studsvik’s next generation nodal code simulate-5

Advances in Nuclear Fuel Management IV (ANFM 2009) Hilton Head Island, South Carolina, USA, April 12-15, 2009, on CD-ROM, American Nuclear Society, LaGrange Park, IL (2009)

© ANS 2009, Topical Meeting ANFM 2009, p. 1/12

STUDSVIK’S NEXT GENERATION NODAL CODE SIMULATE-5

Tamer Bahadir Studsvik Scandpower, Inc.

1087 Beacon St. Suite #301, Newton, MA 02459-1700, USA [email protected]

Sten-Örjan Lindahl

Studsvik Scandpower AB Hantverkargatan 2A, SE-722 12 Västerås, Sweden

[email protected]

Keywords: SIMULATE-5, Nodal Method, Core-Tracking, Gamma Scan

ABSTRACT

SIMULATE-5 is Studsvik’s next generation nodal code which has been developed along with CASMO-5, Studsvik’s next generation lattice physics code, to address the deficiencies of existing reactor physics tools for today’s aggressive core designs with more heterogeneous fuel, extended cycle lengths, and operation strategies. This paper discusses the new neutronic and thermal-hydraulics models employed in SIMULATE-5 as well as benchmarking activities against theoretical problems, critical experiments, core follow calculations, and gamma scan evaluations.

1. INTRODUCTION

Studsvik’s nodal code SIMULATE-31 has been in use for LWR reactor analysis for nearly 25 years. The model has been proven to be accurate for typical LWR applications. Today’s aggressive core designs with more heterogeneous fuel and extended cycle lengths and operation strategies require more comprehensive reactor physics tools. New emerging issues in the industry also demand improved accuracy in pin powers and hence, thermal margins related to pin loads. To address these concerns Studsvik has developed the next generation lattice physics code CASMO-52 and the nodal code SIMULATE-53,4,5. (Note that the development of the new nodal code was started under the product name SIMULATE-4, and following the completion of the development, the new code was renamed SIMULATE-5.)

SIMULATE-5 is largely based on true physics and true geometry and avoids ad hoc models. In addition to improving accuracy, SIMULATE-5 provides more detailed information about the reactor core and its components than SIMULATE-3.

T. Bahadir, et al, Studsvik’s Next Generation Nodal Code SIMULATE-5


In the first part of this paper, the geometry, cross section, neutronic, and thermal-hydraulics models employed in SIMULATE-5 are summarized. Various validation results are presented in the second part.

2. MODELS

2.1 Geometry

The basic geometry unit in SIMULATE-3 is the node, defined as 1/N:th of an assembly (typically, N=24). The ‘conventional’ node may contain several material zones and hence, be strongly heterogeneous.

In SIMULATE-5 the basic unit is the ‘subnode’, defined such that it is materially homogeneous in the axial direction. The complexities of spacers, control rods and their zonings, enrichment/BA zoning, and reflector material at the assembly end points are thus all taken into account.

The subnode layout may differ from one assembly design to another. Also, as control rods move, the boundaries move.

Burnup and nuclide data are stored subnode-wise. Thermal margins are evaluated in this geometry. Hence, the uncertainty of the meaning of burnup or thermal margin in a heterogeneous, conventional node is avoided.

In the x/y direction, an assembly is divided into N×N ‘submeshes’ (typically N=5) where the submeshes follow pin cell boundaries. For BWRs, the outer level of submeshes is made up of the water gap region. For PWRs, the outer submesh layer is chosen to capture intra-assembly mismatch effects. Submesh cross sections and discontinuity factors are generated from CASMO-5 parallel to assembly average data.

The conventional ‘node’ concept (sometimes also called 3D nodes below) is used also for SIMULATE-5. The node is used for coupling the axial subnode and the radial submesh calculations as will be described later.

2.2 Cross section model

The cross sections are evaluated by a hybrid macroscopic/microscopic model:

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Approximately 50 isotopes (17 actinides and 30+ fission products/burnable absorbers) are tracked for each node. Nuclides have been chosen based on their impact on reactivity and on their interest from a safeguard point of view (special nuclear materials).

In addition to these 50 isotopes, five important actinides (U235, U238, Pu239, Pu240, Pu241) are tracked for each submesh in order to accurately describe the internal radial skewness of the nuclide distribution of an assembly. The cross section model faithfully reproduces history effects. The shutdown cooling phenomenon is automatically taken



into account and as-built enrichment and fuel weight appear naturally in the hybrid model.

SIMULATE-5 includes a new control rod depletion module for BWRs. The new model keeps track of the depletion of active absorber material (either B10 and/or Hf) and the fast fluence for each wing of the control rod with detailed axial zoning. The reactivity effect of the depletion of an active absorber is considered via cross section feedback to the nodal solution.

2.3 Neutronic Model

SIMULATE-5 solves the multi-group nodal diffusion equation as described below3. Macroscopic and microscopic cross sections and assembly discontinuity factors are generated from single-assembly CASMO-5 calculations with the assumption of a zero current boundary condition. The number of energy groups can be any subset of that used in the two-dimensional transport solver of CASMO-5 (typically 19 for UO2 or 35 for MOX fuel).

In the case of cores loaded with MOX fuel, it is known that methods based on P1 or diffusion theory are not capable of capturing the strong transport effects at the MOX-UO2 interface. Accordingly, the higher order SP3 (simplified P3) method is implemented in SIMULATE-5 as an option.

The core flux is found by a three-step approach:

1) The 1D diffusion equation is solved, one assembly at a time using the ‘subnode’ geometry3. The radial leakage, obtained from the 3D solution (step 3) is converted into an equivalent absorption. The obtained detailed 1D flux is employed to compute flux weighting factors needed when computing homogenized cross sections for the ‘conventional’ nodes. Also, axial discontinuity factors are computed.

2) The 2D diffusion equation is solved, one axial plane at a time, relying on the submesh geometry3. The axial leakage, known from the 3D solution, is converted into an equivalent absorption. The resulting flux solution is then used to compute homogenized cross sections and radial discontinuity factors for the 3D nodes. The 2D submesh approach overcomes the simplifications of the original CASMO-5 solution where homogenized cross sections and discontinuity factors are based on zero net current boundary conditions.

3) The 3D diffusion equation is solved using conventional 3D nodes with homogenized cross sections and discontinuity factors obtained as indicated above. The analytic nodal method (ANM) solution technique is employed. The only approximation made in ANM (within diffusion theory) is for the estimate of the shape of the so-called transverse leakage. However, the shape is known in some detail from the 2D submesh solution, and the traditional quadratic shape approximation need not be made.



Note, that the 3D solution is needed to tie the 1D axial and the 2D radial solutions together.

2.4 Pin Power Reconstruction Model

The flux and power of a pin in a 3D node is computed by synthesizing the 1D, 2D, 3D, and CASMO-5 solutions4.

The 2D solver provides a detailed so-called homogeneous flux in the x/y directions of each submesh. This flux takes into account radial variations of burnup, isotopes, fuel temperature, xenon and density inside the assembly. From the 1D and 3D solvers the homogeneous axial flux shape is known for each pin and subnode of the full 3D node. Finally, CASMO-5 form functions are superimposed subnode-wise on the homogeneous flux to produce pin powers. Note, that the pin power may be discontinuous from one subnode to another if there are material discontinuities.

Pin burnups are found by time integration of pin powers and taking into account the pin wise loadings as opposed to modulating the node homogenous burnups with exposure form functions.

The neutron detector response is computed using the reconstructed flux at the detector position. The gamma detector response in a BWR is found using pin powers of the four surrounding assemblies and detector response functions known from CASMO-5:

!!!!!!!!!!+% " 5%6789: / ;00 <0 (2)

The weights wi depend on the detector-pin distance and the angle of view.

2.5 Thermal-Hydraulics Model

SIMULATE-5’s BWR thermal-hydraulics (TH) models the entire vessel loop: core, chimney (for natural circulation reactors), upper plenum, standpipes, steam separators, down comer, re-circulation pumps, and lower plenum5. The PWR thermal-hydraulics models the region from lower to upper tie plate. The BWR and PWR core portion of the TH models are treated essentially identically, with each assembly having an active channel and a number of parallel water channels. In each axial node of a channel, the total mixture mass, steam mass, mixture enthalpy, and mixture momentum balance equations are solved (making SIMULATE-5 a four-equation TH model). Void fractions are determined by a drift flux model.

Correlations for the boiling process and for two-phase multipliers can be chosen from a library of functions and tied individually to the various assembly types of the core.

The 3-D fuel temperatures are evaluated in the TH module by solving the radial Fourier heat conduction equation for the average pin of each node.



The BWR assembly may be divided into four radial sub-channels, which communicate via cross flow (or are closed for SVEA type fuel). The cross flows are determined by solving lateral momentum equations.

The PWR core treats assembly cross flow by solving the axial and lateral momentum equations as in COBRA IIIC6.

Thermodynamic quantities are evaluated by using the NIST/ASME7 steam/water function library.

3. BENCHMARKING

3.1 Critical experiments

The widely used B&W 18108 series of critical experiments which provide high quality pin-by-pin fission rate (power) measurements are used for validating SIMULATE-5’s pin power reconstruction model. Relatively large cores (five fuel assemblies in diameter) have been built with fresh fuel in two mechanical designs (15x15 fuel with small water rods and 16x16 fuel with large water rods). Fuel assemblies differ in enrichment and the number of gadolinium pins. For several core configurations, each pin in the central assembly has been gamma scanned to infer power distributions, and the pin-power uncertainties have been reported to be very small (~0.5%).

SIMULATE-5 calculations are performed in two, four, eight and 19 energy groups, with data generated from single-assembly CASMO-5 2D transport calculations run in 19 energy groups with the ENDF/B-VII library. Table 1 summarizes the critical eigenvalue calculated with SIMULATE-5 for all core configurations. The values reported for CASMO-5 are calculated from full core 2D transport solution run in 95 energy groups9.

Table 1. SIMULATE-5 Eigenvalues for BAW 1810 Criticals

Fuel!Assm

Design !2!grp 4!grp 8!grp! 19!grp

Average 1.00059 0.99943 0.99984 0.99937 0.99906

Stdev!(pcm) 47 59 59 61 63

Average 1.00239 1.00147 1.00168 1.00127 1.00098

Stdev!(pcm) 27 63 55 52 50

Average 1.00084 0.99971 1.00009 0.99963 0.99932

Stdev!(pcm) 77 92 86 89 90

CASMO"5

SIMULATE"5!

Cores !18"20 16x16

Al l !Cores

Cores !1"17 15x15

Cores

Pin-by-pin fission rates calculated with SIMULATE-5 are compared against the

six B&W core configurations having measured pin fission rates, as summarized in Table 2. Detailed fission rate comparisons for two of these cores are presented in Fig.1. Calculated and measured pin fission rates are normalized to unity for the central assembly. SIMULATE-5 results are generated with eight energy groups in the nodal



solution and with pin power form factors described by two-energy groups. For all core configurations, either setup with 15x15 or 16x16 fuel and different number of Gd bearing pins, the SIMULATE-5 calculations and measurements agree very well, having similar trends as seen between CASMO-5 and measurement.

Table 2. B&W Criticals: Fission rate (pin power) RMS error

Fuel!AssemblyCore !# Design !2!grp 4!grp 8!grp! 19!grp

1 2.46%!15x15!0!Gd 0.51 0.53 0.53 0.53 0.535 2.46%!1515!12!Gd 0.57 0.53 0.55 0.53 0.5412 4.02%!15x15!0!Gd 0.69 0.74 0.74 0.75 0.7614 4.02%!15x51!0!Gd 0.79 0.74 0.74 0.74 0.7518 4.02%!16x16!0!Gd 0.86 0.95 0.94 0.95 0.9520 4.02%!16x16!16!Gd 1.07 1.08 1.08 1.08 1.08

SIMULATE"5CASMO"5

Fig. 1 Fission rate (pin power) comparisons for B&W Cores 18 and 20

3.2 Numerical tests against higher order transport methods

Neutronic parameters calculated with SIMULATE-5 for various loading patterns of BWR and PWR core in 2D quarter-core geometry have been compared against the reference solutions generated from quarter-core 2D CASMO-5 transport calculations.

One of these numerical test cases is the OECD MOX benchmark problem10. The two-dimensional quarter-core PWR with MOX fuel is depleted to 8 MWD/kg at fixed boron concentration. For higher fidelity, the reference solution is obtained from CASMO-5 in 35 energy groups. SIMULATE-5 calculations are performed in two, four, eight and and 19 energy groups with the SP3 model. Fig. 2 presents the error in the core eigenvalue and the RMS error in pin powers where they are normalized to unity core-wide. Fig. 3 shows the error (calculated as [S5-C5]x100) in pin powers of SIMULATE-5 with eight-

!Water Core 18 Water Core 20

1.21 1.08 Measured 1.28 1.11 Measured0.1 0.8 (C5-Meas)*100 "0.1 0.5 (C5-Meas)*1000.0 1.1 (S5-Meas)*100 0.3 1.0 (S5-Meas)*1001.03 1.02 1.07 1.08 1.01 0.170.4 1.3 1.6 0.3 0.6 "0.60.5 1.5 1.6 ! 0.8 1.2 "0.61.00 1.01 1.23 ! 1.06 1.06 1.240.3 1.8 "2.0 ! "0.3 0.5 "1.6 Gd0.5 2.0 "2.2 0.1 1.0 "1.20.98 1.01 1.20 ! ! 1.04 1.08 1.270.6 0.5 "0.8 ! ! 0.7 0.2 0.30.7 0.6 "0.6 0.9 0.6 0.60.96 0.98 1.04 1.18 1.17 1.00 1.04 1.03 1.07 1.28 1.28 1.080.5 "0.2 "0.3 "1.2 "1.1 1.1 "0.5 "0.2 "1.1 "3.0 "1.4 0.60.6 "0.1 0.0 "1.3 "0.9 1.6 "0.4 "0.1 "0.8 "3.0 "1.4 0.70.94 0.95 0.96 0.97 0.97 0.92 0.89 1.02 0.97 0.16 0.97 1.05 0.96 0.160.3 0.1 0.0 "0.4 "1.1 "0.1 "0.2 0.7 "0.3 "0.5 2.7 "0.5 "0.1 "0.50.3 0.2 0.2 "0.3 "1.2 0.0 "0.4 0.6 "0.3 "0.5 3.1 "0.7 "0.2 "0.60.91 0.91 0.93 0.92 0.91 0.89 0.87 0.83 1.01 0.98 0.95 1.00 1.00 0.98 0.90 0.931.2 0.8 "1.0 "1.0 "0.6 "0.1 "0.2 0.9 1.2 1.9 0.2 "0.6 0.3 "0.9 1.4 0.21.2 0.7 "1.0 "1.1 "0.7 "0.2 "0.4 0.6 1.1 1.4 0.0 "0.4 0.0 "1.4 0.9 "0.6



group solution for every pin in the core at BOC and EOC. SIMULATE-5 calculations with four or more energy groups are in excellent agreement with the reference solution. The pin power accuracy obtained at BOC, with the core wide RMS error of 0.5% and the maximum error which is less than 2.5%, is maintained over the entire depletion.

Fig. 2 PWR 2D MOX core: Error in eigenvalue and pin power predictions.

Fig. 3 PWR 2D MOX core: Error in Pin Powers (Calc-Meas)x100 at BOC and EOC

3.3. Core-tracking calculations

As part of validation of SIMULATE-5, core tracking calculations for various operating BWR and PWRs have been performed.

"40

"20

0

20

40

60

80

100

120

0 2 4 6 8

delta!k"eff!(pcm

)

Exposure!(MWd/kg)

S3S5!2grpS5!4grpS5!8!grpS5!19!grp

0

0.2

0.4

0.6

0.8

1

1.2

1.4

1.6

0 2 4 6 8

Pin!RM

S!Error!(%)

Exposure!(MWd/kg)

S3S5!2grpS5!4grpS5!8!grpS5!19!grp

1.0"2.0

0.0"1.0

"1.0"0.0

"2.0""1.0BOCBOC

2.0"3.01.0"2.00.0"1.0"1.0"0.0"2.0""1.0

EOC



Results for Finnish reactor Olkiluoto-1 are shown in Table 3. Olkiliuoto-1, which is operated by Teollisuuden Voima OY (TVO), is an ABB-Westinghouse design BWR with 500 fuel assemblies. For the last 15 cycles of which the results presented here, assembly designs include advanced Areva 9x9 and Atrium 10x10 and GE14 bundles with part length rods. All cycles have gamma tip measurements.

Table 3. Results summary for TVO1

Hot Eigenvalue Cold Eigenvalue Radial RMS Axial RMS Total RMS

Ave Std

(pcm) Ave Std

(pcm) Ave Stdev Ave Stdev Ave Stdev S3 1.0040 170 1.0040 373 2.00 0.50 2.03 0.63 3.36 0.69

S5 2-grp 1.0028 155 1.0038 257 1.79 0.35 2.03 0.45 3.30 0.44 S5 4-grp 1.0030 161 1.0037 263 1.76 0.33 2.04 0.44 3.28 0.43

Data libraries for both SIMULATE-3 and SIMULATE-5 have been generated with CASMO-5 using ENDF/B-VII. The comparisons include more than 150 hot condition burnup steps, corresponding to TIP measurements, and 100 cold critical data points. Fig. 4 presents the hot eigenvalue trend over the last 10 cycles. The excellent accuracy of SIMULATE-3 is reproduced with SIMULATE-5 and is even further improved for the cold-critical eigenvalue. Comparison of the two-group model against the four-group model shows that the predictions are fairly insensitive to the number of energy groups in the solution.

Fig. 4 Hot Eigenvalue for TVO1

0.997

0.999

1.001

1.003

1.005

1.007

0 600 1200 1800 2400 3000 3600

K-e

ff

Cumulative Exposure (EFPD)

S3S5 2grpS5 4grp



3.4. Gamma Scan Comparisons

Oskarshamn-2 is an ABB-Westinghouse BWR with 444 fuel assemblies. In 2007, a total of 48 assemblies (Svea-96 Optima2 and Areva Atrium-10B designs) were gamma scanned11. Measurements also include a number of individual fuel rods from two of these assemblies. These gamma scan evaluations provide an excellent opportunity for validating not only the nodal model but also the pin power reconstruction model employed in SIMULATE-5.

Standard core tracking calculations with SIMULATE-3 and two-group SIMULATE-5 have been performed for the last few cycles before the gamma scan measurements. The nodal Ba140 distributions calculated with SIMULATE are compared to the measurement. The results are summarized in Table 4.

Table 4. Oskarshamn-2 assembly gamma scan statistics

Total RMS Radial RMS Axial RMS

SIMULATE-3 4.5% 1.7% 3.3%

SIMULATE-5 3.9% 1.5% 2.8%

The total RMS (3D) error is calculated using % difference (=[Calc-MSR]x100) on normalized distributions with 23 axial nodes for each bundle (the first and the last axial nodes are removed) and 48 bundles in total. The radial RMS (2D) error is calculated using one axially averaged distribution per bundle. Finally, the axial RMS (1D) error is calculated on normalized axial distributions with one value per axial plane.

It is worthwhile to note that the statistics for the gamma TIPs for the cycle prior to the gamma scans are in good agreement with those of the gamma scans: with SIMULATE-5 gamma TIP results of: 3D RMS=3.4%, 2D=1.2% and 1D=2.3%.

Fig. 5 shows the error in the radial and axial Ba140 distribution calculated with SIMULATE-5. The figure also marks the locations of the inserted rods as well as the two assemblies with pin gamma scans.

The pin Ba140 distributions are calculated by superimposing the Ba140 form factors obtained from CASMO-5 on the node average Ba140 distribution available from SIMULATE-5. Fig. 6 presents the accuracy of SIMULATE-5 for two bundles. Assembly-A has a control rod, 20% inserted. Overall very good agreement is observed between SIMULATE-5 and measurements for all pin locations.



Fig. 5 Assembly gamma scans, 2D and 1D comparison

(SIM-MSR)x100 1.8

1.9

1.1 2.8

0.5 -1.8 -0.1

1.1 -2.1

0.0 1.4 2.0

1.3

0.0

-0.9 1.3 2.7 2.1 -1.3

1.2 -2.5 -1.1 -0.3 -1.5 -0.6 -1.4 -2.8 -1.3 -0.3 0.9

-0.7 1.2

-1.9 0.8 -0.4

-0.9 2.3 -0.5 -0.9

-1.8 -1.8 1.4 1.1

-1.0 -2.2

1.4 -1.1

0.8

3D (Nodal) RMS (%)2D (Radial) RMS (%)1D (Axial) RMS (%)

1.73.3

3.91.52.8

SIM3 SIM54.5

0

5

10

15

20

25

0 0.5 1 1.5

MeasCalc.(S3)Calc.(S5)

-10.0 -5.0 0.0 5.0 10.0

S3-M (%)S5-M (%)

Assembly-A

Assembly-B



Fig. 6 Pin Gamma scans: 1D and 2D comparisons

4. SUMMARY

SIMULATE-5 is Studsvik’s next generation nodal code – now commercially available - which provides an accurate description of the neutronic and the thermal-hydraulics behavior of BWR and PWR cores. It has been developed with consideration for today’s, and future advanced fuel/core designs as well as aggressive operation strategies by taking advantage of current modern computer resources.

A comprehensive validation process has been completed. SIMULATE-5 calculated core eigenvalue and local pin powers are in excellent agreement with measurements in critical experiments and with higher order transport solutions in theoretical problems. Results from real core tracking calculations and gamma scan comparisons have demonstrated that SIMULATE-5 provides improved accuracy, especially concerning the cold critical eigenvalue and in pin powers, compared to existing methods. Combined with Studsvik’s new lattice physics code CASMO-5,

Assembly -A

-2.6 -0.6 0.9

-2.1 -1.2

-2.1 -0.8

1.3 -0.3 0.5

0.9

0.7 1.6

-0.2 -0.3 0.9

-0.2 -1.4 0.6

-0.5 2.1

-0.6 3.4

(SIM5-MSR) x 100

Assembly -B

2.3 0.9 0.1 0.4

-2.1 2.2 0.5 -0.4

-3.0 0.0 -0.3 1.1

-1.0 ... -0.1 0.1 0.0 1.1

-0.1 0.9

-0.5 0.1

-0.1

-2.0 -1.1 -0.2

0.9

(SIM5-MSR) x 100

0

5

10

15

20

25

0 0.5 1 1.5

Meas

Calc. (S3)

Calc. (S5)

-12.0 -6.0 0.0 6.0 12.0

S3-M (%)

S5-M (%)

-10.0 -5.0 0.0 5.0 10.0

S3-M (%)

S5-M (%)

0

5

10

15

20

25

0.3 0.6 0.9 1.2

Meas

Calc. (S3)

Calc. (S5)



SIMULATE-5 will be offering detailed solutions to the improved accuracy demand of the industry for years to come.

ACKNOWLEDGEMENTS

The authors would like to acknowledge TVO for making the operation data for Olkiluoto-1 available and OKG for kindly providing their gamma scan measurements.

REFERENCES

1. K.S. SMITH et al., “SIMULATE-3 Methodology,” Studsvik Report, SOA-95/18 (1995).

2. J.D. RHODES, K.S. SMITH, D. LEE, “ CASMO-5 development and applications,” PHYSOR 2006, Vancouver, Canada, 2006

3. T. BAHADIR, S.-Ö. LINDAHL, S. PALMTAG, “SIMULATE-4 Multi-Group Nodal Transport Code with Microscopic Depletion Model,” ANS Topical Meeting in Mathematics and Computation, Avignon, France (2005).

4. T. BAHADIR, S.-Ö. LINDAHL, “SIMULATE-4 Pin Power Calculations,” PHYSOR 2006, Vancouver, Canada (2006).

5. S.-Ö. LINDAHL, T. BAHADIR, G.M. GRANDI, “SIMULATE-4 developments,” PHYSOR 2008, Interlaken, Switzerland (2008).

6. D.S. ROWE, “COBRA IIIC: A Digital Computer Program for Steady State and Transient Thermal Hydraulic Analysis of Rod Bundle Nuclear Reactor Fuel Elements,” BNWL-1695 (1973).

7. www.nist.gov/srd/nist10.htm

8. L.W. LEWMAN, “Urania-Gadolinia: Nuclear Model Development and Critical Experiment Benchmark,” B&W 1810, DOE/ET/34212-41, Babcock & Wilcox (1984).

9. J.D. RHODES, Z.XU, D. LEE, K.S. SMITH, “Benchmarking of CASMO-5 ENDF/B-VII Nuclear Data Against Critical Experiments,” ANFM 2009, Hilton Head, South Carolina, USA (2009).

10. T. KOZLOWSKI, AND T.J. DOWNAR, “OECD/NEA and U.S. NRC PWR MOX/UO2 core transient benchmark,” NEA/NSC/DOC 20, (2003).

11. C. NETTERBRANT, “Oskarshamn-2 Gamma Scanning Cycle 32- Evaluation with CASMO-4E/SIMULATE-3,” OKG report, 2008-28602 (2008).

studsvik’s next generation nodal code simulate-5

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