Activation calculation of a reactor vessel and application for planning of radiation shielding measures in decommissioning work U. Hesse (GRS), K. Hummelsheim (GRS), R. Nagel (DSR) Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbh, Forschungsinstitute, 85748 Garching near Munich, Germany DSR Ingenieurgesellschaft mbh, Allee der Kosmonauten 28, 12681 Berlin, Germany Abstract: The knowledge of the activation level of materials, which have been exposed to neutron irradiation during lifetime of a nuclear facility, is important for decommissioning and for lifetime extension as well if this is intended. Besides direct measurements of material probes, the calculation of material activation can provide useful and important information with respect to the long term irradiation behaviour of the material of interest. The presentation gives an overview on state of the art of calculation methods for activation and shows examples of application with respect to decommissioning of NPP (Fig. 1.1) Greifswald Stade Obrigheim Rheinsberg Fig. 1.1: Current German decommissioning projects 1 ACTIVATION CALCULATION A PRACTICAL EXAMPLE With the decision to terminate the operational lifetime of nuclear power reactors in Germany until 2025 the performance of activation calculations on an engineering level for reactor pressure vessels (ferritic steel for the main body part and austenitic steel for the inner plates), its internals (austenitic steel) as well as for the surrounding biological shield (mainly concrete) gained increased attention. According to the tools and means available about 15 years ago a rather simple approach has been developed that was applied to practical tasks. Advantages and limitations of the applied approach are reported here.
1.1 The aim The primary aim of the activation calculation is to provide information about the activation of the reactor pressure vessel material and of the material of the reactor internals as well as of the material of the biological shield for planning of cutting, conditioning, packaging, storage and disposal as part of plant dismantling after final shut-down. For this aim only key isotopes need to be considered that are either important for dose rate assessment ( 60 Co) or for fulfillment of interim storage acceptance criteria (e.g. 60 Co, 63 Ni, 55 Fe, 14 C, 3 H). In case acceptance criteria of final repositories or release shall be met additional isotopes need to be declared and become a target for the activation calculation. 1.2 The simplified approach The following approach became the DSR standard one for the performance of an activation calculation for dismantling purposes: - 3d-CAD modelling of the reactor pressure vessel and of the internals (Fig. 1.2-1.3) - Simplification of the 3d-model for 1d-neutron-flux-calculation in r- and in z-direction (ANISN) [1] (Fig. 1.4) - Activation calculation based on 3-group neutron fluxes (DSR code) [2] (Fig. 1.5) - Expansion of 1d-fluxes for main geometry axes towards areas away from main axes by means of applying (r,z)-distributions of measured surface dose rates and/or samples taken at selected positions for analysis of isotope concentration ( 60 Co) (Fig. 1.6) - z complex CAD model r simplified model for 1d flux calculation in r- and z- direction 1d flux calculation in r- and z- direction with deterministic codes (e.g. ANISN) Fig. 1.2: The simplified approach flux calculation 2
Fig. 1.3: The target Example for neutron flux distribution Neutron energy boundaries fast >0,1 MeV; epi >0,414 ev; therm <0,414 ev 1E+15 1E+14 RPV 1E+13 1E+12 Insulation neutron flux [1/cm 2 *s] 1E+11 1E+10 1E+09 baffle fast epi therm 1E+08 barrel 1E+07 Bio-Shield 1000000 100000 150 170 190 210 230 250 270 290 310 Distance from core centre [cm] Fig. 1.4: Results of 1d neutron flux calculation Example for activity distribution, radial 1,00E+10 1,00E+09 1,00E+08 1,00E+07 Mn 54 1,00E+06 Fe 55 Spez. Act. [Bq/g] 1,00E+05 1,00E+04 1,00E+03 1,00E+02 Co 60 Ni 59 Ni 63 1,00E+01 1,00E+00 baffle barrel RPV plating RPV base material Mo 93 Nb 94 1,00E-01 1,00E-02 1600 1700 1800 1900 2000 2100 2200 2300 Distance from core centre [mm] Fig. 1.5: Results of 1d activation calculation 3
? normalised form function for expansion in z-direction 1,E+07 1,E+06 1,E+05? 1,E+04 ff 1,E+03 1,E+02 1,E+01 1,E+00 0 1 2 3 4 5 6 7 8 9 expansion of the 1d results based on dose rate measurements and samples to gain a 3d distribution z [m]? Fig. 1.6: Expansion of 1d results Taking into account that but requesting much more effort- 2d- or 3d- deterministic calculations could have been performed, the chosen approach could be characterized as a 'simplified' one. Some key aspects for the different steps of the calculation shall be mentioned here in order to give an idea about the complexity of the task: Neutron flux calculation - Material composition of reactor pressure vessel and internals - 1d-modelling of the complex 3d-structure, especially in z-direction - Consideration of core loading patterns, varying over years of plant history Activation calculation - Modelling of power production history - Flux condensation and used cross sections (Table of Isotopes, JEFF) Essential part of this procedure is the provision of samples and of dose rate measurements in order to verify and/or adjust the results of the 'simplified' calculation scheme. 1.3 Verification of calculated data In general the number of samples that allow direct comparison between measured and calculated data is limited. It became more common in practice to perform an indirect comparison using dose rate measurements and results of dose rate calculations performed with a source term resulting from the activation calculation. The dose rate calculation is usually performed with computer codes that allow 3d-modelling of the geometric structure and provision of specific source strength distributions (especially important for large structures). 4
1 Drilling of samples for laboratory analysis (built-in structures) 2 3 Dose rate measurement (parts in cast iron 4 Dose rate measurement (parts) Dose rate measurement (parts in concrete container) Fig. 1.7: Proof by dose rate measurements and/or sampling 1.4 Assessment and interpretation of results Main result: The Comparison of measured and calculated data shows it works. Up to now technical measures had to be changed due to the inaccuracy of the activation calculation. So far all by calculation delivered data proved being conservative from the radiation protection point of view without losing too many degrees of freedom for practical work. Measured and predicted data agree quite well (differences in specific activity of 60 Co in steel components and of dose rates less than 50%, mostly much better). No results from sampling of concrete have been analysed yet. Discussing the satisfying results so far obtained by the 'simplified' approach it is always possible to say that the good agreement between measured and calculated data is poor chance. But it seems to be more appropriate to accept that reasonable application of the method produces reasonable results that allow proper planning of dismantling and disposal of activated parts of NPP. 1.5 Conclusion and Recommendation Despite the rather good results of the simple methodology the approach of developing 3d activity distributions from 1d flux calculations on the basis of dose rate measurements or selective sampling its limitations should not be neglected: specific effects of neutron 5
distribution inside a reactor pressure vessel and in its surrounding vicinity cannot be described accurately. Progress in dismantling technology, requests for economic treatment of radioactive waste, minimisation of waste volume in final repositories and the ALARA principle demand progress in calculation techniques also on the engineering level. One way to cope with requests from practice is the provision of qualified 2d or 3d deterministic calculations, ready-made for engineering applications (see part 2 of this paper). From practical experience two additional aspects should be given as lessons learned : - Any calculation needs to be verified by proper sampling and radio-chemical and spectrometric sample analysis. - The awareness of reactor operators (who's facility is to be dismantled at a certain moment of life time) for the complexity of the physical problems of calculation of isotope concentrations in material as a precondition for dismantling and disposal planning should be awakened. In comparison the problems of mechanical engineering in a dismantling project are in general understood much better. 2 DEVELOPMENT OF ADVANCED MOREDIMENSIONAL METHODS AND LIBRARIES FOR ACTIVATION CALCULATION Since several years GRS uses own-developed code systems for material activation calculation, whereby the well known ORNL code ORIGEN is applied as a main tool. The standard method is the GRSAKTIV code system [3], where ORIGEN runs in a loop over multiple material regions with different irradiation conditions of neutron flux strength and spectra, but with the same irradiation time history. Pre-calculated multi-group neutron fluxes and cross sections are used in 84 neutron groups. The ORIGEN libraries inside GRSAKTIV are based on ENDF/B-V with 6 nuclear reaction channels and 3 neutron energy groups up to 10 MeV. Currently an extended version GRS-ORIGENX [3] (<X> for extended) including new updated libraries based on modern nuclear point data files is being developed for practical application with 15 nuclear reaction channels and 6 neutron energy groups up to 20 MeV. In former versions of ORIGEN only 10 irradiation time steps could be used. Now a maximum of 999 time steps can be handled in double precision mode. The decay data are taken from ENDF/B-VI data bases, the cross sections from point data files JEF2.2, ENDF/B-VI, JENDL3.2 and EAF97. Due to known contaminations of structure materials by uranium and thorium, the build-up and depletion chains of the heavy metal isotopes can also completely be recalculated in the same way as the build-up chains of induced fission products. More than 20 fission yield sets are taken from ENDF/B-VI data bases. The new generated ORIGEN libraries have also been successfully checked for reactor decay heat conditions. Fig. 2.1 shows the extended calculation capabilities of the GRS-ORIGENX code, where 14 instead of only 6 standard reaction channels {(n,g), (n,g1), (n,a), (n,p), (n,2n), (n,2n1)} are built in for structure materials up to lead and bismuth. Some problems exist in standard ORIGEN calculations and in its data libraries concerning structural material activation calculations. Now e.g. the Tritium activation 3 H from Boron and the 22 Na, 26 Al, 60 Fe or 93m Nb generating problems can be solved in a satisfying way. The lack of the standard method can shortly be demonstrated at these isotopes: In the first case of 3 H, the necessary reaction channel (n,d) was not available. In the second case of 22 Na an internal numerical problem suppressed the very small (n,2n) cross section (XS). In the third case of 26 Al, the XS starts beyond of the upper energy limit of 10 MeV. In the fourth case of the long living 60 Fe with decay to 60 Co, buildup XS and isotope itself were not present in the library, and in the fifth case of 93m Nb, we missed the buildup reaction channel (n,n1) of 6
inelastic scattering of neutrons on natural Niobium in the standard ORIGEN model. So an improvement of both, of the calculation method ORIGEN and of the related library, was necessary. Standard Extended Fig. 2.1: Neutron induced reaction channels in ORIGENX for structure materials 2.1 The GRS AAA_Activation Sequence for 2d multigroup and multiregion calculations Performing activation calculations in the environment of a nuclear reactor one has to solve three parts of calculations, what is called a full AAA_Sequence. The abbreviation AAA stands for the German words Abbrand as burn-up of fuel, Abschirmung as attenuation of neutrons and gammas and Aktivierung as activation of the irradiated materials. Firstly 1d, 2d or 3d burn-up calculations OREST/KENOREST [3] have to be started to find the neutron flux strength and the fuel composition in the half-burned reactor core region. Secondly one and more dimensional multigroup deep-penetration transport calculations ANISN / DORT [1] (PL=3, 83 groups or PL=5, 175 groups) from the core region to the chosen structure material region must be done to determine the attenuation factors and the neutron spectra. Here the SN-codes run in the <multiplication factor> mode. The up-scatter procedure is calculated in 32 groups to achieve the correct flux shape in the thermal energy range of the neutrons. The neutron spectra are necessary to achieve here the correct problem dependent neutron cross sections. The meshes of the 1d ANISN and the 2d DORT core/vessel/shield model are automatically calculated by the mesh generator, included in the system. And last but not least the activation calculations have to be done by the GRS- ORIGENX code for the structural material regions. For each part of the AAA_sequence a consistent set of burn-up, transport and activation libraries is used: All neutron cross sections will be completely recalculated from the mentioned point data files JEF2.2, ENDF/B-VI, JENDL3.2 and EAF97 for 500 structure material isotopes, heavy metal isotopes and fission product isotopes. The same neutron 7
cross section data will be used in the burn-up, the transport and the activation step of the AAA_sequence. Next fig. 2.2 shows the simplified program flow chart of the AAA_sequence. In a first calculation step the burn-up calculation is done by OREST. A 1d pre-calculation follows with the ANITABLE system [3], where the 1d transport code ANISN works in cylindrical geometry. With these 1d results the 2d transport code DORT in (r,z) geometry is started. In a third step the results (multiplication factors, fluxes) are used for additional calculations of spectra, dose rates or graphical applications. DORTABLE GRSAKTIV 1d OREST 1d ANITABL E 2d R-Z DORT GRS- ORIGENX ABBRAND 1d Burn-up calculation flux strength ABSCHIRMUNG Isotopic XS, 2d radiation transport calculation in the <criticality> mode, 2d neutron spectra calculation, 2d neutron attenuation factors AKTIVIERUNG Activation calculations for multiple material regions Fig 2.2: Program Flow Chart of the burn-up, shielding and activation AAA_Sequence 2.2 Applications with regard to the operation lifetime of a reactor A satisfying validation of the AAA-sequence could be shown in [4]. Here detailed activation calculations for three reactor fuel cycles had been done using the neutron flux strength, spectra and weighted cross sections calculated before. After testing our method in comparison to experimental data for steel [5] and concrete activation [6], the AAA-sequence was applied on calculations based on the operation lifetime of a real NPP. Fig. 2.3 presents the calculated total flux in the core, related to the core averaged value 1.000. The model starts with the coolant inlet on the left side and the heated coolant outlet on the right side. Behind you see the neutron flux in the active regions (yellow bars). At front of the figure the fluxes in the biological shield are presented (grey bars). Note the logarithmic scale of the data. 323 material regions with more than 10,000 meshes had been calculated by (r,z)-dort. The attenuation factors between core and biological shield reach values of 1.E-07. An inter-comparison with other cross section libraries ([4], table 4) showed a spreading of ~ +/- 20 % of the calculated fluxes in the vessel. This is very satisfying in the scientific view of deep penetration problems, but in the case of real decommissioning works of a NPP a re-adjustment of our theoretical 2d results to exact activation measurements would be as necessary, as shown in part 1 of this paper. 8
Lower and upper active part of the Core Normalized Neutron Flux Strength 1,00E+01 1,00E+00 1,00E-01 1,00E-02 1,00E-03 1,00E-04 1,00E-05 1,00E-06 1,00E-07 1,00E-08 Vessel 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 core surround water1 core basket water2 RV1 RV2 RV3 RV4 RV5 Concrete air isolation Bio1 Bio2 Bio3 core Bio4 core basket Bio5 RV2 RV5 Radial Layers Bio1 Bio4 Axial Layers Fig. 2.3: Neutron flux distribution in the RZ-model at core, vessel and biological shield, normalized to the active core region The vessel (blue bars) and the biological shield have been divided into five layers for which the activation was calculated separately by ORIGENX using a constant reactor flux during 40 years. In fig. 2.4 the selected results of most important isotopes of the GRSAKTIV activities are shown up to 100 years decay time. Up to one year a complicated mix of isotope activities is present and a lot of these isotopes ( 59 Fe, 51 Cr, 54 Mn etc.) disappear within the first year. Afterwards the activity is completely dominated by 60 Co, 55 Fe, and especially by 63 Ni (half life 100 a). The tritium concentrations are mainly based on the assumed lithium impurity in the steel composition. 1,00E+07 specific activity Bq/g 1,00E+06 1,00E+05 1,00E+04 1,00E+03 1,00E+02 1,00E+01 1,00E+00 1,00E-01 1,00E-02 1,00E-03 1,00E-04 H 3 C 14 CR 51 MN 54 FE 55 FE 59 CO 58 CO 60 NI 59 NI 63 NB 93M NB 94 MO 93 CS134 EU152 totals 1,00E-05 0 10 20 30 40 50 60 70 80 90 100 decay time Fig. 2.4: Averaged activities of important radioisotopes in the eactor vessel after 40 years operation time in units Bq/g versus cooling time up to 100 years 9
Trace Element Content Nuclide GRSAKTIV STAandard 40 years operation ORIGENX EXTended 40 years operation Difference (STA-EXT)/EXT % (ppm) activated (Bq/g) (Bq/g) 03-Li 1.0 3 H 3.22E+03 3.25E+03-1 06-C 2,100 14 C 1.42E+02 1.37E+02 +4 07-N 590 14 C included included 08-O 0.1 *) 14 C included included 14-Si 3,500 15-P 200 32 P 1.84E+02 1.99E+02-8 16-S 150 S 1.44E+01 1.52E+01-5 P 9.46E-01 8.53E-01 +11 17-Cl 0.04 36 Cl 1.31E-03 1.19E-03 +10 21-Sc 0.1 *) 46 Sc 2.25E+01 2.00E+01 +13 22-Ti 0.01 46 Sc included included 23-V 500 24-Cr 5,000 51 Cr 1.64E+04 1.41E+04 +16 25-Mn 10,000 54 Mn 5.22E+04 6.11E+04-15 26-Fe remainder Fe 9.21E+05 9.67E+05-5 Fe 2.43E+04 1.90E+04 + 27 Fe 0.0 5.52E-08 27-Co 300 60 Co 8.27E+04 8.25E+04 +0.2 28-Ni 9,000 Co 8.05E+03 4.31E+03 + 86 Ni 6.91E+01 6.42E+01 +8 Ni 7.88E+03 6.88E+03 +15 30-Zn 0.1 *) 41-Nb 19.5 42-Mo 7,000 51-Sb 0.1 *) 55-Cs 0.1 *) 58-Ce 0.1 *) 63-Eu 0.1 71-Lu 0.1 *) 72-Hf 0.1 *) 73-Ta 0.1 *) 90-Th 0.1 *) 92-U 0.1 *) additionally assumed 65 Zn 3.92E-01 3.95E-01-1 93m Nb 9.58E+00 8.21E+00 + 17 94 Nb 3.51E-01 3.72E-01-6 93 Mo 1.68E+01 1.40E+01 + 20 99 Mo 1.08E+01 7.49E+00 + 44 99 Tc 2.24E+00 1.58E+00 + 42 124 Sb 5.70E+00 3.05E+00 + 87 134 Cs 2.66E+01 2.75E+01-3 141 Ce 1.31E-01 1.26E-01 +4 152 Eu 4.24E+02 4.49E+02-6 177 Lu 3.12E+00 2.38E+00 + 31 181 Hf 1.18E+00 9.81E-01 + 20 182 Ta 4.47E-04 3.01E-04 + 49 233 Pa 2.10E+00 2.08E+00 +1 90 Sr 2.51E-02 2.45E-02 +2 Cs 2.68E-02 2.65E-02 +1 Pu 9.16E-03 8.79E-03 +4 Alpha-Activity < 1.E-01 < 1.E-01 Total Activity 1.12E+06 1.16E+06 +4 Table 2.1: Estimated contents of steel <22NiMoCr37>, trace elements [6], [7] and important radioisotope activities in the PWR reactor vessel one month after shut down The right column of table 2.1 shows the differences between our standard and our extended method. For the total activities and for the most important isotopes the differences are small with respect to the long irradiation time and different cross section libraries. For some isotopes 58 Co, 93m Nb, 99 Tc and 99 Mo, the differences can reach more than 40 %. The calculated alpha activities, resulting from the thorium and uranium contents and their activation products were found lower than 1.E-01 (Bq/g). 10
2.3 Analyses of important radioactive isotopes in the intermediate time range Our analyses are directed now to the intermediate time region up to 100 years for intermediate storage of the components of the reactor, and to the long term region up to 10,000 years important for final storage considerations. A maximum life time of 40 years of reactor operation has been assumed. For dose calculations outside of the irradiated materials not the activity but the gamma power is more important. Some of the most active isotopes as 3 H, 55 Fe or 63 Ni are weak gamma emitters or produce only beta-particles, X-rays and Bremsstrahlung. Fig. 2.5 shows the development of the most important gamma radiation emitters in the irradiated vessel up to 100 years decay time after shut-down. Here the gamma power is completely dominated by 60 Co (hard gammas with energies up to 1.3 MeV), and in the time range afterwards by 152 Eu (hard gammas with energies up to 1.4 MeV) and the long living isotope 59 Ni with relatively soft X-ray emission spectra up to only 8 kev. Due to the time dependent and very different gamma emission spectra, the resulting dose rates in the environments of such activated materials inside or without a shielding cask have to be calculated in a separate gamma ray transport calculation. This problem is discussed in the next chapter. ORIGENX: Reactor Vessel Gamma Power (W/g) after 40 years operation time specific Gamma Power W/g 1,00E-06 1,00E-07 1,00E-08 1,00E-09 1,00E-10 1,00E-11 1,00E-12 1,00E-13 1,00E-14 1,00E-15 0 10 20 30 40 50 60 70 80 90 100 decay time MN 54 FE 59 CO 58 CO 60 NI 59 NB 93M MO 93 EU152 EU154 EU155 totals Fig. 2.5: Averaged gamma power of important radioisotopes in the reactor vessel after 40 years operation time in units Bq/g versus cooling time up to 100 years 2.4 The future GRS AAA_Sequence for 2d multigroup and multiregion source term and dose rate calculations If the three steps of the AAA_Activation calculations have to be done as described above, in the future an automated module is planned to feed back the multi-region activation results into the DORTABLE system for more-dimensional gamma shielding and dose rate calculations. Fig. 2.6 shows the simplified program flow chart of this fourth step of the AAA_Sequence, where the activated materials could be packed into a shielding cask: 11
DORTABLE GRS-SURF NGSRC 1d ANITABLE 2d R-Z DORT GRS- FLXDOSD ORNL-FALSTF SOURCE n-g source terms for each activated structure region SHIELDING 2d radiation n-g transport calculation running in the <fixed source> mode DOSE Rate 2d/3d n-g dose calculation inside or outside of the DORTABLE geometric model Fig 2.6: The irradiation of reactor pressure vessels of NPP was used as a practical application of activation calculations for planning of radiation shielding measures in decommissioning work. Here 3d activity distributions have been determined from 1d flux calculations on the basis of dose rate measurements or selective sampling. This 1d/3d method should be supported additionally by extended 2d or 3d discrete ordinate calculations. This could be done by the 2d AAA-sequence of GRS. For a typical German PWR more- dimensional deep penetration neutron transport calculations were started for a consistent model of the complete system core water vessel - biological shield. Detailed activation calculations for a reactor operation life time up to 40 years had been done using explicitly the neutron flux strength, spectra and weighted cross sections calculated before for the reactor vessel, for the biological shield and for the upper and lower parts of the assembly structure materials. Program Flow Chart of the AAA_Sequence for source term, shielding and dose rate calculation In a first step the neutron-gamma source term generator NGSRC from our code-collection [3] would be started in a loop over each of the activated material zones. Secondly one and more dimensional multi-group deep-penetration transport calculations with ANISN / DORT (PL=5, 175 neutron and 42 gamma groups) for the chosen activated structure material regions inside of a shielding cask are run in general for both neutron and gamma fluxes. Here the SN-codes run in the <fixed source> mode. In a last third step the results are used for additional calculations of dose rates with the codes GRS-SURF [3], GRS-FLXDOSD [3], ORNL-FALSTF [1] or for graphical applications. This part of the system at the moment has to be started by the user separately, but could be introduced in a coupled loop. A first approach to such a future system was done in the calculation case of a known shielding cask for burned reactor fuel, where an axial burn-up profile was modeled. These 2d/3d neutron-gamma shielding calculations and dose rate estimations are shown in the poster session of this meeting as a comparison between 2d DORTABLE and 3d MCNP results. 3 SUMMARY AND CONCLUSIONS 12
Practical experience of calculations of a reactor vessel activation and the application for radiation shielding in decommissioning work showed, that both of the methods, the 1d/3d DSR approach using ANISN and the 2d GRS activation system using DORT need for comparison proper sampling and radio-chemical and spectrometric sample analysis. This is not only true for 1d, 2d but also for 3d methods because of the spreading of used transport cross section libraries, uncertainties of the reactor operation and of the uncertainty in the knowledge of the impurities and material compositions. The awareness of reactor operators for the complexity of the physical problems of calculation of isotope concentrations in material as a precondition for dismantling and disposal planning should be awakened. 4 REFERENCES [1] DOORS3. 1 Package CCC-0650/01; One, Two- and Three Dimensional Discrete Ordinates Neutron/Photon Transport Code System, RSIC Computer Code Collection, August 1996; ANISN (K-1693, 03/1967); DORT (ORNL/TM-11778, 03/1992) [2] DSR Ingenieurgesellschaft, Programmbeschreibung AKT-2, internal document [3] Hesse, U., Gewehr, K., Moser, E., Hummelsheim, K., Langenbuch, S., Denk, W., Deitenbeck, Sieberer, H.J.: The GRS AAA-Code-Collection for Activation, Shielding and Burn-up GRSAKTIV (GRS-A-2249, 06/1995); GRS-ORIGENX-07 (Technical Report 02/2007); KENOREST (GRS-A-2783, 12/1999, updated 2006), OREST (GRS-63, 11/1986, updated 2006); ANITABLE (GRS-A-3004, 12/2001, unix- Version updated 2004), DORTABLE (Technical Report 03/2007) [4] [5] [6] Pretzsch, G. et.al.; Neutron activation of reactor components during operation lifetime of a NPP, Second International Symposium on Nuclear Power Plant Life Management, Shanghai, China, 15-18 October 2007 Martin, J., Bericht KWU R451/84/38, 13.03.1984, private communication with U. Hesse, 27.10.1994 Ken-ichi Kimura et.al., Compilation of neutron activation cross sections and trace element contents of concrete for estimating the induced radio activities, 8 th International Radiation Shielding Conference, Arlington Texas 1994 [7] Hesse, U., Hummelsheim, K., Bestimmung der Aktivierung des Strukturmaterials in einem Siedewasserreaktor durch dreidimensionale Rechenverfahren, GRS, Technical Report 2001 13