Fuel loading and verification

3.3.3.1 General principles

Implementation of BUC requires a verification of the burnup level of each individual fuel to make sure it satisfies the loading criterion derived from the analysis. This verification can be performed by:

• using the information from the operational (or reactor) records,

• carrying out a measurement (qualitative or quantitative), or

• some combination of both.

Verification obviously should be performed before loading, storing or reprocessing the spent fuel.

Usually, verification is based on the reactor record information. Unless there are other means to ensure keeping of the loading criterion and prevention of misloading events, measurement may be required for checking:

• the reactor records and

• the assignment of the reactor record information to the fuel identification.

Because the measurement is aimed to minimize the risk of misloading events it must be part of the loading sequence. The technology for performing such a measurement is available [3].

The necessity of performing a burnup measurement to check the reactor record information is related to the conditions for which BUC is taken:

• If BUC is taken for normal operation conditions, the necessity of performing a measurement is high;

• If BUC is taken for abnormal operation conditions only, the necessity of performing a measurement becomes less;

• If BUC is taken for design basis accidents only, the necessity to perform a measurement is low or nil;

• If BUC is only necessary for severe accidents with a low probability of occurrence, no measurement is needed.

However, there may be a demand by the public and/or by the relevant regulators to perform checks of the burnup in order to minimize the risk of misloading events, whatever the conditions may be for which BUC is needed.

The accuracy of the burnup measurement is related to the BUC level needed:

• If the BUC taken is less than the burnup, which a fuel assembly gains in one cycle, only a qualitative measurement (e.g. a gross gamma measurement) is necessary giving the information whether the fuel to be loaded in the spent fuel management system of interest is irradiated or not;

• If a higher BUC level is needed, a quantitative measurement of the burnup is necessary.

The outcome of a quantitative measurement confirms the burnup information from the reactor records when the outcome is falling in an interval, named “acceptance interval” in the following, that depends on the burnup value obtained from the reactor records, the uncertainty of this value, the uncertainty of the measurement result, and the probability density of the difference between measurement result and burnup value from the reactor records. It might be possible, that measurement result and burnup value from the reactor records are correlated due to the fact that, dependent on the measurement procedure used, information from the reactor records is necessary for evaluating the measurement result. Such a correlation, if given, has to be taken into account in the estimation of the “acceptance interval”.

The loading criterion is met when the burnup value obtained from the reactor records (and confirmed by measurement if required) is greater than a lower burnup level which depends on the minimum necessary burnup specified by the loading criterion, the uncertainty of the reactor record burnup value and the probability density of the difference between reactor record burnup value and minimum necessary burnup specified by the loading criterion. In other words, the uncertainty of the reactor record information has to be linked to the safety criterion and hence the loading criterion (e.g. by multiplying the loading curve by a possibly burnup-dependent factor adequately representing the uncertainty of the reactor record information at a given confidence level).

A second possibility is, of course, to compare the outcome of the measurement that confirms the reactor record information, directly with the loading criterion. In that case the uncertainty of the measurement has to be linked to the loading criterion. However, a comparison of the measurement outcome with the loading criterion is usually significantly less powerful than a comparison of the reactor record information with the loading criterion, since:

• The precision of the measurement result is usually significantly smaller than the precision of the reactor record information, and

• Usually the measurement result cannot be treated as independent from the reactor record information due to the fact that the procedure to evaluate the measurement outcome relies on some information from the reactor records, and this may lead to a linking of the uncertainty of the reactor record information to the loading criterion in addition to the uncertainty of the measurement.

Therefore, usually the primary quantity is the burnup information from the reactor records.

This quantity has to be compared to the loading criterion. Before that it may be required to confirm this quantity by measurement.

3.3.3.2 Specific fuel handling operations and the need for burnup measurement

As stated in section 3.3.3.1, unless there are other means to ensure keeping of the loading criterion and prevention of misloading events, the necessity of performing a burnup measurement is related to the conditions for which BUC is taken. Thus, the need for measurement depends on the specific spent fuel handling operation to be performed.

Spent fuel movement operations include:

• Movement of fuel within a power plant:

- Transferring fuel assemblies from the core to the spent fuel pool,

- Moving fuel between spent fuel pool locations;

• Loading of fuel into storage/transport casks or canisters and transferring fuel from storage/transport systems into another cask or back into wet storage;

• Loading fuel into the head end process (shear/dissolver) at the reprocessing plant;

• Loading fuel into waste packages for final disposal.

Movement of fuel within a power plant

In this case, reliance on knowledge of the fuel assembly basic parameters (initial enrichment, burnup, etc.) is based on reactor records. Administrative controls, such as verification of the identity, the initial enrichment and the burnup of the fuel assembly being moved, using at least two independent checks, are the subject of well-developed and mature procedures. A check of the reactor record burnup information by burnup measurement is not regarded as a practical option.

Administrative and physical controls of the pool water chemistry, such as boron concentration - if applicable - and verification of pool storage rack conditions - if required - are established to maintain a subcritical configuration in the spent fuel pool for all normal and design basis events.

In the USA and many other countries that apply BUC to PWR wet storage systems, the misloading event is usually regarded as a design basis accident event. As with any other design basis accident scenario, the double contingency principle is applied to misloading events. That means that no other concurrent event has to be postulated when considering the misloading event, and in particular, the possibility of a boron dilution event in a PWR fuel pool does not need to be considered. As a result, neutron absorption provided by the boron diluted in the coolant of PWR spent fuel pools can be credited to the extent guaranteed by the plant technical specifications.

However, in some countries such as Germany a different approach is used. In these countries the misloading event has to be excluded by application of the double contingency principle, i.e. the interpretation of the principle in this case is that two independent, unlikely and concurrent incidents have to happen before a misloading event can occur. Therefore, fuel handling procedures, based on technical measures employing hardware and software controls, have to be set up in such a way that the misloading event is ruled out by virtue of the double contingency principle [3, pp.274-279]. Thus, there is no need to consider the misloading event in the criticality analysis, irrespective of the presence of neutron absorber diluted in the pool water. This approach is used for the following reasons:

• If the misloading scenario is not ruled out by technical measures employing hardware and software controls, the root cause of a misloading event is always an operational error.

Thus, if the event really does occur, there is a high probability of the error remaining undetected. A misloading event that remains undetected and a boron dilution event or any other design basis accident event that takes place at a later time cannot be regarded as

“concurrent events”;

• In addition, since no system designed for BUC application can withstand the misloading of more than one fuel assembly if it cannot withstand the misloading of one fuel assembly, this event needs to be ruled out.

Loading of fuel into dry storage/transport systems

Loading of storage and transport casks or canisters is performed in the pool or in an area adjacent to the pool, i.e. in the presence of water (moderator). The presence of a moderator results in criticality considerations. Depending on the reactor type, the pool water can contain soluble boron. Further, storage-only purpose systems and dual-purpose and dry transport systems are considered.

Storage-only purpose systems

In the USA, the following approach is under discussion: For storage-only purpose casks, criticality safety is mostly considered during cask loading, as it is not an issue once the dry storage cask or canister sits on the storage pad. For the latter, the probability of events leading to re-flooding of the cask cavity is considered to be very low (Note: this may not be the case, if the storage area is located in a zone potentially susceptible to flooding). Therefore, in analogy to the movement of fuel within a power plant, when soluble boron is present in the pool water (such as for PWR spent fuel), loading can be accomplished through administrative controls only, and BUC does not trigger requirements for measurements. When soluble boron is not present in the pool water (such as for BWR spent fuel), BUC levels different from the integral burnable absorber level [4, pp.1] have not been implemented, as yet.

In other countries such as Germany, however, re-flooding of the cask cavity has to be considered as a design basis event. Therefore, the misloading event has to be ruled out by virtue of the double contingency principle, as described for the movement of fuel within a power plant. However, due to the fact that fuel-handling procedures designed for movement of fuel within power plants cannot fully be applied to cask loading, a check of the reactor record burnup information by burnup measurement is regarded as a must.

Dual-purpose and dry transport systems

For dual-purpose (storage and transport) and dry transport systems, criticality safety is considered:

• First, during loading because of the presence of water, as outlined above, and

• Second, in the context of potential accident conditions during transport, which would result in re-flooding of the cask cavity.

In France, BUC has been implemented for spent fuel transport to the La Hague facility. The BUC methodology currently used (actinide-only) consists of two levels of requirements depending on the results of the criticality analysis. If the burnup required is less than the minimum burnup that the reactor operator can guarantee after one cycle of irradiation, a qualitative (go/no-go) measurement is sufficient to prove that the fuel has really been irradiated. Else, a quantitative measurement is required. The measurement has to be made on the last 50 cm of the active fuel length that is the least irradiated, and the average burnup of these last 50 cm is allowed to be taken into account.

In Germany, the BUC methodology currently used (actinide-only) is based either on the French approach (for spent fuel transport to the La Hague facility) or a minimum average burnup of the fuel of 5 MWd/kgU for BWR fuel, if the initial enrichment is higher than 4.2

wt% U-235, and 10 MWd/kgU for PWR fuel, if the initial enrichment is higher than 4.05 wt%

U-235. The respective minimum burnup value needed is guaranteed using the information from the reactor records and checked through the measurement of the γ dose rate of cesium to demonstrate that the fuel has been irradiated.

In the USA, the NRC Interim Staff Guidance (ISG-8) [5] ties acceptance of a BUC methodology (actinide-only) to the determination by measurement of each fuel assembly burnup before loading. This requirement has caused some concern in the USA that the impact on spent fuel pool operation and costs associated with this requirement may discourage the pursuit of BUC implementation by the U.S. operators. Those who are arguing against this requirement present the following rationales:

Operational experience indicates that reactor records provide fairly reliable information with regard to fuel assembly burnup characteristics. Deficiencies in manual data entry, and record archiving and tracking have been the main cause for observed or detected discrepancies between records and measurements. Therefore, the optimum balance among the operators’

desire to minimize operational burden and costs associated with measurement, the benefits deriving from a reduction in the number of shipments, and the risks associated with potential assembly misidentification (misloading) remain to be optimized. Using a risk-informed (what are the risks?), performance based (what is the documented performance?) approach should determine the requirements, at least initially, for any measurement to support the implementation of a specific BUC methodology; any such requirement could be progressively relaxed after a sufficiently long and successful operational experience.

Preparation of spent fuel for reprocessing

Criticality safety is a major consideration in reprocessing facilities using a wet process given the presence of water and organic solvents (moderator), and the resulting separation of the uranium and plutonium species from the fission products. BUC is implemented for reducing analysis conservatism to avoid over-engineering of the facilities, and to keep up processing rates despite increases in initial enrichment of the fuel being reprocessed.

For these reasons, systematic burnup measurement is implemented to validate the data supplied by the reactor operators and set the criticality safety criteria for loading the dissolver.

The data supplied by the operators are regarded as verified by burnup measurement when they fall in a specified interval (or “band”, cf. Section 3.3.3.3).

Preparation of fuel for disposal

The U.S. research program for final disposal of fuel is using a risk-informed, performance-based approach for criticality (see W.H. Lake paper in Session 2.4, Paper No. 1). This approach includes BUC, and accounts for the key actinide and fission product isotopes important to reactivity. The presence of a moderator (water) cannot be ruled out over the time period during which the safety of the repository is assessed. However, if the probability of criticality falls below a regulatory threshold (<10^-4 in 10 000 years), no criticality consequence analyses should be performed. At the time of this writing, the applicant, i.e. the U.S.

Department of Energy, does not plan to perform measurement when it takes ownership of the spent fuel. The applicant proposes reliance on reactors records.

3.3.3.3 Uncertainty of the burnup data

Studies sponsored by EPRI were performed to quantify the magnitude of uncertainties that can be present in burnup value estimates by a PWR utility for their discharged fuel. Uncertainties in average reaction rate in instrumented locations were found to be around 2.2%. The uncertainties of relative assembly power were found to be about 1.8%. Axial evaluation of reaction rate uncertainties indicated that for the top 20%, middle 60%, and bottom 20%, the uncertainty were 6.1%, 2.5%, and 6.7 percent, respectively. After 1 cycle, the assembly average burnup had a 1.5% uncertainty; after 2 cycles, the assembly average burnup had a 1.1% uncertainty, and after 3 cycles, the assembly average burnup had a 1.0% uncertainty (see A. Machiels et al., Paper in Session 2.1, Paper No. 9).

COGEMA’s experience indicates that the value of the burnup as measured at La Hague and the value of the burnup as reported by the French operators differ by a mean average of 5%, which is well within the 15% band constituting COGEMA’s acceptance criterion.