NEUTRON ACTIVATION OF REACTOR COMPONENTS DURING OPERATION LIFETIME OF A NPP

Transcription

1 NEUTRON ACTIVATION OF REACTOR COMPONENTS DURING OPERATION LIFETIME OF A NPP G. Pretzsch, B. Gmal, U. Hesse Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbh, Germany address of main author: prg@grs.de Abstract The knowledge of the activation level of materials, which have been exposed to neutron irradiation during the lifetime of a nuclear facility, is important for decommissioning and for lifetime extension as well if this is intended. Besides direct measurement of material probes, the calculation of material activation can provide useful and important information with respect to the long term irradiation behavior of the material of interest. This presentation gives an overview on state of art calculation methods for activation and shows examples of application with respect to decommissioning of NPP. 1. Introduction Since several years GRS uses own-developed code systems for material activation calculations, whereby the well known ORNL code ORIGEN is applied as a main tool. The standard method is the GRSAKTIV code system [1], 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. 2. Development of advanced methods and libraries for activation calculations Currently an extended version GRS-ORIGENX [1] 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 and 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. 1 shows the extended calculation capabilities of the GRS-ORIGENX code (<X> for extended), where 14 instead of only 6 standard reaction channels {(n,g), (n,g1), (n,α), (n,p), (n,2n), (n,2n1)} are built in for structure materials up to lead and bismuth. Now formerly existing problems, e.g. the Tritium activation from boron and the Na-22, Al-26, Fe-60 or Nb-93m generation can be solved. 1

2 2.1. The GRS AAA-Activation Sequence Calculating activation in the environment of a nuclear reactor one has to solve three parts, what we call 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/3d burn-up calculations KENOREST/OREST [1] 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 multi-group deep-penetration transport calculations ANISN/DORT [2] (level of anisotropy PL=3, 83 groups or PL=5, 175 groups) from the core region to the chosen structure material region must be done to find the attenuation factors and the neutron spectra. The up-scatter procedure will be done in 32 groups to achieve the correct flux shape in the thermal energy range of the neutrons. The neutron spectra are necessary to achieve 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. Lastly the activation calculations have to be done by the GRS-ORIGENX code for the structural material regions. FIG. 1 Neutron induced reaction channels in ORIGENX for structure materials. FIG. 2 Program Flow Chart of the burn-up and shielding sequence DORTABLE. For each part of the AAA-sequence a consistent set of burn-up, transport and activation libraries will be 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 cross section data will be used in the burn-up, the transport and the activation step of the AAAsequence. Fig. 2 shows the program flow chart of the two first parts of AAA-sequence. In a first calculation step the burn-up calculation is done by OREST. A 1d pre-calculation follows with the ANITABLE system, where the 1d transport code ANISN in cylindrical geometry works. With these results the 2d transport code DORT in RZ (R-radial axis, Z-axis) geometry follows. In a third step the results (multiplication factors, fluxes) are used for additional calculations of spectra, dose rates or graphical applications. The first two parts of the AAA- 2

3 sequence are linked together in the GRS code system DORTABLE [1] for practical appliance, which consists of the two main code systems OREST and DORT and some interface tools for data transfer and cross section handling, running in a UNIX or LINUX environment AAA-Calculations of the irradiation of steel and comparison with experiments First we will show the application of the new AAA-sequence for activation calculations of the steel upper and lower parts of the UO 2 fuel assembly BE318, which was irradiated three cycles in the German PWR BIBLIS A in 848 full power days for a burn-up of 33 GWd/tHM. The specific burn-up history of BE318 was considered. The results are compared with measured data [4]. Similar calculations have been done by the Research Center Karlsruhe [3] and independently by GRS [5]. The analyses [4] of activities at the upper part comprise three single measurements, averaged in table 1. Our calculation has been done using a DORT reactor core model in RZ geometry for the neutron transport in 187 material regions and more than 10,000 mesh points. We found in fact the same results as in ref. [3], namely an attenuated neutron flux at the axially lower part of 0.5 % and at the axially upper part of 1 % compared to the flux in the active core region. Using these data, the neutron spectra of the 2d DORT calculation, which were used in GRSAKTIV and ORIGENX, and the impurities found in the analysis [4], the calculated activities can be listed in the following tables. Expressed in the known ORIGEN spectral indices THERM, RES and FAST the neutron spectra were characterized by , and for the lower position of the squeeze lock nut and for the upper position of the head screw , and They symbolize highly thermalized neutron spectra with a neutron temperature of approximately 300 C as the reactor coolant, in contrast to the situation in the core itself. Table 1 shows a comparison of measurements and calculations for a head screw of the same assembly. It can clearly be seen, that the calculated activations may spread up to a factor 2 compared with the measured values. This can be due to the fact that the analyzed sample was irradiated by local flux effects or due to remaining uncertainties in the impurities which were not included in our calculations. If our calculated values are inside or near at the experimental minimum and maximum results or inside or near at the experimental measurement error, the calculated data are bold printed. An over-prediction of 55 Fe and especially the under-prediction of 58 Co can be remarked. For the last mentioned nuclide a small (n,p)-cross section of 58 Ni is responsible, which would be active only for neutron energies above 3 MeV [3]. In our standard system GRSAKTIV an effective fast cross section of 0.11 (barn) is used, whereby in the new data evaluation (European Activation File 97) after the known PWR flux weighting procedure the resulting GRS-ORIGENX value was 0.07 (barn). When ORIGENX is used as calculation tool, some improvements (+) are seen at 54 Mn and especially at 93m Nb against GRSAKTIV due to the new inelastic scatter reaction channel, what is marked by an arrow in the C/E values (ratio of Calculated/Experimental values). With our standard activation method GRSAKTIV, based on ENDF/B-V, the build-up of 93m Nb from Niobium itself could not yet be simulated. On the other side the Niobium impurities in steel were very small in the order of parts per million (ppm). 3

4 Different methods for analyzing these concentrations showed deviations up to a factor 2 [4]. However, in this validation run a satisfying agreement with 7 of 11 isotopes could be found inside the variation of the measurements. Table 1 Comparison of GRS 3-group-calculated and KWU measured activities (Bq/g of sample steel ) of a single head screw of the upper part of UO 2 assembly BE318 at discharge 58 Co 60 Co 54 M Activities n Half life 71 d 5.3 a 312 d <measured > 8 *E+ *E+ *) *E Sc 55 Fe 59 Fe 51 Cr 59 Ni 84 d 2.7 a 45 d 28 d 8E4 a *E+ *E+ *E+ *E+ *E Ni 100 a 14.1 *E Nb 2E4 a 14.7 *E+1 93m N b 16 a 8.63 *E+3 Minimal Maximal exp. Error 15 5 % 3, ~20 ~20 % % % % % % % % % % % Calculations GRSAKTI V 1 C/E ORIGENX 2.84 (-) (+) (-) (+) C/E *) E+7 read as 10 7 to be also used in the calculated values Table 2 shows the analyzed contents of trace elements in two steel types and In cols. 4 you see 6 additional isotopes activated in the mixture but not specified originally. Trace Element Table 2 Contents of trace elements in steel [4] Steel Content Avg. (ppm) Steel Content Avg. (ppm) 6-C Si 10,000 10, P S Ti Cr 158, , Mn 20,000 20, Fe remainder remainder 27-Co 3,200 3,033 Build-up of important Nuclides 14 C not measured 32 P not measured 35 S not measured 46 Sc 51 Cr 54 Mn 55 Fe, 59 Fe, 54 Mn 60 Co 4

5 28-Ni 118,500 93, Nb Mo 22,500 22, Ni, 63 Ni 94 Nb, 93m Nb 93 Mo, 99 Mo, 99 Tc not measured 2.3. AAA-Calculations of the irradiation of concrete and comparison with experiments Concrete is widely used as radiation shield in nuclear power reactors. During the operation lifetime this material is activated by the neutrons. The content of trace elements has e.g. been investigated from the neutron activation analyses at the Japanese JRR-4 thermal reactor of JAERI [6]. Normal concrete with the analyzed trace elements of [7] has been irradiated for one hour and analyzed after 30 days. The GRS AAA-sequence uses a simplified model of U- Al alloy plates 93 % enriched 235 U, surrounded by water. At the irradiation position we assumed room temperature for the thermal neutron spectrum. We found for the strongly thermalized neutron spectra in terms of ORIGEN spectral indices THERM, RES and FAST the values 0.843, and 0.310, so a 3-group-calculation was started. GRSAKTIV and ORIGENX results of 14 important isotopes were compared with the measured results in table 3: Table 3 Contents of trace elements, measured [6] and 3-group-calculated activities in concrete Trace Element Content Avg. Measured activity Nuclide GRSAKTIV ORIGENX (ppm) (Bq/g) activated (Bq/g) C/E (Bq/g) C/E 20-Ca (83,000 Not 37 Ar 90, ,600 assumed) measured 21-Sc , Sc Cr , Cr 1, , Fe 4, Mn (- ) 26-Fe 7,2 4, Fe 3, , Co Co Zn Zn Sb Sb (- ) 55-Cs Cs (- ) 58-Ce Ce Eu Eu Lu Lu (+) 72-Hf Hf Ta Ta (- ) 90-Th Pa (- ) Z1 Not measured Total 2.47E+ Total 4.91E+ 5

6 The analyzed concentrations and measurements [6] in cols. 2 and 3 are averaged from 180 irradiated samples, important for the definition of typical trace element impurities in NPP bioshields. As can be seen in table 3, most of the isotopes agree within ± 20 % with measured activities. Only for 152 Eu both codes underestimate the activities (C/E = 0.62). However, the concentrations of Europium are very low (0.13 ppm in concrete). Comparing the total activities after 30 days a relatively large difference between the old and the new system can be seen, produced by a lot of other hidden isotopes, due to new data or new reaction channels. In the special case of concrete activation the (not measured) 37 Ar production (T 1/2 of 35 d) from natural 40 Ca due to the (n,α) reaction is responsible for the great difference: the new ORIGENX data (updated from JENDL3.2) offered an (n,α)-cross section by factor 4 greater as the elder one. So a satisfying validation of the AAA-sequence OREST/DORTABLE coupled to the standard activation tool GRSAKTIV and the extended ORIGENX code could be shown for concrete activation. 3. Applications with regard to the operation lifetime of a real NPP and analyses of important radioactive isotopes in the short time range After testing our method in comparison with experimental data for steel and concrete activation, the new GRS AAA-sequence was applied to calculations based on the operation lifetime of a real NPP. The sequence was started with the same (simplified) NPP core model as before: For a full core with the dimensions of a typical German PWR of 3000 MW (thermal energy output) the half-burned inventory was calculated by OREST. For simplification an averaged uniform distribution of the spent fuel nuclide inventories has been assumed. The attenuation of the neutron flux in RZ-directions in such a core with core surrounding steel, water, core basket, water, reactor vessel, isolation and biological concrete shielding layers was simulated by RZ-DORT (PL=3, 83 groups) generating realistic neutron flux data inside and out of a NPP core. Fig. 3 shows the calculated flux distributions for the core zone (the five highest yellow columns), vessel region (five layers labeled RV1-RV5) and the biological shield (Bio1-Bio5). The lower part of the core begins on the left side. The data have been normalized to the active core region. Fluxes from the other regions are suppressed. To handle the non-linear deep-penetration effects of boron absorption, the boron concentrations have been set zero outward of the core, although it would vary between zero and 1000 ppm during each reactor cycle. 6

7 1,00E+01 1,00E+00 core coresur 1,00E-01 water1 corecask 1,00E-02 water2 RV1 rel. Flux Strength 1,00E-03 1,00E-04 RV2 RV3 RV4 RV5 1,00E- air isolat. 1,00E-06 1,00E-07 1,00E-08 Bio1 RV5 core corecask RV2 Radial Regions Bio1 Bio2 Bio3 Bio4 Bio Bio4 Axial Regions FIG. 3 Normalized flux distribution in the RZ-model at core, vessel and biological shield Activation of the NPP vessel with multigroup and multi-region methods The vessel was divided into five layers for which the activation was calculated separately by ORIGENX using a constant flux during 40 years. The variations of the neutron flux fraction compared to the active core and the spectra inside the vessel, calculated by DORT and interface tools, are shown at different geometric points in table 4. All data are flux-volume averaged inside the corresponding layer. Corresponding flux is the averaged core flux. At layer 5 the scattering effects of thermalized neutrons from the adjacent biological concrete shield back to the vessel can be seen in the softening of the epithermal and fast values of RES and FAST and the increasing of the THERM factor due to a lowering of the neutron temperatures. The ORIGEN standard irradiation calculations use LWR type cross section library with only the three group spectral indices (table 4). Due to the resulting hard neutron spectra in layers 3 and 4 strongly different from the reactor core spectra in line 3 a pre-weighting all ORIGEN cross sections by local multi-group fluxes was necessary. Due to the strong variations of flux spectra and flux strength inside the vessel a separate layer activation calculation method was needed considering non-linear build-up effects of the radioisotopes. The transport calculations yield attenuation factors of the initial flux reach in the vessel 1E-04 up to 1E-5. The last three columns show the shielding effect of additional libraries (ANISN-22, EURLIB-78 and EURLIB-99). The agreement within only ± 20 % is very satisfying. In the reactor irradiation history an averaged value of 80 % of full reactor power was used for 39 years and for the last year before shut down the power was increased to 100 %. Evaluation of the results with respect to nuclides of interest, e.g. from the point of view of radiation 7

8 protection and waste management, will be presented in table 5 for the short time range, one month after reactor shut down. The results of table 4, ORIGEN indices cols.2-4, the attenuation factor of the total flux cols. 6 and a total flux of 3.5E+14 at 100 % power can be used directly to calculate the local fast neutron fluency above 1 MeV during reactor life time, listed in table 4 at cols. 5. The trace elements in the steel type <22NiMoCr37>, used as reactor vessel material in German PWR, were taken from table 2, combined with other measured or assumed values [7]. Table 4 Neutron spectral data and attenuation factors for the reactor vessel ORIGEN Indices Fast neutron Flux*Time Flux strength Position THERM RES FAST (n/cm 2 ) PL=3 83 grps. PL=3 13 grps. PL=5 100 grps. PL=5 175 grps. Core ,81E Layer ,42E E- 5.75E- 7.10E- 5.90E- Layer ,44E E- 2.98E- 2.25E- 2.44E- Layer ,86E E- 2.09E- 1.46E- 1.76E- Layer ,10E E- 1.44E- 1.02E- 1.24E- Layer ,94E E E E E- 06 In fig. 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.) will vanish during one 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 impurities in the steel mix. 8

9 1,00E+07 specific activity Bq/g 1,00E+06 1,00E+ 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 decay time FIG. 4 GRSAKTIV calculated activities of important radioisotopes in the reactor vessel after 40 years operation time in units Bq/g up to 100 years cooling time. The right column of table 5 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 %. 9

10 Table 5 Estimated contents of steel 22NiMoCr37 and trace elements [6], important radioisotopes and the 84-group-calculated activities in the reactor vessel one month after shut down Trace Element Content Nuclide GRSAKTIV Standard 40 years operation ORIGENX Extended 40 years operation Difference (S-E)/E, % (ppm) activated (Bq/g) (Bq/g) 03-Li 1.0 **) 3 H 3.22E E C 2, C 1.42E E N 590 **) 14 C included included 08-O 0.1 *) 14 C included included 14-Si 3, P P 1.84E E S S 1.44E E S 33 P 9.46E E Cl 0.04 **) 36 Cl 1.31E E Sc 0.1 *) 46 Sc 2.25E E Ti Sc included included 23-V Cr 5, Cr 1.64E E Mn 10, Mn 5.22E E Fe remainder 54 Mn included included 26-Fe 55 Fe 9.21E+ 9.67E Fe 59 Fe 2.43E E Fe 60 Fe E Co Co 8.27E E Ni 9, Co 8.E E Ni 59 Ni 6.91E E Ni 63 Ni 7.88E E Zn 0.1 *) 65 Zn 3.92E E Nb m Nb 9.58E E Nb 94 Nb 3.51E E Mo 7, Mo 1.68E E Mo 99 Mo 1.08E E Mo 99 Tc 2.24E E Sb 0.1 *) 124 Sb 5.70E+00 3.E Cs 0.1 *) 134 Cs 2.66E E Ce 0.1 *) 141 Ce 1.31E E Eu 0.1 **) 152 Eu 4.24E E Lu 0.1 *) 177 Lu 3.12E E Hf 0.1 *) 181 Hf 1.18E E Ta 0.1 *) 182 Ta 4.47E E Th 0.1 *) 233 Pa 2.10E E U 0.1 **) 90 Sr 2.51E E U 0.1 **) 137 Cs 2.68E E U 0.1 **) 239 Pu 9.16E E

11 Alpha-Activity < 1.E-01 < 1.E-01 Total Activity 1.12E E *) additionally assumed; **) found in [7] 3.2. Activation of the NPP biological shield with multi-region methods The biological shield has been divided into five layers for which the irradiation has been calculated separately by GRS-ORIGENX. The variations of the neutron flux strength, the neutron spectra and the flux strength are listed in table 6. The data depend strongly on the composition of the biological shield. We assumed 3 wt% of crystal water inside the concrete and 8 wt% of steel of the total mixture. The calculated attenuation factor of the initial flux in the vessel reaches 3E-6 up to 1E-7. The last three columns show the shielding effect of additional libraries as in table 6. The agreement for this deep penetration problem to the biological shield is very satisfying with a spreading of ± 25 %. Table 6 Neutron spectral data and attenuation factors for the reactor bio shield ORIGEN Indices Flux strength Position THERM RES FAST PL=3 83 PL=3 13 groups PL=5 100 PL=5 175 groups groups groups Core Layer E E E E-06 Layer E E E E-06 Layer E E-06 1.E E-06 Layer E E E E-07 Layer E E E E-07 Due to the soft neutron spectra the thermal neutron activation reactions are preferred. Activities of the most important nuclides (average of the five layers) will be presented in table 7 for the short term range. The composition of concrete correlates to the measured impurities of table 3, the analogous impurities of steel can be found in table 5. In fig. 5 the selected results of most important isotopes of the GRSAKTIV activities are shown up to 100 years decay time. The right column of table 7 shows the differences between our standard and our extended method. For some isotopes 37 Ar, 54 Mn and 58 Co, the differences can reach more than 40 %. Compared with measurements the calculated 58 Co activities could be underestimated by factors three up to five by both methods. 11

12 1,00E+ specific activity Bq/g 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 99 CS134 EU152 totals 1,00E decay time FIG. 5 GRSAKTIV calculated activities of important radioisotopes in the reactor biological shield after 40 years operation time in units Bq/g versus cooling time up to 100 years. 12

13 Trace Element Table 7 Estimated contents of steel armed concrete and trace elements, important radioisotopes and the 3-group-calculated activities in the biological shield one month after shut down Contents 92 % Concrete Contents 8 % Steel Nuclide GRSAKTIV-I 40 years operation ORIGENX 40 years operation Difference (S-E)/E, % (ppm) (ppm) activated (Bq/g) (Bq/g) 01-H 3, Li 1.0 **) 3 H 3.29E E C C 1.73E E N 590 **) 14 C included included 08-O 380, C included included 13-AL Si remainder 3, P P 1.22E E S S 1.46E E S 33 P 1.00E E Cl 0.04 **) 36 Cl 1.22E- 1.26E Ca 195, Ar 4.86E E Ca 41 Ca 2.17E E Ca 45 Ca 3.99E+03 4.E Sc **) 46 Sc 6.76E E V Cr Cr 1.45E E Mn 10, Mn 5.52E E Fe 10,500 remainder 54 Mn included included 26-Fe 55 Fe 9.47E E Fe 59 Fe 2.27E E Co Co 9.43E E Ni 9, Co 7.55E+00 5.E Ni 59 Ni 6.37E E Ni 63 Ni 7.26E E Cu 2, Zn Zn 3.30E E Nb m Nb 4.69E E Nb 94 Nb 2.73E E Mo 7, Mo 8.22E E Mo 99 Mo 5.09E E Mo 99 Tc 1.06E E Sb Sb 2.90E E Cs Cs 1.82E E Ce Ce 8.08E E Eu Eu 7.77E E Eu Eu 8.53E E Lu Lu 4.18E E Hf Hf 8.10E E Ta Ta 1.21E E Th Pa 2.97E E

14 92-U 0.1 *) *) additionally assumed; **) found in [7] 239 Pu 1.02E E Alpha-Activity < 1.E-01 < 1.E-01 Total Activity 1.56E E+4-15 IAEA-CN Analyses of important radioactive isotopes in the intermediate and long time range Our analyses is 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 Analyses of important radioactive isotopes in the intermediate time range 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 decay time MN 54 FE 59 CO 58 CO 60 NI 59 NB 93M MO 93 EU152 EU154 EU155 totals FIG. 6 GRSAKTIV calculated 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. 14

15 ORIGENX: Reactor Biological Shield Gamma Power (W/g) after 40 years operation time 1,00E-08 specific Gamma Power W/g 1,00E-09 1,00E-10 1,00E-11 1,00E-12 1,00E-13 1,00E-14 1,00E decay time SC 46 MN 54 FE 59 CO 58 CO 60 EU152 totals FIG. 7 GRSAKTIV calculated gamma power of important radioisotopes in the reactor biological shield after 40 years operation time in units Bq/g versus cooling time up to 100 years. For dose calculations outside of the irradiated materials not the information on the activity but on the gamma power is the most important one. Some of the most active isotopes as 3 H, 55 Fe or 63 Ni are weak gamma emitters or non gamma emitters, which produce only beta-particles and bremsstrahlung. Fig. 6 shows the development of the most important gamma power emitters in the irradiated vessel. The most important isotope is 60 Co (from the steel), which at least in the time period up to 100 years dominates over all other radioisotopes. Its gamma power activity lies directly on the curve of the total values. Fig. 7 shows analogously the development of the most important gamma power emitters in the irradiated concrete. The most important isotope is 60 Co (from the arming steel), which again up to 40 years dominates over all other radioisotopes. After this time period the next important isotope is 152 Eu from concrete, where an impurity concentration of 0.13 ppm [6] was used Analyses of important radioactive isotopes in the long term range Table 8 shows the contributors of activity or gamma power of the irradiated vessel material. The most important isotope for the gamma power is 94 Nb (T 1/2 of 20,300 a) with two gammalines at 0.7 and 0.87 MeV. The gamma power of all other isotopes consists of the internal bremsstrahlung or X-rays of the isotopes. Bremsstrahlung from beta particles is omitted in this evaluation. 15

16 Table 8 Important radioisotopes in the reactor vessel in the long time range Activity Bq/g vessel material Radioisotop decay time in years 1,000 2,000 5,000 10, C 1.22E E E E Ni 6.36E E E E Ni 6.80E E E-12 93m Nb 9.47E E E E Nb 3.59E E E E Mo 1.15E E E E Tc 1.57E E E E+00 Total 2.15E E E+02 1.E+02 Gamma power in W/g vessel material 14 C 1.66E E E E Ni 2.66E E E E-14 93m Nb 2.86E E E E Nb 9.04E E E E Mo 2.29E E E E-15 Total 1.43E E E E-14 Table 9 Important radioisotopes in the biological concrete shield in the long time range Most important radioisotopes in the long time range related to Activity Bq/g Decay Time Years 1,000 2,000 5,000 10, C % Ca % Ni % sum % Totals 2.48E E E E+01 Most important radioisotopes in the long time range related to Gamma Power W/g 41 Ca % Ni % Nb % Mo % Tl % Pb % Bi % Ac % sum % Totals 4.81E E E E-15 Table 9 shows the most important isotopes of the biological shield as contributors of activity or gamma power. 16

17 Three isotopes dominate the residual activity from 1000 to 10,000 years, i.e. 14 C, 41 Ca and 59 Ni, summing up to 99 % or more. 41 Ca (T 1/2 of 103,000 a) is generated by activation of Ca contents in the concrete and 59 Ni (T 1/2 of 75,000 a) by irradiation of the nickel containing structural reinforcements. The situation is more complex for gamma emission in concrete. The nuclides 41 Ca, 59 Ni and 93 Mo (T 1/2 of 350,000 a) have an emission of weak internal bremsstrahlung and some X-rays, which can easily be shielded. The contributions of 208 Tl (2.6 MeV gamma rays), 212 Pb, 212 Bi and 228 Ac as decay daughters of natural thorium traces in the concrete (table 7) do not result from neutron activation processes. 4. Summary and conclusions A satisfying validation of the AAA-sequence OREST/DORTABLE coupled to the standard activation tool GRSAKTIV and the extended ORIGENX code could be shown for steel activation. One of the new additional reaction channels could successfully be tested against experiment. For a recalculation of an irradiation experiment of concrete for biological shields most of the isotopes agree within ± 20 % with measured activities. But generally additional new completely documented experiments are needed for future validation. For a typical German PWR more-dimensional deep penetration neutron transport calculations were performed 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 have been done using explicitly the neutron flux strength, spectra and weighted cross sections calculated before. Analyzing of the most important isotopes in the short, the intermediate and the long time range have been done. The results of standard and extended calculation methods have been compared and analyzed. 5. References [1] Hesse, U., Gewehr, K., Moser, E., Hummelsheim, K., Langenbuch, S., Denk, W., Deitenbeck, H., GRS AAA-Code-Collection for Activation, Shielding and Burn-up: GRSAKTIV (GRS-A-2249, 06/1995); GRS-ORIGENX-07, Version 2 (Technical Report 02/2007); KENOREST (GRS-A-2783, 12/1999), OREST (GRS-63, 11/1986, updated 2006); DORTABLE (Technical Report 03/2007) [2] Rhoades, W.A., Childs, R.L., The DORT Two-Dimensional Discrete Ordinates Transport Code, Nuclear Science&Engeneering 99, pp.88-89, May 1988 [3] Fischer, U., Aktivierung von Endstücken eines DWR-Brennelements, Atomwirtschaft, pp.38, January 1987 [4] Martin, J., Bericht KWU R451/84/38, , private communication with U. Hesse, [5] Hesse, U., Die Aktivierung der Strukturmaterialien von DWR-UO 2 - Brennelementen, GRS, Technical Report 02/1987 [6] 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