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Progress in Nuclear Energy 51 (2009) 334–338

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Progress in Nuclear Energy

journal homepage: www.elsevier .com/locate/pnucene

Influence of evaluated data of fission product poisons on criticality

Ammar Ahmad, Siraj-ul-Islam Ahmad*, Nasir Ahmad, Khurrum Saleem Chaudri,Tasveer Muhammad Sahibzada, Masroor AhmadPakistan Institute of Engineering and Applied Sciences, Islamabad 45650, Pakistan

Keywords:Fission productsCross-section evaluationsNeutron poisons

* Corresponding author. Tel.: þ92 51 9209032.E-mail address: sirajisl@yahoo.co.uk (S.-I. Ahmad)

0149-1970/$ – see front matter � 2008 Elsevier Ltd.doi:10.1016/j.pnucene.2008.06.004

a b s t r a c t

The main objective of this research is to study the influence of cross-section differences of fission productpoisons among various newly released evaluated cross-section libraries ENDFB-VI.8, JENDL3.2, JEF2.2,IAEA, ENDFB-VII and JEFF3.1 on criticality of an MTR type research reactor. The effect of cross-sections ofpoisons on the reactivity was studied with the help of WIMSD and CITATION codes. Various cross-sectionlibraries were used in SARC (System for Analysis of Reactor Core) code for the fuel cycle analysis. It wasfound that the negative reactivity induced due to 135Xe for the equilibrium core is around 36.00 mkwhereas for 149Sm it ranges from 6.65 to 7.06 mk. The three libraries (JENDL3.2, IAEA and ENDFB-VII)resulted in small increase in the Xenon worth as compared to the other three libraries. For Samarium,JEFF3.1 gives the highest worth whereas ENDFB-VI.8 gives the least worth among the six libraries.

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1. Introduction

During reactor operation, the concentration of fission productsand actinides continuously changes with time and space due to fueldepletion and nuclear reactions of neutrons with other materials. Inthe reactor core, actinides are produced as a result of neutroncapture by heavy atoms. Among these, the fissile isotopes such as239Pu and 241Pu contribute positively in reactivity and henceincrease the operating life of the system. On the other hand thefission fragments generated at the time of fission, decay to producea variety of fission products. Fission products are of importance inreactors because they become parasitic absorbers of neutrons andresult in long term sources of heat.

The materials that capture neutrons without leading to anysubsequent fission are called neutron poisons. They can be eithernaturally occurring elements or produced in the core due tofission, e.g. Boron and Cadmium are naturally occurring elementswhereas 135Xe and 149Sm are the fission products. Although neu-tron absorption cross-sections for several fission products havesignificant effect on reactivity, 135Xe and 149Sm have the mostconsiderable impact on reactor design and operation due to theirrelatively high fission yield and absorption cross-sections forthermal neutrons. As they remove neutrons from the reactor, theyhave an impact on the thermal utilization factor and thusreactivity (Integrated Publishing, 2007). As a result, the presenceof these isotopes decreases the reactivity and hence the operating

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life of the reactor. The code WIMSD has been used for theoreticalcomputations in the work related to material test reactors likecore neutronics, conversion from HEU to LEU fuel, thermalhydraulics, transient analysis, and determination of the neutronenergy spectrum. Most of these calculations have been carried outby using old 1981 cross-section library which is of UK origin.Release of different WIMSD libraries by IAEA which are based onrecent cross-section data from ENDFB-VI.8 (McLane et al., 1995),JENDL3.2 (Nakagawa et al., 1995), JEF2.2 (IAEA, 1993), IAEA,ENDFB-VII and JEFF3.1 (Nuclear Energy Agency, 2007) enabled theusers of WIMSD to use new cross-section evaluations for theirresearch and analysis. Some parameters have been analyzed withthe new evaluations, e.g. the effects on reactivity and neutronenergy spectrum have been studied in the article by Ahmad et al.(2004) and Ahmad and Ahmad (2005), the core burnup has beenanalyzed in Ahmad and Ahmad (2006a). 135Xe and 149Sm are ofthe primarily concerned fission product poisons having theirstrong impact on reactivity and reactor operation. This research isrelated to the study of influence of these cross-sections of fissionproduct poisons like 135Xe and 149Sm on reactivity during opera-tion of typical material test reactor. It has been observed that thecross-section differences of materials among various nuclear datalibraries have strong impact on the prediction of actinides con-centration in burnt fuel (Ahmad and Ahmad, 2006b).

The analyses were carried out using SARC code (System forAnalysis of Reactor Core) (Ahmad and Ahmad, 2006a). The fuelcycle analysis was carried out for equilibrium core of a typicalmaterial test research reactor. Negative worth of 135Xe and 149Smwas analyzed by using the data of these fission products from sixlibraries.

F E D C B A

5 290.00 233.32 163.91 248.28 261.36 277.83

6 267.07 253.40 220.59 237.40 137.53 274.96

7 245.56 126.62 218.15 WB 230.82 290.00

8 259.70 230.63 215.05 213.19 114.88 277.06

9 290.00 276.35 150.98 247.63 262.51 290.00

Fig. 1. U-235 contents (g) at BOC.

F E D C B A

5 277.86 220.63 150.97 233.39 248.33 267.13

6 253.47 237.45 204.63 218.15 126.64 261.38

7 230.64 114.86 198.16 WB 213.16 274.95

8 245.58 215.07 198.31 194.32 104.94 262.48

9 276.34 259.67 137.44 230.76 247.59 277.04

Fig. 2. U-235 contents (g) at EOC.

A. Ahmad et al. / Progress in Nuclear Energy 51 (2009) 334–338 335

2. Material and methods

2.1. Reactor description

The core used in the analysis is the equilibrium core of a typicalmaterial test reactor. 235U content in equilibrium core is shown inFig. 1. There are 24 standard fuel elements, 5 control fuel elementsand 2 flux traps. The detailed description of the equilibrium core isgiven in Table 1. The total Uranium per standard fuel element is1451 g of which 290 g is 235U. The fuel is uniformly distributedamong 23 flat plates. The physical dimensions of the fuel elementare 79.63 mm� 75.92 mm, so that there is a water gap of 1.19 mmbetween the side plates of two adjacent fuel elements. Similarly,there is a water gap of 1.37 mm between the two fuel elements inthe direction perpendicular to the fuel plates (Khan et al., 1992;Ahmad et al., 2004).

The plates in CFE are similar to those of the plates of SFE. In thecontrol region, i.e. between the guide plates, the side plates have anincreased thickness (0.7085 cm instead of 0.45 cm) to obtain goodcoolant flow distribution. The overall physical dimension of controlfuel element is same as that of standard fuel element. The Uraniumloading of control fuel element is 820 g, of which 163.9 g is 235U whichis uniformly distributed among 13 flat plates (Khan et al., 1992).

2.2. Methodology

Burnup analysis has been carried out using the code SARC. Thedetail for this code can be seen elsewhere (Ahmad and Ahmad,2006a). Initially the analysis was done using JENDL3.2 library. Tosee the effects of 135Xe, the cross-sections of 135Xe from different

Table 1Description for equilibrium core.

Weight of 235U at BOC 6804.9 gWeight of 238U at BOC 31,034.4 gNumber of standard fuel element 24Number of control fuel element 5Number of flux trap 2Core dimensionsActive height 60.000 cmLength 46.266 cmWidth 40.500 cmNumber of fresh fuel elements 5Number of fuel elements (one cycle burnt) 5Number of fuel elements (two cycle burnt) 5Number of fuel elements (three cycle burnt) 5Number of fuel elements (four cycle burnt) 5Number of fuel elements (five cycle burnt) 4Reflector Water

libraries were used in JENDL3.2 library, and negative worth for135Xe was calculated using SARC. Similarly cross-sections of 149Smfrom other libraries were replaced in JENDL3.2 library; then againby using SARC negative worth for 149Sm was calculated. Theamount of 235U in the core at BOC (beginning of cycle) and EOC (endof cycle) is shown in Figs. 1 and 2, respectively. The minor differ-ences with the earlier results reported by Ahmad and Ahmad(2006a) are due to different number of small burnup steps takennear BOC. The equilibrium 135Xe and 149Sm concentration in thecore is also shown in Figs. 3 and 4.

3. Results

3.1. Effect of 135Xe cross-sections

Firstly the worth of 135Xe was calculated using various librariesand is tabulated in Table 2. It can be seen in the table that worth for135Xe is nearly same for all the libraries with minor differences.Three libraries, i.e. JENDL3.2, IAEA and ENDFB-VII, are showingsimilar results while the results from other three libraries (ENDFB-VI.8, JEF2.2, and JEFF3.1) are same. To analyze this difference inresults, the absorption cross-sections of 135Xe from various WIMSDlibraries were plotted and are shown in Fig. 5. From figure, it can beseen that three libraries JENDL3.2, IAEA and ENDFB-VII have sameabsorption cross-sections for 135Xe taken from same evaluation(evaluated by JNDC FP Nuclear Data W.G.) while absorption cross-sections from other three libraries are same and are taken from

F E D C B A

5 1.510 1.330 0.953 1.430 1.440 1.430

6 1.480 1.470 1.330 1.430 0.822 1.500

7 1.420 0.775 1.350 WB 1.390 1.580

8 1.460 1.370 1.31 1.320 0.708 1.530

9 1.550 1.570 0.900 1.450 1.480 1.530

Fig. 3. Mass of Xe-135 (mg) at EOC.

F E D C B A

5 14.30 13.60 8.060 13.90 14.10 14.40

6 14.20 13.90 13.30 13.60 7.740 14.30

7 13.80 7.500 13.20 WB 13.50 14.40

8 14.10 13.50 13.10 13.10 7.220 14.30

9 14.40 14.30 7.920 13.80 14.10 14.40

Fig. 4. Mass of Sm-149 (mg) at EOC.

10-2 10-1 100 101 102 103 104 105 106 10710-510-410-310-210-1100101102103104105106107108

10-2 10-1 100 101 102 103 104 105 106 107

10-5

10-410-310-210-1100101102103104105106107108

Ab

so

rp

tio

n C

ro

ss-sectio

n (b

arn

s)

Energy(eV)

ENDFB6ENDFB7IAEAJEF22

JEFF31JENDL

Fig. 5. Absorption cross-sections of Xe-135.

10-2 10-1 100 101 102 103 104 105 106 107

0

2

4

6

8

1010-2 10-1 100 101 102 103 104 105 106 107

0

2

4

6

8

10

Ratio

o

f ab

so

rp

tio

n cro

ss-sectio

n

Energy (eV)

Fig. 6. Ratio of absorption cross-sections of Xe-135 obtained using JEF2.2 to the ab-sorption cross-sections obtained using IAEA.

Table 2Xe-worth by changing Xe-135 cross-sections.

Library 135Xe-worth (mk) Ratio of 135Xe-worth with JENDL3.2

JENDL3.2 36.05 1.000IAEAENDFB-VII

JEFF3.1 36.00 0.998JEF2.2ENDFB-VI.8

Table 3Sm-worth by changing Sm-149 cross-sections.

Library 149Sm-worth (mk) Ratio of 149Sm-worth with JENDL3.2

JENDL 6.71 1.000IAEA 6.86 1.022ENDFB-VII 6.65 0.991ENDFB-VI.8 6.86 1.022JEF2.2 6.85 1.020JEFF3.1 7.06 1.052

Table 4Absorption cross-section for 135Xe.

Library One groupeffective absorptioncross-sections (sa barns)

Ratio of absorptioncross-section withJENDL3.2

JENDL3.2 2.34� 106 1.000IAEAENDFB-VII

ENDFB-VI.8 2.32� 106 0.991JEF2.2JEFF3.1

A. Ahmad et al. / Progress in Nuclear Energy 51 (2009) 334–338336

ENDB-VI evaluations. The difference in absorption cross-sectionsamong two sets of cross-sections is plotted in Fig. 6.

To analyze the differences more clearly we have given repre-sentative one group effective cross-section for fuel element D6, thiselement has 220.59 g of 235U, 1152.95 g of 238U, 1.402 mg of 135Xe

10-2 10-1 100 101 102 103 104 105 106 10710-3

10-2

10-1

100

101

102

103

104

105

10610-2 10-1 100 101 102 103 104 105 106 107

10-3

10-2

10-1

100

101

102

103

104

105

106

Ab

so

rp

tio

n C

ro

ss-sectio

n (b

arn

s)

Energy(eV)

JEF22 ENDFB6

JEFF31 ENDFB7

IAEA JENDL

Fig. 7. Absorption cross-section of Sm-149.

10-2 10-1 100 101 102 103 104 105 106 1070.5

1.0

1.5

2.0

2.5

3.0

3.510-2 10-1 100 101 102 103 104 105 106 107

0.5

1.0

1.5

2.0

2.5

3.0

3.5a

Ratio

o

f ab

so

rp

tio

n cro

ss-sectio

n

Ratio

o

f ab

so

rp

tio

n cro

ss-sectio

n

Energy(eV) Energy(eV)

10-2 10-1 100 101 102 103 104 105 106 107

0.995

1.000

1.005

1.010

1.015

1.020

1.025

1.030

10-2 10-1 100 101 102 103 104 105 106 107

0.995

1.000

1.005

1.010

1.015

1.020

1.025

1.030

b

Fig. 8. Ratio of absorption cross-sections of Sm-149 obtained using (a) JEF2.2 and (b) JEFF3.1 to the absorption cross-sections obtained using IAEA library.

A. Ahmad et al. / Progress in Nuclear Energy 51 (2009) 334–338 337

and 13.58 mg of 149Sm at BOC. The power density for this element is244.14 MW/TE. The effective one group cross-section for 135Xe isgiven in Table 4. It can be seen from Table 2 that the ratio of 135Xe-worth from two sets is 0.998. The ratio of absorption cross-sectionsis 0.991. It is clear that the difference is due to absorption cross-section of 135Xe. However, it should be noted that we have used onegroup effective cross-sections for only one element to discuss thedifferences, whereas different elements in the core have differentflux spectrum resulting in differences in the effective one groupabsorption cross-sections for different elements.

3.2. Effect of 149Sm cross-sections

The 149Sm induced negative reactivity is given in Table 3, whichshows that negative worth of 149Sm calculated from all libraries isdifferent. The absorption cross-sections form all the libraries wereplotted and are shown in Fig. 7 which shows that the absorptioncross-sections from all the libraries are different. This difference canalso be seen in Figs. 8 and 9. In these figures ratio of absorptioncross-sections of 149Sm from different libraries to IAEA library isplotted. The significant difference in absorption cross-sections of149Sm among all libraries can be seen there. The differences incalculated values of negative worth of 149Sm are due to the differ-ence of one group effective absorption cross-sections of 149Sm inseveral libraries. To analyze the differences we have calculated

10-2 10-1 100 101 102 103 104 105 106 1070.6

0.7

0.8

0.9

1.0

1.1

1.2

1.3

1.4

1.510-2 10-1 100 101 102 103 104 105 106 107

0.6

0.7

0.8

0.9

1.0

1.1

1.2

1.3

1.4

1.5a

Ratio

o

f A

bso

rp

tio

n cro

ss-sectio

n

Energy(eV)

Fig. 9. Ratio of absorption cross-sections of Sm-149 obtained using (a) ENDFB-VI

effective one group absorption cross-sections from all libraries forthe element D6 of the core. These one group effective absorptioncross-sections are tabulated in Table 5. ENDFB-VI.8 and IAEA havesame absorption cross-sections, but difference in negative worthcalculated from these two libraries is due to the difference of onegroup effective microscopic scattering cross-sections of these twolibraries. The one group effective microscopic scattering cross-section for IAEA and ENDFB-VI.8 library is 104.02 and 104.04 barns,respectively. The ratio of effective one group scattering cross-sections for these two libraries is 0.999. The ratios of 149Sm-worthand effective one group absorption cross-sections from all the li-braries to JENDL3.2 are tabulated in Tables 3 and 5. The differencebetween the ratios of respective libraries is due to the difference ofabsorption cross-sections. Moreover, one group effective cross-sections for only one element were calculated which explains thedifference between the ratio of 149Sm-worth and effective onegroup absorption cross-sections for respective libraries. JEFF3.1library has greater one group effective absorption cross-sectionwhich is why it gives more negative reactivity for 149Sm.

4. Conclusion

The calculated Xenon worth of the selected core of MTR is nearlysame for all the 135Xe data in various libraries. The minor differ-ences arise due to differences in absorption cross-sections. There

Ratio

o

f A

bso

rp

tio

n cro

ss-sectio

n

Energy(eV)

10-2 10-1 100 101 102 103 104 105 106 107

0.6

0.8

1.0

1.2

1.4

1.6

1.8

2.0

2.2

10-2 10-1 100 101 102 103 104 105 106 107

0.6

0.8

1.0

1.2

1.4

1.6

1.8

2.0

2.2

b

I and (b) JENDL to the absorption cross-sections obtained using IAEA library.

Table 5Absorption cross-section for 149Sm.

Library One groupabsorptioncross-section (sa barns)

Ratio of absorptioncross-sectionwith JENDL3.2

JENDL 6.10� 104 1.000IAEA 6.25� 104 1.024ENDFB-VII 6.05� 104 1.024ENDFB-VI.8 6.25� 104 0.991JEF2.2 6.24� 104 1.022JEFF3.1 6.43� 104 1.054

A. Ahmad et al. / Progress in Nuclear Energy 51 (2009) 334–338338

are two sets of absorption cross-sections of 135Xe in six libraries.JENDL3.2, IAEA and ENDFB-VII have the same cross-sections whilethe remaining three libraries use the same data for 135Xe. In spite ofthe difference of cross-sections of 135Xe in various libraries, thenegative worth of 135Xe is nearly the same from all the libraries;this is due to the fact that one group absorption cross-sections for135Xe are approximately same for all libraries.

In case of 149Sm, the calculated Samarium worth for the selectedcore is again nearly close to each other for all Samarium data invarious libraries. The small differences are due to the difference ofabsorption cross-sections of 149Sm (Figs. 8 and 9) in all the cross-section data sets.

References

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Ahmad, S.I., Ahmad, N., 2006a. Burnup-dependent core neutronics analysis andcalculation of actinide and fission product inventories in discharged fuel ofa material test research reactor. Annals of Nuclear Energy 48, 599–616.

Ahmad, S.I., Ahmad, N., 2006b. Plutonium build-up credits for a material testresearch reactor and influence of cross section differences on actinide pro-duction. Annals of Nuclear Energy 236, 2537–2546.

Ahmad, S.I., Ahmad, N., Aslam, 2004. Effect of new cross section evaluations oncriticality and neutron energy spectrum of a typical material test reactor. Annalsof Nuclear Energy 31, 1867–1881.

IAEA, 1993. The evaluated nuclear data library of NEA Data Bank (JEF-2). IAEA-NDS-120, Rev.2.

Integrated Publishing, 2007. 135Xe Response to Reactor Shutdown. Integrated Pub-lishing, 9438US Hwy 19 N. # 311 Port Richey, FL 34668. http://www.tpub.com/content/doe/h1019v2/css/h1019v2_63.htm.

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McLane, V., Dunford, C.L., Rose, P.F., 1995. ENDF-102 Data Formats and Proceduresfor the Evaluated Nuclear Data File. ENDF-6, BNL-NCS-44945, Rev. 11/95.National Nuclear Data Center, Brookhaven National Laboratory.

Nakagawa, T., Shibata, K., Chiba, S., 1995. Japanese evaluated nuclear data libraryVersion 3 Rev.-2, JENDL-3.2. Journal of Nuclear Science and Technology 32, 1259.

Nuclear Energy Agency, 2007. The JEFF-3.1 Project. OECD Nuclear Energy Agency, LeSeine Saint-Germain, 12, boulevard des Iles, F-92130 Issy-les-Moulineaux,France. http://www.nea.fr/html/dbdata/JEFF/.