, and for both cases: 1. The ACRO and criticality case (or keff ¼1.076977 0.00038 and keff ¼1.00038 70.00044) for the U3O8-Al original fuel (See Fig. 7). 2. The ACRO and the criticality case (or keff ¼1.07230 70.00042and keff ¼1.000757 0.00039) for the MOX fuel (See Fig. 8). From Figs. 7 and 8 the following considerations could be made: - The hottest FE in the reactor core is located in positions 24 and 10 (See Fig. 7) of the U3O8-Al original fuel for the ACRO and criticality cases. -. The hottest FE in the reactor core is located in positions 24 and 15 (See Fig. 8) of the MOX fuel for the ACRO and criticality cases. The values of the hottest fuel element power peaking factor (Ppf) and the axial peaking factor (Apf) of the hottest fuel element are calculated as (IAEA,1980): 1. The Ppf is defined as the ratio between the maximum power released from one fuel element (PFE)max and the average power per element in the reactor core (Pcore). 2. The Apf is defined as the ratio between the axial peak of the hottest fuel element and the average ratio for a typical element. In this case the hottest FE was divided into 12 segments along the FE height, then the power was calculated in each segment using the MCNP4C2 code. Table 11 Calculated values of the power peaking factors Ppf and Apf of the hottest fuel element for the ETRR-2 reactor for the various core configurations. Parameter

Ppf

The MCNP4C2 code calculations of the Ppf and Apf of the U3O8-Al original fuel and the MOX fuel for the ONT and the TNTs are given in Table 11. Where as, the reference and calculated value of the Apf of the current ETRR-2 core fueled by the U3O8-Al original fuel for the ONT is given in the Table 8. The criticality safety, power distribution and the neutronic calculations were performed by executing the input file of the ETRR-2 reactor by the MCNP4C2 code using three hundred million neutron histories (106 neutron particle source histories per cycle and 300 cycles with an initial criticality keff guess of 1 and twenty passive cycles before the active ones begin) and using all fuel elements as fission source points, where the fission source is located in the middle of each FE.

5. Fuel burn-up calculation for the ETRR-2 reactor core The primary objective of the modified core nuclear design and management analysis is the verification of the safe operational conditions of the core. All the parameters required to verify the accomplishment of the design limits should be calculated. The excess core reactivity, the SM of the FSS, the SM of the FSS with SF, the SRF, the CRW and the reactivity at the end of the fuel cycle. The values of these parameters should be within the limiting conditions specified in the operational limits and the safe operation, where the reference values of these parameters are summarized in Table 12. From Table 12 it can be seen that the reactivity at the end of the fuel cycle is greater than 1.33$. Therefore, the MCNP4C2 and the ORIGEN-S codes (a depletion code (Hermann and Westfall, 1998)), and the PFCOM system (Shaaban and Albarhoum, 2015a,b) were used to calculate the excess core reactivity at the end of the fuel cycle (19 day) for the both the U3O8-Al original fuel and the MOX fuel. In the calculation: The actinides 233U, 234U, 235U, 236U, 238U, 237Np, 239Np, 238 Pu,239Pu,240Pu,241Pu,242Pu, 232Th,. a. The major fission products: 91Zr, 115In, 110Cd, 111Cd, 112Cd, 113Cd, 114 Cd, 116Cd, 154Gd, 155Gd, 156Gd, 157Gd, 158Gd, 160Gd, 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, 151 Eu, 153Eu, 154Eu, 155Eu, b. The light elements 1H, 6Li, 13C, 16O and the 27Al, were taken in the calculation of the reactivity at the end of the fuel cycle. The obtained results for both the U3O8-Al original fuel and the MOX fuel are given in Table 12 for the ONT and the TNTs.

Apf

Current ETRR-2 core fueled by the U3O8-Al original fuel for the ONT (See Fig. 6) keff ¼1.076977 0.00038 1.432 70.002 1.428 7 0.010 keff ¼1.00038 7 0.00044 1.4337 0.004 1.6707 0.014 Current ETRR-2 core re-fueled by the MOX fuel for the ONT (See Fig. 6) 1.4337 0.004 1.393 7 0.012 keff ¼1.07230 7 0.00042 keff ¼1.00075 70.00044 1.538 70.004 1.718 70.013 Current ETRR-2 core re-fueled by the MOX fuel for the TNTs (See Fig. 3) 1.4617 0.004 1.4127 0.011 keff ¼1.074277 0.00044 keff ¼1.00092 7 0.00044 1.650 70.007 2.001 70.011 Ratio of the calculated values of the U3O8-Al original fuel to the calculated values of the MOX fuel of the Ppf and Apf parameters For the ONT (See Fig. 6) Parameter

99

Ppf (U3O8-Al)/MOX

All control plates full out of the ETRR- 0.999 2 core Criticality case 0.932

Apf (U3O8-Al)/MOX 1.025 0.972

6. Results and discussion 6.1. Validation of theMCNP4C2 model of the ETRR-2 reactor The MCNP4C2 code world wide was used to model the ETRR-2 reactor. It has the capability to model and treat different complicated geometries in 3-D and also to simulate the transport behavior of different particles and nuclear interaction processes. Good and accurate modeling of the different zones and diverse geometries of the ETRR-2 reactor is important for realizing good neutronic and criticality calculations, particle transport simulation, and physics analysis. For these reasons, the MCNP4C2 code was employed to simulate the ETRR-2 reactor and to perform the criticality and neutronic analysis. 6.1.1. Results of the MCNP4C2 calculations for the criticality

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Table 12 Safety limits (acceptance values) of the criticality parameters of the ETRR-2 core. Parameter

Safety limitsa (acceptance values)

SM of the FSS ($) Z 4.00 SM of FSS with SF ($) Z 1.33 PD (MW) 22.0 RSF Z 1.50 End of cycle reactivity ($) 4 1.33 Parameter SM of the FSS ($) Z 4.00 SM of FSS with SF ($) Z 1.33 PD (MW) 22.0 RSF Z 1.50 End of cycle reactivity ($) 4 1.33 a

Calculated values of the current core with U3O8-Al original fuel and one NT (See Fig. 6)

15.1417 0.036 8.429 7 0.036 21.2017 0.003 2.6317 0.004 3.346 7 0.029 Calculated values of the MOX fuel for ONT (See Fig. 6) 15.5197 0.040 8.329 7 0.040 21.3017 0.004 2.7267 0.005 3.3557 0.003

Calculated values of the MOX fuel for TNT (See Fig. 3) 14.405 70.038 8.456 7 0.038 21.802 7 0.004 2.563 7 0.004 4.721 70.004

Acceptance values were taken from the reference (Nagy et al., 2004; Gaheen, 2010; Hussein et al., 2011).

parameters The results of the MCNP4C2 code for the criticality parameters such as: the excess core reactivity, the SM and the SM of the FSS with SF of the 1/98 ETRR-2 core (See Fig. 2) fueled by U3O8-Al original fuel as compared with experimental measurements are tabulated in Table 4. From Table 6 it can be seen that the MCNP4C2 calculations differ from the measured ones by not more than 4.2% on average. 6.1.2. Results of the MCNP4C2 calculations for the neutron flux and the axial power peaking factor (Apf) Table 8 shows that the maximum relative error between theMCNP4C2 code calculations for the TNF in the CIB, the ATNF in the NT and the ATNF in the Be reflector for the current ETRR-2 core (See Fig. 6) and the corresponding reference values of those parameters is 1.2%. The relative error is calculated as: (reference value – calculated value)/reference value. It results also from Table 8 that the relative error in the calculated value of the βeff and the Apf differs from the reference value by 2.3% and 5.11%, respectively as well. The good agreement between the MCNP4C2 code calculations and the reference values confirms the fact that the model is good and that Monte Carlo method can be effectively employed to model the ETRR-2 reactor. Table 2 shows that: 1. The replacement of the U3O8-Al original fuel by the MOX (Note that the sum of the U3O8-Al and the MOX fuels percentages make 100%) fuel reduces the enrichment with 235U from 19.7– 7.1% and 9.7% and leads to save 2421.31 g and 1058.21 g of 235U for the ONT and the TNTs, respectively, 2. Plutonium is burnt in the reactor, 3. The MOX fuel could be used to design new research reactors working only on MOX fuels. As can be seen from Table 7 the calculations of the criticality parameters of the current ETRR-2 core (See Fig. 6) using the U3O8-Al original fuel differ from those for the 1/98 ETRR-2 core (See Fig. 2) by 4.39%, 4.47% and 5.60% for the case of the excess core reactivity, the SM and the SM of the FSS with SF, respectively. This difference is due to the change in the structure of the ETRR-2 core where 13 Be cubes were added around the current ETRR-2 core (See Figs. 2 and 6). As known the Be cubes will increase the value of the core reactivity as a result of reflecting neutrons into it. 6.2. Results of the MCNP4C2calculationsfor the safety criticality parameters and the neutron flux for the ETRR-2 reactor re-fueled by

MOX fuel with the ONT and the TNTs 6.2.1. Results of the MCNP4C2 calculations for the safety criticality parameters The MCNP4C2 results for the safety criticality parameters for the current ETRR-2 core (See Fig. 6) re-fueled by the MOX fuel are tabulated in Table 7. It results from Table 7 that the safety criticality parameters (the excess core reactivity, the SM, the SM of the FSS with SF, the CRW and the βeff) for the current ETRR-2 core fueled by the U3O8-Al original fuel are different, from those for the MOX fuel by 4.11%, 2.43%, 1.18%, 1.87% and 2.9%, respectively, and by 1.68, 4.86, 0.32,4.25% and 5.0% for the excess core reactivity, the SM and the SM of the FSS with SF, the CRW and the βeff, respectively for the current ETRR-2 core with the ONT and the TNTs, respectively. These results indicate that the MOX fuel can be safely used to substitute the U3O8-Al original fuel in the ETRR-2 reactor without any negative effect on the safety criticality parameters of the reactor. 6.2.2. Results of the MCNP4C2 calculations for the neutronic parameters The MCNP4C2 results for the neutronic parameters for the current ETRR-2 core re-fueled by the MOX fuel are given in Table 9. Table 9 shows that: 1. The MCNP4C2 code results of the neutronic parameters (the TNF in the CIB, the ATNF in the NT and the ATNF in the Be reflector) for the case ACRO of the modified ETRR-2 core re-fueled by the MOX fuel with the ONT differ by 1.13%, 0.81% and 1.7% from the reference values of the TNF in the CIB, the ATNF in the NT and the ATNF in the Be reflector, respectively. This result indicates that the replacement of the U3O8-Al original fuel by the MOX fuel with the limits tabulated in Table 2, Table 3 and Table 4 dos not have negative effects on the neutronic parameters of the ETRR-2 reactor. This leads to say that all the scientific applications which were available in the ETRR-2 reactor fueled by U3O8-Al original fuel are still available for the ETRR-2 core re-fueled by the MOX fuel for the ONT. Additionally, for the criticality case the TNF in the CIB is reduced by 6.86%whereas the ATNF in the NT is increased by 7.33% 2. The MCNP4C2 code results for the TNF in the CIB and the ATNF in the NT for the modified ETRR-2 core re-fueled by the MOX fuel with the TNTs are: - Higher than the 2.0  1014 n/cm2.s for the ACRO case. This value is sufficient to produce 99Mo and other radioisotopes such as: 14C,32S, 51Cr,60Co, 89Sr, 153Sm,169Yb, 170Tm, 192Ir. (IAEA, 2003; SAR, 1997; Shaaban, 2010; NEA, 2010). This

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result leads to say that the modified ETRR-2 core using this MOX fuel with TNTs allows to enlarge the scientific applications of the ETRR-2 reactor. - Higher than the 2.0  1014 n/cm2 s in the sites 1, 2 and 3 for the CIB, and in the site 1 for the ATNF in the NT for the criticality case. This value is sufficient to produce the radioisotopes which are mentioned above. As well as, there is a depression of the ATNF in the NT for the site 2 and site 3 but it remains of the order of 1014 n/cm2 s that is enough to produce other radioisotopes such as: 90Y, 99Mo, 125I, 131I and 133 Xe (Shaaban, 2010). 6.3. Results of the MCNP4C2 calculations of the Ppf and the Apf for the hottest fuel element From Table 11 it results that the maximum difference between the calculated values of the Ppf and the Apf for the MOX fuel with ONT and TNTs, and the calculated values of the U3O8-Al original fuel (with ONT) is about 2.5% for the ACRO case. This result indicates that the replacement of the U3O8-Al original fuel by the MOX fuel does not create a big change in the value of the power peaking factors. This change practically does not affect the safety parameters of the reactor. For the criticality case for the MOX fuel with the ONT and the TNTs the calculated values of the Ppf and the Apf differ from the calculated values of the same parameters for the U3O8-Al original fuel (with ONT) by the 6.8% and the 2.8%, and the 13.15% and 16.54% for the ONT and the TNTs, respectively.

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Table 13 Cell constants of the four types of the fuel elements. Type

D Σa ν  Σf

The cell constants of the SFE Fast neutron

Epithermal neutron

Thermal neutron

2.548182E þ 00 1.095226E-03 1.677767E-03

9.620817E-01 6.937002E-03 5.730004E-03

3.089652E-01 9.049527E-02 1.605627E-01

The cell constants of the FE type 1 D 2.571559E þ00 1.000202E þ 00 Σa 6.794126E-04 3.559593E-03 ν  Σf 6.144936E-04 2.245341E-03

2.929203E-01 4.743103E-02 6.951096E-02

The cell constants of the FE type 2 D 2.565720E þ 00 9.896892E-01 Σa 7.780154E-04 4.459639E-03 ν  Σf 8.664972E-04 3.111751E-03

2.981268E-01 5.882315E-02 9.368004E-02

The cell constants of the MOX FE D 2.547207Eþ00 9.570598E-01 Σa 1.027231E-03 8.013948E-03 ν  Σf 1.517660E-03 3.023419E-03

2.955008E-01 6.355886E-02 9.957856E-02

Table 14 MCNP4C2code results of the SM of the FSS, the SM of FSS with SF, the CRW and the SRF parameters of the ETRR-2 core re-fueled by MOX fuel for theAg-In-Cd control palteas. Parameter

Current ETRR-2 core re-fueled by the MOX fuel for the ONT Control plates are Ag-In-Cd

6.4. Safety limits of the criticality parameters of the ETRR-2 core The MCNP4C2 code results for the criticality parameters (the core excess reactivity ρ at the end of the fuel cycle, the SM of the FSS, the SM of the FSS with SF, the RSF and the PD in the reactor core) for the modified ETRR-2 core re-fueled by the MOX fuel for the ONT and the TNTs should be within the safe operational limits of the ETRR-2 reactor. These conditions are summarized in Table 12. From Table 12, it can be noticeably seen that the calculated values for the criticality parameters for the modified ETRR-2 core re-fueled by the MOX fuel with the ONT and the TNTs are in good agreement with the safe operational conditions of the ETRR-2 reactor. This result indicates that there is no negative effect on the safety operational conditions of the reactor with the MOX fuel. 6.5. Results of the MCNP4C2 calculations for the power distribution in the fuel elements From Figs. 7 and 8 it can be seen that: 1. The PD in the ETRR-2 core depends on the fuel composition in the FE and the position of the FE in the reactor core. Therefore, the SFEs have the highest values (See Fig. 7, Sites 1, 6, 12, 24, 25, 29 and 30) for both cases: ACRO and the criticality case. This would be attributed to the SFE that has the highest fission production rate ν  Σf in comparison with all the other fuels used (See Table 13). 2. The MOX FE has a lower value for the fission production rate ν  Σf than the SFE (See Table 13). Therefore, the MOX FEs have lower values (See Fig. 7, Sites 1, 6, 12, 24, 25, 29 and 30) than the SFEs for the ACRO case. 3. The FEs Type 1 and FEs Type 2 have lower values (See Fig. 7, Sites from 2 to 5, from 7 to 11, from 13 to 18 with exception of the site 16, from 19 to 23 and from 26 to 28) than the SFEs and MOX FEs because the fission production rate ν  Σf of those types has a lower value than the SFEs and the MOX FEs as shown in Table 13 for the ACRO case for the ETRR-2 core fueled

SM of the FSS ($) 7.9417 0.020 SM of FSS with SF ($) 3.285 70.020 CRW ($) 16.9317 0.029 SRF 1.883 7 0.003 Parameter

Current ETRR-2 core re-fueled by the MOX fuel for the TNTs Control plates are Ag-In-Cd SM of the FSS ($) 7.3317 0.019 SM of FSS with SF ($) 3.1187 0.019 CRW ($) 16.549 70.030 SRF 1.795 7 0.003

by U3O8-Al original fuel and the MOX fuel for the ONT. 4. The MOX FEs in the sites 4, 5, 11, 18, 23 and 28 have lower values than the Type 2 FEs (See Fig. 8) for the criticality case. 5. The hottest FEs for both the U3O8-Al SFE and the MOX fuel are located in the positions 24 and 10 (See Fig. 7), respectively for the ACRO case while the hottest FEs for the criticality case are located in the positions 24 and 15 for the both the U3O8-Al original fuel and the MOX fuel (See Fig. 8), respectively. The WIMS-D4code was used to calculate the cell constants of the fuel types that are used in this study, the dimensions of the studied cells being taken as: fuel material (0.035 cm thickness), aluminum cladding (0.04 cm) and water channel (0.269 cm), and the neutron energies being selected as: - Fast energy group: 10 MeV to 0.821 MeV, - Epithermal energy group: 0.821 MeV to 1.02 eV, - Thermal energy group: 1.02 eV to 0 eV. The obtained results are given in Table 13. These results were used to justify the PD in the FEs as mentioned above. In the modified core re-fueled by the MOX fuel, the absorber material of the FSS should be changed from Ag–In–Cd plates to B4C plates without any change in the dimensions. The composition of

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Table 15 Calculated values of the neutron flux in the control plates for the MOX fuel. Parameter

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

MOX fuel with B4C as a control plates

Ag-In-Cd/B4C

Plate number 1 (See Fig. 7) Thermal neutron (n/cm2 s)  1012 Epithermal neutron (n/cm2 s)  1013 Fast neutron (n/cm2 s)  1013

5.417 70.016 2.0607 0.014 4.676 70.010

3.148 70.019 1.0057 0.015 4.308 7 0.010

1.720 2.049 1.085

Plate number 2 (See Fig. 7) Thermal neutron (n/cm2 s)  1012 Epithermal neutron (n/cm2 s)  1013 Fast neutron (n/cm2 s)  1013

6.2187 0.015 2.6357 0.013 5.9047 0.007

3.2797 0.019 1.2067 0.013 5.260 7 0.009

1.896 2.185 1.122

Plate number 3 (See Fig. 7) Thermal neutron (n/cm2 s)  1012 Epithermal neutron (n/cm2 s)  1013 Fast neutron (n/cm2 s)  1013

4.942 70.016 1.883 7 0.015 4.286 70.010

2.783 7 0.020 0.850 7 0.016 3.9127 0.009

1.775 2.215 1.095

Plate number 4 (See Fig. 7) Thermal neutron (n/cm2 s)  1012 Epithermal neutron (n/cm2 s)  1013 Fast neutron (n/cm2 s)  1013

5.618 70.016 2.0577 0.014 4.654 70.010

3.0977 0.019 0.9787 0.015 4.3347 0.010

1.814 2.103 1.074

Plate number 5 (See Fig. 7) Thermal neutron (n/cm2 s)  1012 Epithermal neutron (n/cm2 s)  1013 Fast neutron (n/cm2 s)  1013

6.4007 0.015 2.626 70.013 5.828 70.009

3.265 7 0.019 1.2067 0.014 5.312 70.008

1.960 2.177 1.097

Plate number 6 (See Fig. 7) Thermal neutron (n/cm2 s)  1012 Epithermal neutron (n/cm2 s)  1013 Fast neutron (n/cm2 s)  1013

5.184 70.016 1.9417 0.015 4.2617 0.010

2.880 7 0.020 0.880 7 0.016 3.9007 0.010

1.800 2.205 1.092

Table 16 Material constants of the control plates. Type

D Σa

Ag-In-Cd plates Fast neutron

Epithermal neutron

Thermal neutron

 5.14813E þ 10 4.422860E-11

4.30458E þ 04 4.911201E-07

2.57757E-02 1.29321E þ01

6.79918E-01 7.82013E-02

4.10691E-03 8.06122E þ 01

B4C plates D 1.47411E þ00 Σa 2.130270E-02

control material is shown in Table 3. The B4C plates seem to be more effective than the Ag–In–Cd plates for the ETRR-2 core refueled by the MOX fuel with ONT and TNTs. The MOX fuel reduces the effectiveness of the control plates (See Table 14) so they should be substituted by B4C plates. The calculations showed that in the case of the MOX fuel the thermal neutron fluxes reduce in the control plates positions (See Fig. 7 and Table 15) by about 45.18% 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 results of the SM for the FSS, the SM of the FSS with SF, the CRW and SRF of the new control system would be: 7.941 $, 3.285 $, 16.931 $ and 1.883, and 7.331 $, 3.118 $, 16.549 $ and 1.795 for the ONT and the TNTs, respectively as shown in Table 14. In this case the differences between these values and the same values for the current ETRR-2 core fueled by the U3O8-Al original fuel are: 47.6%, 61%, 31.4% and 28.4%, and 51.6%, 63%, 33% and 31.2% for the ONT and the TNTs, respectively.

7. Conclusion The MCNP4C2 computer code is used to model the ETRR-2 reactor when a MOX (U3O8&PuO2) dispersed fuel is used instead of

its original U3O8 fuel. The neutronic parameters for both fuels were compared and good agreement was found. Using the MOX in the MTR-22 MW core leads to reduce the 235U loaded mass in the core up to 34.84% and 15.21% for the ONT and TNTs, respectively, and to extend the scientific research area of the reactor.

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

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