ML20148S777
| ML20148S777 | |
| Person / Time | |
|---|---|
| Issue date: | 11/06/1978 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1585, NUDOCS 7812040036 | |
| Download: ML20148S777 (56) | |
Text
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- $150TbN ACRS SUBCOMMITTEE MEETING ON AD1 e "
@ WASHINGTON, DC
- W SEPTEMBER 27, 1978
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PPR /"/n The ACRS Subcommittee on Advanced Reactors held an open mee,ing erw ten r 27.
1978 in Room 1046, 1717 H St., NW, Washington, DC. The purpose oi ..his ir et-ing was to review matters relating to the NRC-sponsored research o advar ed reactors. Notice of this meeting was published in the Federal Register o.
August 15, and September 12, 1978. Copies of these notices are included ns Attachment A. A list of the attendees for this meeting is included as Attachment B, and the schedule is included as Attachment C. Selec ted portions of the meeting handouts are included as Attachment D. A complete set of the meeting handouts has been included in the ACRS files.
No written :ta temen ts or requests for time to give oral statements were received from members of the publ ic .
The meeting was attended by Dr. M. Carbon, Subcommittee Chairman; Dr. Paul Shewmon, Subcommittee member; Dr. Robert Seale, ACRS consultant; Dr. Sidney Siegel, ACRS consultant; and Dr. Richard Savio, ACRS Staff.
Dr. Richard Savio was the Designated Federal Employee for this meeting.
The meeting was opened at 8:30 a.m. and was adjourned at 6:30 p.m. The meeting was held entirely in open session.
EXECUTIVE SESSION Dr. Carbon opened the Executive Session with a brief summary of the schedule and the goals for the day's meeting. Dr. Carbon then requested that Saul Levine, NRC Office of Nuclear Regulatory Research, be prepared to address the following questions at the full ACRS Committee meeting on October 6th:
(1) the action that RES has taken on each of the recommendations ma Section 6 of NUREG-0392, " Review and Evaluation of the Nuclear Regulatory Commission Safety Research Program - A Report to the Congress of the United States of America," (2) a summary of the FY 77 and FY 78 budgets, the apparent budget for FY 79, and the proposed budgets for FY 80, FY 81. FY 82, and FY 83 ,
and the basis for these. budgets, (3) the adequacy of this budget from a technical standpoint to meet the NRC needs, and (4) the capability of the RES Advanced Reactors Research Prograas to respond to a change in national policy in favor of reprocessing and the expeditious construction of a large number of LMFBRs, and the identification of any weak areas would exist in these programs .should this change in national policy occur.
Dr. Levine noted that given a more generous budget that RES would be able to achieve their'research goals on a
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, ADVANCED REACTORS 9/27/78 accelerated time schedule. Dr. Levine noted that RES had requested mere funding for these programs than would had been granted and he noted t:.at it was difficult to define what scope this Advanced Reactors Program nould have in view of the uncertainty in the national policy on advanced reactors.
He additionally noted that LWRs must be given priority in the allotment of
- research funding. Breeder research is low and will continue to be low on the priority list until the Administration and DOE policies on advanced reactors change.
Dr. Siegel indicated that he wished to reiterate the remarks that he had made at the September 12-13, 1978 Albuquerque meeting in which he stressed the need for the early construction of the safety test facility and the need for participating in which currently are much more vigorous foreign research programs. Dr. Siegel also noted that he believes that the NRC had a responsibility to be prepared to speak to the licensing of advanced reactor concepts when they were being considered in the early decision process.
INTRODUCTION AND OVERVIEW - C. Kelber and R. Curtis, NRC-ARSR Dr. Kelber briefly summarized the schedules and RES goals for this day's discussion. Dr. Curtis gave a brief summary of the accomplishments of the Analysis Research Programs in FY 78. He noted that the distorted core critical experiments at ANL have been completed and that the EPIC Code had been integrated into the SAS Code to produce the SAS-EPIC Code. The 3-D thermal hydraulic code, COMMIX-1, has been completed and the plant systems code, SSC-L, has been completed. The HCDA analysis code, SIMMER-II, has also been completed. These codes are currently in use. Continued veri-fication of these codes will be done in the future. Dr. Curtis indicated that these codes had been well accepted by industry and have produced sig-nificant advances in the ' capabilities for performing an analysis of complex systems.
ADVANCED REACTORS 9/27/78 CRITICAL EXPERIMENT AND ANALYSIS - D. Wade, ANL Dr. Wade summarized the critical experiments on severely damaged cores which had been performed at Argonne National Laboratory in the ZPR-9 Critical Facility. The object of these experiments was to provide experimental data for readily analyzable cores against which the neutronics models for severely damaged cores could be tested. The goals were to separate in the analysis the errors due to the inadequacy of the basic nuclear data and the errors due to the inadequacy in the calculational models. The VIM Monte Carlo Code (htiich allowed an exact representatio's of the physics) was used as a secondary standard for testing the neutronics calculations of the severely damaged cores. For these idealized critical experiments, the VIM Code was subject only to the inaccuracies in the basic nuclear data (ENDF data set)
The ZPR-9 experiments represented a substantial improvement over the pre-viously available critical experiments (see page 1 of Attachment D). The critical configurations studied in the ZPR-9 and the measurements made are summarized on pages 2 and 3 of Attachment D. The ENDF/B-IV data set were used in all of the calculations. In addition to the VIM Monte Carlo Code calculations, diffusion theory, and transport theory calculations were made.
Twenty-nine energy group RZ models were used in both the diffusion theory and transport theory calculations. The S /Po, 4
TWOTRAN, and the DIF 3D/SN Codes were used for the transport theory calculations. A more detailed description of the calculational method.s is given on page 4 of Attachment D.
The summary of the key effective calculations are given on pages 5 through 7 of Attachment D. On page 7 of Attachment D, the results of the diffusion and transport theories are compared to the VIM Monte Carlo calculations.
The premise is that the error in the VIM calculations is due only to the errors in the cross-section data and that the method itself is exact. It is noted that from the standpoint of safety calculations the transport theory yicids conservative results and the diffusion theory yields non-conservative results. .
ADVANCED REACTORS 9/27/78 COMMIX AND BODYFIT COMPUTER PROGRAMS - W. T. Sha, ANL Dr. Sha discussed the COMMIX-1, COMMIX-2, BODYFIT-1, and BODYFIT-2 Codes.
The COMMIX-1 Code is a 3-dimensional transient, single phase, compressible flow with heat transfer code designed to allow detailed flow analyses in reactor components. The COMMIX-2 Code incorporates a two-fluid (liquid and vapor) model with the capability of accounting for non-equilibrium temperatures and in-homogenous velocity distribution. The BODYFIT-1 and BODYFIT-2 Codes are comparable to the COMMIX-1 and COMMIX-2 Codes but incorporate a geometry transformation technique which allows a improved representation of the reactor component geometry. Examples of the BODYFIT codes capability for transforming a component geometry are shown on pages 8 and 9 of Attachment D. A comparison of the form of the transfers momentum equation used in the COMMIX model with models used in other pro-duction codes is given on page 10 of Attachment D. The COMMIX Code represents a significant improvement in the calculational capability. The COMMIX-1 Code is currently operational and it is expected that the COMMIX-2 Code will be released by mid-year of 1979. The programming for BODYFIT-1 is completed and the debugging is now underway. Future plans include the completion and release the BCDYFIT-1 and BODYFIT-2 Codes and the continued validation of the COMMIX-BODYFIT Code series.
REACTOR SAFETY MODELING AND ASSESSMENT - C. Kelber, NRC-RES and P. Abramson, BNL Dr. Kelber indicated that his presentation had been prepared by Dr. H. Hummel .
Dr. Hummel was presently on foreign travel was unable to attend this meeting.
Dr. Kelber summarized the U.S. participation in the LMFBR accident comparison studies being carried by the WAC (Whole-Core Accident Code) Committee. The committee's activities involve both the U.S. and foreign countries. The reactor model being studies represents a LMFB8 with the total power of about 2003 MWt at the beginning of life. A pump coastdown transient and a transient overpower event with a re' a ctivity ramp rate input of $1 per second are being studied. The original SAS 3-D/SLUMPY and SAS 3-D/ EPIC Codes have been used by the UeS. VENUS type calculations are being used by the Europeans for the disassembly calculations. Dr. Kelber noted that the Europeans are generally
ADVANCED REACTORS 9/27/78 not as advanced as the U.S. in their capability to analyze accidents but have very active experimental programs underway related to accidents studies.
The U.K. and NRC have, in addition, a bi-lateral program which is currently emphasizing the U.K. and NRC loss of flow and transient overpower calcu-lations for 3000 MWt model of an LMFBR. The U.K. is using the FAX-2 Code while the NRC is using the SAS/ EPIC Code. A summary of the U.S. parti-cipation in the WAC comparative studies is given on pages 10 and 11 of Attachment D. Some selected results of the analysis done to date is given on page 12 of Attachment D. Problems identified with the use of EPIC through these activities are currently being addressed. A summary of the activities in the U.K./NRC bi-lateral program is given on pages 13 and 14 of Attachment D.
Dr. Kelber also discussed the capabilities of the transient fuel pin failure codes available at ANL. These are the FPIN, LAFM, and FRESS Codes. A summary of the capabilities of these codes is given on page 15 of Attachment D.
Dr. Abramson discussed the FRAM Code system being developed at the Argonne National Laboratory. The SAS 3-D Code framework is employed with the SAS fuel / coolant interaction models being replaced with the EPIC models. The FRESS fission gas release model is included. The code system treats the accident initiation and early transition phase analysis and will provide initial conditions for the SIMMER Code analysis. ANL, in the development of this code system, intends to make maximum use of work done by DOE and available work which has been done by foreign organizations. The current of work emphasizes the comparison of EPIC to the available experiments and the improvement of the EPIC disassembly models to permit the study of transient overpower and loss of flow transients through the prompt burst phase to neutronic shutdown without the use of the VENUS Code.
Two-dimensional boiling and voiding experiments and clad relocation experiments being performed at Purdue and AHL and high void fraction freezing and plugging experiments being performed at ANL are being examined. Near term program goals are to complete the improvements of
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9/27/78 ihe EPIC disassembly models, to develop improved 1-dimensional representations of-boiling and vo3 ding, and to develop improved clad relocation models.
An intermediate term code (2-3 years) is the issuance of the initial version of FRAM.
_ SUPER SYSTEM CODE (SSC) PROJECT - A. K. Agrawal and R. J. Cerbone, BNL Dr. Cerbone summarized the SSC development project. The SSC-L, SSC-P, and SSC-5 Codes are fluid systems analysis codes. The SSC-L has been issued and simulates short-tem (0-1/2 hour) transients in loop type LMFBRs.
The SSC-P will simulate short-tenn transients in pool type LMFBRs, and the SSC-S will simulate intermediate to long-term (beyond 1/2 hour) transients in LMFBRs. The major differences between the features of the SSC-S and the SSC-L and SSC-P are that the SSC-S will have the capability for treating auxiliary heat transfer loops and other modes of removing heat from the primary system. The SSC codes will use lumped representations of system components. Codes such as the COMMIX will be used to generate equilavent lumped representation of system components.
Dr. Agrawal discussed the SSC development. The SSC will be applicable to a class of plans (unlike codes like INSI, ANUS, and DEMO which were developed specifically for FFTF and the CRBR and which do not have the capability to be used for even similar systems with different dimensions without extensive reprogramming). The complexity of the computer model (and the cost of running calculations) can be controlled by the user via the input data in the SSC. The SSC will have the capability of performing the pre-accident calculations and initializing the SSC calculations. The SSC has the capability of using varying time steps at different points in the calculation with the attendant saving computing time. Modular programming with variable dimension schemes are used in the SSC. Standard programming practices have been adhered to and detailed users' manual have been prepared to enhance the capability of the code being installed different computing facilities. Subsample cal-culations performed with SSC-L for a natural circulation transient in the CRBR using a four-channel representation of the reactor core are shown on pages 16 through 18 of Attachment D. The current status and future plans
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7- 9/27/78 for t a hSC s ries 's given on page 19 of Attachment D.
Release of the SSC-P :.nd SSr 6 Cad 3 are planned for FY 80.
_BNL LHFBR EXFiRIMENTAL PRO JAM - 0. C. Jones, BNL Mr. Jcces in/icated that thu purposes of this program were to provide modeling and experimerical ot'servations on areas related to LMFBR safety and to support the development of the SIMMER program. Research is currently directed towards the understanding of thermal / hydrodynamic phenomena. Flow regimes and void dynamics in heat generating boiling mixtures, fuel relocation dynamics, and the heat transfer characteristics of heat generating boiling pools are being studied.
Electric and microwave heated out-of-pile heating techniques are also being studied as to the suitability of microwave and electrical heating sources.
Droplet size has been found to have a significant effect on microwave heating and power density and are thought to yield little benefit over electrical heating methods. Contaminants have been found to greatly affect the microwave heating in a mixture. The electrically heated and microwave heated test apparatus used at the Brookhaven National Laboratory are shown on pages 20 and 21 of Attachment D. Two-phase relocation effects are being studied in the apparatus shown on page 22 of Attachment D. Some selected test results are shown on pages 23 through 26 of Attachment D.
Apparatus used to study heat transfer in boiling pools is shown on page 27 of Attachment D. Selected test results are shown on pages 28 through 30 of Attachment D. The BNL experiments indicate that the two-phase heat transfer in boiling pools can be modeled using simple extensions of single-phase heat transfer concepts. A summary of the SNL program results is given on page 28 of Attachment D.
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. ADVANCED REACTORS 9/27/78 HTGRINTRODUCTIdf -R Sc berger, NRC-RES Mr. Schamberger summar.ized the RES-sponsored HTGR research. Analysis programs are being conducted at BNL and LASL with experimental programs being con-ducted at BNL, LASL, ORNL, PflL. Approximately 60% of RES funding in this area is being spent at BNL. Recent work has placed emphasis on Fort St. Vrain problems. The very recent changes in the proposed budgeting for this work will require the reevaluation of these programs.
_HTGR EXPERIMENTS - D. G. Schweitzer and C. Sastre, BNL Mr. Sastre discussed work conducted at Brookhaven National Laboratory in the areas of catalyzed graphite oxidation and its effect on structural properties, variations in the properties Incoloy 800H with grain size and tramp (impurity) element composition and iodine transport and distribution in HTGR environments.
Mr. Schweitzer indicated that the affect of o.vijat. ion. on the s.trength of graphite is a problem that has been reco'gnized for some time. High purity graphite has been studied and the data will correlate such that reasonable estimates of the effect of oxidation on strength can be made. On the other-hand, relatively little data exists for lower grade graphite such as the type proposed for the core support systems in HTGRs. The role of iron as a catalyst has been of some concern. The BNL work to date indicates that the reduced forms of iron are very effective catalysts. The iron oxides on the otherhand are not effective catalyst.
The work on Incoloy 800H has shown that the creep rate in smalI grain samples is about 100 times faster than the creep rate in coarse grain material.
Solution annealing of Incoloy 800H is difficult because the material must be brought to a temperature very close to melting to insure that a con-sistently large size is achieved. Tramp elements in the Incoloy will affect the melting points of the material.
The problem of iodine transport and distribution is being studied with the intent of obtaining realistic estimates of the release fraction of iodine for various postulated HTGR off nonnal and accident scenarios. The work is primarily directed towards the chemistry of the released iodine. Early capture by the HTGR deals with subsequent re-release is indicated by the work to date. Fission product and fuel migration is being studied in 1
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work to date. Fission product uiprat')n is being studied in simulated core heatup experimen nituil experience indicates that in heating UC2 , U02 , M0C2 , h0 .. paphite that these materials will migrate in the graphite over much' larger distance in much shorter tims than would have been indicated from 3 .' mates chich exist in the current literature. 't 4
Mr. Sastre discussed the BNL work on containment atmosphere mixing graphite oxidation and fission product migration and the HTGR safety code library.
Stagnation of hydrogen and oxygen in the helium rich post-accident atmosphere of the containment is being studied. Experiments on a small-scale (18 inch bell jar) have been perfonned. The results indicated that an analytical treatment of the problem needs to be 3-dimensional. The RICE (Los Alamos) Code is being used at BNL. The GOP-TW0, and OXIDE-3 Codes are being used to study graphite oxidation. These codes have been used to predict oxidation rates and gas compositions in support of the NRR licensing effort. The SORS Code was initially used to study fission product migration. Recent work has, however, indicated weaknesses in this code.
Future calculations at ORNL will be based on the LARC and DAS Codes (LASL) which have been recently made available.
ORNL is maintaining a library of HTGR safety codes. The non-proprietary codes are being issued to potential users upon request and news letters are being issued informing users of errors which have been discovered in issued versions of the codes.
LASL-HTGR PROGRAMS - M. G. Stevenson, LASL The HTGR safety research program at LASL has been directed towards accident delineation and analysis of fission product dynamics and structural investi-gations.
The HTGR accident delineations have been limited to disturbances which could potentially lead to core heat removal system in-balances.A total of 51 initiating events were considered. The accident sequence probabilities were estimated using the WASH-1400 methodology and consequences were ranked using a relative hazard index.
The loss of offsite power leading to a loss of forced cooling was found to be the significant accident sequence .
The CHAP Code is being developed at LASL.for studies of the Fort St. Vrain reactor.
Flow oscillations, reactivity insertions, depressurization scenarios, 1
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9/27/78 steam and air ingress, water transients, ATWS, the loss of core circulation are being studied. Comparisons 4 ,
existing accident analysis codes and Fort St. Vrain systems transient- i .1 >e used in the CHAP qualifi-cation testing. Fission product analysis d i be performed using thm QUIL/QUIC, SUVIUS, LARC-2, and DASH Codes. Character: cics of these codes are summarized on page 29 of Attachment D. Fission produ c release being studied experimentally with the apparatus shown on page 30 of Attachment D. Work to date has confirmed the prior work done by General Atomics.
Mr. Anderson discussed the LASL-HTGR seismic program. The program goals are to develop the FYSMOD models examining the HTGR core block behavior and to develop the methodology for using scale models for the confirmation of HTGR core block behavior. To date, small-scale models have been performed on a small shake table at LASL utilizing plexiglass models and graphite blocks.
A seismic research facility specifically designed for the testing of HTGR core block components is planned. A comparisca of these test results is shown on page 31 of Attachment D.
Work at LASL is being directed towards the development of a computational and compute. code for the analysis of the HTGR concrete pressure vessel .
The code will account for prestressing and 3-dimensional effects. Tests perfonned at the University of Illinois are being used for raadel development and verification. The test geometry is shown on page 31 of Attachment D.
The computer model of this test is on page 32 of Attachment D, and selected experimental results are shown on page 33 of Attachment D. Two-dimensional scale models are under construction and will be used to investigate the effect of block pin connections, friction, and side wali restraints upon HTGR core block seismic behavior, f0RTST.V_RAINCALCULATIONS-S. Ball,ORNL Mr. Ball discussed the ORNL codes and code verification that had been performed using Fort St. Vrain's graph data. ORNL has developed the ORTAP-FSV Code.
The code has been operational for about three years. A description of the characteristics of the code is given on page 34 of Attachment D. Lomparison I
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of the code predictions with a Fort St. Vrain scram. transient it shown on page 35 of Attachment D. The input parameters for this calculation were obtained from Fort St. Vrain operating conditions and plant measurements.
There was some uncertainty in the determination of these parameters. An attempt was made to obtain a better fit to the transient by utilizing a parameter optimization technique. The parameter adjustments are shown on page 35 of Attachment D and the comparison with experimental data is shown on page 36 of Attachment D. Extended code verification programs involving improved measurement analysis and reversed flow tests in Fort St. Vrain have been proposed for the future.
ULTRASONIC DETERMINATION OF THE STRENGTH OF STRUCTURAL GRAPHITE ingey,
- G. PNL Mr. ingey discussed the work at PNL which was directed towards correlating the sonic velocity in graphite with the strength of the graphite. Measure-ments to date would indicate that viable techniques can be developed. A sample of some of the measurement data that has been obtained at BNL is given on pcge 37 of Attachment D.
SUMMARY
- C. Kelber, NRC/RES Dr. Kelber summarized the highlights of the day's meeting. He indicated that RES was sponsoring work which would lead to the development of a family of codes for fast reactor safety analysis and that these codes involve models which utilize an understanding of the basic pher.omena rather than specialized engineering correlations developed for specific application. The SSC systems codes and the COMMIX Codes are a part of this family of codes. The older SAS Code system is being revised and improved to allow the incorporation of new data and calculational models.
RES has initiated some detailed cooperation efforts with foreign countries.
RES expects that these programs will be muturally beneficial . Dr. Kel ber noted that budgetary and political constraints along with the difficulties of convincing key cople to reside a broad were problems.
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The thermal hydraulics experiments at BNL are directed ouart he dissipation phase problem and reinforce the tentative conc' sus ns advanced by LASL at the July HCDA meeting that no inherent tendency to self dispersion appears attributable LKFBR core melt.
The gas cooled programs have established a set of basic programs for addressing safety questions related to the failure modes and limitations of the materials and systems which might either reduce or war,en the effects of an accident and consequences of severe accidents. Budgeta ry constraints required a rather superficial treatment of some of the HTGR safety questions. Dr. Kelber indicated that unlike the LMfBR program, the gas cooled reactor program would require a period of perhaps a year or two before it copes successfully with a commercial plant applicaticn.
In the HTGR area, a significant code library has been established and has been given the benefit of some testing and evaluation. A systematic survey of potential safety problems has also been carried out and basic knowledge has been gained about the properties of graphite in an HTGR environment.
EXECUTIVE SESSION Dr. Carbon asked the ACRS consultant if they had any additional comments which they would like to make at tnis time. The consultants had none and the meeting was adjourned at 6:3r p.m.
NOTE: For additional detait olete transcript of the meeting is available in the N c Document Room, 1717 H St., NW, Washington, DC 20555 ci am Ace-Feceral Reporters Inc.,
444 North Capitol St., h. . Washington, DC.
I
ATTENDANCE LIST ACRS ADVANCED REACTORS SUBCOMMITTEE MEETING WASHINGTON, DC SEPTEMBER 27, 1978
-A NRC STAFF M. Carbon. Chairman C. Kelber P. Shewnon L. Rib R. Seale, ACRS Consultant J. Long S. Siegel, ACRS Consultant T. Speis R. Savio. Designated Federal Employee BROOKHAVEN NATIONAL LAB ~
LOS ALAMOS SCIENTIFIC LAB.
A. Agrawal R. Cerbere C. Anderson J. Guppy P. Secker
- 0. Jones K. Stevenson C. Sastre D. Schweitzer G. Greene ELECTRIC POWER RESEARCH INST.
T. Ginsberg A. Adamantiades ARGONNE NATIONAL LAB.
GENERAL ATOMICS W. Sha D. Wade J. Graham P. Pi22ica Gar;,er Sienicki ENVIRONMENTAL PROTECTION AGENCY Abramson C . Sm 0AK RIDGE NATIONAL LAB
- BABCOCK & WILCOX S. Ball R. Borsum OTHER PACIFIC NORTHWEST LAB.
K. Almenas Univ. of MD G. Tingey M. Moncuso. Ace-Federal Reporters M. Meltzer, Ace-Federal Reporters ATTACHMENT B l
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. ' NOTICES .
40579 In accordance with the procedures Washington, D.C. 20555, to review: (D ppe of meetinr Part 'open-October 8- outlined in the FrDERAI. Rr.CisTEn on A draf t of regulatory guide 8.JD, "Ra-open 9 a m. to 12 noon: closed I p.m. to 5 diation Safety Training at Nuclear p m. Octotwr 6-Clowd 9 a.m. to 5 p.m. October 31,1977 H2 FTL 5G97:1), oral or 3
gn.act person: Mr. Georre W. Tressel. Pro. written statements may be presented Power Plants." (2) certain aspects of gram Dirretor, f*utlie Understandme of by members of the public, recordings emergency planning, and (3) the NRC
" d oy will be permitted only during those research program on radiological ef-ho'na rc dado portions of toe meeting when a tran. fccts for the annual ACRS report to D C. 20550. telephone 202 282-7770. script is being kept, and questions may -
Congress.
.';, p'A1[', yvrpose of subcommittee: To provide adytee be asked only by members of the Sub- In accordance with the procedures and recommendations concerning direc. committee, its consultants, and staff.
70 *- tion and priorities and support for pro- "d t I' s on W jects in the public understanding of sci- Persons. desiring to make oral state. ,
3g 7 5 )- g, ence program. ments should notify the designated written statements may be preschted surnmary minutes: May be obtained fr Federal employee as far in advance as by members of the public, recordings y '*
., L the Committee Manacement Coordinator. practicable so that ' appropriate ar.
rangements can be made to allow the will be permitted only during those ement. Jtoor 2 8 Nat "(
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necessary time during the meeting for portions of the meeting when a tran-43 }toundation, Washington, D.C. 20550.
- script is being kept, and questions may Agenda:
- such statements. be asked only by me nbers of the Sub-
% The agenda for subject meeting committee, its consultants, and stalf.
% Ocronca 5-9 a.u. to 12 noon-Ort" shall be as follows:
- Sossio" Persons desiring to make oral state-Open discussion. Review of the grants and WEDNESDAY, SErTEMBER 27,1978 ments should nottiy the designrted
. pr. Federal employee as far in idvance a5 7.y [. funding pattern of the last fiscal year and s:30 A.M. LTNTIL THE CONCLUs!oN or
- g. .,, diseu.s trends and questions related t -
BUSINESS practicable so that appropriate ar- -
- g PUOS long. range f unding policles. g The' Subcommittee may meet 'n ex. necessary time during the meeting for Octours 5-1 r.u. to 5 r.M. ano Octosta 6-9 ecullVe scSsfon, with any of its consul. g' M. A.M. to 5 r.M.-CLosto StssioNs ,3
"
- The agenda for subject meeting To review and evaluate research proposals and a cir p e ary
-+, as part of the selection process for awarda. lons regarding matters which should shall be as follows: Room 1167 at 1
- ' Rena,on for closing: The proposals being re- be considered during the meeting and p.m., Wednesday, September 27,1978, viewed include information of a proprl- to fqrmulate a report and recommen. until conclusion of business. Room etary or confidential nature, includmg 1046 at 8:30 a.m.. Thursday, Septem- l h dations to the full Committee. ber 28,1978, until conclusion of busi-At the conclusjon of the executive l b2b aNrs nd pc 0 5 l$fo t o$ e"on. I cerning indWiduals associated with the session, the Subcommittee will hear ness.
D The Subcommittee may meet in ex-D proposals These matters are within ex. presentations by and hold discussions ecutive session, with any of its consul-1 emptions (4) and (G) of 5 U.S.C. 552b(c), with representatives of the NRC staff, and their consultants, pertinent to the tants who may be present, to explore
?'4 Oovenunent in the Sunshine Act.
Authority to close meeting: This determina- above topict.. The Subcommittee may and exchange their preliminary opin- l tion saa made by the Committee Manage' then caucus to determine v'hether the ions regarding matters v,hich should
' matters identified in the initial session be considered during the meeting and b sechlon1 of F b 92 4 he on have been adequately covered and to formulate a report and recommen- ,
' C* mittee Manarement Officer was delegated whether the project is ready for l D the authornty to make such determina. dations to the full Committee.
8 tions by the Acting Director. NSF, on Feb. review by the full Committee. At the conclusion of the executive .
'f Further information regarding session, the Subcommittee will hear ruary 18,1977.
Q M. REnEccA WINKLER, topics to be discussed, whether the meeting has been canceled or resche.
presentations by and hold discussions Committee with representatives of the NRC staff, j duled, the Chairman s ruling on re- and their consultants pertinent to Management Coordinator, this review. The Subcomtraittee may l SErrEhrBER 6.1978. "[a#$tateme o 1 is a d t . time al o e -
therefore can b o ained b a e Id then caucus to determine whether the
[FR Doc. 78-25591 Filed 9-11-78; 8.45 arni g eral employee for this meeting. Dr. have been adequately covered and
- whether the project is ready for
~ Richard P. Savio, telephone 202-G34 32G7, between 8.15 a r. . and 5 p.m., review by the full Comraittee.
[7590-01)
- Further information regarding e d t-NUCLEAR REGULATORY topics to be discussed, whether the COMMIS$10N Dated; Sept. 6.1.J78. meeting has been canceled or resche-O JOHN U. IloYLE. duled, the Chairman's ruling on re-ADVISORY COMMITTEE ON REACTOR SAFE- Adelsory Committee QUcsts for the opportunlty to present U GUARD $ $UBCOMMITTEE ON ADVANCED mnagement W/teer. - oral statements and the time allotted REACTOR $
[FR Doc. 78-25580 Filed 9-11-78; 8.45 aml therefore can be obtained by a prepaid telephone call to the designated Fed.
MMn9 eral employee for this meeting. Mr.
> The ACRS Subcommittee on Ad. Ragnwald Muller, telephone 202-634-u vanced Reactorr will hold an open [7590-01]
> meeting on September 27,1978 in ADV150ltY COMMITTEE ON REACTOR $ ATE- #'
room 104G,171711 Street NW., Wa.sh- GUARD 5 5UBCOMMITTEE ON RADIOLOGl-ington, D.C. 20555 to continue itS cat Eff EC15 AND 54E EVALUATION Dated: September 6.1978.
review cf matters relatea to the NRC- JonN C.lioYLE,
. sponsored research on the safety of "8 Adrisory Committee advanced reactor designs. This meet- The ACRS Subcommittee sa Radio- Managemenf O//icer.
Ing was rescheduled from September logical Effects and Site Evaluation will (FR Doc. 78-25574 Filed 9-11-78; 8.45 aml 25,1978, as anno.unced at 43 FR 30152, hold an open meeting on September August 15,1978. 27-28,1978, at 1717 II Street NW., - . ..
l FEDERAL REG 15fER, VOL 43, NO. IT7-TUE5 DAY, SEPTCMBER 12,' 1978 2
!s. .
.4.Mmew6 M 9 '-. ,. .
. . g . .
o.. ,
TENTATIVE SCHEDULE -
ACRS WORKING GROUP 6 SEPTEMBER 27, 1978 FRESENTATION APPR0XIMATE TIME TIME
- 1. EXECUTIVE SESSION 8:30 a - 9:00 a
- 2. NRC/NRC CONTRACTOR PRESENTATION 9:00 a - 9:30 a A. Introduction and Overview (Kelber/Curtis) 30 mins. 9:00 a - 9:30 a
- 1. Critical Experiment & Analysis (ANL) 45 mins. 9:30 a - 10:15 a
- 2. COMMIX (3-D Core Thermal Hydraulic Analysis Under Natural Convection (ANL) 45 mins. 10:15 a - 11:00 a
- 3. Reactor Safety Modeling end Asses- ;
ment (ANL) 45 mins. 11:00 a - 11:45 a j
LUNCH 12:30 p - 1:30 p i
- 5. Thermohydraulic LMFBR Safety Experi- >
ments (BNL) 30 mins. 1:30 p - 2:00 p
- a. HTGR Introduction (Schamberger) 15 mins. 2:00 p - 2:15 p
- 8. HTGR Program (LASL) 45 mins. 3:45 p - 4:30 p
- 9. FSV Calculations (0RNL) 30 mine. 4:30 p - 5:00 p
- 10. Graphite NDT (PNL) 15 mins. 5:00 p - 5:15 p
- 3. CONCLUSION 45 mins. 5:15 p - 6:00 p
- c. MCs b I
i o
' ~
PREVIOUSLY-AVAILABLE CRITICAL EXPERIllEllT.RESULTS: .
- 1. ZPR-3 assemblies 27 and 28 (E1963) .
. Small 2ssU-fueled assemblies
. Severe Damage Rotationally Symmetric ,
J, l
~ .
. : ~
- 2. ZEBRA BG and ZEBRA 12 (*1967) , ,
' ~
. Pu02 /UO2 - fueled assemblies ,
. Sirigle Central Subassembly - -
1
. Small Off-Center Cavity .
1
- 3. 2PPR-2 ($1970) . .
' 1
. Pu02 /U02 fueled assembly i
~
. Single central subassembly . -
l
- 4. ZPPR-5 HCDA Sequence (E1976) .
l
. Pu02/UO2 fueled assembly -
Initial Phases of HCDA Sequence'
. l 9 . e . e
, 9 G *
= '
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f / / / / /,/ /j //////// f_/ /_/ / / / / s V ///////J
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kh %!s,.!{t b a h, @ N l'/ /////// ////////> J //////// / / / ////
- 1. REFERENCE CORE 2. SODIUM VOIDED 3. FUEL- SLUMP-OUT 4. FUEL-SLUMP-IN TEST ZONE -
l
//////// f __ / / ///// / ///// /////////
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l 5. AXIALLY ASYM- 6. AXI ALLY ASYM- 7. AXIALLY ASYM- 8. REFERENCE CORE .
METR 1C FUEL METRIC ' BLANKET METRIC FUEL ^
l . SLUMP-IN COLLAPSE SLUMP-OUT
~"
i 1..
CORE [ REFLECTOR SODIUM .
]
! { BLANKET,-
VOID COMPACTED CORE.
j s
,i . ,
'lIil l :
i, i
- . 4
. . c.
~ '
Pro.o ram of Measurements 'in the Safety Related Critical Experiments Axial **"UID'f) '
R 2300 BuC Control
""[ h " ate
~
Central Reactivity ~
Reactivity Worth S.
Doppler . Rod d e Neutron .Worth Worth s/f. s k a Spectrum Worth Profiles (Axial)(Radici)
, Configuration eff 239Pu, 240Pu 23sPu,}380, / / Core Center Core Center /' )
- 2. Sodium Voided / .
Reference .
" " / .
- 3. Symmetric fuel /* -
Slump-Out .
" " / ~/ Core Center Core Center / -
- 4. Synr.ctric fuel " / . Core Center
- Slump-in .
n / . .
/ -'. 239Pu, 238g, Asymmet'ric fuel . - ..
- 5. SST, 10B .
51 ump-in *
. / . ,
Asymmetric fuel /
- 6. .
Slump-in with -
- 1-s blanket collapse l- -
< " - .. ---- s 239Pu, 238U, / .
- -7. Asymmetric fuel / . - /
- SST .
Slump-out .
..aThe changes in k betwden the configurations gave the worths of these configuration changes, ,
.b The foil locations were determined based on Monte Carlo code validation requirements. - .
cThe control rod. height was kept equal to core height.
j,1 4dRatio of beta effective to prompt neutron lifet'ime. ) , '
~
c .' -
. Beta effective. '
f Stainless Steel, 3
9
. 9
~
CALCULATIONAL METHODS
- 1) ENDF/B-1V DATA USED
- 2) MONTE CARLO VIM CODE CONTINUOUS ENERGY REPRESENTATION EXPLICIT GEOMETRIC DETAIL (ONLY SOURCE OF ERROR IS ENDF DATA ITSELF)
- 3) DIFFUSION THEORY 29 GROUP RZ MODELS MC2-II/SDX' PROCESSING CONVENTIONAL D'S BEN 0IST D'S (STREAMING)
GELBARD D'S (STREAMING AND TRANSPORT CORRECTION)
~
- 4) TRANPORT THEORY 29 GROUP RZ MODEL Sq/P g TWOTRAN, DIF3D/SN GEOMETRY ON CROSS SECTIONS IDENTICAL T0 i
THOSE USED IN' DIFFUSION THEORY l
l l
, Yr
4
. .s RSR Core Meltdown Criticals Progran , ,
Configuration Eigenvalues and C/E Values (in parentheses) k Is tr pi Benoist Gelbard Si,/Po Sr, /Po Monte Carlo
, expt D's D's D's (DIF3D) TWOTRAN 200 K hist Ih/%ak, Reference 1.00093 0.99270 0.98976 0.99093 1.00232 1.00328 0.9939 0.0015 975.61 (0.9918) (0.9888) (0.9900) (1.0014) (1.0023) (0.9930)
Na Void Zone 0.99696 0.99094 0.98379 0.98577 1.00154 1.0025 971.86 m (0.9940) (0.9868) (0.9888) (1.0046) {1.00556)
, Sym. Slump Out 0.98890 0.97620 0.97105 0.97388 1.00271 1.00410 0.9964 0.0015 969.57
, (0.9872) (0.9819) (0.9848) (1.0140) (1.01537) (1.0076) g Asyn. Slump Out 0.9916 0.97875 1.0049 971.17 ,
(0.9870) , (1.01341) b 0.99959 0.99462 0.99824
~
Sym. Slump In 968.18 Sy m. Slump In 1.00115 O.98308 1.01763 1.00823 1.0041 0.0015 965.18
, (0.9820) (1.0165) (1.0071) (1.0029)
$ Asymm. Slump In 0.9862~ 0.97231 1.0058c 969.77
, (0.9859) ,(1.0199)
O
] Asy n. Slump In 0.9920 0.98060 971.17 with Blanket (0.9885) -
Collapse "From Gelbard D Diffusion Calc.
h b This configuration was not physically constructed for operational safety reasons.
"Not fully converged. ,
D
-TAB 1E I. Reactivity Worths of Material Movement ~
s Asymmetric Slump-In
, , With
{ Sodium Voided Symmetric Asymmetric Symetric Asy=etric Blanket j ,
Configuration Zone Sir.mp-Out Slump-Out Slump-In Slump-In Collapse
( Experimental Worth
-386.3 1.9 -781.1 12.8 -518.8 8.5 +2106.1 57.2 +656.4 59.8 +1221.4 1 57.7 (Ih) 1 Calculated Worth e '
l a
a.
Isotropic D's -159.3 -1427.1 +840.7 (0.412) (1.827) (0.399)
- b. Benoist D's -576.8 -1231.5
+1056.1
! (1.492) (1..i77) (0.501)
+
0
- c. Gelbard D's -498.1 -1149.6 -669.3 +1213.1 +142.8 +947.0 f
i (1.289) ,- (1.472) ,
(1.290) (0.576) (0.217) (0.775) ,
{ ? CORE RADIUS 44.2 cm + D CORE RADIUS 42.4 cm L
The reactivity worth of the sodium voided test zone was determined relative to the reference configuration. The worths of all the slumped configurations were determined relative to the sodium voided test zone.
b For the determination of the reactivity worths of the configurations with the smaller core radius, an experimentally deterained corn / radial blanket edge drawer exchange worth was used. The experimental exchange worth was
-53.08 2 1.75 Ih and diffusion theory (Gelbard D's) calculations gave a C/E of 0.877.
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B0UNDARY-FITTED COORDINATE TRANSFORMATION APPROACH
.e FUEL RODS ARE TREATED AS INTERNAL BOUNDARY CONDITIONS e TRANSFORMATION IS NECESSARY (BODYFIT CODE)
=W >
\h// W i 10$ IN
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T7 dj%b9&. A ' '
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PHYSICAL PLANE TRANSFORMED PLANE THE PURPOSE OF THE TRANSFORMATION IS TO TRADE THE ORIGINAL
~~
SIMPL4FIED P'.D.E. WITH COMPLICATED BOUNDARY CONDITIONS TO THE TRANSFORMED COMPLICATED P.D.E. WITH SIMPLIFIED B00iiDARY C0i1DITIONS. THUS THE COMPUTATIONAL GRIDS COINCIDE WITH PHYSICAL B0UNDARIES UNDER C0ilSIDERATION, 2--
a c - .. g e -
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Table 1 - -
COMPARISOF OF TRANSVERSE MOMENTUM EQUATIONS USED IN THE VARIOUS COMPUTER PROCRAMS THI3D-1 -
0=- " "ij + "i"ij - S BP ..
3;,W Sz ax g ax; 13 f.-
COBRA-IIIC
- :-- x , U Sm.. Sw m.. 2 2
- 13 . 13 + E.(p, p,) y.,
8t az 1 1 j 13 X ,u 3,
- y SABRE-1 .
- 2 -
2 3pug v apu; 3p 0=- -1 - +p 3 ug + R.
as ax g ax g ,g 2 i COMMIX-1 e 2 8pu g 3yu u By,ougw ay3 0ug u) - y gp
+p g+y 8v3g+yR Vi yV at az ax. V 3x.i az 2
3x.
2 j
J O O O O @ @@
Computer 1 2 3 4 5 6 7 Program COMMIX-1 yes yes yes yes yes yes yes TIII3D-1 no yes yes yes no no yes COBRA-IIIC yes yes no yes no no yes SABRE-1 no yes ges yes yes no yes l .
g b N .
s 4 s e =',,_sr 412-- ...h..
FIG. 1. ,
E CHRONOLOGY OF US PARTICIPATION IN WAC COMPARATIVE STUDIES -
AUGUST 1977 FIRST RECEIPT OF WAC DOCUMENTS.
1 SEPTEMBER 21, 1977 ATTENDANCE AT BRUSSELS MEETING BY l H. H. HUMMEL AND P. A. PIZZICA, ANL, REPRESENTING USNRC.
JANUARY 25, 1978 ATTENDANCE AT BRUSSELS MEETING BY
.D'. R. 'FERGUSON, ANL, REPRES ENTI NG
'USDOE, AND H. H. HuMMEL. PRELIMINARY REPORT ON NRC LOF AND TOP CALCULATIONS. j l
JyNE 1978 NRC LOF CALCULATIONS COMPLETED. .
l JUNE 28, 1978 ATTENDANCE AT BRUSSELS MEETING BY !
H. H'.. HUMMEL AND P. A. PIZZICA AND BY
- - L. W. DEITRICH, ANL REPRESENTING USDOE. ;
~
PROGRESS REPORT ON NRC TOP CALCULA- l TIONS. PLANS FOR NEXT ROUND OF l
~
CALCULATIONS ON. IRRADIATED CORE.
h t !
. . 1 l
j
~
M
\ -.
. . 4
_,_ q
~
~ Fis,2.
., . PARTICIPANTSINWACCOMPARATIVESTUDIES ,
UKAEA, RISLEY '
CEA', CADARACHE (NOT FOR TOPL f .
KFK,. KARLSRUHE l
BELdONUCLEAIRE, BRUSSELS .
i JOINT RESEARCH CENTER, ISPRA USD0E, ANL -
USNRC, ANL -
. 0THER ORGANIZATIONS REPRESENTED AT MEETINGS' EEC BRUSSELS, CEN - MOL ,_. .-
GRS - COL GNE ._.
INTERATOM - BENSBERG ,
RCN - PETTEN , ,
I e
~~ _m .
= s s
~
FIG. 3.- -
SELECTED RESULTS OF WAC LOF CALCULATIONS KFK BELGIUM UK FRANCE USNRC ISPRA CONDITIONS AT ONSET OF DISASSEMBLY
~
REACTIVITY, $ . 2.26 2.31 2.40 1.94 2.18 1.6 E0ILING -
VOIDING ,
RAMP RATE, $/SEc 16 25 - 35 15 12- -
REACTIVITY, $ 0.006 0.009 0.001 0.023 0.002 FUEL .
SLUMPING
~
RAMP RATE, $/SEc ,
2.2 4.5 1.1 4.1 0.5 ,
50 278 297 349 185 265 PEAK NORMALIZED POWER ~
max. FUEL TEMP, OC 4101 4206 4242 4109 4364 4365 -
TOTAL ENERGY RELEASE, MJ 46105 47760 48370 44500 48731 493257,:
WORK FOR EXPANSIGN TO 1 BAR, MJ 42 76 186 131 s
P .
't
\h Fis. 5.
. U
~
CHRON0 LOGY OF UK/NRC BILATERAL PROGRAM JANUARY 1978 AGREEMENT ON DRAFT PLAN FOR PROGRAM REACHED IN MEETING AT RISLEY BETWEEN H. HUMMEL AND UK PERSONNEL. THIS AGREEMENT WAS SUBSEQUENTLY APPROVED BY' UK AND NRC MANAGEMENT. ,
MARCH'1978 ,.
FIRST TECHNICAL DOCUMENTS RELATIVE j
TO' PROGRAM ' RECEIVED FROM UK.
i 1
1 MAY'8-18, 1978' ~ D. BILLINGTON AND R. LESLIE VISIT ANL TO INITIATE SAS CALCULATIONS OF 3000 MWT LMFBR. .
JUNE 1978 ANL LOF CALCULATIONS WITH AND WITHOUT .
AXIAL EXP^NSION FEEDBACK COMPLETED.
RESULTS DISCUSSED AT MEETING AT RISLEY
. BY H. HUMMEL AND P. PIZZICA WITH UK.
~
L. W. DEITRICH REPRESENTING DOE ALSO PRESENT. . .
SEPTEMBER 1978' ANL TOP CALCULATIONS INITI ATED.
- j
.. JANUARY 1979 (PROJECTED) flRST REPORT ON CALCULATIONS TO BE l PREPAREI; . -
. Y' ,
l V - . . - . - . - . . - - . . . . . .
~Fla. 6. .
UK DEVELOPMENTS IN TRANSIENT FUEL PIN MODELING OF PARTICULAR INTEREST FUEL MELT THROUGH ON 75% FUEL MELT FRACTION AND CLAD TEMPERATURE
> 1280 K. THIS MAY BE APPLICABLE ,FOR IRRADI ATED FUEL IN VOIDED REGIONS. .
A CRITERION SEPA' .
R ATION OF CLAD FAILURE AND MOLTEN FUEL MOTION.
j IS PROVIDED FOR BREAKUP OF THE SOLID FUEL ANNULUS AFTER CLAD I FAILURE. .
STRAIN RATE FORMULATION FOR CLAD FAILURE. '
- v" FRUMP CODE DEVELOPMENT.
1 l
egne f
I e
- O O 6 g G
8
' G
- . ~
t I
4
. (
~
. F is .' 7.
SUMMARY
OF TRAtlSIENT FUEL-PIN FAILURE ]
CODES AVAILABLE AT ANL i j
FISSION GAS
~
CRACKING RELEASE CODE C[ TEEP I
. BELOW MELTING FPIN MODELED MODELED NOT MODELED LAFM MODELED NOT SINGLE EXPERI-MODELED MENTAL CURVE FOR RELEASE VS FUEL TEMP.
' h w' ' 9'** #
FRESS 0 t@Cifsi! CAL OLD GRUBER l PARAMETRIC MODELING.,
MODEL
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, 35 1
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0.04 2:
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O.02 -
l -
WITHOUT FLOW REDlSTRIBUTION
~
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g O 80' 160 240 320
- TIME, s r, ..
)
b . = - - - - - - - - - - - -
- - - - - - _ - _ - _ - - --;-__ = --______-___-_:__:-_-_____:_===____
i i
. 1200 i i i i i i .
CRBRP _
~
HOT CHANNEL
. ~' -
110 0 -
. / ..
/ -
N s
N y
s., ' ' s,
'* / 'WITHOUT FLOW ~ -
y ~
f REDISTRIBUTION '
's f -f -
<t 1000 -
'f cc to ,
I CL g
/ -
WITH FLOW
% *~
REDlSTRIBUTION -
900 -- mv.~g c a w 9" -
9
~ .
800 -
~
i i i i i - i i i 320
' O 80 160 240 4 TIME, s i
) _- - - - - - _ -
I I I i i i 1
! 1 CRBRP s -
AVERAGE CHANNEL IN PEAK ASSEMBLY Q
1000 -
_ WITHOUT FLOW ' REDISTRIBUTION - 1300 f -~~'s k
/
y /
's s. .
1200 g g 900 -
/ .
's @,
x ;
'~ q
/ lloo
$ L g e a.
W WITH :E 1000 p
$ 800 -
~
FLOW REDISTRIBUTION w
'r-900
~
700 ' ' ' ' 800 0 80 16 0 240 320
' 'G" ,
TIME, s
7 ,
f CURRENT STATUS'AND FUTURE PLANS ....
TASK FY 78 FY 79 FY 80 CODE DEVELOPMENT SSC-L CODE TESTING -
CODE APPLICATION RELEASE UPGRADED-AND RELEASE AND MODIFICATION CODE DHX MODELED INTERFACE DHX COMPLETE PLANT COMPLETE CODE CODE DOCUMENTATION
. SSC-P INITIALIZATION ASSEMBLY .
AND RELEASE, CODE TESTING AND. APPLICATION SSC-S DEVELOP MODELS FOR AUXILIARY COMPLETE CODE HEAT REMOVAL SYSTEM ASSEMBLY AND RELEASE _
CODE VALIDATION- ,
FFTF SOME COMPARISON PRE- AND POST-TEST DEFINE FUTURE BETWEEN DIFFER- ANALYSES OF EXPERIMENTS ENT CODES ACCEPTANCE TESTS
] BROOKHAVEN NATIONAL LABORATORY {} g)l A5500ATED UNIVERSITIES, INC.(ILII
{ -
i - - - - -
6
DISPERSION IN INTERNALLY HEATED BOILING POOLS-ELECTRICAL ,
(INITIAL SCOPING TESTS) s-PURPOSE g
s
\ RECTEE OBTAIN VISUAL AND MACROSCOPIC OBSERVATIONS OF VOID BEHAVIOR AND FLOW REGIMES IN NEAR-
' ih, '
PROTOTYPIC GEOMETRIES l i i
-,8 i 11 l I '
I i .
' '/LIj ,
i, i 8
STATUS li i i I
'l ' e EXPERIMENTS COMPLETE a i i' ' J. 'I ) j [ ZNS04
,, u e BNL REPORT ISSUED li I (BNL-NUREG-2fl270) it 's sl l I
lI'
- l
) l ~
! . '1 ' p .)
g \ l D '4 BROOKHAVEN NATIONAL LABORATORY l} g){
E GAMNMRE,INC.GlH MAKEUP FLOW e
T 4 - + - , , , - -
y' .
DISPERSION IN INTERNALLY HEATED. BOILING POOLS-MICROWAVE .
S .
PURPOSE OBTAIN Flow REGIME AND VOID DYNAMIC BEHAVIOR WITH UNIFORMLY HEATED.
DISPERSED Flows (COUPLED DROPS)
I l
$l1..;
\ t;\; .
\\
l.: :9 STATUS
/
~
..3 e ENGINEERING DESIGN PARTIALLY COMPLETE L
k h hb ",.5 -
)t g 3
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e CHARACTERISTICS OF MICR0 WAVE-FLUID COUPLING BEING STUDIED
~
j .
e TRANSIENT POOL-DYNAMICS MODEL
- $ BASICALLY COMPLETE 4 SCHETATIC 0F MICROWAVE DISPERSAL TEST FACILITY BROOKHAVEN NATIONAL LABORATORY l} g)l A5500ATED UNIVERSITIES, INC.(IIll g .
TWO-PHASE RELOCATION - simul. ANTS '.
-4 PURPOSE
----- e DEVELOP.A MODEL'FOR TRANSIENT FREEZING OF GAS-LIQUID Flows wows METAL
~
s _E, RESERVOIR 9 IDENTIFY APPR0'PRIATE SCALING CRITERIA '
iE=ts_ _e -
?~5 55 e EVALUATE MODELING AND SCALING CRITERIA WIT
- SIMPLE TESTS N2 INLET ~ .]
'c TWO PHASE MIXTURE p g TO HEAT EXCHANGER .-
./ 7 SOLID RADIAL CRUST To EVALUATE SIMMER.
~
' .' 000BLE PIPE CONDENSER STATUS 7
- I.;
e Low PRANDTL NUMBER (M00D'S METAL PR = 0.02)
TESTS VIRTUALLY COMPLETE- REPORT ISSUED (BNL-NUREG-24486)
! .:: y WATER COOLANT
{
y
- ,1 8 HIGH PRANDTL NUMBER (PARAFFIN wax PR = 45).
BEING EVALUATED I
-4 l _-_
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M A5500ATED UNIVERSITIES, INC.(Itil i -
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SIMULATION EXPERIMENTS AND TRANSITION PHASE :
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TRANSITION PHASE EXPERIMENTS UO / steel vapor 2
Wood's Metal /N2 1 Fluid System pin tube Geometry yes no Wall Melting maybe yes Crust Stability 1 - 10
<1 Differential Pressure (bar) 2800 - 3100 75-80 Fluid Temperature Range ( C) 1500 15
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2800
\ 74
Fusion iemperature ( C) O.2 1.1 -
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0.0 - 0.4 Void Fraction, a 2.0 0.2 Phase Change Number, Npch 0.8 0.6 Conduction-Convection Number, $ i BROOKHAVEN NATIONAL LABORATORY [3 gj g h A5500ATED UNIVERSITIES, INC.(Illi W
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TIl1E t SEC 1 COMPARISON OF TWO-PHASE GAS-LIQUID RELOCATION RESULTS WITH VARIOUS VOID FRACTIONS FOR LOW AND HIGH PRANDTL NUMBERS h
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CALCUU\TED BEHAVIOR OF SUPERFICIAL FUEL PEL0 CATION VELOCITY
, AND FUEL RELOCATION FRACTION FOR 2U0 /5 ' VAPOR SYSTEM .
BR00XHAVEN NATIONAL LABORATORY l} g) {
ASSOCIATED UNIVERSITIES, INC.(I ElI 1
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1
ilEAT TRANSFER FROM B0ILfNG P0OLS - NON-DISPERSED ,
. (COMPLETION OF PAHR-RELATED WORK) .
i PURPOSE .
OUTLET
~I ~ PERFORM ANALYSIS AND EXPERIMENTS jl h YIELDING MUCH IMPROVED ACCURACY IN b* EM 'U" D" D #
Boilin a V
/
/ J poot POWER DECAY HEAT LEVELS FOR VERTICAL AND-
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j e PREVIOUS RESULTS ANALYZED IN TERMS OF OjI -2 N 1 # ~~~~
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O LOCAL NUSSELT NUMBER AND LOCAL Two-f]r O ;t O _
PHASE RAYLEIGH NUMBER.
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j e APPARATUS CONSTRUCTED AND BEING IESTED.
- co O O TO BE COMPLETED SHORTLY.
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INLET-. v ( e REPORT ISSUED - (BNL-NUREG-50759)
I p o( c W A y, 8ASE BROOKHAVEN NATIONAL LABORATORY l} g)l l4 i
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i OBSERVATIONS e LARGE SCATTER DUE TO
. . . INACCURACY IN LOCAL COEFFICIENT
. . . INACCURACY IN LOCAL VOID FRACTION .
H 0 . GENERAL AGREEMENT WITH SINGLE PHASE .
4 Nu(z) = 0.7.6 [ Gr(z)Pr]1
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. BROOKHAVEN NAll0NAL LABORATORY l} g)l ASSOCIATED UNIVERSITIES, INC.(Illl h -
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DISPERSAL
/
e MICROWAVE TECHNIQUES PROBABLY YIELD LITTLE BENEFIT OVER ELECTRICAL HEATING METHODS FOR WATER.
e TEST RESULTS INDICATE DISPERSED DROPLET FLOWS DO NOT EXIST AT DECAY HEAT POWER LEVELS.
e UNRESOLVED OUESTION REGARDING CONTAMINANTS AND STABILIZED F0AMS.
FUEL RELOCATION DYNAMICS e VaiDS SIGNIFICANTLY REDUCE DUCT PLUGGING IIMES AND MASS RELOCATION QUANTITIES, e SCALING PARAMETERS IDENTIFIED AND INITIALLY VERIFIED WITH SIMULANTS OVER THE PRANDTL NUMBER RANGE OF 0.02 TO 45, e RECRITICALITY CAN NOT BE DISCOUNTED AT THIS TIME.
e EFFECTS OF HIGH VOID CONTENT AND WALL MELTING REMAIN TO BE ASSESSED, ESPECIALLY WITH REACTOR MATERIALS AND PROTOTYPIC GEOMETRIES.
BOILING POOL HEAT TPANSFER e Two-PHASE APPEARS TO BE A SIMPLE EXTENSION OF SINGLE-PHASE HEAT IRANSFER CONCEPTS. >
BROOKHAVEN NATIONAL LABORATORY l} g)l A5500ATED UNIVERSITIES, INC.(1151 3
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1 e NONBOILING Go -9 BOILING ,
En ~8 '
r9 RUN frJMBER 602A 3-
- RUN 101BER 6305A 10lRL PouERIKul 2A.0
- 10TRL POWERIKul 2. 0 ,] e 0.0 9 RVERAGE VOID FRRCT10N 0.A Uo* RVEhAGE VOID FRAC 710N o.
" POWER DENSITYICRL/CM=3 SI 0.07 f3$_ _ _ _ _ POWER DENSITYtCRL/CM=3 SI 0.6 tD 3' C
RNGLE OF INCL INAT IONI DEG I 33.0 27.5
$~ o RNGLE OF INCLINATIONIDEGI 30.0 INITIRL POOL DEPTHICMI 15.0 INITIRL POOL DEPTHICni
- 16.0 Um BOILING POOL DEPTHttni 27.5 U9 DOILING POOL DEPTHtCMI t1J9
- SUPERF ICIRL VRPOR VELOClTY 0.09 SUPERFICIAL VAPOR VELOCITY 1.1 (D - N.O_ ~
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TYPICAL BEHAVIOR OBSERVED IN CURRENT BOILING POOL AFPARATUS BROOKHAVEN NATIONAL LABORATORY l} g)l g A5500ATED UNIVERSITIES,INC.(IElI 4
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I FISSION PRODUCT ANALYSIS CODES :
QUIL/0UIC - CHEMICAL E'0UILIBRIUM AND KINETIC ANALYS1S SUVIUS - STEADY-STATE PRIMARY COOLANT CIRCULATING ACTIVITY -
LARC-2 - FISSION PRODUCT RELEASE AND DECAY l
l DASH - MULTICOMPONENT DIFFUSION AND DECAY IN !
PLANE, CYLINDRICAL, OR SPHERICAL GEOMETRY
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-- - , , _ _ _ _ _ _ _ _ _ '- _ - _ _ _ _ . . _ _ _ _ - _ _ - _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _____________J____________-
CONCLUSIONS TO DATE e VERTICAL WALL HEAT TRANSFER FROM Two-PHASE BOILING POOLS BEHAVES SIMILARLY TO LAMINAR SINGLE-PHASE HEAT IRANSFER IF LOCAL TWO-PHASE DENSITY DIFFERENCE IS TAKEN AS THE DRIVING POTENTIAL. ,
- e INCLINED WALL HEAT TRANSFER TESTS IN PROGRESS AND TO BE COMPLETED SHORTLY. PRELIMINARY SINGLE Ph.4SE RESULTS AGREE WITH THOSE FOUND IN THE LITERATURE; WHERE GRI(z) = GRV(z) os 0 l
BROOKHAVEN NATIONAL LABORATORY l} g)l ASSOCIATED UNIVERSITIES, INC. (B 101 1 30
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THE BUGT STEAM GEITMTOR CODE IESULTS FOR A DBDA TEST CASE CLOSELY MTCHED A GA AMLYSIS ,.
THE ORTAP-FSV CODC, milch IS LIKE A p
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MAKE USE OF THE ORECA CODE
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. BE TSED FOR EMERGENCY COOLING ANALYSES
. ORECA CAN EITHER BE USED. AS A SUBROUTINE IN ORTAP OR AS A STAND-ALONE CODE DERIVING
, INPUT PARAMETERS FROM OTHER CODES OR PLANT DATA , .
. RESULTS Or ORECA ACCIDENT ANAL *!LES HAVE '
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. ORECA HAS BEEN
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- 5) . AS$tmED TIME OF SCRAn 30 SEC LATER.
s - / -
.= :: .-
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WITil P.F. = 1.335 F5Y 28% Scrarn Test 8/6/77 I-Region i 1
~ = ISV Data s
- ORECA Case i 11 .
I- .
- t
- L I-h' si-e
[.
' i. .
3, i" " " * ~~ ~
L.ut .
DATA FROM THREE OTHER SCRAM TESTS (30, 40, s 50% POWER)
HAVE ALSO BEEN ANALYZED AND " FIT" BY ORECA.
CONCLUSIONS TO DATE:
- 1) RESULTS ARE SENSITIVE TO ASSUMPTIONS OF CORE BYPASS FLOW FRACTION
- 2) ADJUSTMENTS MUST BE MADE IN MANY CALCULATED REGION PEAKING FACTORS
- 3) THE REGION OUTLET THERMOCOUPLE RESPONSE MODEl. NEEDS TO BE VERIFIED
Y.-
. . .L ... . . . _ . . . .
'N' *
, 2 *. . .
4 STRENGTH VERSUS S0NIC VELOCITY. .
,,. FOR OXIDIZED GRAPHITE SAMPLES-
- I 1
el .
. 70 .
i 60 -
50 G
W D M O
[ 40 -
5 E. . .
g e 30 o
.7 E A '
9 o.
%o
-@ CnD o - C O. .
3 .
g e MEAN OF UNOXIDlZED SAMPLES 1 g o UNSTRESSED SAMPLES O PRESTRESSED SAMPLES o a STRESSED 5AMPLES A
A A.
10 ^' ' ' ' ' '
E4 a5 a6 a7 a8 a9 LO
- ' RELATIVE VELOCITY WNo)
FIGURE 4 - Compressive Strength of 1 in. Diameter oxidized Samples vs Velocity Change Ratio-O
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, , . , _ _ _ . , . . , . . _. m__. - . - . - . , . . .. . . . . . ,