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At least 361 records · Page 20
All sourcesNASA NTRSDOE OSTI
NASA NTRSDate not supplied

Plans for Ice Crystal Icing Tests Using a 3D Heated Test Article at the NASA Icing Research Tunnel

This presentation describes the plans for ice crystal icing and supercooled water icing tests to be conducted in Fall/Winter 2021 at the NASA Icing Research Tunnel. A test article, whose geometry is representative of an inter-compressor duct and strut region of a turbofan engine, has been designed and fabricated for the study of ice crystal icing physics tests. The Simulated Inter-compressor Duct Research Model (SIDRM) will have the ability to heat surfaces to simulate the warm surfaces of the inter-compressor duct within an engine. The test article is instrumented with heaters, heat flux gauges, thermocouples, and pressure taps to measure icing behavior. The aim of these tests is to generate ice shapes on the SIDRM test article under well-characterized conditions. The results of the testing with the SIDRM model will be used to provide icing physics and validation data for development of the engine icing simulation capabilities in GlennICE, the 3D ice prediction tool. This is intended to be a presentation only, and does not have an accompanying paper.

ice-crystal icing↗
DOE OSTI2019

Test Plan for Reirradiation and 1600°C Heating Test of AGR-3/4 Compact 3-1

Post irradiation examination (PIE) and heating tests of tristructural isotropic (TRISO)-coated fuel particles, fuel compacts, and graphite from Advanced Gas Reactor (AGR) irradiation experiment AGR 3/4 are in progress. The AGR-3/4 test train consisted of twelve separate capsules irradiated in the northeast flux trap of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) from December 14, 2011, to April 12, 2014 (Collin 2015b). This test plan describes the reirradiation and subsequent heating test of AGR-3/4 Compact 3-1. One of the main goals of this test is to assess the retention of short-lived fission products I-131 (t1/2 = 8.02 d) and Xe-133 (t1/2 = 5.2 d) in fuel kernels. To generate short-lived fission products, the compact will first be reirradiated in the Neutron Radiography Reactor (NRAD). Following reirradiation, the compact will be removed from NRAD, gamma scanned on the Precision Gamma Scanner (PGS), and then loaded into the Fuel Accident Condition Simulator (FACS) furnace for heating at 1600°C. Fission gases and condensable fission products released from the compact will be collected in the Fission Gas Monitoring System (FGMS) and the FACS condensation plates, respectively. The planned uses of AGR-3/4 compacts (including heating testing and destructive analysis) are discussed in the AGR-3/4 PIE plan (Demkowicz 2017).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Fabrication and Testing of DOE Standard Canister Closure Leak Test Assembly

DOE manages over 300 types of spent nuclear fuel (SNF), many of which are located at the Idaho National Laboratory (INL) site. Managing this large variety of SNF for storage, transportation, and disposal poses a challenge to DOE. The Idaho Cleanup Project and INL are collaborating on the DOE SNF Road-Ready Demonstration (“Road-Ready Demonstration”), which will develop and demonstrate the designs, technology, processes, and regulatory framework for packaging DOE-managed SNF for “road-ready dry storage.” Road-ready dry storage is an SNF management concept in which SNF is packaged into dry, sealed canisters that are then placed in onsite storage in anticipation of later transport and disposition. The forward-looking goal of the Road-Ready Demonstration is to establish the foundation for a large-scale road-ready dry storage program at the INL site. In support of establishing a large-scale road-ready dry storage program at the INL site, the Road-Ready Demonstration will first package Fort St. Vrain SNF currently in dry storage at INL into several DOE Standard Canisters (DOESCs). These DOESCs will in turn be loaded into another containment similar to commercial multi-purpose canisters. This multi-purpose canister will then be compatible with a transportation or storage system, such as a storage cask for interim storage or transportation package for offsite transport. These DOESCs will remain sealed over the course of their storage, transportation, and applicable disposal functions. The closure process for the DOESC will include fuel and basket loading, welding, inspection, leak testing, and, if needed, repair. As a follow-up to previous discussions on the design of the DOE Closure Leak Test Assembly (LTA), this report describes recent fabrication and testing efforts performed at INL. DOESCs are sealed by two sequential gas tungsten arc welds, both of which are performed by remotely operated and semiautomatic welding systems. The first weld is a circumferential pipe weld that completes the assembly of the canister body and lid assembly. The second and final closure weld attaches the vent plug to the vent socket via a butt joint. After the second weld is performed, the welds are helium leak tested using an evacuated envelope technique. The LTA was designed for both remote and manual operation. This report describes the fabrication and performance testing associated with the evacuated envelope technique. INL staff designed, fabricated, and tested the LTA at INL facilities. This testing included establishing technique and system sensitivities in accordance with ASME and American National Standards Institute N14.5 requirements. Forthcoming work will cover such areas as design optimization, process and personnel qualification, and implementation in Road-Ready Demonstration operations.

42 ENGINEERING↗
DOE OSTI2024

Fabrication and Testing of DOE Standard Canister Closure Leak Test Assembly

DOE manages over 300 types of SNF, most of which are located at the INL site. “Road-ready dry storage” is a management concept where SNF is packaged into dry, sealed canisters, which are then placed in on-site storage in anticipation of later removal. The Idaho Cleanup Project and INL are collaborating on the Road-Ready Capability Demonstration Project, which will develop and demonstrate the designs, technology, processes, and regulatory framework for packaging DOE SNF for road-ready dry storage. In support of establishing a large-scale road-ready dry storage program at the INL site, the demonstration will package a select amount of DOE-managed SNF into DOE Standard Canisters. The closure process for the DOE Standard Canisters will include fuel and basket loading, welding, inspection, leak testing, and if needed, repair. As a follow-up to previous discussion on the design of the DOE Closure Leak Test Assembly, this report describes recent fabrication and testing efforts performed at INL. DOE Standard Canisters are sealed by two sequential gas tungsten arc welds. Both are performed by remotely operated and semi-autonomous welding systems. The first weld is a circumferential pipe weld that completes assembly of the canister body and lid assembly. The second and final closure weld connects the vent plug to the vent port with an identical butt joint to the circumferential pipe weld. After the second weld is performed on the vent port, these welds are helium leak tested using an inside-out technique. In addition to the commercially available vacuum and leak detector systems, the DOE Standard Canister Closure Leak Test Assembly was designed for both remote and manual operation. This report describes fabrication and performance testing associated with the inside-out technique. INL staff designed and tested systems to accomplish these tasks. Hardware fabrication occurred at INL facilities. Forthcoming work includes design optimization, integration to existing systems, and implementation to packaging demonstration operations.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Fabrication and Testing of DOE Standard Canister Closure Leak Test Assembly – 24139

DOE manages over 300 types of spent nuclear fuel (SNF), many of which are located at the Idaho National Laboratory (INL) site. Managing this large variety of SNF for storage, transportation, and disposal poses a challenge to DOE. The Idaho Cleanup Project and INL are collaborating on the DOE SNF Road-Ready Demonstration (“Road-Ready Demonstration”), which will develop and demonstrate the designs, technology, processes, and regulatory framework for packaging DOE-managed SNF for “road-ready dry storage.” Road-ready dry storage is an SNF management concept in which SNF is packaged into dry, sealed canisters that are then placed in onsite storage in anticipation of later transport and disposition. The forward-looking goal of the Road-Ready Demonstration is to establish the foundation for a large-scale road-ready dry storage program at the INL site. In support of establishing a large-scale road-ready dry storage program at the INL site, the Road-Ready Demonstration will first package Fort St. Vrain SNF currently in dry storage at INL into several DOE Standard Canisters (DOESCs). These DOESCs will in turn be loaded into another containment similar to commercial multi-purpose canisters. This multi-purpose canister will then be compatible with a transportation or storage system, such as a storage cask for interim storage or transportation package for offsite transport. These DOESCs will remain sealed over the course of their storage, transportation, and applicable disposal functions. The closure process for the DOESC will include fuel and basket loading, welding, inspection, leak testing, and, if needed, repair. As a follow-up to previous discussions on the design of the DOE Closure Leak Test Assembly (LTA), this report describes recent fabrication and testing efforts performed at INL. DOESCs are sealed by two sequential gas tungsten arc welds, both of which are performed by remotely operated and semiautomatic welding systems. The first weld is a circumferential pipe weld that completes the assembly of the canister body and lid assembly. The second and final closure weld attaches the vent plug to the vent socket via a butt joint. After the second weld is performed, the welds are helium leak tested using an evacuated envelope technique. The LTA was designed for both remote and manual operation. This report describes the fabrication and performance testing associated with the evacuated envelope technique. INL staff designed, fabricated, and tested the LTA at INL facilities. This testing included establishing technique and system sensitivities in accordance with ASME and American National Standards Institute N14.5 requirements. Forthcoming work will cover such areas as design optimization, process and personnel qualification, and implementation in Road-Ready Demonstration operations.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Demonstration of the surveillance test article for an advanced reactor surveillance test program

Operational environments in generation IV reactors involve corrosive and irradiative conditions at elevated temperatures. Typical reactor operations consist of transients which impose cyclic loads on reactor components. These cyclic loads, combined with corrosive and irradiative environments, result in synergistic degradation of component materials. However, limited data exists on the coupled damage effects on materials for reactor environments. While the surrogate material surveillance concept has been used in light water reactors to assess irradiation damage, existing material surveillance technologies are not suitable for in-situ monitoring of coupled material degradation. The materials surveillance program focuses on material degradation management and the estimation of remaining life of reactor components through surveillance test articles. This paper presents the design and analysis methodology of a bi-metal surveillance test article, which uses difference in thermal expansion coefficient between two metals to induce in-situ cyclic loads. This report presents the work conducted in FY 25, to test the surveillance test article in air and salt environments. These test specimens were evaluated after thermal cycle exposure and remaining life is measured through creep test.

36 - MATERIALS SCIENCE↗
DOE OSTI2025

Development of a Miniature Small Punch Creep Test for Post Surveillance Test Specimens

Summary: This poster outlines a new small punch creep test (SPCT) for examining the mechanical properties of post surveillance test specimens. A surveillance test article was previously designed to study the effects of Molten-Salt conditions on the structural materials in the reactor. To better understand the post-test mechanical properties of the specimen, a miniature SPCT is proposed. The principle, testing standards and challenges of the method is introduced in the poster. An initial design of the testing apparatus is shown, and an experiment plan is outlined.

36 MATERIALS SCIENCE↗
DOE OSTI2014

First elevated-temperature performance testing of coated particle fuel compacts from the AGR-1 irradiation experiment

In the AGR-1 irradiation experiment, 72 coated-particle fuel compacts were taken to a peak burnup of 19.5% fissions per initial metal atom with no in-pile failures. This paper discusses the first post-irradiation test of these mixed uranium oxide/uranium carbide fuel compacts at elevated temperature to examine the fuel performance under a simulated depressurized conduction cooldown event. A compact was heated for 400 h at 1600 degrees C. Release of 85Kr was monitored throughout the furnace test as an indicator of coating failure, while other fission product releases from the compact were periodically measured by capturing them on exchangeable, water-cooled deposition cups. No coating failure was detected during the furnace test, and this result was verified by subsequent electrolytic deconsolidation and acid leaching of the compact, which showed that all SiC layers were still intact. However, the deposition cups recovered significant quantities of silver, europium, and strontium. Based on comparison of calculated compact inventories at the end of irradiation versus analysis of these fission products released to the deposition cups and furnace internals, the minimum estimated fractional losses from the compact during the furnace test were 1.9 x 10-2 for silver, 1.4 x 10-3 for europium, and 1.1 x 10-5 for strontium. Other post-irradiation examination of AGR-1 compacts indicates that similar fractions of europium and silver may have already been released by the intact coated particles during irradiation, and it is therefore likely that the detected fission products released from the compact in this 1600 degrees C furnace test were from residual fission products in the matrix. Gamma analysis of coated particles deconsolidated from the compact after the heating test revealed that silver content within each particle varied considerably; a result that is probably not related to the furnace test, because it has also been observed in other as-irradiated AGR-1 compacts. X-ray imaging of selected particles was performed to examine the internal microstructure. This examination revealed variable irradiation performance of the coating layers, but sufficient statistical sampling is not yet available to identify any possible correlation to variation in individual particle fission product retention.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2018

Biochemical methane potential (BMP) tests: Reducing test time by early parameter estimation

Highlights: • Early parameters prediction of BMP model is achieved by using sensitivity functions. • Minimum testing time for early parameter estimation is related to kinetic constant. • Balanced regression improves the accuracy of estimated parameters. - Abstract: Biochemical methane potential (BMP) test is a key analytical technique to assess the implementation and optimisation of anaerobic biotechnologies. However, this technique is characterised by long testing times (from 20 to >100 days), which is not suitable for waste utilities, consulting companies or plants operators whose decision-making processes cannot be held for such a long time. This study develops a statistically robust mathematical strategy using sensitivity functions for early prediction of BMP first-order model parameters, i.e. methane yield (B{sub 0}) and kinetic constant rate (k). The minimum testing time for early parameter estimation showed a potential correlation with the k value, where (i) slowly biodegradable substrates (k ≤ 0.1 d{sup −1}) have a minimum testing times of ≥15 days, (ii) moderately biodegradable substrates (0.1 −1 ) have a minimum testing times between 8 and 15 days, and (iii) rapidly biodegradable substrates (k ≥ 0.2 d{sup −1}) have testing times lower than 7 days.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Design, testing and emplacement of sand-bentonite for the construction of a gas-permeable seal test (gast)

The main aim of a gas-permeable seal is to increase the gas transport capacity of the backfilled underground structures without compromising the radionuclide retention capacity of the engineered barrier system or the host rock. Such a seal, proposed by NAGRA as part of the 'Engineered Gas Transport System' in a L/ILW repository, considers specially designed backfill and sealing materials such as sand/bentonite (S/B) mixtures with a bentonite content of 20- 30%. NAGRA's RD and D plan foresees demonstrating the construction and performance of repository seals and improving the understanding and the database for reliably predicting water and gas transport through these systems. The fluid flow and gas transport properties of these backfills have been determined at the laboratory scale and through modelling the maximum gas pressures in the near field of a repository system and the gas flow rates have been evaluated. Within this context, the Gas-permeable Seal Test (GAST) was constructed at Grimsel Test Site (GTS) to validate the effective functioning of gas-permeable seals at realistic scale. The intrinsic permeability of such seals should be in the order of 10-18 m2. Because the construction of S/B seals is not common practice for construction companies, a stepwise approach was followed to evaluate different construction and quality assurance methods. As a first step, an investigation campaign with simple tests in the laboratory and in the field followed by 1:1 scale pre-tests at GTS was performed. Through this gradual increase of the degree of complexity, practical experience was gained and confidence in the methods and procedures to be used was built, which allowed reliably producing and working with S/B mixtures at a realistic scale. During the whole pre-testing phase, a quality assurance (QA) programme for S/B mixtures was developed and different methods were assessed. They helped to evaluate and choose appropriate emplacement techniques and methodologies to achieve the target S/B dry density of 1.70 g/cm{sup 3}, which results in the desired intrinsic permeability throughout the experiment. The final QA methodology was targeted at engineering measures to decide if the work can proceed, and at producing high resolution material properties database for future water and gas transport modelling activities. The different applied QA techniques included standard core cutter tests, the application of neutron-gamma (Troxler) probes and two mass balance methods (2D and 3D). The methods, looking at different representative scales, have provided only slightly different results and showed that the average density of the emplaced S/B plug was between 1.65 and 1.73 g/cm{sup 3}. Spatial variability of dry densities was observed at decimeter scale. Overall, the pre-testing and QA programme performed for the GAST project demonstrated how the given design criteria and requirements can be met by appropriately planning and designing the material emplacement. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

TEST REPORT #21: Compressor Calorimeter Test of R-404A Alternatives ARM-31a, D2Y-65, L-40, and R32/ R-134a (50/50)

As a contribution to the AHRI Low-GWP Alternative Refrigerants Evaluation Program (AREP), this study compares the performance of four lower-GWP alternative refrigerants, ARM-31a, D2Y-65, L-40, and R-32 + R-134a mixture, to that of refrigerant R-404A (baseline) in a scroll compressor designed for medium temperature refrigeration applications. These comparisons were carried out via compressor calorimeter tests performed on a compressor designed for refrigerant R-404A and having a nominal rated capacity of 23,500 Btu/hr. Tests were conducted over a suction dew point temperature range of -10°F to 35°F in 5°F increments and a discharge dew point temperature range of 70°F to 140°F in 10°F increments. All the tests were performed with 20°F superheat, 40°F superheat, and 65°F suction temperature. A liquid subcooling level of 10°F to 15°F was maintained for all the test conditions. However, the cooling capacities reported in this study are normalized for 0°F subcooling. The tests showed that the compressor energy efficiency ratio (EER) and cooling capacity with all four alternative refrigerants tested are higher at higher saturation suction and saturation discharge temperature and lower at lower saturation suction and saturation discharge temperature, compared to that of R-404A. Discharge temperatures of all the alternative refrigerants were higher than that of R-404A at all test conditions.

Alternative Refrigerants↗
DOE OSTI2017

Shakedown Tests for Refurbished and Upgraded Frames and Initiation of Alloy 709 Creep Rupture Tests

This report describes the shakedown tests conducted on the upgraded frames, and initiation of creep rupture tests on refurbished frames. SS316H, a reference material for Alloy 709, was used in shakedown tests, and the tests were conducted at 816 degree C under three stress levels to accumulate 1% creep strain. 1/4” gage diameter specimen design was used. The creep rupture tests on Alloy 709 were initiated at 600 degree C under 330 MPa to target 1,500 h rupture time. 12 specimens with 3/8” gage diameter were prepared from the materials with 6 heat treatment conditions, 2 from each. The required mechanical load under 330MPa was calculated to be 5,286 lb for the 3/8” gage diameter specimen. Among the ART frames, 7 frames are equipped with 10,000 lb load cell including #5 to 8 and #88 to 90, and can be used. 7 tests were thus started in this stage of project, and remaining 5 will be continued whenever any of the 7 tests is completed.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Rail-Cask Tests: Normal-Conditionsof- Transport Tests of Surrogate PWR Fuel Assemblies in an ENSA ENUN 32P Cask.

This report describes tests conducted using a full-size rail cask, the ENSA ENUN 32P, involving handling of the cask and transport of the cask via truck, ships, and rail. The purpose of the tests was to measure strains and accelerations on surrogate pressurized water reactor fuel rods when the fuel assemblies were subjected to Normal Conditions of Transport within the rail cask. In addition, accelerations were measured on the transport platform, the cask cradle, the cask, and the basket within the cask holding the assemblies. These tests were an international collaboration that included Equipos Nucleares S.A., Sandia National Laboratories, Pacific Northwest National Laboratory, Coordinadora Internacional de Cargas S.A., the Transportation Technology Center, Inc., the Korea Radioactive Waste Agency, and the Korea Atomic Energy Research Institute. All test results in this report are PRELIMINARY – complete analyses of test data will be completed and reported in FY18. However, preliminarily: The strains were exceedingly low on the surrogate fuel rods during the rail-cask tests for all the transport and handling modes. The test results provide a compelling technical basis for the safe transport of spent fuel.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Length Dependence of Severe Accident Test Station Integral Testing

The safety of accident-tolerant fuel (ATF) candidates must be tested under simulated loss-of-coolant accident (LOCA) conditions to evaluate their improved safety margins. This evaluation eventually must include testing of irradiated materials. This report describes the fiscal year (FY) 2019 activities to (1) improve the information yield of integral design–basis LOCA tests by using digital image correlation (DIC) calculations on the engraved specimens’ surfaces and (2) to investigate the effects of reducing the length of nuclear-grade FeCrAl specimens tested in the Severe Accident Test Station (SATS) to improve the costeffectiveness of irradiation campaigns for future work. Analysis showed the DIC technique to be effective at calculating strains, and that specimens as short as 4 inches could be used for LOCA experiments while maintaining similar balloon and burst behavior as 12-inch specimens. With these results, an irradiation vehicle design effort was initiated to encapsulate multiple 4-inch thin-walled tube cladding specimens and test them under irradiation in the flux trap of the High Flux Isotope Reactor. By utilizing thermal modeling and optimizing the design parameters, calculated specimen temperatures were predicted to vary less than 2% during irradiation. Future work is planned to finalize the design and safety basis approvals and initiate a HFIR irradiation in support of in-cell integral LOCA tests at the Irradiated Fuels Examination Laboratory (IFEL).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Mechanisms Engineering Test Loop - Flexicask Design and Testing Report

This report documents the design development and testing of the Flexicask system for the Mechanisms Engineering Test Loop (METL). Currently, METL has completed Phase I construction and has been operating for over a year. The Flexicask system (Figure 1) was designed to allow for the removal and insertion of test articles into a METL test vessel while maintaining an acceptable inert atmosphere to minimize contamination of the METL sodium. This report discusses the design development of the Flexicask system, the commissioning and testing of the system, and the initial use of the Flexicask system to remove the Gear Test Assembly (GTA) from the METL Test Vessel 1.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Mechanisms Engineering Test Loop (METL) Operations and Testing Report (FY2021)

This report documents the operations and testing that was performed at the Mechanisms Engineering Test Loop (METL) during FY2021. The METL facility had a very successful third year of operations having logged about 944 days of operations with molten sodium either in a flowing or static condition. Operations were paused in April 20, 2021 to accommodate the refurbishing of building 308’s alkali metal passivation booth and scrubber (AMPB&S). METL is being prepared to support a second larger experiment, the Thermal Hydraulic Experimental Test Article (THETA) and expects to resume the Gear Test Assembly (GTA) testing. The technology development team is also developing two additional experiments, a gripper test article and a flow sensor test article that are expected to go into METL in late FY22 or early FY23.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Report on Thermal Cycling Testing and Bimaterial Weld Development for a Passively Actuated Materials Surveillance Test Article

This report describes the continuation and conclusion of a thermal cycling test program designed to evaluate a family of passively actuated material surveillance test articles. Additionally, the report details a preliminary study of bimetallic welding between 316 stainless steel and TZM (tungsten-zirconium-molybdenum) alloy. The test program was ultimately successful in demonstrating the basic viability of the passively actuated test article concept as well as specifically validating the procedures developed to size specimens to mimic the mechanical response of operating components. The report discusses and summarizes the results of the complete test program which tested samples for more than 1,000 controlled thermal cycles. The weld development program included both stir friction and electron beam joining. The stir friction welding was ultimately successful, while the electron beam welding would require additional development. Ultimately, these test programs pave the way for implementing a materials surveillance program in future molten salt reactors (MSRs). The report conclusions outline a plan for transforming the research and developed described here into a definite set of practices and procedures for a surveillance program at a future operating reactor.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Test Report for DPP-1 Regulatory Compliance Testing. Volume 1 - Main Report

Consolidated Nuclear Security, LLC (CNS), the management and operating contractor for the Y-12 National Security Complex (Y-12) for the National Nuclear Security Administration (NNSA), is currently responsible for the regulatory testing of a new Type B fissile material shipping package called the Defense Programs Package (DPP)-1. The DPP-1 is a drum packaging with an inner liner and a removable lid. Both the drum body and the lid are filled with an impact-limiting, thermal-insulating material that protects the inner containment vessel (CV) during normal conditions of transport (NCT) and hypothetical accident conditions (HAC) as defined by 10 US code of Federal Regulations (CFR) 71. In fiscal year 2021, the Oak Ridge National Laboratory (ORNL) Package Testing Program (PTP), located at the National Transportation Research Center (NTRC), conducted Type B performance tests as prescribed in Title 10 Code of Federal Regulations Part 71 (10 CFR 71) on seven DPP-1 prototype units. This test report documents test unit (TU) preparation, pretest condition, conditioning of each TU, NCT and HAC testing, and posttest measurements and observations made of the damage resulting from these tests.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗