The characteristics of LWR fuel at high burnup and their relevance to AGR spent fuel

58 Pages·2011·3.93 MB·English

The characteristics of
LWR fuel at high
burnup and their
relevance to AGR
spent fuel

NNL (10) 10930
Issue 2

A report prepared for and on behalf of
the Nuclear Decommissioning Authority (NDA)

Page 3 of 58

NNL (10) 10930
Issue 2

EXECUTIVE SUMMARY

With emphasis on high burnup behaviour, the available information on the characteristics
of spent LWR fuel have been reviewed, and the relevance of the resulting information to
spent AGR fuel has been assessed. The characteristics covered were those that have the
potential to affect the release of radionuclides after final disposal of AGR fuel, and during
the pre-disposal transport and any re-packaging of such fuel. These are: (i) stable fission
gas inventory in the rod free volume prior to disposal; (ii) radioactive fission product
inventory in the rod free volume and fuel grain boundaries prior to disposal;
(iii) solubility of fuel matrix in groundwater.
Due to the similar maximum fuel centre temperatures and radial temperature gradients
in AGR and LWR fuel pellets, the fission gas release process in AGR UO at low to
2
intermediate burnup is expected to be phenomenologically the same as that in LWR UO .
2
The presence of the pellet bore in AGR fuel is not observed to have a direct effect on
fission gas release.
Although phenomenologically the same, there are two important differences between gas
release in AGRs and LWRs: (i) deposition of carbonaceous material onto the outer surface
of the AGR pin can significantly increase fuel temperatures, and hence fission gas
release; (ii) the low hydrostatic stress imposed upon AGR fuel pellets throughout life due
to the collapsible can concept leads to low morphological relaxation of grain face bubbles.
The effects of carbonaceous deposition on fission gas release can be significant — at
~ 20 MWd/kgU, the fission gas release fraction can be increased from ~ 0.1% to ~ 5%.
This would give significantly increased grain boundary plus pin free volume inventories of
the important radionuclides 129I and 137Cs. More significant increases in gas release, and
hence also in 129I and 137Cs inventories, could be expected at higher burnups given
similar levels of deposition.
Low morphological relaxation of grain face bubbles will generally only be important given
well developed interlinkage when fission gas release fractions are relatively high (above
~ 3%). Thus, in the absence of carbonaceous deposition, this should not be a significant
issue given current and future AGR irradiation strategies.
As for LWR fuel, columnar grain growth is not expected in normal operation for
commercially irradiated AGR fuel, due to the maximum fuel centre temperatures being
significantly lower than the threshold for columnar grain formation of ~ 1800°C. Due to
the similar fuel centre temperatures, equi-axed grain growth in AGR fuel is — both
qualitatively and quantitatively — similar to that in LWR fuel.
The maximum pellet burnup for all current and future commercial AGR operation is
~ 50 MWd/kgU. At this pellet average value, the rim burnup would be expected to be
~ 65 MWd/kgU, which suggests that there will be no high burnup structure in AGR fuel.
However, given the uncertainties, it is possible that there is some partial high burnup
structure, although complete high burnup structure formation is unlikely. Post-irradiation
examination data for high burnup AGR fuel (which is currently proprietary) and/or more
detailed calculations, would be required to investigate this further.
The currently available published data for fission gas release in commercially irradiated
AGR fuel are for pin average burnups up to 38 MWd/kgU. Given no carbonaceous
deposition, the measured fission gas release fraction increases from ~ 0.01% at
beginning of life to ~ 0.5% at 38 MWd/kgU. Scoping calculations using the ENIGMA fuel
performance code have shown that the ENIGMA fission gas release predictions are in
good agreement with these data. Extending the predictions to 50 MWd/kgU gives 1.5%
fission gas release at this burnup.
Page 4 of 58

NNL (10) 10930
Issue 2

Given the similarities in maximum fuel centre temperatures and radial temperature
gradients in AGR and LWR fuel pellets, and the use of a helium fill gas in both AGR pins
and LWR rods, the average concentrations and chemical forms of the fission products for
fuel of a given burnup are expected to be broadly similar. The differences in neutron
spectrum, in pellet dimensions, and in chemical interaction between fuel and clad/can will
give some differences, but these are not expected to be particularly significant.
The softer neutron spectrum in the AGR means that, for a given burnup: the average
heavy metal fraction of plutonium will be less than in an LWR; and the radial distribution
of plutonium will be less peaked towards the rim.
Leaching tests on spent MOX and UO fuel suggest that the presence of plutonium (and
2
higher actinides) increases radioactive fission product inventories in the fuel grain
boundaries and rod free volume, but has little effect on matrix solubility.
Given the discussion in the previous three paragraphs, together with the expected lack of
high burnup structure in AGR fuel, and the uncertainties in the understanding and
measurement of leaching and dissolution, the conclusions drawn for LWR fuel with
respect to the radioactive fission product inventory in the rod free volume and fuel grain
boundaries prior to disposal, and with respect to the solubility of the fuel matrix in
groundwater, can be considered to also apply to AGR fuel in a conservative manner.
Thus:
(a) both the measured rod free volume and grain boundary inventories of 90Sr and
99Tc should be very low at between 0.01 and 0.4% of the total amounts of the
nuclides;
(b) the combined rod free volume and grain boundary inventories of both 129I and
137Cs are expected to be correlated to fission gas release (with a 1:1
correspondence for 129I, and a ratio for 137Cs which is roughly 1:1 for fission gas
release below ~ 1%, and tends towards 1:3 as fission gas release increases);
(c) matrix dissolution is best modelled with a fractional mass loss of 10-7 per year.
However, these tentative conclusions all require confirmation from experiments on AGR
fuel which take account of the nature of spent AGR fuel and of the geological disposal
facility environment in which AGR elements would be emplaced. Since AGR pins with
carbonaceous deposition would be expected to have significantly higher fission gas
release than pins with no deposition, they would also be expected to have significantly
higher grain boundary plus pin free volume inventories of 129I and 137Cs.
Page 5 of 58

NNL (10) 10930
Issue 2

VERIFICATION STATEMENT

This document has been verified and is fit for purpose. An auditable record has been
made of the verification process. The scope of the verification was to confirm that: -

(cid:127) The document meets the requirements as defined in the task specification/scope
statement
(cid:127) The constraints are valid
(cid:127) The assumptions are reasonable
(cid:127) The document demonstrates that the project is using the latest company approved
data
(cid:127) The document is internally self consistent

HISTORY SHEET

Issue Number Date Comments
Draft 1 25/03/10 Initial draft for verification
Draft 2 31/03/10 Verified draft for approval
Issue 1 31/03/10 Verified and approved version for issue to
customer
Issue 2 Draft 1 24/06/11 Draft of Issue 2 for verification and approval
Issue 2 28/06/11 For issue to customer after verification and
approval of changes from Issue 1 to Issue 2
Page 6 of 58

NNL (10) 10930
Issue 2

CONTENTS
Page
1. INTRODUCTION.............................................................................................9
2. DESCRIPTION OF LWR AND AGR FUEL DESIGNS..........................................10
2.1. Key Parameters for Fuel Designs ..............................................................15
3. LWR AND AGR CORE DESIGNS AND IRRADIATION CONDITIONS.................16
4. EFFECTS OF IRRADIATION CONDITIONS ON FUEL CHEMISTRY...................19
5. REVIEW OF CHARACTERISTICS OF SPENT LWR FUEL...................................20
5.1. Stable Fission Gas Inventory in the Rod Free Volume Prior to Disposal..........20
5.1.1. Fission gas generation rate ..............................................................20
5.1.2. Fission gas swelling and release at low to intermediate burnup.............21
5.1.3. Effects of fuel restructuring..............................................................27
5.1.3.1. Grain growth..........................................................................27
5.1.3.2. High burnup structure .............................................................28
5.1.4. Fission gas swelling and release at high burnup..................................36
5.2. Radioactive Fission Product Inventory in the Rod Free Volume and Fuel Grain
Boundaries Prior to Disposal.....................................................................37
5.2.1. Important radioactive fission products for spent AGR fuel ....................37
5.2.2. Important radioactive fission products for spent AGR fuel: inventories in
spent LWR fuel...............................................................................39
5.3. Solubility of Fuel Matrix in Groundwater.....................................................41
6. CHEMICAL COMPOSITION OF IRRADIATED AGR FUEL.................................43
7. RELEVANCE OF SPENT LWR FUEL INFORMATION TO SPENT AGR FUEL........47
7.1. Stable Fission Gas Inventory in the Rod Free Volume Prior to Disposal..........47
7.1.1. Measurements of fission gas release from commercial AGR fuel............49
7.1.2. ENIGMA calculations of fission gas release from commercial AGR fuel....49
7.2. Radioactive Fission Product Behaviour.......................................................50
8. CONCLUSIONS.............................................................................................52
9. REFERENCES................................................................................................54
Page 7 of 58

NNL (10) 10930
Issue 2

LIST OF TABLES
Page
Table 1: Key parameters for PWR, BWR and AGR fuel assemblies...................15
Table 2: Typical LWR and AGR core design parameters and irradiation conditions
........................................................................................................18
Table 3: Cumulative fission yields of stable fission gas isotopes .....................21
Table 4: The twenty most radiotoxic nuclides in a spent AGR fuel element at
31 MWd/kgU [14]............................................................................38
Table 5: Cumulative fission yields and half-lives of important radioactive fission
products for AGR fuel.......................................................................39
Table 6: Irradiation data for AGR fuel samples examined by EPMA .................44

LIST OF FIGURES
Page
Figure 1: Schematic of typical LWR fuel rod ....................................................11
Figure 2: Typical 17x17 PWR fuel assembly ....................................................11
Figure 3: General Electric 7x7 BWR fuel assembly [1].....................................12
Figure 4: Commercial AGR fuel element (single sleeve design) [4].................14
Figure 5: Halden threshold for significant fission gas release [24]..................23
Figure 6: Fission gas release versus burnup data for commercially irradiated
FRAGEMA and BNFL PWR fuel [25,26].............................................24
Figure 7: Fission gas release versus burnup data for commercially irradiated
ABB BWR fuel [27]...........................................................................25
Figure 8: Fission gas release versus burnup data for commercially irradiated
VVER fuel [28].................................................................................26
Figure 9: CEA fission gas release versus burnup data for French PWR fuel [15]
........................................................................................................27
Figure 10: SEM radial scan of fractured fuel sample at 73 MWd/kgU [34].......29
Figure 11: Enlargement of rightmost boxed section in Figure 10 [34].............29
Figure 12: Local temperature and burnup for HBS formation [36]...................30
Figure 13: Local xenon concentration versus local burnup for commercially
irradiated LWR fuel [33]..................................................................31
Figure 14: Rim width versus pellet average burnup for commercially irradiated
PWR fuel [37]..................................................................................32
Figure 15: Rim width versus pellet average burnup [38].................................32
Figure 16: Measured and predicted (TUBRNP) radial profiles of burnup and
plutonium concentration in irradiated BWR fuel [35] ......................33
Figure 17: Appearance of HBS porosity in PWR fuel with pellet average burnups
of 67 and 102 MWd/kgU [37]..........................................................34
Figure 18: Rim porosity volume fraction versus local burnup for commercial
PWR fuel [37]..................................................................................35
Page 8 of 58

NNL (10) 10930
Issue 2

Figure 19: Rim porosity volume fraction versus local burnup for various fuel
restraint conditions [40]..................................................................35
Figure 20: 129I inventories versus fission gas release for spent LWR UO , as
2
derived from leaching tests [15] .....................................................40
Figure 21: 137Cs inventories versus fission gas release for spent LWR UO , as
2
derived from leaching tests [15] .....................................................40
Figure 22: Typical radial distributions of caesium and xenon concentration for
LWR fuel subjected to a power ramp [46] .......................................41
Figure 23: Xe, Cs and Pu radial concentration profiles for AGR fuel pellets [49]
........................................................................................................45
Figure 24: Comparison of Xe and Cs radial concentration distributions for the
intact pin from experimental stringer IE1510/2 [50]......................46
Figure 25: Morphology of grain face bubbles in irradiated LWR fuel (left) and
irradiated AGR fuel (right) [52].......................................................48
Figure 26: Measured and predicted fission gas release versus burnup for AGR
element 6.........................................................................................50
Page 9 of 58

NNL (10) 10930
Issue 2

1. Introduction
The Radioactive Waste Management Directorate (RWMD) of the UK’s Nuclear
Decommissioning Authority (NDA) is undertaking an extensive research programme to
support disposal of the spent fuel, high level waste, intermediate level waste, plutonium
and uranium which has arisen, and which will continue to arise, as part of UK civil nuclear
operations. The UK National Nuclear Laboratory (NNL) has been contracted to perform
work under Phase 1 of this programme.
A significant fraction of the current UK spent fuel inventory is comprised of fuel elements
irradiated in advanced gas-cooled reactor (AGR) plants. Further AGR fuel will be added to
the inventory during the remaining lifetimes of the AGR units. Hence, the characteristics
of spent AGR fuel, and how they would impact the fuel behaviour during final disposal in
a geological disposal facility (GDF), are important. Since commercial AGR systems are
unique to the UK, the primary information on the characteristics of spent AGR fuel comes
from the UK knowledge base, in particular from the data accumulated during post-
irradiation examination (PIE). However, since both AGR fuel pins and light water reactor
(LWR) fuel rods contain uranium dioxide fuel pellets which are subject to similar
irradiation conditions (primarily chemical environment, neutron flux, fission density, and
temperature), secondary information on the characteristics of spent AGR fuel can
potentially be obtained from a study of spent LWR fuel. In particular, since LWR fuel has
been irradiated to considerably higher burnup (energy produced per unit mass) than AGR
fuel, information on high burnup LWR fuel can be used to predict the effects of any future
increases in burnup in AGR fuel. This suggests the following work, with the emphasis on
high burnup behaviour: (i) review the available information on the characteristics of
spent LWR fuel; and (ii) assess the relevance of this information to spent AGR fuel. Such
work is the subject of this report, and forms task SF2 of the NNL’s work scope under the
Phase 1 programme.
The design of LWR and AGR fuel assemblies, and the LWR and AGR core designs and
irradiation conditions, are first described in Sections 2 and 3. The effects of the
irradiation conditions on the fuel chemistry are then briefly discussed in Section 4. The
characteristics of spent LWR fuel are reviewed in Section 5, and the relevance of the
resulting information to spent AGR fuel is assessed in Section 7; the intermediate section
(Section 6) provides a summary of the chemical composition of irradiated AGR fuel, since
this is relevant to the discussion in the subsequent section. Finally, conclusions are drawn
in Section 8.
Page 10 of 58

NNL (10) 10930
Issue 2

2. Description of LWR and AGR Fuel Designs
This section describes the main features of fuel designs for commercial LWRs and AGRs.
With respect to LWR fuel, both pressurised water reactor (PWR) and boiling water reactor
(BWR) designs are discussed.
The fissile material in a light water reactor is in the form of UO or (U,Pu)O (mixed
2 2
oxide, or MOX) fuel pellets. These fuel pellets are loaded into cladding tubes to form fuel
rods. The fuel rods are in turn fastened together to form fuel assemblies.
Some rods have pellets that are doped with burnable neutron absorbers, primarily Gd O ,
2 3
but sometimes Er O . Westinghouse fuel pellets may instead have a ZrB coating on the
2 3 2
surface of the pellets for neutron absorption purposes. With the exception of Russian
design (VVER) fuel, the pellets are solid cylinders, generally with dishes and chamfers at
each end. VVER pellets are hollow cylinders. The stack length is typically 12 ft (3.7 m),
and a typical pellet length is ~ 10 mm, so there are generally ~ 370 pellets per rod.
PWR and BWR cladding tubes are composed of zirconium alloys and are generally ~ 4 m
in length, ~ 1 cm in outer diameter, and less than 1 mm in thickness. With the exception
of Russian design (VVER) cladding, which is Zr-1%Nb, PWR cladding has historically been
of the Zircaloy-4 variant (Zr with 1.5% Sn, 0.2% Fe and 0.1% Cr). In contrast, BWR
cladding has historically been of the Zircaloy-2 variant (Zr with 1.5% Sn, 0.15% Fe,
0.1% Cr and 0.05% Ni)*. Optimised, or low tin, versions of Zircaloy-4, and newer
zirconium alloys such as M5™ and ZIRLO™, have been introduced in more recent times
in PWRs. The cladding tubes are sealed by top and bottom end plugs which are welded to
the tubes. The tubes are then backfilled with helium to maximise heat transfer between
the fuel pellets and the cladding.
A schematic of a typical LWR fuel rod is shown in Figure 1. In addition to the cladding
and fuel pellets, there is a plenum spring (typically of Inconel) to hold down the fuel
pellets during handling and transport, and top and bottom end plugs which are welded to
the cladding tube (and are generally of the same zirconium alloy as the cladding). There
may also be a fuel support tube below the fuel stack (for rods with both an upper and a
lower plenum), and/or insulation pellets (typically of alumina) above and/or below the
active fuel stack (to prevent chemical interaction of the upper and/or lower fuel pellets
with the rod components they would otherwise be in contact with).
The fuel rods in a PWR assembly are held within the assembly skeleton, which, with the
exception of VVER assemblies, consists of a top and bottom nozzle, a number of guide
tubes/thimbles (into which are inserted control rods and other core components), an
instrumentation tube/thimble, and a number of grids. A typical modern 17x17 assembly
(with 264 fuel rods, 24 guide tubes and an instrumentation tube in a square lattice),
complete with debris filter bottom nozzle, is illustrated in Figure 2 (together with a
control rod assembly on the right). VVER assemblies are significantly different in that:
(a) there are no guide tubes (control assemblies which move between fuel assemblies
are used instead of control rods); (b) the fuel rods and instrumentation tube form an
hexagonal lattice; (c) an assembly shroud tube is attached to the top and bottom nozzles
to form a coolant flow channel.

* In some LWR designs a thin layer of a different zirconium alloy is applied to the Zircaloy base material to form
duplex cladding (with the thin layer on the outside to reduce corrosion) or liner cladding (with the thin layer on
the inside to reduce the propensity for pellet-cladding interaction (PCI) failures).