2017 Water Reactor Fuel Performance Meeting September 10 (Sun) ~ 14 (Thu), 2017 Ramada Plaza Jeju Jeju Island, Korea

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1 HALDEN S IN-PILE TEST TECHNOLOGY FOR DEMONSTRATING THE ENHANCED SAFETY OF WATER REACTOR FUELS Margaret A. McGrath 1 1 OECD Halden Reactor Project, IFE: Os Alle 5/P.O. Box 173, 1751 Halden, Norway, Margaret.mcgrath@ife.no ABSTRACT: The Halden Reactor (HBWR) is a research reactor operated by the Institute for Energy Technology (IFE) in Norway. The main research programme of the HBWR is the internationally sponsored OECD Halden Reactor Project (HRP) which includes experimental investigations of different types of fuel under different operating modes, including transient and accident conditions. The member organizations of the HRP represent the whole nuclear community: licensing and regulatory bodies, technical support organizations, utilities, fuel vendors, and research organizations, all of whom are involved in formulating the content of the Joint Research Programme. Based on reactor experimental capabilities, experience and high quality instrumentation developed by IFE for use in the HRP Joint Programme, many bilateral projects are also performed in the HBWR with the aim of generating in-pile data for validating specific models developed for fuel performance codes or contributing to safety and licensing assessments. The HBWR is an ideal test reactor for testing a wide variety of fuels under diverse conditions with its many experimental systems enabling the implementation of transient and accident conditions as well as being able to provide the thermalhydraulic and ALHR conditions prototypic for fuel rods in PWRs, VVERs and BWRs. Normal and innovative BWR and PWR coolant chemistry conditions can also be applied. Test rigs are designed to enable instrumented fuel rods to be subjected to different modes of operation while in-situ, real time measurements of key performance parameters are made. Such integral tests designed to study operation and safety limits are supplemented by separate or focused effects tests aimed at better revealing a particular behaviour. The paper will give an overview of the capabilities of the HBWR for performing fuel safety related R&D with examples given of past experiments performed as part of the HRP Joint Research Programme. KEYWORDS: Halden Reactor, fuel behavior, in-reactor testing, thermal conductivity, FGR, lift-off, LOCA, ATF I. THE OECD HALDEN REACTOR PROJECT The safe and reliable operation of nuclear power plants (NPPs) is supported by long-term national and international R&D that has the ability to provide solutions for current and future technical challenges. The OECD Halden Reactor Project (HRP) is a leading, internationally sponsored project organized under the auspices of the Nuclear Energy Agency (NEA) and operated by the Institute for Energy Technology (IFE) in Norway, with the specific aim of improving safety and reliability of NPPs. The HRP started in 1958 and since then a new 3-year research programme has been devised by the Project s management together with the participating organizations for every subsequent 3-year programme period. The organizations participating in the HRP have varied over the last 59 years but have always fully represented the nuclear community, encompassing licensing and regulatory bodies, utilities, vendors and technical support and research organizations. Twenty countries have membership in the HRP today, with more than 100 organizations participating. The active guidance and scrutiny exerted by the membership on the HRP research programme ensures that it remains focused on issues of direct and practical relevance to the nuclear community and delivers solutions in a direct and effective manner. The HRP delivers research results under three main topic areas for the membership, generating key information for safety and licensing assessments: Nuclear fuels performance under normal operation and transient conditions, as well as under accident scenarios, with emphasis on increased fuel utilization. Nuclear power plant materials behaviour under the combined deleterious effects of coolant water chemistry and neutron irradiation, with emphasis on effective ageing management. 1

2 Human Factors and Digital Systems research for existing and new reactors including human reliability, humansystem interface design, computerised surveillance systems and software dependability. The collaborative nature of the HRP is further fostered by member organizations sending guest scientists to Halden (there have been more than 400 since 1958), who participate directly in research activities and gain training and knowledge within an international research environment. II. NUCLEAR FUELS SAFETY RESEARCH Safe and economical operation of NPPs remains a priority for the nuclear industry, and understanding the phenomena that challenge fuel integrity and component lifetime is essential. Test reactor experiments can generate qualified performance data for model development, validation of fuel performance codes and the assessment of safety criteria. The Halden reactor (HBWR), the basis for the HRP, is unique amongst test reactors with its extensive range of experimental systems and pressurized water loops providing thermal-hydraulic and coolant chemistry conditions prototypic of commercial LWRs. A suite of test rigs, all designed and fabricated in-house, enable nuclear fuels and materials to be subjected to different modes of operation within the HBWR such as normal conditions for long periods of time, short transient conditions and accident scenarios. The HRP has also developed a range of reliable in-pile instrumentation able to measure all the key performance characteristics of fuel such as rod internal pressure (for fission gas release), fuel and cladding temperatures, fuel and cladding elongation (swelling, densification, irradiation growth and stress relaxation) and cladding diameter (clad strain and CRUD deposition). All can be attached to fresh or commercially irradiated samples. The HBWR with its various experimental facilities and capabilities has developed into a powerful tool for international nuclear safety research; the Halden reactor is regarded in many countries as a strategic asset for testing nuclear fuel and reactor components. Fig. 1. Illustration of some of the main fuel rod instrumentation used in the HBWR to monitor online fuel performance. 2

3 Within the HRP nuclear fuels topic area, numerous separate research projects are performed with focus on fuel safety and operational margins. All the projects take the form of highly instrumented irradiation testing carried out in the HBWR followed by post irradiation examination (PIE). II.A. Integral Fuel Performance Tests Integral fuel performance tests are designed to measure combinations of interrelated properties and phenomena such as fuel pellet operating temperature, fuel rod heat transfer characteristics, dimensional stability, fission gas release and cladding strain. To study FGR and PCMI behavior in the HBWR, a test rig containing two fuel rods made from re-fabricated segments pre-irradiated in a commercial NPP is used. The rods are instrumented with a pressure transducer (PF) or a cladding elongation detector (EC) in one end and with a fuel thermocouple in the other end. This instrumentation allows FGR to be detected at the fuel center temperature at which it occurs. The rig has a He-3 coil surrounding the test channel in order to control the power of the fuel in the test assembly independent of the reactor power. Such a test is usually performed with a step-wise increase in fuel rod power - and thus temperature - with steps of around 50 o C held for hours to allow the fission gases to be thermally activated and released if the thermal threshold is exceeded. An example of the in-pile measurements obtained during such a test is shown in Fig. 2. Fig. 2. Example of in-pile measurements made during an integral fuel rod study to determine FGR threshold. Power reductions are usually performed at the end of each hold period in order to ensure that all gas released during the preceding hold period - but trapped due to tight fuel-clad contact or within fuel cracks - actually reaches the pressure detector that is positioned in the fuel rod plenum. With this experimental approach, whereby there is a simultaneous measurement of fuel temperature and gas pressure, the thermally activated FGR threshold can be determined. The empirical correlation for the 1% FGR threshold as a function of fuel temperature and burn-up, developed from data obtained from tests performed both in the Halden reactor and other research reactors to a maximum of ~35 MWd/kg oxide, is shown in Fig. 3. Data obtained later in Halden show that the temperature of fission gas release onset exhibits a decreasing trend at higher burn-ups. It should also be noted that the thermal FGR threshold may be dependent on fuel type (UO 2 or MOX, Gd-doped) but independent of fuel design (PWR, BWR and VVER). 3

4 Fig. 3. Halden empirical thermal threshold for 1% FGR as a function of burn-up. Current emphasis within the HRP Joint Programme is on enhanced performance fuels (ATF) which are aimed at reducing the production or release of fission gases from the fuel pellets and/or reducing the operational temperature of the fuel by improving its thermal conductivity. An example of such a fuel type already tested by the HRP is a BeO additive fuel produced by the Ulba Metallurgical Plant in Kazakhstan. This fuel has been tested comparatively with several other fuel types produced by AREVA, and demonstrated to operate at ~30% lower temperature for the same power level. Lower fuel temperature provides a general increase in operational safety margins but in particular reduces the probability of fission gas release from the fuel pellets, which in turn reduces the inventory of radioactive fission products inside the fuel rod available for release to the reactor environment in the event of a fuel rod failure (cladding beach). Fig. 4. Example from the HRP Joint Programme experiment to study the in-pile performance of several different fuel, showing the effect of BeO doping on the thermal conductivity of UO 2 fuel. 4

5 II.B. Separate or Focused Effects Tests Separate effect tests are where one phenomenon is studied in focus in order to gain a deeper level of knowledge and understanding of the complex and interconnected processes occurring in fuel rods as individually as possible. Such tests also include subjecting fuel to conditions in excess of normal operating conditions in order to magnify an effect and are ideal for aiding development of models of fuel behaviour, including those used in the latest 3D fuel performance codes. An example of using such an approach is the study of fuel rod behavior under high internal overpressure carried out within the HRP where the aim was to determine the extent if any of a lift-off effect: potential fuel-clad gap re-opening leading to increased fuel temperature and thus increased potential for fission gas release, in turn inducing a further fuel temperature increase. The study was designed to evaluate what overpressure would induce an onset of fuel temperature increase. A refabricated pre-irradiated fuel rod was instrumented with a fuel centerline thermocouple, loaded in a pressure flask test rig connected to a water loop system in the HBWR, with the fuel rod connected to an ultra-high pressure gas system. This enabled the unique possibility to study fuel rod behavior with a rod internal pressure of up to 450 bar (overpressure of 300 bar), well beyond normal rod pressure in commercial NPPs. The rod was subjected to increasing levels of rod overpressure, with hold times of about 500 hours, while fuel temperature and cladding elongation were measured online. In order to gain a more complete picture of the fuel performance under rod overpressure conditions, supplementary measurements of hydraulic diameter, fission gas release by means of gamma-spectrometry, and noise analysis were combined with the primary measurements. A test rig with advanced design features including a diameter gauge has since been developed enabling a combination of lift-off by temperature measurements and cladding creep-out by diameter measurements to be studied. Fig. 5. In-pile measurements from a lift-off study carried out within the HRP Joint Programme. In order to detect any clad lift-off, increases in fuel temperature are looked for and in order to reveal what are going to be small changes, the temperature measurements have to be normalized to power. From the example of normalized temperature data shown in Fig. 5 it can be seen that fuel temperature does increase at a slow rate which is proportional to the level of overpressure. Further tests enabled the conclusion that there is no lift-off effect in PWR fuels with Zr-4 cladding, BWR fuel with Zr-2 cladding and VVER fuel with E110 cladding until about 150 bar overpressure. The complete analysis of these tests shows that many factors are at play: cladding temperature; power rating before and during the tests; fuel burn-up and fuel-clad bonding; and the type of cladding including its texture and level of hardening under irradiation. 5

6 II.C. Transient Conditions and Accident Scenarios LOCA testing within the HRP Joint Programme was established with the aim of creating a knowledge base for safety assessments and to demonstrate performance of different fuels in accident situations. The main objective of the experiment series is to study the effect of fuel fragmentation, relocation and dispersal with secondary hydriding of the cladding and fission product release as additional objectives. A specially designed test rig and water loop, with simulation of blow-down and a radioactive dump-tank, were developed for the planned series of experiments. The schematic view of the rig and an example of the fuel gamma scan obtained just after one of the LOCA tests are shown in Fig. 6. The experiment setup consists of a single fuel rod inserted into a pressure flask connected to a water loop. A low level of nuclear power generation in the fuel rod is used to simulate decay heat (10 25 W/cm depending on required peak clad temperature to be reached). The electrical heater surrounding the rod is simulating the heat from neighbour rods. The instrumentation consists of two cladding thermocouples (TC) at the upper part of the rod, one cladding TC at the lower part, three heater thermocouples at different axial elevations, a cladding extensometer and a rod pressure sensor. After the test, the main focus is on detecting fuel fragmentation and relocation, both of which are examined first using non-destructive gamma scanning and neutron radiography followed by PIE. New gamma scanning equipment with tomography capability has now been installed for these tests to investigate fuel relocation inside the ballooned region of the fuel cladding, in particular with the aim of determining the packing fraction to thereby evaluate cladding temperature in the ballooned region. Neutron radiography reveals more details than the initial gamma scanning, and intact pellets as well as coarse fragments can be clearly revealed. Ceramography is used to confirm the fragmentation with more detail at selected locations and sieving of the fuel inventory collected from inside and, to the extent possible, outside the rod gives the overall fuel fragment size population. Fig. 6. LOCA test rig and gamma scan of the rod tested. 6

7 What has been found by implementing in-pile semi-integral rod testing is that Zircaloy cladding balloons and bursts as it approaches about 800 o C and the high burn-up UO 2 fuel pellets inside the fuel rod fragment into a range of particle sizes, including in the micrometre range, which relocate to the ballooned region and are then expelled from the fuel rod. Fuel fragmentation is not an obvious function of a single parameter, although it is obviously affected by burn-up and it would seem the formation of high burnup structure plays an important role. Another important influence on fuel fragmentation seems to be cladding distension. Where the pellets stay in contact with the cladding, they can remain quite unaffected by the LOCA transient and it can be surmised that spacers will thus play an important role. The effect of a spacer on fuel rod behaviour under LOCA conditions is to be studied in the next HRP LOCA test planned for September Such LOCA testing can lead to changes in safety criteria or their implementation, with much emphasis placed on limiting the radiological consequences of an accident even if it means restricting discharge burn-up levels. Currently there is much focus on innovations within cladding design (ATF) aimed at reducing the likelihood of cladding ballooning and burst and on possibly delaying onset of high burn-up structure formation. III. CONCLUSIONS The results from the many different types of tests performed in the Halden reactor are used for integral investigations of fuel behavior within design limits. These tests are important for understanding fuel behavior and can provide information enabling operational design margins to be safely reduced and the utilization of fuel to be increased. Other types of the tests performed beyond operational and safety limits are done in order to assess operational and safety limits. These tests may well be used for confirmation of the effectiveness of some innovative fuel developments aimed at extending safe utilization. ACKNOWLEDGMENTS All the research projects in the Halden reactor are financially supported by the HRP member organizations within the Joint Programme or on a bilateral basis. The technical solutions, design of the test rigs and evaluation of the results are provided by the qualified and experienced staff at IFE. Thanks to all, an extensive test programme has been performed continuously over the last six decades which will continue in the future with the continued international co-operation. 7