THE RERTR-2006 INTERNATIONAL MEETING ON REDUCED ENRICHMENT FOR RESEARCH AND TEST REACTORS October 29 – November 2, 2006 Southern Sun Cape Sun Cape Town, Republic of South Africa

STEADY-STATE AND LOSS-OF FLOW TRANSIENT MODELING FOR TAJOURA NUCLEAR RESEARCH REACTOR

F. M. Bsebsu

Reactor Department Renewable Energies and Water Desalinization Research Center, Tajoura – Libya

and

P. L. Garner and N. A. Hanan Nuclear Engineering Division

Argonne National Laboratory, Argonne, Illinois, 60439 – United States of America

ABSTRACT

Design information is presented for high enriched uranium (HEU, using IRT-2M fuel) and low enriched uranium (LEU, using IRT-4M fuel) cores in Tajoura (IRT-1) reactor in Libya. Experiment results are presented for coolant flow rate versus pressure drop at steady state and flow during pump coast down. Steady state operational parameters at maximum power of 10 MW are presented. Calculated results are presented for several flow-induced accident situations. In no case does fuel approach conditions which might lead to damage.

1. Introduction The Tajoura (IRT-1) reactor [1] is a pool type reactor, moderated and cooled by light water located at the Renewable Energies and Water Desalinization Research Center (REWDRC). The reactor is designated to carry out experiments in field of nuclear physics and nuclear engineering, neutron activation analysis, solid state physics and isotope production. The reactor was put into operation at a power level of 10 MW in September 1983. The old fuel of the reactor is of the IRT-2M type: High Enriched Uranium (HEU, 80% of 235U); the fuel is an alloy (matrix) of aluminum and uranium-aluminum eutectic (UAlx–Al) with aluminum cladding. The reactor is being converted to Low Enriched Uranium (LEU, 19.7% of 235U) fuel of type IRT-4M [2]; the new fuel is an alloy (matrix) of aluminum and uranium-dioxide (UO2–Al) with aluminum cladding. This paper presents the analysis of the Tajoura core for power level (maximum) of 10 MW. The IRTCO [3] computer code and standard correlations have been employed to calculate different parameters such as coolant, cladding and centerline temperatures, and other thermal hydraulic critical parameters at steady state conditions. Results are shown from PARET [4] and RELAP [5] codes for response to flow-induced accidents. A companion paper [6] presents results for reactivity-induced transients.

2. Tajoura Reactor Core Design Parameters The base of the Tajoura core is a square grid plate with 36 identically formed places with a lattice pitch of 71.5 mm. The fuel assemblies (FA), the removable beryllium units, and guide tubes of the control rods (8 shim control rods, 2 safety rods and one automatic regulating rod) can be put into theses places. The compact core loading of Tajoura consists of 16 FAs. The FAs are surrounded by 20 removable beryllium units. Stationary beryllium reflector surrounds the removable core units. The horizontal cross section of the core is shown in Figure 1. The active fuel length is 0.60 m. The fuel is cooled by the pumped flow of water from top to bottom of the core. The HEU core is composed of IRT-2M FAs: 10 × 3-tube fuel assemblies (3TFA) and 6 × 4-tube fuel assemblies (4TFA). The IRT-2M FA geometries are shown on the left side of Figure 2. The 4TFA consists of 4 fuel elements (FE) and 5 coolant channels; the 3TFA is identical to the 4TFA except the innermost fuel tube is replaced with control rod guide tube. The fuel material is UAlx – Al matrix with 80% of 235U enrichment and its thickness is 0.4 mm covered by cladding material type SAV-1 (Russian Al alloy material) and its thickness is 0.8 mm. The 4TFA contains 164 g of 235U and the 3TFA contains 148 g of 235U. The LEU core is composed of IRT-4M FAs: 10 × 6-tube fuel assemblies (6TFA) and 6 × 8-tube fuel assemblies (8TFA). The IRT-4M FA geometries are shown on the right side of Figure 2. The 8TFA consists of 8 fuel elements and 9 coolant channels; the 6TFA is identical to the 8TFA except the 2 innermost fuel tubes are replaced with control rod guide tube. The fuel material is UO2 – Al matrix with 19.7% of 235U enrichment and its thickness is 0.7 mm covered by cladding material type SAV-1 (Russian Al alloy material) and its thickness is 0.45 mm. The 8TFA contains 300 g of 235U and the 6TFA contains 264 g of 235U. 3. Reactor Core Hydraulics Experiment Two hydraulic experiments have been performed recently in the Tajoura reactor at zero power. These experiments and their results are described below. 3.1. Experiment Procedures The Tajoura Nuclear Research Reactor consists of reactor core and three cooling system loops. The reactor pool depth is 8.5 m and the top of reactor core is located 7.0 m below the surface of water in the reactor pool, which it is open to atmospheric pressure. The

Figure 1 HEU (up) and LEU (down) Compact Core Loading for Tajoura

Reactor.

total power of the reactor core is 10 MWth. The primary cooling system is composed of the reactor pool (68 m3), fuel storage pool, emergency tank (7 m3), delay tank (21 m3), four main circulation pumps (each pump delivers 650 m3/hr as maximum rating), four heat exchangers, and the connecting piping. The secondary cooling system is closed system and it is composed of cooling piping connected the circuit, main circulation pumps, heat exchanger. The tertiary cooling circuit is open circuit, removing the heat using a cooling tower. After the loading of the reactor core with fuel, the hydraulic tests are essential to set up the reactor core operating parameters. The procedures of the experiment are as follows:

1. Prepare all the reactor electronic and analogue instruments.

2. Activate the secondary and tertiary reactor cooling loops. 3. Switch on primary pump no. 202-1. 4. Record the main parameters such as pump volume flow rate, level of water in

reactor and storage pools, pressure drop, and pressure under the reactor core, flow rate in shielding and horizontal beams.

5. Repeat step 3 and 4 for primary pumps 202-2 and 202-3. (Only 3 pumps are run simultaneously; the fourth pump is a spare to be used if another pump can not be used.)

6. After certain period of time switch off all pumps; get the opening time of natural circulation valves and filling time of emergency tank.

Steps 1 through 5 provide data on steady state coolant flow. Step 6 provides information during loss of pumping transient. 3.2. Experiment Results The experiment was repeated for many times for both IRT-2M and IRT-4M fuel cores. The steady-state results are shown in Figure 3. The main conclusion is that the maximum reactor core pressure drop must not be more than 0.7 kgf/cm2 and the pressure under the core not less than 0.12 kgf/cm2 in order to avoid air entering into the primary loop through the emergency tank. Based on these tests, we conclude that the upper limit value of coolant volume flow rate for TAJOURA reactor core with HEU fuel is equal to 1872 m3/hr, giving a pressure drop 0.0657 MPa, and pressure under the reactor core of 0.0123 MPa; the upper limit flow with LEU fuel is equal to 1350 m3/hr, giving a pressure drop 0.0657 MPa, and pressure under the reactor core of 0.01177 MPa.

Figure 2 IRT-2M (left) and IRT-4M (right) Fuel Assembly

Cross Section

400 600 800 1000 1200 14000.00

0.01

0.02

0.03

0.04

0.05

0.06

0.07

Pressure under Core

Pressure Drop

Pres

sure

[MPa

]

Volume Flow Rate [m3/hr]

Figure 3 Core Pressure Drop (∆P) and Pressure under Core (PUC) for Compact

Core of IRT-4M Fuel Assemblies

Experiment step 6 is associated with loss of electrical power to the pumps in primary loop. As the primary flow rate decreases, primary coolant moves into the emergency tank (ET). At some time during this transient, the pressure in the delay tank (DT, under the reactor pool) and the bottom of the reactor pool will become equal, and two natural circulation valves (NCV) will automatically open allowing natural circulation of coolant. The results for experiment step 6 are shown in Figure 4 for both cores. For the HEU core, the NCVs open at 63.76 s and water stops entering the ET in 82 s. For the LEU core, the NCVs open at 79.9 s and water stops entering the ET in 155 s. 4. Steady-State Theoretical Results The results of the first experiment (pressure drop versus coolant volume flow rate as shown in Figure 3) are used as an input parameter of IRTCO computer code at reactor thermal power level of 10 MWth for HEU core and 8.6 MWth for LEU core. The theoretical results of Tajoura reactor core are given in Table 1. Results for the peak power FA of each type are shown in the following subsections.

Table 1 Tajoura Reactor Core Thermal Hydraulic Parameters

Parameter HEU LEU Reactor Core Power, MW 10 8.6 Inlet Pressure , MPa 0.16932 0.16932 Pressure Drop, MPa 0.06570 0.06570 Active Mass Flow Rate , kg/s 171 147 4/8TFA Coolant Volume Flow Rate, m3/h 70.1 33.89 3/6TFA Coolant Volume Flow Rate, m3/h 62.3 31.93 Active Coolant Volume Flow Rate, m3/h 1044 533 Primary Coolant Volume Flow Rate, m3/h 1850 1350 Primary Inlet Temperature, oC 45 45 Primary Out Temperature, oC 49.86 50.5 Primary Temperature Difference, oC 4.86 5.5

Table 2 shows the main operating hydraulic parameters of 4TFA such as coolant mass flow rate ( m& ), and coolant velocity (V), outlet pressure (Pout) and temperature (Tout), and critical conditions such as departure from nucleate boiling ratio (DNBR), onset of nucleate boiling factor (ONBF), temperature for ONB (TONB), coolant saturation temperature (Tsat), and coolant boiling temperature (Tboil) for each coolant channel of the fuel assembly. (In tables, FE and coolant channel numbers are “1” for outer and increase moving inward.) The coolant volume flow rate per 4TFA is equal to 70.1 m3/hr and its mass flow rate 19.3 kg/s. Table 3 shows the maximum fuel centerline temperature (Tfuel), clad surface temperature (Tclad), and clad surface heat flux (q”) on both sides of fuel elements of 4TFA. Table 4 shows the main operating hydraulic parameters of 3TFA. The coolant volume flow rate per 3TFA is equal to

0 20 40 60 80 100 120 140 160 180 200-0.5

0.0

0.5

1.0

1.5

2.0

2.5

3.0

NCV Opening Time

79.9 s

63.76 s

Emer

genc

y Ta

nk L

evel

[m]

Time,[s]

Experimental Results for ET Filling Time HEU Core LEU Core

Figure 4 Water Level in Emergency Tank versus Time Measured for HEU and LEU Core During

Primary Pumps Stop.

62.3 m3/hr and its mass flow rate 17.1 kg/s. Table 5 shows the relevant temperatures for 3TFA.

Table 2

4TFA Operating Hydraulic Parameters Ch. No.

m& [kg/s]

V [m/s]

Pout [MPa]

Tout [°C] DNBR ONBF TONB

[°C] Tsat [°C]

Tboil [°C]

5 3.821 5.451 0.120 53.1

2.376 1.557 121.6 107.3 133.4 4 5.401 5.120 0.122 55.3 3 4.235 5.153 0.122 55.2 2 3.070 5.212 0.122 55.1 1 2.178 5.499 0.120 50.7

Table 3 4TFA Maximum Heat Flux and Temperature

FE Inner Side Tfuel [°C] Outer Side q” [MW/m2] Tclad [°C] Tclad [°C] q” [MW/m2]

4 1.50069 101.3 104.0 99.2 1.44281 3 1.51336 101.5 105.1 101.2 1.44049 2 1.53498 101.8 105.1 101.0 1.43664 1 1.58009 99.5 103.8 100.2 1.42900

Table 4 3TFA Operating Hydraulic Parameters

Ch. No.

m& [kg/s]

V [m/s]

Pout [MPa]

Tout [°C] DNBR ONBF TONB

[°C] Tsat [°C]

Tboil [°C]

4 3.872 5.523 0.120 51.9

2.197 1.490 122.2 107.3 134.0 3 5.471 5.185 0.122 53.82 4.290 5.218 0.121 53.81 3.103 5.262 0.121 50.2

Table 5 3TFA Maximum Heat Flux and Temperature

FE Inner Side Tfuel [°C] Outer Side q” [MW/m2] Tclad [°C] Tclad [°C] q” [MW/m2]

3 1.64261 104.5 105.7 100.5 1.57925 2 1.65647 104.7 106.7 102.6 1.57671 1 1.68014 103.1 105.8 102.3 1.57249

Table 6 shows the main operating hydraulic parameters of 8TFA. The coolant volume flow rate per 8TFA is equal to 33.89 m3/hr and its mass flow rate 9.321 kg/s. Table 7 shows the relevant temperatures for 8TFA. Table 8 shows the main operating hydraulic parameters of 6TFA. The coolant volume flow rate per 6TFA is equal to 31.93 m3/hr and its mass flow rate 8.785 kg/s. Table 9 shows the relevant temperatures for 6TFA. Temperatures and other parameters are within allowed limits for operation.

Table 6

8TFA Operating Hydraulic Parameters Ch. No.

m& [kg/s]

V [m/s]

Pout [MPa]

Tout [°C] DNBR ONBF TONB

[°C] Tsat [°C]

Tboil [°C]

9 0.319 3.133 0.125 59.7

2.574 1.484 118.5 107.2 129.7

8 0.980 3.785 0.122 56.27 0.724 3.222 0.124 65.06 0.889 3.220 0.124 65.05 1.058 3.222 0.124 64.74 1.222 3.220 0.124 64.83 1.387 3.219 0.124 64.92 1.542 3.213 0.124 65.01 1.199 3.551 0.123 58.6

Table 7

8TFA Maximum Heat Flux and Temperature FE Inner Side Tfuel [°C] Outer Side

q” [MW/m2] Tclad [°C] Tclad [°C] q” [MW/m2] 8 0.99777 100.4 100.5 91.2 0.87084 7 0.97249 95.1 100.8 99.2 0.87957 6 0.95631 102.1 104.1 99.5 0.88550 5 0.94848 101.8 103.9 99.4 0.88847 4 0.94202 101.3 103.8 99.5 0.89097 3 0.93726 101.2 103.8 99.6 0.89285 2 0.93340 101.1 103.8 99.8 0.89439 1 0.93060 101.2 101.9 94.3 0.89552

Table 8

6TFA Operating Hydraulic Parameters Ch. No.

m& [kg/s]

V [m/s]

Pout [MPa]

Tout [°C] DNBR ONBF TONB

[°C] Tsat [°C]

Tboil [°C]

7 1.081 3.183 0.122 51.4

2.940 1.649 118.3 107.2 128.9

6 0.940 3.402 0.122 60.95 1.115 3.400 0.122 60.74 1.290 3.399 0.122 60.73 1.464 3.399 0.122 60.82 1.629 3.392 0.122 60.91 1.266 3.743 0.120 55.8

Table 9 6TFA Maximum Heat Flux and Temperature

FE Inner Side Tfuel [°C] Outer Side q” [MW/m2] Tclad [°C] Tclad [°C] q” [MW/m2]

6 0.89709 90.2 94.6 93.7 0.83066 5 0.88975 95.9 97.2 93.6 0.83345 4 0.88369 95.5 97.0 93.7 0.83580 3 0.87922 95.4 97.0 93.9 0.83756 2 0.87560 95.3 97.1 94.0 0.83900 1 0.87297 95.3 95.6 89.5 0.84007

5. Loss of Flow Analysis The most severe Loss-Of-Flow Accident (LOFA) is due to loss of off-site electrical

power which will lead to stopping of the pumps. The transition from forced flow to natural circulation is aided by the presence of the Emergency Tank, as described in the next subsection. This transition is then used to set the flow boundary condition to use in the PARET analysis shown in the second subsection.

5.1. Emergency Tank Filling Model The reactor fuel is normally cooled by downward forced circulation of coolant; the primary pumps also provide cooling of the reflector, the experiment tube structures, and the pool walls. Following loss of electric power to the primary pumps, the primary flow rate decreases, the pressure increases in the delay tank (DT) under the reactor pool, and the water level rises in the emergency tank (ET). The 7 m3 ET is located in the fuel storage portion of the reactor pool and connected to the DT (which is the exit path for coolant from the core) by a pipe; during normal operation, the ET has a low water level. During a loss of flow, the water level rise in the ET helps maintain the downward flow of coolant in the core for a longer time than would be otherwise be associated with the decrease in revolution speed of the primary pumps. Once the pressure in the DT increases to be the same as the pressure in the bottom of the reactor pool, the natural circulation valves (NCV) open, allowing a path for natural circulation of coolant from the DT up through the reactor core, down through the reactor pool, through the NCVs, to the DT. Two NCVs are provided for redundancy; the flow through either NCV is sufficient for reactor cooling. Various relations among coolant flow and water level may be obtained by applying and solving the mass, momentum, and energy conservation equations. One such set of relations follows. [7] 1. The coolant velocity, V(t) [m/s], leaving the reactor pool as function of time, t [s], is

assumed to decrease linearly in time during loss of primary pumps: V(t) = ao – bo × t where: ao = 7.5910, bo = 0.119100 for HEU core and ao = 8.5625, bo = 0.107165 for LEU core.

The values of the constants are set by knowing the velocity at t=0 and requiring V=0 at the time determined by experiment when the natural circulation values open (i.e., 63.76 s for HEU and 79.90 s for LEU).

2. The reactor pool coolant level, ZRP(t) [m], variation as a function of time is found by integrating V(t) multiplied by a flow area ratio factor:

ZRP(t) = LRP + co × t2 – do × t where: co =1.58929×10-4 , do = 0.020262982 for HEU core, co =1.39546×10-4 , do = 0.022299568 for LEU core, and LRP = ZRP(0) = 8 m. 3. At any time the emergency tank coolant level, ZET(t) [m], differs from the reactor pool

coolant level by a factor proportional to the coolant velocity: ZET(t) = ZRP(t) – eo × V2(t) where: eo = 0.04760 for HEU core and eo = 0.05456 for LEU core. The water levels from these equations as a function of time are shown in Figure 5 for the HEU and LEU cores. Figure 6 shows the variation of water level in the emergency tank as a function of filling time for theoretical and experimental results.

5.2. PARET Analysis Loss of off-site electrical power will result in an immediate scram and the primary pumps will begin to coast down. For the purposes of modeling this using the PARET code [4] the coolant flow rate through the core is assumed to decrease linearly over the time required for the NCV to open, as described in the previous section; after that time, the core flow transitions (over an assumed time of 5 s) from pump-forced downward flow to buoyancy-driven upward flow. The peak coolant and clad surface temperatures in the highest power channel computed using PARET for this transient in the HEU core are shown in Figure 7. The pump trip occurs at time of 10 s in the PARET cases. There is no initial temperature increase since the decrease in coolant flow rate is sufficiently slow and the control rods are moving into core 0.5 s after start of the flow decrease. Maximum temperature decreases during the power decrease. As the flow rate approaches zero, the maximum clad temperature rises, peaking about 5 s later at 93°C, which is 16°C lower than the value at the beginning of the transient. The clad temperature remains much lower than the temperature at which clad damage might occur. The buoyancy driven upward flow is 90-110 kg/m2-s at 2 minutes after the start of the transient. The reactor period is always negative.

0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 75 80 853.5

4.0

4.5

5.0

5.5

6.0

6.5

7.0

7.5

8.0

8.5

(79.90, 7.11)

(63.75, 7.35)

H2O

Lev

el [m

]

Time [sec]

IRT-4M Core Reactor Pool Level Emergency Tank Level

IRT-2M Core Reactor Pool Level Emergency Tank Level

Figure 5 Computed Water Levels during LOF

0 20 40 60 80 100 120 140 160 180 2000.0

0.5

1.0

1.5

2.0

2.5

3.0

Emer

genc

y Ta

nk H

2O L

evel

[m]

Time [sec]

Experimental Theoritcal (IRT-4M) Theoritcal (IRT-2M)

Figure 6 Measured and Calculated Emergency Tank

Water Level during LOF

TNR-IRT-1: HEU: LOF 2.2: LOOP 3 pumps in 70 s, with scram, PARETv5

0

20

40

60

80

100

120

140

10 20 30 40 50 60 70 80 90 100 110 120 130

Time [s]

Tem

pera

ture

[C]

TcoMax-1 TclMax-1

Figure 7 Computed Temperatures during LOF in

HEU Core

TNR-IRT-1: LEU: LOF 2.2: LOOP 3 pumps in 80 s, with scram, PARETv5

0

20

40

60

80

100

120

140

10 20 30 40 50 60 70 80 90 100 110 120 130

Time [s]

Tem

pera

ture

[C]

TcoMax-1 TclMax-1

Figure 8 Computed Temperatures during LOF in

LEU Core

The equivalent results for the LEU core are shown in Figure 8 and have the same general behavior as computed for the HEU core; a few specific values follow. As the flow rate approaches zero, the maximum clad temperature rises, peaking about 8 s later at 114°C, which is 2°C higher than the value at the beginning of the transient and remains much lower than the temperature at which clad damage might occur. The buoyancy driven upward flow is 50-75 kg/m2-s at 2 minutes after the start of the transient. 6. Partial Blockage of Fuel Assembly The consequence of blocking coolant flow to the coolant channel which is interior to the innermost fuel element in the peak power fuel assembly has been examined. Flow to the other coolant channels is not modified. The blockage would not cause a reactor scram; therefore, the analysis is performed at a constant power of 10 MW. The RELAP5/MOD3 code [5] was used. The model treats all fuel tubes as parallel plates separated by their coolant channels. The results for the clad surface temperature for the inner FE after the blocking the inner channel of the HEU 4TFA with highest power density are provided in Figure 9; the results for the normal flow (i.e. unblocked) case are also included for comparison. These results show that the peak clad surface temperature for the inner tube is increased (to 160°C) but is still much lower than the temperature at which clad damage might occur. The equivalent results after blocking the inner channel of the LEU 8TFA with highest power density are provided in Figure 10. As for HEU, the peak clad surface temperature for the inner tube of the LEU fuel is increased (to less than 130°C) but is still much lower than the temperature at which clad damage might occur. Although not shown as a Figure, the temperatures of the outer fuel tubes are not significantly affected by the blockage interior to the inner fuel tube for either the HEU or LEU case. Additional studies indicated that one could approximate this blocked channel situation by using a full (not the usual half) fuel plate thickness in PARET with the same power density rather than requiring the use of other codes such as RELAP5 or PLTEMP. 7. Conclusions Design information has been presented for HEU and LEU cores in Tajoura reactor. Coolant flow versus pressure drop at steady state and flow during pump coast down have

IRT-1 HEU: SS Inner Channel Blocked vs Open - Tube 4 (Inner Tube)

60

70

80

90

100

110

120

130

140

150

160

170

0 10 20 30 40 50 60Distance from Bottom of Active Fuel, cm

Cla

d Su

rfac

e Te

mp,

C

normal-left normal-right blocked-left blocked-right

Figure 9 Effect of 1 Channel Blockage on

Temperature in IRT-2M FA

IRT-1 LEU: SS Inner Channel Blocked vs Open - Tube 8 (Inner Tube)

50

60

70

80

90

100

110

120

130

0 10 20 30 40 50 60Distance from Bottom of Active Fuel, cm

Cla

d Su

rfac

e Te

mpe

ratu

re, C

open-left open-right blocked-left blocked-right

Figure 10 Effect of 1 Channel Blockage on

Temperature in IRT-4M FA

been measured experimentally. Steady state operational parameters have been calculated. Calculated results have also been presented for several flow-induced accident situations. In no case does fuel approach conditions which might lead to damage. References [1] Tajoura Reactor Core Design Documents, Tajoura (Tripoli) Libya (1980). [2] IRT-4M Fuel Assemblies, Catalogue Description – Л 0019.20.00.000 ДКО,

Novosibirsk Chemical Concentrates Plant, 630110 Novosibirsk-110, B. Khmelnitsky St., 94, Russia, (28.12.2004).

[3] F. M. Bsebsu, Tajoura Nuclear Research Reactor Core with LEU Fuel Thermal Hydraulic Operating Parameters Setup, REWDRC: Technical Report No.: R-CU2-05-2006, Tajoura (Tripoli), Libya (2006).

[4] W. L. Woodruff and R. S. Smith, “A Users Guide for the ANL Version of the PARET Code, PARET/ANL (2001 Rev.)”, ANL/RERTR/TM-16, Argonne National Laboratory, Argonne, Illinois (March 2001); Woodruff, W.L., “A User’s Guide for the ANL Version of the PARET Code”, National Energy Software Center, Argonne, Illinois (1984); and Obenchain, C.F., “PARET – A Program for the Analysis of Reactor Transients”, IDO-17282, AEC Research and Development Report, Phillips Petroleum Company (January 1969).

[5] “RELAP5/MOD3.2 Code Manual”, NUREG/CR-5335 or INEL-95-0174, Idaho National Engineering Laboratory (June 1995)

[6] F. M. Bsebsu, P. L. Garner, and N. A. Hanan, “Transient Thermal Hydraulic Analysis for the Tajoura Nuclear Reactor with HEU and LEU Fuels”, Proceedings of the RERTR-2006 International Meeting on Reduced Enrichment for Research and Test Reactors, Cape Town, South Africa (October 29 – November 2, 2006).

[7] F. M. Bsebsu, Mathematical Model of Tajoura Nuclear Research Reactor Core with LEU Fuel LOFA Accident through Horizontal Beam Channels, REWDRC: Technical Report No.: R-CU2-06-2006, Tajoura (Tripoli), Libya (2006).