Fuels and Materials Programme Achievements 2013

Transcription

1 Institutt for energiteknikk OECD HALDEN REACTOR PROJECT HP-1416 vol. 1 For use within the Halden Project Member Organisations only OECD HALDEN REACTOR PROJECT Fuels and Materials Programme Achievements 2013 February 2014

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3 FOREWORD The experimental operation of the Halden Boiling Water Reactor, the Halden Man-Machine Laboratory, HAMMLAB, the Virtual Reality Laboratory, the Future Lab and the 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 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 of 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 privat nuclear organisations in Sweden, the Swiss Federal Nuclear Safety Inspectorate ENSI, representing also the Swiss nuclear utilities (Swissnuclear) and the Paul Scherrer Institute, 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), the French Institut de Radioprotection et de Sûreté Nucléaire (IRSN), Commissariat á l energie atomique (CEA), France, EU JRC Institute for Transuranium Elements, Karlsruhe, Germany, the Hungarian Academy of Sciences, KFKI Atomic Energy Research Institute, the Italian National Agency for new Technologies, Energy and Sustainable Economic Development (ENEA), representing a group of Italian companies, 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) the Ulba Metallurgical Plant JSC in Kazakhstan, the JSC "TVEL" and NRC "Kurchatov Institute", Russia, All-Russian Research Institute for Nuclear Power Plants Operation (VNIIAES), Russia, the Slovakian VUJE - Nuclear Power Plant Research Institute, the Federal Authority for Nuclear Regulation (FANR), United Arab Emirates, and associated parties from USA: the Westinghouse Electric Power Company, LLC (WEC), the Electric Power Research Institute (EPRI), the Global Nuclear Fuel (GNF) Americas, LLC and GE-Hitachi Nuclear Energy, LLC, 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 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 2013, 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 2013 The OECD Halden Reactor Project started the new agreement period ( ) with nineteen member countries. 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. In 2013, the United Arab Emirates (UAE) became the 20 th member country of the Halden Project. The Federal Authority of Nuclear regulation (FANR) is representing the UAE. 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 2013 was Hans Larsen from DTU, Denmark. The chairman of the Halden Programme Group in 2013 was Jing Xing from the US NRC. 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 was held at Storefjell, Norway in March The second EHPG meeting will be held at Røros, 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 2013 was within the topic of Principle of Fuel Behaviour Modelling and Practical Applications 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 2013: Workshop on «Using Simulator Data to Improve Human Reliability Analysis», Halden, May Workshop on «Nordic Cooperation IEC/TC45», Halden, October NKS Decommissioning Seminar, Halden, November 6-7 Workshop on Data Uncertainties in Experiments and Modelling, Halden, September 4-5

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, 10 test rigs were under irradiation at any one time during 2013 as part of the Halden Reactor Project Joint Programme, with a total of 12 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 2013 was 57%. 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 2013, a third set of creep data were obtained at a burn-up of ~25 MWd/kg UO 2 and compared to the behaviour observed earlier at burn-ups of 8 and 18 MWd/kg UO 2. The creep rate continues to show a linear dependence on stress and fission rate, and was independent of temperature, consistent with data from open literature. From the three data sets, there is also no clear burn-up effect on the creep rate (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-1416 vol. 1 OECD Halden Reactor Project HALDEN PROJECT PROGRAMME ACHIEVEMENTS 2013 CONTENTS Page OVERVIEW OF FUELS AND MATERIALS EXPERIMENTS DURING PIE ON TEST RODS FROM IFA HIGH BURN-UP FUEL DISK IRRADIATION (IFA-655.2)... 3 PWR OVERPRESSURE/LIFT-OFF EXPERIMENT (IFA )... 4 FGR MECHANISMS (IFA-716)... 5 FUEL CREEP TEST (IFA-701)... 6 VVER FUEL BEHAVIOUR (IFA-676)... 7 PIE ON THE BWR LOCA TEST ROD FROM IFA BWR LOCA TEST IFA : IN PILE RESULTS... 9 BWR LOCA TEST IFA : GAMMA SCANNING PWR CLADDING CREEP (IFA-741) PIE ON IFA-708 TEST RODS BWR CRACK GROWTH RATE TEST IFA CRACK INITIATION STUDY (IFA-733) CHARACTERISATION OF IASCC TEST MATERIALS IRRADIATION CREEP AND STRESS RELAXATION STUDY (IFA-669.2) PRESSURE VESSEL AGING (SMALL PUNCH) REPORTING February 2014 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 HP-1416 vol. 1 OVERVIEW OF FUELS AND MATERIALS EXPERIMENTS DURING 2013 MEASUREMENTS APPLICATIONS IFA fuel (f) clad (c) or material type and origin # of rods / specimens burnup MWd/ kg oxide fluence n/cm 2 (x10 20 ) Recent reports Comment UO 2 /PWR 1 65 X x x x x x x HP-1397, HP-1409 PWR Overpressure UO 2 disks X x x x HP-1397, HP-1409 FGR ramp PIE UO 2 disks X x x x HP-1397, HP-1409, HWR-1041 FGR ramp PIE BWR 1 64 X x x x x x x HP-1397, HP-1409, HPR-380 LOCA test, PIE BWR 64 HP-1397, HP-1409, HWR-1084 LOCA test UO x x x HP-1397, HP-1409 Disks, rim 669 SS/Inconel 30 - X HP-1397, HP-1409, HWR-1047 Stress relaxation 676 VVER/UO X x x x x x x x HP-1397, HP-1409 Base irradiation 681 Gd/UO X x x x x x x x HP-1397, HWR-1038 Gd-fuel, PIE 701 UO 2 / Cr-doped 4 27 X x x x HP-1397, HP-1409, HWR-1039 Fuel creep 708 M5, Zirlo, M-MDA-SR 6 24 x x x HP-1397, HP-1409 Cladding corrosion PIE 716 UO 2 / fresh 6 24 X x x x x x HP-1397, HP-1409 High LHR, FGR 731 Zr-2, Zr-4 5 x x HP-1397, HP-1409 On-line corrosion L dpa x x HP-1397, HP-1409 BWR crack init. 741 E110, M5, M-MDA, Opt. zirlo 4 19 x x x x HP-1397, HP-1409 Cladding creep SS, 304L SS 7 36 HP-1397, HP-1409, HWR-1079 BWR crack growth SPT SS 210 HP-1397, HP-1409 RPV testing

10 - 2 - HP-1416 vol. 1 PIE ON TEST RODS FROM IFA-629 Measure volume of, and determine nuclides present in, the fission gas produced in the test rods from the FGR tests IFA and The rods were punctured in the plenum region in the lower end. The expansion pressure became stable after puncturing, i.e. no problems were experienced with degraded gas communication under puncturing of the rods. Kr-85 was the only radioactive fission product identified. The released fission gas was 11.6% and 14.6% respectively for Rods IFA R and IFA R. These values compare well with the fission gas release determined from the on-line measurements during the IFA-629 tests (12 and 13%). The gas amount of xenon divided by gas amount of krypton was similar for the two rods, i.e. 7.9 which reflects high burn-up UO 2 fuel. Main ref. Remaining work Rod internal pressure history and FGR during IFA Two test rods containing UO 2 and Mo disks irradiated in IFA to a burn-up of 113 MWd/ kg U in order to produce the high burn-up (HBS) structure. The rods were refabricated with PF and EFs and the FGR behaviour was studied during ramp testing in IFA-629. Rod 655-R-1 had medium grain size fuel and was tested in IFA-629.5, while Rod 655-R-2 had large grain size fuel and was tested in IFA HWR-1010 (IFA-629.5) HWR-1041 (IFA-629.6) HP-1409 Perform ceramography on the test rods.

11 - 3 - HP-1416 vol. 1 HIGH BURN-UP FUEL DISK IRRADIATION (IFA-655.2) To irradiate fuel disks to high burn-up under low temperature conditions for investigation of high burn-up structure. During 2013, an updated depletion calculation was performed with HELIOS, extending the burnup calculation range beyond the target burnup. The depletion function for this extended burnup range was lower than the function used previously, resulting in a calculated burnup ~10% lower than that reported previously. With the resulting power history from the HELIOS calculation to extended burnup, the total burnup of the fuel rods at the end of December 2013 IFA was around 154 MWd/kgUO 2. Main ref. Remaining work Figure 1 Updated power and burnup history after extended burnup depletion calculation. The two test rods in IFA were originally loaded in IFA together with four other test rods, where the irradiation started in August After unloading the other four rods, continuing irradiation of the test rods and in the same rig started in June 2008 as IFA-655.2, when the total irradiation period was around 881 FPD and the burn-up of the two rods was around 113 MWd/kgUO 2. Each fuel rod contains 25 fuel disks (~20% enriched UO 2 ), 1 mm thick, and constrained by Mo disks on both sides. The density of the fuel pellets in both rods is 95 %T.D. The grain size of the fuel pellets in the rod No.3 is 10 µm, while that in the rod No.4 is 50 µm. Both the test rods were equipped with fuel extensometer (EF) and pressure transducer (PF) from the start of operation. However the rig instrumentation has since experienced failure. Status Report July December 2013, HP To continue irradiation to around 200 MWd/kgHM.

12 - 4 - HP-1416 vol. 1 PWR OVERPRESSURE/LIFT-OFF EXPERIMENT (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 for significant fission gas release. The rod has been subjected to overpressure of up to 150 bar. During the period with constant ALHR (from about 1600 fph), the normalized fuel temperature has decreased and then stabilized. Conversely, the normalized cladding elongation was almost constant or increasing slowly during the same period. These observations may be due to improved heat transfer from the fuel to the coolant as a result of increased PCMI. No sign of lift-off has been detected to date. Main ref. Remaining work Average linear heat rate (ALHR), LHR at the thermocouple position (LHRTF), overpressure, normalized fuel temperature at 12kW/m (Norm.TF) and normalized elongation of the test rod at 12kW/m (Norm.EC) The test rod for IFA is a part of a father rod with recrystallized 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. 70 MWd/kg U (rod average burnup). The test rod was fabricated at Kjeller with a fuel thermocouple (TF) and a clad extensometer (EC). A gas line is mounted at each end of the test rod and connected to the gas flow control system which enables 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. HP The overpressure will be increased to +150, +175, +200 bar sequentially according to the planned scheme.

13 - 5 - HP-1416 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. Following a slow power increase in Dec 2012, during which fuel temperatures approached the Halden 1% fission gas release threshold in the five rods without BeO dopant, fission gas release was seen in these five rods. Operation in 2013 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 is in the order 3-4% for the various rods (the sensor attached to Rod 2, indicating 1% release is faulty). The temperatures in the rod with 3.0 wt% BeO are ~ C lower on account of the higher thermal conductivity of this fuel. No fission gas release has been seen for this rod. The burnup in December 2013 was about 25 MWd/kg Oxide. Main ref. Remaining work Fission gas release and peak temperatures during 2013 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). Status Report July December 2013, HP Irradiation is planned to continue at about C peak fuel centre temperature. Periods with power increase will be inserted at appropriate points during operation in order to observe fission gas release kinetics. An interim work report will be prepared for the September 2014 EHPG meeting.

14 - 6 - HP-1416 vol. 1 FUEL CREEP TEST (IFA-701) To study the fission induced creep of UO 2 and Cr-doped fuel in the temperature range where thermal creep is not significant (below 900 o C) by measuring the irradiation-induced creep rate of the fuel as a function of fuel temperature and applied compressive stress at a fixed fission rate. The second creep test was initiated in the middle of July 2013 at the average burnup of ~ 25 MWd/kg UO 2, which is ~ 5 MWd/kg UO 2 higher than the previous measurements. The target fuel temperatures were 400, 600 and 800 o C, while the applied stress was 30, 45 and 60 MPa during each temperature period. The creep test was finished after ~ 60 days of operation. At this point the average burnup was ~ 26.5 MWd/kg oxide. The creep rates were obtained from the difference between the measured fuel stack extensions from the creep rods and reference rods. The results show that the creep rates increased with the increase of fission rate and stress, but there was no dependence on temperature. These findings are in good agreement with literature data. Summary of measured fuel creep rates Main ref. Remaining work The assembly contains one dummy rod (replacing the failed UO 2 reference rod) and three test rods: two Cr-doped rods (one a reference rod and one a creep rod) and one UO 2 creep rod, loaded in a single cluster. The rods comprise alternating fuel and molybdenum (Mo) disks, and all are fitted with fuel centreline thermocouples, fuel stack elongation detectors and a loading device. The loading device uses a bellows with gas line connection to generate the compressive stresses in the creep rods and a hold down stress in the reference rods. All rods are connected to the He/Ar gas system in order to control fuel temperature. HP-1409, HWR-1039 Move the test rig to a higher power position to examine the feasibility of testing at higher temperatures (1200 C).

15 - 7 - HP-1416 vol. 1 VVER FUEL BEHAVIOUR (IFA-676) To determine the FGR threshold for large grain and standard VVER UO 2 fuels at high burnup (60 MWd/ kg oxide). An experiment was performed with successive power ramps (step-wise temperature increases) and 24 hours hold time at each power level. The first power level corresponded to the maximum measured fuel centre temperature of about ~800 C. Then the maximum measured temperature was increased in steps of ~50 C until 900 C, and continued in steps of 30 C until 1% FGR was achieved in both rods. The 1% FGR threshold for both fuel types was about 980 C, against 1040 C calculated from the Halden threshold. Additional gas release was detected following the subsequent power reduction, which may be explained by the release of the fission gas that was trapped in the fuel cracks at power The total FGR (6 % for standard grain and 2.5% for the large grain fuel) was dependent on the maximum level of the fuel temperatures reached during the step wise power ramps. Measured gas pressure and fuel temperatures, and FGR evaluation Main ref. Remaining work Six test rods, with pairs of (i) VVER fuel with aluminium silicate additives, which enhance grain size to µm; (ii) standard VVER fuel (about 11 µm grain size); and (iii) VVER fuel with 5 wt% Gd. All test rods clad with Zr-1% Nb and irradiated in the HBWR moderator. HP Irradiate the UO 2 rods to a burnup exceeding 65 MWd/kg oxide, and the Gd-bearing rods to 30 MWd/kg oxide. PIE will then be performed.

16 - 8 - HP-1416 vol. 1 PIE ON THE BWR LOCA TEST ROD FROM IFA To obtain information on the LOCA-induced state of the rod in terms of thresholds for cladding ductility, cladding failure, and fuel fragmentation. Neutron radiography revealed six intact fuel pellets in the lower part of the rod, while in the upper end, the top two pellets were intact. A fuel gap was observed in the upper end of the fuel stack. Fine fragments of the fuel had dropped down to the lower end of fuel stack. An increase in clad deformation was seen from the lower end towards mid rod location. Cracking of the cladding outer surface oxide was mainly observed above pellet-pellet-interfaces (PPI) and fuel relocation sites. In these locations a higher cladding deformation dominated. Fuel fragmentation with millimetre-sized fuel fragments and fuel relocation were observed in a >25 cm-long zone above and below mid-rod location and burst. Neutron radiography of the IFA test rod Main ref. HP-1397, Vol. 1. Remaining work A test segment prepared from a rod irradiated in the KKL BWR to a burn-up of ~72 MWd/kgU. The rod plenum volume was 15 cm 3. The cladding material was LK3 with liner. The oxide thickness was ~20 µm and the hydrogen content ~300 ppm. In the LOCA test, ballooning and cladding failure occurred (burst at ~827 C), consistent with pre-test code calculations made by PSI. The non-destructive PIE program included gamma scanning, visual inspection, neutron-radiography, and measurement of the rod diameter. Metallography, determination of the size distribution of the fuel fragments.

17 - 9 - HP-1416 vol. 1 BWR LOCA TEST IFA : IN PILE RESULTS Perform a LOCA test on a pre-irradiated BWR fuel segment. Interrupt the temperature transient in the ballooning progress by a reactor scram before clad rupture when the rod pressure had decreased to % of the maximum measured hot pressure. The target/maximum allowed cladding temperature (PCT) was 800 C. The LOCA sequence was run as planned. Rod fill pressure at RT was 20 bar, which increased to 77.3 bar at hot conditions. Significant ballooning occurred above C. The cladding was ductile and the reduction in rod pressure to 74 percent of the maximum value was achieved at a temperature of 792 C. The rod was intact when the reactor was scrammed. The maximum measured clad temperature was 792 C. No rupture occurred, and the pressure stabilized at 21 bar after the test. Rod pressure, clad temperatures and elongation, gamma monitor response in the blow-down line and heater power Main ref. Remaining work A test fuel segment prepared from a rod which had been irradiated in the KKL BWR to a burn-up of ~72 MWd/kgU. The length of the fuel stack was ~ 360 mm and no end pellets were inserted. The rod was filled with a 95% Ar/5% He mixture at 20 bar (RT). The rod plenum volume was made relatively small, 1.9 cm³, to allow a significant ballooning without burst. The cladding material was Zry-2 (LK3/L) with an inside liner. HWR PIE to be performed on the fuel segment, including neutron radiography and sieving of the fuel fragments.

18 HP-1416 vol. 1 BWR LOCA TEST IFA : GAMMA SCANNING Perform gamma scanning on the test rod following LOCA testing in IFA Gamma scanning was performed in Halden approximately two weeks after the test to examine the state of the rod after the LOCA test. The scans indicate that the rod was slightly bent and that significant ballooning had occurred. There were strong indications of fuel fragmentation and relocation since the balloon was filled with fuel. A small axial gap can be seen at elevation 118 mm. Gamma scans of IFA two weeks after the test. Full scan at 0 and 90 Main ref. Remaining work A test fuel segment prepared from a rod which had been irradiated in the KKL BWR to a burn-up of ~72 MWd/kgU. The length of the fuel stack was ~ 360 mm and no end pellets were inserted. The rod was filled with a 95% Ar/5% He mixture at 20 bar (RT). The rod plenum volume was made relatively small, 1.9 cm³, to allow a significant ballooning without burst. The cladding material was Zry-2 (LK3/L) with an inside liner. HWR PIE to be performed on the fuel segment, including neutron radiography and sieving of the fuel fragments.

19 HP-1416 vol. 1 PWR 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 lower test rod (carried over from IFA-699 to IFA-741), and containing M5 and M-MDA, has experienced hoop stress conditions of -50 MPa and +30 MPa twice. For the -50 MPa levels, the total diameter change during the later period is larger than that during the first period. Conversely, for the +30 MPa level, the opposite behaviour was seen. From these results, an irradiation hardening effect on creep behaviour cannot be clearly seen. In addition, the total diameter change was significant both when the magnitude of the hoop stress was increased in the same direction, i.e. from +30 MPa to +110 MPa, and when the sign of stress was reversed, i.e. from - 50 MPa to +30 MPa and from +30 MPa to -50 MPa. Start of 741 Main ref. Remaining work Total diameter change for the lower test rod segments 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 a gas line and a diameter scanning gauge. The rig is connected to a loop system providing PWR thermal-hydraulic conditions. HP The hoop stress condition will be changed to +110 MPa for the upper test rod and +/-0 MPa for the lower test rod.

20 HP-1416 vol. 1 PIE ON IFA-708 TEST RODS The main objective of IFA-708 is to study the in-pile corrosion and hydriding behaviour of modern Zircaloy-based PWR cladding materials in aggressive water chemistry and thermal hydraulic conditions exceeding those currently allowable in operating PWRs. Diameter and bow measurements of rod 9 and rod 12 were performed in four angular orientations, i.e. 0, 45, 90 and 135. The results show that rod 9 is still relatively straight, while rod 12 has a maximum bow of approximately 1 mm at 0. The 0 degree orientation points out from the rig normal to the flask inner surface. Visual inspection was performed in four different angular orientations, i.e. 0, 90, 180 and 270. No scratches were observed on the two rods, which shows that the rods were not in direct mechanical contact with the flask during irradiation or under removal of the rig after irradiation. Dry-out marks were observed at the 0 orientation and nearby areas in segment 3 for both rods, which correlates to diameter increases. Fig. 1. Visual inspection of Segment 3, rod 9 and 12, at 0 degree orientation after irradiation in the HBWR. For rod 12, segment 2 at the upper end, a diameter decrease is observed for 3-4 pellets. Irradiation of IFA-708 was restarted in June 2012 after the second interim inspection. The rods were operated at power range of kw/m, with outlet mass evaporation rates of up to 4200 kg/m 2 h. After 29 FPD of irradiation, an indication of rod failure was given by a gamma monitor in the loop. The assembly was discharged and visual inspection showed failures in Segment 3 of rods (M5) and (J2). The failed rods were transported to Kjeller for PIE. Main ref. HP-1342, Status Report, July-December Remaining work R. Szőke from the second interim inspection of the PWR cladding corrosion test IFA-708, HWR Irradiation of IFA-708 will be restarted in June The remaining four unfailed rods will be used in the new loading if they do not exhibit elevated bowing.

21 HP-1416 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. During irradiation, CGRs were measured on all the CTs in O 2 (high ECP) and H 2 (low ECP) water conditions at ~280 C and on five of the CTs at ~320 C. The CGRs were in the range of mm/s at K levels of 9 48 MPa m. Most of the CGRs in H 2 conditions decreased by about one order of magnitude relative to those in O 2 conditions at the same K levels. For the ~280 to ~320 C switches, the CGRs increased for 5.9 dpa 304L SS (CT1) and 6.2 dpa 304 SS (CT2). The CGRs for the PIA treated 7.7 dpa 304L SS CTs were similar to those of non-annealed samples prepared from the same material. For CT7, indications of a fault in the bellows loading device were found. Main ref. Example of crack growth rates measured in 6.2 dpa 304 SS CT at 280 and 320 C IFA-745 contained 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) were applied to two CT specimens. Six irradiated CT specimens were instrumented for crack growth monitoring with the dc potential drop (dcpd) method. One unirradiated CT specimen was 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 were also equipped with bellows for load application. HWR-1079, Final Report on the BWR Crack Growth Rate Investigation IFA- 745, Y. Chimi, J. Balak, V. Andersson, October 2013

22 HP-1416 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. Loads are currently increased by 5 % YS every ~2000 full power hours (FPH). One additional specimen failed as load was increased to 100 % YS and three as loads were increased to 105 % YS. Fracture surfaces of the first four failed specimens have been inspected and intergranular fracture was observed on one while the three others indicated completely ductile failures. Fractography of the final three specimens is pending. Fracture surfaces on the shorter half-pieces of failed specimens. The surface on specimen 54 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% 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 Remaining work Testing will continue in the programme period. Loads are increased stepwise by ~5% every ~2000 FPH.

23 HP-1416 vol. 1 CHARACTERISATION OF IASCC TEST MATERIALS Characterisation of the microstructure and microchemistry of irradiated IASCC test materials. Both low and high magnification SEM images were taken of the fracture surfaces of the six dcpd-instrumented CTs from IFA-745. The depths of the fatigue pre-cracks and stress corrosion crack depths were measured and the results were used to adjust the data that were recorded on-line with the dcpd method. IFA-745 CT 1 Details of IGSCC Direct observation location C of CT1- half after splitting at Halden A B C Details of IGSCC at location D A B A: Pre-crack Crack length: 1: 1.7 mm 2: 2.1 mm 3: 2.1 mm 4: 2.2 mm 5: 2.1 mm 1) Detail of IGSCC at location C, 100X 4) Detail of IGSCC at location D, 100X D C C: Stress corrosion crack Crack length: 1: 8.0 mm 2: 7.9 mm 3: 8.0 mm 4: 7.7 mm 5: 7.8 mm E F 1a) CT1-half used for replica 2) Detail of IGSCC at location C, 250X 5) Detail of IGSCC at location D, 250X D F E 3) Detail of IGSCC at location C, 500X 6) Detail of IGSCC at location D, 500X Main ref. Low magnification image (left) of 5.9 dpa 304L SS CT from IFA-745 showing fatigue pre-crack and SCC regions and high magnification (right) images of region C. After in-pile testing the CT specimens from IFA-745 were broken open and replicas were taken of the specimen fracture surfaces. The replicas were examined in a Scanning Electron Microscope (SEM). HWR-1079, Final Report on the BWR Crack Growth Rate Investigation IFA- 745, Y. Chimi, J. Balak, V. Andersson, October Remaining work Transition Electron Microscope (TEM) 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 (i.e. the same PIA treatments that were applied to two CTs from the BWR crack growth rate test IFA-745). TEM characterisation of additional irradiated materials to be used in crack growth rate studies.

24 HP-1416 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 was irradiated January 2006 to December 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. Stress relaxation data for CW 316 and CW 316N lot specimens Main ref. Remaining work 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. 16 th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors, Asheville North Carolina John Paul Foster and Torill Karlsen, Reduction of Creep in Cold Worked 316 Stainless Steel by Compositional Modification J.P. Foster and T.M. Karlsen, Augsut PIE will be performed on selected samples.

25 HP-1416 vol. 1 PRESSURE VESSEL AGING (SMALL PUNCH) Evaluation of use of the Small Punch Test (SPT) method for determining the mechanical properties of reactor pressure vessel and core component materials. Irradiation began in March 2012 and the 5 x n/cm 2 capsule was discharged in November The high dose capsule was discharged in June Mechanical testing of the unirradiated materials and low dose materials has been completed and the low dose results are being analysed. The results from the unirradiated material testing shows that the ferritic reactor pressure vessel materials (WM2 and BM2 in the figures below show) a distinct transition temperature while there are only slight changes of SPT energy for the austenitic steels (AC and VPKO in the figures below). Main ref. Temperature dependence of SPT energy for ferritic reactor pressure vessel materials (WM2 and BM2) and austenitic steels (AC and VPKO). SPT samples and tensile specimens prepared from reactor pressure vessel base and weld material, austenitic cladding and two austenitic steels were irradiated in an inert environment at 275 C to fluences of 5 x n/cm 2 and 10 x n/cm 2 (> 1MeV). In addition to the specimens the dry irradiation capsules also contain melting alloy temperature monitors (MATMs) and flux wires for post-test determination of temperature and fluence respectively. HPR-378-Volume 2, EHPG Meeting 2013, Determination of Mechanical Properties of WWER-440 Reactor Pressure Vessel Steels before Irradiation in the Halden Reactor, M. Březina, J. Petzová, L. Kupča Remaining work Testing of the high dose samples will take place in 2014.

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27 Author Title HP-1416 vol. 1 REPORTING HPR-380 W. Overview of Halden LOCA tests: evaluation report Wiesenack HWR-1043 K. Eitrheim Radioactive iodine and cesium release measurements (HWR-1044) B. PIE on the test rod from LOCA test IFA Oberländer (HWR-1040) R. Tradotti Status report on the fission gas release test IFA-716 HWR-1041 T. Tverberg Fission gas release of high burnup fuel disks in IFA HWR-1038 F. Khattout The Gadolinia fuel test IFA-681: Overview of in-pile measurements from beginning of irradiation to unloading (HPR-377) R. Szőke Comparison of Halden Zry-4 corrosion tests HWR-1045 R. Szőke from the second interim inspection of the PWR cladding corrosion test IFA-708 HWR-1046 P. Bennett from the on-line PWR cladding corrosion test IFA-731 HWR-1047 T. M. Karlsen Update on the stress relaxation test IFA-669 HWR-1042 F. Khattout The BWR LOCA test IFA : in-pile measurements HWR-1049 J. C. Kim Core physics calculations for the Halden Reactor HWR-1039 K. Sakai The fuel creep test IFA-701: results after four irradiation cycles HWR-1079 Y. Chimi Final Report on the BWR Crack Growth Rate Investigation IFA-745 HWR-1080 W. Wiesenack Data uncertainties in experiments and modelling workshop minutes Papers presented at Storefjell, March 2013 Author Title F2.1 R. Szőke Update on the VVER fuel test IFA-676 F2.8 S. Holcombe A non-destructive gamma-spectroscopy based method for investigating the radial location of released fission gasses from within high burnup fuel pellets F3.7 Y. Shinohara Initial results from the cladding creep test IFA-741 F4.3 Y. Chimi Initial results from the BWR IASCC test IFA-745 F4.4 M. Lundgren Initial results from the BWR crack initiation test IFA-733 F5.2 B. Preliminary PIE results from the LOCA test IFA Oberländer F7.1 Y. Shinohara Initial results from the overpressure test with M-MDA-RX cladding (IFA ) F8.3 T. Tverberg Plans for a power ramp test on Gd fuel in IFA-720

28 Paper presented outside the Halden series IFE/HR/E 2013/ HP-1416 vol. 1 Storage of Spent Nuclear Fuel in Norway: Status and Prospects, P. Bennett, E. Larsen, CSNI International Workshop on Safety of long Term Interim Storage Facilities, May, Munich, Germany 2013 Der Halden-Reaktor- Experimentelle Möglichkeiten und Beispiele von Ergebnissen unter besonderer Berücksichtigung von Fragen zum Hochabbrand, Wolfgang Wiesenack. Presented at Reaktorsicherheit und technik der RWTH Aschen, Aachen, Germany, January 2013 Water chemistry measurements in the Halden Reactor and its experimental circuits, Peter Bennett and Kari Lye Moum. Presented at the 16th International Conference on the Properties of Water and Steam, University of Greenwich, UK, 1-5 September Fuel Fragmentation, Relocation and Dispersal under LOCA Conditions: Observations, M. Flanagan, B. C. Oberlander and A. Puranen. Presented at TOPFUEL 2013, Charlotte, North Carolina, September 15-19, 2013 Overview of PCMI results from the Halden Reactor Project, Wolfgang Wiesenack. Presented at Towards nuclear fuel modelling in the various types across Europe, Karlsruhe, Germany, June 2013