Neutronic and thermal hydraulic analysis for production of fission molybdenum-99 at Pakistan Research Reactor1

ARTICLE IN PRESS

www.elsevier.com/locate/anucene

Annals of Nuclear Energy xxx (2007) xxx–xxx

annals of

NUCLEAR ENERGY

Technical Note

Neutronic and thermal hydraulic analysis for productionof fission molybdenum-99 at Pakistan Research Reactor-1

A. Mushtaq a,*, Massod Iqbal b, Ishtiaq Hussain Bokhari b, Tariq Mahmood b,Tayyab Mahmood b, Zahoor Ahmad b, Qamar Zaman b

a Isotope Production Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad, Pakistanb Nuclear Engineering Division, Pakistan Institute of Nuclear Science and Technology, P.O. Nilore, Islamabad, Pakistan

Received 14 December 2006; received in revised form 8 June 2007; accepted 8 June 2007

Abstract

Neutronic and thermal hydraulic analysis for the fission molybdenum-99 production at PARR-1 has been performed. Low enricheduranium foil (<20% 235U) will be used as target material. Annular target designed by ANL (USA) will be irradiated in PARR-1 for theproduction of 100 Ci of molybdenum-99 at the end of irradiation, which will be sufficient to prepare required 99Mo/99mTc generators atPINSTECH and its supply in the country. Neutronic and thermal hydraulic analysis were performed using various codes. Data showsthat annular targets can be safely irradiated in PARR-1 for production of required amount of fission molybdenum-99.� 2007 Elsevier Ltd. All rights reserved.

1. Introduction

Techetium-99m (T1/2 = 6.02 h) is accounting for morethan 80% of all diagnostic nuclear medicine procedures inthe world. It is a unique isotope that can be incorporatedinto a number of radiopharmaceuticals to assist the diag-nosis of problems in different parts of the human bodyincluding, brain, heart, liver, thyroid gland, lungs, kidneysand bone. The popularity of 99mTc is also due to its lowcost, low radiation exposure to patients, high quality imag-ing and reliable availability in the form of 99Mo/99mTc gen-erators. Currently 99mTc is exclusively produced from thedecay of its 66 h half-life parent molybdenum-99 (IAEA-TECDOC-1065, 1999). Sterile PAKGEN� 99Mo/99mTcgenerators are being fabricated weekly using importedmolybdenum-99 from South Africa in a 99Mo loadingclean room facility created under an IAEA TechnicalCooperation Project at PINSTECH, Islamabad. The num-ber of hospitals practicing nuclear medicine in Pakistan is>25. Presently, the molybdenum-99 is shipped from South

0306-4549/$ - see front matter � 2007 Elsevier Ltd. All rights reserved.

doi:10.1016/j.anucene.2007.06.006

* Corresponding author. Tel.: +92 51 2208010; fax: +92 51 9290275.E-mail address: [email protected] (A. Mushtaq).

Please cite this article in press as: Mushtaq, A. et al., Neutronic and(2007), doi:10.1016/j.anucene.2007.06.006

Africa to Pakistan in a plastic bottle shielded by depleteduranium, in amount of 11 Ci/407 GBq (reference date).Various problems faced by PINSTECH with imports of fis-sion molybdenum include uncertainty of arrival of bulk99Mo, custom clearance at air port, activity variation, sizeof 99Mo plastic bottle, steel container of 99Mo plastic bottleand finally the mismatch of long holidays in Pakistan withthe supplier country.

To overcome the problems associated with the import oftechnetium-99m generators or fission 99Mo, such as avail-ability of hard currency, increase in the price of molybde-num-99, import policies, delay and changes in supplyschedule, etc. the indigenous production of molybdenum-99 in the country is justified. The country’s demand of fis-sion 99Mo for 99mTc generator is �12 Ci/444 GBq at refer-ence date. Under Co-ordinated Research Project,sponsored by International Atomic Energy Agency(IAEA), Vienna, instigations are underway to adopt thetechnology developed by ANL (USA) for small scale pro-duction of 99Mo in Chile, Libya, Pakistan and Romania.For preparation of safety documents, neutronic, thermaland hydraulic analysis for production of 100 Ci/3700 GBq 99Mo using LEU annular foil target and newly

thermal hydraulic analysis for production ..., Ann. Nucl. Energ.

2 A. Mushtaq et al. / Annals of Nuclear Energy xxx (2007) xxx–xxx

ARTICLE IN PRESS

designed target holder in PARR-1 was performed. Impor-tant data is presented in this article.

2. Pakistan Research Reactor-1 (PARR-1)

Pakistan Institute of Nuclear Science and Technology(PINSTECH), Islamabad is operating PARR-1 to provideservices to the users for the production of radioisotopesand for neutron irradiation. Since initial criticality,PARR-1 has rendered invaluable service in the trainingof manpower, production of radioisotopes and as a sourceof neutrons for basic and applied research. To reducenuclear-proliferation concerns it became essential that itscore be converted for operation with low enriched uranium(<20% 235U) fuel. The PARR-1 is a swimming pool typeresearch reactor originally designed for a thermal powerof 5 MW. Its core has been redesigned to operate withLEU fuel at a power level of 9 MW in 1992 and 10 MWin 1998. Technical data and fuel data of PARR-1 are pre-sented in Tables 1 and 2, respectively.

2.1. Reactor core assembly

The PARR-1 core consists of an assembly of standardand control fuel elements mounted on the grid plate. Thefuel elements can be assembled in different core configura-tions. The core is immersed in demineralized water whichacts as coolant, moderator and reflector. However, usingspecially designed reflector elements the light water canbe replaced on one or more sides with other reflectors suchas graphite, beryllium or heavy water.

2.2. Grid plate

The PARR-1 grid plate is made of 127 mm thick alumi-num. It has 54 holes in 9 · 6 patterns with a lattice spacing

Table 1Technical data of PARR-1

Reactor type: PoolThermal power, steady-state, kW: 10,000Maximum flux steady-state, thermal, n/cm2-s: 1.5 · 1014

Max flux steady-state, fast, n/cm2-s: 6.0 · 1013

Moderator and coolant: Light water (total flow rate = 950 m3/h andmaximum coolant temperature = 46 �C)

Reflector: Graphite, waterControl rod material: Ag, In, CdCriticality with LEU: October 1991Power increase: 10 MW in February 1998

Table 2Fuel data of PARR-1

Min critical mass, kg U-235: 4.42Normal core loading, kg U-235: 6.59Fuel material: U3Si2�AlEnrichment: 19.99%Origin of fissile material: USA, China

Please cite this article in press as: Mushtaq, A. et al., Neutronic and(2007), doi:10.1016/j.anucene.2007.06.006

of 81 · 77.11 mm. These holes accommodate the end fit-tings of the fuel elements. During operation the unusedholes are closed using plugs so that maximum coolantpasses through the fuel elements. In between the fuel ele-ment bearing holes there are 40 smaller holes (17.5 mmdiameter) for coolant flow which passes through the sideand the outer fuel plates. The PARR-1 core consists oflow enriched uranium (LEU) fuel having 19.99% 235U.The fuel material used is uranium silicide U3Si2�Al.

2.3. Core configuration No. 98

Core configuration # 98 is shown in Fig. 1. This coreconsists of 29 standard and five control fuel elements.There are two irradiation positions inside the core at theposition C7 and C4. The core is reflected by graphite ontwo sides and water on other sides.

3. Low enriched uranium (19.99% 235U) annular target

The goal in the LEU target design is to produce 100 Ciof molybdenum-99 at the end of irradiation in PARR-1 tomeet the current demands of 99Mo/99 mTc generators inPakistan. The LEU foil target selected has followingcharacteristics.

Uranium foil (19.99% 235U) of 125 lm thickness isenveloped in 15 lm thick Nickel foil and placed betweentwo aluminum tubes that are welded from both ends. Thegeometry of annular target is shown in Fig. 2, while dimen-sions are given below:

Outer aluminum tube external diameter = 30.00 mm.Outer aluminum tube internal diameter = 28.22 mm.Nickel foil = 15 lm.Uranium foil = 125 lm.Nickel foil = 15 lm.Inner aluminum tube external diameter = 27.99 mm.Inner aluminum tube internal diameter = 26.21 mm.

A B C D E F

Water Box

Graphite

9

8

7

6

5

4

3

2

1

Fission Chamber

Standard Fuel element

Control Fuel Element

Thermal Column

Key:s s s ss

s

s

s

s

s

s

s

s

s

s

s

s

s

s

s

s

s

s

s

s

s

s

ss s

Fig. 1. Core configuration No. 98 (PARR-1).

thermal hydraulic analysis for production ..., Ann. Nucl. Energ.

Al

Ni

U

Fig. 2. Schematic arrangement of aluminum tubes and foils in annulartarget.

Fig. 3. Annular target h

A. Mushtaq et al. / Annals of Nuclear Energy xxx (2007) xxx–xxx 3

ARTICLE IN PRESS

Please cite this article in press as: Mushtaq, A. et al., Neutronic and(2007), doi:10.1016/j.anucene.2007.06.006

Aluminum tube length = 162 mm.Uranium foil dimension = 76 ± 2 · 88 ± 2 mm = 16 guranium, 235U contents = 3.19 g.

4. Annular target holder

The annular target holder is made of reactor grade alu-minum metal. It has two tubes one outer and the otherinner which is larger. The extended part of inner tubeguides the insertion of annular target. The upper and lowerplates have same dimension. Holes and groves have beenprovided in upper plate for fixing the stainless steel flexiblewire. For steady flow of coolant (water) fins has been fixedon inner tube to separate the annular target from innertube. Details and dimensions of annular target holder aredepicted in Figs. 3–5.

older (general view).

thermal hydraulic analysis for production ..., Ann. Nucl. Energ.

4 A. Mushtaq et al. / Annals of Nuclear Energy xxx (2007) xxx–xxx

ARTICLE IN PRESS

5. Neutronic analysis

Neutronic analysis of the target holder for 99Mo produc-tion was performed. Computer codes WIMSD/4 alongwith BORGES and CITATION were used to performthe calculations. In the current study target holder wasplaced at the central water box facility and was irradiatedat five different axial positions. To generate the microscopiccross-sections for different regions of the core, the MTR-PC26 package was used. This package is a collection ofstandard computer codes and libraries for conducting reac-tor core static, depletion, transient; thermal hydraulic and

Fig. 4. Annular target holder, f

Please cite this article in press as: Mushtaq, A. et al., Neutronic and(2007), doi:10.1016/j.anucene.2007.06.006

shielding studies for MTR type reactors using personalcomputers. In current research, only neutronic option forcross-section generation has been used from this package.For this purpose, computer codes WIMSD/4 along withBORGES were used from MTR-PC26 package (Deenet al., 1995; Halsall, 1980; Rubio, 1993). Employing thesecodes 10 group microscopic cross-sections and numberdensity calculations were performed. These calculationswere performed for fuel, structure, control absorber, con-trol follower, end plugs, water reflector, and graphitereflector regions. Also cross-sections were generated forannular target holder. Representative unit cell models for

ront, top and bottom view.

thermal hydraulic analysis for production ..., Ann. Nucl. Energ.

Fig. 5. Annular target holder showing location of annular target.

1.4195cm(Ni Clad)

1.3105 cm (Water)

1.3995 cm (Al)

1.404 cm (Air) 1.40555cm

(Ni Clad)1.418 cm (fuel)

1.424 cm(Air)

1.5 cm (Al)

4.459 cm (Water)

Fig. 6. Unit cell model employed in WIMSD/4 for annular target holder.

0.17609cm

0.17609cm

Fuel Clad Moderator Extra 0.1685cm Region Region Region Region

(Al) (H 2O) (H2O+Al)

0.0635cm

0.0255cm

Fig. 7. Unit cell model employed in WIMSD/4 for fuel meat of PARR-1.

A. Mushtaq et al. / Annals of Nuclear Energy xxx (2007) xxx–xxx 5

ARTICLE IN PRESS

fuel meat of PARR-1 and for annular target holder areshown in Figs. 6 and 7. The incremental radii for eachregion are shown with their material specified in parenthe-sis. Burn up option of WIMSD/4 was used for fuel regionto generate microscopic cross-sections for fission products.Computer code BORGES is used to read the output ofWIMS/D4, as per instructions of the user, and then writesit in a form that is readily usable in the multi-dimensional,diffusion theory code CITATION. Hence, these cross-sec-tions and number densities at the output of BORGES wereemployed in CITATION. The equilibrium core No. 98(Fig. 1) was modeled employing 3D option of CITATION(Fowler et al., 1971). The target holder contains the annu-lar target with 16 g of total uranium whereas contents of235U are 3.19 g. Analysis of the target holder was

Please cite this article in press as: Mushtaq, A. et al., Neutronic and(2007), doi:10.1016/j.anucene.2007.06.006

performed by irradiating it in the central water box at C-7 location in core configuration # 98 at the beginning ofequilibrium cycle as shown in Fig. 1. Effect of irradiationwas studied by placing target holder at five different axialpositions. Excess reactivity in the core for these cases wascalculated and is shown in Fig. 8. The target reactivitiesare summarized in Table 3. It is evident from figure thatmaximum excess reactivity of 120 pcm is present in the corewhen target holder is placed in the fourth plane from thetop of the core. This is due to the neutron flux variationinside the core. It has been observed that location of max-imum power density in the core remains unchanged by

thermal hydraulic analysis for production ..., Ann. Nucl. Energ.

Rea

ctiv

ity

(pcm

)

0

30

60

90

120

150

0 20 40 60 80Distance in cm of Target Holder from top of the core

Fig. 8. Excess reactivity in the core after placing LEU annular target.

44

45

6 A. Mushtaq et al. / Annals of Nuclear Energy xxx (2007) xxx–xxx

ARTICLE IN PRESS

placing target holder in any of five axial planes. However,value of maximum power density varies slightly by movingtarget holder in different axial planes. The variation inpeaking factor calculated is from 2.928 to 2.961. This max-imum peaking value is less than the safe value calculatedfor the PARR-1 core configuration (FSAR of PARR-1,1992). Based on these calculations, the target holder canbe irradiated in the central flux trap of the core. Duringreactor operation the target holder should not be movedbecause it can add reactivity in the system of the order of120 pcm. No significant effect would be on peaking factor.

6. Thermal hydraulic analysis

Thermal hydraulic analysis has been carried out for theproposed design (shown in Figs. 3–5) of target holder forfission molybdenum-99 production at PARR-1. Parameterof interest is the maximum temperature gained when thereactor is in operation at full power level (10 MW).

It is intended to mount the target holder inside the waterbox in the core. Provision would be made that it could beinserted at any desired vertical position (with middle posi-tion for maximum neutron targeting). Calculations wereperformed to assess its impact on the overall core coolantflow rate as well as to estimate its plate surface temperaturerise.

Hydraulic calculations were performed to estimate thecoolant flow rate through the target. Computer code DPwas employed for this purpose. In the code, pressure drops,velocity distribution and flow rates through different chan-nels of the core, effective and bypass flow are calculated.

Table 3Reactivity of fission molybdenum-99 target at C-7 in PARR-1

Annular target holder

Plane U-235(g)

Power(kW)

Net reactivity(pcm)

Irradiation Time(h)

1 3.19 5.42 7 37.772 3.19 10.05 34 20.363 3.19 14.37 73 14.254 3.19 17.41 120 11.765 3.19 14.62 81 14.00

Please cite this article in press as: Mushtaq, A. et al., Neutronic and(2007), doi:10.1016/j.anucene.2007.06.006

The code computes velocities through an iterative proce-dure after converging to the same pressure drop acrossthe core for each channel.

The hole size at water box exit was varied to check itsimpact on coolant velocity distribution in the core. Varioussizes were considered in calculations. In order to minimizeits effect on standard fuel channel velocity, hole size waskept the same as that of regular water box (i.e. 1 cm diam-eter). Calculation of pressure drops along the holder inwater box revealed that hole of such size would allow cool-ant flow rate that would be less than 0.5% of total effectiveflow rate through the core (Bokhari et al., 1999). Watervelocity has been calculated along the active channel ofholder to be 1.63 m/s.

Neutronic calculations predicted various power levels atdifferent vertical positions shown in Table 3. For the anal-ysis, maximum power level (17.58 kW) has been consid-ered. Axial distribution of power has been assumed tohave cosine shape. To account for the uncertainties, anengineering hot channel factor (1.584) was incorporatedusing the conservative multiplicative method (Bokhariet al., 2002). This factor is the product of three compo-nents: (i) a factor 1.2 for the coolant temperature rise dueto manufacturing tolerances in the coolant channel spac-ing, (ii) a factor of 1.2 for the film temperature rise dueto uncertainties in the heat transfer coefficient and inhomo-geneities in U-235 distribution, etc. and (iii) a factor of 1.1for uncertainties in the calculated power distribution.

Target holder was modeled for one-dimensional hydro-dynamics and one-dimensional heat transfer. Local cladsurface temperature was computed with the help of heattransfer coefficient (calculated by using Dittus–Boelter cor-relation) and the local coolant temperature variation as peraxial power distribution.

As the holder has good flow area (704 mm2) along activeplate with appreciable coolant velocity, calculated maxi-mum temperature rise is nominal (Tmax,surface = 44.1 �C)with Tinlet as 38 �C. Heat transfer coefficient along the platewas calculated to be 6910 W/m2 K. Temperature distribu-tion in axial direction is shown in Fig. 9.

0 20 40 60 80 100 12038

39

40

41

42

43

Tem

pera

ture

(C)

Distance from top (mm)

Fig. 9. Temperature distribution along the plate surface.

thermal hydraulic analysis for production ..., Ann. Nucl. Energ.

Table 4High enriched uranium targets used for 99Mo production

Target 235Uenrichment (%)

Dissolution Producer

UO2, (on inside of stainless steel cylinder) (i) 93 (i)HNO3 + H2SO4

(i) (Cintichem)/Indonesian National Atomic Energy Agency(BATAN), Indonesia

(ii) 93 (ii)HNO3 + H2SO4

(ii) Sandia National Laboratories (SNL), USA

Extruded Al-clad U/Al, alloy pins 93 HNO3 AECL/Nordian CanadaAluminum clad U/Al/Al dispersion fuel plates (i) 89–93 (i)

NaOH + NaNO3

(i) Institut National des Radioelements IRE, Belgium

(ii) 46 (ii)NaOH + NaNO3

(ii) Atomic Energy Corporation of South Africa Limited AEC,South Africa

(iii) 90–93 (iii) NaOH (iii) Mallinckrodt, Netherlands

UO2 pressed with MgO in the form of hollowcylinder (SS clad)

90 HNO3 IPPE, Russian Federation

A. Mushtaq et al. / Annals of Nuclear Energy xxx (2007) xxx–xxx 7

ARTICLE IN PRESS

At PARR-1, safety analysis performed earlier (Bokhariet al., 2002) considered various probable reactivity-inducedaccidents such as, maximum start-up accident, drop of afuel element on the core, flooding of beam-tube, removalof in-pile experiment, movement of core towards thermalcolumn. Amongst these accidents (with overpower trip set-ting at 115% of steady-state power level), a maximum ofpower level (@40 MW) could be attained in case of accidentdue to drop of fuel element on the core. Calculations showthat during such worst possible reactivity-induced accidentat PARR-1 (resulting in heat flux increase by four times),plate temperature would only increase to a maximum of62.5 �C, which is far less than the saturation temperature(@113 �C) at the core pressure level.

7. Discussion

Although the potential use of accelerators for the pro-duction of 99Mo is underway, but the present sourceresearch reactors will continue their role in foreseeablefuture. The present world demand for high specific activity99Mo are met by utilizing fissioning of 235U, while somecountries also produce low specific activity via 98Mo (n,c)99Mo in a nuclear reactor. At present almost entire demandfor 99Mo is supplied by a few commercial producers byusing thermal neutron induced fission of highly enricheduranium-235 (IAEA-TECDOC-1051, 1998). In October1992 the USA Congress passed an amendment to AtomicEnergy Act of 1954. This amendment prohibits the exportof HEU for use as a fuel or target in a research or test reac-tor unless several conditions are met: (i) no alternativeLEU fuel or target can be used, (ii) the US is actively devel-oping an LEU fuel or target for the reactor and (iii) theproposed recipient of HEU provides assurance that, when-ever possible, an LEU fuel or target will be used in thatreactor. To reduce nuclear-proliferation concerns, the USReduced Enrichment for Research and Test Reactors(RERTR) Program is working to reduce the use of highenriched uranium (HEU) by substituting low enriched

Please cite this article in press as: Mushtaq, A. et al., Neutronic and(2007), doi:10.1016/j.anucene.2007.06.006

uranium (LEU) fuel and targets. Low enriched uraniumcontains <20% 235U. Currently, most of the world’s supplyof 99Mo is produced generally by fissioning of 93% 235U.Some details about targets for the production of 99Moare given in Table 4 (IAEA-TECDOC-1065, 1999).

In the end of year 2005 International Atomic EnergyAgency (IAEA), Vienna with financial support of USDepartment of Energy initiated a co-ordinated researchproject (CRP), in which Pakistan is participating as a con-tract holder. The CRP aims to assist recipients to research,test, and evaluate the LEU modified Cintichem processwith LEU foil targets, and neutron activation of naturalmolybdenum oxide targets and utilization of gel genera-tors. so-called ‘‘Freeware Technology’’ for LEU fissionand gel moly production of Mo-99.

To yield equivalent amounts of 99Mo, the LEU targetsmust contain P5 times as much uranium as the HEU tar-gets they replace. The target developed by ANL (USA)consisted of a uranium metal foil wrapped in Ni foil andfinally sandwiched between concentric cylinders of alumi-num (Conner et al., 1999). After irradiation the uraniumfoil is removed from the target body for processing. Thisremoves the aluminum which then does not need to be dis-solved, thus reducing the volume of dissolving mixture/liquid waste from any processing technique. In order toprevent spot welding of the uranium foil to the aluminumcylinders by fission fragments the target foil is wrappedin Ni foil for acid dissolution. Since weekly demand of fis-sion 99Mo for preparation of Technetium generator is �12Ci, the required amount of 99Mo at the end of irradiationin PAAR-1 was estimated in a following way.

Time elapsed during:

(i) Cooling of target and chemical separation of99Mo = 2 days.

(ii) Preparation of 99mTc generators = 2 days.(iii) Transportation of 99mTc generators = 1 day.(iv) Arrival at hospital (early) = 2 days.(v) Total time elapsed after irradiation = 7 days.

thermal hydraulic analysis for production ..., Ann. Nucl. Energ.

8 A. Mushtaq et al. / Annals of Nuclear Energy xxx (2007) xxx–xxx

ARTICLE IN PRESS

Decay factor of Mo-99 for seven days= e�(0.693·7·24)/66 = 0.171.

Considering chemical processing yield �73%.100 Ci · 0.171 · 0.73 = 12.48 Ci (desired activity of

Mo-99 at reference date).

8. Conclusion

Before irradiating annular LEU foil target, neutronicand thermal hydraulic calculations were performed forthe production of 100 Ci {3700 GBq) 99Mo in PARR-1and it is concluded that the proposed annular target andits holder designs could be safely adopted for molybde-num-99 production project at PARR-1, without compro-mising reactor safety.

Acknowledgements

This work was sponsored by International Atomic En-ergy Agency (IAEA), Vienna under research contract No.13362/RB. We are also thankful to Ira. Goldman (IAEAProject Officer), Dr. M. Jehangir (Director General PINS-TECH) and Showkat Pervez (Head Nuclear EngineeringDivision, PINSTECH), for their constant support andstimulating discussions.

Please cite this article in press as: Mushtaq, A. et al., Neutronic and(2007), doi:10.1016/j.anucene.2007.06.006

References

Bokhari, I.H., Israr, M., Pervez, S., 1999. Thermal Hydraulic andSafety Analyses for Pakistan Research Reactor-1, In: Proceedings of22nd International Meeting on Reduced Enrichment for Researchand Test Reactors (RERTR), 3–8 October, 1999, Budapest,Hungary.

Bokhari, I.H., Israr, M., Pervez, S., 2002. Analysis of reactivity inducedaccidents at Pakistan Research Reactor-1. Annals of Nuclear Energy29, 2225–2234.

Conner, C., Lewandowski, E.F., Snelgrove, J.L., Liberatore, M.W.,Walker, D.E., Wiencek, T.C., McGann, D.J., Hofman, G.L., Vande-grift, G.F., 1999. Development of annular targets for 99Mo produc-tion. In: International Meeting on reduced Enrichment for Researchand Test Reactors, October 3–8, 1999, Budapest, Hungary.

Deen, J.R., Woodruff, W.L., Costescu, C.I., 1995. WIMS/D4 UserManual REV.0, ANL/RERTR/TM-23.

Fowler, T.B., Vondy, D.R., Cunningham, G.W., 1971. Nuclear ReactorCore Analyses Code CITATION, ORNL – TM – 2496 Rev. 2.

Halsall, M.J., 1980. A Summary of WIMSD4 Input Options, AEEW-M1327.

IAEA-TECDOC-1065, 1999. Production technologies for molybdenum-99 and technetium-99m, International Atomic Energy AgencyVienna.

IAEA-TECDOC-1051, 1998. Management of radioactive waste from99Mo production. International Atomic Energy Agency Vienna.

Pakistan Research Reactor-1, September 1992, Final Safety AnalysisReport (FSAR) for Conversion to LEU Fuel and Power Upgradation.Revision 1.

Rubio, R.O., 1993. INVAP SE ‘‘BORGRS v3.0’’.

thermal hydraulic analysis for production ..., Ann. Nucl. Energ.