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From 2017, Forum and FAQ posts will be hosted together in the Blog category FAQs, see


Pre-2017 Forum and FAQ posts are listed below.


Frequently Asked Questions (FAQs)

What is the Standalone version of Nucleonica?

Some time ago we were contacted by the Federal Office of Radiation Protection in Germany (BfS). They were interested in a standalone version of Nucleonica for the following reason: They needed Nucleonica on a number of notebook computers for their field workers. These persons are active at border controls etc. and cannot be assured of an internet connection in the case of an emergency.

So for the above mentioned reasons we developed a Standalone version of Nucleonica. It is based on virtualisation technology (on the use of VMWare) and on the use of a virtual server. The whole package comes on a DVD. The program is then installed on the notebook computer and it works very well. It is also much faster the the internet version!

Because the version on the DVD is "frozen" it needs regular updates since we are making changes to Nucleonica on a daily basis. So every six months or so an update is required.

Qu.: Databases in Gamma Library++

(Anika Sander)

I created some Gamma Libraries. What happens, if there is a change in nuclide specific data? Is this change automatically integrated in my library? And is there a possibility to get always updated, if there is a change in any nuclide specific data? Are news like this mentioned in the newsletter on the main page?

Ans.:Individual nuclide data is never changed. The user has, however, the possibility to change the underlying database. Currently the databases in Nucleonica are..

- JEFF-3.1 (European database from 2006)

- ENDF/B-VII.1 (American Brookhaven database from 2012)

- or the older 8th Table of Isotopes.

You can select which database to use in the Gamma library++ Options. We recommend ENDF/B-VII.1 since it is the most recent. In the future we will be adding additional databases such as ENSDF.

Regarding getting informed on updates etc.: the latest information is given in the Nucleonica Blog. We recommend users to subscribe to the blog via email (this is explained in blog page). As soon as new information is posted, the user will receive an alert by email.

Qu.: Calibrating a gammaspectrometry system using the X-ray emissions of Ba-133 and Eu-152


I use Ba-133, Eu-152 and Am-241 for the calibration of a gammaspectrometry system. (Since the counting efficiency is quite small, I don't have to care about cascade coincidence summing.) For a calibration at the lower end of the spectrum I would like to use the X-ray emissions of Ba-133 and Eu-152. Nucleonica has the following abundances for the X-ray emissions of the two nuclides: Ba-133:

31 keV 98.6%

35 keV 23.0%


40 keV 59.1%

45 keV 14.8%

Other sources such as nucleide.org have very similar values so I do not doubt that they are correct.

The calibration points of Eu-152 do not fit to those of Ba-133 (of which I don't know either if they are correct), nor do they fit to each other. The counting uncertainty is in the range of a few percent, 10% for the peak at 45 keV. A crosscheck on a laboratory gammaspectrometry system with higher counting efficiency showed the same result. Thank you for any information about the reason for this behaviour!

The calibration is that of a whole-body counter. We use so-called rod sources where the radionuclides are contained in an ion exchange resin, which is sealed in thin plastic tubes. These tubes are distributed throughout a phantom, which is made of polyethylen.

The detector is a p-type HPGe detector with 80% relative efficiency. The front window is made of carbon fibre covered by a plastic cap.

I think self-absorption can be the reason in cases where it gets stronger at higher energies. This is only the case around the absorption edges. But absorption edges occur at these energies of about 40 keV only with quite heavy elements. For example, tin has its last edge only at 30 keV. I can't see any such heavy elements in the setup (except of the lead-shielded walls).

I just simulated the spectrum with Gamma Spectrum Generator Pro++ and calculated the efficiencies. There I could not see that effect.

Qu.: Updating gamma library files

We are looking for a solution that can help update our Genie2000 .nlb gamma library files. We want to use traceable and updated data for our gamma spectrometry. I found a description of a “converter” under the “Gamma Library +++” application, that may transfer a csv text-file into .nlb. We would like to be able to include data (gamma energies, yields and uncertainty) from various sources (DDEP, NUDAT…) that we prepare in a csv text-file. Can this “converter” handle external data, or is it limited to using the Gamma Library output?

Ans.: With the Gamma Library++ application in Nucleonica, you can indeed update your Genie2000 library files.

The Gamma Library++ application uses the data from JEFF3.1 or ENDF/B-VII.1 (energies, emission probabilities, etc.). This data is then stored in an intermediate Genie2000 format. This file can then be downloaded and used with the Genie2000 converter (which we supply separately) to produce the updated Genie2000 data files with the latest international data.

In principle we could extend the Gamma Library ++ to include data not only from JEFF-3.1 or ENDF/B-VII.1 to other data sources (available for example in csv files). But this feature would first need to be implemented.

For further information, please see... http://www.nucleonica.com/wiki/index.php?title=Help%3AGamma_Library%2B%2B#Downloading_libraries

Qu.: Polonium isotopes in waste decay calculations

I have uploaded a nuclide vector (mixture) with about 80 nuclides and then decayed this using the Decay Engine ++ application in 1400 days. The total initial activity was 1E + 12 Bq. The result has surprised me that the daughters of the U/Np/Th - decay series Po211, Po212, Po213, Po214, Po215 and At217 (possibly even more.. ) have an activity of 0 Bq. Why do these nuclides have zero activity –they should be in equilibrium with their parents?

Ans.: The problem arises in the Decay Engine due to the fact that the numbers of atoms is less than 1. If the number of atoms is > 0.5 bet <1 the number is rounded to 1. If it is less than 0.5, the number is rounded to 0! We introduced this since we thought that it is physically not meaningful to have a number of atoms <1.

This is the source of your problem.

By increasing the activity to 1e18, you are increasing all the activities and also the number of atoms. Then you see all the Po isotopes (click on the nuclides header then scroll to 84 (Po)). Note that even here the Po-213 is still 0, hence you will need to increase the total activity again to see Po-213.

Note also that you can see directly the ratio of atoms by checking the "activity ratio" box in the Options. Now in the results grid, if you click on a particular nuclide in the column activity ratio (A/Atotal), all nuclides are normalised to this value. If you click on the total activity at the bottom of the Activity ratio column all values are normalised to the total activity.

Finally, to check the accuracy of the calculation, you should increase the accuracy of the calculation (change default value from 0.001 to 0.0001 or lower). If the ratios do not change then you have the final result.

Another way to do the decay calculations in Nucleonica - an alternative to using the DE++ - is to use the application Decay Engine for Large Nuclide Sets (DELNUS) specially developed for large numbers of nuclides.

Qu.: Pu-241 -> U-237

In Nuclide datasheet++, I can see that U-237 is a decay product of Pu-241 but when I use Decay Engine++ U-237 is not present. Can somebody explain why?

Ans.: The quick answer is just reduce the Accuracy Factor in the Options to 1e-5. You will then see the U-237.

Discussion: From the Nucleonica datasheets++ you can see that the branching ratio for decay to U-237 is quite small i.e. 2.46E-05.

To “see” U-237 in the decay you therefore need to increase the accuracy of the decay calculation. To do this, go to the Options in the Decay Engine++ and change the Accuracy Factor.

The default value for the Accuracy Factor is 0.001. This basically means that if the branching ratio for a particular daughter is less than this value, it will be omitted in the calculation. This is exactly the case with U-237. The branching ratio for U-237 is 2.46E-5 i.e. much smaller than the default 0.001. Hence it is omitted in the calculation.

To “see” the U-237, the Accuracy Factor should be reduced to a value smaller than the branching ratio for U-237. If you change the Accuracy Factor, for example, to 1e-5 and restart the decay calculation, you will now “see” the U-237 in the lists of daughters.

In the Decay Tree tab you can also see all the “chains” included in the calculation depending on the Accuracy Factor set in the Options.

Finally, if you use the Gamma Spectrum Generator (GSG) to generate a spectrum for Am-241 and its daughters, you will see the 208 keV line from U-237 provided you have used the value of 1e-5 for the Accuracy Factor in the GSG settings.

Qu.: In-114m effective dose inhalation coefficient

(Chris Theis, CERN)

I have recently looked at the effective dose inhalation coefficient for In-114m as this is critical for one of our applications. In Nucleonica it is given with 0.0093 (Sv/Bq) which refers to ICRP 72. As the value seemed rather high I've gone back to the source in in ICRP (Table A.2) I find 9.3E-9 Sv/Bq, which is the same value that is also given in ICRP 119 (Appendix G). I'm now wondering which value is correct and if there could have been a transcription error in Nucleonica?

Ans.: Chris, You are right. This is clearly a typo in Nucleonica. Thanks for noticing this and informing us. We have updated the value in the database to 9.3e-9 Sv/Bq. This is the value given in ICRP 119 table G.1.

Qu.: 2.223 MeV ???

(Niko, CERN)

A peak at 2.223 MeV appears in my gamma spectrometry measurements of irradiated water. After a thorough search of radionuclides I don't find a possible candidate. Inside literature, I found that line can be results from neutron capture by hydrogen. Do you know if gamma is issued directly after the absorption of a neutron? Do you have another idea for me?

Ans.: Yes, indeed Niko - this looks very much like the neutron capture reaction in hydrogen 1H(n,γ)2H. The (n,gamma) reaction in hydrogen releases a capture photon at 2.225 MeV. This is a prompt gamma ray release in which the prompt gammas are emitted at the instant of the neutron capture to relieve the excess energy associated with highly excited compound nucleus.

Reply.: That's true, is due to this reaction I've seen appear this peak! It's probably due to the cosmic neutrons who interact in water. We haven't habit to measure this big quantity of water, but when we make a test with the same quanttity of distilled water, this peak also appear :-)

Thanks for your help.

Qu.: Alpha, beta and gamma share in PWR spent fuel (Vincenzo Romanello)

How do I identify the main alpha, beta and gamma emitters (separately) in a PWR fuel after 1 year cooling using the Nucleonica applications? What is the best procedure to follow?

Ans.: Thanks for the interesting question Vincenzo. Here are the steps to follow:

1. Do the calculation with webKORIGEN with 1 y cooling.

2. Download the Excel spreadsheet

3. Delete the unnecessary columns and rows such that you just have two columns: nuclide and activity (or mass). Make sure the first line of the file is Nuclide; Activity(Bq) or Nuclide; Mass(g). Save the file as csv.









4. Use the Nuclide Mixtures and upload this csv file. Then in the Radiations tab of the Nuclide Mixtures you get everything you want. There you will see the alpha and beta activities, and the gamma dose rates listed for each nuclide. In this radiations tab the data can be arranged such that the nuclides with highest activities etc. are listed at the top of the grid.

Qu.: Half- and Tenth- Value Shield Thicknesses Question

I have one question: I am looking for the water thickness which halves the radiation field of Cs-137 (Ba-137m) gamma dose rate. I used the “Dosimetry and Shielding++” tool, and with a Cs-137 source aged 20 years, with a thickness of 100 cm in water, I get:

Half-Value Shield Thickness(cm) 27.25

Tenth-Value Shield Thickness(cm) 57.74

First question: how is it possible that after 27.25 cm I get half of the radiation, while crossing (almost) the same thickness the value is reduced to 1/10th? Shouldn’t it be 1/4th?

According to this reference: https://books.google.it/books?id=a05...0water&f=false, page 516, the half value thickness for Cs-137 is 3.75 in. (i.e. 9.5 cm). Also this link seems conferming this value (here 500 keV photons are considered in various materials, including water):


or this:


Am I missing something?

Ans:. You have overlooked the problem of multiple scattering!

Narrow beams

The normal formula for the intensity reduction in shield materials is…

I = I0 · exp(-µx)

where x is the shield material thickness, µ is the attenuation coefficient, I0 the original intensity and I the intensity after traversing the shield thickness x. It then follows that µx = ln(I0/I)

For the half-value and tenth-value shield thicknesses

x = ln(I0/I)/µ

For Cs 137 (the dominant emission at 661.6 keV (from Ba137m), µ = 8.57e-2 cm-1. Hence

x1/2 = ln2 / 8.57e-2 = 8.09 cm

x1/10 = ln10 / 8.57e-2 = 26.87 cm

These values agree with what you find in the literature. However, it should be noted that they apply for “well collimated beams” or “narrow beams”. Indeed in the reference… https://en.wikibooks.org/wiki/Basic_..._of_Gamma-Rays it states specifically… “Finally it is important to appreciate that our analysis above is only strictly true when we are dealing with narrow radiation beams. Other factors need to be taken into account when broad radiation beams are involved.”

Broad beams

In general, one usually has broad radiation beams. This is shown schematically in the setup in Nucleonica’s Dosimetry& Shielding application as shown in the link... http://www.nucleonica.com/wiki/index...llimated_beams

With broad beams, radiation can be scattered in the detector by so called multiple scattering. This multiple scattering is shown in the above diagram. The formula governing broad beam situations is given by…

I = I0 · B(µx,E) · exp(-µx)

where B is the scattering coefficient and depends on the shield thickness x and the attenuation coefficient µ. In this more general case, it follows

µx = ln(B·I0/I)

For the half-value and tenth-value shield thicknesses,

x = ln(B·I0/I)/µ

For Cs 137 (the dominant emission at 661.6 keV (from Ba137m), µ = 8.57e-2 cm-1. The B values have to be interpolated form tables of data given in Nucleonica. For the half-value thickness B= 5.22, for the tenth-value thickness B = 14.1, hence

x1/2 = ln(10.44) / 8.57e-2 = 27.37 cm

x1/10 = ln(141) / 8.57e-2 = 57.75 cm

Note that the results for the half and tenth value thickness for Cs 137 radiation in water shields is strongly dependent on multiple scattering as the results above show. Stated alternatively, the half and tenth value thickness for water shields depends strongly on whether we have a narrow beam or a broad beam setup.

Qu.: Energy Level Schemes / Reduced Decay Schemes in Nucleonica

Where can I find Energy Level Schemes / Reduced Decay Schemes in Nucleonica?

Ans.: Reduced Decay Schemes are available in Nucleonica in the application Reduced Decay Schemes. Users need to be registered to have access. More information on this applicaiton is giev in the Nucleonica wiki at... http://www.nucleonica.com/wiki/index.php?title=Help%3AReduced_Decay_Schemes

This list is updated on a regular basis. Reduced Decay Schemes for approximately 50 nuclides can be found in the latest edition of the Karlsruhe Nuclide Chart which can be ordered from our online shop… https://shop.marktdienste.de/shoppages/produktuebersicht.aspx

Qu:. Schülerfrage zu Atomen und Elemente Klasse 9B

im Chemie-Unterricht haben wir den Atomaufbau besprochen und haben eine Frage an Sie. Frage: Kann die Anzahl der Neutronen eines Atoms kleiner sein als die Anzal der Protonen in einem Kern? Und wenn, warum stoßen sich die Protonen nicht wieder ab?

(Ans. Nucleonica Team) Natürlich gibt es solche Nuklide, die mehr Protonen als Neutronen im Kern aufweisen.

Ein Blick auf die Karlsruher Nuklidkarte zeigt etliche solche Nuklide: sie befinden sich oberhalb der ersten Winkelhalbierenden. Abgesehen von Wasserstoff H1 (1 Proton, 0 Neutron) gibt es sogar ein stabiles Nuklid darunter, nämlich He3 (2 Protonen, 1 Neutron).

Der Grund dafür ist, dass bei leichten Nukleonen die nuklearen Kräfte, die den Kern zusammenhalten, stärker als die elektrischen Abstoßungskräfte sind. Erst bei höheren Atomzahlen überwiegen die elektrischen Abstoßungskräfte, so dass die Nuklidenverteilung sich allmählich unterhalb der ersten Winkelhalbierenden verschiebt.

Qu.: Blanket salt neutron capture

(Thomas Jam Pedersen)

I am trying to make simulations on molten salt reactors. My question is if I can use one of the Nucleonica applications to calculate the following: Given a blanket salt of thickness W = 50 cm, consisting of 78%LiF-22%ThF4 salt. Assuming I have a given neutron flux on one side of the blanket. Can I calculate the neutron flux that will occurs on the other side of the blanket. And as a result of the above also calculate the number of neutrons absorbed in the blanket salt.

Ans.: Unfortunately the answer to your question is no – or at least not directly. We do not have the possibility to do neutron shielding calculations directly at the moment.

One application that might be of help is webKORIGEN. There is a mode in webKORIGEN (Reactor irradiation, flux mode) which allows you to do neutron activation calculations. This basically allows you to irradiate a sample in different types of neutron flux (thermal, PWR, BWR, FR).

The first step is to create a nuclide mixture corresponding to the 78%LiF-22%ThF4 salt (this is relatively easy to do by adding elements in the nuclide mixture application – see also the wiki Help on this). In the nuclide mixture application you then need to specify a mass to be activated. This could correspond to 1m x 1m x 50cm thickness of the salt.

Then in the flux mode you can do an activation calculation (by specifying the neutron flux, the neutron spectrum (thermal, PWR, BWR, FR) and the time of the irradiation.

In the results grid you can basically see all the activation products. With this information you can at least get a qualitative idea of how much interaction took place between the neutron flux and the salt: if few activation products are produced this implies little interaction; if many activation products are product this implies a strong interaction. Maybe one then then work “backwards” to obtain the number of neutrons absorbed.

Qu.: Database for Gamma Library++

Why is it not possible to select ENDF as database within Gamma Library++?

Ans.: As you say, it is not possible to select the ENDF database within the Gamma Library++. It is currently only possible to select either the JEFF-3.1 or 8th Table of Isotopes databases. It is planned to include the ENDF in a future Nucleonica update. Is there any particular reason why you would like access to ENDF in the Gamma Library++ application?

Reply: In the JEFF database the emmision probabilities and their uncertainties for Ac228 seem to be wrong, whereas the ENDF data appear to be correct.

Qu.: Isotopes of fermium

I have found a site: http://myonverse.wikia.com/wiki/Fermium. What is interesting, they describe an isotope of fermium (Z=100), Fm-260, with an extremely long half-time of 6E+5 years! Another one, Fm-262, is also long-living, with T(1/2) of 116 days. Both cannot be found in other sources. Moreover, Fm-259 and -258, elsewhere reported to be short-living (370 mcs and 1.5 sec., respectively), have also half-times in order of several days. Is it true? Is there anybody, who could share the newest and most recent news about fermium? How is it possible to synthesize fermium nuclides heavier than 260?

Ans.: The currently experimentally observed heaviest Fm isotope is Fm-259. It was first observed in 1980. It is an SF (spontaneously fissioning) isotope with a half-life of 1.5 s. http://dx.doi.org/10.1103/PhysRevC.21.966

Based on systematics, alpha and ß- decays are also possible for Fm-259. Model calculations have been used to estimate partial half-lives (note that partial half-lives are not the same as the actual half-lives). The table you mention presents only the alpha decay partial half-lives for Fm-259 and also for Fm-258. The real observed half-lives are much more shorter because of the dominant SF. For the nuclides Fm-260 and Fm-262 it was not possible to find any publications. One can suppose that in the table presented values are based also on some models, but no reference is given there on the actual calculations.


Fm-259 was produced through Fm-257(t,p).

There was some information from 1982 on the reaction "O-18 + Cm-248" which could lead to Fm isotopes - but no detailed literature was found.

Qu.: Multiplication of the isotopes lists

(Niko, CERN)

Why e-ship can't used the data from an existing "nuclides mixture" and we must create a new "package"? It will be more efficient if we can use the same "Library" in all aplications of Nucleonica, you don't think?

Ans.: You can indeed use an existing mixture in e-Ship:

• From the My Packages tab, click create a new package

• At the bottom of the edit form select a mixture from the drop down list

• Click Add Mixture

• Set nuclide’s physiological properties if needed

Add the specific data needed for a package so you can save and work with the package.

Qu.: Alpha or not alpha...

(Niko, CERN)

Why is Gd-149 identified like an alpha emitter in e-SHIP++ ? His alpha decay is < 1E-3, and his decays products have an half life > 10 days. So, reading the § I.54. ET I.58. of AIEA SAFETY GUIDE No. TS-G-1.1 (ST-2) Gd-149 not corespond to the definition alpha emitter.

Ans.:Thank you for the post and this very good question.

Table lists the A1 and A2 values for “unknown” nuclides, in particular alpha emitters.

The definition of an alpha emitter is given in § I.59 of the IAEA SAFETY GUIDE No. S-G-1.1, but this definition does not mention any restriction about the half-life of the nuclide or of its parent…

I.59. A radionuclide is defined as an alpha emitter if in greater than 10–3 of its decays it emits alpha particles or it decays to an alpha emitter. For example, Np-235, which decays by alpha emission in 1.4 × 10–5 of its decays, is not an alpha emitter for the purpose of the special forms consideration. Similarly Pb-212 is an alpha emitter since its daughter Bi-212 undergoes alpha decay.

So according to the above § I.59, Gd-149 should be considered as an alpha emitter. However, the situation is somewhat more complicated than that stated in § I.59 alone. In fact one has to consider the daughter nuclides and their half-lives. In particular, one has to read §I.55 and §I.56 of the Consideration of parent and progeny radionuclides of the above IAEA SAFETY GUIDE i.e.:

I.55. The earlier Q system assumed a maximum transport time of 50 d, and thus radioactive decay products with half-lives less than 10 d were assumed to be in equilibrium with their longer lived parents. In such cases the Q values were calculated for the parent and its progeny, and the limiting value was used in determining A1 and A2 of the parent. In cases where a daughter radionuclide has a half-life either greater than 10 d or greater than that of the parent nuclide, such progeny, with the parent, were considered to be a mixture.

I.56. The 10 d half-life criterion is retained. Progeny radionuclides products with half-lives less than 10 d are assumed to be in secular equilibrium with the longer lived parent; however, the daughter’s contribution to each Q value is summed with that of the parent. This provides a means of accounting for progeny with branching fractions less than one; for example, Ba-137m is produced in 0.946 of the decays of its parent Cs-137. If the parent’s half-life is less than 10 d and the daughter’s half-life is greater than 10 d then the mixture rule is to be used by the consignor. For example, a package containing Ca-47 (4.53 d) has been evaluated with its Sc-47 (3.351 d) daughter in transient equilibrium with the parent. A package containing Ge-77 (11.3 h) will be evaluated by the consignor as a mixture of Ge-77 and its daughter As-77 (38.8 h).

In the meantime the 10d criterion has been taken into account and the database is updated accordingly which means:

A daughter nuclide is not considered as an alpha emitter when • The half-life of this nuclide is greater than or equal to 10 days or • The half-life of this nuclide is greater than the half-life of its parent

As a result Gd-149 is no longer an alpha emitter but is classified as a beta/gamma emitter.

Qu.: Possible errors in the KARLSRUHER NUKLIDKARTE Ausg. 7, 2006?


In some other publications, I have read that the cross-section for production of Lu-177m by neutron irradiation of Lu-176 is higher than for production of Lu-177 (groundstate). In the "Lu-176" cell, there is a value "2+2100", in the legend, the sequence is following: "metastable and ground state" - i.e., it shows that thousand times more of Lu-177 than Lu-177m should be produced by neutron irradiation of Lu-176. Is this correct?

Ans.: We believe the values given in the Karlsruhe Nuclide Chart 2006 are correct. Here are some details... The original experimental paper in Nuclear Instruments and Methods in Physics Research A 521 (2004) 5–11 http://www.sciencedirect.com/science/article/pii/S0168900203030614 in the paragraph 2.2.1 reports the following:

As the production of 177mLu strongly depends on the cross-section of the 176Lu(n; g)177mLu reaction, we performed a test measurement of this reaction using a natural Lu sample at the T4 neutron facility at ILL. The obtained value of neutron capture cross-section is equal to 2.8070:17 b [6] for a flux of thermal neutrons which confirms the usual adopted value2.870:7 b [7].

We used as scientific source the summary report: N.E.Holden, Neutron Scattering and Absorption Properties (Revised 2003), Handbook of Chemistry and Physics on CD-ROM, Version 2006, 11-185 Where is reported (2+2100) barn for Lu-176 with the following comment:

Parentheses with two or more numbers indicate values to the excited state(s) and to the ground state of the product nucleus.

Unless more recent data is available than the above mentioned report, we believe the values quoted in the KNC to be correct.

Qu.: Nuclide Datasheets++ - Differences in metastable nuclides Nubase 2012 versus others

The Nuclide Datasheets++ is a very practical application to compare world-wide used evaluated nuclear data. In some metastable nuclides I have found big differences in half-life and spin between Nubase 2012 and other data files (e. g. ENDF/B-VII.1. )

For example:


Half-life, spin in Nubase: 380 ns 1/2+,

others: 4.69 m, 11/2-



660 ns, 2- versus 249 day, 6+


Is this a mistake of evaluators?

Ans.:Isomer Notation in different Databases: The problem of comparing nuclide data from different databases such as ENDF/B-VII.1, JEFF3.1 or NuBase2012, is the identification of the isomeric states of the nuclides. Depending on the criteria applied by the evaluators to consider an excited state as metastable or not, the number of metastable states of a given ground state nuclide may differ. It follows that the name of the isomers (m, n, etc.) may differ between databases.

As an example, Pd-109m with a half-life 380 ns is considered as metastable in Nubase2012. It follows that the next metastable state with a half-life of 4.696 m is called Pd-109n.

In the JEFF and ENDF databases metastable states with Isomeric Transition (IT) have a half-life longer than 1 µs. It follows that the first considered metastable state is the 4.69 m half-life denoted by Pd-109m.

In the example above, one should compare Pd-109m from ENDF/B-VII.1 or JEFF3.1 to Pd-109n from NuBase2012 which are the same nuclide. This problem of inconsistency of notation between ENDF/B-VII.1, JEFF3.1 and NuBase2012 will be addressed in a future Nucleonica release.

Qu.: Nuclide Search / Radiation Search, question on ß- emitters

In the new Nuclide Search / Radiation Search application, I obtain 658 ß- nuclides in the JEFF3.1 database. However, in the Nuclide Explorer I find 1308 nuclides with ß- emission as their main decay mode. What is the reason for this difference?

Ans.:Isomer Notation in different Databases: The problem of comparing nuclide data from different databases such as ENDF/B-VII.1, JEFF3.1 or NuBase2012, is the identification of the isomeric states of the nuclides. Depending on the criteria applied by the evaluators to consider an excited state as metastable or not, the number of metastable states of a given ground state nuclide may differ. It follows that the name of the isomers (m, n, etc.) may differ between databases.

As an example, Pd-109m with a half-life 380 ns is considered as metastable in Nubase2012. It follows that the next metastable state with a half-life of 4.696 m is called Pd-109n.

In the JEFF and ENDF databases metastable states with Isomeric Transition (IT) have a half-life longer than 1 µs. It follows that the first considered metastable state is the 4.69 m half-life denoted by Pd-109m.

In the example above, one should compare Pd-109m from ENDF/B-VII.1 or JEFF3.1 to Pd-109n from NuBase2012 which are the same nuclide. This problem of inconsistency of notation between ENDF/B-VII.1, JEFF3.1 and NuBase2012 will be addressed in a future Nucleonica release.

Qu.: JEFF3.1 vs 8thTORI database: question about gamma lines

(Francesco Damati)

I would need some explanations about the reporting of gamma lines in the Nucleonica databases. I am using the JEFF3.1 as a reference to revise and update some Genie2K libraries. This database does not always match the gamma lines with the nuclide to which they are commonly referred to. For example, the 661.657 keV line of Cs-137 is displayed in the Ba-137m keV datasheet. This makes sense, as Ba-137m is a daughter product of Cs-137 decay and it is the nuclide from which the gamma ray is effectively emitted. Is there any basic criterion according to which some lines are matched to the "parent" nuclide and others to the "daughter" nuclide?

Also, I have seen that the 8th Table Of Radioactive Isotopes reports some (few) lines that JEFF3.1 reports in the daughter product's datasheet (like the above mentioned example) or does not reports at all: as an example of the latter, the 63.83 keV for Th-232, that I have not found in the JEFF3.1 database (nor in the Th-232 nor in the Ra-228 datasheets).

The JEFF3.1 is a recent database, but it seems that some (very few) lines are missing and that some of them are stored in the daughter product (so asking the user to correct the intensity for the decay's branching ratio). On the other hand, the 8th TORI is a more dated database, but it seems to be easier to browse when looking for gamma lines from a given isotope.

According to your experience, which database is more reliable and should be used as a reference for checking Genie2K libraries? JEFF3.1, 8th TORI or both of them?

Ans.: In Nucleonica there are two main databases: JEFF3.1 (European, 2007) and ENDF/B-VII.1 (American, 2011). There is also the 8th TORI (table of radioactive Isotopes, 1997) for comparison purposes. The 8th TORI is older and less reliable than JEFF3.1 and ENDF/B-VII.1

One feature of the TORI is that the gamma lines of short-lived daughters are grouped together with the parent nuclide. An example of this is the 661 keV line of Ba-137m. This can be found in the TORI under the parent Cs-137m. This “grouping” of lines in TORI is a source of much confusion – it is not done in JEFF3.1 and ENDF/B-VII.1. In JEFF3.1 and ENDF/B-VII.1, the 661 keV line is not listed under Cs-137 since it is associated with the daughter Ba-137m decay. So when using JEFF3.1 and ENDF/B-VII.1, one has to be aware that gamma lines of short-lived daughters are not listed under the parent radiations.

To obtain the “effective” emission probability of the 661 keV Ba-137m line, consider the following: from the Nucleonica Datasheets++, the branching ratio of Cs-137 to Ba-137m is 0.944. Now the emission probability of Ba-137m for the 661.657 is 0.9007. So to get the "effective" branching ratio of the 661.657 line from Cs you have to multiply both these numbers together i.e. 0.944*0.9004= 0.8499.

This means physically that for every 100 disintegrations of Cs-137, you will get 85 lines at 661.657 keV.

Finally we recommend using both the JEFF3.1 and ENDF/B-VII.1 databases as given in Nucleonica. There one can compare the gamma emissions on a line by line basis. You only have to be aware, however, that the gamma lines of short-lived daughters are NOT listed together with the parent emissions.

Qu.:Decay engine: no negative times?

Is there a reason why the decay engine can only handle positive decay times? I occasionally have to calculate what activity a source had some time before it was recalibrated. Of course it's easy to do by hand, but it doesn't seem like it would be hard to add to the engine. Obviously it would not calculate progenitor activities, but just the prior activity of a source.

Ans.: Thanks for the interesting question. The Decay Engine++ can only handle positive times. If a negative time is used, an error message will appear “The decay time should be a positive number”.

However, it is very easy to investigate “negative” times by using the rescale feature. Consider the following example: a Gd-148 source has an activity now of 1.73 kBq. What was the activity 23 months ago?

1. In the Decay Engine++ select the nuclide Gd-148. Use the default activity of 1e6 Bq and set the decay time to 23 months (if larger decay times are used the requried results can be obtained using the slider control in the Graph). The calculation results show that the activity drops from 1e6 Bq to 9.82e5 Bq after 23 months.

2. Use the rescale tool, to rescale the end activity from 9.82e5 Bq to the activity now 1.73e5 Bq.

3. Rescale by pressing the Start button. Both in the table and in the graph, it can be seen that the activity 23 months ago was 1.76e3 Bq.

It should be noted that this procedure only works for the parent nuclide. For more details see our wiki page on "negative" decay times

Qu.: half-life uncertainty (Nucleonica nuclides datasheets)

(Francesco Damati)

I am using Nucleonica data to compile some libraries for Genie2K and I would need an explanation about the notation of the half-life's uncertainty reported by the nuclide datasheets. As far as I have understood, I should apply the uncertainty starting from to the last digit of the half-life, so that, for example, the half-life of Na-22, which is reported as:

2.6027 (+10)y

may range between 2.6017y and 2.6037y.

Or, if we consider Nb-94:

20.0 (± 24) ky

this would mean that its half-life ranges between 20024y and 19976y.

Is my interpretation correct?

Ans.: for Na-22 your interpretation is exactly correct. The half-life 2.6027 (±10)y should be interpreted as ranging between 2.6017y and 2.6037y.

However for Nb-94 it is not correct. Since the uncertainty concerns the last two digits, the half-life 20.0 (± 24) ky ranges between 22.4 ky and 17.6 ky. These are the values from JEFF3.1. The US datafile ENDF/BVII.1 gives slightly different values of 20.3 (± 16) ky which ranges from 21.9 ky to 18.7 ky.

Both JEFF3.1 and ENDF/BVII.1 values are given in the Nuclide Datasheets++. You can see these by activating the Compare databases in the Options tab. There you will also see the value given by Nubase 2012 as 20.3 ky 1.6. The latter notation may be somewhat clearer.

Qu.: Import of own nuclide mixture failed


I tried to upload a csv-file into nucleonica for creating an own nuclide mixture. But it did not work. A warning pops up: Data at the root level is invalid. Line 1, position 1. The file cannot be processed. In the nucleonica help I found that currently only the mass could be accepted as input. Is that right? Beside mass I need to work with the activity.

Ans.: We have been working on this problem for some time now. In the latest Nucleonica deployment it is now possilbe to upload/import csv-files into the Nuclide Mixtures application.

How do I find the polonium "Factfile" in Nucleonica?

(Qu.) I am trying to find information on poloniun-210 in Nucleonica. In particular I would like to find the po-210 "Factfile" - how do I find this information?

(Ans.) Just go to the wiki main page and enter polonium into the search engine - then press "search". The search engine will then return a list of links on the subject of polonium.

Alternativly, just click on the link below... http://www.nucleonica.com/wiki/index.php/ReadingRoom:Polonium-210

Many other interesting article can be found in the wiki ReadingRoom. This can be accessed from the wiki page by click on "Main" and then "ReadingRoom".

Qu.: Genie2000 Library Conversion Program gives error "comdlg32.ocx missing"


I tried to work with the Genie2000 library conversion program to convert a Nucleonica created nuclide library to use it with our Canberra G2k software. Installation worked fine but when I tried to start it, I got the “Component ‘comdlg32.ocx’ missing” error (see attached images). After some google research I copied a comdlg32.ocx from microsoft.com to my system and registered it. But after that I got a run-time error. Since my first thought was that there might be a problem with me running Win7 64-bit I tried to run it on a WinXP 32-bit machine with exactly the same errors. Any idea on that problem?

Ans.(Spectro): Do you know that once you create your G2K library on Nucleonica portal, you need to use a conversion program to convert it (provided by Nucleonica Team on request) ? Then, after this conversion, you need to edit the Library first before using it (otherwise you might have problems). Did you follow these steps ?

WARNING: you need to use a computer where G2K is installed to be able to run the conversion program

Do you have any error when opening the library in Library editor ?

Reply: thank you very much for the fast reply. I got the mentioned errors by starting the external conversion program I got from the Nucleonica Team on request. So using that tool itself is/was the problem:

WARNING: you need to use a computer where G2K is installed to be able to run the conversion program

That was the info I was missing! I tried to use the conversion program on my office PC where no G2k is installed (but where I use the Nucleonica applications most of the time). After your post I installed the library conversion tool on the gamma lab PC down in the basement and voilà, the coversion program works without any errors. Thank you very much for your help again.

Some note for other new users of the conversion program: My first converted library had wrong values for energies and probabilities (wrong magnitude) because of my regional setting for the decimal separator ( , instead of . ). So either change the regional setting on the PC used or (as I did) replace all "." by "," in the created nuclide library downloaded from Nucleonica before using the conversion program on it.

Qu.: Whats the difference between the Decay Engine++ and the Decay Engine applications?

Recently I've seen there is a new Decay Engine++ application. What's the difference between this and the previous Decay Engine?

Ans.: We recommend that you use the Decay Engine++ application. This is the latest version of the application and has been optimized for speed and flexibility. For a list of improvements in the Decay Engine++ see our blog posts:




The name of the application has been changed to Decay Engine++ to distinguish it from the previous version. We continue to support the previous version - Decay Engine – for users who have not yet switched over to the new version.

Qu.: webKORIGEN - Initial enrichment of U-234

(C. Lehman) I was calculating the spectrum of a PWR with webKORIGEN. Here I noticed that the calculated value for U-234 is very low. Can it be that is not considered initial enrichment for U-234 when calculating? Is there a reason? Usualy in the enrichment of U-235, U-234 is getting enriched too.

Ans.: We have looked into the problem you raise. Indeed you are correct - thanks for pointing this out. The U-234 is not included in the initial enrichment. You can see this clearly by doing a calculation and then plotting the uranium nuclides U-238, U-235, U-234. We are correcting this right at the moment but it may take some time before this is implemented. We will keep you posted via the forum.

Ans.: After a long delay, we have now included the initial enrichment of U-234 in uranium calculations. To estimate the amount of U-234 in the fresh fuel, the following analysis has been made:

Denoting the masses of uranium isotopes by m4 (mass U-234), m5 (mass U-235), and m8 (mass U-238), the enrichment x5 of U-235 in a uranium sample is defined as:

x5 = m5 / mU = m5 / (m4 + m5 + m8) (1)

In the enriched uranium, we assume the ratio of masses m4/m5 is the same as in natural U i.e.

m4/m5 = 7.46e-3 (2)

Relation (2) is now used in webKORIGEN to determine the inital mass of U-234. Combining relations (1) and (2) leads to expressions for m4 and m5 from a starting mass of m8. Consider m8 = 100 g and the U-235 enrichment x5 = 0.04 (4%). It follows that m5 = 4.168 g, and m4 = 3.109e-2 g.

Rescaling (using the nuclide mixture application) to a total mass of 100 g gives:

m8 = 95.97 g; m5 = 4 g; m4 = 0.02984 g.

Qu.: Gamma Spectrum Generator: Dead (inactive) layer thickness and low energy peaks

Using a standard HPGe (coaxial, p-type, rel. eff. 50%) I could not see the low energy peaks under 20 keV for Co-57. If I switch to a lower efficiency standard detector (BEGe, 30mm x 50 cm2, rel. eff. 30%) I do see the lower energy peaks. Why does a lower efficiency detector have a higher sensitivity in the low energy range?

Ans.: The problem is that HPGe 50% is a coaxial type detector with the 0.5 mm dead layer (this can be seen in the settings) that absorbs nearly all the photons below 40 keV. The one with the 30% relative efficiency is a so-called broad-energy HPGe with only 0.0003 mm dead layer. With such detector one may detect photons down to several keV.

One should be aware of the fact that the relative efficiency is defined for the 1.33 MeV photons and, thus, it represents the detector sensitivity (efficiency) for high-energy photons. Here, the dimensions of the detector sensitive volume (diameter and length) are the parameters influencing the detection efficiency.

For the low energy photons (e.g. < 100 keV), which are normally absorbed in a thin front layer of a detector crystal, the dead-layer thickness, endcap material and thickness, as well as crystal area are much more relevant parameteres for describing the detector efficiency.

Qu.: Gamma Spectrum Generator Pro

I already put some effort in simulating one of our detectors using the Gamma Spectrum Generator Pro. Unfortunately there are always big discrepancies on the low energy side of the peaks increasing towards lower energies. Maybe you can give me some idea what I am doing wrong. First some data on the detector. It is a HPGe detector in a big lead housing:

Even with measurement times of 3600 s no peaks can be detected in the background spectrum. All data given next are derived either by measurement or by the producer protocols. In my Nucleonica account the detector is named “Det 30 – Probe 30 cm 002”.


type Germanium n-type

length in mm 30.1

diameter in mm 60.2

relative efficiency in % 25


length in mm unkown, I use 20 mm, but effect on result is negligible

diameter in mm unkown, I use 10 mm, but effect on result is negligible

entrance window

material Berylium

thickness in mm 0.5

vacuum layer

thickness in mm 3

inactive layer

material Germanium

thickness in mm 0.0003


Number of spectrum channels 8192

Channel-to-energy factor 0.3249

Energy resolution

at 122 keV 0.982

at 1332 keV 2.07

Background count rates [cps]

Peak at 185,7 keV 0.00778

Peak at 238.6 keV 0.01639

Peak at 511 keV 0.01472

Peak at 609.3 keV 0.00889

Peak at 661.6 keV 0.00167

Peak at 1332.5 keV 0.00111

Peak at 1460.8 keV 0.00167

continuum (0 - 3 MeV) 3.6075

Playing in the options menu with the settings for Bremsstrahlung (factor 5) and external scattering effects (factor 1.75) the best fits to measured data were achieved by these factors, but the discrepancy is still too large (to my opinion). Attached you will find some results of these playing for Co-57. Click image for larger version

I would greatly appreciate if somebody can give me an advice how I can improve the simulations (for the attached example specially in the range 50 keV to 100 keV).

Ans. (aberlizov): Note that both GSG and GSG Pro treat the source as a point, which means no gamma-ray scattering inside the source is taken into account. In your case (as seen from the photograph) the source is voluminous, so that photons can be scattered inside the source prior entering the detector sensitive volume. Furthermore, the Compton profiles used in the GSG modeling were generated using a generic detector model shown in Fig. 3.6 in the GSG manual. This model does not include lots of bulky material that surrounds your detector - radiation shielding, sample holder etc. These scatter the source photons back to the detector, thus, forming a very intense backscatter (plus multiple scatter) continuum in the 50 keV - 110 keV energy interval in the measured spectrum. The intense contribution of the detector shielding to the spectrum is clearly recognizable also on the Pb KX-rays at about 75 keV, which do not appear in the simulated but are present in the measured spectrum. So, basically, there is no remedy for the discrepancy in your particular case, as this is related to features in the actual measurement geometry that are not considered in the GSG modelling. You could have obtained a much better agreement, if the detector were positioned outside shielding. Using a point-like Co-57 source would have improved the agreement even further. Although, I understand, this is not the geometry intended for your low background measurements...

Reply: For the measuring result shown, a Co-57 point source was used. I was aware of the features you described, but hoped doing something wrong. Anyway, Nucleonica (and the GSG) is a great tool helping a lot in daily business.

Qu.: Sn-115 metastable state or not?

(Niko, CERN)

Why in some nuclear tables i found an isométric transition and gamma emission for the tin-115 and here not ?

Ans.: The nuclide Sn-115 (stable) has very many excited states. These states can be excited or activated by a variety of physical processes involving for example protons, alphas etc. The entire energy levels diagrams can be found for example in the Nuclear Data Sheets.

However in Nucleonica and the Karlsruhe Nuclide Chart, we are primarily interested in levels which can be populated only by radioactive decay from parent nuclides to excited states of the daughter. These are not all the levels of Sn-115 but only a subset.

In the case of the daughter Sn-115, for example, the parents are Sb-115 or In-115 and In-115m. In the radioactive decay of the parent Sb-115 to Sn-115, not all excited states of Sn-115 will be populated. The levels which are populated are determined from spin considerations (for the transition from the ground state of the parent to the excited stated of the daughter).

In Nucleonica's datasheets you will find all gamma transitions from the parent Sb-115 to the daughter Sn-115. There are in fact 38 such transitions listed. The highest emission probability is at 497 keV.

In the case on the Karlsruhe Nuclide Chart, metastable states which do not undergo alpha, or beta decay, or spontaneous fission, i.e. decay only by gamma emission, are included only if their half-life is larger than 1 s. This is a general rule for nuclides in the Karlsruhe Nuclide Chart. Hence you do not see any metastable states for Sn-115. If this rule were not implemented there would be many thousands of additional lines to be shown on the Chart.

In summary only radioactive decay data is given in Nucleonica - not the full nuclear reaction data.

Qu.: The Karlsruhe Nuclide Chart : paper version VS computer version

(Niko, CERN)

Why in the new Karlsruhe Nuclide Chart (8th Edition 2012) some nuclides are stable (like Os-184, Hg-196...) but not in the Nuclide Explorer?

Ans.: The short answer to your question is that the Karlsruhe Nuclide Chart shows only experimentally observed nuclide data.

Various attempts have been made to show that both Os-184 and Hg-196 are radioactive. There have also been some calculations made on these nuclides. The current state of knowledge is that only lower limits for half-lives can be given (see Nuclear Data Sheets). In the case of Os-184, the half-life quoted in Nucleonica is a lower limit for alpha decay. For Hg-196, the half-life quoted is a lower limit for double beta emission. Note that the Nucleonica database is based on the JEFF3.1 OECD NEA radioactive decay datafile. The evaluators of this file therefore expect that both Os-184 and Hg-196 are radioactive. However as stated above, this has not yet been verified experimentally. This is the reason why this information does not appear in the Karlsruhe Nuclide Chart but does appear in the Nucleonica database.

Qu.: DECAY Tool - need clarifications...


We have played a little bit with the DECAY Tool. We used Mo-99 (2.75 d) --> Tc99 (6.01 h). We introduced A1=1 MBq (Mo-99) and a decay of 50 days. After 50 days, the activity of A2 (Tc-99m) is 3.22 Bq while A1 (Mo-99) is 3.32 Bq Which means that A2 < A1.

To our opinion, there is a problem, since A2 (Tc-99m) should be higher than A1 (Mo-99). We done the calculation manually, and A2 = A1 x 1.10 (A2 > A1)

How could we explain this difference ?

Ans.: The reason for the apparent discrepancy lies is the branching ratios. In your calculation you have assumed the branching ratio for the decay from Mo99 to Tc99m is 1. In fact it is smaller. For the decay of Mo99, the branching ratios are:

Mo99 -> Tc99m (branching ratio is 0.881 - see Nucleonica datasheets)

Mo99 - >Tc99 (branching ratio is 0.119)

If you take this into account, you will see that the calculations in the Decay Engine are correct.

Qu.: Complex nuclide mixtures creation

(Jeremie Muswema)

Given these characteristics for fuel elements in a Triga MK II Research Reactor: 8.5% wt 20% enriched U-235, mixed with 91.5%wt ZrH1.6; how can I create the mixture to use as input in running KORIGEN?

Ans.: Here is how to create the mixture you asked for: fuel elements : 8.5% wt 20% enriched U-235, mixed with 91.5%wt ZrH1.6

Step 1: The element Zr consists of 5 stable (or almost) isotopes with the following abundancies

Zr-90 = 51.45

Zr-91 = 11.22

Zr-92 = 17.15

Zr.94 = 17.38

Zr-96 = 2.80

so in this first step we create, using the nuclide mixtures application, 100 atoms of Zr with 51.45 atoms of Zr-90, 11.22 atoms of Zr-91, etc. We now have 100 atoms of Zr. To this we have to add 160 atoms of H-1 (we neglect H-2 deuterium). So add 100 atoms of H-1. We now have a mixture of ZrH1.6 with 100 atoms Zr and 160 atoms H. Now rescale the 260 atoms to 91.5 g. To do this click on Total in the Nuclide column. Now change the 260 atoms to 91.5 g.and update. Save the mixture.

Step 2: Create a new mixture containing 100g of 20% enriched uranium using

U-238 = 80 g

U-235 = 20 g

(neglect U-234)

Now rescale this result to 8.5 g total mass using the method described above. Save the mixture. Make a note of the masses of U-235 (1.7g) and U-238 (6.8g) Step 3: Open the previously created mixture of ZrH and add the masses of U-235 and U-238. You will then see that the total mass is 100 g.

It sounds somewhat complicated but is actually quite straightforward. This mixture is now available in the nuclide mixture tab "Sample Mixtures". You can select the mixtures from the list and then click on Send to my Mixtures.

Ans.: Zircaloy-4

Using the procedures described above, even more complicated mixtures can be created. An interesting example is the creation of the mixture Zircaloy-4 which is a cladding material used in power reactors. The irradiation of this material can then be simulated using webKORIGEN in the flux mode to obtain activation products. The decay of these activation products can be further investigated using the module Decay Engine for Large Nuclide Sets.

Qu.: Abundance and uranium enrichment...


I have a question which drives me crazy: Is there any difference between abundance number (for U-235) and uranium enrichment or depleted uranium number. To my opinion, there is one difference, and the following apply:

  • Abundance = atomic fraction
  • Enrichment = weight fraction


X5 = N5 / (N4+N5+N8)

X = abundance (atomic fraction)

N = number of atoms

If we say that we have 0.72% U5 in nature, then:

For 100 atoms of theelements U

No. of atoms U238 = 99.2742

No. of atoms U235 = 0.7204

No. of atoms U234 = 0.0054

ENRICHMENT (or depleted uranium as well):

x5 = m5/m8 or x5 = m5/(m5+m8) ???

Can someone confirm, or give additionnal explanations?

Ans.: enrichment definitely refers to mass. The first formula (x5=m5/m8) is not correct, the second one is nearly correct: the exact one is x5=m5/mU often simplified as m5/(m5+m8). Abundance is the number reported in the Karlsruher Nuklidkarte and is the atom percent.

Qu.: Core Inventory in Bq?

(Jeremie Muswema) How can I obtain outputs of core inventory in Bq? If I do an activation or irradiation calculation in webKORIGEN (Mode 1: Flux), I get only the masses or moles in the output grid. I need this information in terms of activities

Ans.: In webKORIGEN Mode 1 (power or flux mode), the output shows only the masses or number of moles. To obtain for example, the activities or radiotoxicities, you need to use additionally the application "Decay Engine for Large Nuclide Sets". With this application you can always obtain the most recent webKORIGEN calculation.

In detail: if you are in the webKORIGEN application, click on the item "New Browser" in the taskbar. This opens a new tab in your browser. In this new tab, select the application "Decay Engine for Large Nuclide Sets". If you then click on the tab "Sample Nuclide Sets", the recent webKORIGEN results are contained in the item "Nuclide Set". Just select this item and then press the buttom Load Sample. Your recent webKORIGEN results should now be visible in the main tab of the application.

Reply: Nuclide mixture irradiation

Thanks. It works perfectly, as I wanted.

1/However, when running KORIGEN at step 2 (Reactor/Operation) nowhere is it specified U235+U238 for nuclide mixture irradiation, but only HEU. It it the presumed specified mixture? What about a LEU in this case then? Assuming one has a 4% LEU, for nuclide mixture irradiation, what option to choose?

2/ In step 3 (Input summary and run), the cross section library gives 2 options: one for PWR UOX (4% U235) and one for PWR UOX high burnup (ORNL). Question: in a Research Reactor, LEU or HEU respectively, which option is the best, and why?

Reply: reply to point 1.: You have to create the mixture U235 + U238 using the Nuclide Mixtures application with the appropriate composition. This mixture is then available in the drop down list of mixtures in webKORIGEN Step 2.

Reply: reply to point 2: (spectrum in a research reactor)

The neutron spectrum in a nuclear facility is determined by the spectrum of the neutron source (thermal or fast fission, spallation), by the moderator (light or heavy water, sodium, helium), by the size of the facility (large leakage in small ones), and by the fuel.

The spectrum changes during irradiation due to formation of new fissile material (plutonium in case of UOX fuel), and due to formation of neutron poisons (Xe and other neutron absorbing fission products). In PWRs the formation of Pu240 with its 1eV capture resonance plays an important role.

For recommending a cross-section set, spectrum governing details of the considered facility have to be known. Information on the following is required: formation of special nuclides, fuel properties after discharge etc ? For a reliable irradiation calculation, a spectral calculation with subsequent cross-section averaging for leading nuclides is the first step. By comparison of the obtained cross-sections with those of the available libraries can give a hint on which libraries should be used.

A simpler way would be just to use the different cross-section libraries (PWR, BWR, FR) and look at the results, i.e. making a sensitivity analysis with respect to the data. The spread of the results will give a rough estimation of the calculation’s reliability.

Qu.: Core Inventory calculations

My target is a HEU Low Power Tank-in-pool closed core Reactor (MNSR), 30 kW thermal power, with neutron flux of 1.2×10^12 s-1.cm-2. Cooling system being done by natural conversion; critical mass: 0.998 kg; 10.2 cm thickness of Be reflector and 1 Cd control rod. Assumption is made that the reactor had operated on continuous basis at its 1/ full power, 2/ half power, to achieve burnup of a given percentage, how or which component of Nucleonica maybe helpful for me to compute: - Activity inventories of important radionuclides in the reactor core;

- release fractions; and

- activity released (to the atmosphere)?

I see also that webKORIGEN interest is more focused on NPP (PWR, BWR or EFR); what about Research reactors (more than 1 MW), such as LEU TRIGA which are of my second interest?

Ans.: Basically your problem is the irradiation or activation of the highly enriched uranium in a neutron flux. This can be simulated in Nucleonica as follows:

1. The module you should use is webKORIGEN. In this module there are various modes of operation. One of these modes is Irradiation (or Flux mode). With a Premium account you can access this mode. For further information on this mode see our brochure pages 28-30... http://www.nucleonica.com/wiki/images/1/12/NucleonicaBrochure_2010.pdf

2. In the irradiation mode you need to specify the neutron flux, spectrum and the material to be irradiated and the irradiation time:

2a. Neutron flux: this is easy - just enter 1.2e12 in the appropriate box.

2b. Neutron spectrum: here you can select a pure thermal spectrum or a spectrum characteristic of for example a PWR (essentially thermal with a fast tail).

2c. Material to be irradiated: Since you mention HEU, then a first choice would be just to select U-235. If you want to add some U-238 then you need to create a nuclide mixture (using the Nuclide Mixture module) where any composition can be specified.

2d. Irradiation time: Here you can enter any time e.g. 1y

3. Once everything has been defined as above, then you can run the calculation. In the output grid, you can then plot the quantities of interest (e.g. fission products). In a worst case scenario, these activities would be fully released into the environment. To some extent the problem is similar to the Fukushima reactor accident. We have already done a case study on this which may be of interest to you.

Finally, from the above, if you understand the module properly you can simulate not just PWRs, BWRs and EFRs but almost any nuclear reactor.

Qu.: Functionality of the "Gamma Library"

(Niko, CERN)

Congratulations for your site who contain a lot of data, and applications very useful! I have just a question about the "Gamma Library": Would it be possible to add some functionality? It will be really nice if we can print the "Radiation Lybrary" and export to a file. "txt" or ". Xxs".

Ans.: You can print the radiation library. To do this, go to the Summary tab. In the grid, you will see your gamma library (provided you have already saved it). In the column Download you will see two download icons (i.e. disk icons). The left-most icon allows you to download the library in GammaVision format. This is a binary format and not directly suitable for viewing. The second download icon allows you to download the library in "Identify" format. This is a readable text file and the .lib file can be viewed with a simple text editor like Notepad. On opening with Notepad you will see all the energies and emission probabilities as shown in the gamma library application. You can then save this as a .txt file. Hope this helps.

Qu.: webKORIGEN output in EXCEL format

Is there any possibility to transfer the complete output of a webKORIGEN calculation into a spreadsheet?

Ans.: In many of the Nucleonica applications, one has the possibility to download the data as an Excel spreadsheet. This feature is, however, not yet available for webKORIGEN output.

The main reason for this is due to the powerful "filter" features available in webKORIGEN. When you do a calculation, you will see the 3-column output grid. If you, for example, select Actinides and then click on Filter, a list of all actinides is shown in the central grid column. From this central grid column you can select specific Elements and again click on Filter. In the first column you obtain then all the nuclides belonging to these selected elements. In this first column you can then select the nuclides you are interest in. These can then be plotted using the plot buttons below the gird. This is a very powerful way of plotting the output data in webKORIGEN.

Now, of course, there are situations where indeed one would like to have all the output data in an Excel spreadsheet. So, following your remark, we will introduce this feature in the next deployment. This will result in an Excel spreadsheet with a list of all nuclides, masses, number of atoms, activities.

Qu.: Why does HEU produce fewer counts than pure U235 in the Gamma Spectrum Generator?

Most of the activity of HEU is U235. 10 kg is about 8E+8 Bq. The spectrum for 8E8 Bq of HEU gives 6.7E4 counts/s from U235 (other contributions are less), but the spectrum for 10 kg of pure U235 gives 1.7E6 counts/s. The latter number is about right, using the size of the default 1" X 2" NaI detector and the decay rate. Why does the HEU produce so many fewer counts?

Ans.: I assume that you are using a pre-defined HEU mixture containing 90.1 wt.% U-235 and 0.77 wt% U-234. The total activity of the mixture is mostly (approx 95.6%) defined by the activity of U-234, while U-235 activity contributes only about 4%. Therefore, if one takes HEU and pure U-235 samples of the same activity, the activities of U-235 in these samples will relate as 1/25. The U-235 gamma-ray count rates from these samples should therefore also relate as 1/25, which is exactly the ratio reported above: 6.7E4/1.7E6=4e-2=1/25.

Qu.: Decay engine with Pu-241

U-237 is a daughternuclide of Pu-241 and it has a branching ratio of 2.46E-5. But this daughternuclide has no appearance in the decay engine of Pu-241 or I don't find it. Is there something missing or what should i do?

Ans.: The Decay Engine in Nucleonica is based on the Bateman solution for radiactive decay. This is a mathematically exact solution to the differential equations so in principle all decay products (no matter how small the branching ratio) should be calculated. For the decay of Pu-241, the standard settings in the Decay Engine for the accuracy factor is 1E-2. This means that daughter products with a branching ratio BR less than 1E-2 are ignored in the calculation. For most cases this is satisfactory. But in your case, you are interested in U-237 which has a BR of 2.46E-5. Since this BR is less than the default 1E-2 it is not included in the calculation. To see the U-237 daughter you must enter an accuracy factor less than the BR e.g. 1E-5. You will then see the daughter U-237. In addition, if you set the accuracy factor to 0, then all daughters are included in the calculation (the results then show 23 linear chains). For more information see our wiki article on this at http://www.nucleonica.com/wiki/index.php?title=Help%3ADecay_Engine%2B%2B

Qu.: How to use decay engine for spontaneous fission of Cf-252?

I need to determine how long it will take to reach an equilibrium amount of Xe135 resulting from SF of CF252 to Xe135 and SF of Cf252 to I135 with subsequent decay to Xe135. I want to check my analytical solution. I cannot find an option for this in decay engine.

Ans.: Just use the Decay Engine++. You can increase the accuracy by changing the Accuracy Factor in the Options from the default 0.001 to lower values. Then you will see Xe 135 in the list of nuclides.

Qu.: Am-242m Gamma Ray Dose in a MA mixture unexpected!

(Vincenzo Romanello)

Why if I evaluate the gamma dose with the following composition (expressed in %):

Am241 31.46

Am242M 2.95

Am243 20.06

Np237 33.60

Cm242 0.01

Cm243 0.13

Cm244 9.36

Cm245 2.42

Looking at gamma rays dose I get a contribution from Am-242m of ca. 47%??? It appears a little bit strange, as Am-242m emits only few low energy gammas (and in fact its corresponding isotopic power is low). Should it be a bug?

I performed the decay calculation on a 100 grams composition, after 1 second (all other numbers were left as default). I am trying to evaluate it in order to assess the radiological risk in different keypoints of a fuel cycle (fabrication and reprocessing plants, disposal, etc.).

Ans.: First of all you can do a Dosimetry & Shielding calculation for the single nuclide Am242m with a mass of 2.95g. You will see that this gives a dose rate at 1m of 2.24e4 µSv/h (as you correctlty say this is almost half the total gamma dose rate for the whole mixture).

When you look at the dose contributions from each individual gamma ray, you see that the most important contribution come from the Am242m line at 15.2 keV (X-ray).

Now, if you look at the Options (in the D&S module), you will see that the threshold energy used in the calculation is 15 keV. The reason for this is that photons with energies below 15 keV are normally absorbed in the outer layers of the skin. It just happens that Am232m has an X-ray at just above this value i.e. 15.2 keV. This makes a large contribution to the dose rate if counted in the calculation. Further details on the Threshold Energy are given in the Nucleonica wiki.

Now because it is very close to 15 keV, the Am242m photon at 15.2 keV will also be absorbed in the outer layers of the skin and so practically can be ignored in the calculation.

So you can do the following: 1) increase the threshold to 16 keV or 2) switch off the X-rays in the Options. If you do this you will see that the dose rate for 2.95 g Am242m decreases from 2.24e4 to 57.6 µSv/hr!

In summary, the Nucleonica calculations are correct - there is no bug. However, in all calculations, you must be aware that the threshold energy is 15 keV. If you have a nuclide with gamma or X-ray photon energy near to this value - this should be eliminated from the calculation (since it is absorbed in the outer layers of the skin).

Qu.: What does "rel.eff. xx.x%" really mean for a HPGe configuration?

(Mikhail Morev)

I've got a difficulty to understand the reason for descripancy between GSG simulations and experimental measurements of HPGe full-peak efficiency. The HPGe detector is Ortec, model GMX-35195-P-S, rel.eff. 35%, geometry parameters:

Crystal diameter - 59 mm

Cristal length - 72.4 mm

Contact diameter - 9.2 mm

Contact length - 64 mm

EC to crystal gap - 3 mm

EC window - 0.5 mm, Beryllium

Outer contact (dead) layer - 0.3 mkm, Boron

I put the parameters into GSG and get a spectrum. The title in the output spectrum says "Gamma-Spectrum Simulated for HPGe (rel.eff. 53.6%". This value (53.6%) is much bigger than 35% (certified by manufacturer). The experimentally measured efficiency for Co-60 is some 20-30 % less than calculated with GSG. The discrepancy is explainable by the difference in "rel.eff." which is also some 25%. The trouble is that I do not know what is the value (53.6%) really stands for.


1) How does GSG estimate this "rel.eff." based on detector geometry? I have read the manual - nothing.

2) Does GSG use this "rel.eff." in any way when simulating efficiencies, or just reports it for convenience of the user?

Ans.(A. Berlizov): Here are the answers to your questions:

1) in every simulation run the GSG (in addition to a user’s task) calculates the Full Energy Peak (FEP) efficiency for a monoenergetic 1.33 MeV point-like gamma source located at the 25 cm distance from a detector end-cap. This efficiency is then normalized to 0.0012, which is generally accepted as the FEP efficiency for a “standard” 3´´ x 3´´ NaI detector and 1.33 MeV photons emitted by a point-like Co-60 source at the 25 cm distance. The obtained ratio is then multiplied by 100% to yield an estimate for the relative efficiency, which is finally reported by the GSG in the spectrum title.

2) as it follows from the previous explanation, the relative efficiency itself is not used anyhow in the calculations. However, it does reflect the quality of the simulation approach as well as the appropriateness of the detector model parameters used. Therefore, primarily for the convenience of users, this simple “quality indicator” was added to the spectrum output.

Now about the discrepancy between the declared relative efficiency and that that is evaluated by the GSG for your GMX detector. The discrepancy is really huge. To find out a reason for this, I performed an independent calculation using the MCNP code and your detector parameters. This gave me a value of 49.1%, which is close to the GSG result (53.6%) and quite far away from the relative efficiency declared in your detector certificate (35%). If we take the MCNP efficiency as a reference, then there are two other questions that need to be answered:

Question #1: Why is the GSG value higher than the MCNP result by 4.5%?

This may be explained by a deficiency in the modelling approach used in the GSG. The FEP efficiencies are calculated in GSG by multiplying the total efficiency value (TOT), which is very accurately calculated, by the FEP/TOT efficiency ratio, which is interpolated for given detector dimensions from database values. The database values were obtained for a detector without a rear contact (see GSG manual) and this is the reason for the discrepancy. You may see it from the efficiency values below, which were calculated using the MCNP for your detector:

TOT = 0.0034 (MCNP with contact)

FEP/TOT = 0.1732 (MCNP with contact)

Rel. Eff. = 49.1% (MCNP with contact)

TOT= 0.0034 (MCNP without contact)

FEP/TOT = 0.1894 (MCNP without contact)

Rel. Eff. = 53.7% (MCNP without contact)

It is seen that the presence of a contact decreases the FEP/TOT ratio, compared to a no-contact-detector. It however does not change noticeably the total efficiency value (as the contact volume constitutes only a few percent of the total crystal volume). As a result, the calculated relative efficiency decreases when the crystal contact is taken into account. Since the GSG’s FEP/TOT ratios do not take into account the presence of a contact, the GSG’s relative efficiency is therefore somewhat higher. Note, that the MCNP relative efficiency for a no-contact-crystal (53.7%) agrees perfectly with the value reported by the GSG (53.6%).

Conclusion: The relative efficiency for a coaxial HPGe detector may be somewhat overestimated by the GSG as a result of the presence of a rear contact. It is because the FEP/TOT ratios used in the GSG were calculated and tabulated for a detector model, which did not include a contact at the rear side of a Ge crystal. Note, that it is not only the relative efficiency can be affected by the presence of a contact but the whole simulated spectrum as well.

Question #2: Why is the MCNP result significantly higher than the declared relative efficiency for the GMX detector?

A possible reason may be that the crystal dimensions provided in detector certificates (normally) does tell about the external dimensions of a crystal but (very often) does not tell about the dimensions of a sensitive (active) volume of a crystal. This is exactly the reason why a special detector characterization procedure has to be usually performed, if one wants to establish an accurate model for his detector.

This is exactly what CANBERRA does for its detectors, which are to be used with the ISOCS/LabSOCS application. Can you imaging – they do not trust THEIR OWN detector specifications and do perform their tuning in order to establish a validated detector model that would reproduce a set of reference experimental measurements.

Now concerning ORTEC. I have just visited their web-site and found a GMX detector in the detector stocklist (at http://www.ortec-online.com/Solution...stocklist.aspx) with almost the same crystal diameter (59.4 mm) as in your case, much shorter (by 16.5 mm) crystal length (55.9 mm) and (WHAT A BIG SURPRISE) higher relative efficiency (37%). I will not give any further comments on this but just draw your attention to papers written by ORTEC guys, where you can find more information and examples concerning possible discrepancies between the declared geometrical dimensions of Ge crystals and actual dimensions of their active volume:

R.M. Keyser, W.K. Hensley, Efficiency of germanium detectors as a function of energy and incident geometry: Comparison of measurements and calculations, Journal of Radioanalytical and Nuclear Chemistry, 264, No.1 (2005) 259-264.

R.M. Keyser, W.K. Hensley, Efficiency and Resolution of Germanium Detectors as a Function of Energy and Incident Geometry, http://www.ortec-online.com/download...4-b492edc90b5c

Concerning your case, my opinion is that the crystal length of your detector is either incorrect or does not correspond to the actual length of the active volume inside the crystal.

Qu.: How is Mo99 produced?

In nuclear medicine, the nuclide technetium-99m, Tc99m, is often mentioned in connection with in-vivo gamma irradiation. The parent nuclide of Tc99m is Mo99 which is produced from the thermal fission of U235. The fission yield of Mo99 is around 6%. I would like to know the particular reaction path responsible for the production of Mo99, i.e. U-235 + n --> U-236 --> Mo-99 + ? + ? + ?n

Ans.: Lets start by assuming Mo99 is produced directly as one of the two fission products. Typically the number of neutrons produced in the fission process is small, either 2 or 3. Let's assume 3, then, in order to conserve charge and mass, the reaction must be

U-235 + n --> U-236 --> Mo-99 + Sn-134 + 3n

If the number of neutron were only 2, then instead of Sn134, Sn135 would be produced. However if you look at the individual yield of Mo99 from the thermal fission of U235, it is only 1.8e-3% (this can be obtained in the Nucleonica Fission Yield module). This implies that diect fission to Mo99 is not very probable. The fact that the cumulative fission yield is around 6% gives a hint that the Mo99 is being produced by other fission reactions followed by decay.

A full analysis of this problem has been given in the Nucleonica Ask an Expert page. It turns out that the main reaction paths are...

U-235 + n --> U-236 --> Zr-99 + Te-134 + 3n


U-235 + n --> U-236 --> Y-99 + I-134 + 3n

The Zr99 and Y99 then decay by beta emission to Mo99.

Qu.: Range & Stopping Power of fission product ions in materials

Can the Range & Stopping Power module be used to calculate the ranges of fission product ions in materials?

Ans.: The underlying calculation engine in the Range and Stopping Power (R&SP) module for ions, SRIM, assumes that all ions are in an equilibrium charge state which depends on the ion, its speed, and the target material. Fission fragments, however, are fully stripped of their electrons. For this reason, the R&SP module will not give a very accurate value of the range and stopping power.

Since, however, highly ionised ions will "attract" more electrons in the medium than less ionised ions, they will experience a greater slowing down effect. For this reason, one can expect the R&SP with "equilibrium" ions to give a larger range. This can then be regarded as an upper limit of the range for the fission fragment ions.

For more information, see the Nucleonica wiki... http://www.nucleonica.com/wiki/index.php?title=Help%3ARange_%26_Stopping_Power%2B%2B

Qu.: Fission fragments versus fission products: what is the difference?

In Nucleonica, webKorigen computes the quantities of fission products, while the Fission Product yields module displays and compares yields from different fission parents and various libraries. But sometimes we use the term Fission Fragments: what is the difference, if any?

Ans.: The daughter nuclei showing up right at scission of a fissioning mother nucleus are called primary fission “fragments”. In the large majority of the cases, the fragments will be sufficiently excited to evaporate neutrons in times less than 1E−15s. This means that the nuclei detected in experiments are not the primary fragments, but instead secondary fragments having lost a varying number of neutrons. The secondary fragments are called the fission products (i.e. fission fragments after prompt neutron evaporation).

Qu.: Why do electrons and positrons not have the same stopping power?

The interaction between charged particles and matter depends on the absolute value of the electrical charge of the incoming particle, thus one should expect that the stopping power and the range of a particle and its antiparticle in matter are equal. However, with the module Range & Stopping power of Nucleonica, it appears that a positron for example is slightly more decelerated than an electron. What is the reason of this small difference?

Ans.: The Barkas Effect: According to Bethe, the stopping power of a fast charged particle penetrating through matter is proportional only to the square of its charge. In 1953 Smith and colleagues observed a small difference between ranges of positive and negative pions in emulsion. This difference between the range and stopping power of a particle and its anti-particle is called the Barkas effect and is attributed to the polarisation of the medium caused by the particle or anti-particle. Qualitatively, a positively charged particle will attract electrons and experience a greater resistance to motion. For this reason positively charged particles have a smaller range than the equivalent negative anti-particle.

More information http://www.nucleonica.net/wiki/index.php/Barkas_Effect

Qu.: webKORIGEN: Single Group Cross Sections?

The neutron spectrum in a reactor depends on the composition of the fuel. Since the fuel composition changes with time, so too must the neutron spectra. But in the webKORIGEN calculations, single group cross sections are used - independent of fuel composition and time. Can you explain what is the origin of these cross sections?

Ans.: For important heavy metal nuclides: Th232, Pa233, U232,233,234,235,236,238, Np237,239, Pu238,239,240,241,242, Am241, Am242,242m, Am243, Cm242,244, spectrum-dependant single-group capture and fission x-sections are used. All other heavy metal nuclides and ca. 70 important fission products are provided with constant x-sections, the mid-irradiation-time spectrum taken for x-section averaging. The spectra depend on the reactor type (PWR, BWR, Fast Reactor), fuel (UOX, MOX) and initial enrichment. For the remaining nuclides, x-sections are determined from thermal x-sections, resonance integrals and a fast-neutron x-section, combined by spectral indices typcal for PWR and BWR UOX and MOX fuels.

Qu.: Does an alpha particle have excited states?

When a radionuclide undergoes radioactive decay, the resulting daughter nuclides are generally produced in excited states which then decay to the ground states. What happens if the “daughter” nucleus is an alpha particle? Can this alpha particle be produced in an excited state following radioactive decay?

Ans.: The neutron separation energy for He-4 is more than 20 MeV. This is the energy required to remove a neutron from a He-4 nucleus and can be obtained from the differences in the mass of He-4 compare to the masses of (He-3 + a neutron). Not quite sure if intermediate excited states can be produced - but probably not in the radioactive decay process.

Ans.: Good question indeed. It can be excited with some nuclear reactions e.g (e,e') (p,P') , but all excited levels are above the particle separation threshold. So.. once excited, the resonances decay preferably via p or n channels. Also, 4 He systems can be formed as short lived (excited) resonances of lighter systems. So .. that aint really an alpha particle... as such... This in a way, though being an excited state, is different to what one would think of as excited (non-resonance) states in heavy nuclei... Its down to interpretation. .. in simple terms, I would be inclined to say that an alpha particle as such cannot be excited, however there are resonances that have the 4He composition. Surely these resonances wont be found in the aftermath of an alpha decay. So the nuclear core of an alpha partice emerging from a alpha decay wont be excited.

I'm using Microsoft Excel. Why are the greek characters missing in the CSV file I downloaded?

The charset of the CSV file was not set correctly by Microsoft Excel. Do not open the CSV file directly but use the import functionality and choose Unicode as charset.

Qu.: How the backscatter portion of the spectrum is calculated

(Trvor Balint)

I was wondering if you could explain exactly how the backscatter portion of the spectrum is calculated when the normalization factor is nonzero? And what exactly the backscatter norm factor represents? My lab is trying to cover all our bases before we fully engage in using this, and this info would really help in looking at the GSG.

Ans.: The backscatter photon contribution (the so-called "backscatter peak") is modelled based on the interpolation of pre-calculated backscatter photon spectra, which were simulated using a Monte-Carlo method implemented in a dedicated simulation code DRGen. The modelling approach used is described in detail in the GSG Model Description document, available in the Nucleonica Wiki at http://www.nucleonica.com/wiki/index.php?title=Help%3AGamma_Spectrum_Generator%2B%2B. The backscatter spectrum represents a contribution of photons scattered on the detector elements, which are included in the respective NaI and HPGe detector models (see Fig.3.6 in the Model Description). These elements include a detector cap, a crystal holder, a reflector etc. The backscatter scaling factor on the "Options" tab allows to improve the agreement with the experimental spectra, which normally exhibit an increased contribution from the backscattered photons in comparison with the pre-calculated backscatter spectra. This is due to the presence of additional (scattering) objects in the detector surroundings (walls, floor, table, source holder etc.), which were not taken into account in the detector models shown in Fig.3.6. It is however important to note that this scaling is intended for fine tuning, i.e. in cases when discrepancy between the simulation and experiment is not too large. Normally, the scaling factor should not exceed 3. If it turns out that higher values are required, this must indicate a significant difference between the modelled and actual measurement geometry. In this case a coupled EMC-GSG simulation is a more suitable approach, This simulation scheme includes a Monte Carlo simulation of the physical photon spectrum at the first step (using an Easy Monte Carlo tool), and then the convolution of the physical spectrum with a detector response matrix on the second step (using the GSG tool). This simulation scheme is currently under testing and, hopefully, it will be available for the Nucleonica users in the near future.

Qu.: Some questions on the GSG

When I select the detector LEGe (which I assume refers to a low efficiency germanium dectector, and then start the calculation, the results heading reads: "Gamma-Spectrum simulted for HPGE (rel. eff. 8%)". The reason for this is the although a low efficiency detector has been seleced, it is nevertheless consists of high purity Ge (HPGe). (Note: BEin BEGe refers to "broad energy").

Ans.(A.Berlizov): You are right: it is still a High Purity (or Hyper Pure) Germanium detector (i.e. a detector made of HPGe) disregarding the actual efficiency value. The LEGe normally means the Low Energy (not Low Efficiency !) Germanium detector, i.e. a germanium detector for low-energy photons and X-rays. The useful energy range of such detectors is < 200-300 keV, where they offer good detection efficiency and superior energy resolution.

Qu.: how to make available the additional settings for the GSG detector?

how to make available the additional settings for the detector? They appear unactive for me so far...

Ans.: You must select the <...Edit...> option in "Current Configuration". Also tick the "Show More Settings" box at the bottom right hand corner of the screen. Then make all the changes you want, then click "Save As" which is to the right of "Current Configuration".

Qu.: How can I change my profile photo?

When I try to change my profile photo I receive this message

" Error We are very sorry for the inconvenience caused to you...Internal server error The server encountered an unexpected condition that prevented it from fulfilling the request." !

Can the photo be changed ?

Thank you for helping me !

Ans.: I have tried to change my profile picture and it does not seem to be a general error. Did you make sure the picture you attempted to upload was less than 2 mb ? You should normally be able to change the your profile picture, have you tried another time? If it is still not possible to change your picture please let me know and I will investigate the matter further.

Qu.: webKORIGEN questions

(D. Freis)

I have two questions:

1. From time to time we have to calculate inventories for high BU fuel up to 100 GWd/t. Are there existing effective cross sections for these high BUs? And if yes, will they be implemented into webKORIGEN in the future?

2. I need to determine the inventory of HTR fuel (pebbles) as good as possible. Several codes offer this possibility but with some effort. Are there any effective cross sections for high temperature reactors available for webKORIGEN? For the light water reactors webKORIGEN offers preformated irradiation histories with load factors and cycles. It should be possible to do that for HTRs as well (roughly). In general, is it planed to include more types of reactors in webKORIGEN?

Ans. (A. Schwenk-Ferrero): in reference to your questions concerning high burn-up data you posed on July the 5 th following remarks:

1. The KORIGEN code is able to tackle thermal and fast, critical and subcritical systems. It must be supplied however with a suitable set of problem dependent input : facility data and associated nuclear data libraries.

Without any doubt the effective cross-sections for fuel irradiated in thermal spectrum with the high burn-up of 100 GWd/tHM exist. However they have not yet been generated by KORIGEN team in FzK. WebKORIGEN libraries which can be used by NUCLEONICA community address UOX and MOX fuel loaded to German thermal critical reactors (PWR, BWR). The burn-up is limited to 60GWd/tHM, simply because no power plant was licensed untill now for fuel with the higher burn-up.

Facility data like: Fuel composition, irradiation time, thermal power history etc. can be introduced to KORIGEN easily once they are known in detail from the design and/or the operation of power plants. Effective cross-sections for fuels irradiated in the core of the facility under consideration must be generated separately, employing coupled spectral and burn up procedures. The reason of this burn-up dependent coupling is the strong impact of some generated fission products (highly absorbing isotopes) on the neutron spectrum in a thermal core . Generation of a new cross-section library requires numerical evaluation of energy dependent data at first and then an extensive validation (benchmarking) to irradiation experiments or against the results of other codes performing fuel evolution assessments. If you have well benchmarked data available together with the core characteristics, plant operational parameters and fuel management schemes on hand we could consider implementing new facilities and fuels into KORIGEN or webKORIGEN in the mid-term.

2. No, we have only fast critical EFR facility with a set of effective cross-sections implemented at present. The upgrade of KORIGEN and webKORIGEN is however envisaged in mid- and long term .

Qu.: atw Press Release

Summary: Die siebte, aktualisierte Auflage der „Karlsruher Nuklidkarte“ ist erschienen. Die vollständig überarbeitete Ausgabe zeigt die bekannten Atomkerne aller Elemente und deren radioaktive Daten. Die Nuklidkarte wurde seit 1958 vom Forschungszentrum Karlsruhe und seinen Vorläuferinstitutionen herausgegeben. Die Gemeinsame Forschungsstelle der Europäischen Kommission, das Institut für Transurane in Karlsruhe, setzt diese Tradition mit der vorliegenden und weiteren Ausgaben der Nuklidkarte zusammen mit dem Forschungszentrum fort.

The full text can be seen at... http://www.nucleonica.com/docs/atw-2007-02-publications.pdf

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