ML19260B142

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Summary of ACRS ECCS Subcommittee 790319-20 Meeting W/Exxon Nuclear Co,S&W,Bechtel & Westinghouse in Los Angeles,Ca Re Review of Status of NRC Research Programs on Analytical Work Related to Loca/Eccs & ECCS-related Experiments
ML19260B142
Person / Time
Issue date: 06/14/1979
From:
Advisory Committee on Reactor Safeguards
To:
Advisory Committee on Reactor Safeguards
References
ACRS-1625, NUDOCS 7912070302
Download: ML19260B142 (75)
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{{#Wiki_filter:. ,' l 5 DATE ISSUED: 6/14 /79 h" h MINUTES OF ME ACRS ECCS ,M SUBCCMMITTEE MEETING 6//8/M MARCH 19-20, 1979 LDS ANGELES, CALIFORNIA We ECCS Subcomittee of the ACRS met on March 19-20, 1979, at the Travelodge International Hotel, 9750 Airport Boulevard, Los Angeles, California. Se main purpose of the meeting was to review the status of NRC Research Programs on analytical werk relating to LOCA/ECCS, ECCS related experiments, and NRC licensing activities related to ECCS. Notice of the meeting was published in the Federal Register on March 3, 1979. Copies of the notice, meeting attendees, and the schedule are included as Attachments A, B and C, respectively. B e Designated Federal Employee for the meeting was Dr. Andrew Bates. W request for time to make oral statements were received frem members of the public and no written statements were received. INTRODUCTORY STATEMENT BY THE SUBCCMMITTEE CHAIRMAN Dr. Plesset, Lubcommittee Chairman, convened the meeting at 8:30 a.m., introduced the ACRS members and consultants, who were present and indicated that the day's discussion would cover topics related to NRC research programs on advanced ECCS/LOCA codes, the experimental data needed for the physical code inputs, analysis of the LCFT L2-2 test, and the status of ECCS related research programs in 2D-3D, BWR core spray test programs, the Creare Downcomer program and EDBI. PRESENTATIONS BY THE NRC STAFF AND CONIPACTCRS Dr. S. Fabic reviewed the status of the NRC Advanced ECCS/LCCA Code Development Program. Prior to December 1978, WRSR sponsored advanced codes included WAC, WCR, REIAP-5, and C RA-TF. S e Code effort was reduced with the elimina-tion of the programs on WCR and REIAP-5. since January 1979, programs are being revised or continued and include the development of the PWR 'IRAC code at IASL, BWR WAC at INEL, and C3RA-TF at IHL. 1511 209 791207 03o2_.

ECCS Mtg March 19-20, 19~9 TRAC-PIA was released in March 1979; TRAC-P2 which includes a faster running time, collapse of the 3-D vessel to 2-D or 1-D, two fluid models in all 1-D components, improved constitutive relations, flow regime recognition, heat transfer improvements and fuel module improvements, will be released in about 1 year. Plans for 'IRAC-P3 include a non-condensable gas field, 3-D neutron kinetics (A' INS), and additional constitutive equation improvements. INEL will develop the EW 'IRAC models. IASL will provide INEL with a completed version of TRAC-PIA for conversion to 'IRAC-Bl. Iater 'IRAC-P2 will be developed into 'IRAC-B2. 'Ihe droplet field to be included in 'IRAC-P2 will probably be necessary in order to do an adequate job of modeling the EWR upper plenum region. C RA-TF is being developed at PNL with the goal of being able to calculate the effects of local flow blockage, unheated ods, lateral flows and void migrations, on a subchannel basis. 'Ihe code will be capable of intimate coupling with FRAP-T (Attacnment D) and will provide a detailed analysis of the vesse'. bahrior. UHI capability is included. 'Ihe vessel boundries must be supplied by a system code calculation. Attempts are to be made to ecuple C RA to the 'IRAC code to provide the system effects r.eeded for the vessel boundry conditions. DISCUSSION CN 'IHE MATHEMATICAL ISSUES RELATED TO TRAC Dr. B. Wendroff, EASL, reviewed the issues involved in the questions raised about the well-posedness of the 'IRAC equations. Dr. Wendroff indicated that when modeling real-physical systems errors may be introduced in t1e ways, one from the approximations made in the mathematical formulation of the analytical model and cne from the approximations introduced in the computer code to solm the equation numerically. In order to determine the accuracy at both steps comparison with experiments and exact analytical solutions. must be made. Checks and comparisons with different numerical methods can be made and checks on internal consistancy of tha models can be made. In some cases one may not be able to prove with a rigorous mathematical analysis that various models faith-fully represent the system being described. Because of the complexity of the

ECCS Mtg March 19-20, 1979 flows within a reactor system one must resort to averaging techniques for describing the various physical phenomena. When one averages the local con-tinuum equation one does not obtain equations which depend only upon the averaged state variables (Slide F-1). Che may be able to construct a model which is suf41ciently correct using the average variables, but be unable to prove it rigorously. For a set of well posed equations (as shown in Slide F-3) there exists a unique solution for each data set. Sufficiently small changes in the data set produce small changes in the solution. However, a slightly different problem, Slide F-5, shows a large change in solution from a small coefficient chr.nge and in fact could produce a situation of no solution. 21s problem is well posed but ill-conditioned in that it is extremely sensitive when one wants to compute a solution. In the case of an ill-posed problem, Slide F-6, one cannot obtain an exact solution for a and b unless certain condtions apply to Xi, Yi, i.e. they lie on a line. However if one changes one's requr,t to finding the least squares solution to [a,b] the problem becoces well-posed. One then can find himself in a situation where a well-po' ed problem presents difficulty in solution because of the sensitivity of the coefficients, as well as having an apparently ill-posed problem with useful solution if one asks for a sufficiently general solution. The partial differential equations of gas dynamics are apparently well posed and well conditioned (at least in one dimension) provided one introduces some generalization in the notions of their solution and uniqueness. Eis however has not.been rigorously proven. Be equations of gas dynamics do pass a test indicating that the solution depends continously on the data (Slide F-7). In WAC the 2-phase flow models will fail the test if the phase velocity differences are less then the sceed of sound in the gas (Slide F-8). Bus under most flow conditions the equations in WAC cannot be proven to be well i v i A,/ J y

a ECCS Mtg March 19-20, 1979 posed using the tests available. In general, the solution to the 'mAC equations contain 2 real and 4 complex eigenvalves for the gas and fluid velocities under consideration. The lack of ability to prove the equation well-posed has produced numerous attempts to develop,ystems of equations which are well wed. Dr. Wendroff reported that he has found at least 10 models in them literature for 2-phase flow that he breaks into two groups. Se one group provides a set of hyper-bolic equations but probably does not correctly model the stysics, the other group attempts to increase the region of space where the generalizations and well-posedness apply. Were is a fair amount of disagreement as to the degree of importance of the added terms used to improve the region where the equations are hyperbolic. In closing, Dr. Wendroff made four points indicating that: (1) the basic model has wave propagation properties analogous to those in ordinary gas dynamics (the two real eigenvalves correspond to the rarefaction wave solution to the problem). (2) If a very large interfacial drag term is used the solution obtained collapses to the solutions of the equal velocity model. (3) te growth factors of waves on the finite difference mesh used are small enotqh so that the problem does not blowup rapidly; and (4) While ill-posedness becomes important on a fine scale and for small wave leryths, viscosity also becomes increasingly important but is not included in the basic model. With the inclusion of viscous terms the model becomes wall-posed. On a course mech (cf1 cm) the 'mAC models are stable. S e validity of the model is proven throtqh solutions to the problem and comparisons to experimental tests. Dr. K. Williams, LASL reviewed the work ongoing in the developnental assessment of 'mAC. Mznerically tnere are no proble:ns in obtaining a ' steady state ccndition. The steady state conditions reached are also accurate provided that accurate information en the experimental system flow resistances and heat transfer surfaces have been provided. Slides of the 'mAC calculaion done for the T.CFr L2-2 test were shown (Attachment G). In general the agreement was quite good although the early rewet on the hot rods was not prediated. 1511 212

~ a. March 19-20, 1979 ECCS Mtg IASL is also performing a sensitivity annlysis on IRAC with Standard Problem Ten different inputs are varied simu-45 (Semiscale heated blowdown test). taneously Esing a statistical approach in order to develop an envelop of values for the various code outputs. One can then develop a partial rank correlation of the various coefficients as We coefficient's a measure of the degree of association between input and output. The techni-imprtance may vary with time during the calculation (Attachment H). ques developed in this study will be helpful in calculating full size reactor problems where one needs to know the importance of uncertainties in various A second type of error that will be addressed is the inherent input parameters. error antained within the code which exists because of approximation in the The NRC Staff will address their plans for assessing mathematical models. this error in a future ACRS meeting. Dr. Liles reviewed the Model Development work that is ongoing with 1RAC at He indicated that they are awars of a number of sources of error in LASL. Rese areas include non-the code due to inadequate or a lack of modeling. condensable gases and bubble nucleation. They believe that the numerical methods in IRAC are stable provided the mesh size does not shrink down to less than about 1/2 centimeter. Vary large mesh sizes may also produce some truncation error in the code. An important area of interest in the further developnent of the code lies with There is a need to develop improved data on inter-the constitutive equations. facial interactions (momentum and heat transfer, entrainment, flow regimes) and wall interaction (shear, heat transfer). We present code models are Better data from based upon relatively simple steady state flow regime maps. A larger scale prototypic experiments is needed as well as transient data. These number of programs are underway which will M p provide better data. programs include wrk in the US, Germany, and France (see Dr. Fabic's prese below for more detail). 1511 713

ECCS Mtg March 19-20, 1979 Dr. Liles indicated that the damping in the TRAC numerical schemes could be reduced with the use of other solution techniques for the partial differential equations; however, the interfacial data in the code also provides damping of the solutions. The separations of the relative degree of damping between the two contributions is difficult and a large reduction in the numerical damping may not make a difference. The results when com-pared with experimental and analytic results appear good. Dr. W. Kirchner discussed the 11W' heat transfer modeling (Attachment J). The calculations are based upon a generalized boiling curve which correlates surface heat flux with the surface temperature, and fluid temperature and pressure condition. Various correlations are used depending upon which region of the boiling curve the heated surface is in. Areas where additional data are needed include down flow CHF, minimum film boiling temperature, and quench conditions and processes during reflood. Dr. Catton indicated that data on quench front propagation needs to include both the precusor heat transfer ahead of the quench front as well as the heat transfer at the quench front. Both processes take place in a very small region on a fuel rod and are difficult to measure. The TRAC models also need to improve the fuel model for gap hetat transfer.. Dr. Fabic reviewed slides (Attachment K) indicating a number of tests underway or to be conducted in the US, Germany, Italy', Canada, Sweden, and England, which will provide needed information on interfacial heat and mass transfer terms under a number of flow regimes and conditions of temperature pressure and void fraction. Sme of the date. will be used in the code developnent work, other data will be kept locked up by the NRC for use in code assessment problems. (Theh (ganma} terms on the slides are mass transfer terms,Ds (taus) are interfacial drag terms, and K's are momentum transfer terms. 1511 2l4

ECCS Mtg March 19-20, 1979 Dr. L. I4aach, EG&G, reviewed the post test analysis of IDFT Test L2-2 and the pretest predictions of IDFT Test L2-3 (Attachment L). Pretest predic-tions with REIAP 4 showed a peak temperature of 1015 K (actual 782 K). The rewet at 4-8 sr-7 was also not predicted. His rewet was due to the upward flow of a slug vf water through the core as break flow from the hot and cold legs caused a change in direction of AP across the vessel. Post test analysis work revealed that there were several input errors in the calculation. Corrections included initial test conoitions and break flow multipliers. During the post test analysis they also varied the break flow transition quality from.02 to.0025 and used the Condie-Benston, Groeneveld, Berenson, and Dougall-Rosenow transition toiling heat transfer correlation. Se main effect on the calculation was produced by correction of the-break flow multipliers while some of the other variations produced some improvement. Following the test the peak RELAP 4 prediction was 825 K. Dr. Leach agreed that a great deal of additional work was needed on understanding rewet phenomena. The L2-3 test was re--predicted with the correction made to the prior test. They had some uncertaint.y as to which changes w re appropriate and which could be identified as code tuning. Dr. Fabic indicated some concern with changing the transition quality on the break flow from.02 to.0025. His improves the hydraulic calculations and provides for a better understanding of the core heat transfer taking place. Dr. K. Williams presented the results oC the 'IRAC pre and post test calculations of IDFT L2-2. W e post test prediction used the correct test operating condi-tions and corrected some problems in the code calculations of fluid enthalpies. We calculations show delayed INB on the core rods and some quenching and dryout of the lower power rods. Se hot rod does not rewet at 6-8 seconds as occurred ~ in the test. 'IRAC models on rewet do not allow quench at temperatures above approximately 750 K. Following the test, calculated peak temperatures were 0 about 840 K. IASL personnel'believe that the external fin thermocouples on the fuel rods contribute to increased heat transfer, perhaps causing the '1511 215

ECCS Mtg March 19-20, 1979 quench. Se TRAC calculatiuns do not take the fin heat transfer into account. Se region of the calculation follo'.dng the rewet produces the wrong temperature due to the lack of quench prediction. *he total core reflood times were about right. Work is being carried wt to include data from MIT (Iloege) on high temperature quench in the TRAC models. (A later calculation with that model shows the cuench at the proper time.) Dr. L. Imach indicated that INEL had done extensive studies on the IDFT fuel both wica and without thermocuples; they did not see a large difference in quench behavior due to the thermocuples. Dr. A. Levine, GE, reviewed the program on BWR core spray. A cooperative program is planned between GE, NRC, and EPRI to extend the work to BWR CCFL tests at the Lynn Mass. facility. Foreign tests with spray nozzles indicated that the water distribution changed in a steam environment. In response to the new information GE idstituted a multiphase program of nozzle-testing in steam and in air with single and multiple nozzles. A portion of the program involves the Lyn.1, Mass. facility where a full scale 30 sector off a BWR upper plenum and upper channel boxes have been constructed. A steam supply is available and water distriubtion to the various channels will be measured under various steam upflow conditions. he extension of this program in cooperation with the NRC and EPRI will look at CCFL in the parallel channels of the EWR (see Attachment M). Other infor-mation obtained would include transient blowdown condition, and three-dimensional plenum effects. Dr. Tong reviewed the status of the 2D-3D program and the test facilities involved (Attachment N). Technical negotiation of the agreements has been completed, the lawyers are still wrking on the agreements. Construction of ~ CCTF has been completed, the slab core test facility will be done in June 1979, and the Upper Plenum Test Facility preliminary design will be firmed up in June 1979. Se PKL, CCFf and SCTF all have upper pressures of about 90 PSIA. me UPfF will be about 130 PSIA. All pressures are at the very end of blowdown and reflood region. n i /,

m March 19-20, 1979 ECCS Mtg Mr. D. Hansen, EG&G-INEL, reviewed the characteristic of the Italian LOBI Se ISBI facility bas facility in relation to Semiscale (Attachment 0). a full lenght 8x8 bundle with skin heated fuel rods of a size typical of a Initial tests are to be defined and run by the European countries, PWR. intially by the ImG, later by the larger European Community. We tests are Se stored primarily blowdown oriented and do not include reflood or refill. energy in the rods is not adequate to look at the later phases of a LOCA ~ Bere are also geometry differences between the IDBI facility, transient. Semiscale, ROSA-II and a IMR as far as the steam generator heights and A large portion of the electrical heating also goes into the leads sizes. to the rods which presents problems in heat transfer analysis and flow analysis. The meeting was recessed at 6:00 p.m. to reconvene at 8:30 a.m., on Mar 1979. Dr. W. Beckner reviewed the small scale ECC Bypass programs conducted by Se program is presently being completed and there the NRC (Attachment P). d will be a major reduction in effect during W 1980 with minimal support beyon. The program includes tests at Battelle Columbus at 1/15 and 2/15 scale, Creare, Inc. at 1/30 and 1/15 scale, and at Cart:nouth College with air / wat flooding in tubes (2" - 10") and annuli (1/7 and 1/10 scale). Research Plans to issue a formal Research Information Letter sumarizing in tne results of the bypass work in June 1979 (a draft letter for contnent Se highlights of that letter were summarited (see Attachment P). April). Bis Future work at small scale is to be completed in 1979 and early W 80. includes flow topology maping and tests with expanded instrumentatio f scale at BCL, and lower plenum voiding, combined frC penetration / refil developnene, and level swll tests at Creare, Inc. 1511 2U

ECCS Mtg March 19-20, 1979 Mr. L. Philips, NRC, reviwed the present Standard Problem Program (Attachment Q). mis program involves voluntary participation by the nuclear industry in a series of EDCA/ECCS related experiments. Parti-cipants are asked to pre-predict test results that are then compared to the actual test data. M e objectives include evaluation of code pre-dictive accuracy, identification of needed mdel improvements, and quantifi-cations of safety margins in licensing calculations. It is hoped the program will contribute to better understanding of the IDCA event. Se program has been ongoing for a number of years and over the last 18 mnths the NRC and the participants have attempted to work out a mutually agreeable program plan that defines the overall objectives and provides guidelines. Se plan addresses the licensing role, selection of the standard problem, report of results, post-test analysit, and issuance of a final report. An attempt will be made to have two tests (problems) per year at least one of which is applicable to BWR's. Eight Standard problems have been conducted and additional ones are planned. Se results of Standard Problem 47 were reviewed (LOET test L1-4). The major area of difficulty in the Program is the incentive for the various vendors to participate in the program. Se vendors generally have Evaluation Model ECCS codes which are not appropriate for many of the best estimate calculations that must be done in the standard problem program. Participation in the program requires time, effort, and money which is sometimes difficult to justify particularly if pressing items involving reactor licensing arise. The voluntary nature of the program that makes participation less complete than the NRC would like. Se NRC has debated about making the program mandatory as a basis for confirmation of the adequacy of the EECS models but has not done so. Mr. Tedesco and Mr. Odar reviewed the results of the NRC review of the GE ODYN code. In 1977 GE ran transient tests on the Peach Bottom Reactor. Re-sults of those tests indicated that the GE REDY code may not be adequately 7'n r4 4

ECCS Mtg March 19-20, 1979 conservative. Se new ODYN code was developed to accurately model the reactor system transient during main steam line isolation and other events which perturb the reactor or steamline pressure (Attachment R). The new code is an overall system model with one-dimensional thermal hydraulics, neutronics and fuel elements in the core. Se system includes the core, recirculation loop, steamline dynamics, and the control system. We mo?als included in the core were briefly reviewd (Attachment R). We Staff evaluated the uncertanties asrociated with various portions of the code and compared calculations to test results at Peach Bottom and KKM. Rey have also preformed audit calculations. The Staff has concluded that GAN predicts peak pressurea conservatively and predicts CPR valves -in a generally conservative manner. Based upon the limited data available the Staff has concluded that they cannot give credit for the conservatism at the present time and that further tests are needed to reduce uncertainties. Se Staff has concluded that the code can be used in tansient analysis and that if necescarv a factor of conservatism will be provided later. We final part of the review is still on going. Mr. B. Sheron reviewed the status of the evaluation of loads on the reactor pressure vessel, core internals and shield walls and subecmpart-ments due to blowdown hydraulic forces and pressure waves. Westinghouse has subnitted WCAP-8708P in February 1976 and the NRC issued a topical report evaluation in June 1977. Se NRC evaluation required several changes to the code as a result of their evaluation. Mere has been no new developnents with regard to W plants. S&W submitted BAW-10132P in March 1977 and the NRC staff issued a topical Report Evaluation on October 31, 1978. 2e Staff concluded that the loads calculated were conservative. 1511 2i9

ECCS Mtg March 19-20, 1979 CE sut:mitted their topical CENPD-252P in December 1977 and the Staff issued their report on February 12, 1979. W e Staff concluded that the CE loads that were calculated are conservative. The GE subnittal was made in September 1978 (NEDO-24048) and the first round of questions on it is to be subnitted cn June 1,1979. In addition, the Garman HDR Test Program will address a number of the questions on fluid structure interaction which have arisen. IASL also will be releasing their SOIA-FLX and K-Fix (FLX) codes which can perform these calcu_ s vs. Dr. Catton indicated that he was interested in the ability to calculate the travel of a pressure wave through the core barrel and across the core. Mr. Odar presented the status of wrk on the asynmetric loads on the outside of the reactor vessel. In general, each vendor is developing a generic approach to the problem based upon the geometrical characteristics of their plants. Following the generic model developnent plant specific analysis will be performed as necessary. Generic analysis should be ccrnpleted in June 1979 for B&W and CE, January 1980 for W, and the GE schedule is under developnent. Mr. D. Hansen reviewed the present status of the analysis work on Semiscale test S-07-6. Bis test was initially reviewed at the January 19, 1979 ACRS ECCS Subcomitteee meeting. Subsequent to that meeting additional work has been done in an attempt to explain the observed phenomena. It is now believed that the delayed quench and oscillatory refill behavior is caused by excess heat transfer in the downcomer walls causing a mass depletion. 21s initiating event along with the system interaction of steam condensation on the subcooled ECC injection is thought to produce the phenomena. A calculation with TRAC out to about 250 seconds shows some oscillating behavior but the magnitude is not right and the perod may or may not correspond exactly. Calculations have 1511 220

Msrch 19-20, 1979 ECCS Mt9 In some respects the code is not adequate to been done with REIAP. model the observed phenomena due to water packing problems and hydraulic INEL did some studies with varying parameters, modes, flow balances. flow rates in order to more correctly model the observed conditions and try to understand what the system was doing. S e new calculations tend to show Mdi-the oscillatory behavior but are not of proper magnitude or frequer.cy. tional work is needed to clearly understand and model the system behaviour. A review of other data available also indicates that mass depletion effects have been present in other tests (Semiscale, LOFT) but was generally not noted due to its minor effects on system performance. Based Mr. B. Sheron reviewed the status of NRC Staff Action on S-07-6. upon the work done by INEL and IASL with MAC and REIAP, the Staff believes that depletion is real and based on well-known principles, prese* t indications tre that depletion Fhenomena can occur in EMR's, and additional work is W e Staff has asked necessary before final conclusions can be reached. LASL to make modifications to W AC to better predict bypass, then S-07-6 will We need to request be recalculated and a fine node EWR may be calculated. information from applicants or vendors is being considered. Mr. L. Phillips reviewed the present status of the W Upper Plenum Injection. The UPI system injects HPIS and LPIS water directly into the upper plenum. In November 1976 the staff Accumulator water is injected into the cold legs. reopened the review of UPI plants based upon the UHI review (The old ( UPI During 1977 and early model assumed all water was iniected into the cold legs). 1978 a series of meetings and reviews were held with the NRC, W and'the affected utilities. In January 1978 the Utilities committed to revise their ECCS evacation models, An interim bases for con-to treat NRC concerns on injection water location. tinued plant operation was provided. Operation of the plants has continued g\\ \\ 22\\

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rurch 19-20, 1979 ECCS Mtg based upon the interim calculaticn. Se vendor medel develognent should be complete during the spring and summer of 1979 and the NRC Statf expects to complete their review in September 1979. Se NRC Staff Mr. B. Shoron gave a brief review of the WI review status. issited an SER in April 1978 indicating that the model was acceptable but required model verification by appropriate experiments to be conducted. He first WI test will be Semiscale MOD 3 was to be usef. for this test. evaluated by NRC and its contractors to determine if it is appropriate for Subcommittee consultants and members indicated that the a W calculaticn. one-dimensional nature of Semiscale MOD 3 may produce atypical results not applicable to W WI plants. PNL is performing COBRA calculations for blowoown and reflood in the WI We CTBA boundary conditions were obtained frczn a calculation system. performed by W_ with their system code. Bere A movie of the comp 2ter calculation with CTPA has shown by PNL. IASL were strong three-dimensional effects apparent in the fluid flow. will be performing a calculation of the McGuire plant with 'mAC to evaluate the WI behavior. Dr. Plesset thanked the participants and adjourned the meeting at 4:30 p.m. For those who desire more detailed information a complete transcript is available at the NRC Public Document Room, at 1717 H St., N.W., Washington, D.C., or from ACE Federal Reporters, 444 North Capital Street, Washington, D.C. A complete copy of all vu-graphs presented is on file in the ACRS office with the record copy of the minutes. I511 222

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11870 committee. Its consultants and Staff. "J590-01,M] .-f regulatory efforts in accornplishing Persons desiring to make oral state. Acvisoty CeuxiTTit CH ttACTCt $A71.-j. the general purposes set forth in the ments should notify the Designated custos HucttAt RIGut.ATCRY CCMws.$ Clean Air Actt -Appropriate automobile emisslan' Federal Employee as far in advance as 3, 510H stancards and best available technol. practicable so that appropriate ar. rangements esa be made to allow the g,,g,g g,g,,, y,go, a ogles needed to meet them -Most. appropriate and practical nece=sary time during the meeting for Regarding the previous FtstaAr #' means of preserving air quality in such statements. Rzctsizst Notice (published on Febru ad. areas in which the air is now cleaner The agenda for subject meeting Volume 44. P.10557) for 4 ary 21,1973. the meeting of the Advisory Commit. D than the national ambient air quality shall be as follows: -Most appropriate and practical MONDAY.,Manc t 19. AND TUESDAY. ted on Reactor Safeguards to,De held "In Wash standards; 1Aac2t 20.1973 on March S-10.1979. means of enhancing air quality in those areas in which estatushed air s:30 A.M.UMn. NIX CoNCt.UsloN OF D.C A change for the items Deing dis. M. cussed on Friday March 9.1979 has y 8 SINES 3 EAC)! AT been made as follows. g pecla! roc e of sm business The Subcommittee may meet in Ex. and govenmental agencies in obtain. ecutive Session, with any of its consul. Fatmay.Maac2s 9.1979 Ing nductions of emissions from exist. tants who may be present. to expicre 3:J0 com.-IO:00 a.m.I Erceutfre Ses. -F ing sources to offset increased emis. and exchange their preliminary opin. lons regarding matters which should sion (Open)-The Committee will dis. C stons from new sources; -Alternatives to regulation as a be considered during the meeting and cuss the role and responsibilities of 71-means of reducing pollution to formulate a report and recorn:nen-the ACRS in the NdC regulatory 4 -Inherent problems in efforts to di. dations to the full Com=1ttee. At the conclusion of the Executive. *rocess. minish pollution in high altitude 10:00 a.m.1:15 p.m. and 115 p.m.. Meeting with NP.C Staff D areas; and Session. the Subcommittee will hear 100 p.m. d -Relationship of established env1 regulations to national. presentations by and hold discussions (Open)-The Committee will meet with representatives of the NRC Staff. with members of the NRC Staff to m ronmental ener:y policies. These wishing to testify should and their consultants, regarding the hear reports on and to discuss recent w noti.fy Paul Freeman at (202) 634-7138 zono ving topics: operating experience and !! censing ac-s by March 7 in order to schedule a time (1) Code Work on Transient Two-t!cns including proposed changes in if for submission of prepared oral testl. Phase Flow the Technical Specifications for the i and should send at least 50 (2) Status of Physical Inputs to Dresden Nuclest Power Station Unit 4 many copies of such testi=ocy no later than Codes

2. The Committee will also hear and
  • Nr.rch 14 to the attention of Paul (3) Analysis of LOFT L2-2 Test T

Freeman at the office of the National (4) Status of ECCS Related Re. discuss reports regarding generic mat. ters related to the regulatory process .m Commission on Air Quauty 1730 K search Progra=3 (5) Standard Problem Program including the enteria and procedures M Street. N.W., Suite 007. Washingten (6) ODYN Code Review for imposition of Civil Penalties (10 =; e D.C. 0006. (7) Status of Analysis of Assymetric CFR Part 2.005), the Nonproliferation NAneN A1. Co3CdissioN oN Blowdown Forces Alternative Syste=s Assessment Pro-a h Ara CNT. (8) Status of Current Licensing Ac' gram and the shipment of spent reac. Wtu.zA:4 E. Lrwts. Jr tor fuel elements through densely + addition, it may be necessary for populated areas. the Subco=mittee to 1;old one or more The future schedule for ACRS activ. Im Doc.19-6596 Pled 3-1-73: 11:31 aml closed sessions for the purpose of ex. it!es Mll also be discussed. ploring matters involving proprieta.ry 100 p.m.-rJ0 p.m.: Exteuttre Ses. information. I have determined. In ac. sion (O;e.2)-The Committee will hest [7590-01-M) cordance with Subsection 10(d) of and discuss the report of its Subcom. ?j Pub.1. 92-463, that, should such ses. NUCLEAR REGUt.ATCRY sions be required, it is necessary to mittee and Consultants who may be s i COMMIS$1CN close these sessions to protect propr!- present regarding the ret;uest for an etary infor=stion (5 U.S.C. 55:b Operating License for the WiU!a:n H. E Zim=er Nuclear Generating Station -7 AovtsetY COMMrfill CN ttAC7Ct sArt. M 4)). cusacs, susccMxinit CH tuttotscy Further information regarding g nig 1, M coat CccuMG sYsitM5 (Kc.3) topics to be discussed, whether the Portiens of this session n11 te closed 5 meeting has been cancelled or resched* as required to discuss Proprietary In. uled, the Chair ~an's ruling on re. formation related to this facinty and 5 W+the zency Core Coolin{ Syste=s will hold {g 0[ 8jppe t I arrange =e ts r the physical seennty @, The ACRS Subcommittee on E=er. 3 + dt of th:s station. J JO p.m.-7:02 p.m.: Witzicm g. j a meeting on Marc.. 9-20.1973 at the

    • 0 Travelodge Internatiooal Hotel. 9730 f' 0[ 0*

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g.:st 1 (Cpen)-The Co=mittee v.11 i Airport Blvd., Los Angeles. CA 90045. ral Em3i0 f0[g,th {tice of this meeting was pubushed 3:57.) between 3:15 a.m. and 5:00 p.m.._ hear reports from and hcid discussions 3, 30c, g;,g34 _ j with representatives of the NRC Staff -a out!med.r.Ence ~ rith the procedures October 4.1978 (43 FR 459:S), oral or Dated: February OS.1979. ~ and the AppHeant regard!nd propcsed 4 o ES** .n the Pcow. Rrcism on operation of this unit. 2 wruten stater:ents may te p esented Jome C. Hov:.r Portiens of this session MII be e!os*d 0 as required to discuss Preptfetary In. y by members of the pubuc. recordings N.uancpeme"st O// ice. for=stion related to this factuty and 4 I will be pernutted only during those portions of the meeting when a tran. La Occ. Nm N M-;s: Sa5 aml+ ar angegnts for the phystes! protec. ? script is being sept, and questions may tion of th23 station. j be as.ked only by members of the Sut>. .7 FlotA.At alGIST12. VCL. 44. NC. 43 10 DAY. M.AsG L W s u m wlw ai Ju 'DL r , n

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..1, v. %.,,._.,,i.,,,. ....j TENTATIVE SCHEDULE ACRS ECCS SUBCCri4ITTES MEET. I!G MNtCH 19-20,1979 Mondav, March 19, 1979_ 8:30 a I. Opening Comments - Executive Se. sion a 8:40 e II. Review of Code Work on Transient Pao-Phase Flow (a) Mathematical Aspects of Advanced Codes i. Stability of numerical techniques 11. Well Posedness of equations (1 hr. 50 min ) 111. Sensitivity to input /init;'il conditions iiii. Ability to Jalculate steady s u te pretest condit. ions 10:30 a - 11:30 a ~ (b) Status of Physical Inputs to Codes i. Thermal hydraulic dat a available and needed ii. Data uncertainties (1 hr.) iii. Sensitivity of caletia' lons to data c uncertainties liii. Flow regimec (c) Presentation by LASL on TRAC regarding (a) and 11:30 a - 12:30 p (b) above 1:30 p - 2:30 p (2 hr.) 12:30 p - 1:30 p LUNCH III. Analysis of ELFT L2-2 Test .:30 p - 3:30 p (1 hr.) (a) Post test calculations with %\\C and REEAP Heat transfer correlations and ability to calculate rewet/ quench phenomena (b) Pretest predictions for L2-3 (if available) e 1511 224 r

- -., :.. :...e:, c. n.... v... .. ;,,...~ su.,.. o. .i..... r.. - e:.....: u.w ;...... +. 1 ~ f s.e ~ ; EECS Meeting March 19-20, 1979 IV. Status of ECCS Related Research Programs on 3:30 p - 5:15 p (a) 2D-3D Program (20 mins.) (b) BWR Core Spray Test Program (Lynn, Mass, and NUS) (45 mins.) (c) Creare Downcomer Program (20 mins.) (d) LOBI (Facility description and planned tests, (20 mins.) relationship to Semiscale and Rosa) Tuesday, March 20, 1979 I. Opening Comments - Executive 8:30 a - 8:40 a II. Standard Problem Program (1 1/2 hr.) - 8:40 a - 10:10 a III. CDYN Code Review (1 1/2 hr.) - 10:10 a - 11:40 a IV. Status of Analysis of Assymetric BloWown Forces (1 1/2 hr.) - 11:40 a - 1:00 p LUNCH 1:00 p - 2:00 p V.. Status of NRR Action and work to be done on: 2:00 p - 4:30 p (a) Analysis of Semiscale Test S-07-6 (1 hr.)

  • (b) W - 2 Loop Plants (1/2 hr.)
  • (c) UHI (1/* hr.)

(* Interested in 3-dimensional flow considerations in the above.) 1511 225 ,... n. _.n.. .n. s

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............ :,5.. ACRS ECCS SUBCCMMITTEE MEETItG G CH 19-20, 1979 II)S A!GELES, CALIFORNIA ATTENDEES LI E ACRS NRC M. Plesset, Chairman S. Fabic D. Okrent, Member L. Tong H. Etherington, Member L. Shotkin, I. Catton, Consultant W. Bixby Z. Zudans, Consultant W. Bechner H. Sullivan, Consultant L. Phillips R. Alleman, Consultant R. Tedesco R. Shumway, Consultant F. Odar T. M1, Consultant B. Sheron F. Zaloudek, Consultant K. Garlid, Consultant f.ASL T. McCreless, Staff A. Bates, Staff

  • B. Wendroff D. Liles
  • 0esignated Federal Employee J. Jackson K. Williams EXXON NUCLEAR CO R. Pryor E. Chapyak R. Collingham W. Kirchner R. Kelley J. Vigil S&W INEL C. Lin W. Grush L. Leach BECHTEL PChER CCRP T. Larson S. Field PNL hESTI3CHCUSE ELECTRIC CCRP M. Thurgood R. Alench i5!1 226

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matrix. 1.et u be an eigeavalue af A with aigertvector r. h complaz .m. .' Consider a sequence '~ '- ~ b conjugata e is also se sig,esvalue with alaenvector of initial data 7. na +.- -ikx' .J.. v (z.0) = A {t [ k = 1.2..... re re g g i, .o w. f...' ' obsce 2 u,. lim e km .I. h initial v (z.0) evolve into the solutions v (z.t). where h. g 7, g ikW) + r e-ik(iMst) 1 (P 1; ' v (z.t) - A re .g: g g p< i" 'If a is real, then lim (x,t) = 0. which is another way of saying that the h+= .h.... solution depends continucesly on the,taitial data. This, them, is the test: s ,1.. All eizenvalues of A should be res! .7. '.- ,.n= Ar .~ -- .T - 6.9' 3 4. t ."g.. .1 '~ OC '.E."*.,. g. p , l y- ' tf r ~ A-U3 L. '. A.Wd' cd i f.1: ; 3 u. t + ^ 4 s- ,j.,, ~) 1,. - . y,, -,, ...x.~p*.... s t i,.,'*' - g e .? g 7,

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..;;; % *;.-^ ? :.'. is - y,,*,. ..4 .t.. s 4 f.t : . u (s. 6 s. 1. ,, n PRESSURE ^^ s. votD FRACT 1.0 _ o.205 t = i .) .\\ r

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..a -

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1. :

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-1 L2-2 STEADY STATE h1 ,e PARAMETER L2-2 DATA TRAC (POSTTEST) TRAC (PRETEST) F IllTACT HOT-LEG TEMPERATURE (K) 580.6 580.8 593.0 IllTACT COLD-LEG TEMPERATURE (K) 558.8 559.0 566.0 k .g CORE AT (K) 21.8 21.8 26.6 IllTACT LOOP MASS FLOW (KG/S) 197.5 2 07.1 18E.6 3.. f. PUMP AP (PA) 9.1 x 104 9.2 x 100 7.8 x 304 155 x 105 I', PRESSURIZER PRESSURE (PA) 155 x 105 155 x 105 L - STEAM GEtlERATOR SEC0flDARY 5 PRESSURE (PA) 63 x 10 62.0 x 105 5 'f. 63 x 10 tn g 2 U i ~u ~

...,, :.... f.: x . c... ' ~ ...ii:2.<. 5:,. :. G, TRAC L2-2 POSTTEST CALCULATION AGREEf1ENT WIiH EXPERIMENT 1. ALL SYSTEM PRESSURES 2. DOWNCOMER BEHAVIOR SIMILAR TO LIQUID LEVEL PROBES 3. I!ME TO FIRST DNB (~ 1.5 S) 4. BIG BLAST OF POSITIVE FLOW THROUGH CORE JUST BEFORE PCT 5. PCT AT CORRET TIME (~ 6 S). TRAC PCT 840 K, L2-2 790 K 6. MULTIPLE REWET/DRYOUT ON SOME LOW POWER RODS 7. ECC (HPIS, ACCUMULATOR, LPIS) INJECTION TIMES AND RATES 8. PRESSURIZER EMPTIES AT 35 S 9. CORE VOIDS AT 20 S -- LOW-POWER RODS DRYOUT 10. REFILL START 3 AT 23 S, DATA 25-30 S 11. CORE REFLOODING BEGHINS AT 40 S 12. LOW-POWER RODS QUENCH AT 40 S 13. HIGH-POWER ROD QUENCEe AT LOCATIONS BELOW CORE MIDPLANE AT 40-45 S 1511 240

, t;, .. :... " : ~.4... ].. l?a... -. .. v iYa i.. ..... crsz. a.-.., -s. z G-.3 TRAC L2-2 POSTTEST CALCULATION {-2PHENOMENANOTPREDICTED: 1. EARLY REWET AT 6 S OF HIGHEST POWER LOCATION 2. MULTIPLE DRYOUT/REWET OF HOT ROD BEFORE 30 S 3. HOLD UP OF WATER IN UPPER PLENUM ? 4. QUENCH BY DOWNFLOW FROM UPPER PLENUM BEFORE 30 S-5. TIME TO FINAL QUENCH OF HOT RODS G 9 \\5\\\\ 2'\\\\

.....:......~..v u. w 1. ....: a. a.. ae - '.<. a, .u. ....._-~ .r. n...:....._... (,. - Y oo o g g-S.d _ c o Ju o .3 150. I i i. i i i i FR-BL-als i i i l ! l 7 e ! ; i, It ' l l i ! I i i i i l : i i i i l l i i l i ' !l --- T F AC POST TEST I ! I i ' ' i I I ' ' ' ' I i I ' i 100. 'i .i. i 8 i i i, I i e i

i l i ; i s

I en i i ' i I : i i

i I

i i i i ! i l l l i l

' i i

i i i i i i i i i i i, y I, ' I I I ' Q 50. 8 I ii ' n i i e i i e l i I ! i I 2 .-J J Yd' 3 %. i i i 7i O i ! t i i d l l9 I l'dk i I 3 h _m't ,.. _ _1 m 4 i m 0' "I W i@# e i ! l ! 6 i

  • i i

4 I l i i I i e i i i i ~ t I t I l l 4 i i t i .t i I i i -50. -20. O. 20, 40. 60, 80,

100, TIME AFTER RUPTURE (s)

Mass flow rate in broken loop hot leg calculated frcm (FR-8L-216). IJlI &dl

.s-, 23...- ~. :... a - .... i:.u.,.:...i.. ~ x, c 0 -f" IE i e - - RELAP4 PRETEST - - TRAC PRETEST! M

L2 -2 TE-5F4-21

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j 2

~ =" [/ I, i i w \\ & 500 a 's -- - - - -(.l 250 I -50 0 50 100 15 0 200 TI M E (S) LOFT Test L2-2: Temperature of cladding of fuel assembly 5, rod F4. 800. T E-1 dl l -028 3 e o TE-I B i l - 032 A TE-4GO2-030 2 O T E-4G14-030 --- TRAC POSTTEST-700. w (I D r F i 'i i 600. 1 3 2 N s - 1 w \\ l }- t I 500. a g 4 / J ( U Ws-- ~ I I I 400. 0 25 50 75 i00 TI M E (S) LOFT Test L2-2: Temperature of cladding of fuel assertly 1, rod Bil, and fuel assertly 4, rods G2 and G14. 1511 243

, i... -...a ...s.:.:.u..M '.... ~........ art.l e.W s. .....a.>.,-a :...... ~ ;: SEVERAL RUNS e e e< op Q FPIPE l i e e el SP/5 e e e o e e' TIME INPUTS VARIED SIMULTANEOUSLY ENVELOPE OF ACCORDING TO LHS OUTPUT VALUES PRIMARY QUESTIONS TECHNIQUE FOR ANSWER 1. WHICH IltPUTS CONTRIBUTE PARTI AL RANK CORRELATION $16MIFICANTLY TO THE SPREAD IN 00TPUT YALUES? 2. WHAT RANGE OF VARIATION TOLERANCE LIMITS IN OUTPUT VALUES CAN BE EXPECTED? Ij j i ts 15\\\\ 244

.~ c,.. ~....... ;...:a u1.03. '...... ~. u a.... : w... &:...- ' n. :::.. ~........ = -u-s. 1. /y t I IPPORTMIT INPUTS +1 C TIME I e HL / / / / / MV i s- -1 l. t. I i PAJtTIAL RANK CORRELATION COEFFICIENTS MEASURE DEtJtEE OF MONOTONIC ASSOCI ATION BETWEEN t OUTPUT AMD INPUTS h hj b 9 0 } J{' ll 443 j ., f -

h i. c TABLE I '^ IIEAT TIWJSFER CORRELATIONS Eqs. Ref. Regirre Flag Correlation 0 135 7 Forced convection to 1 laminar flow : constant IJusselt number 136 41 i turbulent flow : Dittus-Doelter single-phase liquid 137 6,36,42 lAncleate 1: oiling airl 2 Chen forced convection va prization 126 33 Critical heat flux low flow : Zuber pool boiling 127,128 34,35,6 b high flow : Biasi 138,139,140 36 Transition boiling 3 log-log interpolation 131,132 38,39 14initiamn stable film low pressure : llenry-Derenson 133 36,40 honogeneous nucleation C Loiling high pressure : 141,142,143 43,44,45 Film boil ng 4 rnlified Bromley 144,145 46 Dougall-Rohsenow M 146 47 tbrced convoction to 6 free convection : McAdams ~ turbulent flow : Dittus-Boelter 147 41 single-phase vapor ) 148 7 Forced convection to 7 laminar flow : constant flusselt number 149 41 turbulent flow : Dittus-Boelter Ltx> phase mixture 150 6,48 h llorizontal film 11 Chato i corvlensation 151 6 5 Vertical film 12 tAssselt theory N corxlensation ^;.r C= i' 152 49

  • Turbulent film 13 Carpenter arxl Colburn j

condensation L s;' '-I s

0 GENERALIZED DOILING CURVE g. I I l 8 i 1 I I 1 l TEMPERATURE SUDSCRIPTS l l l l o;{ I l SINGLE-PHASE , NUCLEATE I TRANSITION I FILM 00lLING 1. I $ = SATURATION LIQUID I BOILING l DOILING y.- 3 l l I x i CHF = CRITICAL llEAT 8 3 l l FLUX i e g 8 a l l l MIN = MINIMUM FILM t-a l 4 e l DOILING y l I i l w 'A l l I e i 8 l l a i i i I l i I - y 1 I I 8 I I i l .a. l l I i l > c. i 1 I f T ,g T T g 3 cg, T, - SURFACE TEMPERATURE Ln a 'c. N 4 N \\ 1P Fig. 6. Generalized boiling curve.

+. .&.. w.. 2-.

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c. 5
V.'

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...,........:H 7 . :.. e' h ) h ) l CON D GN.ch rio6/ Suoeg 3yg,g E ) h f2 ; W fd E) o Nb w-u~ 9 M CREARE' bjn MATHM/ktw SCC h/)(/H/- Cd44 g/4 l T 1511 249

BA S /C TE S 7"$ ~ $ N d e r 3 ~g_3 ~ ~ ~ ~ V. HeksrpN 1 1 fg (A/R/WAran) $ "FL FILM rHict<NGss, RCS sin H er4Rs (srGAM/ ) wraa. , ~ (- =

==, (g) CREARG 4" M 't CCFL A/AlwarER f k A I bA AllEL N ATMottrH K. THB5$ ggggggg CCFL (A/RlvATHX) SP 't t t t N WR, SIE vd PLATE CCFL 5pRr ~ ~. -

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Af ec .I 6 f, c i 1 k if PPER CORE $4f?rRr TNEL CCFL <- ~ ~~:,,*x 4-PL A TE A W, K VW GG0HGray f&(( g $8 $e & ,smpta gguc45 o u.o o o 1511 250

-a 4. + c ~ , y y g - T G 3 T5~ ' ** *"' ' K-y r $f. 4GH VM/Al _ g.y...-~ q.Q W ~ ~ ~.~

.....~x.,S-.Ydb 6caar-ORIFICE l0[2); Ql2) ocrz,s) 4-CANpN P/Ps (CEN/C-)

Slow = b 98 val AwRG-(g) 0y

fewtnen, g.,g, _

Y & fa, t) c/se> -nevr ^ " " ' ErrBE }9 ~ ' (R 'o y MARV/KEM C A / 7: c D/sCH. J -f,T,,i-cer, ~ ~ ~ ~ - BArrGU*- ~ ~' ~ FRANKFWAT,* C005 ALS EL5h* %,d C s E, m B 577/$ = FLACM tg k s,s, 9 9 V#'O I (g) FR/G \\ y=f (swebGH) b/.57' RIG. Cd [h> 'E) sreA b y PLossi & (A s) Gy ns.sers. M2 i 4^$ T [1) con re ot.t.ao ra A M.u a n 75 ] HEATED GWNOLG t o,, t t it \\ Noh A PI, yp/D AIR /wsreg gpj$ bisitif t wcuo k ~ e W HARWeLL 4' et, (arrs) ? ft 3 h 1511 251

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l HPPGR PLEN. 's i 2 RESP . q 7 tg-y u-- (caou now) &M [Q. P. '.".7 E *: ! *.~ : - ?v a W "I - 15!! ' l ill! ', lI ' ! " *h l/ w = ma i tt .itt itt F 1 r7 p FLpW yppyggg y A CKoss R (Hrps) op pas -*f ARR.Avs 6 i _ _ _ _.. _+ f2-bormo 15 1 252

/, t r Initial Condition Changes i Parameter Units EP Post-Analysis Power (MW) 25.295 24.88 [ Pressure - pressurizer (MPa) 15.60 15.64 l Primary coolant pump speed Pump 1 (rpm) 1190.0 1269.7 Pump 2 (rpm) 1220.0 1276.4 INEL-S-17 486 N Ln b

  • i U

g ~

Initial Condition Changes ~ . (cont'd) t Parameter Units EP Post-Analysis f Coolant iemperature Intact hot leg (K) 587.6 581.5 Intact cold leg (K) 563.7 557.7 i

t intact loop differential (K) 23.9 23.8 Broken hot leg (K) 566.3 545.1 4

Broken cold leg (K) 563.7 532.8 Steam generator secondary (K) 556.0 552.4 f S Accumulator A (K) 305.4 300.8 I7 Steam generator secondary level (m) 0 0.19 If INEL-S-17 487

IV l_O FT L2-2: Pre-and Post-Analysis initial Condition Comparison i 850 i .i h6 800 e La i 750 ue -Experiment prediction i y~700 3

Postlest 650 0)

O 600 s-a s 550. M Z 0 5 10 15 20 25 30 35 40 y Time (s) { INEL-S-17 498

.. D ~ OFT L2-2: Post-Analysis Cornparison of Filrn Boiling Correlations ~ 850-i i i i x " 800 ~'~~,- ',/ s a s's' = m 750 N l g o.700 E / t e m / -New Groenveld ~ e 650 i '~ o j

- C o n d.ie-Bengston m

a w s 600 / %%) i: f (j) l I i r-550 ~ K 0 3 6 9 12 15 ic Time (s) INEL-S-17 497 Vi?

. l' LOFT L2-2 Post-Test Analysis: System Model Clad y Temperature for Various [ Critical Flow Choices j; 850-i i i i i; M '1 [800 EP--Best PT [ --- PT-A a E 750 -- PT-B = [ s,, / e + -~~~,x~N ~~- - / a. / E 700 g ~ ~ ' t e y I 650 / \\ \\ [ 5 8 I./ l 600 5 M \\. E ~ W b l-- 3 I 550. O 3 6 9 12 15 8 h Time (s) O. lHEt -S-17 491 -)f/f,:

L-6 o k 3 Q i n I ( 3 ) N D E e A. s g' ) ~ I ) 1 z h \\ 4 R 0 RR 3 s ~ --c /n..\\ N m ~ 'a 7 ff D A o 1 8 G P 1 / e Nbh h $ l k w 9$ I n 'U l d X

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A x I k D N'DO 9 0 4-m s jGG 4 ii o 4 e Q 'M d R kh, ..E m .aa, a.d i.u. A. o = > *a s 15!1 258

6U6T/E bSS IlGl TV qmm m cru y[ [ 0 0 0 i J' _ E ,0 me H d t \\ g34 / I i g' u M.h3 - - ll.1_y$'\\ = = = - - - 'h_1 c. 7 7 $D J __ M .5 e CJ, t==, . - t._ e N! C-s, E h 1 g e>;- '( f / G) \\x Y-j s' V "l l I! 1 a S ,a h 5' f u s$ JN I // 3 \\ hn f _( b .1 & R=== Q-]-M 5 4.?er :tt,,y,li o \\J E J n-u

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  • ).# ~.".'.~. 3 J

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pl - t-SIDE VIEW OF TEST SECTION -I .e STEAM SEPARATOR STEAM DOME MOCKUP \\ r' /.. j/ UPPER PLENUM s MOCKUP ~ i / i, [l ~ [hi1 ~b7 / l / i.I: , Vl l r i i 9 x,f V HPCS <,3, s . N I LPCS N: N o t" r 3 c l ) fD ri- ~ -F ) f l'- i ). ip' yj l i I 4 EXTENDED ARRAY FUEL BUNDLE MOCKUP w .o f AJL 3/19/79 1511 200

M -3 PROPOSED (NRC/EPRI/GE) REFI;LL/REFLOOD e ADDITIONAL SYSTEM COMPONENTS e TRANSIENT BLOWDOWN EXPERIMENTS e THREE-DIMENSIONAL PLENUM EFFECTS e PARALLEL CHANNEL CORE EFFECTS = AJL 3/19/79 1511 261

it.k, 'Ihh, u.. jl.lu. f. 'h[a -jh jhi f*!,'jf,h-h !M ii]-{ h- (j{h((ph*[5[$,})5:!dIfY)0b8 N N N P " M N iT T M # # # E f i 5 [d2A -~ '- a a.,_ ~. n 20/3D RESEARCll PROGRAM SCllEltlLE JAERI REFILL-REFLOGO FACILITIES CALENDAR YEAR r 1973 1979 1980 1981 1982 1983 1984 [w 7 } Test CYLINDRICAL CORE TEST Constructio; Te-t FACILITY TCCTF A kedown

  • d Wu Core 2 Fab, Test Core 1 e

t TOKAI, JAPAN g _ 9 Test St A'l CORE TEST 005I9"IDU a Construs Lion n Test Core 3 { Shakedown FfCTITTG$Ufl n P ' Siral 'edown J Shakedown Core 2 Te: ts Core 1 Tests TOKAI, JAPAN Tests (Tobe" I ' N f 5c11-1 g CCTF 1 a M bSCTF 2 Sp. Pc., LLDs, 8 = SCTF ' THs, 00s & ET)s t. d SCTF 1 % 7 I CCTF 2 17,] ORNL_ a r , SCIF 2 y .g Film-Ispedance SCTF 'l Y L z E Probes 1 m U 9 Stereo leri ; I and I ost-Te st Analysis Xf r m LASL /\\ Cylindr cal Core Pri g Y 4 Stereo Lens & Slab Co e Design Pre ani) Post-Te t Analysis ,) l Analysis I, Na fT

w m.n _-_..,. g-- -r-., m.,~,..,. '., ,j n g. 6.n 1 "i.:,. - y,;., - s. ,;3- . o.. .U;.5't-['l-Y;;:;.lj,, 20/3D RESEARCil PROGRAM SCifE00LE T!! ~.' 'c :..., T.,- FRG Refill-Reflood Facilities Calendar Year ~ 1984 1981 1982 1983 1978 1979 1980 .in 1 Core inst _ m core 2 h sts i t i P_KL TEST FACILITY _ s ~Lest checkout KW cor F Erlangen,fRG Manufacturt & constructnog Test a UPIf TEST FACILITE ~ b KW/GKH Mannheim FRG 1 PKL flo-ins r. INEL L - O i Sp. Pc., LLDS' tjPTF fin instr. ~ IHs. 00S li Ds } LD ~ ) P prot >es ~ j OllHL o OPTF probes i N %L - 1,, film-Inped. tlPTF DAS o g; j y a _ r 2 Probes & DAS .I pKI-tiP1f stereo len-I 0 3 Stereo Lens UPTF design jpre & post-est analys1< j a i y p i a Analysis p i i l i_ g

J VOLUME AND VOLUME DISTRIBUTION FOR-LOBI, SEMISCALE MOD-3, AND ROSA-II LOBI MOD-3 ROSA II U.S. PWR 3 TOTAL SYSTEM VOLUME (M ) 0.817 0.195 0.836 321.2 VOLUME DISTRIBUTION (% TOTAL VOLUME) INTACT LOOP (%) 3 11 fl7 53 11 8 B94 KEN LOOP (%) 8 15 13 13 VESSEL TOTAL (%) 58 38 3 11 -39 UPPER PLENUM / UPPER HEAD (%) 10 lli 13 - Ifl CORE (%) 11 5 6 6 LOWER PLENUM (%) 7 8 8 9 DOWNCOMER (%) 37 11 7 10 ~ L '~ to L

d d i IMPORTANT COR PARAMETERS i.0Bl,SEMISCALEMOD-3,ANDROSA-Il ~ 1 LOBI MOD-3 ROSA-II U.S. PWR .1 HEATER ROD T'/PE SHEATH FILAMENT FILAMENT NA ilEATED NOMINAL CORE POWER (MW) 11.56 2.0 2.'2I1 3'111 ~ LINEAR HEAT GENERATION RATE ((KW/M) AVERAGE 18.0+ 23.8 10.7 23.6 MAXIMUM 21.2+ 58 60 55* 3 POWER / CORE VOLUME (MW/M ) 127.0 1S0.5 I16.8 186.5 3 POWER / SYSTEM VOLUME (MW/M ) 5.6 10.3 2.7 10.6.

  • APPROXIMATE - DEPENDS ON' PLANT.

+FROM STANDARD PROBLEM INFORMATION.

~ 7 a' ~ ~ ~ O ~ CORE GE0 METRY FOR o G i.0BI,SEMISCALEMOD-3,ANDROSA-ll ~ LOBI MOD-3 ROSA-II U.S. PWR BUNDLtARRANGEMENT 8x8 5x5

  • 9x9&
15x15, 17x17 NtlMBak0FRODS HEATED Bil 23 "96 39372+

2 13 11053+ UNilEATED R0D GE0 METRY HtATEDLtNGTH(H) i.9 3.66 1.5 3,66 RODDIAMETER(MM) 10.7 10.7 10,7,9,5 10,7, 9.5 R0D PITCil (MM) 1 14, 3 111.3 114.3, 12.6 1f1.3, 12.6

  • ROSA-Il15x15 BUNDLE

+ 15x15 ASSEMBLY e

77"Em:iifiCM31LTUMLC%7iWSuaamwn...m-. c; "i.D ,9 ' 1 P_EELIMINARY HIGilLIGHTS ECC BYPASS RIL 'l: 5' ?.

  • PRESSURE EFFECT - ACCOUNTED FOR THROUGil STEAli P p@.

!D-

  • EVIDEllCE TilAT J' MAY BE BREAKING DOWN AT 2/15 SCALE c:

Ik.

5. -
  • EVIDENCE THAT CONDENSATION MAY BE INCREASI

(. K. m INCREASING SCALE SIZE E.li '

  • HOST UNCERTAINIES IN TRANSIENT 110T HALL N
.~.,

PRODUCE RESULTS CONSERVATIVE WITil RESPEC s.- 'j i' CALCULATIONS g.- " DEFINITIVE RESULTS MUST AWAIT LARGE SCALE TEST $.j) w- ?.' t:: i!.'. Un [.:= h% ~ t k l*EY, w Ch N W:' 5 ( Y.V ~ ,a - { v t 3 g 1

,- ;. r.l ',.*

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\\\\o ?_m (q;;e ~ n w l...- \\~" m EU,.$... w 4 CONSTANT J' wO s ,c.m 4 m\\ 0.20 m 4;.u r) 6/6 o w \\ e. eo \\ o -o i o V. :. H [./... >N. L z t m o -r.- ~ 'P V O 0.15 N s !lQ.li. f.. i m o z m yu.r w n..' o r:0NSTANT MOMENTUM FLUX' g,o O gg,. zo q?!:. ' l.- i 0.10 o s. s M 3i {9;. 95% CONF 8DENCE LIMITS --eg O AIR-WATER DATA O AIR-WATER AND STEAM-WATER DATA I(' 0.05 'S STEAM-WATER DATA p..

  • a*

~' U~1 .).} ~ a : R. 1* rx; g l 15 gjz. O-0.00 10 5 co 0 g PERCENT FULL SCALE I I,,h' FIGURE 15 - EMPIRlCALLY DETERMINED G AS FLOW FOR 100% BYPASS W/O CONDEN [3 A FUNCTION OF SCALE SIZE. S _s. 1i

. ~.&..L % n.:s.~..:... u...*- - '.:. % n....%' V:..-;..'c M,.,:.....:1^.i*:.& -:. ........ ~.. ,c,,..

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w;.;..,:(t:Ch...

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STNGJO PR03.Ei PROGPRi PIE DEFIES 03)ECTIVd5 #6 PRC/ IDES GJIELIfES U .g .m1. ~ FOR BE C0!DJCT T BE ST#0ARD PROBLBi PROGPMi LICBGillG RDE 9.ECTIGi 0F A STNUJO PFORBi n. x FSORT T hcautio .4*. FOST-TEST #MLYSIS .1' FItML REPORT e P . I .3 . g a,. a. 4

r.

.'T. k,. .f 1511247

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. 4-Q t,,. a. .~ STARIS_ COMPLETED ST AND ARD PROBLEMS "5 .y PARTICIP#fiS u

1. ENRD3 PIE EXERIENT 6

a .,1

2. SEMISCALE IS0THEWAL BLOWDOWN (TEST 101D 5
    • i
3. S MISCAlf ISOTHEWAL RDe0WN 5

WIW ECC INJECTION CTEST 1009) -( .S LI. WO LOOP TEST APPAPAWS GEST 6) 5 4

5. SEMISCAl.E El.0EGil W1E El.ECTRICAU.Y 5

3. EATED CDE (TEST S-02-8) e% .s 5 5

6. SEMISCAl.E 9'ALL BEAK AREA
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.s. ' i-a3 ..lh i 's I ENM PR0aRI5 SOEUlE Cl1979 l 1 I;. gg s @BtBi 7 (UFT L 1 - O .s. gaga mas a (s - m - E - -i Ex. mas e es= sTRm 2B84 10 (uFT L 2 - E .). s. .g. 1511 27; g

ALA }"' tEUTPONICS ASS 11PT10tG

1. OE BERGY GFOUP DIFFUSION THEORY
2. OE DIfESIO"AL (AXIAD, TIE DEF90ENT

-- VARIATIG{ 3, SIX DELAYED EUTRG{ GROUPS

4. RADIAL EFFECTS ARE TAKEN CAE OF BY C011APSItlG A, FDDEPATOR B. DOFRER C

CONTROL STATE Ze = Zg ca (1 + x e ue - V ca] + ~ c o e [uco - v co7 2}, bc o '6 HEE pg UO = P1000 PO o U,0 = P1000 1511 272

g _1-FULL HEAT TRANSFER MODEL 1. AVERAGE CYLINDICAL FUEL AND CLADDING MODEL 2. GAP CONDUCTANCE IS INPUT 3. CONDUCTION PARAMETERS ARE TEMPERATURE DEPENDENT 4. UNIFORM POWER GENERATION 5. HEAT CONDUCTION AND CONSERVATION OF ENERGY EQUATIONS ARE SOLVED SIMULTANEOUSLY 6. VALID ONLY ~FOR SHORT TERM TRANSIENTS - NO CREEP OR SWELLING 1511 273

g3 II. RECIRCULATION SYSTEM INCLUDES: 1. RECIRCULATIOM LOOPS 2. RECIRCULATION PUMPS 3. JET PUMPS 4. PLENUMS 5. SEPARATORS 6. BULKWATER MASS, ENERGY AND MOMENTUM CONSERVATION EQUATIONS 1511 274

g-V / STEAM LINE MODEL 1. SINGLE PHASE MASS AND MOMENTUM CONSERVATION EQUATIONS 2. EXPLICIT FINITE DIFFERENCING 3. ISENTRIPIC EXPANSION - COMPRESSIBLE FLOW COMPARISON WITH MOODY'S METHOD OF CHARACTERISTICS SHOWS THAT SELECTION OF 8 N0 DES IS SUFFICIENT. GE ESTIMATES FOLLOWING UNCERTAINTIES IN TERMS OF ACPR/ICPR: 1. PRESSURE LOSS COEFFICIENTS 0.010 2. SPECIFIC HEAT RATIO 0.010 STAFF FINDS THESE UNCERTAINTIES ACCEPTABLE. 1511 275

gf PEACH BOTTOM TESTS PEAK MEASUREMENTS DURING TESTS TEST P8ESSU8E net)"RON FLUX ACPR/ICPR (PSIAJ d RATED) TT1 1042 239 .170 TT2 1052 280 .136 TT3 1069 339 .132 CALCULATION OF PEAK VALUES WITH ODYN TEST PRESSURE NE RON Flux aCPR/ICPR (PSIA) RATED) TT1 1070 338 .173 TT2 1072 438 .129 TT3 1100 400 .141 1511 276

gsb o ci \\ ( t f' ~ m i Li n ei C ~ ej \\. e s g E a. e e a 8 Q ) n t ? n_ .9 ~ e w c o2 4 en 3 N 8 c a Mc .= N I A E I ci \\.N. m ci I t t o e o o o e e 2 8 o o N o e e e 8 2 2 s e e (!sd) 380SS38d 3 WOO 1511 277

600 625 450 MODEL - DATA O 375 F <f " a00 .I _e x au $225 n. >= y150 i O \\ 'E \\ 75 e------- '% m.-- * -. 0 ._.~.~._._._._, -75 l l l l l l l l l 1 I I I I -I 0 0 0.15 0.30 0.45 0.60 0.75 0.90 1.05 1.20 1.35 1.50 1.65 1.80 1.95 2.10 2.25 2.40 TIME (sec) n Peach Bottom-2 Turbine Trip 3 Prompt Neutron Power m t Q N -i s a

9r g/ CONCLUSION FROM TEST RESULTS: 1. ODYN PREDICTED ALL PEAK PRESSURES CONSERVATIVELY 2. ODYN PREDICTED TWO OUT OF THREE ACPR/ICPR VALUES CONSERVATIVELY THE DIFFERENCES ARE BETWEEN -5.1% AND +6.8% (+) MEANS CONSERVATIVE (-) MEANS NONCONSERVATIVE 3. FROM SPREAD OF THESE DIFFERENCES /' = 1.14% Cr = 4.94% VERY LIMITED DATA STAFF JUDGES NO CREDIT FOR CODE CONSERVATISM AT THIS TIME 4. FURTHER TESTS ARE NEEDED TO REDUCE UNCERTAINTIES 15\\\\ 279

R-7 / KKM TEST PEAK MEASUREMENTS DURING TEST P8ESSuBE NggTacN Flux aCPR/ICPR (PSIA) (A RATED) 1065* 220' .077 CALCULATIONS OF PEAK VALUES. WITH ODYN PRESSuBE Ng% RATED) uracN Ft.ux aCPR/ICPR (PSIA) ( 1095* 250" .084

  • ESTIMATED FROM FIGURES 9

1511 280

O gI CONCLUSIONS FROM TEST RESULTS: 1. ODYN PREDICTED PEAK PRESSURE CONSERVATIVELY 2. ODYN PREDICTED aCPR/ICPR CONSERVATIVELY 3. IT IS NOT A DEMANDING TEST g = 0.077 (MEASURED) ICPR = 1.279 g'=0.084(CALCULATED) q a(ACPR) = 0.007 4 THIS CORRESPONDS TO 10% CONSERVATISM 5. GE STATES PRACTICAL ACCURACY IN ACPR IS 0.01 6. STAFF JUDGES NO CREDIT FOR CODE CONSERVATISM AT THIS TIME e

,( t I AUDIT' CALCULATIONS: 1. TEST PROBLEM TURBINE IRIP WITHOUT BYPASS PEACH BOTTOM EOC2 2. USE RELAP 3B AND BNL-TWIGL 3. USE PEACH BOTTOM TESTS AS BENCH MARK 4. TWO BNL MODELS 5. THERE IS ABOUT 20% DIFFERENCE IN POWER INTEGRAL BETWEEN TWO BNL CALCULATIONS OR BETWEEN GE AND THE FIRST BNL CALCULATION 6. THIS UNCERTAINTY IS CONSISTENT WITH CODE UNCERTAINTIES W 4

C'= r p EVAUATICfl 0F FAPGIN

1. STill %8 E5 2

A. TEE IS to GWEE IN COE B. C[E CN1 BE lJSs IN TtWSIER N% LYSIS C. A FACT m (F CONS EVATISM, IF ANY, Will BE ppovite IATER 1511 283}}

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