The mission of ITU is to provide the scientific foundation for the protection of the European citizen against risks associated with the handling and storage of highly radioactive material. ITU’s prime objectives are to serve as a reference centre for basic actinide research, to contribute to an effective safety and safeguards system for the nuclear fuel cycle, and to study technological and medical applications of radionuclides/actinides.
Report -No: JRC-ITU-TN-2008/22
Classification: Unclassified
Type of Report: Technical Report
Unit: Actinide Research / Nuclear Safeguards and Security
Action No: 53103 (FACIL)
Name / Date / Signaturereviewed by the
project coordinator /
or action leader / (A. Berlizov)
approved by the
project leader / (J. Magill)
approved by the head of unit / (K. Lützenkirchen)
released by the
director / (T. Fanghänel)
European Commission
Joint Research Centre
Institute for Transuranium Elements
Contact information
Address: Dr. Andrey Berlizov, Institute for Transuranium Elements, Joint Research Centre, Postfach 2340, 76125 Karlsruhe
E-mail:
Tel.: +49 7247 951 586
Fax: +49 7247 961 99 586
http://itu.jrc.ec.europa.eu
http://www.jrc.ec.europa.eu
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1. Introduction
This work has been done on the request of the AGS ITU to evaluate the neutron dose rate from the TU-B-100 container with spent nuclear fuel wastes. The purpose of the calculation was to verify the compliance of the measured neutron dose rate (35 mSv/h ± 30% at 7.5 cm distance from the container sidewall, see Appendix A) with the declared quantities of the neutron emitting nuclides, obtained based on the burnup calculations. As provided by the AGS staff, these quantities are as follows: 238Pu - 1.7×1010 Bq, 240Pu - 2.3×109 Bq, 242Pu - 1.5×107 Bq, 244Cm - 5.6×1010 Bq, and 252Cf - 2.4×105 Bq.
2. Summary of the approach
Dose rates were evaluated by means of Monte Carlo calculations with application of the general purpose Monte Carlo N-Particle Transport Code MCNP4c (Briesmeister, 1997). Both neutron dose rate and prompt photon (from (n,gx)-reactions in the container walls) dose rate were evaluated. The calculations involve the convolution of calculated neutron and photon flux energy distributions with the respective fluence-to-effective dose conversion coefficients. The neutron and photon flux energy distributions were computed using non-analog (next-event-estimator or F5 tally) approach.
The anterior-posterior irradiation condition was assumed and the respective fluence-to-effective dose conversion coefficients were utilized (Pelliccioni, 2000). The neutron doses were evaluated in three different spectral regions, from 0 to 0.55 eV, from 0.55 to 105 eV, and from 105 to 2×106 eV, corresponding to thermal, epithermal and fast neutrons respectively.
3. Source properties
The pictures and drawing of the TU-B-100 container are shown in Figure 1 and 2. The respective MCNP model of the container setup is shown in Figure 3. The container is modeled by a cylindrical shell made of Pb (r = 11.342 g/cm3) and Fe (r = 7.875 g/cm3). The external dimensions of the shell are Æ64.5 cm ´ 95 cm, and the sidewall thickness is 16.75 cm (14 cm of Pb and 2.75 cm of Fe). From the open sides, the shell is plugged with the 10.5 cm Pb disk and the 85´100´14.5 cm3 Pb slab. A point-like source is placed inside the container in the middle of the cylindrical cavity, which has dimensions Æ31 cm ´ 95 cm. The dose rate was evaluated at 7.5 cm distance from the container sidewall at the point, closest to the source.
The neutron source properties are summarized in Table 1. The nuclide specific neutron emission rates and cumulative source strength shown in the Table were evaluated using data from Nucleonica (Magill et al., 2007) and from Perry and Wilson, 1981. The cumulative strength of the source turned out to be 2.3×105 n/s with the major contribution from spontaneous fission of 244Cm. The neutron spectrum p(E) was modeled using a composite Watt fission energy spectrum (Beckurts and Wirtz, 1964)
,
where sum is over the neutron emitting nuclides, wi is the relative contribution of the i-th nuclide to the total neutron yield (second column in Table 2), ai and bi are the nuclide specific parameters (shown in the third and forth columns of Table 2), and C is the normalization constant.
4. Results and discussion
The spectral distributions of the neutron and secondary photon fluxes are shown in Figure 4. The integrated over the spectrum neutron and secondary photon flux values at the point of the detector are 15.6 cm-2s-1 and 1.1 cm-2s-1, respectively.
The calculated partial and cumulative dose rates are summarized in Table 3. As it follows from the data shown, the prompt neutron capture photons negligibly contribute to the total dose rate outside the container. The major contribution to the neutron dose rate comes from the fast fission neutrons. This is because the lead shielding does not provide any significant thermalization effect on the primary fission energy spectrum.
The obtained neutron dose rate 12.55 mSv/hr explains nearly 36% of the measured value. Taking into account the possible biasing of the burnup calculations and simplicity of the geometry used in the Monte Carlo calculations (e.g. point-like source in the middle of the container), this result can be considered as a good support for the declared composition of the neutron emitting nuclides inside the waste. The remaining part of the measured dose rate can be attributed to neutrons from (a,n)-reactions, which occur in the waste material, and neutrons scattered in the measurement room. The neutron multiplication inside the waste can also contribute to the resulting dose rate outside the container. This contribution will depend on the amount of the waste and neutron moderation properties of the waste material (i.e. the amount of light elements present).
5. Conclusions
The calculations show that the declared composition of the neutron emitting nuclides in the spent nuclear fuel wastes agrees reasonably well with the neutron dose measurements.
If the elevated neutron dose rate from the container is an issue, then the introduction of an additional shielding layer made of some light material (e.g., polyethylene or paraphine) inside the lead container can be suggested.
Development of a Monte Carlo based neutron dosimetry and shielding module in Nucleonica is highly desirable, so that the AGS staff can use it for performing routine calculations.
6. References
K.H. Beckurts, K. Wirtz, Neutron physics, Springer-Verlag, 1964, p.456
J.F. Briesmeister. MCNP – a general Monte Carlo N-particle transport code. Los Alamos National Laboratory Report, 1997, LA-12625-M.
J. Magill, J. Galy, R. Dreher, D. Hamilton, M. Tufan, C. Normand, A. Schwenk-Ferrero, H. W. Wiese, NUCLEONICA: A Nuclear Science Portal, http://www.euronuclear.org/e-news/e-news-17/nucleonica.htm.
M. Pelliccioni, Overview of Fluence-to-Effective Dose and Fluence-to-Ambient Dose Equivalent Conversion Coefficients for High Energy Radiation Calculated Using Fluka Code, Radiation Protection Dosimetry, 88(4) (2000) 279–297.
R.T. Perry, W.B. Wilson, Neutron Production from (a,n) Reactions and Spontaneous Fission in ThO2, UO2, and (U,Pu)O2 Fuels, Los Alamos National Laboratory report LA-8869-MS (June 1981).
Fig. 1. Views of the TU-B-100 container.
Fig. 2. Drawing of the TU-B-100 container.
Fig.3. Calculation geometry (dimensions in mm).
Fig.4. Neutron and photon flux energy distributions at the detector position.
Table 1. Properties of the spent fuel waste as a neutron source.
Nuclide / Activity, Bq / Specific activity, Bq/g / Specific fission neutron yield, n/g×s / Neutron emissionrate, n/s
238Pu / 1.70E+10 / 6.34E+11 / 2.59E+03 / 6.94E+01
240Pu / 2.30E+09 / 8.40E+09 / 1.02E+03 / 2.79E+02
242Pu / 1.50E+07 / 1.46E+08 / 1.72E+03 / 1.77E+02
244Cm / 5.60E+10 / 3.01E+12 / 1.08E+07 / 2.01E+05
252Cf / 2.40E+05 / 1.98E+13 / 2.34E+12 / 2.84E+04
Total / 7.53E+10 / 2.30E+05
Table 2. The evaluated relative neutron emission rates and parameters of the Watt spontaneous fission energy spectrum (Briesmeister, 1997) used in the calculation.
Nuclide / Relative neutronemission rate, w, % / a, MeV / b, MeV-1
238Pu / 0.030 / 0.799* / 4.903*
240Pu / 0.122 / 0.799 / 4.903
242Pu / 0.077 / 0.834 / 4.432
244Cm / 87.430 / 0.906 / 3.848
252Cf / 12.341 / 1.025 / 2.926
* Because of the lack of data, the values of the Watt spectrum parameters for 238Pu were taken the same as for 240Pu
Table 3. Calculated partial and cumulative dose rates at 7.5 cm distance from the container sidewall.
Neutron dose rate: / Value, mSv/h / Statistical uncertainty, %Thermal (< 0.55 eV) / 0.00 / 0.00
Epithermal (0.55 - 105 eV) / 0.108 / 0.36
Fast (> 105 eV) / 12.44 / 0.08
Total / 12.55 / 0.08
Photon dose rate: / Value, nSv/h / Statistical uncertainty, %
(n,xg)-reactions / 13.9 / 0.37
Appendix A: The neutron dose rate measurement report (provided by the AGS staff)
Appendix B: The MCNP input for neutron dose rate calculations using F5 point-detector tally
Pu-Cm-Cf Spent Nuclear Waste: shielding calculations using F5 tally
c
c Source location volume (void)
1 0 -1 5 -6
c Cylindrical shell (Pb)
2 1 -11.342 1 -2 5 -6
c Cylindrical shell (Fe)
3 2 -7.875 2 -3 5 -6
c Pb disk
4 1 -11.342 4 -5 -3
c Pb slab
5 1 -11.342 6 -7 8 -9 10 -11
c Volume containing the geometry
6 0 -12 (3 : -4 : 7) #5
c The universe
7 0 12
1 cz 15.50
2 cz 29.50
3 cz 32.25
4 pz -58.00
5 pz -47.50
6 pz 47.50
7 pz 62.00
8 px -50.00
9 px 50.00
10 py -42.50
11 py 42.50
12 so 150.00
imp:p 1 5R 0
imp:n 1 5R 0
imp:e 1 5R 0
mode n p
sdef par=1 pos=0 0 0 erg=D1
SI1 S 11 12 13 14 15 16
SP1 D 0.030 0.122 0.077 0.000 87.430 12.342
c Neutron spectrum for Pu-238
SP11 -3 0.799 4.903
c Neutron spectrum for Pu-240
SP12 -3 0.799 4.903
c Neutron spectrum for Pu-242
SP13 -3 0.834 4.432
c Neutron spectrum for Cm-242
SP14 -3 0.891 4.046
c Neutron spectrum for Cm-244
SP15 -3 0.906 3.848
c Neutron spectrum for Cf-252
SP16 -3 1.025 2.926
c Neutron flux energy distribution
f5:n 39.75 0 0 1
e5 1-9 1.7782794-9 3.1622777-9 5.6234133-9
1-8 1.7782794-8 3.1622777-8 5.6234133-8
1-7 1.7782794-7 3.1622777-7 5.6234133-7
1-6 1.7782794-6 3.1622777-6 5.6234133-6
1-5 1.7782794-5 3.1622777-5 5.6234133-5
1-4 1.7782794-4 3.1622777-4 5.6234133-4
1-3 1.7782794-3 3.1622777-3 5.6234133-3
1-2 1.7782794-2 3.1622777-2 5.6234133-2
1-1 1.7782794-1 3.1622777-1 5.6234133-1
1+0 1.7782794+0 3.1622777+0 5.6234133+0
1+1 1.7782794+1 3.1622777+1 5.6234133+1
c
c Neutron dose in Sv/hr
f15:n 39.75 0 0 1
e15 0.55e-6 0.1 20 $ thermal, epithermal and fast neutrons
c Neutron fluence-to-effective dose conversion coefficients (Sv.cm2)
c for AP (aposterior-posterior) irradiation conditions
c Radiation Protection Dosimetry, V.88, 2000
# de15 df15
2.5-8 8.65-12
1.0-4 1.73-11
1.0-2 3.31-11
1.0-1 4.50-11
1.0+0 2.91-10
1.0+1 5.89-10
3.0+1 3.93-10
c Normalization: I(n/s)*3600(s/hr)=2.3e5*3600=8.28e8
em15 8.28+8 8.28+8 8.28+8
c
c Secondary gamma-ray flux energy distribution
f25:p 39.75 0 0 1
e25 1e-6 100I 20
c
c Dose associated with secondary gamma-rays in Sv/hr
f35:p 39.75 0 0 1
e35 1e-6 100I 20
c Photon fluence-to-effective dose conversion coefficients (Sv.cm2)
c for AP (aposterior-posterior) irradiation conditions
c Radiation Protection Dosimetry, V.88, 2000
# de35 df35
0.05 3.68-13
0.10 5.13-13
0.50 2.48-12
1.00 4.47-12
1.50 6.13-12
2.00 7.47-12
3.00 9.94-12
4.00 1.22-11
5.00 1.36-11
6.00 1.52-11
8.00 1.82-11
10.00 2.16-11
20.00 3.44-11
30.00 4.54-11
c Normalization: I(n/s)*3600(s/hr)=2.3e5*3600=8.28e8
em35 0 8.28+8 100R
c
c Materials
m1 82000 1.000 $ Pb
m2 26000 1.000 $ Fe
c void
c
c Number of histories
nps 2+8
European Commission – Joint Research Centre – Institute for Transuranium Elements
Title: Neutron Dose Rate Calculation for a Shielded Spent Nuclear Fuel Waste
Author(s): Andrey Berlizov
2008 – 10 pp. – 21.0 x 29.7 cm
Abstract
The neutron dose rate from a shielded spent nuclear fuel waste was evaluated using the MCNP-4C code. The calculations showed a reasonable agreement between the calculated and measured neutron dose rates. Through this agreement, an additional support was provided for the declared contents of the neutron emitting nuclides (238,240,242Pu, 244Cm and 252Cf) in the waste, which were established based on the burnup calculations.
Distribution List
T. Fanghänel (Director) / ITU / 1xF. Wastin (Programme) / ITU / 1x
G. Weber (Registration & Archives) / ITU / 2x
W. Wagner / ITU / 1x
R. Caciuffo / ITU / 1x
K. Lützenkirchen / ITU / 1x
A. Berlizov / ITU / 1x
J. Magill / ITU / 1x
3
The mission of the JRC is to provide customer-driven scientific and technical support for the conception, development, implementation and monitoring of EU policies. As a service of the European Commission, the JRC functions as a reference centre of science and technology for the Union. Close to the policy-making process, it serves the common interest of the Member States, while being independent of special interests, whether private or national.3