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JAERI-Research—98-057

JAERI-Research98-057 JP9901046

MM

CALCULATION OF NEUTRON FLUX CHARACTERISTICSOF DALAT REACTOR USING MCNP4A CODE

October 1 9 9 8

Tran Van HUNG, Yukio SAKAMOTO and Hideshi YASUDA

Japan Atomic Energy Research Institute

3 0 - 0 8 j)

- Hi,

(¥319-1195

T319-1195

This report is issued irregularly.

Inquiries about availability of the reports should be addressed to Research

Information Division, Department of Intellectual Resources, Japan Atomic Energy

Research Institure, Tokai-mura, Naka-gun, Ibaraki-ken 319-1195, Japan.

© Japan Atomic Energy Research Institute, 1998

JAERI-Research 98-057

Calculation of Neutron Flux Characteristics

of Dalat Reactor Using MCNP4A Code

Tran Van HUNG*, Yukio SAKAMOTO and Hideshi YASUDA

Center for Neutron Science

Tokai Research Establishment

Japan Atomic Energy Research Institute

Tokai-mura, Naka-gun, Ibaraki-ken

(Received September 9, 1998)

The neutron flux characteristics of the Dalat reactor such as energy spectra,absolute neutron flux and neutron flux distribution along an irradiation chan-nel were calculated by using MCNP4P code. All computations were done on apersonal computer with the running time about 2 days for every case. Thecalculation configuration is similar to real one of reactor at the nominal power500 kW. The calculated results of neutron flux and a spectrum fittingparameter, a , are in good agreement with experimental ones within 5%.

By using the calculated energy spectra, the cadmium ratios (Red) and effec-tive cross-section of 197Au in the irradiation channels were calculated. In thiscalculation, were used the multi-group cross-sections of 197Au from JapaneseEvaluated Nuclear Data Library (JENDL) and International Reactor DosimetryFile 82 (IRDF82). The comparison of the calculated results shows that : (1)the difference of RCd values between the experiment and calculation usingcross-section of 197Au(n, y ) 198Au reaction is 1 to 6% for IRDF82 and 4 to 8%for JENDL, and (2) the effective cross-sections of 197Au(n, y ) 198Au reactionfrom IRDF82 and JENDL dosimetry files are completely in good agreementwith each other.

Keyword : Calculation, Experiment, Configuration, Distribution, Neutron Flux, Dalat

Reactor, MCNP4A Code, Cadmium Ratio, Neutron Spectrum, Per-

sonal Computer

'STA Scientist Exchange Program (Dalat Nuclear Research Institute, VIETNAM)


MCNP4A 3 - Viz X & Vy y h

Tran Van HUNG"* • Wf #5fc • ^ffl

(1998^9 n 9

D a l a t ^ f t T m # t t ^ ^ ^ i f i S

^ i b , (l)* K 5 •> A

6%. JENDL<7?^-g-tC 4 - 8 % T * 1?, (2)197Au

(n, y) 198Au^S&Bfffi«fi JENDL *fcl

: T 319-1195


Contents

1. Introduction** 12. Calculation 1

2.1 Calculation Code and Computer 1

2.2 Dalat Reactor Description 1

2.3 MCNP Model Description 23. Calculation Reaults and Discussions • 4

3.1 The Neutron Flux in Irradiation Channels 43.2 Energy Spectra 6

4. Conclusion 8Acknowledgments 9

References 9Appendix 21

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1. Introduction

The Dalat reactor is a pool-type research reactor, which uses water as the coolant andthe moderator. It was rebuilt from 250 kW TRIGA MARK II reactor in 1984. By thereconstruction, the nominal power was increased to 500 kW. The present workingconfiguration of the reactor core includes 100 fuel elements, 2 safety rods, 4 shim rods, 1regulating rod, 2 dry irradiation channels, and 2 wet irradiation channels. The central channelis encircled by a beryllium layer, which traps thermal neutron. Therefore, it is called "neutrontrap". The outside of active core is a graphite reflector. All of reactor core is submerged in awater tank of 6.26 m height and 1.98 m diameter.

The characteristics of the neutron flux in irradiation channels have been determined byexperiments. The experimental method and results used in this paper were reported in aprevious work [1]. However, the reliable calculations of these characteristics are necessary toexamine the experimental results, specially, for absolute values. In the experiments, thedetermination of the absolute values of neutron flux in the neutron trap and 1-4 channel have ahindrance, due to the fact that the irradiation sample is placed and removed manually, withoutpneumatic system, so that the accurate control of irradiation time is difficult.

In this report, the neutron flux characteristics such as energy spectra, absolute neutronflux and distribution of neutron flux in irradiation channels are calculated by using a MonteCarlo code.

2. Calculation

2.1 Calculation Code and Computer

The MCNP4A code is chosen for the calculation in this problem. The MCNP4A [2] isa useful nuclear physics computer code and used widely in the world. At present, this code canbe run on a personal computer and is not limited to a super computer since a few years ago [3].

In the present work, the calculations were performed on the personal computer (MMXPentium 200 MHz).

2.2 Dalat Reactor Description

The Dalat reactor is a pool type research reactor. Distilled water is used as moderator,coolant, reflector and biological protection. The cylindrical reactor is arranged symmetricallyby several kinds of material elements. In the center of active region is a neutron trap of radius3.2 cm. The fuel elements, beryllium and graphite blocks, control rods and irradiation channelsare the active region of reactor and form a hexagonal prism lattice in the cylindrical reactorwith radius 20.8 cm and height 60.0 cm. Outside of the active region is a graphite reflectorwith thickness of 32.6 cm. The working configuration of reactor core is shown in Fig. 1. Thestructure and dimension of fuel elements, control rods and others are as follows:

1


2.2.1 Fuel Elements

The structure and dimension of fuel elements are shown in Fig. 2. The fuel element isbundle type and composed of 3 concentric tubes: 2 circular inner tubes and a hexagonal outertube. The fuel tubes are 0.7 mm thick and are wrapped with the 0.9 mm thick aluminumcladding. Outside of the aluminum cladding layer is coolant water. The fuel material isuranium with the enrichment of 235U of 36% and has a density of 1.51 g/cm3. The active lengthof a fuel element is 60 cm.

2.2.2 Control Rods

There are three types of control rods. They are shim, safety and regulating rods. Theouter structure and dimension of control rods are same as those of fuel elements. In Fig.l, the 4shim rods (marked by KC1, KC2, KC3, KC4) and 2 safety rods (marked by AZ1, AZ2) arearranged symmetrically in the active region. The regulating rod is made of stainless steel andlocated at (7,11) element (marked by AR). The shim and safety rods are made of boron carbidewith density of 2.05 g/cm3 wrapped with 1 mm thick aluminum.

2.2.3 Irradiation Channels

The Dalat reactor has two types of irradiation channels; wet and dry irradiation channels.The wet irradiation channel is a cylindrical water column of 1.4 cm radius and encircled by analuminum tube. The dry irradiation channel includes two parts; a central air column of lcmradius and a hexagonal surrounding water layer.

In Fig. 1, the address of an element is indicated by two numbers, one is the order fromleft to right and another is from top to bottom. So that, first irradiation channel is located at(1,4) element and called 1-4 channel. It is a wet irradiation channel. Similarly, 13-2 and 7-1channels are located at (13,2) and (7,1) elements, respectively. They are dry irradiationchannels. The neutron trap is also a wet channel, but it's structure is different from 1-4 channel.The neutron trap is located at the center of the active region. It is water column of 3.2 cmradius and is encircled by the beryllium layer.

2.2.4 Beryllium and Graphite Blocks

Beryllium and graphite blocks are simply solid material with density of 1.875 g/cm3 and1.68 g/cm3, respectively.

2.3 MCNP Model Description

2.3.1 Calculation Model

- 2 -


The reactor core configuration was modeled following the real one realistically.However, for simplicity in model description, the structure of fuel element along the length wasdivided into three zones. The central zone; fuel contained zone, is described accurately andboth end zones are described with the homogenous mixture of aluminum and water with anaverage density of 1.319 g/cm3. The actual and model structures of the fuel elements are shownin Fig.2.

When reactor was operated at 500 kW without the irradiation sample, 4 shim rods wereelevated up to 16 cm from the bottom of active region, therefore, this condition was taken intoaccount. Two safety rods and the regulating rod are pulled up over the active region. Therefore,they are water columns in the calculation model.

The model structures of graphite and beryllium block and graphite reflector arecompletely the same as the actual ones.

The material data used in the calculation model are presented in Table 1.

2.3.2 Source and Tally

In the present problem, KCODE (criticality problem) and KSRC (initial spatialdistribution of source points) cards are used for the source description. In this case, MCNPoffers several options for the Keff value estimate, including the combined average of theabsorption, collision and track-length estimator.

For all calculations of neutron flux and spectra, the F4 tally was used. F4 is a track-length tally card for calculating average neutron flux over a cell. However, in some case, forcomparison, the F5 tally was also used. It is a point detector tally for the calculation of neutronflux at a point in space. It can provide a good estimate of the neutron flux where would bealmost impossible to get by either track-length (F4) or other tallies. The flux at a point cannotbe estimated by a track-length tally (F4) or a surface flux tally (F2), because the probability of atrack entering the volume or crossing the surface of a point is zero.

The experimental results are given in terms of flux, neutron/cm2/s at 500 kW operation,on the other hand, MCNP tallies give fluence in unit, 1 neutron/cm2, caused by one sourceneutron. For the comparison with experiment, the MCNP results can be properly normalizedusing the following conversion factor:

ljoule/s"^ f lMeV ^ f lfission "̂= 3.12xl010(fission/Ws)

W J M.602xl013joule^ *- 200MeV J (1)

For the Dalat reactor at nominal power of 500 kW, the scaling factor is

r l neutron/cm2 •> r2.4source-\f 106W ^ p . 12xlO10 fission"^(0.5 MW) = 3.744xl016n/cm2s

L source J ^ fission J ^ 1 MW ^ W'S ^ (2)

- 3 -


It means that the value of neutron flux (1 neutron/cm2) in OUPUT of MCNP4A has tobe multiplied by the scaling factor 3.744xlO16 for the Dalat reactor at 500 kW operation.

In the calculation, the neutron energy range from 0 to 20 MeV was divided into 112groups.

2.3.3 Cross-section Libraries

Nuclide cross-sections were taken from JENDL-3.2 [4]. In calculation, the S(a,P)thermal treatment was also used for H2O, beryllium and graphite.

A plan view of the calculation configuration obtained from INPUT deck description isshown in Fig. 3. We can confirm that INPUT deck described faithfully the workingconfiguration shown in Fig. 1. An INPUT sample for calculation of energy spectrum, neutronflux and its distribution in neutron trap and 1-4 channel is shown in Appendix.

3 Calculation Results and Discussions

3.1 The Neutron Flux in Irradiation Channels

The calculated keff value was 1.0010 ± 0.0005. The calculation simulated a critical state.Therefore, this value is quite satisfactory. Absolute values of maximal thermal neutron flux at1-4, 13-2, 7-1 channels and neutron trap are presented in Table 2. In this table, neutron flux isevaluated at experimentally maximal positions for neutron trap and 1-4 channel and at thesample irradiation positions for 7-1 and 13-2 channels.

In the experiment [1], the activation method was used for the determination of neutronflux characteristics in irradiation channel. This method is described in detail in paper [5]. Au-foils were used as the neutron flux monitors, because gold has excellent properties as a capturestandard. The sample preparation is particularly easy, since the material is monoisotopic andthe product nucleus has a simple decay scheme, especially, it's cross sections has beendetermined in detail and with high accuracy. For determination of neutron flux distributionalong the vertical irradiation channel, the bare Au-foils were fixed at different positions alongthe length of an organic glass sheet and irradiated for about 5 minutes at the reactor power of500 kW. The gamma-ray intensity of 198Au was measured by a high purity Ge detector of 70cm3 coupled to a multi-channel spectrometer. The relative neutron flux at position (/) wasdetermined by:

(3)T m

_ 4 -


where, A; and Amax are gamma-ray intensities of 198Au at (i) and maximal flux positions,<|)max are neutron flux at position (i) and maximal flux, respectively. In the experiment, themaximal flux position was at 25 cm from the bottom of active region.

Absolute thermal neutron flux was determined by using the formula:

(4)

where m\ weight of gold in foil (g), <yacl; activation cross section of !97Au (barn), No; Avogadronumber (6.023xl023), A; mass number of gold, X ; decay constant of 198Au , /,-„.; irradiation time,tc; cooling time, /„,; measuring time, e, efficiency of detector, y\ emission rate of the gammaray, S; area of the gamma ray peak.

In formula (4), the error of the absolute value of the neutron flux in the experimentdepends on many factors. Therefore, the accurate determination of absolute neutron flux isdifficult, especially, in 1-4 channel and neutron trap due to the inaccurate irradiation time.

As shown in Table 2, the absolute thermal neutron flux in the irradiation channel iscalculated with high level of accuracy. As the calculation check, the F4 and F5 tally cards weresimultaneously used in some calculations. The results obtained from both tallies were similar.For example, when calculating the neutron flux at position of 22 cm from the bottom of activeregion in neutron trap, the calculated results are (1.23±0.07)xl0"3 neutron/cm2 and(1.18±0.007)xl0'3 neutron/cm2 by using F5 and F4 tallies, respectively. The relative uncertaintyof calculated results using F5 tally is ten times larger.

The comparison of calculated and experimental values of the absolute neutron fluxshows that both values are in good agreement with each other. The difference betweencalculated and experimental values are less than 5%. The relative error in the Table 2 is lessthan 1% for MCNP calculation and 5% for experimental data.

The distributions of neutron flux along 1-4 channel and neutron trap are shown in Fig.4and Fig.5, respectively. In these figures, neutron flux distributions are normalized at 30 cmposition. The neutron flux distributions in irradiation channels also correspond to experimentalones by their appearance and positions of maximal values. The maximal flux position in theexperimental data locates at 25 cm from the bottom of the active region. In the MCNPcalculation results, it locates at about 24 cm. In the Fig. 4 and Fig. 5, the appearances of thedistributions in 1-4 channel and neutron trap are quite similar. At the height about 15 cm fromthe bottom of active region, the calculated and experimental neutron fluxes are lower thanfitting curve and the neutron flux distribution is not a symmetric cosine function. If can beexplained by the effect of control rods. During the reactor operation, the control rods werepulled up to 16 cm from the bottom of active region.


3.2 Energy Spectra

The important characteristics of irradiation positions are thermal, epi-thermal neutronfluxes and cc-factor. We know that, in many cases, the epi-thermal neutron flux distribution isnot proportional to 1/E, but rather 1/E1+ a, where a is a small positive or negative constantbetween -0.1 and 0.1. The neutron energy spectrum was calculated at different positions alongthe vertical irradiation channels. In Fig. 6 and Fig. 7 are shown the calculated energy spectra atposition 22cm in the neutron trap and 1-4 channel.

Neutron spectrum at an irradiation position can be expressed as a sum of a thermalequilibrium spectrum; (p,h(E) (Maxwellian distribution, neutron energy from 0 to 0.5 eV) andepi-thermal spectrum in the slowing down region; (pepi(E) (neutron energy from 0.5 eV tomaximal). In the paper [5], the neutron spectra were fit on a semi-empirical function:

cpth(E) = Ith. (E/E\).e-E/ET (5)

andcpepi(E) = Iepi.(E/leVr.l/E (6)

where, I,h(E) and Iepi(E) are scaling constants for the thermal and epi-thermal portion, ET is acharacteristic energy of the Maxwellian portion of the spectrum.

In the present work, I,h(E), Iepj(E), ET and a are determined by fitting the calculatedneutron spectra on Eqs. (5) and (6). The fitting results are shown in Tables 3 and 4.

The relative error of the calculated spectrum intensity in each energy group was lessthan 5%, especially, in thermal neutron region was less than 2%. Consequently, the fittingparameters Ith and ET are obtained with relative error of about 2%. In the epi-thermal region,fitting parameter Iepi was obtained with the similar relative error of 2%, but a factor with therelative error about 10%. From the experimental data, a factor was also obtained with a lessaccuracy. Perhaps, this is because a is small power constant standing for a wide energy region,in addition, the proportion of 1/E1+CC is an approximation in the formula (6). The method for thedetermination of experimental a-factor is described in detail in Ref. [6].

The ET-fitting values in these tables demonstrate that neutron thermalization in neutrontrap is better than in 1-4 channel. This is because the neutron trap is well moderated with waterand beryllium layers.

The comparison of the a-factors in Table 5 shows that the calculated values agree wellwith experimental ones. The difference between both data is 6%.

In order to examine the calculated spectra, the cadmium ratios (R^) of I97Au in theirradiation channels were also calculated. Rcd of an isotope is defined as the ratio of themeasured activities of the sample with and without a cadmium cover. It can be easily evaluatedfrom the experiment by activating monitor foils. Au-foils were used as monitors in thisexperiment. The experimental cadmium ratio at the irradiation position is determined by:

Abar

Red = (?)

6 -

JAERI- Research 98-057

where, Acd and Abarare gamma-ray intensities of 19SAu produced in the irradiated foils with andwithout a cadmium filter, respectively. In the experiment, cadmium filter thickness was 0.8mm.

In the calculation, Rcdof gold can be approximated as follows:

oc

J c (E) . <J>(E) dE0

Rcc = (8)oc

Ja(E).<|>(E)dE

where, a(E) and <|>(E) are (n,y) cross-section of gold and neutron spectrum, Ecd is the cadmiumcut-off energy. In many references, ECd are chosen to about 0.5 eV. However, the chosen cut-off energy Ecd depends on the shape and thickness of the Cd filter. For that reason, in this workthe cadmium ratio is calculated by:

oc

o(E). <|>(E) dE0

Red = (9)oc

Ja(E).r,(E).<t>(E)dE0

and r|(E) = EXP{-N.ot(E).d} (10)

where, r|(E) is the transmission factor of neutron beam of energy E through Cd filter ofthickness d (cm) with N atoms per cm3, at(E) is total cross-section of cadmium.

The calculated result of the neutron transmission factor r|(E) for cadmium with 0.8mmthickness is shown in Fig.8. In this figure, we can see that cadmium filter completely absorbsthermal neutrons, transmitting the neutrons of above leV energy. However, in resonance regionthe neutrons are not completely transmitted through the cadmium filter, especially, at theresonance peaks of cadmium. For comparison, the cadmium ratios were also calculated byusing formula (8) with the chosen cut-off energy ECd = 0.5 eV. However, the difference of Rcd

between formula (8) and (9) were less than 3%.The calculated R^ ratios are presented in Table 6. The cross-sections of gold were

taken from the International Reactor Dosimetry File 82 (IRDF-82) [7,8] and JENDL DosimetryFile [9]. For simplicity, the JENDL Dosimetry File and IRDF-82 are denoted as JENDL and

- 7 -


IRDF in the following descriptions. The multi-group cross-sections of 197Au(n,y)198Au reactionin IRDF and JENDL are shown in Fig.9. In the energy region from 0 to 100 eV, the cross-section values of two libraries are very close each other. But in the energy region above 100eV there is a little difference between them, especially, in the En >3.5 MeV region. However,the difference of cross-section of 197Au in energy region above 3.5 MeV will not affect thecalculated results of cadmium ratios because of the small cross-section and the low flux.

The uncertainty of the calculated RCd ratio should depend on the uncertainties of thecalculated neutron intensity and cross-section of gold in each energy group. However, theerrors in Table 6 are 5 to 8% which takes account only of the uncertainty of calculated neutronintensity. The comparisons of R^ ratios showed that all the calculated results are higher thanthe experimental ones. In the experiment, the Rcd ratio was determined by irradiating two Au-foils of with and without cadmium covers and adjacent each other at a time. The RQJ ratio wasnot paid attention to the thermal neutron flux disturbance. Therefore, the activity of Au-foilwithout cadmium cover is lower. That conduced the cadmium ratio obtained from formula (7)is lower than actual value. Thus, the experimental values of the cadmium ratios must becorrected for the disturbance of neutron field caused by cadmium filter.

The difference of R^ values between the experiment and calculation using cross-sectionof 197Au(n,y)198Au reaction is 1 to 6% for IRDF and 4 to 8% for JENDL, and the differencebetween two calculated values is only 2%. This difference is mainly caused by the differenceof cross-sections in the resonance energy region.

Through the comparison of cadmium ratios between the calculations and experiments, itshould be noted that the energy spectra calculated with MCNP are accurate from this viewpoint.

To evaluate the practical difference of multi-group cross-sections between two libraries,the effective cross-sections of 197Au(n,y)198Au reaction in the neutron trap and 1-4 channel werecalculated by using the calculated neutron spectra and the multi-group cross-sections ofI97Au(n,y)'98Au reaction from IRDF and JENDL following the definition:

oc

J a(E). ((>(E) dE0

a«ff =~ (11)oc

J<KE)dE0

The calculated effective cross-sections are presented in Table 7. The results show thatthe effective cross-sections of 197Au(n,y)198Au reaction in two these libraries agreed well eachother within the difference of 1%.

4. Conclusion

The use of MCNP4A code on a personal computer Pentium 200 MHz, 32 Mbitememory with MMX is very successful for the calculation of the neutron flux and energy


spectrum of Dalat reactor. The obtained results are accurate with the relative error of <1% forneutron flux and 5% for energy spectrum by the running time of the computer from 2 to 3 days.All calculated results are in good agreement with experimental values.

The simplification of the calculation model for the fuel element did not lose thefaithfulness of MCNP simulation performance. The value obtained from these MCNPcalculations can be used to estimate and examine the experimental results.

From the calculation of the cadmium ratios and the effective cross-sections of197Au(n,y)198Au reaction at the irradiation positions, it was revealed that the uses of the multi-group cross-sections from IRDF82 and JENDL Dosimetry File are completely equivalent.

Acknowledgments

The authors would like to thank Dr. Takehiko MUKAIYAMA, director of Center forNeutron Science, JAERI, for his interest and support in all the work. The authors also wouldlike to thank Dr. Makoto TESHIGAWARA, Mr. Tadashi NAGAO, Mr. Shin-ichiro MEIGO,Mr. Yoshihiro NAKANE, Mr. Masanobu IBARAKI for their discussions and effectiveassistance in computations.

References

(1) HUNG. T. V: " Determination of Some Characteristics of Neutron Flux of Dalat Reactorfor Neutron Activation Analysis and Radioisotope Production", Proceedings of the firstNational Conference on Nuclear Physics and Techniques, Hanoi, 15-16 May, p.85 (1996)(In Vietnamese).

(2) BRIESMEISTER R. F: " MCNP -A General Monte Carlo N-Particle Transport Code,Version 4A," LA-12625, Ed., Los Alamos National Laboratory Report (1993).

(3) HENDRICKS J. S., BROCKHOFF R. C: Nucl. Sci. Eng., 116, p. 269-277 (1994).(4) SHIBATA K., et al.: " Japanese Evaluated Nuclear Data Library, Version 3 -JENDL-3-,"

JAERI 1319, Japan Atomic Energy Research Institute (1990).(5) CARPENTER J. M., et al.: Nucl. Instrum. Methods Phys Res., A234, 542 (1985).(6) De COTRTE F., et al.: J. Radioanal. Chem., 62, 209 (1981).(7) GRYNTAKIS E., et al.: Thermal Neutron Cross-sections and Infinite Dilution Resonance

Integrals, Technical Reports Series No. 273, IAEA, VIENNA, p. 199-256 (1987).(8) CULLEN D. E., et al.: Nucl. Sci. Eng.,Vol. 83(4), p.497-503 (1993).(9) NAKAZAWA M., et al.: JENDL Dosimetry File, JAERI 1325, March (1992).

- 9 -


Table 1. Composition of materials used in the MCNP model

Material Unit

Fuel element(Central part)

Fuel element(End parts)

Graphite block

Beryllium block

Coolant

Dry channel

Control rod

Wet channel

Materials

- Uranium-235U_238U

- Aluminum-Al-

-H2O+A1-Al-H- 0

- Graphite-C-Fe-S

- Beryllium-Be

- Water-O-H

-Air-N-0-AT

- Aluminum-Al

-Boron Carbide-10B-HB- C

- Aluminum-Al

- Water-H-O

Symbol

M6

M4

M8

M5

M2

Ml

M7

M4

M3

M4

Ml

Density (g/cm3)

1.51

2.7

1.319

1.68

1.875

1.0

0.0012

2.7

2.05

2.7

1.0

Weight Fraction

0.360.64

1.0

0.38460.068310.5469

0.9950.00340.0016

1.0

0.8890.111

0.78180.20980.0085

1.0

0.68700.08400.2290

1.0

0.8890.111

- 1 0 -

JAERI - Research 98 - 057

Table 2. The calculated and experimental absolute values of thermal neutron flux inthe irradiation channels of Dalat reactor

Irradiation channel Calculated neutronflux (n/cm2/s)

Experimental neutronflux (n/cm2/s)

Cal./Exp.

Neutron trap )

1-4 channel(1)

7-1 channel(2)

13-2channel(3)

(2.10±0.01)xl013

(L21±O.OO6)xlO(6.93 ± 0.04)xl0

(5.1±O.O3)xlO

13

12

12

(2.15±0.08)xl013

(1.25±0.05)xl013

(7.0±0.3)xl012

(4.9 ± 0.2)xl012

(1) Maximum flux point, h = 25cm(2) Irradiation point, h = 8 cm(3) Irradiation point, h = 4 cm

0.980.970.991.04

Table 3. The fitting parameters of neutron spectrum at various heights in neutron trap

Height (cm)

2610141822263034384246505458

Ith(10-Vcm2)

1.27 ±0.022.80 ±0.034.05 ± 0.055.04 ± 0.065.78 ±0.076.29 ± 0.086.31 ±0.096.21 ±0.096.01 ± 0.085.78 ± 0.075.18 ±0.064.46 ± 0.053.46 ± 0.042.37 ± 0.031.03 ±0.02

ET (meV)

26.7 ± 0.326.9 ± 0.326.9 ± 0.327.2 ± 0.327.1 ±0.327.1 ±0.327.1 ±0.327.3 ± 0.327.1+0.327.1 ±0.327.1 ± 0.327.1 ± 0.327.0 + 0.327.2 ± 0.327.3 ±0.3

IepidO"5 n/cm2)

0.36 + 0.070.80 ±0.161.19 ±0.021.53 ±0.031.81 ±0.031.93 ±0.042.06 + 0.041.96 ±0.041.93 ±0.041.81+0.031.52 ±0.031.20 + 0.021.09 + 0.020.70 ± 0.020.28 ± 0.07

a(102)

-4.1 ±0.4-3.8 ±0.4-3.6 ±0.4-3.4 ±0.3-3.2 ±0.3-3.3 ±0.3-3.0 ±0.3-2.9 ±0.3-3.3 ±0.3-3.2 ±0.3-3.4 ±0.3-3.3 ± 0.3-3.5 ±0.3-3.7 ±0.4-.3.9 ±0.4

- 1 1 -


Table 4. The fitting parameters of neutron spectrum at various heights in 1-4 channel

Height (cm)

2610141822263034384654

I,h (104 n/cm2)

0.396 + 0.0060.880 ±0.0101.231 ±0.0121.502 ±0.0161.761 ±0.0181.929 ±0.0201.910 ±0.0201.900 ±0.0201.897 ±0.0191.618 ±0.0171.074 ±0.0110.338 ±0.005

ET (meV)

27.7 ± 0.328.0 ±0.327.9 ± 0.328.2 ±0.328.0 + 0.327.8 ± 0.328.2 ± 0.328.2 ±0.328.1 ±0.327.9 ± 0.328.3 ±0.328.0 ± 0.3

Iep (lO^n/cm2)

2.33 ±0.055.30 ±0.117.74 ±0.159.85 ± 0.2011.9 ±0.2213.6 ±0.2615.5 + 0.2814.7 ±0.2712.8 ±0.2410.0 ± 0.206.35 ±0.121.93 ±0.04

a (102)

-4.3 ±0.4-3.8 + 0.4-3.3 ±0.3-3.2 ±0.3-3.1+0.3-3.4 ±0.3-3.3+0.2-3.6 + 0.3-3.8+0.3-3.7 ±0.3-3.9 + 0.4-4.1+0.4

Table 5 a-factors in irradiation channels determined fromMCNP calculation and experiment

Irradiation channels

Neutron trap(*}

1-4 channel^

7-1 channel(+)

13-2channel(+)

Calculated a (102)

-3.3±0.3

-3.4 ±0.4

-4.8 ± 0.5

-7.2 ± 0.7

Experimental a (102)

-3.1±0.3

-3.6 + 0.4

-4.5 ±0.5

-7.5 ±0.8

Cal/Exp.

1.06

0.94

1.06

0.96

At position of maximal neutron fluxAt irradiation position

- 1 2 -


Table 6. Calculated and experimental cadmium ratios of gold in irradiation channels

Irradiationchannels

Neutron trap

1-4 channel

7-1 channel

13-2 channel

Calculated[RcdliRDF

2.65 ± 0.08

1.73 ±0.05

1.92 ±0.06

2.33 + 0.07

Calculated[RcdJjENDL

2.75 ± 0.08

1.77 ±0.05

1.98 ±0.06

2.38 ± 0.07

Experimental[Rcdtaxp

2.55 + 0.07

1.65 + 0.04

1.90 ±0.05

2.20 ± 0.06

rRcdllRDF

[RcdlEXP

1.06

1.05

1.01

1.06

[RrdliRNni,

[Rcd]EXP

1.08

1.07

1.04

1.08

and [RcdJiENDL are cadmium ratio of gold calculated using cross-section from IRDF82 and JENDLlibraries, respectively.

is experimental cadmium ratio.

Table 7. The calculated effective cross-section of 197Au(n,y)198Au reaction inneutron trap and 1-4 channel using the multi-groups cross-section from

IRDF82 and JENDL

Irradiation channels

Neutron trap

1-4 channel

[C?eff]jENDL

79.41

64.90

[<7eff][RDF

78.92

64.33

Dif.(%)

0.62

0.89

- 1 3 -


Element (1,1)

Element (2,

Fig. 1 Working configaratioa of Dalat reactor

O Fuel element

KC) \ Control rod

Irradiation channel

Regulating rod

Beryllium block

Graphite block

Neutron trap

- 1 4 -


Actual Model

Water + Al composition

Water+ A1 composition

500

Fig. 2 The fuel element structure of Dalat reactor

( Actual structure and calculation model)

- 1 5 -


Graphite

Fig. 3 Calculation configuration received from MCNP model description

Fuel element

Control rod

Irradiation channel

Beryllium block

Beryllium

T Neutron trap

- 1 6

JAERI- Research 98 - 057

II

1.2

1 -

0 . 8 -

0 . 6 -

0.4

0.2

0

1 ' ' '-

- - /

\

1 I > > 1 1 1 1

- ^

1 1 1 1

tET

X,r JSE-*

— * — MCNP4A calculation- •— Experiment

. . . . J

, , , , i , , , , i , , , . i , . , ,

' ' ' '

-

X :' ^V l —

, , , , •

10 20 30 40h(cm)

50 60

Fig.4 Distribution of neutron flux in neutron trap(MCNP calculation and experiment)

MCNP4A calculationExperimental

0 10 20 30 40 50h(cm)

60

Fig.5 Distribution of neutron flux in 1-4 channel(MCNP4A calculation and experimental)

- 1 7 -


I I I I inii| i-4-iJiTn] I I I IIIIII—i i i mill—I I 1111 ill—I I I null—i I I IIIIII—i I I mill—I I 11 MI li—i i i MINI—i i HUB- i ' T^~L ' ' ' —

1 100

s

| 0 .01

&0.0001

10"D r

10',"8 F i l l llllll 1 I I llllll 1 I I llllll I I I mill I I I mill I I I I I I I I I Ill I I I llllll I ll il I I H U M

10-9 10-7 10.-5 0.001Energy (MeV)

0.1 10

Fig.6 Calculated neutron spectrum in neutron trap(h = 22cm)

10"c

I 100s

1 r

£0.01M

0.0001

6 -10° r

10"

s i i i [inij i i 11 liiii i i i inn

mill I

E

i

I

1 llllll

\

\

•HIM

1

=

MINI i

\

\

i m i , , ,

1 1 1 HIM

I I I lllll

1 1 1 ll l l l

1 1 1 Mil l

1 1 1 l l l l l 1 1 1 ll l l l

i—|

1 1 1 lll l l

1-nO" ^

, , i,,,,

i I I inn

• ^ - ^

i i i i n n

i I Mini

\ - - -

L

i i i inn

i i l i n e

i IIIIII

E

s

1

1 lll

lll

a

i niim

I

10-9 107 10-5 0.001Energy (MeV)

0.1 10

Fig.7 Calculated neutron spectrum in 1-4 channel(h = 22cm)

- 1 8 -


..1.2o

I 1CO

CO

I-I-0.8

0.6

0.4

0.2

0

-0.2

i i i i mi

-

-

-

_

1 1 1 M i l t

1 1 1 1 1 III

(i\...

1 1 1 1 Mi l

1 1

j

mi

/

1 1 1 1 ,,,

1 '

!

1 1 1 1 1 1 II

i i i I mi

n.i

i

i M M in 11111 in

< 1 1

1 1 1 1 Illl

—e

-

i i

i

-

-

Energy (MeV)

10"9 10'8 10'7 10'6 10"5 0.0001 0.001 0.01 0.1

Fig.8 Neutron transmission factor of cadmium filter(thickness 0.8 mm)

- 1 9 -


1000

100

" " " I ' ""'" ' ' " " " I ' ' " " " I ' I ' ' I ' ' ' I I ' I ' '"""I "I

ill i i Mint i i linn i i IMIII i i Mini! t i 11nill i < i i linn! J 11 Mm i i

10

0.1

0.01

10-6 0.0001 0.01 100 104

Energy (eV)

106

Fig.9 Cross-sections of gold from JENDL and IRDF82

(-•— JENDL, -B— IRDF82 )

2 0 -


Appendix An INPUT sample for calculation of energy spectrumand neutron flux in neutron trap and 1-4 channel

name=inp311- Probl - Calculation neutron flux and it's energy spectrum using MCNP4A code2-3-4-5-6-7-8-9-10-11-12-13-14-15-16-17-18-19-20-21-22-23-24-25-26-27-28-29-30-31-32-33-34-35-36-37-38-39-40-41

123456789101112131415161718192021222324252627282930313233343536373839

111111111111111111111111111111187785005

-1.0 -10-10120-1.0 -10 101-102-1.0 -10 102-103-1.0 -10 103-104-1.0 -10 104-105-1.0 -10 105-106-1.0 -10 106-107-1.0 -10 107-108-1.0 -10 108-109-1.0 -10 109-110-1.0 -10 110-111-1.0 -10 111-112-1.0 -10 112-113-1.0 -10 113-114-1.0 -10-30-1.0 10-120-30-1.0 -40 20-101-1.0 ̂ 0 101 -102-1.0 -40 102-103-1.0-40 103-104-1.0 -40 104-105-1.0 -40 105-106-1.0 ̂ 0 106-107-1.0 -40 107-108-1.0 -40 108-109-1.0 ̂ 0 109-110-1.0 -40 110-111-1.0-40 111-112-1.0 -40 112-113-1.0 -40 113-114-1.0 ̂ K) 114-30-1.319 -50 12-20-0.0012-60 20-30-0.0012-70 20-30-1.319 -50 30-13-1.68 20-30 50-100-50 20 -30 1 40 60 70 FILL=1100:-12: 13-1.68 -51 52 -53 54 55 -56 U=l LAT=2 FILL=-7:9 -10:6 0:01 16R

- 2 1 -


42434445464748495051525354555657585960616263 •64656667686970717273747576777879

808182838485

C4041C183184185186187188189190191192193194C205206C210211

11012132021

1 16R1 8R 2 4 2 2R 1 2R16R2 7R 1 1R15R23R423R1 1R1 4R 2 1R 3 2 3R 3 2 1R 1 1R13R2 10R1 1R12R523R51R23R51 1R1 2R 4 2 2R 5 2R 2 2R 4 1 2R1 1R523R5 1R23R512R1 1R 2 10R 1 3R1 1R2 1R323R32 1R 14R1 1R23R423R15R1 1R 2 7R 1 6R1 2R 2 2R 4 2 1 8R1 16R1 16RUniverse 2: structure of control rod3 -2.05 -80 81 u=31 -1.0 80: -81 u=3Universe 3: structure of fuel rod1 -1.0 -31u=24 -2.7 31-32u=26 -1.51 32-33u=24 -2.7 33-34u=21 -1.0 34-35u=24 -2.7 35-36u=26 -1.51 36-37u=24 -2.7 37-38u=21 -1.0 38-61-63 65 62 64-66 u=24 -2.7 (61:-62:63:-64:-65:66) 71 72 -73 74 75 -76 u=26 -1.51 (71:-72:73:-74:-75:76) -91 92 -93 94 95 -96 u=24 -2.7 91:-92:93:-94:-95:96u=2Universe 4: Water channel1 -1.0 -222 u=44 -2.7 222 u=4Universe 5: beryllium rod2 -1.875 -21 22 -23 24 25 -26 u=55 -1.68 21:-22:23:-24:-25:26u=5

c/z 0 -5.542725174 3.2c/z 0 -5.542725174 1.0pz -5.0pz 63pz 0.0px 1.59

- 2 2 -


8687888990919293949596979899100101102103104105106107108109110111112113114115116117118119120121122123124125126127128129130

2223242526303132333435363738405051525354555660616263646566707172737475768081919293949596100

pxPPPPpzczczczczczczczczc/zc/zpzpzppppc/zpxpxppppc/zpxpxppppczpzPXpxppppc/z

-1.591.0 1.732 0 3.191.0 1.732 0 -3.191.0 -1.732 0 -3.191.0 -1.732 0 3.1960.00.170.270.330.420.760.850.921.013.2 11.08545035 1.00.0 -5.542725174 20.81.6-1.61.0 1.732 0 3.201.0 1.732 0 -3.21.0 -1.732 0-3.21.0 -1.732 0 3.2-16.0 -5.542725174 1.1.35-1.351.0 1.732 0 2.911.0 1.732 0 -2.911.0 -1.732 0 -2.911.0 -1.732 0 2.91-3.2-22.1709 1.01.44-1.441.0 1.732 0 3.011.0 1.732 0 -3.011.0 -1.732 0 -3.011.0 -1.732 0 3.010.716.01.51-1.511.0 1.732 0 3.091.0 1.732 0 -3.091.0 -1.732 0 -3.091.0 -1.732 0 3.09-5.542725174 30.8

- 2 3 -


131 101 pz 4.0132 102 pz 8.0133 103 pz 12.0134 104 pz 16.0135 105 pz 20.0136 106 pz 24.0137 107 pz 28.0138 108 pz 32.0139 109 pz 36.0140 110 pz 40.0141 111 pz 44.0142 112 pz 48.0143 113 pz 52.0144 114 pz 56.0145 222 cz 1.51

146 imp:n 1 36r0 1 18r147 ml 1001.37c 2 8016.37c 1148 m2 4009.37c 1149 m3 5010.37c-0.6870 5011.37c-0.0840 6012.37c-0.2290150 m4 13027.37c 1151 m5 6012.37c 1152 m6 92235.37c -0.36 92238.37c -0.64153 ml 1001.37c 3 8016.37c 1154 m8 13027.37c-0.3846 1001.37c -0.06831 8016.37c -0.5469155 mtl lwtr.Olt156 mt2 be.Olt157 mt5 grph.Olt158 kcode 4000 1 40 3000159 ksrc 1.5 11.08545 0.1 -1.5 11.08545 20 0.27 11.08545 11-0.27

11.08545 11160 eO 0.00le-6 0.003e-6 0.006e-6

0.01e-6 0.02e-6 0.03e-6 0.04e-6 0.05e-6 0.06e-6 0.07e-6 0.08e-60.09e-6 O.le-6 0.15e-6 0.2e-6 0.25e-6 0.3e-6 0.35e-6 0.4e-60.45e-6 0.5e-6 0.6e-6 0.7e-6 0.8e-6 0.9e-6 le-6 1.5e-6 2e-62.5e-6 3e-6 3.5e-6 4e-6 4.5e-6 5e-6 5.5e-6 6e-6 6.5e-6 7e-68e-6 9e-6 le-5 1.5e-5 2e-5 2.5e-5 3e-5 3.5e-5 4e-5 4.5e-5 5e-56e-5 7e-5 8e-5 9e-5 le-4 1.5e-4 2e-4 2.5e-4 3e-4 3.5e-4 4e-45e-4 6e-4 7e-4 8e-4 9e-4 le-3 1.5e-3 2e-3 2.5e-3 3e-3 3.5e-34e-3 5e-3 6e-3 7e-3 8e-3 9e-3 le-2 1.5e-2 2e-2 2.5e-2 3e-2 3.5e-24e-2 5e-2 6e-2 7e-2 8e-2 9e-2 le-1 1.5e-l 2e-l 2.5e-l 3e-l 3.5e-l4e-l 5e-l 6e-l 7e-l 8e-l 9e-l 1.0 1.5 2.0 2.5 3.0 3.5 4.0 5.0 6.07.0 8.0 9.0 10.0 20.0

161 f4:n 12 3 4 5 6 7 8 9 10 1112 13 14 1517 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33

24-


In the above geometrical description, the co-ordinate origin is chosen at thecenter of (5,6) element.

The cells from cell 1 to cell 35 describe ones used in the calculation of neutronspectra and flux (neutron trap, 1-4 and 13-2 channels). Cell 37 describes graphitereflector. Cell 39 describes structure of reactor core and materials. The detailedstructures of materials in the reactor core, including control rods, fuel rods, beryllium,graphite blocks and water are described in cells from 40 to 221.

- 2 5 -

am(si)t«i;

m.s sK mm m

« i i

P̂ Si fti (* ft

*

T

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7

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7

b"

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r

71/

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Ty

7U

7

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y

IE ^

mkgs

A

Kmolcd

radsr

mEX

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m58A-b

ftM

ftiR

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& IK? ? y x9 x S S

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s"1

J/kgJ/kg

i; ^ h ,u

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min, h, d

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leV=1.60218x]0-"J1 u= 1.66054x10"" kg

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w y h

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y

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barGalCiR

radrein

1 A=0.1nm=10-'0mlb=100fm2=10-28m2

lbar=0.1MPa=10sPa

lCi=3.7xlO'°BqlR=2.58xlO-'C/kglrad = lcGy=10-2Gy1 rem=lcSv=10~2Sv

1 0 "

1O 1 S

10'2

109

106

103

102

10'

10"'io-2

10"3

10"6

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J0-18

mmX ?/<

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J ~J

3. barli,

T-fr,

r, barnfcj;

N( = 10sdyn)

1

9.80665

4.44822

kgf

0.101972

1

0.453592

lbf

0.224809

2.20462

1

tt S 1 Pa-s(N-s/m2)=10P(#TX')(g/(cm-s))

lmVs=10'St(x h -?x)(cm !/s)

E

•ft

MPa(=10bar )

1

0.0980665

0.101325

1.33322 x 10-'

6.89476 x 10"3

kgf/cm2

10.1972

1

1.03323

1.35951 x 10"3

7.03070 x 10-2

atm

9.86923

0.967841

1

1.31579 x lO'3

6.80460 x 10-2

mmHg(Torr)

7.50062 x 103

735.559

760

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51.7149

Ibf/in2(psi)

145.038

14.2233

14.6959

1.93368 x 10-2

1

X

*IV

¥1

tt

ft

J( = 107erg)

1

9.80665

3.6 x 106

4.18605

1055.06

1.35582

1.60218x10-"

kgf*m

0.101972

1

3.67098 x 10s

0.426858

107.586

0.138255

1.63377 x 10"M

kW- h

2.77778 x 10'7

2.72407 x lO"6

1

1.16279 x 10"'

2.93072 x 10-'

3.76616 x 10"'

4.45050 x 10""

cal (l+ftiS)

0.238889

2.34270

8.59999 x 10s

1

252.042

0.323890

3.82743 x 10"20

Btu

9.47813 x 10"'

9.29487 x lO"3

3412.13

3.96759 xlO"3

1

1.28506 x 10"3

1.51857 xlO-22

ft • lbf

0.737562

7.23301

2.65522 x 106

3.08747

778.172

1

1.18171 xlO"19

eV

6.24150x10"

6.12082x10"

2.24694 xlO2S

2.61272 x ) 0 "

6.58515 x l O 2 '

8.46233 x 10'8

1

1 cal = 4.18605 J(stJftffi)

= 4.184J GSMt¥)

= 4.1855 J (15 X)

<±mm i PS

= 75 kgl-m/s

= 735.499 W

Bq

3.7 x 10"

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2.70270 x 10-

1

Gy

1

0.01

rad

100

1

C/kg

2.58 x 10"'

R

3876

1

Sv

1

0.01

rem

100

1

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