Fixed delayed gamma source (shutdown reactor)

During steady-state reactor operation, prompt neutrons and prompt gamma rays are produced by fission.

When reactor is in operation, contribution of delayed neutrons and gamma rays can be neglected. In this chapter, the focus is on delayed gamma rays since these are the primary source of ionizing radiation when the reactor is shutdown. It is not possible to distinguish between prompt and delayed gamma rays by measuring gamma radiation but they are differentiated by their origin. Prompt gamma rays are produced during fission whereas delayed gamma rays are produced during radioactive decay of fission products and from surrounding activated materials. These sources also contribute to the gamma dose during reactor operation [86].

Since the half-life of the longest living group of delayed neutron precursors is about 56 s, it can be assumed that there are no neutrons in the core 10 minutes after reactor shutdown. In reality, a neutron

source is usually present. In the case of the JSI TRIGA reactor, the radium beryllium neutron source emits about 10⁶ neutrons every second. Besides that, there are always neutrons emitted during spontaneous fission of transuranic elements. When all four control rods are inserted into the core, the multiplication factor is about 0.93. Therefore, the total neutron population in the subcritical core can be estimated by:

and is in the order of 10⁷ neutrons. Their contribution to the delayed neutron dose compared to the delayed gamma rays can be neglected. To summarize, since the number of neutrons in the core when the reactor is shutdown is negligible, the focus of this work is on delayed gamma rays.

There are two origins of delayed gamma rays: i) fission products inside the fuel elements and ii) activated materials surrounding the fuel such as the fuel cladding, graphite pellets in the fuel and components like the graphite reflector, control rods and irradiation channels. The activity of both depends strongly on the operational history of the reactor, therefore attempting to describe the delayed gamma source is much more challenging than describing the source when the reactor is in operation where the intensity depends only on the power of the reactor.

As this is the first time delayed gamma rays are used for safety analyses of a TRIGA reactor and no methodology to determine such source exists, the delayed gamma source is determined in the following way:

 Only gamma rays emitted from the fuel region (fission products and transuranic elements, activation of zirconium is neglected) are taken into account. In this step, the activation of the surrounding components is neglected.

 The spatial distribution was determined by calculating fission rate distribution inside the fuel meat and assuming that emission rate is proportional to the fission rate distribution.

 Energy distribution of gamma ray was determined by using the so called rigorous two-step (R2S) method.

 Total activity of the fuel in the core was determined by simulating complete history of TRIGA reactor with the Serpent code.

All stages are described in detail below. It is assumed and also demonstrated by experiments (described in section 7.3) that such source adequately describes delayed gamma source for LOWE safety analysis.

Activation of fuel surrounding components is neglected by the fact, that only irradiated fuel is considered as highly radioactive material. In addition calculations and measurements by Ambrožič demonstrated that delayed gamma dose rate in-core is higher than delayed gamma dose rate in the region of graphite reflector [86], [87].

The next assumption relates to the spatial distribution of the delayed gamma source. It is assumed that the delayed gamma distribution is the same as the fission rate distribution during reactor operation.

Therefore, the same distribution is used as for the neutron source (Subchapter 6.1).

The delayed gamma ray energy spectrum was determined by the JSI’s Reactor Physics Department developed rigorous two-step (R2S) method [85]. Firstly, the MCNP code is used to determine neutron transport across the reactor core. Secondly, the FISPACT-II code [88] is used to calculate the neutron activation of the materials and delayed gamma ray generation using the EAF 2010 cross-section library [89]. The delayed gamma spectra are obtained for two time steps, i.e. 1 hour and 20 hours after reactor shutdown (Figure 45). The results reveal only subtle changes in the spectra. In the process of method validation, spectrum for 20 h after reactor shutdown is used. When analysing LOWE, a spectrum for delayed gamma one hour after reactor shutdown is used.

Figure 45: A delayed gamma ray spectra normalized to one of irradiated TRIGA fuel 1h and 20 h after shutdown. The integral is normalised to 1.

MCNP calculated fluxes are normalized per source particle. Hence the total activity of the core/fuel must be calculated in order to obtain absolute dose rates due to the delayed gamma rays. In order to calculate total activity of the core a complete operational history of the JSI TRIGA reactor was modelled in the Serpent code to obtain the activity at a specific time after the reactor shut down [71]. Details are presented in [90]. 300 different isotopes (fission products and transuranic elements) are taken into account and transported from the previous fuel cycle (core configuration) into the next one. ENDF/B-VII.0 [61] nuclear library is used.