Fuels and Materials Programme Achievements 2015

Transcription

1 Institutt for energiteknikk OECD HALDEN REACTOR PROJECT HP-1487 vol. 1 For use within the Halden Project Member Organisations only OECD HALDEN REACTOR PROJECT Fuels and Materials Programme Achievements 2015 March 2016

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3 FOREWORD The experimental operation of the Halden Boiling Water Reactor and associated research programmes are sponsored through an international agreement by: the Institutt for energiteknikk (IFE), Norway, the Belgian Nuclear Research Centre SCK CEN, acting also on behalf of other public or private organisations in Belgium, the Czech company UJV Rez a.s. (UJV), acting also on behalf of other public or private organisations in Czech Republic, the Technical University of Denmark (DTU), the Finnish Ministry of Employment and the Economy (MEE), the Electricité de France (EDF), the German Gesellschaft für Anlagen- und Reaktorsicherheit mbh (GRS), representing a group of companies working in agreement with the German Federal Ministry of Economic Affairs and Energy, the Hungarian Centre for Energy Research (MTA EK), Hungarian Academy of Sciences, representing a group of national and industry organisations in Hungary, the Japanese Nuclear Regulation Authority (NRA), the Korea Atomic Energy Research Institute (KAERI), representing a group of national and industry organisations in Korea, the Russian Joint-Stock Company TVEL, representing a group of Russian nuclear industry and research institutes, the Slovak company VUJE, a.s. (VUJE), representing a group of national and industry organisations in the Slovak Republic, the Spanish Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT), representing a group of national and industry organisations in Spain, the Swedish Radiation Safety Authority (SSM), representing a group of national and industry organisations in Sweden, the Swiss Federal Nuclear Safety Inspectorate (ENSI), representing both the Swiss nuclear Utilities and the Paul Scherrer Institute, the National Nuclear Laboratory Limited (NNL), representing a group of nuclear licensing and industry organisations in the United Kingdom, the Federal Authority for Nuclear Regulation (FANR), United Arab Emirates, and the United States Nuclear Regulatory Commission (USNRC), and as associated parties: the French Institut de Radioprotection et de Sûreté Nucléaire (IRSN), Le Commissariat á l energie Atomique et aux Energies Alternatives (CEA), France, EU JRC Institute for Transuranium Elements, Karlsruhe, Germany, Japan Atomic Energy Agency (JAEA), the Central Research Institute of Electric Power Industry (CRIEPI), Japan, the Mitsubishi Nuclear Fuel Ltd. (MNF), Japan, and associated parties from USA: the Westinghouse Electric Power Company, LLC (WEC), the Electric Power Research Institute (EPRI), the Global Nuclear Fuel (GNF) Americas, and the US Department of Energy (DOE) The right to utilise information originating from the research work of the Halden Project is limited to persons and undertakings specifically given this right by one of these Project member organisations. Recipients are invited to use information contained in this report to the discretion normally applied to research and development programmes. Recipients are urged to contact the Project for further and more recent information on programme items of special interest. The present report is part of the series of Halden Work Reports (HWRs) which primarily are addressed to the Halden Programme Group (the technical steering body of the Halden Project) as a for its continuous review of the Project's research programmes. The HWR-series includes work reports on the different research items of the Project, reports on the findings from workshops convened under the auspices of the HPG, and other internal reports to the HPG on the findings from the Project's activities to the extent deemed desirable by the HPG or the Project staff.

4 ABSTRACT This report is intended to summarise the accomplishments of the Fuels and Materials research programme of the Halden Reactor Project during 2015, addressing the most important achievements of the programme. For each work item, the objectives and main results are outlined together with the direction of future activities. This summary is presented in a concise form and serves the purpose of giving an immediate overview of the programme results. For more insights, updated references are given. NOTICE THIS REPORT IS FOR USE BY HALDEN PROJECT PARTICIPANTS ONLY The right to utilise information originating from the research work of the Halden Project is limited to persons and undertakings specifically given the right by one of the Project member organisations in accordance with the Project's rules for "Communication of of Scientific Research and Information". The content of this report should thus neither be disclosed to others nor be reproduced, wholly or partially, unless written permission to do so has been obtained from the appropriate Project member organisation.

5 Membership Executive Summary Governance, events and meetings in the Halden Reactor Project 2015 The OECD Halden Reactor Project started the new agreement period ( ) with eighteen member countries. The member countries comprise Belgium, Czech Republic, Denmark, Finland, France, Germany, Hungary, Japan, Korea, Norway, Russia, Slovak Republic, Spain, Sweden, Switzerland, the United Arab Emirates, the United Kingdom, and the United States. Steering Groups The Halden Project has two international steering groups, the Board of Management and the Programme Group whose function is defined in the Halden Agreement. These groups meet twice a year. The chairman of the Halden Board of Management in 2015 was Kun Woo Song from KAERI, Korea. The chairman of the Halden Programme Group in 2015 was Olli Ventä from the VTT, Finland. Meetings and events The results from the work of the Halden Project are traditionally presented in two Enlarged Halden Programme Group Meetings (EHPG meeting) per programme period. These EHPG meetings also provide an opportunity for participating organisations to present results from their own research. During the programme period, the first EHPG meeting will be held in Sandefjord, Norway in May, The second EHPG meeting will be held at Lillehammer, Norway in September, The OECD-HRP summer schools were initiated in 2000 following a proposal of the Halden Board of Management to facilitate knowledge transfer, especially to the young generation. The Summer School of 2015 was within the topic of Fuel Assembly and Cladding Materials. Workshop meetings on selected subjects related to the Joint Programme are a means to evaluate and guide the work of the Halden Project. The following workshops were arranged in 2015: Workshop on Fuel cladding behaviour in interim dry storage conditions, Espoo (Finland), April 8, Workshop on «Accident Management and Operation», Halden, May 20 21, Workshop on Fuel fragmentation, relocation and dispersal (FFRD) experimental, mechanisms and modelling approaches, Aix-en-Provence (France), May 20-21, HOLMUG meeting, Smolenice, Slovakia, October 7-8, 2015.

6 Executive summary of Fuels and Materials achievements The Fuels & Materials programme was defined and executed under two main chapters: Fuel safety and operational margins Plant Ageing and Degradation Approximately 9 test rigs were under irradiation at any one time during 2015 as part of the Halden Reactor Project Joint Programme, with a total of 11 unique in-pile experiments being performed during In addition to this there were 13 tests either undergoing PIE or in preparation for starting irradiation. Reactor availability throughout 2015 was 50%. Fuel safety and operational margins is related to fuels and claddings in use in light water reactors (PWR, BWR, VVER), comprising, for the programme, standard UO 2, Gd-bearing UO 2 and Cr-doped UO 2 as well as UO 2 fuel with addition of BeO. Cladding materials include Zircaloys, ZIRLO, M5 and M-MDA alloys. The objective is to provide fuel and cladding property data for design and licensing from zero to MWd/kg, including commercially irradiated fuels. Research activities focus on: Integral fuel performance Fuel behaviour separate effects Fuel behaviour in transients and under accident scenarios Cladding performance and behaviour A highlight from these studies is the start of irradiation of IFA-744, which has the objective to investigate fuel thermal conductivity degradation and recovery mechanisms, in particular assess the significance of fission density on thermal conductivity and study whether irradiation damage annealing and thermal conductivity recovery are also activated in-pile. The experiment is designed such that the fuel temperature can be varied independently of the power, and temperature increases can be introduced after some burnup accumulation in order to observe a conductivity recovery effect, if any. Another special feature is the use of segments where the pellet-clad gap is filled with liquid metal which leads to lower fuel temperatures. The purpose of this feature is to facilitate data analysis by removing the uncertainty associated with gap conductance. Data from the start of irradiation indicate that conductivity degradation induced by irradiation damage can be seen at low temperature. Plant ageing and degradation studies focus on the generation of validated data on stress corrosion cracking of reactor component materials at representative stress, temperature, neutron flux and water chemistry conditions. Stress relaxation is also addressed as well as a study related to RPV. A highlight from these studies is completion of the PWR IASCC test IFA-772, which included compact tension (CT) specimens fabricated from a variety of stainless steels. It was observed that the overall crack growth rates of 316 SS specimens were about an order of magnitude greater than those of the 321 SS samples.

7 HP-1487 vol. 1 OECD Halden Reactor Project HALDEN PROJECT PROGRAMME ACHIEVEMENTS 2013 CONTENTS Page OVERVIEW OF FUELS AND MATERIALS EXPERIMENTS DURING PIE ON GADOLINIA FUEL FROM IFA IRRADIATION OF VVER GADOLINIA FUEL (IFA-676)... 3 PIE ON VVER FUELS (IFA-676)... 4 FGR MECHANISMS (IFA-716)... 5 PIE ON GADOLINIA FUEL FROM THE FGR TEST IN IFA OVERPRESSURE TESTING ON SR-M MDA CLADDING IN IFA FUEL THERMAL CONDUCTIVITY AND RECOVERY MECHANISMS (IFA-744)... 8 LOCA TEST IFA GAMMA SCANNING FROM LOCA TEST IFA HIGH BURN-UP FUEL DISK IRRADIATION (IFA-655.2) FGR FROM HIGH BURNUP MOX DISKS (IFA-629.8) STEADY STATE AND TRANSIENT CLADDING CREEP (IFA-741) PWR CLADDING CORROSION TEST IFA PIE ON THE COATED CLADDINGS TEST IFA PWR CRACK GROWTH RATE TEST (IFA-772) CRACK INITIATION STUDY (IFA-733) CREEP OF ODS STEELS IN IFA REPORTING March 2016 HALDEN PROJECT USE ONLY The information contained in this report is to be communicated only to persons and undertakings authorised to receive it by one of the organisations participating in the OECD Halden Reactor Project in accordance with the Project s rules for communication of information

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9 - 1 - HP-1487 vol. 1 OVERVIEW OF FUELS AND MATERIALS EXPERIMENTS DURING 2015 MEASUREMENTS APPLICATIONS IFA Test materials # of rods / specimens Burnup MWd/ kg oxide fluence n/cm 2 (x10 20 ) temperature pressure clad elongation fuel elongation gas flow oxide thickness crack length clad diameter ECP thermal conductivity fission gas release densification, swelling PCMI clad creep corrosion IASCC irr. ind. mat. changes Recent reports Comment UO 2 /RX-M-MDA/PWR 1 65 X X X X X X X HWR-1089 PWR Overpressure, PIE UO 2 /SR-M-MDA/PWR 1 65 X X X X X X X HP-1485 PWR Overpressure MOX discs X X X X HP-1474 FGR ramp, PIE MOX discs X X X X HP-1485 FGR ramp UO 2 /M5/PWR 1 65 X X X X X X X HP-1485 LOCA test UO X X X HP-1485 Disks, rim 676 UO 2 /Gd-doped/Al 2 SiO 5 -doped 6 44 X X X X X X X X HWR-1093 Base irradiation 681 UO 2 /Gd-doped 6 44 X X X X X X X X HWR-1038, HWR-1091 Gd-fuel, PIE 708/785 M5, Zirlo, M-MDA-SR 6 24 X X X HWR-1098 Cladding corrosion, PIE 716 UO 2 /Cr-doped/BeO-doped 6 24 X X X X X X HWR-1090, HP-1474 High LHR, FGR UO 2 /E-110/VVER 1 50 X X X X HWR-1011, HWR-1012 Ramp, PIE UO 2 /Gd-doped/Zirlo/PWR 2 54 X X X X HWR-1087 Ramp, PIE 731 Zry-2, Zry-4 5 X X HWR-1046, HP-1409 On-line corrosion, PIE L dpa X X X HP-1485 BWR crack initiation 741 E110, M5, M-MDA, Opt. zirlo 4 19 X X X X X HWR-1097, HP-1485 Cladding creep 744 UO 2 6 X X X X HP-1485 Thermal conductivity L, 321, 304, CW316 SS dpa X X X HP-1485 PWR crack growth 774 UO 2, coated-zry X X HP-1485 Coated fuel cladding

10 - 2 - HP-1487 vol. 1 PIE ON GADOLINIA FUEL FROM IFA-681 Quantify the impact of different degrees of Gd 2 O 3 in oxide solution on thermal-mechanical operation characteristics, such as in-pile densification, solid selling, fission gas release and thermal performance under representative irradiation conditions. The test assembly was unloaded for PIE in 2012 after ~1300 FPD of irradiation. The discharge burnup was ~47 MWd/kgOx for the UO 2 and 2% Gd rods, and ~34 MWd/kgOx for the 8% Gd rods. Axial gamma scanning was performed on all six rods together with a calibration fuel rod for determination of fuel rod burn-up. Density, ceramography, puncturing and fission gas analysis were performed on the three rods with solid fuel pellets. The measured densities were 10.20, and g/cm3 respectively for UO 2, 2% Gd and 8% Gd rods. The burn-up estimation coincides well with the results obtained in-pile. Cross section of 2% Gd test rod (681-2) in polished and etched condition Main ref. Remaining work Six rods in one cluster. Three rod pairs of UO 2, and 2 wt% Gd and 8 wt% Gd fuel. Each pair consists of one rod loaded with solid and one loaded with hollow fuel pellets, which were delivered by ENUSA, Spain. PIE on the three solid pellet rods with pressure measurements, to replicate the examinations performed on IFA-636. The PIE includes profilometry, axial gamma scan, puncturing, ceramography, chemical burnup analysis and density measurements. HWR-1038; HWR in preparation. Complete HWR.

11 - 3 - HP-1487 vol. 1 IRRADIATION OF VVER GADOLINIA FUEL (IFA-676) Commercial Gd-doped VVER fuel with 5 wt% absorbing isotopes and standard enrichment is being irradiated in IFA-676 to investigate the effect of the burnable poison on the fuel behaviour at BOL and with burnup. In July 2015 the rig was moved to a higher flux position in the HBWR core to increase the burnup accumulation rate. The rod power was increased to 8-10 kw/m. The measured fuel temperatures suggest that the fuel-cladding gap is still about mm. The rod burnups were increased to about 33 MWd/kg oxide. The measured fuel temperatures were in the range C. The fuel temperatures normalised to 8 kw/m suggest stable thermal performance, with the fuel swelling (leading to cladding-fuel gap closure) being compensated by the fuel conductivity degradation. Power history and fuel temperatures (measured and normalised to 8 kw/m) for rod 5 Main ref. Remaining work Two rods with commercial 5 wt% Gd-doped VVER fuel have been irradiated in IFA-676 together with two reference and two large grain VVER fuel rods. The rods are instrumented with fuel stack elongation detectors (EFs), cladding elongation detectors (ECs), expansion thermometers (ETs), pressure transducers (PFs), and pairs of fuel thermocouples (TFs). All test rods are clad with Zr-1% Nb and irradiated in the HBWR moderator. Status Report July-December 2015, HP-1485 Continue irradiation to > 35 MWd/kg oxide. Perform PIE and report.

12 - 4 - HP-1487 vol. 1 PIE ON VVER FUELS (IFA-676) Compare the behaviour of VVER-1000 fuel with aluminium silicate additive, which increases grain size to µm, with standard VVER fuel. The two test rods were unloaded from the reactor in 2013 after ~1300 FPD of irradiation, at a burnup of ~ 60 MWd/kg UO 2. Gamma scanning of the two rods was performed to obtain the isotopic activity distribution along the rods. The rod with silicate additive has released relatively more fission gas compared to the reference rod, i.e. ~17% and ~9% respectively. The additive fuel has also more caesium accumulated in pellet-pellet interfaces. Gamma scanning of Rod for Cs-134 and Cs-137 activities Main ref. Remaining work Two test rods: (i) Rod 676-4, VVER-1000 fuel with aluminium silicate additive, which increases grain size to µm; (ii) Rod 676-6, standard VVER fuel with about 11 µm grain size. Both test rods originally with 10% enrichment, clad with Zr-1% Nb and irradiated in the HBWR moderator. PIE includes neutron radiography of Rod to verify whether the rod EF is stuck, and on both rods: rod puncturing and fission gas analysis, gamma scanning, metallography and ceramography. Status Report July December 2015, HP-1485; HWR-1094, in preparation. Complete PIE and issue HWR.

13 - 5 - HP-1487 vol. 1 FGR MECHANISMS (IFA-716) To investigate fuel dimensional stability, thermal performance and fission gas release with variations in grain size and dopant concentration, including Cr 2 O 3 and a BeO dopant. Operation during the first half of 2015 continued at temperatures close to the Halden 1% FGR threshold, and some additional FGR was seen during the periods at highest power. The estimated FGR for the Cr-doped and large grain UO 2 rods is in the order % for the various rods. The temperatures in the rod with 3.0 wt% BeO have been consistently ~ C lower on account of the higher thermal conductivity of this fuel. No fission gas release has been seen for this rod. Due to deteriorating rig instrumentation for the pressure transducers, it was decided to unload the rig for PIE after the end of the reactor cycle in the spring of After PIE, four of the rods will be subjected to refabrication and further irradiation for fission gas release studies. The burnup at the time of unloading was about 32 MWd/kg Oxide. Evaluated fission gas release from the time of the power uprate in Dec Six rods in one cluster, each instrumented with PF, EF and TF: (i) two rods with standard UO 2 (provided by AREVA), one with normal and one with large sized grains; (ii) two rods doped with Cr (provided by AREVA), 0.16 and 0.1 wt%; (iii) one rod with large grain size and high density UO 2 (provided by ULBA); and (iv) one rod doped with BeO (provided by ULBA). Main ref. HWR-1155, to be issued Remaining work Irradiation is complete and the rods have been shipped to the Kjeller hotcells for PIE. PIE will include axial gamma scanning, rod puncturing and fission gas release analysis, and ceramography. Four of the rods are planned to be refabricated and further irradiated in the Halden reactor for continued studies of fission gas release of Cr-doped and BeO doped fuel in comparison with large grain size UO 2 fuel.

14 - 6 - HP-1487 vol. 1 PIE ON GADOLINIA FUEL FROM THE FGR TEST IN IFA To study thermal performance, pellet-cladding mechanical interaction and fission gas release from high burnup 8% Gd doped UO 2 fuel irradiated in a commercial PWR. The in-core experiment was performed with six successive power ramps with about 36 hours hold time at each level The first power level corresponded to a measured fuel centreline temperature of 700 C, and the last level to about 950 C. FGR was observed at 1210 C (peak temperature). The total fission gas release at the end of the test (70 operation days) was~6.5%. The test rods are undergoing PIE at the Kjeller hot lab. Visual inspection and rod puncturing / fission gas analysis are completed. Visual inspection of IFA-720.2, rod 1 at 0 degree orientation Two re-fabricated PWR fuel rods with 8% gadolinia were tested in IFA The father rod was irradiated for 4 cycles in the Vandellos II reactor in Spain to burn-up to 54 MWd/kgU (47.6 MWd/kgOx). Both test rods were fitted with a fuel centreline thermocouple at the upper end. At the lower end, one rod was fitted with a bellows pressure transducer and the other with a clad extensometer. The test assembly was fitted with a He-3 coil for power control. The in-core test was operated under PWR thermal-hydraulic and water chemistry conditions. Main ref. HP-1485, Vol. 1, Status Report July December 2015 Remaining work Complete PIE (neutron radiography, gamma scanning, profilometry and metallography / ceramography) and issue HWR-1154.

15 - 7 - HP-1487 vol. 1 OVERPRESSURE TESTING ON SR-M MDA CLADDING IN IFA To investigate the effect of overpressure on the thermal feedback behaviour of high burn-up fuel rods. The tests address a safety issue related to extended irradiation of high burnup fuel rods in which there is the potential of fission gas release. An RX M-MDA segment was tested in IFA , while an SR M-MDA segment is being tested in IFA Irradiation started in July 2014, and operated for ca. 2,300 full power hours to the end of Overpressure levels from +100 to +250 bar have been applied. The normalized fuel temperature increased and the measured elongation of the cladding decreased at the beginning of the test. These phenomena may be due to decreased PCMI as a result of the cladding creep-out. Above an overpressure of 200 bar, a further increase in the normalized fuel temperatures was observed. Due to a leak in the test rod, the irradiation was ended in April The cause of the leak, and an evaluation of whether repair and re-irradiation is feasible, will be determined from the initiated PIE work. Main ref. Remaining work The test rod is a part of a father rod with stress relieved M-MDA cladding and UO 2 fuel which was irradiated in the Vandellós-II PWR in Spain for four 18-month cycles up to ca. 65 MWd/kg U. The test rod was fabricated at Kjeller with a TF, an EC and two gas lines to enable inner pressure control of the test rod, flushing, hydraulic diameter measurement and selection of fill gas (argon or helium). The rig is connected to a loop system providing PWR thermal-hydraulic conditions. Status Report January June 2015, HP HWR-1158 (To be issued, April 2016) Issue HWR-1158 summarising the in-pile observations. Initiate PIE and evaluate the feasibility of re-fabrication and re-irradiation.

16 - 8 - HP-1487 vol. 1 FUEL THERMAL CONDUCTIVITY AND RECOVERY MECHANISMS (IFA-744) Investigate fuel thermal conductivity degradation and recovery mechanisms, in particular assess the significance of fission-density on thermal conductivity and study whether irradiation damage annealing and thermal conductivity recovery are also activated in-pile. The test comprises six rods, three of which contain liquid metal in the fuelclad gap to facilitate the data analysis by removing the uncertainty associated with gap conductance. The fuel temperature can be varied independently of the power. Four temperature increases were introduced during the first start-up in March 2015, and four more during the first month of operation at about 22 kw/m rod power as shown in the figure. IFA-744 irradiation history with temperature cycles An example of the response to a temperature cycle is shown to the right. After the temperature increase, the resistivity (inverse conductivity) goes down because irradiation damage is annealed. After the temperature reduction, the resistivity increases again because more irradiation damage is accumulated at the lower temperature. IFA-744 irradiation in the HBWR; on-line measurement of fuel centre and cladding temperature. Main ref. Status reports HP-1474 and HP 1485 Remaining work Summary of irradiation in 2015 (HWR-1162); continue irradiation in 2016.

17 - 9 - HP-1487 vol. 1 LOCA TEST IFA Assess the burnup threshold of fuel fragmentation and relocation using fuel with 65 MWd/kgU burnup which is between the 60 and 72 MWd/kgU burnup of previous tests with little and more pronounced fragmentation, resp. The important design parameters plenum volume and rod pressure were determined such that IFA , which showed little fragmentation, was mimicked in terms of pressure development, ballooning and time of burst as well as peak clad temperature (850 C). The test was instrumented with two cladding thermocouples (TC), a TC for measuring the plenum temperature, a fuel pressure sensor and a cladding elongation detector. The test was executed on Nov. 26, 2015, after irradiation for about 100 hours in PWR conditions and approximately at the EOL power in Ringhals 4. The development of rod pressure and temperature are shown in the figure. Main ref. Remaining work Development of pressure and temperatures during the LOCA transient The development of pressure and temperatures is similar to those seen in other tests. The semi-slow pressure drop after burst indicates that the gas flow from the plenum to the burst opening is restricted. No spray was applied during the transient. The test was terminated with a scram and left to cool down without reflood as usual in the IFA-650 series. Segment from rod 36U-N05 irradiated in Ringhals 4 (PWR), burn-up 65 MWd/kgU. Cladding material M5, fuel stack length 460 mm; plenum volume of 17 cm³ similar to IFA Status report HP-1485, HWRs 1163 and 1164 (in preparation) Post irradiation examination (profilometry, neutron radiography, fragment size determination)

18 HP-1487 vol. 1 GAMMA SCANNING FROM LOCA TEST IFA Continue the LOCA fuel fragmentation and relocation studies. Use gamma scanning to document the state of the segment shortly after unloading. After a few days of cooling, the LOCA rig was unloaded from the HBWR and moved to the gamma scanning compartment. During this operation, the segment is kept in a vertical position and handled gently in order not to change the fragment configuration caused by the transient. The gamma scans taken in 0 and 90 degrees orientation are shown in the figure. Main ref. Remaining work Gamma scan of IFA ; left: 0 orientation; right: 90 orientation The gamma scan shows a plug of fuel at the upper end (about one pellet length) as the likely cause of the relatively slow pressure drop registered by the in-core sensor after burst. A well-developed balloon has provided space for fuel to relocate into, and about four cm of cladding tube below the plug are void of fuel. Fuel dispersal (fuel at the flask bottom in the 0 degree scan) is not as pronounced as seen in and with similar gamma scan appearance, but higher burnup. In contrast to these previous tests, the ragged contour along the ballooned cladding indicates predominantly coarse fragments. Segment from rod 36U-N05 irradiated in Ringhals 4 (PWR), burn-up 65 MWd/kgU. Cladding material M5, fuel stack length 460 mm. HWR-1163, in preparation. Post irradiation examination (profilometry, neutron radiography, fragment size determination)

19 HP-1487 vol. 1 HIGH BURN-UP FUEL DISK IRRADIATION (IFA-655.2) To irradiate fuel disks to high burn-up at low temperature for subsequent investigation of the behaviour of fuel with high burn-up structure. The total burnup of the two UO 2 fuel rods at the end of December 2015 in IFA was around 160 MWd/kgUO 2 (180 MWd/kgHM). The test rig and rods have been unloaded from the reactor and will be sent to the Kjeller hot-cells for rod puncturing and refabrication in preparation for transient testing. Main ref. Remaining work Power and burnup history for the two rods in IFA The two test rods in IFA (Rod 3 and Rod 4) started irradiation in August 2001 in IFA-655.1, where they were loaded with ten other test rods and irradiated for 881 FPD to a burnup of 113 MWd/kgUO 2. After the other ten rods were unloaded, the irradiation continued as IFA from June 2008 onwards. Each fuel rod contains 25 fuel disks (~20% enriched UO 2 ), 1 mm thick, and constrained by 3 mm thick Mo disks on both sides. The density of the fuel in both rods is 95 %T.D. and the fabricated grain size of the fuel pellets in Rod 3 is 10 µm, while that in Rod 4 is 50 µm Status Report July December 2015, HP-1485 Both the test rods were equipped with fuel extensometers (EF) and pressure transducers (PF) from the start of operation. However the rig instrumentation has since experienced failure. Both the unloaded rods will be sent to the Kjeller hot-cells for rod puncturing and fission gas release determination. After this, the rods will be refabricated to facilitate transient testing in a He-3 type rig in the Halden reactor. The first of these tests is planned for the autumn in 2016.

20 HP-1487 vol. 1 FGR FROM HIGH BURNUP MOX DISKS (IFA-629.8) The main objective of the current IFA-629 test series is to study fission gas release behaviour of fuel irradiated to high burn-up representative of that in the rim region, i.e. with the high burn-up structure (HBS). Initially, the power was taken up to a level corresponding to a fuel temperature of ca. 550 C, which is approximately 50 C below the end of life irradiation temperature for IFA The power was kept at this level for about three days, before the transient test was initiated by first performing a small power adjustment to a calculated fuel temperature of 580 C. During the actual transient, the power was increased to a maximum calculated fuel disk centre temperature of 900 C. The transient was conducted over about 4.5 minutes, corresponding to a heat-up rate close to 1 C/s. The hold time at maximum temperature was ~5 hours, after which the test was terminated with a reactor scram. The first indication of fission gas release from the pressure sensor during the transient was at a temperature of C. During the transient, the rod pressure continued to increase at a near constant rate. No additional pressure increase was seen during the subsequent hold at maximum temperature. The calculated fission gas release at the end of the test was about 7%. Main ref. Rod internal pressure history and FGR during IFA IFA was performed with the heterogeneous MOX rod R from the disk irradiation test IFA Status Report July December 2015, HP-1485 Remaining work HWR-1159, To be issued (April 2016)

21 HP-1487 vol. 1 STEADY STATE AND TRANSIENT CLADDING CREEP (IFA-741) To study creep behaviour of modern fuel cladding alloys, and specifically to assess whether cladding creep is symmetrical under tensile and compressive loading and reversals, and whether mechanistic changes occur due to fast fluence effects on clad microstructure. The test rod has accumulated over 12,000 full power hours (FPH) of irradiation. In 2015 data were obtained at +30, -50, 0 and +30 MPa stress levels. In general, a diameter increase was observed as the stress level was increased. Diameter changes for E-110 segment Main ref. Remaining work Two test rods, each consisting of two fuelled segments. The upper test rod consists of new cladding tubes, E110-M and Optimized ZIRLO, while the lower test rod contains the M5 and M-MDA segments that were previously irradiated in IFA-699. Each test rod is instrumented with a gas line for hoop stress control and a diameter scanning gauge for OD measurement. The rig is connected to a loop system providing PWR thermal-hydraulic conditions. In April 2014 the diameter gauge for the lower test rod failed; the rod was unloaded and replaced with a dummy, and the irradiation was continued for the upper test rod as planned. The irradiation for the unloaded rod will be resumed after the irradiation of the upper rod is completed, and will use the single operational diameter gauge. Status Report July-December 2015, HP HWR Complete in-pile testing according to the specified stress history.

22 HP-1487 vol. 1 PWR CLADDING CORROSION TEST IFA-785 IFA-785 is a continuation of IFA-708 in a new test rig. The main objective is to study the in-pile corrosion and hydriding behaviour of modern Zircaloybased PWR cladding materials in aggressive water chemistry and thermal hydraulic conditions exceeding those currently allowable in operating PWRs. IFA-785 commenced irradiation in March 2015 and has accumulated approximately 150 full power days. Outlet maximum mass evaporation rates values ranged from ~ 3380 to ~ 3550 kg/m 2 /h; the linear heat rate in the segments was in the range of ~ 29 to ~ 34 kw/m. Water chemistry conditions are 10 ppm Li and 1580 ppm B, resulting in a ph 300 of 7.4, and 3 ppm H 2. The average burnup of the segments transferred from IFA-708 was around 32.8 MWd/kg oxide at the end of December 2015, while the burnup of the replaced segments varied between 8.2 to 9.6 MWd/kg oxide. Coolant temperatures and maximum mass evaporation rate Six test rods, each consisting of four fuel 120 mm-long segments. Three reference rods comprise commercially available claddings, and three rods contain developmental alloys. In IFA-785, the four unfailed rods from IFA-708 are used, while Segments 3 and 4 from Rod and Segments 2 and 3 from Rod have been replaced with new segments - consisting of fresh fuel and fresh cladding. Corrosion is being assessed by means of interim inspections, comprising photography and oxide thickness measurements. Main ref. HP-1485, Status Report July-December 2015 Remaining work Irradiate to assembly burn-up in excess of 40 MWd/kg oxide. Annual interim inspections; the next to occur in early 2016.

23 HP-1487 vol. 1 PIE ON THE COATED CLADDINGS TEST IFA-774 Investigate the corrosion behaviour of coatings, applied by Physical Vapour Deposition (PVD), on Zircaloy cladding. The un-coated Zr-4 reference had a porous oxide layer with many cracks, which indicates higher than target cladding surface temperatures. At least 80 % of the CrN coating remained intact, although the coating had local damages in terms of cracks and missing parts. Under the damaged coating there was observed corrosion in the Zry-4. Some pores were found in the coating, which likely originate from fabrication. Pores, notches and other irregularities are weak spots prone to crack initiation. The PIE results indicate that micro cracks have propagated through the coating with subsequent water ingress and Zry-4 corrosion. The Zry-4 corrosion results in a volumetric expansion which leads to further cracking of the coating. The TiAlN and AlCrN coatings dissolved. Cross section of rod with CrN coating showing local corrosion of Zr-4 under the coating Main ref. Remaining work Four test rods, each with 5 wt % enrichment and fuel stack length of 93 mm. The rods had an inner cladding of Inconel 600 and an outer cladding of Zry-4. One of the rods (Rod 1) was uncoated and the remaining three had 2-3 µm thick coatings of TiAlN (Rod 2), CrN (Rod 3) and AlCrN (Rod 4). The rods were exposed to PWR conditions (4.6 ppm Li / 1000 ppm and 2-3 ppm H 2 ). Status Report July-December 2015, HP HWR-1157 (in preparation). Cladding hydrogen measurements on segments from rod 1 (un-coated) and rod 3 (CrN-coated)

24 HP-1487 vol. 1 PWR CRACK GROWTH RATE TEST (IFA-772) The main objectives of the PWR crack growth rate (CGR) test IFA-772 were to generate long-term CGR data for irradiated Compact Tension (CT) specimens in PWR conditions with varying Li/B ratios and at high and low H 2 concentrations. During irradiation the specimens were exposed to 0.5 ppm Li/0 B and ~2 ppm H 2 and a coolant temperature of ~320 C. Crack growth rate data were obtained for 5 of the 6 CTs (no data were obtained for the 6.2 dpa 304 SS CT). The applied stress intensity (K) levels on the specimens ranged from ~10 to 30 MPa m and growth rates ranged from 10-8 to 10-5 mm/s. The highest crack growth rates were recorded for the ~6-10 dpa CW 316 SS specimen. The crack growth rates for the two 5.2 dpa samples were similar and lay on the BWR-VIP 99 disposition curve for hydrogen water chemistry (HWC). Comparison of CGR data measured for 5.2-dpa 321 SS (CT2 and CT5) in IFA BWRVIP-99 disposition curves for NWC and HWC are also shown. Main ref. Remaining work IFA-772 contained six CT specimens prepared from irradiated stainless steels from commercial reactors. Two CTs were prepared from 5.2 dpa 321 SS. The remaining CTs were prepared from 5.9 dpa 304L SS, 6.2 dpa 304 SS, ~6-10 dpa CW 316 SS ~4 dpa CW 316 Ti SS. The CTs were instrumented for crack growth monitoring with the dc potential drop (dcpd) method and equipped with bellows for load application. HWR-1140, Minutes of Halden IASCC Review Meeting November HWR-1150, in preparation.

25 HP-1487 vol. 1 CRACK INITIATION STUDY (IFA-733) To develop a protocol for crack initiation testing and evaluate the effectiveness of HWC in mitigating the initiation of cracks in irradiated material by comparing the number of failures occurring in tensile specimens in NWC and HWC. Irradiation of IFA-733 began in July 2011 and in total ten specimen failures have been recorded. Nine failures occurred at loads of % yield strength (YS). One failure occurred at ~95 % YS. Fracture surfaces of the failed specimens have been inspected and intergranular fracture was observed only on one specimen. Eight (8) specimens showed ductile fracture in the gauge region. One specimen failed in one of the loading holes and was therefore not subjected to PIE. Main ref. Remaining work Example of the fracture surface of a failed specimen. Typical ductile tensile fracture; shear lips, dimples and heavily deformed flat crack surface and necking at the gauge area are observed. Eighteen miniature tensile specimens prepared from 304L SS with a dose of 13 dpa. Nine of the specimens were transferred from the previous integrated time-to-failure study, IFA-660. Load (originally low load (80 %) and high load (100%) of the 718 MPa irradiated yield strength of the material) is applied by means of system pressure acting on the outside of bellows attached to the upper end of the specimens. The specimens, which are equipped with LVDTs to monitor failures on-line, are exposed to BWR conditions with 5 ppm O 2 ( NWC ). HWR-1140, Minutes of Halden IASCC Review Meeting November Testing of the remaining nine low load specimens will continue until mid

26 HP-1487 vol. 1 CREEP OF ODS STEELS IN IFA-744 The main objective is to measure and compare the creep of two specimens prepared from oxide dispersion strengthened (ODS) steel. Irradiation of IFA-744 began in March 2015 and the deformation recorded for the two specimens during 2015 (up to ~0.057 dpa) is shown in Figures 1 and 2. The creep of PM2000 was found to be negligible while the strain rate of the 12%CrODS was on average 1.7x 10-6 /h. Stress, temperature and elongation history for ODS specimen prepared from PM2000 material (top) and the Kobelco material (bottom). Main ref. - Creep is being measured on two samples prepared from ODS materials (one from Plansee (PM2000) and the other from Kobelco (12%CrODS)). The samples, in the form of tensile specimens, are being irradiated in an inert environment in test units that allow on-line monitoring of specimen elongation by means of LVDTs and temperature and applied stress on the specimens are controlled by means of gas lines connected to an external system. The specimens are being exposed to a temperature of 400 C and a stress level of 350 MPa. Remaining work Irradiation of the ODS creep specimens in IFA-744 will continue in 2016.

27 HP-1487 vol. 1 REPORTING Author Title HWR-1098 H. Jenssen Post Irradiation Examination of IFA-708 fuel rods HWR-1141 W. Wiesenack Monte Carlo uncertainty and sensitivity analysis of the power determination in a Halden test rig containing multiple rods HWR-1147 W. Wiesenack Minutes of the HRP-WGFS workshop on fuel fragmentation, relocation and dispersal HWR-1189 V. Tulkki Analysis and modelling of Halden cladding creep experiments Paper presented outside the Halden series Applicability of a set of tomographic reconstruction algorithms for quantitative SPECT on irradiated nuclear fuel assemblies. Staffan Jacobsson Svärd, Scott Holcombe, Sophie Grape. Nuclear Instrumentation and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, vol. 783, pp , May 2015 Tomographic determination of spent fuel assembly pin-wise burnup and cooling time for detection of anomalies. Staffan Jacobsson Svärd, Peter Andersson, Anna Davour, Sophie Grape, Scott Holcombe, Peter Jansson. ESARDA Symposium, Manchester, U.K., May, 2015 In-pile Crack Growth Rate Testing of Irradiated 304L and 316L Stainless Steels in PWR and BWR Environments T.M. Karlsen, M. Helin. J. Nakano. 17 th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors, August , Ottawa, Canada. In-pile experimental investigations of fuels and materials degradation in the Halden reactor. B.Yu. Volkov, M.A. McGrath, T. Tverberg. International Conference on WWER Fuel Performance, Modelling and Support, 11, September 26 October , Varna, Bulgaria. Modeling of axial distribution of released fission gas in KKL BWR fuel rods during base irradiation. V. Brankov, G. Khvostov, K. Mikityuk, A. Pautz, R. Restani, S. Abolhassani, G. Ledergerber, W. Wiesenack, September Test Reactor Evaluation of Zirconium Alloy Shadow Corrosion Behavior. D. Lutz, Y-P. Lin, P. Cantonwine, J. Varela, A. Kucuk, K. Edsinger, M. McGrath TopFuel, September , Zurich, Switzerland. from In-Pile Creep Testing of High Purity Polycrystalline Sic and Select Fecral Alloys K Terrani, Y Katoh, Y Yamamoto, L Snead and T M Karlsen TopFuel, September , Zurich, Switzerland. CSNI Working Group on Fuel Safety (WGFS): an Overview. Wolfgang Wiesenack Meeting of the IAEA Technical Working Group on Fuel Performance and Technology Vienna, Austria, April 2015

28 Selected results from HRP IFA-650 LOCA experiments Wolfgang Wiesenack American Nuclear Society Winter Meeting Washington DC, USA, November HP-1487 vol. 1 Investigation of the impact of coatings on the corrosion of nuclear components K Daub, R Van Nieuwenhove and H Nordin 17 th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Water, Ontario, Canada, August 9-12, 2015 Development of oxygen sensor for use in liquid metal Rudi Van Nieuwenhove ANIMMA, Lisbon, Portugal, April 20-24, 2015 Irradiation capabilities at the Halden reactor and testing possibilities under supercritical water conditions Rudi Van Nieuwenhove ISSCWR-7, Helsinki, Finland, March 15-18, 2015 Status of testing of commercially available PVD coatings at the Halden Reactor Project Rudi Van Nieuwenhove NEA (Nuclear Energy Agency) ATF (Accident Tolerant Fuel) meeting Paris, France, March 3-5, 2015 Corrosion protection for steel and zircaloy by PVD coatings and instrument developments for Gen-4 reactors Rudi Van Nieuwenhove JPNM (Joint Programme of Nuclear Materials) EERA (European Energy Research Alliance) Madrid, Spain, February 1-4, 2015