Summary of measurement/verification techniques

There is a well established range of spent fuel measurement techniques available to support BUC verification. A useful set of these, along with application examples is presented in a burnup credit training course that was run in Beijing [5].

The principal types of these non-destructive measurement techniques may be summarized as:

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Gamma ray counting

o

Gross gamma ray counting:

This technique may be used to provide a quick verification that an assembly has been irradiated. With information on cooling time a crude measure of irradiation can be achieved.

o

Gamma spectrometry:

This powerful technique measures the intensity of emissions from individual fission and activation products and may be used to measure burnup, cooling time and, to a limited extent, enrichment under certain conditions. The technique may be applied with high, medium and low resolution detectors and their associated pulse processing electronic systems. The choice of detector type will depend on a number of parameters linked to the desired measurement performance, i.e. accuracy and precision, as well as local dose rate conditions and budgetary constraints.

Dependent on the cooling time the decay or the activity ratio of the following isotopes can be used for determining the burnup of spent fuel: Cs-134, Cs-137, Eu-154, and Cs-134/Cs-137, Eu-154/Cs-137, (Ru-106

Cs-137)/(Cs-134)2 respectively, cf.

References [5] and [6].

Since gamma rays can be collimated, gamma ray counting is also used for determination of axial burnup profiles [6].

Measurement of isotopic ratios by means of high resolution gamma spectrometry does not require knowledge of the detector yield, but the measurement result significantly depends on the irradiation history of the fuel assembly of interest.

In addition, due to shielding of gamma rays in the fuel pins only a peripheral assay of the fuel assembly can be achieved. This is different from neutron counting.

Accordingly, placing two neutron detectors (e.g. fission chambers) at opposite lateral faces of the fuel assembly and averaging the results from the two detectors leads to a very low sensitivity to horizontal burnup gradients in the fuel assembly. There are in fact detector designs with at least two measurement arms embracing the fuel assembly; each arm is equipped with one neutron detector at least (see below).

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Neutron measurements

o

Passive neutron counting:

This is a very sensitive technique that measures spontaneous fission neutrons from

higher actinides, mainly Cm-244. The technique has a strong count rate dependency

on burnup, cooling time and initial enrichment so that, if the enrichment and cooling

time are known and there is good geometrical control during measurement, an

accurate measure of burnup can be made.

o

Active neutron counting:

This technique measures the enhanced neutron flux from induced fission events that occur in the residual fissile material. The induced fissions are induced or “activated”

by an external source, normally Cf-252, though a development of the technique has been explored that uses the assembly’s inherent passive neutron emission to act as the activation or interrogating source [7]. This development is called “self-interrogation”.

Usually an axial section of the fuel zone of several centimeters in length contributes to the signals recorded in a neutron detector. Consequently, neutron signals are usually used to evaluate the average burnup, whereas burnup profiles are measured with collimated gamma rays.

Since the first derivate of the correlation between count rate and burnup is smaller for the gamma emitters than for neutron emission, application of passive neutron counting results in a more accurate estimate of the burnup than use of gamma measurement methods. For example, the following accuracies have been reported:

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For the BNFL Spent Fuel Monitor, which takes measurements on Cs-137 (i.e. on the 662 keV gamma ray following the decay of Cs-137) and the activity ratio Cs-134/Cs-137, accuracies of 10% at 15 MWd/kg and 5% at 33 MWd/kg [5]

•

For FORK and FORK+ detectors, having two measurement arms each quipped with on thermal neutron detector, one epithermal detector, and one gross gamma detector, accuracies of about 2% [5] (the FORK+ design is equipped with one detector more, a CZT – Cadmium – Zinc – Telluride gamma detector, usually adjusted to the 662 keV ray from the Cs-137 decay)

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For the PYTHON device applied to PWR (using passive neutron and collimated gross gamma counting) 2% in prototype testing and up to 5% under operation conditions [6]

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For the NAJA device (developed to segregate UOX fuel from MOX fuel and to estimate initial enrichment, burnup and reactivity, and using passive and active neutron measurements) up to 2.5% for average burnup, 0.5% for initial U-235 enrichment for fresh UOX fuel between 3 and 4 wt.-% initial enrichment, and about 0.3% in keff [8].

As regards the “assayability” on U-235, U-238, fissile Pu isotopes and total Pu content accuracies between 1% and 5% are listed in Ref. [9] for applications of active and passive neutron measurements to fuel rods (including neutron coincidence collar and rod scanning techniques).

The use of active neutron counting, using a source or self interrogation, has had only limited development as a procedure for the measurement of k

. Depending on the industry’s need development in this area may be useful.

Developments in small solid-state gamma spectrometry detectors, e.g. CZT and combined gamma and neutron counter, e.g. silicon carbide, could potentially find applications in

“in built” cask monitoring devices.

Other measurement data also available from reactor monitoring systems include:

•

In-core detectors

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Ex-core detectors (limited use, particularly for assemblies away from edge of core)

•

Core following data.

One important feature of the techniques mentioned above is that they all involve some level of interpretation or calibration. For example, due to the multiplication effect in a fuel assembly the measured count rate in a passive neutron measurement is not only dependent on the neutron emission and the detector yield but is also impacted by the neutron multiplication factor k

of the fuel assembly. Since k

depends on the initial enrichment, the burnup, the irradiation history, and the measurement environment (measurement in pure or borated water, air, etc.) determination of the fuel assembly’s burnup from the measured count rate requires to make use of the information about the initial enrichment and the irradiation history from the reactor records to obtain a priori known relations between k

and burnup or count rate and burnup at given initial enrichment, irradiation history, and given measurement environment.

Such relations can be for instance obtained and used as follows:

•

Use of a measured calibration curve: For a set of fuel assemblies with very similar irradiation histories, known initial enrichments and known burnups the count rates are measured under the given measurement environment conditions. From the results calibration curves are derived for the correlation between count rate and burnup at given initial enrichment and irradiation history. The irradiation histories of the fuel assemblies for which the burnup has to be determined with the aid of the calibration curves have to be very similar to the irradiation histories on which the calibration curves are based.

•

Use of a calculated correlation curve: By means of a validated online depletion calculation code the isotopic content is calculated as a function of burnup for the initial enrichment and the irradiation history of the fuel assembly to be measured.

Then, considering spontaneous fission and (alpha, neutron) reactions the online code calculates the correlation between burnup and actual neutron emission (from Cm-244 for instance) for the given initial enrichment and irradiation history. In addition, by means of a validated reactivity calculation code the neutron multiplication factor keff of the fuel assembly under examination has to be calculated as a function if the burnup (B) for the given initial enrichment (e), irradiation history (IH) and measurement environment conditions (MEC), keff = keff (B|e,IH,MEC). Using this function and the correlation between burnup and actual neutron emission the measured count rate of the fuel assembly can be evaluated. This is usually done in an iterative way: Using the measured count rate and a first guess of keff a first estimate of the burnup is obtained.

With this first estimate a new estimate of keff is obtained from keff = keff (B|e,IH,MEC). With this new estimate and the measured count rate a further estimate for burnup is obtained. The procedure is continued till convergence in keff is achieved.

In any case it is required to use information from the reactor records for the evaluation of the

measurement results. (The more complex the applied measurement technique is the smaller is

usually the amount of information needed from the reactor records.) The evaluation of the

measurement results cannot be performed independently from reactor record data, but the

measurement results are of course completely independent from the reactor record data. So

therefore, the evaluation of the measurements results is an independent consistency check of

variations in reactor record data are discussed in detail in Ref. [10].

4.3.5. Observations and recommendations