Los Alamos National Laboratory is a participant in the 2D/3D program. Activities conducted at Los Alamos National Laboratory in support of 2D/3D program goals include analysis support of facility design, construction, and operation; provision of boundary and initial conditions for test-facility operations based on analysis of pressurized water re~ctors; performance of pretest and posttest predictions and analyses: and use of experimental results to validate and assess the singleand multi-dimensional, nonequilibrium features in the Transient Reactor Analysis Code (T RAC). During fiscal year 1987, Los Alamos conducted analytical assessment activities using data from the Slab Core Test Facility, the Cylindrical Core Test Facility, and the Upper F’ienum Test Facility. Finally, Los Alamos continued work to provide TRAC improvements. in this paper, Los Alamos activities during fiscal year 1987 will be summarized; several significant accomplishments will bc described in more detail to illustrate the work activities at Los Alamos. .— INTRODUCTION The 2D/3D program is sponsored jointly by Japan, the Federal Republic of Germany, and the U,lited States (US). The safety-related objectives of the 2D/3D program are as follows: first, to provide an improved understanding ot the effectiveness of various emergency corecooling (ECC) systems in Iimlting peak fuel rod cladding temperatures during vessel refill and core reflood for mediumtc large-break loss-of-coolant ac;idents (LOCAS) in pressurized water reactors (P WRS); second, to reveal core-coolant inventory and system flow characteristics during the refill and reflood phas~s of a medium to large-break LOCA; third, to study convective flow and temperature distributions inside a heated core during reflood for a mediumto large break LOCA; fourth, to assess the predictive capability of best-estimate computer codes and the conservatism of evaluation model computer codes; ond fifth, to obtain informatio~l which * Work performed under the auspices of the US Nuclear f?eEulatory Conlmissior~ may be used to improve thermal-hydraulic models In best-estimate, evaluation-model and other computer codes. Activities conducted at Los Alamos National Laboratory in support of 2D/3D program goals include analysis support of faci!lty design, construction, and operation: provision of boundary ~nd initial conditions for test facility operations based on analysis of PWRS: performance of pretest and postt~st predictions and analyses: and use of experimental results to validate and assess the singleand multidimensional, nonequilibrium features in the Transient Reactor Analysis Code (TRAC), Three experimental facilities provide data to 2D/3D program participants. The Slab Core Test Facility (SCTF) is a separate-effects reflood facility located in Japan. It models a full-height 1 /21-scale section of the core, one fuel element wide from core centerline to outer periphery. This facility began testing with its third electrically heated core during 1986; Los Alamos will continue analysis of Core-ill tests in FY-1988. The Cylindrical Core Test Facility (CCTF) is an approximately 1/21-scale facility, also located in Japan; this facility has completed its test program and the Los Alamos counterpart analysis program is nearing completion. The Upper Plenum Test Facility (U PTF), located in the Federal Republic of Germany, is a l/1-scale integral test facility focusing on phenomena in the downcomer, lower plenum, upper plenum, and primary-system loops of a PWR. Los Alamos analytical efforts to date have largely supported test design and specification: posttest analyses of UPTF experiments have started and will be emphasized during FY-1988; in part because of the importance of these efforts in supporting an effort to quantify the uncertainty associated witfi TRAC predictions of peak cladding temperatures for large-break LOCAS. During FY-1987, Los Alamos conducted analytical assessment activities using data from the SCTF, CCTF, and UPTF facilities, Finally, Los Alamos continued dforts to provide improvements to the TRAC code, This paper summarizes Los Alamos activities during F Y1987: several significant accomplishments are described in more detail to illustrate the work activities at Los Alamos, TRAC ASSESSMENT ACTIVITIES A few comments are appropriate to introduce the summary comments that will be provided to describe our TRAC-PF1 /MOD1 (Ref. 1 ) assessment activities. When performing a code assessmer~t, understanding must be developeJ, cataloged, and reported in three vital areas, These three areas are (1) sufficiency of knowledge about the as-built arid asoperated state of the facility providing the data to be used for assessment, (2) the adequacy of the input model prepared to describe the facility, and (3) the adequacy of the closure models and correlations within the code, Now consider a situation in which some significant feature of the plant configuration or operation is either unknown or undetected by the individual performing a posttest cssessmcnt. This deficiency of knowledge will be reflected in the input model and in the calculatwf ~csult, ‘.acking knowledge that the deficiency exists, the analyst will tend to assign fault incorrectly to either the adequacy of the input model or the adequacy of the closure models and corre!;~t!ons within the code. Consider, for example, a second example in which the overall knowledge of tht’ facility and its operation is good; i.e,, the perception of th~ plant configuration and oper.~tiw] is accurate, but the calculated and measured resclts do not agrtw In significant respe~ t% 1 Iw ,. cause for the disagreement(s) can lie with wther the adequw y of the Input model or thv adequacy of the code closure models and correlations or with some combination of the two. Care must be taken to determine the cause Problems associated with the input model can frequently be remedied and user guidelines issued to alert others to the problem. Problems associated with the code closure models and correlations frequently require more effort to correct, A decision must be reached as to whether code model and correlation improvement should be attempted or whether the deficiency should be accepted as part of the quantified code uncertainty for related transients. As we summarize our CCTF, SCTF, and UPTF posttest assessment activities, we will attempt to use the framework identified above. It is hoped that this will provide a cohesive structure for identifying the ‘“lesr,ons learned”’ during these assessment activities. SCTF PROGRAM SUPPORT We will summarize results for three SCTF posttest assessment activities at L05 Alamos, Runs 704, 713, and 714. The versions of TRAC-PFl/MODl used for these analyses were version 13.1 for Run 704. 13.0 for Run 713, and 13.1 for Run 714. A detailed analysis discussion is provided for Run 704, with more concise discussions for Runs 713 and 714, For all figures in this paper comparing calculated and test results, the calculated results are shown as a solid line and the data as a dashed line, In addition, the figures frequently carry a legend identifying the JAERI identification number for that data item. In the legend, the corresponding TRAC calculated value carries the prefix “’C.” SCTF Run 704 (Ref. 2) is a German PWR (GPWR) evaluation-model integral orientation test. Run 704 was one of the first GPWR tests. Important features of the test specification included a blowdown of the initial primary pressure from 0.6 MPa to 0.3 MPa and multiple ECC-system injections into the cold leg, four locations in the upper plenum and two locations above the upper core support plate. Many interesting phenomena occurred during the test. The posttest calculation and analysis effort for Run 704 are also interesting because information was obtained in each of the three assessment areas identified above, Figure 1 displays the measured and calculated differential pressure in bundle 5 over the full core height, The differential pressure can also be considered as a direct measure of the liquid level in the bundle. During the test, the liquid level in the bundle generally increased until about 200 s, when the liquid level in the core was severely depressed. The liquid level subsequently began to recover at about 245 s, The TRAC-calculated liquid level trace showed a similar trend but was noticeably different in magnitude, In particular, the increase in liquid level stalled at about 170 s and the calculated depression in liquid level occurred l~ter {at about 220 s) and was deeper than measured, In fact, liquid displaced from the core passed into the lower plenum, up the downcomer, and out the broken cold leg on the pressure vesse! side, This liquid was lost from the system and not available for subsequent core cooling, One consequence of the greater Ilquid-level depression and loss of core coolant calculated by TRAC was a dryout and heating of the high-elevation cl.]dding not seen in the test, as shown in Fig, 2, A number of lessons were learned during the ccmrse of the posttest analysis, These are summarized below using the categories previously discussed. Overall, our knowledge of the ;ac.ility configuration and operation is very good However, the GPWR integral orientation test Run 704 was among the first III a new test ~eries h,~virlg conditions rllarkedly different than tests previously analyzed In particular, the quar,tlty of ECC Ilquid injected Into the upper plenum was Iargc; much of thif liquid was carried out

The objective of this document is to ease the task of adding new system components to the Transient Reactor Analysis Code (TRAC) or altering old ones. Sufficient information is provided to permit replacement or modification of physical models and correlations. Within TRAC, information is passed at two levels. At the upper level, information is passed by system-wide and component-specific data modules at and above the level of component subroutines. At the lower level, information is passed through a combination of module-based data structures and argument lists. This document describes the basic mechanics involved in the flow of information within the code. The discussion of interfaces in the body of this document has been kept to a general level to highlight key considerations. The appendices cover instructions for obtaining a detailed list of variables used to communicate in each subprogram, definitions and locations of key variables, and proposed improvements to intercomponent interfaces that are not available in the first level of code modernization.

Scaling has been identified as a particularly important element of the Severe Accident Research Program because of its relevance not only to experimentation, but also to analyses based on code calculations or special models. Recognizing the central importance of severe accident scaling issues, the United States Regulatory Commission implemented a Severe Accident Scaling Methodology (SASM) development program involving a lead laboratory contractor and a Technical Program Group to guide the development and to demonstrate its practicality via a challenging application. The Technical Program Group recognized that the Severe Accident Scaling Methodology was an integral part of a larger structure for technical issue resolution and, therefore, found the need to define and document this larger structure, the Integrated Structure for Technical Issue Resolution (ISTIR). The larger part of the efforts have been devoted to the development and demonstration of the Severe Accident Scaling Methodology, which is Component II of the ISTIR. The ISTIR and the SASM have been tested and demonstrated, by their application to a postulated direct containment heating scenario. The ISTIR objectives and process are summarized in this paper, as is its demonstration associated directly with the SASM. The objectives, processes and demonstration for the SASM are also summarized in the paper. The full body of work is referenced.

In September 1988, the United States Nuclear Regulatory Commission issued a revised emergency core cooling system rule for light water reactors that allows, as an option, the use of best estimate plus uncertainty methods in safety analysis. To support the 1988 licensing revision, the United States Nuclear Regulatory Commission and its contractors developed the code scaling, applicability and uncertainty evaluation methodology to demonstrate the feasibility of the best estimate plus uncertainty approach. The phenomena identification and ranking table (PIRT) process, Step 3 in the code scaling, applicability and uncertainty methodology, was originally formulated to support the best estimate plus uncertainty licensing option. Through further development and application, the PIRT process has shown additional utility as a robust means to establish safety analysis computer code phenomenological requirements in their order of importance to such analyses. The generic PIRT process, including typical and common illustrations from prior applications that promoted further development of the process, are described. Analysis of the results of the prior applications is also described. The analysis results provide information that can help guide future applications of the process in a graded approach based on phenomena relative importance. (C) 1998 Elsevier Science S.A. All rights reserved.

The Nuclear Regulatory Commission has issued a Regulatory Bulletin and accompanying Regulatory Guide (1.82, Rev. 2) which requires licensees of boiling water reactors to develop a specific plan of action (including hardware backfits, if necessary) to preclude the possibility of early emergency core cooling system strainer blockage following a postulated loss-of-coolant-accident. The postulated mechanism for strainer blockage is destruction of piping insulation in the vicinity of the break and subsequent transport of fragmented insulation to the wetwell. In the absence of more definitive information, the Regulatory Guide recommends that licensees assume a drywell debris transport fraction of 1.0. Accordingly, the Nuclear Regulatory Commission initiated research focused toward developing a technical basis to provide insights useful to regulatory oversight of licensee submittals associated with resolution of the postulated strainer blockage issue. Part of this program was directed towards experimental and analytical research leading to a more realistic specification of the debris transport through the drywell to the wetwell. To help focus this development into a cost effective effort, a panel, with broad based knowledge and experience, was formed to address the relative importance of the various phenomena that can be expected in plant response to postulated accidents that may produce strainer blockage. The resulting phenomena identification and ranking tables reported herein were used to help guide research. The phenomena occurring in boiling water reactors drywells was the specific focus of the panel, although supporting experimental data and calculations of debris transport fractions were considered.

This document briefly describes the elements of the Nuclear Regulatory Commission`s (NRC`s) software quality assurance program leading to software (code) qualification and identifies a test matrix for qualifying Transient Reactor Analysis Code (TRAC)-Pressurized Water Reactor Version (-P), or TRAC-P, to the NRC`s software quality assurance requirements. Code qualification is the outcome of several software life-cycle activities, specifically, (1) Requirements Definition, (2) Design, (3) Implementation, and (4) Qualification Testing. The major objective of this document is to define the TRAC-P Qualification Testing effort.

The Programmatic Environmental Impact Statement for Tritium Supply and Recycling considers several methods for the production of tritium. One of these methods is the Accelerator Production of Tritium. This report summarizes the design characteristics of APT including the accelerator, target/blanket, tritium extraction facility, and the balance of plant. Two spallation targets are considered: (1) a tungsten neutron-source target and (2) a lead neutron-source target. In the tungsten target concept, the neutrons are captured by the circulating He-3, thus producing tritium; in the lead target concept, the tritium is produced by neutron capture by Li-6 in a surrounding lithium-aluminum blanket. This report also provides information to support the PEIS including construction and operational resource needs, waste generation, and potential routine and accidental releases of radioactive material. The focus of the report is on the impacts of a facility that will produce 3/8th of the baseline goal of tritium. However, some information is provided on the impacts of APT facilities that would produce smaller quantities.

VELCOR is an integrated, engineering-level computer code that models the progression of severe accidents in light water reactor (LWR) nuclear power plants. The entire spectrum of severe accident phenomena, including reactor coolant system and containment thermal-hydraulic response, core heatup, degradation and relocation, and fission product release and transport is treated in MELCOR in a unified framework for both boiling water reactors (BWRs) and pressurized water reactors (PWRs). Its current uses include the estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. Independent assessment efforts have been successfully completed by the US and international MELCOR user communities. Most of these independent assessment efforts have been conducted to support the needs and fulfill the requirements of the individual user organizations. The resources required to perform an extensive set of model and integral code assessments are large. A prudent approach to fostering code development and maturation is to coordinate the individual assessment efforts of the MELCOR user community. While retaining individual control over assessment resources, each organization using the MELCOR code could work with the other users to broaden assessment coverage and minimize duplication. In recognition of these considerations, the US Nuclear Regulatory Commission (US NRC) has initiated the MELCOR Cooperative Assessment Program (MCAP), a vehicle for coordinating and standardizing the assessment practices of the various MELCOR users. In addition, the user community will have a forum to better communicate lessons learned regarding MELCOR applications, capabilities, and user guidelines and limitations and to provide a user community perspective on code development needs and priorities. This second Annual Report builds on the foundation laid with the first Annual Report.

A postulated double-ended guillotine break of a direct-vessel-injection line in an AP600 plant has been analyzed. This event is characterized as an intermediate break loss-of-coolant accident (IBLOCA). Most of the insights regarding the response of the AP600 safety systems to the postulated accident are derived from calculations performed with the TRAC-PFl/MOD2 code. However, complementary insights derived from a scaled experiment conducted in the ROSA facility, as well as insights based upon calculations by other codes, are also presented. The key processes occurring in an AP600 during a IBLOCA are primary coolant system depressurization, inventory depletion, inventory replacement via emergency core coolant injection, continuous core cooling, and long-term decay heat rejection to the atmosphere. Based upon the calculated and experimental results, the AP600 will not experience a core heat up and will reach a safe shutdown state using only safety-class equipment. Only the early part of the long-term cooling period initiated by In-containment Refueling Water Storage Tank injection was evaluated Thus, the observation that the core is continuously cooled should be verified for the latter phase of the long-term cooling period, the interval when sump injection and containment cooling processes are important.

The PIUS Advanced Reactor is a 640-MW(e) pressurized-water reactor developed by Asea Brown Boveri. A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity normally is controlled by the boron concentration in the coolant and the temperature of the moderator coolant. Analyses of five initiating events have been completed on the basis of calculations per formed with the system neutronic and thermal-hydraulic analysis code TRAC-PF1/MOD2. The initiating events analyzed are (1) reactor scram, (2) loss of off-site power (3) main steam-line break, (4) small-break loss of coolant, and (5) large-break loss of coolant. In addition to the baseline calculation for each sequence, sensitivity studies were performed to explore the response of the PIUS reactor to severe off-normal conditions having a very low probability of occurrence. The sensitivity studies provide insights into the robustness of the design.

A postulated double-ended guillotine break of an AP600 direct-vessel-injection line has been analyzed. This event is characterized as an intermediate-break loss-of-coolant accident. Most of the insights regarding the response of the AP600 safety systems to the postulated accident are derived from calculations preformed with the TRAC-PF1/MOD2 code. However, complementary insights derived from a scaled experiment conducted in the ROSA facility, as well as insights based upon calculations by other codes, are also presented. Based upon the calculated and experimental results, the AP600 will not experience a core heat up and will reach a safe shutdown state using only safety-class equipment. Only the early part of the long-term cooling period initiated by In-containment Refueling Water Storage Tank injection was evaluated. Thus, the observation that the core is continuously cooled should be verified for the later phase of the long-term cooling period when sump injection and containment cooling processes are important.

The CONTAIN code was developed by Sandia National Laboratories under the sponsorship of the US Nuclear Regulatory Commission (NRC) to provide integrated analyses of containment phenomena. It is used to predict nuclear reactor containment loads, radiological source terms, and associated physical phenomena for a range of accident conditions encompassing both design-basis and severe accidents. The code`s targeted applications include support for containment-related experimental programs, light water and advanced light water reactor plant analysis, and analytical support for resolution of specific technical issues such as direct containment heating. The NRC decided that a broad technical review of the code should be performed by technical experts to determine its overall technical adequacy. For this purpose, a six-member CONTAIN Peer Review Committee was organized and a peer review as conducted. While the review was in progress, the NRC issued a draft ``Revised Severe Accident Code Strategy`` that incorporated revised design objectives and targeted applications for the CONTAIN code. The committee continued its effort to develop findings relative to the original NRC statement of design objectives and targeted applications. However, the revised CONTAIN design objectives and targeted applications. However, the revised CONTAIN design objectives and targeted applications were considered by the Committee in assigning priorities to the Committee`s recommendations. The Committee determined some improvements are warranted and provided recommendations in five code-related areas: (1) documentation, (2) user guidance, (3) modeling capability, (4) code assessment, and (5) technical assessment.

An updated TRAC 80% large-break loss-of-coolant accident (LBLOCA) has been calculated for the Westinghouse AP600 advanced reactor design, The updated calculation incorporates major code error corrections, model corrections, and plant design changes. The 80% break size was calculated by Westinghouse to be the most severe large-break size for the AP600 design. The LBLOCA transient was calculated to 144 s. Peak cladding temperatures (PCTS) were well below the Appendix K limit of 1,478 K (2,200 F), but very near the cladding oxidation temperature of 1,200 K (1,700 F). Transient event times and PCT for the TRAC calculation were in reasonable agreement with those calculated by Westinghouse using their {und W}COBRA/TRAC code. However, there were significant differences in the detailed phenomena calculated by the two codes, particularly during the blowdown phase. The reasons for these differences are still being investigated. Additional break sizes and break locations need to be analyzed to confirm the most severe break postulated by Westinghouse.

The PIUS advanced reactor is a 640-MWe pressurized water reactor concept developed by Asea Brown Boveri. A unique feature of PIUS is the absence of mechanical control and shutdown rods. Reactivity is controlled by coolant boron concentration and the temperature of the moderator coolant. Los Alamos is supporting the US Nuclear Regulatory Commission`s preapplication review of the PIUS reactor. Baseline calculations of the PIUS design were performed for a loss of offsite power initiator using TRAC-PF1/MOD2. Additional sensitivity studies examined flow blockage and boron dilution events to explore the robustness of the PIUS concept for low-probability combination events following a loss of offsite power.

a four-volume set of documentation on TRAC-PF1/MOD2. This guide was developed to assist the TRAC programmer and contains information on the TRAC code and data structure, the TRAC calculational sequence, memory management, and various machine configurations supported by TRAC.

The PIUS advanced reactor is a 640-MWe pressurized water reactor concept developed by Asea Brown Boveri. A unique feature of PIUS is the absence of mechanical control and shutdown rods. Reactivity is controlled by coolant boron concentration and the temperature of the moderator coolant. Los Alamos supported the US Nuclear Regulatory Commission`s preapplication review of the PIUS reactor. Baseline calculations of the PIUS design were performed for active and passive reactor scrams using TRAC-PF1/MOD2. Additional sensitivity studies examined flow blockage and boron dilution events to explore the robustness of the PIUS concept for low-probability combination events following active-system scrams.