Fuels and Materials Programme Achievements 2012

Transcription

1 Institutt for energiteknikk OECD HALDEN REACTOR PROJECT HP-1378 vol. 1 For use within the Halden Project Member Organisations only OECD HALDEN REACTOR PROJECT Fuels and Materials Programme Achievements 2012 February 2013

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3 FOREWORD The experimental operation of the Halden Boiling Water Reactor, the Halden Man-Machine Laboratory, HAMMLAB, and the Virtual Reality Laboratory and the associated research programmes are sponsored through an international agreement for the period by the Institutt for energiteknikk (IFE), Norway, the Belgian Nuclear Research Centre SCKCEN, acting also on behalf of other public or private organisations in Belgium, the Technical University of Denmark, the Finnish Ministry of Employment and the Economy (TYÖ), the Electricité de France (EDF), the Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbh, representing a German group of companies working in agreement with the German Federal Ministry for Economics and Technology, the Japan Nuclear Energy Safety Organization (JNES), the Korean Atomic Energy Research Institute (KAERI), acting also on behalf of other public or private organisations in Korea, 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 public and private nuclear organisations in Sweden, the Swiss Federal Nuclear Safety Inspectorate ENSI, representing also the Swiss nuclear utilities (Swissnuclear) and the Paul Scherrer Institute, (PSI) the National Nuclear Laboratory (NNL), representing a group of nuclear licensing and industry organisations in the United Kingdom, and the United States Nuclear Regulatory Commission (USNRC), and as associated parties: the Czech Nuclear Research Institute (NRI), EU JRC Institute for Transuranium Elements, Karlsruhe, the French Institut de Radioprotection et de Sûreté Nucléaire (IRSN), Commissariat à l energie atomique (CEA), France, the Hungarian Academy of Sciences, KFKI Atomic Energy Research Institute, the Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA), Japan Atomic Energy Agency (JAEA), the Central Research Institute of Electric Power Industry (CRIEPI), representing a group of nuclear research and industry organisations in Japan, the Mitsubishi Nuclear Fuel Co., Ltd. (MNF), Japan, the Ulba Metallurgical Plant JSC in Kazakhstan, JSC TVEL and RRC Kurchatov Institute, Russia, Public Joint Stock Company All-Russian Scientific and Research Institute for Nuclear Power Plant Operation (VNIIAES), Russia, and the Slovakian VUJE - Nuclear Power Plant Research Institute, the Electric Power Research Institute (EPRI), the Global Nuclear Fuel (GNF) Americas, LLC and GE-Hitachi Nuclear Energy, LLC, the US Department of Energy (DOE), and the Westinghouse Electric Power Company, LLC (WEC) 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 programme items of special interest.

4 ABSTRACT This report is intended to summarise the accomplishments of the Fuels and Materials research programme of the Halden Reactor Project during 2012, 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 2012 The OECD Halden Reactor Project started the new agreement period ( ) with nineteen member countries. This number was two up from 17 in the previous period because Kazakhstan joined the Project from the beginning of 2009 and Italy during The member countries comprise Belgium, Czech Republic, Denmark, Finland, France, Germany, Hungary, Italy, Japan, Kazakhstan, Korea, Norway, Russia, Slovak Republic, Spain, Sweden, Switzerland, the United Kingdom, and the Unites States. Furthermore, important nuclear organisations such as VNIIAES, Russia; DOE, USA and CEA, France joined the Project as Associated Parties during the program period, making a total of about 60 directly participating organisations from nuclear regulatory bodies, utilities, vendors and research centres. 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 2012 was: 2012: Keijo Valtonen, Finnish Radiation and Nuclear Safety Authority STUK The chairman of the Halden Programme Group in 2012 was: 2012: Virginie Colombier, Electricité de France EDF 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 at Storefjell in March 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. One such summer school was conducted in 2012 with the following subject: 2012: Software Systems Dependability, August 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 4 workshops were arranged in 2012: 2012: LOCA workshop, Lyon, May IASCC workshop, Halden, October Halden On-line Monitoring User Group, Italy, October Workshop on Data needs and testing ideas for demanding operation conditions and innovative fuels and claddings, Halden, October 29

6 Executive summary of Fuels and Materials achievements The Fuels & Materials programme was defined and executed under three main chapters: Fuel Safety and Operational Margins Plant Ageing and Degradation Contribution to International Gen-IV Research On average, test rigs were under irradiation at any one time as part of the Halden Reactor Project Joint Programme, with a total of 16 unique in-pile experiments being performed during Reactor availability throughout 2012 was 50%. Fuel safety and operational margins is related to fuels 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 with addition of BeO. The objective is to provide fuel property data for design and licensing from zero to MWd/kg, including commercially irradiated fuels. Research activities focus on: Gas release behaviour from fuel under normal irradiation conditions Fuel thermo-mechanical behaviour under normal irradiation conditions Fuel behaviour under accident scenarios (LOCA) Fuel behaviour under demanding operation conditions A highlight from these studies is: Creep of UO 2 fuel under irradiation is being studied in an experiment aimed at generating data for improved modelling of fuel pellet periphery behaviour during PCMI. Standard UO 2 and Cr-doped pellets are subjected to a range of stresses and temperatures consistent with pellet periphery conditions, while dimensional change in the pellets is recorded. During 2012, a second set of creep data were obtained at a burn-up of 18 MWd/kg UO 2 and the behaviour observed earlier at a burn-up of 8 MWd/kg UO 2 continued to be shown, namely that the creep rate increased linearly with stress and fission rate, and was independent of temperature, consistent with data from open literature. From the two data sets, it was also observed that there was no burn-up effect (in the range investigated) and that in the temperature range investigated the Cr-doped fuel behaved similarly to the UO 2 fuel. 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: The long-term creep and stress relaxation study of materials used in PWR and BWR plants is continuing to provide data for component lifetime assessments. Some materials are showing more creep/stress relaxation resistance than others, which could lead to improved alloy selection in future component designs. These data are unique by being obtained online from samples irradiated under a neutron flux prototypic of commercial nuclear plants rather than under accelerated conditions. The HRP aims to contribute to international Gen-IV research by developing instruments able to withstand GEN-IV reactor concept conditions as well as investigating the efficacy of coatings for corrosion resistance in such environments. A highlight from these studies is: A prototype instrument for monitoring crack growth on a CT specimen by measuring crack mouth opening displacement has been developed and tested in-pile. This method could be used in a highly conducting coolant such as liquid lead for which the potential drop method for crack growth monitoring could not be used.

7 HP-1378 vol. 1 CONTENTS Page OVERVIEW OF FUELS AND MATERIALS EXPERIMENTS DURING PIE ON TEST ROD FROM THE HIGH BURNUP FUEL DISKS FGR TEST (IFA-629.5)... 2 OECD Halden Reactor Project FGR FROM HIGH BURNUP FUEL DISKS (IFA-629.6)... 3 PWR OVERPRESSURE/LIFT-OFF EXPERIMENT (IFA )... 4 FISSION GAS RELEASE MECHANISMS (IFA-716)... 5 FUEL CREEP TEST (IFA-701) (1)... 6 FUEL CREEP TEST (IFA-701) (2)... 7 COMPARATIVE INTEGRAL IRRADIATION TEST ON GD FUEL (IFA-681)... 8 GD-DOPED VVER FUEL BEHAVIOUR (IFA-676.1) (I)... 9 LARGE GRAIN VVER FUEL BEHAVIOUR (IFA-676.1) (II) HALDEN PROJECT PROGRAMME ACHIEVEMENTS 2012 BWR LOCA TEST (IFA ) (IN-PILE) LOCA TEST (IFA ) (GAMMA SCANNING) BWR LOCA TEST (IFA ) (IODINE AND CESIUM RELEASE) BWR LOCA TEST (IFA ) (PIE) PWR CLADDING CREEP (IFA-741) ND INTERIM EXAMINATION OF RODS FROM THE PWR CORROSION TEST (IFA-708) ON-LINE, IN-CORE CLADDING CORROSION (IFA-731) TEM ANALYSIS OF CLADDING SAMPLES (FROM IFA-638) BWR CRACK GROWTH RATE TEST (IFA-745) CRACK INITIATION STUDY (IFA-733) CHARACTERISATION OF IASCC TEST MATERIALS (I) CHARACTERISATION OF IASCC TEST MATERIALS (II) IRRADIATION CREEP AND STRESS RELAXATION STUDY (IFA-669.2) February 2013 PRESSURE VESSEL AGING (SMALL PUNCH TEST) INSTRUMENT DEVELOPMENT IASCC REVIEW MEETING OCTOBER SUMMARY OF WORKSHOP ON DATA NEEDS AND TESTING IDEAS FOR DEMANDING OPERATION CONDITIONS AND INNOVATIVE FUELS AND CLADDINGS, 29TH OCTOBER 2012, HALDEN REPORTING 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 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 IFA fuel (f) clad (c) or material type and origin HP-1378 vol. 1 OVERVIEW OF FUELS AND MATERIALS EXPERIMENTS DURING 2012 MEASUREMENTS APPLICATIONS # of rods / specimens burnup MWd/ kg oxide fluence n/cm 2 (x10 20 ) Recent reports Comment UO 2 /PWR 1 62 x x x x x x x HP-1366, HP-1377 PWR Overpressure UO 2 disks x x x x HWR-1010, HP-1366 FGR ramp UO 2 disks x x x x HP-1377 FGR ramp BWR /irradiation 1 64 x x x x x x x HP-1366, HPR-374, HWR-1009 LOCA test PIE BWR/irradiation 1 64 x x x x x x x HP-1377, HWR-1042 LOCA test UO x x x HP-1366, HP-1377, HPR-371 Disks, rim 669 SS/Inconel 30 - x HP-1377, HWR-1047 Stress relaxation 676 VVER/UO x x x x x x x x HWR- 1007, HP-1366, HP-1377 Base irradiation 681 Gd/UO x x x x x x x x HPR-374, HWR-1038 Gd-fuel 701 UO 2 / Cr-doped 4 14 x x x x HP-1377, HWR-1039 Fuel creep 708 M5, Zirlo, M-MDA-SR 6 17 x x x HP-1366, HP-1377, HWR-1045 Cladding corrosion 716 UO 2 / fresh 6 5 x x x x x x HWR-1008, HP-1377 High LHR, FGR 720 VVER fuel 1 50 x x x x HWR-1011, 1012 Ramp PIE 731 Zr-2, Zr-4 5 x x HP-1377, HWR-1046 On-line corrosion L dpa x x HPR- 377 BWR crack init. 741 E110, M5, M-MDA, Opt. zirlo 4 x x x x HP-1377 Cladding creep SS, 304L SS 7 HP-1377 BWR crack growth SPT SS 210 HP-1366 RPV testing D06 CGR specimens

10 - 2 - HP-1378 vol. 1 PIE ON TEST ROD FROM THE HIGH BURNUP FUEL DISKS FGR TEST (IFA-629.5) 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. In IFA-629.5, onset of fission gas release was detected at the temperature step from 710 to 770 C. The measured pressure at the end of the temperature ramp was 11.4 bar (at 20 C) corresponding to a fission gas release of 12.5 %. The rod was then transported to Kjeller for PIE. The rod has been subjected to rod puncturing, measuring the pressure, free volume and gas composition. The puncture pressure was in good agreement with what was measured in-pile at the end of the test (11.3 bar puncture pressure vs bar in-pile at 20 C). Lower Cluster rodlets in IFA-655 Rod Number Fuel Type UO 2 MOX UO 2 Grain Size / Homogeneity Medium Large Hom. Het. Mediu m Large Constraining Rings x x x x x x Fuel Extensometer x x x x x x Pressure Transducer x x x x x x Burnup [MWd/kgOx] Status Tested in IFA Tested in IFA Re-fabricated for loading in IFA-629 Burnup extension in IFA Test matrix for current IFA-629 test series IFA was performed with the standard grain size UO 2 rod R from the disk irradiation test IFA-655. A step-wise power up-rate was performed from ~600 C to ~1025 C. Main ref. Status Report January June 2012, HP-1366, Vol. 1 HWR-1010, September 2011 Mass spectrometry will be performed on a gas sample extracted during puncturing to determine the gas composition. Ceramography will be performed on selected disks from the rod and compared against twin disks from the upper cluster irradiation in IFA-655 that have not undergone any temperature ramp in the IFA-629 rig. Chemical burnup analysis of one UO 2 and one MOX disk from these rods is also planned.

11 - 3 - HP-1378 vol. 1 FGR FROM HIGH BURNUP FUEL DISKS (IFA-629.6) s 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. Initially, the power was taken up to a level corresponding to a fuel temperature of ca. 650 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 700 C. During the actual transient, the power was increased to a maximum calculated fuel disk centre temperature of 1120 C. The transient was conducted over 7.5 minutes, corresponding to a heat-up rate of close 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. During the subsequent hold at maximum temperature, the gauge indicated a continued pressure increase at a continuously decreasing rate. The calculated fission gas release at the end of the test was about 13%. Rod internal pressure history and FGR during IFA-629.6s IFA was performed with the large grain size UO 2 rod R from the disk irradiation test IFA-655. Main ref. Status Report July December 2012, HP-1377, Vol. 1 HWR-1041, in preparation Mass spectrometry will be performed on a gas sample extracted during puncturing to determine the gas composition. PIE similar to that being performed on the earlier test IFA (see previous page)

12 - 4 - HP-1378 vol. 1 PWR OVERPRESSURE/LIFT-OFF EXPERIMENT (IFA ) The objective of the overpressure test series IFA-610 is to provide data for understanding the behaviour of recrystallised (RX) vs. recrystallized stressrelieved (RX-SR) cladding under internal overpressure, and for determining cladding lift-off conditions. In IFA a RX M-MDA segment is being tested. An SR-RX M-MDA segment will be tested in the subsequent loading. The assembly commenced irradiation in June However, shortly afterwards, loss of rig instrumentation necessary for analysis of the experiment required that the assembly was unloaded from the reactor. The test rod was reloaded in a new test in December As of the end of December 2012, the rod overpressure reached +125 bar and no indications of lift-off were observed. History of fuel temperature, power and overpressure A UO 2 fuel segment with recrystallized M-MDA cladding was irradiated for four 18-month cycles up to burnup of 69.8 MWd/kgU in the Vandellós-II PWR in Spain. The rod instrumentation includes a clad extensometer (EC) and a fuel thermocouple (TF). Two gas lines are mounted at each end of the rod and connected to the two gas systems to control rod pressure and gas type. The first system is used for rod flushing and hydraulic diameter measurements whereas the second one is used for the rod pressurisation with argon or helium. Main ref. Status report July December 2012, HP-1377, Vol. 1. The overpressure testing will follow the agreed test scheme, in which the overpressure is increased step wise up to +300 bar (max). The holding time at each overpressure level is set to be about 3 weeks, and the duration of the experiment will be two reactor cycles of approximately 90 FPD each.

13 - 5 - HP-1378 vol. 1 FISSION GAS RELEASE MECHANISMS (IFA-716) To investigate fuel thermal performance and fission gas release with variations in grain size and dopant concentration, including the effect of Cr 2 O 3 dopant concentration and UO 2 grain size on densification, thermal behaviour and fission gas release; and the effect of BeO dopant on thermal conductivity. The burnup in December 2012 was about ~17 MWd/kg Oxide. Up to December 2012, the rods have operated at average linear heat rates from 25 to 30 kw/m with measured temperatures from 1100 to 1200 C and peak fuel temperatures ~ C below the Halden 1% fission gas release threshold. The temperatures measured in the rod with 3.0% BeO are ~100 C lower on account of the higher conductivity of the dopant. The two doped Cr fuel variants show the same behaviour, with limited densification until a burnup of about 1 MWd/kg Oxide and a steady increase with solid fission product swelling after this. For the two UO 2 fuel rods the densification is counterbalanced by the solid swelling at a burnup of between 3 and 5 MWd/kgUO 2.For the 3% BeO rod, densification is complete at a burnup of ~3.5 MWd/kgOxide. The swelling rate for this fuel is comparable to that of the other fuel types. Linear heat rates and peak temperatures Six rods in one cluster, each instrumented with PF, EF and TF: (i) two rods with standard UO 2 (provided by AREVA), one each with normal and large grains; (ii) two rods doped with Cr (provided by AREVA), 0.16 and 0.1 %; (iii) one rod with large grain and high density UO 2 (provided by ULBA); and (iv) one rod doped with BeO (provided by ULBA). Main ref. Status Report July December 2012, HP-1377, Vol. 1. A slow power uprate started in December 2012 in order investigate the FGR of the different fuel types as they cross the FGR threshold. Irradiation is planned to continue to an average burnup of 30 MWd/kg Oxide, after which the rods will undergo PIE.

14 - 6 - HP-1378 vol. 1 FUEL CREEP TEST (IFA-701) (1) IFA-701 is designed to investigate fission induced creep in fuel, which is observed in-reactor at temperatures below 1000 C. Two creep testing periods have been conducted, at burn-ups of ca 8 and 18 MWd/kg UO 2. In each period, the planned target fuel temperatures were 400, 600 and 800 C, and the planned applied stress levels were 30, 45 and 60 MPa during each temperature period. In both periods, the creep rate showed a linear increase with stress but was independent of temperature. Fission rate dependency was almost linear, but a burnup effect was not observed. The creep rate of the Cr-doped pellets was almost the same as that of standard UO 2 pellets. The fuel creep rates were similar to literature data. Comparison of creep rate with literature data Main ref. Two fuel types are being tested in IFA-701.1: standard UO 2 and Cr-doped fuel. For each fuel type, a test rod and a reference rod are loaded. All rods are manufactured from 44 fuel disks and 45 molybdenum (Mo) disks, fitted with fuel centre line thermocouples (TF), fuel stack elongation detectors (EF) and a loading device which generates the compressive stress in the fuel. All rods are also connected to a gas line which can change the gas ratio and pressure in the rod, and to bellows in order to control fuel temperature and the compressive stress. Status report July December 2012, HP-1377, Vol. 1. HWR-1039, in preparation One failed rod has been replaced by a dummy. A third set of creep measurements will be performed in mid-2013.

15 - 7 - HP-1378 vol. 1 FUEL CREEP TEST (IFA-701) (2) IFA-701 is designed to investigate fission induced creep in fuel, which is observed in-reactor at temperatures below 1000 C. In the second creep testing period, the creep data obtained at temperatures below 800 C were similar to those obtained in the first creep test period. However, the data obtained at 800 C were unreliable due to coolant leaking into Rod 701-3, UO 2 test rod. Therefore, the failed rod was replaced by a dummy and the test was continued with the remaining rods for evaluating the effect of burnup on creep behaviour. Two irradiation cycles are planned: the first for burnup accumulation (started in November 2012), and the second for a further creep test. Fuel temperature and LHR history of IFA Main ref. Two fuel types are being tested in IFA-701.1: standard UO 2 and Cr-doped fuel. For each fuel type, a test rod and a reference rod are loaded. All rods are manufactured from 44 fuel disks and 45 molybdenum (Mo) disks, fitted with fuel centre line thermocouples (TF), fuel stack elongation detectors (EF) and a loading device which generates the compressive stress in the fuel. All rods are also connected to a gas line which can change the gas ratio and pressure in the rod, and to bellows in order to control fuel temperature and the compressive stress. Status report July December 2012, HP-1377, Vol. 1. HWR-1039, in preparation A third set of creep measurements will be performed in mid-2013.

16 - 8 - HP-1378 vol. 1 COMPARATIVE INTEGRAL IRRADIATION TEST ON GD FUEL (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 rods operating with peak fuel temperatures close to or above the empirical FGR threshold curve showed gradual but moderate pressure increases. A power uprate at the end of the fourth cycle brought about small stepwise FGR increases in two rods (UO 2 rod 1 and 2% Gd rod 2). Unlike the UO 2 rods, the Gd-rods have shown little or no densification, while swelling rates are similar for both types of fuel. Strong PCMI has been absent throughout the entire irradiation, except during a higher power period in cycle 4. In 2011 the rig was moved to a higher flux position in the core, but the power increase was limited. A new power calibration was carried out in this new core position, with an agreement with the HELIOS predicted power at this burnup of ~10%. Since rod instrumentation had deteriorated over the ~7 years of in-pile service, it was decided to unload the rig in 2012 and transfer the rods to Kjeller for PIE. The discharge burnup was ~47 MWd/kgOx for the UO 2 and 2% Gd rods, and ~34 MWd/kgOx for the 8% Gd rods. Normalised fuel elongation at hot standby Main ref. 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. F Khattout, HWR-1038, in preparation. PIE at Kjeller on the three solid pellet rods with pressure measurements, which will replicate the examinations performed in the rods from IFA-636. Scope of PIE will include profilometry, axial gamma scan, puncturing, ceramography, chemical burnup analysis and density measurements.

17 - 9 - HP-1378 vol. 1 GD-DOPED VVER FUEL BEHAVIOUR (IFA-676.1) (I) Commercial Gd-doped VVER fuel with 5 wt% absorbing isotopes and standard enrichment is being tested to investigate the effect of the burnable poison on the fuel behaviour at BOL and with burnup. At BOL the thermal behaviour of Gd-doped fuel was influenced by the neutron flux depression, which changes with burning up of the absorbing isotopes. The fuel temperature measured as a function of heat rating indicated lower thermal conductivity of the Gd-doped fuel in comparison with reference VVER fuel. After six years of irradiation, the current burnup of the Gd-doped fuels is approximately 25 MWd/kg oxide. Fuel elongation measurements indicated no densification in Gd-doped VVER fuel at BOL. Fuel swelling, at a rate of about 0.5 vol% per 10 MWd/kg oxide, was detected from the beginning of the irradiation. No specific features related to Gd fuel behaviour were observed from the cladding elongations, which showed irradiation growth and soft PCMI. Gas pressure measured in the rods indicated slight fuel densification at BOL and subsequent swelling during further irradiation. Main ref. Two rods with commercial Gd-doped VVER fuel consisting of 5 wt% of the absorbing isotopes have been irradiated in IFA 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). S. Koike and B. Volkov, Progress report on the irradiation of VVER large grain and Gd-doped fuel in IFA-676.1, HWR-1007, The irradiation will continue to a burnup of MWd/kg oxide.

18 HP-1378 vol. 1 LARGE GRAIN VVER FUEL BEHAVIOUR (IFA-676.1) (II) The main objective of the test is to study the behavior of VVER-1000 fuel with additives (aluminium silicate) that enhance grain size to µm. Two rods with large grain VVER-1000 fuel are being tested in comparison with reference VVER fuel. There is no indication of any substantial effect of the large grains on thermal stability or fuel thermal behaviour. However, the large grain fuel exhibits better axial dimensional stability at power, which may be attributed to a higher creep resistance of this type of fuel. After six years of irradiation, the current burnup of the UO 2 fuels is ~ 57 MWd/kg oxide. IFA-676 was initially loaded in reactor position (5-15) but in order to follow the power history specified by JSC TVEL, the rig was moved to a lower flux position in June In that period, the measured gas pressure increased and the fuel temperatures exceeded the FGR threshold for the UO 2 rods. A higher level of FGR was detected in the large grain fuel due to higher rod power and corresponding fuel center temperature. The subsequent power reduction allowed FGR to be suppressed. A second pressure increase was observed in both fuels at a burnup of MWd/kg oxide, despite the fuel temperatures being below the thermal FGR threshold. See Achievements Report IFA (I) Main ref. S. Koike and B. Volkov, Progress report on the irradiation of VVER large grain and Gd-doped fuel in IFA-676.1, HWR-1007, The irradiation is continuing to study FGR and PCMI to the target burnup of 60 MWd/kg oxide. Rod 4 (large grain) and Rod 6 (reference) will then be sent for PIE to investigate FGR and to examine the fuel stack in the top of the rods. Rod 1 (large grain) and Rod 3 (reference) will be moved to a higher reactor position in order to perform a power uprating to study FGR behaviour.

19 HP-1378 vol. 1 BWR LOCA TEST (IFA ) (IN-PILE) To study the behaviour of a high burn-up BWR rod during a simulated LOCA temperature transient. IFA was intended to be a cladding failure test. The target peak cladding temperature (PCT) was C. The test rod was a BWR segment (AEB072-E4-D), very similar to the one in IFA , pre-irradiated in the KKL plant (Switzerland) for seven cycles to a segment average burn-up of 72 MWd/kg U. Cladding type was LK3/L (with a 70 µm liner). Oxide thickness was ~ 20 µm and hydrogen content about 300 wppm. The rod had been refabricated with a normal plenum volume of 15 cc. The 38 cm long rod was pressurised to 20 bar (RT) and fitted with standard instrumentation. The standard single-sided blow-down procedure was adopted and no spray was used. Before the test the rod had been irradiated at W/cm for about 3.5 days. The test was terminated after 6 minutess, counted from the end of the blow-down phase. The three cladding TCs showed fairly similar temperatures during the transient. As in IFA , the two upper TCs were positioned somewhat lower down (closer to the power peak) due to the shorter rod length. The rod pressure increased to 42 bar during the temperature transient (35 bar before the blowdown) and dropped by some 7-8 bar in the ballooning phase prior to clad burst, which resulted in a rapid drop. Judging from the pressure measurements significant cladding creep deformation and ballooning occurred above C. The rod failed at a clad temperature of C, close to predicted values by PSI/KKL. The maximum measured clad temperature was 860 C. The gamma monitor showed release of activity around 20 seconds after rod failure. Measurements of the release of iodine (I-131) and cesium (Cs-137) were made using improved methods Main ref. Rod pressure (PF1), cladding temperature (TCs1-3), rod elongation measurements(ec2) and gamma monitor response to cladding burst. In IFA-650 one rod is located in a standard high-pressure flask, connected to a heavy water loop and a blow-down system. The key rod and rig instrumentation comprised three clad TCs at two elevations, a clad extensometer, a rod pressure transducer, two heater TCs and inlet and outlet coolant TCs. A gamma monitor was mounted on the blow-down line to the dump tank HWR-1042, The BWR LOCA test IFA ; in-pile measurements. F. Khattout. October In-pile data evaluation is completed. The rod will be shipped to Kjeller hot cells for PIE

20 HP-1378 vol. 1 LOCA TEST (IFA ) (GAMMA SCANNING) To confirm fuel relocation/dispersal and study rod deformation and ballooning characteristics. Gamma spectroscopy assay was carried out in October 2012, resulting in 2D images with two orientations of the fuel in the pressure flask. The images indicate significant fuel deformation and fragmentation in the axial midheight of the fuel stack. Fuel particles are observed at the bottom of the pressure flask. Gamma ray (662 kev/cs 137 ) 2D images of the IFA pressure flask in two orientations, 0 (left) and 90 (right). Each pixel represents a gamma spectroscopy measurement using a point collimator, and the colour reflects the count rate of the selected gamma peak (white=low, red=high). The gamma spectroscopy assay resulting in 2D representations of the fuel rod was carried out in the handling compartment at the reactor site 6 days after completion of the in-pile LOCA test. Two axial scans were completed, 90 degrees apart. Both scans included the bottom part and the full cross section of the pressure flask. Main ref. HP-?. Status report, July-December Following the gamma spectroscopy assay at the Halden reactor facility, the rod will be shipped to the Kjeller hot cell laboratories for further examinations.

21 HP-1378 vol. 1 BWR LOCA TEST (IFA ) (IODINE AND CESIUM RELEASE) To study release of Iodine and Cesium from the test rod and test rig. The analysis of the samples retrieved from the blow-down line indicates that less than 1 % of the total rod inventory of Cesium was released from the test flask. Examinations of the fuel rod using gamma-ray spectroscopy assay techniques indicate that the fuel rod has released approximately 3 % of Cesium and 1 % of Iodine during the LOCA test. These results point to low release for the fragmented high burnup fuel employed in the experiments. Gamma-ray assay of fuel rod and blow-down line (BD-line) examination results, in fraction (%) of calculated total fuel inventory. Nuclide In rod BD-line (id) (%) (%) 137 Cs I In the LOCA test series radioactive material may be released to either the pressure flask in the test rig, or the blow-down line and dump tank. In order to examine the release of Cesium and Iodine the activity of 137 Cs and 131 I in the blow-down line is investigated. Post-LOCA 131 I and 137 Cs inventory in the fuel rod is investigated using non-destructive gamma-ray spectroscopy assay. The technique relies on high resolution gamma-ray spectroscopy on the test pressure flask, using a High Purity Germanium (HPGe) detector with accompanying electronics, spectroscopy software and hardware. The collected data are correlated with calibration standard rods, in order to obtain an assessment of the post-test activity inventories of 137 Cs and 131 I. Finally, all measurement data are correlated with the results of fuel inventory calculations, using Origen software by the Oak Ridge National Laboratory (USA). In the case of IFA , the gamma-ray spectroscopy assay on the fuel rod was carried out in the handling compartment at the reactor site 4 days after completion of the in-pile LOCA test. 19 data points were collected over the axial length of the fuel rod. Main ref. HP Status report, July-December Following the gamma spectroscopy assay at the Halden reactor facility, the rod will be shipped to the Kjeller hot cell laboratories for further examinations.

22 HP-1378 vol. 1 BWR LOCA TEST (IFA ) (PIE) The in-pile measurements and gamma scanning suggested rod failure and considerable cladding deformations. The planned PIE is extensive and comprises visual inspection, neutron radiography, dimensional measurements, failure site location and characterisation, ceramography, metallography, clad H-content analyses and fragment size distribution by ASTM sieve analysis. The rod was subjected to a temperature transient over a period of 5 mins. Maximum allowed temperature was 850 C. No spray was used. This was planned to be a no-failure test. The built-in plenum volume was thus reduced to 1.8 cc in order to better control the test, namely to terminate the transient by a scram before rod burst, when the rod pressure had decreased to 50 % of the maximum hot pressure. Maximum measured clad temperature was ~ 800 C. The rod was sound when the scram occurred, but it failed 5-10 seconds into the cooling-down period, evidenced by rod pressure drop and a low activity release. PIE to date has covered gamma scanning, visual, dimensional measurements, neutron radiography and some ceramography. The profilometry showed 42 % maximum ballooning strain (in good agreement with pre-calculations). The burst opening was small and tight. Extensive pellet cracking and fragmentation occurred over nearly the entire rod length, except for the 7-8 lowermost pellets, where the cladding deformations were limited. Significant fuel fragmentation into small size particles were observed and relocation of particles into the balloon zone. Neutron radiography of test rod Main ref. The test segment was from a BWR rod pre-irradiated in the KKL NPP to a burn-up of 72 MWd/kg U. Cladding type was LK3/L (with a 70 µm liner) with an outer diameter of 9.62 mm and a wall thickness of 0.63 mm. Oxide thickness was ~ 40 µm and H-content ~300 wppm. The rod was 38 cm long and pressurized to 20 bar (RT). B. C. Oberlander PIE on high burn-up BWR rod in LOCA test IFA presentation at HPG meeting, Lyon, May 2012 Status report July-December HP-1377 The further PIE will include determination of fragment size distribution by ASTM sieve analysis, burn-up determination of fine particles, oxide thickness and H-content determinations at various positions.

23 HP-1378 vol. 1 PWR CLADDING CREEP (IFA-741) To obtain in-reactor creep data on modern cladding materials under several tensile and compressive stress levels and stress reversals, and assess whether mechanistic changes occur due to fast fluence effects on clad microstructure. The design of IFA-741 is basically the same as the rig it replaces (IFA-699). Loading was deferred because out-of-pile operation and function tests in the ATL (Assembly Test Loop) showed irregular behaviour at higher temperatures and poor repeatability of the DG-2 signal. After repair, the final qualification and commissioning tests are completed, and the rig is working well in-pile. Typical cladding profile measurements are shown in the figure below First in-reactor diameter profile measurements in IFA-741 The lower rod consists of two segments (transferred from IFA-699), M-MDA and M5, and the upper consists of two fresh segments, E110-M and Opt.Zirlo. All segments are fuelled with 8 w% UO 2 and operated under PWR conditions. The rods have a large pellet-clad gap to avoid PCMI. The test rig is housed in a pressure flask surrounded by twelve booster rods to enhance the fast flux. The rods are independently connected to the gas pressurization systems to control cladding hoop stress. Two 3-point contact diameter gauges are used to measure diameter changes of the segments. Main ref. HP Status report, July-December Creep testing will start in December.

24 HP-1378 vol. 1 2ND INTERIM EXAMINATION OF RODS FROM THE PWR CORROSION TEST (IFA-708) The effects of elevated ph (7.4 at 300 C), high power rating (35 kw/m) and significant subcooled boiling on the performance of the cladding materials are being studied to ascertain if sufficient margin is available for further increases in these parameters for future PWR operation. The second interim inspection on IFA-708 was performed in May 2012 after 320 full power days (assembly burnup: 22 MWd/kg UO 2 ). The oxide thicknesses of the different cladding materials varied between ~ 7 and 16 µm. The results indicate a possible synergic effect on corrosion between Sn and Fe. Oxide thickness profile of OPT. ZIRLO cladding material after the first and second interim inspections Main ref. Six test rods, each consisting of four fuel segments fuelled with wt% enriched UO 2. Water chemistry conditions are 10 ppm Li and 1580 ppm B, resulting in a ph 300 of 7.4, and 3 ppm H 2. Maximum mass evaporation rates ~4500 kg/h/m 2. Corrosion is being assessed by means of interim inspections, comprising photography and oxide thickness measurements. Réka Szőke from the second interim inspection of the PWR cladding corrosion test IFA-708. HWR-1045, 2012 Irradiation of IFA-708 was restarted in June After a further 29 FPD, indication of rod failure was observed. The assembly was discharged, and visual inspection showed failures in Segment 3 of rods (M5) and (J2). The causes of these failures will be investigated and reported separately. A decision will be made on whether to continue irradiation of the four unfailed rods.

25 HP-1378 vol. 1 ON-LINE, IN-CORE CLADDING CORROSION (IFA-731) Perform on-line, in-core measurements of the corrosion of Zircaloy-2 and Zircaloy-4 cladding under PWR water chemistry conditions, using Electrochemical Impedance Spectroscopy (EIS) and Potential Drop (PD) measurements. The test assembly was loaded in March 2012 and operated for 125 Full Power Days before the assembly was discharged in October. Daily measurements have been made of the solution and polarisation resistances of each cell, together with the double layer capacitance. These results, together with standard electrochemical parameters, have been used to calculate the corrosion rates of Zry-2 and Zry-4 test rods. Calculated corrosion rates of Zircaloy-4 (Rcorr3) and Zircaloy-2 (Rcorr4) Main ref. Five test fuel rods: four are instrumented for EIS measurements and the other for PD measurements. For EIS, two rods each with Zry-4 and Zry-2 cladding. The PD rod has Zry-4 cladding. For the EIS measurements, a three-electrode assembly is used, consisting of the fuel rod, a counter electrode and a reference. This standard set-up employs the counter electrode to allow current to flow in the circuit while the reference is employed to control the potential between it and the fuel rod. The assembly was operated under PWR thermal-hydraulic and water chemistry conditions. P J Bennett and R Szöke, HWR-1046, in preparation. The oxide layer thicknesses on the test rods will be determined by PIE for comparison / calibration against the in-core data. Equivalent circuit modelling of the EIS results is in progress.

26 HP-1378 vol. 1 TEM Analysis of Cladding Samples (From IFA-638) To study the metal-oxide interface region and the microstructure on both sides of the interface of six Zirconium alloys from IFA-638 by transmission electron microscopy (TEM). The investigations will focus on the geometry of the metal-oxide interface, the composition and microstructure of precipitates in the metal and oxide, the oxide microstructure and the presence of hydrides. The work is being performed at PSI in Switzerland. The pre-irradiated and initially fresh M5 showed a sub-micron zigzagged interface with circumferentially oriented hydrides. EDS analysis of the precipitate composition in the pre-irradiated M5 showed a Nb/Zr ratio of 1.2 and ratios up to 4.2 in the initially fresh M5. The initially fresh E635 and ZIRLO showed undulated metal-oxide interfaces. EDS analysis of the precipitates in the E635 showed primarily a composition of ~40% Zr, 40% Nb, 20% Fe. The precipitates in the ZIRLO were of the β-nb type (with a Nb/Zr ratio of up to 4.5) and amorphous precipitates with Fe, Nb and Zr. Main ref. TEM dark field contrast of metal-oxide interface of E635 with interface marked by arrows. Hydrides in the metal can be observed, precipitate in the oxide marked with P. The TEM examination (using a JEOL2010 equipped with a LaB6 cathode and an EDS system) is performed on cladding materials from IFA-638. The materials are initially fresh and pre-irradiated M5 and ZIRLO and initially fresh E638 and Alloy A. ENSI Research Report 2011, IFA-638-TEM Examinations of Metal-oxide Interface of Zirconium alloys. TEM examination of the pre-irradiated ZIRLO and initially fresh Alloy A.

27 HP-1378 vol. 1 BWR CRACK GROWTH RATE TEST (IFA-745) To generate long-term crack growth rate (CGR) data for irradiated Compact Tension (CT) specimens in simulated BWR conditions and to compare the cracking response as a function of material, dose, electrochemical corrosion potential (ECP), temperature, stress intensity (K) level and post irradiation annealing (PIA) treatment. Seven CT specimens were irradiated from July to September CGRs were measured on CT1 CT6 in ~5 ppm O 2 and ~2 ppm H 2 water conditions at 280 C. The CGRs were in the range of mm/s at K levels of MPa m. Most of the CGRs in H 2 conditions were reduced by about one order of magnitude relative to those in O 2 conditions at the same K levels. Similar CGRs were measured for the 7.7-dpa 304L SS CTs with different PIA conditions. For CT7, indications of a fault in the bellows loading device were found. Main ref. Summary of CGR vs. K data for 7.7-dpa 304L SS CTs with different PIA conditions. IFA-745 contains six CT specimens prepared from irradiated stainless steels from commercial reactors and one CT specimen prepared from unirradiated stainless steel. Two PIA treatments (500 C for 25 hours and 550 C for 25 hours) are applied to two CT specimens. Six irradiated CT specimens are instrumented for crack growth monitoring with the dc potential drop (dcpd) method. One unirradiated CT specimen is equipped with a linear variable differential transformer (LVDT) for crack growth monitoring by means of displacement in load line (DiLL) measurements instead of the ordinary dcpd method. All the CTs are also equipped with bellows for load application. HWR-1072, Minutes of Halden IASCC Review Meeting October 2012., M. Lundgren, November Continue irradiation and continue to measure the CGRs in O 2 and H 2 water conditions at 280 and 325 C to evaluate temperature effects.

28 HP-1378 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 At the start of the test failures were recorded for three specimens with a target load of 100 % yield strength (YS). The target load was subsequently reduced to 90 % YS and thereafter increased to 95 % YS. Fracture surfaces of failed specimens have been inspected and intergranular fracture was observed on one while the two other indicated completely ductile failures. Fracture surface on the shorter half-piece of specimen in unit 3. The surface displays a ~10% intergranular region. 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 80 % and 100%, thereafter lowered to ~70% and 90 % and subsequently raised to ~75% and 95 % 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 ). Main ref. HWR-1072, Minutes of Halden IASCC Review Meeting October Testing will continue in the programme period. Loads will be increased stepwise by ~5% every ~2000 FPH.

29 HP-1378 vol. 1 CHARACTERISATION OF IASCC TEST MATERIALS (I) Characterisation of the microstructure and microchemistry of irradiated IASCC test materials. include determination of Frank loop size distributions and densities and imaging of precipitate populations. The radiation induced segregation (RIS) profiles of the materials were analysed and the extent of the segregation (e.g. of Ni and Si) or depletion (e.g. of Cr and Fe) at the grain boundaries was measured. The profiles showed that PIA only slightly altered the amount of RIS. Main ref. TEM characterisation of samples of as-irradiated and post irradiation annealed (500 C for 6 hours) 7.7 dpa 304L SS was made using a Philips CM200 FEG-STEM operating at 200 kv. Energy-dispersive x-ray spectroscopy (EDS) was performed at suitable grain boundaries within the materials in order to characterize the spatial distribution of elements in the materials. VTT-R , TEM examination of the effect of post-irradiation annealing on 7.7 dpa AISI 304 stainless steel, J. Pakarinen, September 2012 VVT-R , ATEM characterization of a slice from a 12 dpa AISI 316 stainless steel baffle bolt, J. Pakarinen, October 2012 Characterisation of samples of 7.7 dpa 304L SS with Post Irradiation Annealing (PIA) treatments of 500 C for 25 hours and 550 C for 25 hours. Characterisation of additional irradiated materials to be used in crack growth rate studies.

30 HP-1378 vol. 1 CHARACTERISATION OF IASCC TEST MATERIALS (II) Characterisation of the microstructure and microchemistry of irradiated IASCC test materials. Frank loops with a concentration of loop/m 3 were found. The average loop size was 13.2 nm. GBs showed clear RIS, as well as the presence of precipitates rich in Si and Ni. One ppt phase was identified as ɣ (Ni 3 Si) and had a lattice relation with the fcc matrix. Voids/bubbles ~5-10 nm were present in the microstructure. Void walls also showed clear RIS but did not appear to be constant at the void surfaces. Voids and precipitates were found to often be associated with one another. Dark and bright field images showing grain boundary (GB) and matrix precipitates and EDS results from the GB precipitates TEM characterisation of a ~12 dpa CW 316 SS baffle bolt sample was made using a Philips CM200 FEG-STEM operating at 200 kv. Energy-dispersive x-ray spectroscopy (EDS) was performed in order to characterize the spatial distribution of elements in the materials. Main ref. VVT-R , ATEM characterization of a slice from a 12 dpa AISI 316 stainless steel baffle bolt, J. Pakarinen, October 2012 Characterisation of 7.7 dpa 304L SS samples with Post Irradiation Annealing (PIA) treatments of 500 C for 25 hours and 550 C for 25 hours. Characterisation of additional irradiated materials to be used in crack growth rate studies.

31 HP-1378 vol. 1 IRRADIATION CREEP AND STRESS RELAXATION STUDY (IFA-669.2) Measure creep and stress relaxation of materials used in PWR and BWR plants. IFA-669 has been irradiated since January 2006 and irradiation creep and stress relaxation data have been obtained for the specimens. For the CW 316 SS samples, irradiation creep and stress relaxation data are consistent. The CW 316 N lot and aged alloy 718 specimens exhibit higher stress relaxation than CW 316 SS. The CW 316LN and SA 304L SS are more creep resistant than CW 316 SS. Summary of creep data for CW 316 SS and CW 316LN samples. The CW 316LN is more creep/stress relaxation resistant than the CW 316 SS. Main ref. The rig contains 12 instrumented tensile specimens prepared from CW 316 SS, CW 316LN, CW 316N lot, SA 304L SS and aged alloy 718. The specimens are installed 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. During irradiation (in an inert environment) the specimens are exposed to temperatures of 290, 330 or 370 C and applied stress levels range from 92 to 345 MPa. 15 th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors, Colorado Springs, Irradiation Creep and Irradiation Stress Relaxation of 316 and 304L Stainless Steels in Thermal and Fast Neutron Spectrum Reactors, J.P. Foster and T.M. Karlsen August 2011.