United Nations Educational Scientific and Cultural Organization and International Atomic Energy Agency

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1 Available at: IC/26/2 United Nations Educational Scientific and Cultural Organization and International Atomic Energy Agency THE ABDUS SALAM INTERNATIONAL CENTRE FOR THEORETICAL PHYSICS THE LOW POWER MINIATURE NEUTRON SOURCE REACTORS: DESIGN, SAFETY AND APPLICATIONS Y.A. Ahmed 1 Centre for Energy Research and Training, Ahmadu Bello University, PMB 114, Zaria, Nigeria and The Abdus Salam International Centre for Theoretical Physics, Trieste, Italy, I.O.B. Ewa, I.M. Umar Centre for Energy Research and Training, Ahmadu Bello University, PMB 114, Zaria, Nigeria, T. Bezboruah 1 Department of Electronics Science, Gauhati University, Guwahati-78114, Assam, India and The Abdus Salam International Centre for Theoretical Physics, Trieste, Italy, M. Johri 1 Department of Physics and Electronics, DAV College, CSJM University, Kanpur-281, UP, India and The Abdus Salam International Centre for Theoretical Physics, Trieste, Italy and E.H.K. Akaho Nuclear Research Institute, Ghana Atomic Energy Commission, Legon-Accra, Ghana. MIRAMARE TRIESTE April 26 1 Junior Associate of ICTP. Corresponding author: yaahmed2@yahoo.com

2 Abstract The Chinese Miniature Neutron Source Reactor (MNSR) is a low power research reactor with maximum thermal neutron flux of 1 x 1 12 n.cm -2.s -1 in one of its inner irradiation channels and thermal power of approximately 3kW. The MNSR is designed based on the Canadian SLOWPOKE reactor and is one of the smallest commercial research reactors presently available in the world. Its commercial versions currently in operation in China, Ghana, Iran, Nigeria, Pakistan and Syria, is considered as an excellent tool for Neutron Activation Analysis (NAA), training of Scientist, and Engineers in nuclear science and technology and small scale radioisotope production. The paper highlights the basic design and theory of the commercial MNSR, its safety features, applications and advantages over the Chinese Prototype. The experimental flux characteristics determined in this work and in similar studies by other authors reveal that the commercial MNSR has more flux stability, longer life span, higher negative temperature coefficient of reactivity and low under-moderation compared to its prototype in China. The result shows that the facility is safe for reactor physics experiments, teaching and training of students and also ideal for application of NAA for the determination of elemental composition of biological and environmental samples. It can also be a useful tool for geochemical and soil fertility mapping. 1

3 1. INTRODUCTION Research reactors comprise a wide range of civil and commercial nuclear reactors which are generally not used for power generation. The primary purpose of research reactors is to provide a neutron source for research and other purposes. They are small relative to power reactors whose primary function is to produce heat to make electricity and power is designated in megawatts (or kilowatts) thermal. Research reactors are simpler than power reactors and operate at lower temperatures. They need far less fuel, and far less fission products build up as the fuel is used. On the other hand, their fuel requires more highly enriched uranium, typically up to 92% U-235. They also have a very high power density in the core, which requires special design features. Like power reactors, the core needs cooling, and usually a moderator is required to slow down the neutrons and enhance fission. As neutron production is their main function, most research reactors also need a reflector to reduce neutron loss from the core. The Chinese built MNSR (Yang 1992) is a small and compact research reactor based on the Canadian SLOWPOKE reactor design. The MNSR is getting acceptance over the world because of its inherent safety (negative temperature coefficient of reactivity), simple auxiliary facilities, reliable cooling and shielding mechanism, and most of all low radioactivity release to the environment. The stability and reproducibility of its neutron flux and the large flux to power ratio gives it an analytical advantage over the bigger types. MNSR is designed to be situated at hospitals, universities, research institutions, and training centers located in densely populated area. 2. DESIGN CONSIDERATIONS The MNSR uses highly enriched U-235 fuel in an Al cladding. A tank is immersed in a large pool of water, and the core is, in turn, immersed in the tank- an arrangement popularly called tank-in-pool type (Yang, 1992). Their maximum nominal power is ~ 3 kw which is equivalent to a thermal neutron flux of 1 12 n.cm -2.s -1 in one of the inner irradiation channels (Balogun 23; Ahmed 24). For a typical MNSR, the central core consists of 347 fuel rods, with 4 tie rods and 3 dummy elements uniformly distributed on a total of ten concentric circles, each having a number of fuel rods ranging between 6 and 62 (Table 1) with a thick beryllium reflector (~ 1 cm) surrounding the core radially (Yang 1992). 2

4 Table 1: Technical Specifications of MNSR Parameters Description Reactor type Tank-in-pool Rated thermal power 3 kw Fuel UA 4 dispersed in Al U-234 Enrichment 9.2% Core shape Cylinder Core diameter 23.1 cm Core height 23. cm No. of fuel elements 344 Weight of U g No. of irradiation channels 1 Inner channels 5 Flux in inner channel 1 x 1 12 n cm -2 s -1 Flux in outer channel 5 x 1 11 n cm -2 s -1 Reactor cooling mode Natural convection Height of inlet orifice 6 mm Height of outlet orifice 7.5 mm Diameter of fuel meat 4.3 mm Diameter of fuel element 5.5 mm Excess reactivity 3.99 mk Length of Cd control rod 23 mm The MNSR has one central control rod made of cadmium and cladded with stainless steel. The rod performs shim, safety and regulatory control functions (Qazi et al. 1996). This makes the reactor safe and compact enabling it to be located in densely populated area without harm to the public (Yang 1992). The geometrical dimensions of the reactor fuel lattice are shown in Table 2 and a schematic diagram of the reactor is shown in Figure 1. A detailed description of the operating characteristics of MNSR has been presented elsewhere (Akaho et al., 2; Ahmed et al., 22; Balogun et al., 24). 3

5 Table 2: Geometrical dimension of the MNSR core lattice Fuel Ring No. No. of Fuel Rods Diameter (mm) Radial Pitch (mm) Outer channel Inner channel Pool Water 2.7 m Reactor Water Top Beryllium shim 6.5 m Side Beryllium reflector Reactor Core Support frame Reactor Vessel Figure 1: Schematic Diagram of MNSR (Side view) Shielding concrete Due to its inherent safety features, stability of flux and moderate cost, the MNSR has recently found enormous application in various fields of science (Kennedy et al., 2; Akaho and Nyarko 22), particularly in trace elements in matrices of biological and environmental samples (Valkavic 1975; Su-De 1984) and soil fertility studies and geochemical mapping (Umar 23, Jonah et al. 26). 4

6 3. SAFETY FEATURES The MNSR is designed to incorporate some inherent safety measures in order to prevent radiation leakage. The objective of its design is to achieve a small core structure with undermoderation, achieved by choosing highly enriched uranium as fuel material with light water serving as both moderator and coolant (Yang 1992). The core structure is simplified by arranging the irradiation channels in a beryllium reflector (Fig 1). This reduces the effect of flux distribution on the irradiation experiments due to movement of the central control rod (Balogun 23), thus, enhancing the performance of the reactor from both mechanical and computational physics points of view. Two major provisions were made in the design to increase the safety margin and to enhance the inherent safety of the reactors as mentioned below: (i) The 197 ratio of hydrogen to Uranium 235 enlarges the degree of under-moderation process of the core and thereby increases the negative temperature coefficient and enhances the temperature feed back. (ii) The ratio of the core height to diameter was optimized to 1. to take care of axial neutron leakage fraction, to increase the coupling of outlet and inlet flow, mixing of the water coolant and to enable part of the warm coolant to directly enter the core, to shorten the time of temperature feedback and to increase the worth of the top beryllium reflectors-used for the extension of the lifetime of the core. The two factors above became the new features that the commercial MNSR possed over its prototype counterpart. 4. THEORETICAL CONSIDERATIONS The total reaction rate per target atom for (n, γ) reactions induced by both thermal and epithermal neutrons in a reactor is given (IAEA, 199) by: where Φ th Φ epi σ th R σ (1) = R th + R epi = Φ th th + Φ epiι = thermal neutron flux = epithermal neutron flux = average thermal neutron cross-section I = effective resonance integral For an ideal (1/E) flux distribution I is defined (Simonits et al., 1984) as: 5

7 = de I σ ( E) (2) E E E =.55eV (effective Cd cut-off energy σ (E) = (n,γ) cross-section However, for the non-ideal (real) reactor situations (Figure 2), the resonance integral I needs to be modified with an -dependent term because the I values, which are valid only for ideal spectra, is not true for a deviating spectra (Meons et al., 1979). For the non-ideal conditions I () values ought to be used instead of I. The conversion of I to -dependent terms takes the form (Jovanovic et al, 1987): where = I.429σ.429σ Ea I ( ) = + E de E ( ) (3) a E 1+ (2 + 1) E r E E r = effective resonance energy Ea E σ = = 1eV arbitrary energy =.55eV-effective cadmium cut-off energy = 22ms -1 (n, γ) Cross section an experimentally determinable characteristics of the reactor channel. E Although the epithermal neutrons represent only a small fraction of the total reactor neutrons, they are sometimes useful in NAA for several elements (e.g. Br, Rb, Sr, Mo, Ba, Ta and U) that have higher relative reaction rates for epithermal neutrons than for thermal neutrons. The technique of taking advantage of those (n, γ) reactions with high resonance integrals through the irradiation of samples under a cadmium cover to shield out the thermal neutrons is commonly known as Epithermal Neutron Activation Analysis (ENAA). The large number of resonance peaks for most nuclides makes a calculation of effective cross sections for the epithermal neutrons slightly complicated. In order to avoid these resonances, a gold standard is used (Nisle, 1963) because the reaction 197 Au (n, γ) 198 Au has a single resonance (411KeV) peak and it has been well investigated (De Corte et al., 1987; Kennedy et al., 2) and found to be 155 barns. The activity ratio for an infinitely thin gold foil or alloy irradiated with and without cadmium covers (Cadmium ratio method) is used (De Corte et al., 1981; Jovanovic et al., 1989; Akaho and Nyarko, 22) for measuring epithermal flux-shaping factor and thermal to epithermal flux ratio. This method is also used as a calibration standard to measure the resonance integrals for (n, γ) reactions (Moens et al., 1979). 6

8 Figure 2: Neutron Cross-section showing the non-ideality of flux in the epithermal region The equation describing the cadmium ratio method is: R Equation (4) can be also written as: A ( φ I ) thσ th + φ bare epi = = (4) F A φ I R R = A epi bare = (5) Repi F A F = cadmium transmission factor Abare = activity of bare monitor A = activity of cadmium covered monitor Using equation (1) in (5) we obtain: ( R φth 1) = φepi σ I th = The ratio of thermal to epithermal flux is defined as: f Q φth f = φ epi (6) 7

9 and the ratio of resonance integral to thermal cross-section is given as: Q = I σ th Thus f = ( R 1) Q (7) R = Cadmium ratio If the single comparator technique is adopted for routine NAA, the effect of the non-ideal epithermal spectrum should not be neglected. Thus, to be accurate in all relevant expressions, Q should be replaced by Q ( ) [where Q ( ) is the -corrected Q to take care of non-ideality of the epithermal spectrum] as defined by De Corte, 1987 (Chapter V): I ( ) Q Q ( ) = = + (8) a σ E (2 + 1) E th r Experimental determination of, using the cadmium ratio method involves irradiating two or more monitors (e.g. Au, Zr, Co) with and without cadmium alternately at the same channel (De Corte et al., 1981). When monitors such as 197 Au(1) and 94 Zr(2) are the irradiated under uniform neutron flux (f 1 ~f 2 ), equation (7) transforms to: ( ) 1 (,2 1) ( ) 2 ( R 1) Q = R Q (9) The epithermal deviation factor could be obtained either from equation (9) (using iteration method) or mathematically as a negative slope of the straight line by a plotting graph of equation (1): log k, Au ( E ) ( A ) r, i sp, i () i. ε p, i. F, i. Q, i ( ). Ge, i Versus ( ) log (1) E r, i where i = isotope 1,2,3..N G e,i = epithermal neutron self-shielding factor for the i th monitor, G th,i = thermal neutron self-shielding factor for the i th monitor. 5. EXPERIMENTAL TECHNIQUE Standard Gold solution and Zirconium foils with masses between 4.7 and 5. mg were sent to the outer irradiation channel 7 and the inner irradiation channel 5 of GHARR-1 research reactor for one-hour irradiation. The samples were irradiated with the reactor operating at half thermal power (15 kw) with neutron flux of 5x 1 11 n cm -2 s -1 in one of the inner irradiation channels. The nuclear parameters of the flux monitor foils were obtained from literature (Erdtmann and Soyka 1979). The irradiated samples after an appropriate decay period were assayed for 3 minutes at the geometry of 7.2 cm using a gamma-ray spectroscopy system. The system consists of an N-Type High Purity Germanium (HPGe) detector (Model: GR2518) with relative efficiency of 25% and an energy resolution of 1.8KeV (FWHM) at KeV gamma ray of 6 Co and a SPAN-spectrum 8

10 accusation and simulation software (Wang, 1998). The efficiency parameters of this detector were obtained by fitting the efficiency curve with polynomial function using parameters determined and presented elsewhere (Osae et al., 1999). 6. RESULTS AND DISCUSSION The result of the measurements of thermal-epithermal flux ratio and epithermal flux-shaping factor carried out in this study and the ones reported earlier by other authors are presented and discussed below. 6.1 Measurement of thermal to epithermal flux ratio The determination of flux parameters in an irradiation channel is necessary to monitor the continuous stability of the reactor flux (required for activation analysis) and to characterize a new channel (Jonah et al., 25), re-characterize an old channel (De Corte and De Wispelaere, 23) after refueling and/or core configuration change (Mustra et al., 23). Experimental determination of thermal to epithermal flux is achieved via cadmium ratio or the bare monitor method (Simonits et.al., 1975; De Corte et al., 1981, 1982, 1987). Details of the procedure involved to characterize MNSR channels were discussed elsewhere (Kennedy et al., 2; Akaho and Nyako 22, Ahmed 24, Jonah et al., 25). The results obtained for this work and for similar reactors are presented in Table 3. Table 3: Measured Thermal to Epithermal flux ratios (f-values) in MNSR irradiation channels Reactor Facility (Channels) Values Reference MNSR GHARR-1, Accra, Ghana This work (5 Inner) MNSR NIRR-1, Zaria, Nigeria 19.2 Jonah et al., 25 (B2 Inner) MNSR GHARR-1, Accra, Ghana 18.8 Akaho et al., 22 (1 Inner) Prototype MNSR Beijing, China 19.8 Kennedy et al., 2 (Inner) SLOWPOKE, Dalhousie Univ Akaho et al., 22 Halifax, Canada (5 Inner) MNSR GHARR-1, Accra, Ghana 4.38 This work (7 Outer) MNSR NIRR-1, Zaria, Nigeria 48.3 Jonah et al., 25 (B4 Outer) MNSR GHARR-1, Accra, Ghana 49. Akaho et al., 22 (6 Outer) SLOWPOKE, Dalhousie Univ Akaho et al., 22 Halifax, Canada (1 Outer) MNSR Prototype China (Outer) 58.5 Kennedy et al., 2 9

11 6.2 Measurement of flux deviation in irradiation channels Among the various experimental methods available for determining the epithermal deviation factor (), the cadmium ratio method is known to yield the most accurate results (De Corte et al., 1981). Determination of the flux deviation factor in this study was carried out using the cadmium ratio method. This involves irradiation of Gold (Au) and Zirconium (Zr) monitors with and without cadmium cover alternately at the same channel under uniform neutron flux. A detailed procedure for this method has been discussed elsewhere (Ahmed et al., 22; Ahmed 24) and the results are shown in Table 4. The results obtained in this work and the ones reported earlier by other authors (Tables 3 and 4) shows that the Prototype MNSR has higher flux parameters indicating that it has a better thermalization than its commercial counterpart. This is in agreement with the 1:1 Zero-power Audit Experimental results for the MNSR which indicates that relative to the prototype, the commercial MNSR is designed with the value of H/U-235 atoms reduced from 24 to 197 in the fuel lattice (Jonah et al., 25). This has been done to further enhance the life span and to improve safety measures by ensuring a relatively higher negative temperature coefficient of reactivity. This fact places the prototype on the disadvantage side when safety and life span is considered as well as applicability in research and training. Table 4: Epithermal flux shaping factor ( ) in MNSR irradiation channels Reactor Facility (Channels) Values Reference MNSR GHARR-1, Accra, Ghana -.17 This work (5 Inner) MNSR NIRR-1, Zaria, Nigeria -.52 Jonah et al., 25 (B2 Inner) MNSR GHARR-1, Accra, Ghana -.14 Akaho et al., 22 (1 Inner) Prototype MNSR Beijing, China -.9 Kennedy et al., 2 (Inner) SLOWPOKE, Dalhousie Univ Akaho et al., 22 Halifax, Canada (5 Inner) MNSR GHARR-1 Accra, Ghana This work (7 Outer) MNSR NIRR-1, Zaria, Nigeria +.29 Jonah et al., 25 (B4 Outer) MNSR GHARR-1 Accra, Ghana Akaho et al., 22 (6 Outer) Prototype MNSR Beijing, China -.23 Kennedy et al., 2 (Outer) SLOWPOKE, Dalhousie Univ. Halifax, Canada (1 outer) -.98 Akaho et al., 22 1

12 7. POTENTIALS FOR TRAINING AND RESEARCH The interest in training is the main concern for developing countries with high priority national program in nuclear science and technology. The low power reactors discussed in this study find a wide variety of applications and have great potentials in achieving the training and research needs of these countries. The facility was found useful to current needs, such as characterization of pollutants and determination of the source and method of reducing these in environment using the NAA technique (Witkowska et al., 25). The technique can also be utilized in the analysis of semiconductor materials to measure ultra trace elements, impurities and to determine the methods for reducing or eliminating impurities from the final products. Forensic studies can be performed as a nondestructive method to analyze samples tendered as evidence in investigation and prosecution of criminal cases. Further, archeological studies to fingerprints artifacts to determine the place of origin as a way to understand the activities of human in the past can also be done along with nutritional epidemiological studies to investigate the contribution of diet, occupation and life styles in chronic diseases (Filby, 1995). In general, the commercial MNSR can be said to be ideal for research and training and education in universities in the areas such as activation analysis and non-destructive testing by neutrons. Training equipment like pneumatic transfer systems, radiation detectors, neutron monitors and reactor simulators are associated facilities of MNSR that assist in teaching nuclear science and technology. The power of this reactor (~3kW) allows training in the areas such as nuclear radiation measurement, neutron transport by using spectrometers, reactor kinetics, relative and absolute flux measurements and spectrum measurements. The MNSR is also a source for thermal and epithermal neutrons, production of weak and short lived isotopes, shielding experiments in experimental channels of thermal column, and measurement of the radiation field in the vicinity of the reactor. Radioactive tracer production, radiation source to demonstrate radiography, criticality experiment and control rod calibration could also be exploited with the reactor. The reactors are designed in such a way that they are user friendly for the purpose of providing training to students and can be operated by individuals under supervision. The reactor is safe with a large negative temperature coefficient, a small excess reactivity (about 3-4 mk), and a safe shutdown system (SCRAM). It also has a shut down margin that guarantees its life span and flexibility. 8. CONCLUSION The thermal-epithermal flux ratio and the epithermal flux shaping factor reported in this work for channels 5 and 7 of the GHARR-1 show a similar trend with the one reported earlier by Akaho and Nyako for channels 1 and 6 of the same reactor and it compares well with those reported 11

13 in the literature for similar reactors. The result is an indication that the flux parameters of the commercial MNSR is stable over a period of time. It also shows that the facility is safe for reactor physics experiments, teaching and training of students and also ideal for application in NAA for the determination of elemental composition of biological and environmental samples. It can also be a useful tool for geochemical and soil fertility mapping. The negative flux deviation parameter predisposes the MNSR as an under-moderated reactor type which is another advantage for reactor experiments under safe operation. The effect of the adjustment of Hydrogen to Uranium ratio in the design of the commercial MNSR-from the prototype value of 24 to 197 leads to a longer life span and a relatively higher negative temperature coefficient of reactivity for the commercial MNSR. This could be seen in the lower flux parameters of the latter as compared to the former (Tables 3 and 4). ACKNOWLEDGMENTS The authors are grateful to the Associate Scheme Office of the Abdus Salam International Centre for Theoretical Physics (ICTP), Trieste, Italy for supporting the scientific visit of the first author to ICTP and for providing the literature and excellent computing facilities for preparation of the manuscript. The authors are also grateful to the International Atomic Energy Agency (IAEA) for supporting the mission of Y.A. Ahmed and I.O.B. Ewa to the Ghana Reactor Centre, GAEC, Accra, Ghana without which this work could not have been possible. The hospitality and kind assistance of entire staff of GHARR-1 Centre is highly acknowledged. REFERENCES Ahmed, Y. A., M.Sc. thesis, 24. Characterization of Irradiation Channels of the GHARR-I Miniature Neutron Source Reactor, Physics Department, Ahmadu Bello University, Zaria, Nigeria. Ahmed, Y. A., Ewa, I. O. B., Umar, I.M., 22. Effective resonance energy and non ideality of epithermal neutron flux distribution in neutron activation analysis, Nigerian Journal of Physics, 14(1), Akaho, E. H. K. and Nyako, B.J. B., 22. Characterization of neutron flux spectra in irradiation channel of MNSR reactor, using the Westcott-formalism for the k neutron activation analysis method, Applied Radiation and Isotopes, 57, Akaho, E.H.K., Maakuu, B. T., Qazi, M.K., 2. Comparison for some measured and calculated nuclear parameters for Ghana Research Reactor-1 core, Journal of Applied Science and Technology, 5 (1-2),

14 Balogun, G. I., Jonah, S. A., Ahmed, Y.A., Sa adu, N., 24. Results of On-site Zero-power and Criticality Experiments for the Nigerian Research Reactor-1, Internal Report CERT/NIRR- 1/ZP/1. Balogun, G. I., 23. Automating some analysis and design calculation of Miniature Neutron Source Reactors at CERT (1), Annals of Nuclear Energy, 3, De Corte, F. and De Wispelaere, A., 23. Recalibration of the irradiation facilities in the Thetis reactor with an examination of the versus E behavior in kev neutron energy range, 257(3), De Corte, F., The k -standardization method a move to optimization of NAA, PhD Thesis, University of Gent, Belgium. De Corte, F., Simonits, A., De Wispelaere, A., Hoste, J., Accuracy and applicability of k method, Journal of Radioanalytical Chemistry, 113, De Corte, F., Moens, L., Simonits, A., Hammami, K.S., De Wispelaere, A., Hoste, J., The effect of epithermal neutron flux distribution on the accuracy of absolute and comparator standardization methods in (n, γ) activation analysis, Journal of Radioanalytical Chemistry, 72, De Corte, F., Hammami, K. S, Moens, L., Simontis, A., DE Wispelarere, A., Hoste, J., The accuracy of the experimental alpha-determination in the 1/E (1+) epithermal reactor neutron spectrum, Journal of Radioanalytical Chemistry, 62, Erdtmann G. and Soyka W., The Gamma Rays of the Radionuclides, Verlag Chemie, Weinheim, New York. Filby, R.H., Isotopic and nuclear analytical techniques in Biological Systems: A Critical study (Part-IX. neutron activation analysis) Technical Report, Pure and Appl. Chem.,67(11), IAEA, 199. International Atomic Energy IAEA-TEC DOC 564: Practical aspect of operating a neutron activation analysis laboratory, IAEA, Vienna. Jonah, S.A., Balogun, G.I, Mati, A.A., Yusuf, I., Ahmed, Y.A., Saad, N., Nkom, B., Oladipo, M.O.A., Umar, I.M., Ogunlaye, P.O., 26. The Nigerian Research Reactor-1(NIRR-1) and its potential Applications in Geosciences and mining, 42 nd Annual International Conference of NMGS, Kaduna, Nigeria. Jonah, S.A., Balogun, G.I., Umar, I.M., Mayaki, M.C., 25. Neutron spectrum parameters in irradiation channels of the Nigeria Research Reactor-1(NIRR-1) for the k -NAA Standardization, Journal of Radioanalytical and Nuclear Chemistry, 266(1),

15 Jovanovic, S., Vukotic, P. Smodis, B., Jacimovic, R., Mihaljevic, N,.Stegnar, P., 1989 Epithermal neutron flux characterizationof the RTIGA II Reactor, Ljubljana, Yogoslavia for use in NAA, Journal of Radioanalytical Chemistry, 129(2), Jovanovic, S., De Cotre, F., Simonits, A., Moens, L., Vukotic, P., Hoste, J., The effective resonance energy as a parameter in (n, γ) activation analysis with reactor neutrons, Journal of Radioanalytical Chemistry, 113(1), Kennedy, G., St. Pierre, J., Wang, K., Zang, Y., Preston, J., Grant, C., Vutchkov, M., 2, Activation constant for SLOWPLOKE and MNS Reactors calculated from the neutron spectrum and k and Q values, Journal of Radioanalytical Chemistry, 245(1), Moens, L., De Corte, F., Simonits, A., Wispelaere, A., Hoste, J., The effective resonance energy E r as a parameter for the correction of resonance integrals, 1/E (1+) epithermal neutron spectra: tabulation of E r values for 96 isotopes, Journal of Radioanalytical Chemistry, 52(2), Mustra, C. O., Freitas, M.C., Almeida, S.M., 23. Neutron flux and associated k parameters in the RPI after the last configuration change, Journal of Radioanalytical and Nuclear Chemistry, 257(3), Nisle, R.G., A unified formulation for the specification of neutron flux spectra in reactors, Neutron Dosimetry, IAEA-TECH REPORT 1, Osae, E.K., Nyako, B.J.B., Serfor-Armah, Y., Darko, E.K., An empirical expression for full energy peak efficiency for an n-type high purity Ge detector calibration, Journal of Radioanalytical nuclear chemistry, 242, Qazi, M.K., Akaho, E. H. K., Maakuu, B. T., Anim-Sampong, S., Nuclear Core Design Analysis of Ghana Research Reactor 1, NNRI Technical Report, GAEC-NNRI-RT-35. Simonitis, A., Jovanovic, S., De Corte, F., Moens, L., Hoste, J., A method for experimental determination of effective resonance energies related to (n, γ) reaction, Journal of Radioanalytical Nuclear Chemistry, 82, 169. Simonits, A., De Cotre, F., Hoste, J., Single comparator methods in reactor neutron activation analysis, Journal of Radioanalytical Nuclear Chemistry, 24, Su-De, T., Multi-elemental analysis of Chinese Biological standard reference materials by Monostandard instrumental neutron activation analysis, Journal of Radioanalytical Nuclear Chemistry, 81, 345. Umar, I.M., 23. The potentials of MNSR in the Socioeconomic Development of Nigeria, ICENS- IAEA Small Research Reactor Workshop, University of West Indies, Jamaika. Valkovic, V., Trace element analysis, Taylor and Francis Ltd, London. 14

16 Wang, L., Multipurpose Gamma Ray spectrum and NAA application software, SPAN 98 users manual, China Institute of Atomic Energy, Beijing, China. Witkowska, E., Szczeppaniak, K., Bizuik, M., 25. Some applications of neutron activation analysis: A review, Journal of Radioanalytical and Nuclear Chemistry, 265(1), Yang, Y.W., MNSR Reactor Complex Manual, CIAE, Beijing, China. 15