This report describes the first full-length high-temperature test (FLHT-1) performed by Pacific Northwest Laboratory (PNL) in the National Research Universal (NRU) reactor at Chalk River, Ontario, Canada. The test is part of a series of experiments being performed for the NRC as a part of their Severe Fuel Damage Program and is one of several planned for PNL`s Coolant Boilaway and Damage Progression Program. The report summarizes the test design and test plan. it also provides a summary and discussion of the data collected during the test and of the photos taken during the post-test examination. All objectives for the test were met. The key objective was to demonstrate that severe fuel damage tests on full-length fuel bundles can be safely conducted in the NRU reactor.

This is a report on one of a series of experiments to simulates a loss-of-coolant accident (LOCA) using full-length fuel rods for pressurized water reactors (PWR). The experiments were conducted by Pacific Northwest Laboratory (PNL) under the LOCA simulation Program sponsored by the US Nuclear Regulatory Commission (NRC). The major objective of this program was causing the maximum possible expansion of the cladding on the fuel rods from a short-term adiabatic temperature transient to 1200 K (1700 F) leading to the rupture of the cladding; and second, by reflooding the fuel rods to determine the rate at which the fuel bundle is cooled.

Hazardous conditions associated with performing the Full-Length High- Temperature (FLHT). Severe Fuel Damage Test No. 2 experiment have been analyzed. Major hazards that could cause harm or damage are (1) radioactive fission products, (2) radiation fields, (3) reactivity changes, (4) hydrogen generation, (5) materials at high temperature, (6) steam explosion, and (7) steam pressure pulse. As a result of this analysis, it is concluded that with proper precautions the FLHT- 2 test can be safely conducted.

Battelle, Pacific Northwest Laboratory is conducting severe fuel damage experiments for the U.S. Nuclear Regulatory Commission to provide data that are being used to bench-mark or validate computer codes like SCDAP and MELPROG. These experiments are designed to study the early phase of a light water reactor (LWR) core uncovery accident from the time of initial uncovery (boil-away) until peak temperatures reach [approximately]2500K. A description of four in-reactor, full axial length, high-temperature (FLHT) LWR fuel bundle tests is given including the geometries and test conditions. The test conditions increase in severity from test to test by increasing the peak temperature attained in the fuel bundle and/or by increasing the operating period at the peak temperatures. Thus the degree of damage progresses from slight in the early test to extensive in later tests. Salient features of these tests include the use of twelve 3.7-m-long fresh and preirradiated fuel rods that are held in position by standard 17 [times] 17 (either Inconel or Zircaloy-4) grid spacers. Two of the fuel bundles were operated for brief periods at typical commercial power ratings just prior to the boil-away transient.

This report presents the final safety analysis for the preparation, conduct, and post-test discharge operation for the Full-Length High Temperature Experiment-5 (FLHT-5) to be conducted in the L-24 position of the National Research Universal (NRU) Reactor at Chalk River Nuclear Laboratories (CRNL), Ontario, Canada. The test is sponsored by an international group organized by the US Nuclear Regulatory Commission. The test is designed and conducted by staff from Pacific Northwest Laboratory with CRNL staff support. The test will study the consequences of loss-of-coolant and the progression of severe fuel damage.

This document presents an assessment of the severe accident phenomena observed from four Full-Length High-Temperature (FLHT) tests that were performed by the Pacific Northwest Laboratory (PNL) in the National Research Universal (NRU) reactor at Chalk River, Ontario, Canada. These tests were conducted for the US Nuclear Regulatory Commission (NRC) as part of the Severe Accident Research Program. The objectives of the test were to simulate conditions and provide information on the behavior of full-length fuel rods during hypothetical, small-break, loss-of-coolant severe accidents, in commercial light water reactors.

GT2R2 calculates the thermal behavior of a nuclear fuel rod during normal steady-state operation. The program was developed as a tool for estimating fuel-cladding gap conductances and fuel-stored energy. Models used include power history, fission gas generation and release, fuel relocation and densification, and fuel-cladding gap conductance. The gas release and relocation models can be used to make either best-estimate or conservative predictions. The code is used by the United States Nuclear Regulatory Commission for audit calculations of nuclear fuel thermal performance computer codes.

Computations of power history effects on the pre-loss-of-coolant accident (LOCA) conditions of generic pressurized water reactor (PWR) and boiling water reactor (BWR) fuel rods were performed at Pacific Northwest Laboratory using the U.S. Nuclear Regulatory Commission (NRC) code FRAPCON-2. Comparisons were made between cases where the fuel operated at a high ( 11 LOCA-limited") power throughout life (20,000 MWd/MTU) and those where the fuel was at a lower power for most of its burnup and ramped to the high power at 10,000 or 20,000 MWd/MTU burnup. The PWR rod was calculated to have more cladding creepdown during the lower power cases, which resulted in slightly lower centerline temperatures (as much as 100{degrees}C). This result was insensitive to the method used to increase the power during the ramps (i.e., by increasing the average rod power or by changing the peak-to-average (P/A} ratio of the axial power shape). The calculations also indicate that the highest fuel centerline temperatures were reached at startup. The BWR rod, however, demonstrated a substantial dependence on the power history. In this case, the constant high-power rod released considerably more fission gas than the lower power cases (21% versus 0.4%), which resulted in temperature differences of up to 350°C. The hiqhest temperature was reached at end-of-life (EOL) in the constant high-power case.

The ANS5.4 committee developed fission gas release models for stable and radioactive species for both low and high temperatures. This report reproduces these models and presents a FORTRAN coding of the models. The coding comprises subroutine ANS54 which is in a format suitable for direct insertion into a code like FRAPCON. Also shown are comparisons of model predictions with light water reactor low burnup data and liquid metal fast breeder reactor high burnup data that were used in the model development.

A description is presented of the computer code GAPCON THERMAL-2, a light water reactor (LWR) fuel thermal performance prediction code. GAPCON- THERMAL-2, is intended to be used as a calculational tool for reactor fuel steady- state thermal performance and to provide input for accident analyses. Some models used in the code provide best estimate as well as conservative predictions. Each of the individual models in the code is based on the best available data. (auth)

Page 3 4 5 uo 2 -4 wt% Puo 2 Fuel.6 Longitudinal Section of the Bottom End Region of a Vibrationally Compacted uo 2 -Pu0 2 PRTR Fuel Rod.Transverse Sections at Different Planes Along the Length of a Vibrationally Compacted uo 2 -4 wt% Puo 2 PRTR Fuel Rod.Photomosaic of a Transverse Section Through an Intentionally Defected 00 2 -2 wt% Pu0 2 PRTR Fuel Rod.Transverse Sections Through an Intentionally Defected and a Nondefected Vibrationally Compacted uo 2 -2 wt% Pu0 2 PRTR Fuel Rod.Cladding Rupture in a Vibrationally Compacted uo 2 -4 wt% Pu0 2 PRTR Fuel Rod.Transverse Section from the Upper Portion of the Ruptured Intentionally 12 Defected PRTR Fuel Rod.Fission Product and Plutonium Distribution in the Intentionally Defected uo 2 -2 wt% Puo 2 PRTR Fuel Rod Shown in Figures 7 and 8.Fission Gas Release as a Function of