Annals of Nuclear Energy 76 (2015) 530–539

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Annals of Nuclear Energy

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Use of the ETA-1 reactor for the validation of the multi-groupAPOLLO2–MORET 5 code and the Monte Carlo continuous energyMORET 5 code

http://dx.doi.org/10.1016/j.anucene.2014.10.0330306-4549/� 2014 Elsevier Ltd. All rights reserved.

⇑ Corresponding author.E-mail address: nicolas.leclaire@irsn.fr (N. Leclaire).

N. Leclaire a,⇑, B. Cochet a, F.X. Le Dauphin a, W. Haeck a, O. Jacquet b

a Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-EXP, SNC, LNC, Fontenay-Aux-Roses 92262, Franceb Millennium Company, 16, avenue du Québec, Silic 628, 91945 Villebon sur Yvette Cedex, France

a r t i c l e i n f o a b s t r a c t

Article history:Received 10 June 2014Accepted 27 October 2014

Keywords:ValidationETA reactorThO2

Heavy waterProbability tablesThermal-scattering

The present paper aims at providing experimental validation for the use of the MORET 5 code foradvanced concepts of reactor involving thorium and heavy water. It therefore constitutes an opportunityto test and improve the thermal-scattering data of heavy water and also to test the recent implementa-tion of probability tables in the MORET 5 code.

� 2014 Elsevier Ltd. All rights reserved.

1. Introduction

The Epithermal Test Assembly (ETA) experiments (NEA/NSC/DOC(95)03, 2011) were performed in support of the Light WaterBreeder Reactor (LWBR) Program from July 1969 to March 1971at Bettis Atomic Power Laboratory (USA). The experiments weredesigned to provide a clean test of the adequacy of the epithermalcross section data for 235U, 233U and 232Th. The ETA experimentsconsisted of a heavy water moderated central test region latticeof test fuel rods surrounded by a light water moderated annular dri-ver region lattice of aluminum alloy clad UO2 rods. The ETA-1 crit-ical experiment consisted of a test region of 2304 UO2–ThO2 fuelrods (6.7 wt.% UO2 in UO2–ThO2, 93 wt.% 235U in U) moderated byheavy water surrounded by a driver region of 2580 TRX high den-sity UO2 fuel rods (1.3 wt.% 235U in U) moderated by light water.

Until now, no documented benchmark experiment with tho-rium in the fuel region and heavy water as moderator or coolantwas available in the literature such as the ICSBEP Handbook(NEA/NSC/DOC(95)03, 2011). This set of experiment, referred toas HEU-COMP-THERM-018 in the ICSBEP Handbook (NEA/NSC/DOC(95)03, 2011), is therefore valuable for validating the calcula-tion of this type of configurations.

The aim of the paper is to test and validate the geometricalmodels and the treatment of nuclear data performed by the MORET5 (Cochet et al., 2013) code used either in the APOLLO2–MORET 5multi-group route or in the continuous energy route. Indeed, theseexperiments are an opportunity to test the physical models (self-shielding of cross sections, 281-group energy mesh) used in theAPOLLO2 cell code of the APOLLO2–MORET 5 route but also to testthe implementation of probability tables in the MORET 5 code andthe treatment of thermal-scattering S (a,b) data from the JEFF-3.1and ENDF/B-VII.0 libraries done by the continuous energy codes.

2. Description of the experiments

2.1. Geometry

As shown in Fig. 1, the reactor was composed of two zones: acentral test zone with a lattice of 256 bundles containing nineUO2–ThO2 fuel rods per bundle and a driver zone composed ofUO2 rods. The central zone was set in a double-walled tank; theouter diameter of the outer wall was 71.12 cm while the outerdiameter of the inner wall was 67.31 cm. The thickness of the wallswas 1.27 cm.

This driver lattice zone was located in the reactor tank, whoseouter diameter was 168.91 cm.

A cross section view from above of the two lattices is given inFig. 2. A lateral view (see Fig. 3) shows the base plates on which

N. Leclaire et al. / Annals of Nuclear Energy 76 (2015) 530–539 531

the reactor stands and the dimensions of the two tanks. The heightof the reflector tank was 182.88 cm, whereas the internal height ofthe central tank was 194.0052 cm (Fig. 4).

Curved plates designed as ‘‘control rods’’ were placed betweenthe inner and outer tank walls. Four winged-tee shaped 70Ag–30Cd safety rods were located in the test tank to ensure the assem-bly would remain sub-critical if the test region were filled withlight water instead of heavy water.

The water temperature was maintained constant in the reactortank through a steam heat exchanger. Similarly, the dump tanks oflight and heavy water included thermostatically controlled electri-cal heaters to maintain the temperature of water at about 25 �C.

The 256 fuel bundles were made of aluminum canisters. Theirexternal length was 2.5933 cm and the thickness of their wallswas 0.08128 cm. The pitch of the lattice of bundles was equal to3.3782 cm.

Each bundle contained nine ETA-1 UO2–ThO2 rods clad withaluminum. The outer radius of the clad was 0.39243 cm, its thick-ness was 0.03556 cm. The diameter of the fuel pellets was0.3302 cm. The fuel rods are presented in Fig. 5. The fuel bundleis described in Fig. 6.

Fig. 1. Schematic view of t

The driver lattice comprised 2580 TRX UO2 rods. The UO2 rods(Fig. 8) were clad with aluminum. The outer radius of the clad was0.57531 cm; its thickness 0.07112 cm. The radius of the fuel pelletwas 0.46841 cm and the pitch of the lattice was 1.5367 cm. The rodswere maintained by four 0.635-cm thick grids spaced vertically by10.16, 23.495, 59.055 and 39.37 cm. A picture of these grids isreported in Fig. 7. A picture of the TRX UO2 rods is reported in Fig. 8.

2.2. Chemical media

2.2.1. UO2 TRX rodsThe UO2 TRX fuel rods of the peripheral zone (driver lattice) had

a 235U enrichment of 1.31 wt.%. They were immersed in light water,whose density was 0.99855 g/cm3.

The density of the fuel pellets was 10.53 g/cm3.

2.2.2. UO2–ThO2. ETA-1 rodsThe UO2–ThO2 ETA-1 fuel rods of the central zone had a 235U

enrichment of 93.21 wt.%. They were immersed in heavy water,whose density was 1.10385 g/cm3. The weight proportion of ThO2

in the fuel is 93.189%. The density of the fuel pellets was 8.35 g/cm3.

he assembled reactor.

TRX rods

UO2-ThO2 rods

Fig. 2. Cross section view from above of the ETA-1 reactor.

532 N. Leclaire et al. / Annals of Nuclear Energy 76 (2015) 530–539

2.3. Critical approach

The initial approach to criticality was made step by step. Theinitial fuel loading in the test region consisted of a four-by-eigh-teen rectangular lattice of ETA-1 fuel clusters in the center of thetest region. The lattice length corresponded to the full diameterof the test region.

The initial loading of the driver lattice region resulted in a rect-angular core that was approximately 12.7 cm wide and 111.76 cmlong. Both the test and driver regions were dry (un-moderated)during fuel loading. Then an approach to criticality was performed.

The approach to critical procedure was initiated by deploying a30-Ci 238PuBe neutron source to a position just below the testregion lattice base plate.

The test region tank was filled with heavy water until the testregion water level indicated ‘‘full up’’; then the driver region wasfilled with light water until the driver region water level indicated‘‘full up’’. The safety rods were then withdrawn followed by thecontrol rods being withdrawn using a pull and wait procedure.

The neutron count rate was measured at approximately every12.7 cm surface level increase during the heavy water and lightwater filling procedures and as a function of safety and controlrod position in order to predict the critical control rod positionand to ensure the core had the required shutdown margin.

For the ETA-1 experiment, the final critical fuel loading con-sisted of 256 clusters of fuel rods (total of 2304 rods) in the heavywater moderated test region and 2580 TRX high-density UO2 fuelrods in the light water moderated driver region.

The keff of the experiment was finally 0.9995 ± 0.00200.

2.4. Experimental uncertainties

The uncertainty calculations were performed using the MCNP5(version 1.5.1) Monte Carlo code with the ENDF/B-VII.0 cross

section library (see ICSBEP Handbook NEA/NSC/DOC(95)03,2011). For each calculation, a standard deviation in keff of approx-imately 0.00011 was targeted.

The uncertainties were calculated separately for each indepen-dent parameter. The global uncertainty was then calculated asbeing the square root of the quadratic sum of each component.

The detail of the main uncertainties is given in Table 1.

3. Codes and libraries

The MORET 5 Monte Carlo code is used in two main calculationroutes:

� A multi-group route, called APOLLO2–MORET 5 associated withthe CRISTAL V2.0 b package, involving the deterministicAPOLLO2 cell code and the MORET transport code. TheAPOLLO2–MORET 5 route is intended to be the ‘‘industrial’’route used by criticality practitioners inside the CRISTAL V2.0package, which will be delivered no later than end of 2014. Thisroute is an update of the APOLLO2–MORET 4 route associatedwith the CRISTAL V1.2 package.� A continuous energy route using the 3D Monte Carlo MORET 5

code and various nuclear data libraries.

The MCNPX 2.6 code is used as a ‘‘reference route’’ for a purposeof comparison with the MORET 5 continuous energy route, whoseversion 5.B.1 will be delivered very soon and whose version 5.C.1development is still in progress.

These codes and libraries are gathered in Table 2.

3.1. Multi-group route

The multi-group route, APOLLO2–MORET 5, proceeds in twosteps. First, it produces through the APOLLO2.8.3 deterministic cell

Fig. 3. Lateral view of the ETA-1 reactor.

N. Leclaire et al. / Annals of Nuclear Energy 76 (2015) 530–539 533

code, homogenized and self-shielded 281-energy-group cross sec-tions using the collision probability method to solve the Boltzmannequation. These macroscopic cross sections are then used by theMonte Carlo MORET 5.B.1 (or 5.C.1) code to calculate the fluxesand keff in a 3D configuration. The nuclear data used in theAPOLLO2 calculations are coming from the JEFF-3.1 library.

It should be noted that, as the CRISTAL V2.0 package is stillunder development at the time of this study, in the calculationsperformed with the APOLLO2.8.3 code, the final calculation optionsof the CRISTAL (Gomit et al., 2003, 2011) V2.0 package were notconsidered (especially the self-shielding of structural materials).

The multi-group route, APOLLO2–MORET 4, was the ‘‘indus-trial’’ route used in the CRISTAL V1.2 package. The APOLLO2.5.5code produces 172-energy-group, macroscopic, homogenized andself-shielded cross sections coming from the JEF2.2 library. Thesecross sections are then used by the Monte Carlo MORET 4.B.4 codeto calculate the fluxes and keff in a 3D configuration.

3.2. Continuous energy routes

The continuous energy route involves the MORET 5 code. Itsolves the transport equation to calculate fluxes and keff of the

3D configuration. It uses nuclear data processed in the ACE formatcoming either from the JEF2.2, the JEFF-3.1 or the ENDF/B-VII.0libraries.

Two versions are considered: 5.B.1 and 5.C.1. The MORET 5.B.1version, which will be released very soon within the CRISTAL pack-age; the thermal scattering data are implemented but the probabil-ity tables are not available. However, they are already available inthe MORET 5.C.1 version, which is still under development.

The MCNPX 2.6 code is also used to make comparisons with theMORET 5 continuous energy code. It is a 3D Monte Carlo code thatsolves the transport equation. The MORET 5 code uses indeed thesame nuclear data, processed in the same ACE format as theMCNPX 2.6 code.

4. Methodology for experimental validation

The validation of the aforementioned codes is done through thecomparison of the calculated keff with the benchmark keff. If thediscrepancy between these two keff values is lower than the com-bined standard deviation of the benchmark uncertainty and theMonte Carlo standard deviation (see Eq. (1)), then no bias can behighlighted.

Fig. 4. Cross section view from above of the internal part of the ETA-1 reactor.

Fig. 5. ETA-1 UO2–ThO2 fuel rod.

Fig. 6. Fuel bundle cross section view from above.

534 N. Leclaire et al. / Annals of Nuclear Energy 76 (2015) 530–539

Fig. 7. Photo of the driver lattice grids.

Fig. 8. TRX UO2 fuel rod.

Table 1Main uncertainties in keff corresponding to 1 parameter uncertainties.

Parameter Dkeff (1)

UO2–ThO2 fuelThO2 mass 0.00051Fuel Density 0.00050Boron equivalent impurities 0.00042Rod Pitch 0.00055

TRX fuelEnrichment (235U) 0.00081Outer diameter/Clad thickness 0.00094Rod Pitch 0.00075

N. Leclaire et al. / Annals of Nuclear Energy 76 (2015) 530–539 535

rcombined ¼ffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffir2

MC þ Dk2eff benchmark

q� �ð1Þ

Various libraries are tested for continuous energy codes toassess the impact of nuclear data.

The discrepancies are judged significant if it exceeds the qua-dratic sum of Monte Carlo standard deviations for the two codesffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffiffi

r2MC code1 þ r2

MC code2

q

. When it is not stated, the Monte Carlo standard deviation is equalto 30 pcm, which leads to a combined standard deviation at 3r ofabout 130 pcm.

Moreover, whenever the reactivity worth of an elementreplaced with another is tested, the calculated keff corresponds tothe keff with the replaced element minus the keff of the referencecase.

In a second step, the keff results of multi-group codes(APOLLO2–MORET 4 and/or APOLLO2–MORET 5) is compared tothe keff of continuous energy codes (MORET 5 and MCNP 5) usingthe same libraries. It allows enhancing biases due to the multi-group treatment and physical models implemented in multi-groupcodes.

Finally, the validation of D2O is done comparing to other bench-marks with D2O and without thorium; then the same principle isapplied to the thorium, for which other experiments with thoriumand without D2O are searched.

5. Validation results

When it is not specified, the MORET 5 calculations are run bydefault using the 5.C.1 release of the MORET code, with the thermal

Table 2Codes and libraries used for the validation process.

Code Energy groupstructure

Library

APOLLO2–MORET 5 (b version ofCRISTAL V2.0 package)

281 energygroups

JEFF-3.1

MORET 5.B.1 and 5.C.1 (Cochet et al.,2013)

Continuousenergy

JEF2.2, JEFF-3.1 andENDF/B-VII.0

MCNPX 2.6 (Booth et al., 2003)

Table 4Reactivity worth of the central Zone.

State keff (MORET 5 JEFF-3.1 library)(rMC in pcm)

Central zone filled with UO2–ThO2 rods 0.99141 (30)Central zone without UO2–ThO2 rods

and bundles0.89396 (30)

Reactivity worth (pcm) +9745

Table 5Reactivity worth of heavy water.

D2Oreplaced by

Dkeff (MCNPX 2.6 – JEFF-3.1) (pcm)

Dkeff (MORET 5 – JEFF-3.1) (pcm)

H2O 23,712 23,496

Table 6Reactivity worth of thorium in UO2–ThO2 rods.

Thoriumreplaced by

Dkeff (MCNPX 2.6-JEFF –3.1) (pcm)

Dkeff (MORET 5 – JEFF-3.1) (pcm)

235U +50,181 +49,882238U +1968 +1906Void +20,590 +20,336

Table 7Discrepancy (pcm) associated to the nuclear data of thorium and uranium –MCNPX 2.6 code.

All elements are JEFF-3.1 except elementbelow

DiscrepancyJEF2.2 – JEFF-3.1(pcm)

Discrepancy ENDF/B-VII.0 – JEFF-3.1(pcm)

Thorium 454 502Uranium 158 �19

Table 8Discrepancy (pcm) associated to the nuclear data of thorium, uranium,aluminum and deuterium – MORET 5 code.

All elements are JEFF-3.1 except elementbelow

DiscrepancyJEF2.2 – JEFF-3.1(pcm)

Discrepancy ENDF/B-VII.0 – JEFF-3.1(pcm)

Thorium 421 425Uranium 73 �36Aluminum 140 �312H in D2O �46 �96

536 N. Leclaire et al. / Annals of Nuclear Energy 76 (2015) 530–539

scattering data of heavy water and the probability tables activatedin the unresolved energy range of cross sections.

5.1. keff results with continuous energy codes and effects of libraries

The keff results are reported in Table 3. The reactivity worth ofthe different materials encountered in the two fissile zones isreported in Table 4, through Table 6. The influence of nuclear datalibraries is studied in Table 7 through Table 8.

5.1.1. keff resultsThe keff results calculated with the continuous energy codes

mentioned in Section 3 and for various libraries are reported inTable 3.

It appears that there is an effect of cross sections libraries, theJEF2.2 library leading to keff results closer to the benchmark keff

(0.9995 ± 0.00200) than the ENDF/B-VII.0 and more especially thanthe JEFF-3.1 library. The comparison with the benchmark keff sug-gests that, given the amount of the experimental uncertainties,there is a significant negative bias associated with the calculationof these experiments either with the ENDF/B-VII.0 or the JEFF-3.1evaluation of cross sections.

Moreover, a systematic discrepancy of the order of 200 pcmbetween the MORET 5 code results and the MCNPX 2.6 resultscan be observed. The origin of this small bias could be attributedto the discrepancy in the models used to implement the probabilitytables in the code or even in the description of the geometry in theinput decks. The aim of the following sections is to investigate thisproblem.

5.1.2. Reactivity worth of thoriumIn order to assess the influence of the central part of the reactor

containing the UO2–ThO2 fuel bundles, MORET 5 calculations wererun removing the UO2–ThO2 fuel bundles of the central zone. Theresults are gathered in Table 4.

It appears that the UO2–ThO2 rods contained in the fuel bundlesof the central zone have a significant impact on keff of about10,000 pcm, allowing the validation of the UO2–ThO2 rods.

In addition, the thorium is first replaced by 235U, then by 238Uand then finally by void in the composition of the UO2–ThO2 rods.The results with the MCNPX 2.6 and MORET 5 Monte Carlo codesare provided in Table 6. It emphasizes the non negligible impactof thorium in UO2–ThO2 rods, showing that these experimentscan allow contributing to the validation of thorium.

These keff changes can be understood when having a look at thethorium and uranium cross sections (see Fig. 9). For example, thethorium capture cross section is quite comparable to the 238U

Table 3keff results with MORET 5 and MCNPX 2.6 codes and various libraries (rMC in pcm).

Code Library (rMC in pcm)

JEF2.2 JEFF-3.1 ENDF/B-VII.0

MORET 5 0.99935 (30) 0.99141 (30) 0.99659 (30)MCNPX 2.6 0.98892 (10) 0.99480 (11)

capture cross section, which justifies the small discrepancy(1900 pcm) observed. On the contrary to 238U and 232Th thatmostly capture neutrons, 235U is fissile in the thermal energy range.This is the reason for the large keff increase when 232Th is replacedby 235U.

5.1.3. Reactivity worth of heavy water in the central lattice zone (fuelbundles)

The heavy water in the fuel bundle zone has been replaced bylight water. It leads to a keff increase by around 23,000 pcm (SeeTable 5). This increase of keff finds its origin in the higher ratio offission, when D2O is replaced by H2O, in the central zone wherea better slowing down of neutrons is obtained. In fact, even ifhydrogen is more absorbent, the contribution of its cross sectionsto elastic scattering is more important than it is for deuterium.

Th232 capture

U235 capture U238 scattering U238 capture

U235 fission

Fig. 9. Cross sections of thorium and uranium (JEFF-3.1 evaluation).

Fig. 10. Capture cross sections of thorium (JEFF-3.1 and ENDF/B-VII.0 evaluations).

Table 9keff results with various libraries – APOLLO2–MORET 5 multi-group code (rMC inpcm).

Code Library (rMC in pcm)

172-group JEF2.2 281-group JEFF-3.1

APOLLO2–MORET 5 1.00694 (30) 0.99488 (30)

N. Leclaire et al. / Annals of Nuclear Energy 76 (2015) 530–539 537

5.1.4. Effect of nuclear dataThe evaluation of thorium, uranium, aluminum and deuterium

is tested with MCNPX 2.6 (see Table 7) and MORET 5 (see Table 8)

calculations. As a result, all elements are taken in the JEFF-3.1library except those named previously, for which other librariesare tested.

The impact of the evaluation is quite negligible for uranium,aluminum and deuterium compared with the Monte Carlo stan-dard deviation of the calculations (3rcombined = 130 pcm). However,it is significant for thorium since a 400 pcm effect is observed. Thiseffect is quite understandable when looking at Fig. 10, where it canbe seen that the JEFF-3.1 and ENDF/B-VII.0 evaluations of capturecross sections are quite different between 1 eV and 10 keV.

Table 10Spectral (q4eV) indices calculated with APOLLO2 and MORET 5.

Code q4eV

APOLLO2 at zero leakage – central lattice 0.047APOLLO2 at zero leakage – driver lattice 0.817APOLLO2–MORET 5 0.717

Table 11keff results with various codes and libraries – continuous energy codes.

Code Library (rMC in pcm)

JEF2.2 JEFF-3.1 ENDF/B-VII.0

Thermal scattering data for 2D not taken into accountMORET 5 1.00241 (30) 0.99482 (30) 1.00044 (30)MCNPX 2.6 1.00174 (30) 0.99286 (30) 0.99868 (30)

Scattering data of 2D taken into accountMORET 5 0.99935 (30) 0.99141 (30) 0.99659 (30)MCNPX 2.6 0.99786 (10) 0.98892 (10) 0.99480 (11)

Effect in pcmMORET 5 �306 �341 �385MCNPX 2.6 �388 �394 �388

Table 12Effect (in pcm) of probability tables with various codes and libraries.

Code Library

JEF2.2 JEFF-3.1 ENDF/B-VII.0

MORET 5 165 201 142MCNPX 2.6 149 147 158

538 N. Leclaire et al. / Annals of Nuclear Energy 76 (2015) 530–539

Table 13keff results for ICSBEP benchmarks.

Benchmark (q4eV) Case Benchmark keff (1runcertainty in pcm)

keff (rMC) MORET 5JEFF-3.1

HST-020 (0.644) 5 0.9966 (1160) 1.01162 (50)HST-004 (0.584) 1 1.0000 (325) 0.98427 (100)LMT-001 (0.805) 1 0.9990 (570) 0.99666 (100)

Finally, the impact of the thorium cross sections evaluationmainly explains the tendencies found in Table 3.

5.2. keff results with multi-group codes

The keff results calculated with the multi-group route of thecodes mentioned in Section 3 are reported in Table 9.

The spectral indices calculated by the APOLLO2 and MORET 5codes are given in Table 10.

5.2.1. Spectral indicesAt first, the attention is focused on spectral indices such as the

slowing down factor at 4 eV. It is calculated separately for eachzone for the infinite fissile medium by the APOLLO2 cell code

Table 14Selection of experiments with heavy water.

Benchmark identifier Title

HEU-SOL-INTER-001 Reflected uranyl-fluoride solutions in heavyHEU-SOL-THERM-004 Reflected uranyl-fluoride solutions in heavyHEU-SOL-THERM-020 Unreflected cylinders of uranyl-fluoride soluHEU-COMP-INTER-006 Highly enriched uranium dioxide cylinders iHEU-COMP-THERM-017 RB reactor: lattices of 80%-enriched uraniumLEU-MET-THERM-001 RB reactor: natural-uranium rods in heavy wHEU-COMP-MIXED-002 Highly enriched uranium dioxide cylinders i

and then for the total configuration by the MORET 5 code. Thisslowing down factor is an estimate of the proportion of fission neu-trons slowing at energies lower than 4 eV. It appears that the con-figuration is globally thermal and that the neutrons of the centralzone are slowed down in the peripheral zone or in the reflector.

5.2.2. Effect of multi-group treatmentWhen comparing the continuous energy MORET 5 keff results

with the ones obtained with the APOLLO2–MORET 5 multi-grouproute using the same evaluation, it can be concluded that themulti-group treatment of cross sections only has a small impact(around 300 pcm) for the new calculation options retained forCRISTAL V2.0, whereas a bigger one is observed with JEF2.2 (usedin the former criticality package).

5.2.3. Effect of nuclear dataWhen shifting from the 172-energy-group structure using the

JEF2.2 library to the 281-energy-group structure using the JEFF-3.1 library, the keff is decreased by 1200 pcm. Since the energygroup structure influence on keff is marginal, it shows the potentialinfluence of cross section data.

Indeed, the value of the discrepancy of keff is very close to theone highlighted for the continuous energy codes (800 pcm).

5.3. Impact of thermal-scattering data of heavy water

The thermal-scattering of the chemical link between deuteriumand oxygen is tested by switching on or off the thermalizationmatrices available in MCNPX 2.6 and MORET 5. It can be shownfrom Table 11 that its impact on keff is significant but lower than500 pcm.

Moreover, the effect is the same either with the MCNPX 2.6code or with the MORET 5 code, which tends to show that thetreatment of thermal-scattering of cross sections is correctly donein the MORET 5 code and cannot explain the 200 pcm discrepancyobserved between MORET and MCNP in Table 3.

5.4. Impact of the probability tables

In this paragraph, the attention is focused on the influence ofthe probability tables, which have been recently implemented inthe release 5.C.1 of MORET 5 code. Calculations were performedwith the MORET 5.B.1 (no probability tables), MORET 5.C.1 (prob-ability tables switched on) and MCNPX 2.6 codes.

The effect is expected to be small since the energy spectrum ofthe experiment is thermal.

From Table 12, it can be concluded that the order of magnitudeof the effect is code independent. The implementation of themethod to calculate the probability tables in MORET 5 seems tobe validated, as the results are similar to the MCNP ones. However,this work is a first step and needs further developments to fullyvalidate the implementation.

Finally, the treatment of the thermal-scattering data, based onACE files being the same between the MCNPX 2.6 and MORET 5codes, and the impact of the probability tables being quite similar,

Number of cases

water 6water 6tions in heavy water 5mmersed in mixtures of light and heavy water 23

elements in heavy water 9ater 1

mmersed in mixtures of light and heavy water 23

N. Leclaire et al. / Annals of Nuclear Energy 76 (2015) 530–539 539

one cannot exclude a discrepancy in the description of the modelwith the two codes to explain the systematic 200 pcm discrepan-cies between the MCNP and MORET 5 codes.

5.5. Comparison with other experimental programs

It was shown previously that the reactivity worth of heavywater in the central lattice was significant. As a consequence, itis paramount to have critical experiments that validate thecalculated results.

Thus, the keff results and observed tendencies are comparedwith the keff results of other programs involving thorium or heavywater in the same energy spectrum. The objective is to separate thebiases associated with the calculation of heavy water and thoseassociated with the calculation of thorium.

5.5.1. Benchmarks involving heavy waterWhen looking in the DICE database (using the graphical inter-

face) for the experimental programs (see ICSBEP Handbook NEA/NSC/DOC(95)03, 2011) involving either UO2 rods with a highenrichment in 235U (HEU) or solutions highly enriched in uranium,moderated by heavy water and with a thermal, mixed or interme-diate energy spectrum, the identifiers of Table 14 are returned.

No experiment with highly enriched rods was available. How-ever, some experiments involving solutions (HEU-SOL-THERM-004 and HEU-SOL-THERM-020) were calculated with the MORET5 continuous energy Monte Carlo code (See Table 13).

Moreover, a critical experiment involving uranium-metal rodswith a low 235U enrichment, moderated by heavy water isavailable. The ICSBEP identifier is LEU-MET-THERM-001. Theseexperiments present a neutron thermal spectrum comparable tothe studied configuration .

No clear tendency can be derived from HST-020 since the exper-imental uncertainty at the 1r level is high. Regarding the HST-004series, it was outlined that the tendency to underestimate keff couldbe due to the presence of heavy water as a reflector. This is not thecase in our configuration.

Moreover, the closest benchmark in terms of representativenessto the studied configuration is LMT-001. A good agreement beingobserved with the MORET 5 code using the JEFF-3.1 evaluationfor this type of rods in heavy water, as well as in light water forother experimental programs, it can be concluded that heavy wateris not the source of a bias.

As a result, the bias observed in the calculation of the ETA-1reactor should be explained by the calculation of thorium. As amatter of fact, thorium has a great impact on keff and the two

ENDF/B-VII.0 and JEFF-3.1 libraries are not necessarily inaccordance with one another, as far as thorium is concerned, asdiscussed in previous sections (see Fig. 10).

5.5.2. Benchmarks involving thoriumWhen looking in the DICE database at the experimental pro-

gram (see ICSBEP Handbook NEA/NSC/DOC(95)03, 2011) involvingthorium in the fuel region, moderated by light water and with athermal, mixed or inter energy spectrum, only two series of exper-iments are available (HCT-015 and HCT-021). These series wereidentified as being of interest for the MORET 5 validation database.However, none of them has already been calculated. It will be theobject of future works.

6. Conclusion

The experiments realized on the ETA-1 reactor allow validatingthe APOLLO2–MORET 5 and MORET 5 codes and deriving calcula-tion biases for nuclear systems involving thorium and heavy waterin the thermal energy range.

The influence of various sets of cross sections is alsoinvestigated. A noticeable impact of the thorium evaluation ishighlighted, the JEF2.2 library leading to results closer from thebenchmark.

Moreover, it is shown that the impact of the thermal-scatteringdata of 2D is significant. As for the use of the probability tables forthe determination of cross sections in the unresolved energy range,it has a lesser impact.

Finally, the comparison of the results between the MCNPX 2.6and MORET 5 continuous energy Monte Carlo codes using the samedata libraries contributes to the validation of the models imple-mented in the MORET 5 code.

References

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Gomit, J. M., Cousinou, P., Diop, C., Fernandez de Grado, G., Gantenbein, F., Grouiller,J. P., et al., 2003. CRISTAL V1: criticality package for burnup credit calculations.In: Proc. ICNC2003, Tokaï.

Gomit, J.M., et al., 2011. CRISTAL criticality package twelve years later and newfeatures. In: Proc. ICNC2011, Edimburgh.

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