ML20148S806
| ML20148S806 | |
| Person / Time | |
|---|---|
| Issue date: | 10/21/1978 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1577, NUDOCS 7812040039 | |
| Download: ML20148S806 (128) | |
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-MINUTES OF THE ACRS SUBCOMMITTEE
',;,/g j /g MEETING ON EMERGENCY CORE COOLING SYSTEMS WASHINGTON, DC AUGUST 28, 29, AND 30,1978 fb K "fuhr The ACRS Subcommittee on Emergency Core Cooling Systems (ECCS) met with the NRC l
l Staff and their consultants in Room 1046, 1717 H St., NW, Washington, DC to discuss the status of a variety of programs related to ECCS-LOCA research pro-grams. A notice of the meeting appeared in the Federal Register on August 11, 1978 (Attachment A). A copy of the detailed presentation schedule is attached (Attachment B). A list of attendees at the Subcommittee meeting is attached (AttachmentC). A list of documents provided to the Subcommittee for this meeting is attachment (Attachment D). There were no public statements either written or oral..The entire meeting was open to members of the public.
1.0 MEETING WITH NRC STAFF AND THEIR CONSULTANTS TO DISCUSS THE STATUS OF A VARIETY OF PROGRAMS RELATED TO ECCS-LOCA RESEARCH PROGRAMS (OPEN SESSION) 1.1 Subcommittee Chairman's Opening Remarks Dr. Isbin, Subcommittee Chairman, introduced the members of the Sub-committee and noted that the purpose of the meeting was to discuss the status of a variety of programs related to ECCS-LOCA research in order to prepare a section of a report on research to Congress.
He pointed out that the meeting was being conducted in accordance with the provisions of the Federal Advisory Committee Act and the Government in the Sunshine Act and that Dr. Andrew Bates was the Designated Federal Employee for the meeting.
He stated that no requests for oral statements nor written statements from members of the public had been received.
Dr. Isbin provided a detailed introductory statement.
He sum-marized some of the questions and conclusions reached at the August 15, 1978 ECCS Subcommittee meeting with the purpose of stimu-lating additional thought arid input by the participants.
Items mentioned were:
How well are we achieving the research goals?
Do the present licensing policies still remain prudent, cautious, and conservative?
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ECCS 8/28..'9
.0 / G Do we have major areas for concern in following the course of postulated accidents and the effectiveness of the miti-gating systems?
l What specific areas need examination in more detail?
Are the reactor safety research programs and other related industrial and international confirmatory research pro-grams adequate?
What, if any, additional programs might be needed?
Dr. Isbin noted that the British, at the Sixth International Heat Trans-fer Conference in Toronto, August 1978, again raised the question of the Hindle Tests concerning uniform heating and ballooning of the clad over long lengths, producing a potential for major blockage.
This is a topic of discussion on the third day of this meeting.
Dr. Isbin noted that numerous important ECCS LOCA research items are being discussed at meetings but it was his opinion that some formal commitment that proper attention to the suggestions should be carried out.
Dr. Isbin noted that a major function of this meeting is to bring the weight of the Subcommittee's technical judgments into assessing the NRC Staff's presentations and that the Subcommittee's comments are meant to challenge, stimulate, and to pursue why and how and under what bases the NRC Staff justifies its positions. Dr. Isbin recom-mended that the NRC Staff's minutes of meetings contain a postscript of what actions, if any, the NRC Staff plans as a consequence of the meeting.
Dr. Isbin made an observation,from the August 1978 Toronto meeting that several speakers were concerned of how best to accomplish the research goals, recognizing the special interests.
He questioned how sophisticated do we have to be? He suggested that leadership is needed in recognizing which course of action needs additional work and which can be terminated.
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ECCS 8/28, ) / ?! '78 Dr. Isbin discussed the matter of alternative ECC systems and questioned how the objectives could be defined for incorporating alternate ECC systems inte the confirmatory research programs. He suggested that we need more realistic definitions of what is expected from the 2-D and 3 D programs and a fallback position if these programs are not completed.
Dr. Isbin questioned how RSR was assuring itself that there is effective
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communication among the major research groups. He commented on the two-level code assessment, the second level which is an independent evaluation. Dr. Isbin referred to NUREG-0224 which deals with code assessments and suggested the questions raised in that Report might be important on how one approaches code assessments.
1.2 Summary of LOCA-ECCS Research - NRC Staff Dr. L. S. Tong, NRC Staff, provided a summary of the NRC Staff's LOCA-ECCS Rescarch Program. Tong discussed the experimental pro-grams to be presented by the contractors and NRC Staff during this session. Attachment E-1 provides a LOCA-ECC Experimental Program Matrix showing the areas each of the contractors are involved in ECC separate effects and integral research.
Dr. Tong discussed the LOCA-ECC research approach. The approach was to provide an understanding of the nature, threshhold, and sensitivity effects of important features of LOCA-ECC by testing simulated systems. Modeling is being used to develop a quanti-tative description of the important features and to determine scaling effects.
Code analysis is being used to predict the consequences of LOCA-ECC and to study sensitivity effects of important features by code calculations. Tong stated that vali-dation of predictions was necessary to determine the uncertainty of code predictions for application to LWRs by comparing the prediction with well-simulated experimental data. Additional uncertainties due to problems of simulating or scaling individual features can be bracketed by sensitivity studies. Dr. Tong requested ACRS advice with regard to determining permissible uncertainty bands.
,ECCS 8/28, 29 &30/78 Dr. Tong discussed the important separate effects features in the PWR LOCA-ECCS noting that a one-second delay in DNB will reduce the peak clad temperature by =125 F.
He noted that the model is presently allowed to use only a 0.1 second delay, but that ORNL Semiscale indi-cates a 0.5 to 2 second delay is appropriate. Tong noted that a prelimi-nary estimate for reflood heat transfer influence on peak clad temperature 2
is about 35 F per one BTU /hr.ft.F. Another item of importance was ECC bypass which affects PCT 28 to 50 F at realistic. conditions. Tong added that tw-phase flow pump degradation and steam generator heat trans-fer are important items. Another separate effects item of importance is the sensitivity of steam binding which, by preliminary estimate, is about 400 F PCT change / inch /sec injection rate.
It wa s Dr. Tong 's opinion that if the bottom flood injection rate is greater than 2.5 inches /sec that the heat transfer rate is high enough to reverse clad temperature increase.
1.3 Blowdown and Reflood Heat Transfer Summary - NRC Staff Mr. Edwin Davidson, NRC Staff, provided a summary of the PWR and BWR blow-down and heat transfer research programs and discussed the principal features of each. He noted that their primary objective is to study transient heat transfer during the depressurization phase following the assumed LOCA in PWR and BWR systems. He added that principal interests are variables influencing the following:
Time to CHF/BT Pre CHF/BT heat transfer Post CHF/BT heat transfer Rewet heat transfer Peak clad temperature Mr. Davidson noted that their principal interests in the FLECHT-SEASET pro-gram are tests and analyses on flow blockage effects, system feedback, and separate effects of steam generator and upper plenum.
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ECCS 8/28, 29 &30/78 In response to a question from Dr. Catton on how the differences in fuel pin material and method of heating was to be considered between the tests and actual reactor conditions with regard to clad thermal characteristics, oxidation, and gap conductance, Mr. Davidson said that a facility in Canada and one overseas would be utilized to help evaluate these differences.
Dr. Tong added that the Halden reactor will be used to compare heat trans-fer mechanisms with heater rods and nuclear fuel rods as well as stainless steel and zirc clad differences. Dr. Hedrick commented that they have analytical simulators whereby one can take a nuclear fuel pin, with all I
the models, and run a transient.
By back-solving the equations for electric pins you can program to get an electric pin calculation to produce' the same surface or coolant conditions as a nuclear pin. He noted that the results of such calculations show that a lot of experi-mental data is not applicable to nuclear systems.
In response to a question from Dr. Zaloudek concerning information received at a recent ACRS Subcommittee meeting where PBF data showed CHF occurred at 12% to 15% lower than W-3 correlation predictions, Dr. Tong stated that the PBF data did not consider cold wall effect. Tong also questioned the power measurements and the bowing situation of the two rods in question.
1.4 Summary of NRC/EPRI/GE Blowdown /ECC Program - General Electric Co.
Mr. G. Sozzi, General Electric Company, provided a summary of the NRC, EPRI.
and GE sponsored ECCS blowdown program (Attachments 16-30). Objectives of this program were to obtain information on BWR thermal hydraulic and bundle heat transfer response during a LOCA simulation, and to provide a data basis for evaluation of BWR LOCA models and assumptions. Their two-loop test apparatus provides integrated system affects tests on one full size bundle of 7 x 7 or 8 x 8 assemblies. Sozzi discussed updated results of blowdown heat transfer tests noting that due to slower blowdown and the larger inventory remaining that they do have substantially better heat transfer than previously reported.
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ECCS 8/28, 29 & 30/78 Program activities since previous meetings with ACRS included a 100% break test and the effects of the ECC System with an average power of 5.05 MW and a minimum spray flow rate of about 5 gpm. Conclusions from these tests were that ECCS is effective in reducing PCT and that no adverse effects are observed by having ECCS water injection and that the water appears to go where it is intended.
In response to a question from Dr. Isbin concerning the availability of a GE best estimate code, Sozzi noted that they have a proposal that flRC, EPRI, and GE perform the CCFL reflood program which includes a best estimate code using the TRAC Code.
1.5 Overview' PWR - Blowdown Heat Transfer Separate-Effects Program - Oak Ridge Nat'l Lab.
Mr. J. White, Oak Ridge flational Laboratory, discussed their work on the PWR Blowdown Heat Transfer Separate - Effects Program. The blowdown heat transfer objectives address areas of concern in hypothetical loss of coolant accidents (LOCA) are as follows:
Determine time to CHF Determine heat transfer coefficient as a function of local fluid conditions Develop computer models required to calculate surface temperatures, heat fluxes, and local fluid conditions Determine the effect of bundle geometry and operating conditions Mr. White discussed the schedule for the short-term and long-term BDHT separate effects program as shown on Attachments F-1 and F-2.
White stated that the BDHT program makercontributions in the areas of experi-mental data, ORNL interpretation, and analysis system.
(SeeAttachments F-3 through F-5).
In response to questioning from Dr. Plesset concerning the significance of a THTF Test with average delay time to CHF of 0.5 seconds or longer (see Attachment F-7) and how this could be used with regard to a reactor LOCA, Davidson commented that he could not conclude that the results of this test would occur in a reactor PWR LOCA.
(CCS 8/28, 29 & 30/78 Mr. White discussed the major accomplishments of their BDHT tests which are summarized on Attachment F-6.
1.6 Full-length Emergency Cooling Heat-Transfer Tests (FLECHT) and System E_ffects and Separate - Effects Test (SEASET) - Westinghouse Electric Corp.
Dr. L. Hochreiter, Westinghouse Corporation, discussed the NRC, EPRI, and Westinghouse Corporation sponsored FLECHT-SEASET research program.
A goal of the program is to enhance the understanding of the physics of reflood phenomena in PWRs. Other goals are to aid in the improve-ment or further development of thermal-hydraulic models and/or computer codes for the reflood phase and to aid in the validation of best esti-mated thermal-hydraulic models and/or computer codes for the reflood phase in PWRs, and aid in improving the understanding of safety margins associated with current licensing evaluation models and criteria.
They also hope to broaden the data base for PWR LOCA-ECCS safety evaluations to permit a coordinated reappraisal of existing licensing criteria and to assess selected alternate ECC injection configurations.
The FLECHT-SEASET Tasks include 17 x 17 unblocked FLECHT tests, 21-rod bundle tests, and 17 x 17 blocked bundle FLECHT tests to address the Appendix K steam cooling / flow blockage rule. Steam generator tests, FLECHT-SEASET Upper Plenum tests, and system effects tests will be performed to assess the safety margin, provide data / analysis for improved reflood system code, TRAC RELAP 4 MOD-6 assessments.
Hochreiter noted that the data analysis plans are to develop a FLECHT type empirical correlation for both 15 x 15 and 17 x 17 data to cal-culate heat transfer and quench times They will reduce the unblocked data to compare with blocked data. They will also perform mass and energy balances on the test, bundle to obtain void distribution, actual and equilibrium quality, and local mass flow.
He noted that they will use high speed movies and still photographs to estimate flow regimes, droplet size, velocity, and distribution.
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ECCS 8/28, 29 & 30/78 Dr. Hochreiter summarized with the following closing remarks:
FLECHT-SEASET will extend the reflood data base to new PWR rod geometries.
The flow blockage and steam cooling Appendix K rule will be addressed. A wide range of inputs are being factored into these tests.
Improved instrumentation and data analysis will assist mechanistic reflood code model development, and assess-ment of advanced reflood codes.
Reflood system effects tests are planned using a building block approach such that component behavior is understood and reflood system response may be analyzed.
Alternate ECCS performance during reflood will be examined.
In response to a question from Dr. Isbin concerning ballooning effects,
Hochreiter stated that they would not have uniform heating in PWR fuel rods as assumed by Hindle; therefore, the ballooning effects shown by Hindle are not applicable.
Dr. Isbin noted that the Springfield metal-lurgists are not convinced of the NRC arguments. This item will be discussed on the last day of this meeting.
In response to questions from Dr. Shumway concerning what droplet size will be used to simulate a PWR, Hockreiter said they would test over a range to see what the effect is on the heat surface.
1.7 NRC Inhouse Reflood Study - NRC Staff Mr. Loren Thompson, NRC Staff, briefly discussed the NRC Inhouse Bottom Reflood Study. The objectives of the study are to detennine cladding temperature behavior using the computer code REFLUX and to develop a correlation for quench front velocity. Attachments G-1 and G-2 provide FLECHT data and REFLUX predictions for cladding temperature versus time and cladding temperature versus inlet velocity.
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ECCS 8/28, 29 & 30/78 Mr. Thompsnr' provided a description of a FWR reflood research information letter which is presently in the approval chain at this time.
Items to be covered by that letter are shown on Attachment G-3.
l 1.8 Instrumentation i v.: Model Development program
- 1. 8.1 NRC Staff Summary Dr. Yin-Yun Hsu, NRC Staff, provided a summary of the instru-mentation and model development program. Attachments H-1 through H-7 are a list of papers written by contractors and outside publications on blowdown and reflood heat transfer, interfacial mass transport, and instrumentation development.
Dr. Hsu stated that he or Mr. Thompson would assist anyone that requested copies of any of these references. Attachments H-8 and H-9 provide a summary list of the contactors and their area of research in the NRC instrumentation and model develop-ment program.
In response to a question from Dr.
Ishir concering how one can asst're that we are effectively using the benefits from these programs which can be derived, Dr. Tong stated that he would personally see that this information is used in the code analysis.
1.8.2 Research on Thermal-Fluids Problems of LWR Safety - Lehigh University _
Dr. J. Chen, Lehigh University, discussed the two-phase instru-mentation and the post-CHF heat transfer model development pro-grams being performed at Lehigh University. The post-CHF heat transfer model development program consists of phenomenological correlation, experimental heat transfer measurements, and non-equilibrium superheat measurements. Chen described the details of their two-phase instrumentation program consisting of vapor superheat probe, liquid film thickness probe, and liquid film velocity probe development. Chen noted that the problem with two-phase probe development was that quenched liquid droplets provide
RGG5 8/28, 29 & 30/78 false vapor tempt.rature readings. He discussed a method using tiot air and water to verify that probes work in a two-phase condition.
1.8.3 Use of Pulsed Neutron Activation (PNA) Techniques in Reactor Safety Research - Argonne National Lab.
Dr. P. Kehler, Argonne National Laboratory, provided details of their work on the use of a pulsed neutron activation (PNA) technique to measure two phase flow. The principal of the PNA measurement is simply to use a small portable neutron source to activate the oxygen in the fluid with known characteristics and then to measure the N-16 activity distribution by b detector located at a downstream location te determine the true mass flow.
PNA uses a data reduction technique of the counts versus time dis-tribution. Attachments I-1 and I-2 provide some of the PNA nomen-clature and equations used in the technique.
Kehler stated that the accuracy of the velocity readings were within 2-5%
of turbine readings and the accuracy of the density readings were within 0.05 grams /cm3 (poor density readings due to presently avail-able equipment). The objective of this proaram is to derive calibration techniques for two-phase flow instrumentation and to measure flow velocities and distribution at the upper plenum (3-D) reflood test.
Kehler noted that although the feasibility of tne PNA technique has been demonstrated, additional develop-ment is necessary to fully utilize the benefits.
In response to a question from Dr. Isbin, Kehler noted that the measurements are an average over the cross section.
Kehler con-sidered this an advantage over other techniques.
It was also noted that in some ccises the technique will separately measure both the liquid and gas flow velocities simultaneously.
ECCS 1
8/28, 29 & 30/78 1.8.4 Experimental and if q SAj:es in 3as Liquid Flow -
University of Hous Dr. A. Dukler, University of Houston, discussed the NRC-supported and non-NRC-supported instrumentation and modeling development programs being performed. Dr Dukler discussed an overview of the University of Houston program in experimental and ulodeling studies in liquid flow since 1974 (See Attachments J-1 and J-2).
Dr. Dukler discussed the approach and techniques used in modeling flow pattern transitions in horizontal tubes under transient flow conditions and flow reversal. The approach includes executing experiments designed to reveal underlying mechanisms and to set up physical models as simple as possible, but sufficiently detailed, to include the essential features of the process.
Dukler discussed transient flow pattern transition theory noting that they have attempted to diagnose what's happening to get some insight into the physics.
Dukler describec a theory of flow transition from one phase to another (See Attachments J-3 through J-5).
He discussed the dependence of the transitior and slug formation on the rate of the transition and indicated a code could be developed to pre-dict these transitions.
1.8.5 Two-Phase Flow Phenomena in Nuclear Reactor Technolooy -
Renssalaer Polytechnic Institute Dr. R. l.ahey, Renssalaer Polytechnic Institute, discussed the instrumentation and model development program being perfont'ed at their facility. This program consists of three main tssks:
Two-phase flow instrumentation Phase separation and distribution phenomena BWR ECC parallel channel effects (PCE)
The two-phase flow instrumentation program at RPl includes the development of several instruments which include:
an infrared device for the measurement of transient steam / air fraction, a digital interferometer for the measurement of global void fraction
t ECOS 8/28, 29 & 30/78 distribution, a radio frequency
..a1 impedence probe, a measurement device for two-phase flow rates using pulsed neutron activition techniques, a high teairierature optical probe for the measurement of local void fraction, and a side-scatter gamma ray system.
Highlights of the RpI phase separation and distribution phenomena are provided on Attachment K-1.
Dr. Lahey noted that with regard to their BWR ECC parallel channel ef fects progran.that they have completed a freon / water sca71ng basis document and have a loop under construction.
1.8.6 A Study of Droplet riydrodynamics Iriportant In Disperse Flow Regime of LOCA Reflood By Laser Doppler Anemometry - SUNY-Stonybrook Dr. R. '.ee, SUNY-Stonybrook, discussed a study of droplet hydro-dyream.cs important in di:perse flow regime of LOCA reflood by laser doppler anemometry being performed at their facility. He noted that the objectives of this program were to experimentally in-vestigate droplet hydrodynamics under those conditions simulating the steam water droplet flow during the reflood of a hypothesized LOCA in both the upward flow in the core and in the cross flow in iha mpec plenum. The purpose of this study is to provide funda-information of droplet hydrodyriamics needed for improved nai understanding of heat and mass transfers in the LOCA reflood.
i The specific tasks were to develop a Laser Doppler Anemomentry (LDA) system for makir.g local measurements in particulate - water tystem, water droplet - air system, and water droplet - vapor system.
Dr. Lee noted that their observations have shown that annular two-phase disperse flow is too complicated to be describable by I
oversimplified pseudo one-phase and/or one-dimensional flow models. Another observation was that small droplets ( 60 microns) are found to play a prominant role with heat and mass transfers resulting from Interactions with the wall.
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ECCS 8/28, is % 30/78 Attachment L-1 shows the optical arrangement of the laser assembly. Attachment L-2 shows a sketch of findings showing the droplet entrainment from the liquid film on the wall.
1.8.7
_ Heat Transfer Coordination for LOCA Programs - Argonne National Lab.
Dr. Paul Lottes, Argonne National Laboratory, discussed the ANL heat transfer coordination activities for LOCA programs. He de-scribed their task as providing coordination and consulting ser-vices. He also participated in the review of other NRC-r.ponsored programs and maintained a local condition heat transfer bank.
In response to Dr. Isbin's opening remarks concerning identiffing areas of work that may need more emphasis, Lottes stated that it was his opinion that in the area of fuel element heat transfer enough work has been done to get heat transfer coefficients to a sufficient accuracy to get maximum fuel tenperatures.
Lottes said that in most areas he has examined, the limiting heat trans-fer is in the fuel element itself (fuel properties and gap conductance).
Dr. Lottes discussed an independent assessment he performed on the effect of gap conductance on heat transfer. Attachments M-1 through M-3 show the results of this assessment. This assessment shows that the accuracy of the heat tranfer coeffi-cient during quenching does not have a significant.effect.
Lottes showed that the heat release rate of the fuel element itself is the significant factor. Relating this to fuel element behavior in a reactor. Lottes stated that beyond one second the temperature curves are nearly parallel regardless of the quench time.
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ECCS V28, 9 & 3L 78 Dr. Lottes discussed a calculation he performed comparing initial energy stored in the fuel element at the time of a LOCA with that produced by decay heat following the IOCA.
He found the amount of energy almost comparabie, at lea st in the early time, such that most energy comes from the initial stored heat in the fuel and very little from the shutdown heat. Lottes noted that as the gap begins to open up, the cladding begins to cool, such that when specifying the coefficient you can have a fairly large error but still reach the right peak clad temperature due to the low slope of the curves (See Attachment M-4).
Dr. Lottes discussed a comparison of the behavior of nuclear heated rods and electrically heated rods, indicating that they behave differently.
Dr. Lottes is continuing his re-search in this area.
1.8.8 Instrumentation Development Programs at LASL, ORNL, and Sandia - NRC Staff Dr. Yin Yun Hsu, NRC Staff, hurriedly discussed the numerous instrumentation development programs at LASL, ORNL, BNL, and Sandia.
A program matrix showing the development programs at each of these organizations, is shown on Attachment N-1.
Dr. Hsu noted that EGAG has a Technical Assistance Program for the cooperative German, Japanese, and United' States Steam Binding Study.
Hsu discussed the LASL development program using the STOP,Z lens for upper plenum de-entrainment tests in multirod facility, downcomer tests for code developer needs, and multi-national tests with the Japanese and Germans. ORNL is involved in the Advanced Instrumentation for Reflood Studies (AIRS) involving several aspects of measurement.
ECCS 8/28, 29 & 30/78 Sandia is developing a neutron generator for two-phase flow 10 calibration with a design goal of 10 neutrons / pulse with a 1 msec. pulse length.
1.8.9 Model Development Programs at Northwestern University, Massachusetts Institute of Technology (MIT), and Wichita State University - NRC Staff Dr. Yin Yun Hsu discussed the model development programs at North-western University, MIT, and Wichita State University. A model development matrix is provided on Attachment 0-1 showing ecch organizations' area of model development.
Northwestern University is working on a model for the impingement of a plane steam jet on a flowing water layer and will measure horizontal and vertical flow and local condensation rates. MIT is involved with the develop-ment of liquid carryover and reficod. heat trensfer correlations which involve thermal hydraulics during reficoc and natural circu-lation in large cores during reflood.
Wichita State University is developing a modal for axial photography for measurement of droplet size and distribution through grids.
1.8.10 Thermal Hydraulic Development Program Facilities, Instrumentation Control Systems, and Data Acquisition - Brookhaven National Laboratory Dr. Yin Yun Hsu discussed the BNL thennal hydraulic development pro-gram which will directly measure the volumetric vapor generation rates in flushing steam-water flows (both steady-state and transient conditions).
Hsu provided details of their test facility which consists of water loop, converging-diverging nozzles, and instru-mentation. Problems have been encountered in interpreting local probe output. These problems consist of electronic response for fast moving small bubbles, hydrodynamic response (wetting of the tip and interface deformation), and bubble dynamics, (bubbles escaping the probe at low velocities).
1.9, Concluding Statements Dr. Isbin thanked the NRC Staff and their contractors for their participation.
The day's meeting was adjourned at 6:55 p.m.
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.EC.CS 8/28, 29 & 30/78 l
2.0 AUGUST 29, 1978 MEETING - CONTINUATION FROM AUGUST 28, 1978 2.1 Chairman's Opening Remarks Dr. H. Isbin, Subcommittee Chairman, opened the meeting by introducing the ACR3 members and consultants. Dr. Andrew Bates was the Designated Federal Employee. There were no written or oral statements from members of the public.
Dr. Isbin commented on the August 28, 1978 meeting noting that in his opinion the talent involved in the instrumentation and model development prngrams appears to be outstandi.ng, and the progress reported in the various programs was impressive. Dr. Isbin discussed several comments andconciusionsmadeatpreviousmeetingsindicatingthatadditional discussion is needed in several areas of the programs. It was Dr. Isbin's opinion that further amplification of the cost effectiveness, desired accuracy, and how the accuracy will be achieved for the advanced best-('
estimate code, are needed.
He elso noted that further discussion and cicrification are needed on how the developer and independent reviever, g
examine and judge the bases for applying codes to a light water reactor.
,,,0 2.2 Code Develoament Programs (NOTE:
Dr. Zaloudek did not participate in the discussion of the COBRA Code Development Program.)
2.2.1 Summary of Selected Code Development Programs - NRC Staff Dr. S. Fabic, NRC Staff, provided a summary of the Code Develop-ment Program. He noted that there is a trend toward winding down work in the code development area and winding up work in the code assessment area (See Attachment P-1).
Dr. Fabic discussed the guidelines the NRC Staff uses for select-ing a code, including, items such as current capabilities, adaptability to future NRC needs, and projections of developmental effort.
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ECCS In response to Dr. Isbin's opening remarks, Fabic suggested that they are taking the majority of the tests performed to date and comparing them against each other as well as with the calculated results of their best-estimate codes.
The Subcommittee discussed the advantages and disadvantages of selecting one fast running best-estimate code.
Fabic noted that they hava asked the three laboratories to calcu-late Standard Problems Nos. 6 and 8, small break and Semiscale integral LOCA, which should sufficiently exercise the codes.
Evaluation of these performances will be considered in the selec-tion of the code to be used.
It was noted that RSR has made a commitment to the Commissioners to make a selection among the three fast running best-estimate codes. The Subcommittee, at the August 14 and 15,1978 meeting in Idaho falls suggested that the ACRS be kept informed.
Dr. Isbin announced that Dr. Theofanous would be in conflict of interest on any dis-cussion concerning the choice of these three codes.
2.2.2 COBRA-1F Computer Code Development - Pacific Northwest Lab.
Mr. M. Thurgood, Pacific Northwest Laboratory, discussed the COBRA-TF Code development program. Attachments Q-1 and Q-2 provide the history of the development and a description of the code. The objective of the code is to model a LOCA transient in a PWR reactor vessel equipped with upper head injection.
Mr. Thurgood discussed the detailed results for UHI by carrying COBRA-TF to 17 seconds.
He also showed a 5-minute movie showing flow movement, including velocf ty and direction, during a blowdown transient.
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ECCS 8/28, 29 & 30/76 Thurgood summarized noting that they were encouraged by their data comparisons. A full UHI simulation is presently underway and significant progress is being made in modeling the UHI transient.
Dr. Plesset commented on Mr. Thurgood's presentation noting that the Subcommittee was interested in reactor safety, not movies or numerics. Dr. Plesset noted that he was disturbed that Mr. Thurgood could not comment on the safety significance of the UHI process. Dr. Is, bin commented that he appreciated W. Thurgood's COBRA illustration of three-dimensional effects in the core.
2.2.3 THOR Code - Brookhaven National Lab.
Dr. W. Wulff, Brookhaven National Laboratory, briefly discussed the TH0R Code development program. The THOR Code is a best-estimate code for developmental verification and a fast licensing code for.msitivity calculations that deals with full-scale power plant transients and experimental test facilities with emphasis on LOCA transients. An objective of the THOR Code is to reduce conservatism and uncertainties with regard to flashing, mixing, phase separation, and flooding.
Attachments R-1 and R-2 provide specific features of the TH0R Code.
Dr. Wulff discussed some calculations obtained using the TH0R Code which show that the computing times are shorter than g
obtained with other methods. He noted that this code is a departure from normal programming and modeling techniques and required considerable work. The first results are now
'/. being produced.
In response to a question from Dr. Shumway concerning a comparison of TH0R running time with other codes, Wulff stated that TH0R was about a factor of four to five faster than RELAP 4 MOD 6 calculations.
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ECCS 8/28, 29 & 30/78 2.2.4 lJpper Hvad Injection Modeling - Sandia Laboratory dr. Berman, Sandia Laboratory, discussed the results of their modi-fied RELAP 4 noting that the new version discovered that with single nodalization they were onable to fill the lower plenum due to sensitivity of water p.k assumption. After changing the nodaliza-tiori, including having two dowmcomers, filling of the lower plenum occurs at about 50 seconds.
In response to a question from Dr. Theofanous concerning the impres-sion that by " fiddling around with nodalization you can produce almost any results you want," Berman indicated that looking back, the mistake was obvious but it has totally obscured at the time.
Dr. Berman summarized the results of the UHI calculations using the RELAP 4 MOD 6 indicating that the modified code, which they know contains errors and instabilities, is able to produce large amounts of quenching in a UHI calculation. Very little quenching is observed with other codes.
In addition, the RELAP 4 MOD 6 takes e. bout half as much running time as RELAP 4 MOD 5.
2.2.5 Statistical Analysis - Sandia Laboratory Dr. M. Berman, Sandia Laboratory, discussed their statistical LOCA study for determining the peak clad temperature probability distribution calculated by RELAP.
The application of this study is intended to quantify the conservatism of the requirements of 10 CFR 50, Appendix K, and to determine peak clad temperature sensitivity to input parameters and their variations.
It is also intended to apply this methodology to other similar prob-lems where sample sizes are small.
.ECCS 8/28, 29 & 30/78 Dr. Berman noted that they have investigated numerous methods l
and htve come upon a good fast one, called response surface I
methat where they replace a complicated function or code with a simule analytic function.
Berman discussed a proposed test of thn new method to estimate peak clad temperature probability distribution for blowdown and reflood for the Zion PWR plant using RELAP 4 MOD 6 (See Attachments T-1 through T-3) having twenty-one variables for the initial blowdown part of the study, of which a few will dominate.
Berman discussed numerous prob-lems they have had with a test of this method against a simulated RELAP function. Berman noted that the major problem faced in this study so far has been the determination of how to initialize the calculation. The Subcommittee members requested several items of clarification but did not find any significant problems with the proposed study.
2.2.6 Water Reactor Analysis Program (WRAP) - Evaluation Model - Savannah River Lab.
Mr. M. Gregory, Savannah River Laboratory (SRL), discussed the topic of how to create a system which automates the LOCA calcu-lational procedure and how to create a system convenient to use.
Mr. Gregory discussed the development of WRAP which is a modular computational system under development by SRL since 1976 for analysis of loss-of-coolant accidents and other transients (See Attachments U-l through U-3).
The development of WRAP was categorized by three phases of work.
The objective of the initial phase was to develop WRAP as a user-oriented version of the RELAP 4 computer code for routine production use. This. phase was completed during 1977. The objective of the second and third phases is the automation of all computational aspects of the LOCA anriyses for BWRs and PWRs, respectively, by implementing other computer codes into WRAP.
The second phase is nearly complete and the third phase has begun. As the second and third phases are completed, WRAP
____.____._______m_____-_-
ECCS 8/28, 29 & 30/78 will be used by SRL and NRC personnel in the evaluation of potential changes in the present licensing requirements.
Mr. Gregory discussed the advantages of " WRAP modularity" being that modules (computer codes) can be developed independently, interchanged, and data transfer between different computational steps facilitated.
Mr. Gregory informed the Subcommittee that the automated EM LOCA analysis for BWRs is operational at SRL and is being evaluated.
Preliminary evaluation indicates that both the RELAP fuel model and top spray injection model may require modification.
The objective of the third phase of WRAPS development is the automa-tion of all computational aspects of EM analysis for PWRs.
In addition, the WRAP blowdown and reflood models will be modified to model the LOFT facility and pretest conditions of LOFT nuclear tests will be performed with WRAP.
Dr. Plesset made several remarks concerning codes. He indicated that he was very negative about sensitivity studies.
He did not think you could use a computer to find out the laws of nature.
Dr. Plesset stressed the need n r proper physics and engineering input into codes if they are to have any real value in reassuring us of reactor safety in accident conditions. It was his opinion that some codes have wrong physics since some codes can't describe the steady state of a reactor. Dr. Plesset was concerned about turning codes with improper physics over to people who are not aware of the difficulties with the code. Ur. Fabic indicated that he shared Dr. Plesset's concerns about plunging ahead with complicated calculations when the physics are not reasonably represented. He indicated that they have overcome the first hurdle of demonstrating that the codes can calculate.
The next hurdle is to improve the physics and make it right.
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8/28, 29 & 30/78
~
2.3 BWR Containment Modelit.s_.id Antj ysis (NOTE:
Dr. Catton did not participate in this portion of the meeting).
2.3.1 Summary - NRC Staff Mr. R. Cudlin, NRC Staff, provided an overview of the BWR containment modeling and analysis program. Cudlin briefly discussed numerous basic experiments (Attachment V-1), and integral tests (V-2),
perfonned or to be performed. Cudlin discussed the steam venting and air venting special effects tests noting that there is good coverage of the spectrum, both with respect to the size of the vents and scale.
Mr. Cudlin noted several areas of interest based on feedback from NRC Licensing people. These included the following:
Investigation of Subcompartment Flows Containment Heat Transfer Coefficients Local Hydrogen Concentrations 2.3.2 Studies at Massachusetts Institute of Technology (MIT)
Dr. A. Sonin, MIT, discussed their program on the scaling laws for modeling containment system type phenomena. He noted that scaling phenomena are quite complex and currently much of the prediction depends on small-scale testing. Sonin discussed the following four aspects of their program:
1.
Fundamental basis for small-scale modeling of pool swell phenomena. This deals with the initial stage of a LOCA when air is ejected before the steam.
2.
Fluid-structure interaction.
3.
Steam condensation chugging.
4.
Distorted-geometry testing. This involves the common practice of using full-scale lengths for testing but reducing the areas and assuming full-scale loads on the system.
Details of the four aspects described above are discussed on Attachments W,1 through W-4.
ECCS 8/28, 29 & 30/78 Dr. Sonin noted that the pew: ace of minute quantities of air bubbles on the surfaces o-boundaries of the pool can distort 2
the behavior significantly. He noted that excessive vapor pres-sure in the air space of the wet well, when scaled down to 1/20, causes changes in sy:, tem behavior because of condensation during compression. Vent volumes and placement of the orificing also influence the scaling laws.
Breakthrough does not necessarily scale with the scaling laws due tc arface tension and instability.
The Subcommittee members had a lengthy discussion on the chugging process in the vent pipe.
It was the opinion of Drs. Plesset and Theofanous that the bubble separates from the vent.
It was the opinion of Dr. Sonin that the bubble did not separate from the vent prior to collapse but is sucked back into the vent.
Dr. Theofanous suggested that if they used high speed photography (20,000 frames /sec), they would see the bubble separate during the chugging process.
In response to a question from Dr. Isbin concerning the impact of the MIT and UCLA tests on the ongoing programs for the operating reactors and the Mark II and Mark III tests, Fabic indicated that they haven't found anything which has an impact on their ongoing containment programs, except they are glad they didn't prematurely go into a 1/5 scale steam-water test at Livermore.
Fabic indicated that they would not be able to make use of the 1/5-scale steam-water data since they do not have appropriate scaling laws.
2.3.3 Studies at the University of California at Los Angeles (UCLA)
Dr. Y. K. Dhir, UCLA,' described the work they have done at UCLA to understand some of the phenomena that exist during transient injection of air in a pool of water or injection of steam into subcooled water.
In air venting they have looked closely at
ECCS 1, 29 & 30/78 vent clearing phenomena and have discovered the role p: ;W by Taylor instability (Attachment X-1) as well as ti, virtual mass of the liquid being pushed into the pool as the vent cl ea rs. They have also looked at the hydrodynamic forces at thb bottom of the test vessel (Attachment X-2) both single and double vents. Dhir noted that they have started some preliminary steam venting experiments and have isolated the various phenomena to look into before proceeding to modeling.
Attachment X-3 shows typical bubble movement results obtained from their experiments using a 2" diameter single tube located in the middle of a cylindrical vessel. This attachment shows the flow rate and pressure histories using air.
Dr. Dhir discussed the results of several steam chugging tests (See Attachments X-4 and X-5).
Dr. Dhir's explanation of the chug-ging phenomena was similar to that discussed by the previous speaker. Movies of the chugging phenomena were taken at 3000 frames /sec which were not fast enough to adequately follow bubble collapse. Dhir indicated that they were trying to get a camera which will go to 10,000 frames /sec.
2.3.4 Review of the 1/5-Scale Mark I BWR Pressure Suppression Program -
Lawrence Livermore Laboratory Dr. E. McCaully, LLL, discussed their 1/5-scale test facility experiments. The objective of the LLL effort is to design and construct a large scale replica of the Peach Bottom Mark I BWR, to develop related scaling laws, and to conduct a detailed program of air tests followed by steam tests for a postulated LOCA (See Attachment Y-1). The facility has been completed for the air clearing tests in the Peach Bottom Geometry. They have completed a series of comprehensive air tests which have been f reported. They have concluded from these tests that they have defined, identified, and quantified the major force sensitivi-ties resulting from the initial drywell pressurization rate, downcomer submergence, initial drywell overpressure, and the
ECCS 8/28,129 & 3:,
vent system loss coefficient. Th re have also boen able to quantify the three-dimensional force distributions. They have not identified any new loads.
In response to a question from Dr. Isbin concerning s.,a-water testing at the facility, Dr. Fabic stated that they do not have any firm plans to continue with steam-water tests at that facility. He added that they have had two professors look into this question and have gained the message that it would not be cost-effective to continue with steam-water tests in
,.N^
this 1/5-scale facility.
2.3.5 Containment Analysis Development - BWR Pool Dynamics - PELE-IC Dr. W. McMaster, LLL, discussed their program to provide an analy-tool for the containment analysis of BWR pool dynamics. The task for FY 1978 was to develop a Eulerian fluid code, a finite element structures code, couple the two codes, and to verify their calculations.
Dr. McMaster described PELE-IC which is a 2-D Eulerian hydrodyna '.-
code with an incompressible fluid. McMaster stated that the basi:
solution strategy is to solve the Navier-Stokes equations for a set of trial velocities, and then iterate on pressure fluid until the divergence satisfies some minimum criteria until they obtain a new pressure and velocity field (See Attachment Z-1).
They have verified this code by looking at a series of classical solutions with known analytical solutions and to the experiments at MIT and UCLA. They have concluded that the fluid dynamic algorithm and surface motion treatment are being handled properly and that the coupling ~ algorithm between the fluid and structure is working right. McMaster noted that they qualitatively agree with an early experiment at MIT and with some UCLA experiments.
I me aeen
ECCS 8/28, 29 & 30/78
. Dr. McMaster described the work they plan to do in FY 79 which includes extending their structure-fluid interface coupling to more general shapes so they can look at the toroidal shell typical of the Mark I containment. They also plan to continue investigation of chugging models and its effects on structural response.
C6 Dr. McMaster summarized stating that with the work they have done, they feel they have a code that accurately couples the fluid and the structure. He felt they have a working algorithm that they can extend to more general structures. He added that simple chugging models look promising and they feel they can cciculate 2-D type problems and effects.
2.4 Concluding Statements Dr. Isbin noted that the ACRS members and consultants had been encouraged during the course of the meeting to give the benefit of their judgments and that they would perhaps determine the net results of these comments at a later meeting.
Dr. Isbin thanked the RSR Staff and their consultants for their presentations.
The meeting was adjourned at 4:20 p.m.
9
3.0 AUGUST 30, 1978 MEETING - CONTINUATION FROM AUGUST 28 AND 29,1978 l
3'.1^ Chairman's Opening Remarks l
Dr. Isbin opened the meeting noting that this is' the third day of the ACRS ECCS Subcommittee meeting. Dr. Andrew Bates is the Designated Federal 4e Employee.
He noted that no requests for oral or written statements had been received from members of the public.
In response to a question from Dr. Isbin concerning whether any new needs for safety research have been identified by UR, Dr. D. Ross noted that they have issued a " user's need" on safety valve critical flow and that they are currently working on several user's needs letters on such things as the 2D-3D programs, fuels related i.tems, and reactor phys.ics.
3.2 Status of NRR Programs - NRC Staff Dr. D. Ross, NRC Staff, provided an update of the ECCS work in NRR for the last year. He noted that the vendors are still requesting changes to Appendix V, and that they had just received a number of GE Topical Re-l ports proposing changes to thei.r model. The Commisston has authorized the Staff to proceed with the changes recommended in SECY-78-26 (Proposed Action Plan for Modifying the Emergency Core Cooling System
(.ECCS) Rule in 10 CFR 50.46 and Appendix K to 10 CFR Part 50, dated January 18, 1978) for which the Federal Register Notice will be out in about a month providing for a 60-day public comment period.
Dr. Ross told the Subcommittee that there is little or no emphasis by vendors on improved ECC Systems and little is being done by vendors on
~
a best-estimate code. He noted that Combustion Engineering is still doing some work on a best-estimate code and would probably provide an update at their annual research meeting with the NRC Staff.
Dr. Ross noted that industry needs a sign from the Commission of the intent to allow use of best-estimate licensing.
Industry does not want to develop best-estimate codes unless the Commission will let them be used. Ross noted that reliability and weakness of the existing ECCS is an area worthy of study. He noted that no dramatic fuel design changes are being proposed by the vendors but they are working on more realistic exotic advanced fuel design codes to account for high burnup, fission gas release, and to gain advantage from improved modeling.
l
ECCS 28 -
8/28, 29 & 30/78 Dr. Ross noted that their proposed Appendix K proposes the use of a con-volution of uncertainties in the nuclear heat source, such as decay heat and peaking factor uncertainties.
He also discussed the status of several Generic Task Action Plans shown on Attachment AA-1.
In response to a question from Dr. Isbin concerning interpretation of the Semiscale Steam Generator Tube Break Simulations, Ross noted that there is competition during reflood between the vapor coming up through the core and the vapor coming down through a leaky tube which tends to interact in a non-conservative way. He added that prior to the Semiscale test it was thought that as you parametrically increase the number of leaky tubes you will get some cooling from the steam flowing downward through the core.
The test showed this was the case.
In response to a request from Dr. Isbin for an update on two-loop Westing-house plants, Ross noted that Westinghouse or the Owners Group is preparing a report which is due September 15, 1978.
Ross felt the NRC would need four to six months to review the study and would probably naed some outside assistance. He estimated the first of the year for a v ritten NRC evaluation of the report.
Numerous miscellaneous items were briefly discussed including the following:
B&W Small Breaks Standard Problems Exxon Reload Core Evaluation for New Models Upper Plenum Injection and the Japanese 2000 Rod Core Tests Effect of Nitrogen Injection on Reflood UHI Modeling and Difficulties With RELAP Adaptation i
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'ECCS 8/28, 29 & 30/78 3.3 Reactor Systems Branch Presentations 3.3.1 ECCS Sumps and Vortex Formation - NRC Staff Mr. J. Watt, NRC Staff, discussed the status of the numerous ECCS recirculation pump tests, both plant and scale tests, on several operating plants (See Attachment BB-1). He noted the success of using grids to provide vortex suppression.
In response to a question from Mr. Etherington concerning whether most tests were an empirical correction or a real study of the vortex problem, Watt indicated that it is the attitude of the laboratories that they can't really predict, with accuracy, the vortex nor its intensity.
If they detect the presence of a vortex in a small-scale model test they will examine the situation further. Novak added that in most cases the utilities are simply running a proof test, due to NRC Staff urging, to demonstrate that their design is acceptable.
Mr. Watt summarized by noting that the test programs have found designs with vortex, air entrainment, and NPSH problems. Scale tests have been useful for identifying tendencies, development work, and demonstration tests.
In-plant tests are best from a preoperational test standpoint which prove the "as built" system works.
In-plant tests do not provide a rigorous evaluation of vortex or air entrainment tendencies.
3.3.2 Failure Modes and Effects Analysis (FMEA) Studies - NRC Sta/f Mr. W. Mills, NRC Staff, noted that the objective of the FMEAs is to provide the NRC with a detailed, systematic, and independent analysis of ECC Systems, RHR systems, and anticipated operational occurrences and to highlight areas for system improvement. Attachment CC-1 shows the approach being taken by EG&G to determine failure modes and liklihood. Attachments CC-2 and CC-3 provide the status of FMEA studies already performed and the schedule for future studies.
ECCs. 8/28, 29 & 30/78
' b.3 ECCS Reliability Studies - NRC Staff Mr. Mills, NRC Staff, discussed the ECCS reliability studies of four plants to develop a quantitative basis for assessing component system reliability. This study is scheduled for FY 79 and FY 80.
No contractor has been assigned at this time. The Subcommittee questioned the usefulness of the " relative" ECCS reliabilities of the four vendors plants.
In response to a question from Dr. Isbin concerning extending l
single failure to passive systems in contrast with the WASH-1400 approach, Novak noted that the NRC Staff does "not quite" apply a single failure criterion consistently in all disciplines since there is no absolute reliability numbers, such that there is some judgment required. He noted that they do not wish to grasp at a new criteria with regard to passive failures but are trying to prepare a data base as to when they should apply the criteria.
3.3.4 Intersystem LOCA - NRC Staff Mr. M. Rubin, NRC Staff, discussed the results of a probabilistic analysis that is currently being conducted to look at a vent path that was identiff i in the Reactor Safety Study as a relatively large contributor M 2 release category) to the risk resulting in a core melt. The "Intersystem LOCA" is discussed in WASH-1400 for Surry and is a non-design basis accident involving rupture of the pressure boundary valves that isolate the primary system from the decay heat removal ejection side. The intersystem LOCA event postulates the failure of two check valves which would subject the low pressure RHR system to full reactor pressure.
This results in a rupture of the RHR outside containment such that there is no coolant available for recirculation from the sumps. Attachments DD-1 and 00-2 show the probability of this event, with and without periodic leak testing. This study is currently utider review by the NRC Staff for possible consideration by the Regulatory Requirements Review Committee.
ECCS 8/28, 29 & 30/78 3.3.5 North Anna Pumps Status - NRC Staff Mr. J. 01shinski, NRC Staff, provided an update of the pump reliability testing conducted at North Anna Nuclear Plant.
Le testing has been conducted and the NRC Staff is reviewing the results. The NRC Staff had questioned reliability of the pumps and certain design changes have been made as a result of the NRC Staff's questions. The NRC Staff noted that the pumps had exhibited a large amount of bearing wear after a relatively short period of time.
The bearing design and material has been changed and the method of alignment has been changed and wear has nearly been eliminated (See Attachments EE-1 and EE-2).
3.4 Analysis Branch 3.4.1
_ECCS Codes - QA Audits - NRC Staff Mr. W. Hodges, NRC Staff, discussed their Safety Analysis Code quality assurance control review and audit requirement. He noted that this requirement was initiated at the request of Commissioner Gilinsky after the Westinghouse Zirc-Water reaction error was discovered. This audit will include all the safety codes. This audit consists of meeting with each reactor vendor to review QA procedures related to code development. The Analysis Branch will accompany I&E Inspectors on the audits. The I&E Inspectors will be provided the findings for use in exit interviews with the vendors. A report will be prepared after all audits are complete.
In response to questions from the Subcommittee concerning the NRC Staff responsibility that a particular plant is being licensed correctly, Novak noted that it is the responsibility of the Appli-cant to see that it meets NRC criteria. The NRC Staff assures itself in a reasonable way that the Apnlicant meets these criteria by perfonning audits and some ir pendent analyses with independent codes.
~ 8/28, 29 & 30/78
, 3.4.2 Rod Bow - NRC Staff Mr'. W. Hodges, NRC Staff, discussed the history of the rod bow problem.
He noted that in June 1977 the NRC Staff discontinued the review of all Topical Reports on rod bow due to inadequacies in the various model s. The NRC Staff has requested each vendor to revise their Topical Reports based on an outline and interim statistical method suggested by the NRC Staff. The NRC Staff has now received some additional data from Westinghouse on the effects of rods bowed 85%
1 of the way to contact which is consistent with the partial bow data from Combustion Engi6eering and confirms that the linear relationship used in the February 1977 model is conservative.
Hodges noted that there does not seem to be a problem with BWRs due to the lower pressures at which they operate. He noted that the PWR data shows that below about 1800 psi the effect of bowing on DNB is negligible even for rods bowed to contact.
In response to a question from Dr. Isbin concerning core spray, Hodges discussed the numerous tests performed by General Electric.
He indicated that even low flow tests on the order of I gpm, only had small effects on heat transfer and that even with only one header they had sufficient spray water to each bundle to assure adequate heat transfer. General Electric will, in the near future, be running a spray test in a full size 30 sector mockup of the BWR 6 upper plenum of a reactor at a test facility in Lynn, Massachusetts.
3.4.3 _ Systems Transient Analysis (BNL Development of the Incid6nt
_and Rapid Transient (IR11 Code) - NRC Staff Mr. M. Odar, NRC Staff, discussed the NRC Staff review prc:edure for transient codes noting the follovdng four steps:
Review Physical Content Make Autif t Calculations Compare Vendor Code Predictions With Actual Experimental Data Perform Standard Problems
ECCS 8/28, 29 & 30/78 Mr. Odat dn
. sed n BNL IRT which is the old Combustion CESEC ATWS Coo, ao.'fied ;.o meet the NRC Staff's needs (See Attachment FF-1). Attachments FF-2 and FF-3 show reactivities versus time for Calvert Cliffs with ar,d without steam generator tube ruptures.
It was noted that the plant does go critical for a short time after 40 seconds to a peak of about 200 MWt until boron injection turns the transient around. Odar had not performed this calculation for Westinghouse reactors but felt that they would be similar.
3.4.4 General Electric Transient Analysis Review: ODYN Code - NRC Staff Mr. M. Odar, NRC Staff, discussed the ODYN review schedule of the GE transient analysis review as shown on Attachmer,t GG-1.
The NRC Staff hopes to finish the review by the end of November 1978.
3.4.5 BNL Audit Calculations - NRC Staff Mr. M. Odar, NRC Staff, discussed the BNL audit calculation method taking the RELAP-3B and BNL-TWIGL and using a certain iterative procedure, verifying that measured versus calculated power reasonably agree (See Attachment HH-1). Odar noted that there was good compari-son during the analysis for several Peach Bottom transients.
In response to a question from Dr. Isbin concerning whether any of the results of these analyses have any effect on safety, Novak indicated that there was no basis to require any changes in the operating requirements for the plants. He noted that the tran-sients were very sophisticated. He added that if one fails to recognize this sophistication and tries to compare this analysis with a best estimate you could come up with suprisingly different results. He noted that the conservatisms, such as no credit for bypass, are significant and sufficient to overcome any deficiencies in the analytical techniques.
ECCS. 8/28, 29 & 30/78 3.4.6 Asymmetric Blowdown I.o,ds on Reactor Vessel - NRC Staff Mr. E. Throm, NRC S.aff. provided a status report on the generic issue of asymmetric blowdown loads on the reactor vessel. Throm discussed the history of meetings held on this subject and the status of each o; the vendors (Attachments JJ-l through JJ-4).
He noted that the Westinghouse methodology described in MULTIPLEX Code has been approved by the NRC Staff. The other vendors '
codes are still under review. The inclusion of General Electric is pending an NRC Staff decision on the safe shutdown earthquake and LOCA loads combinations.
It was suggested that a separate ACRS Subcommittee meeting, with Dr. Plesset as Chairman, be held to review this area in more detail.
3.5 Core Performance Branch 3.5.1 Multi-Rod Burst Test (MRBT) - NRC Staff Mr. D. Powers, NRC Staff, noted that the multi-rod burst test pro-gram at Oak Ridge has been beneficial in providing the NRC Staff with an ability to assess the conservatism in various portions of Vendors' ECCS models, resulting in the identification of a non-conservative Combustion Engineering rupture strain model.
He noted that it is CE's opinion that the use of the revised steam cooling model will offset the use of revised rupture strain model such that the adequacy of the overall ECCS model is unchanged.
Mr. Powers described some of the benefits derived from the MRBT.
He noted that they have learned from the program that the method of heating strongly influences the degree of induced temperature gradients in the cladding and that these gradients directly affect the location and magnitude of bcllooning. More unifonn temperature.
from the surrounding rods, causes the entire rod to undergo more deformation than that observed in the single rod tests. Another benefit derived was the observance of the strong influence of the ramp rate on the temperature at which the cladding fails.
., ~
7.
ECCS 8/28, 29 & 30/78 Increasing the ramp rate leads to an increase in pcessure at which earlier failure occurs.
Mr. Powers noted that the MRBTs have not resulted in any new mechanism for fuel failure propagation nor has there been any j
identified rod-to-rod interaction that affects the burst location.
3.5.2 UKAEA Rod Burst Experiments (Hindle Ballooning Tests) - NRC Staff Mr. D. Powers, NRC Staff, discussed the NRC involvement in some British rod ballooning experiments conducted at Springfield, England by Dr. Ed Hindle. His comments were directed only at Dr. Hindle's single-rod tests in which single-rod specimens 0
were soaked at 600 C, then ramped at about 10 C/sec to test tem-perature.
Post-test measurenents revealed extensive axial ballooning of up to 15" with large diametrical strains typically attaining 70% and occassionally over 100%. Mr. Powers discussed numerous reasons for Hindles test conditions promoting ballooning that were not relevant to PWR LOCA tests (See Attachment KK-1).
This list was based on two workshops and several less formal meetings with the Britsh.
Powers noted that others who partici-pated in these workshops, including British investigators, con-cluded that Hindle's experiments were not relevant to the PWR ECCS analysis.
3.5.3 FRAPCON Code - NRC Sta ff Mr. D. Powers, NRC Staff, discussed the INEL and PNL FRAPCON Code Development program. He noted that NRR uses fuel performance codes i n licensing to audit stored energy and to aid in the review of various submittals. The NRC has had two steadysstate fuel per-formance code development programs, GAPCON and FRAP. The FRAPCON Code integrates the best parts of these two codes to give both best estimate and evaluation model predictions (See Attachment LL-1).
Attachment LL-2 shows what NRR expects to get from this Code coordination program.
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4
ECCS 8/28, 29 & 30/78 3.5.4 Zircaloy Embrittlement - NRC Staff Mr. D. Powers, NRC Staff, discussed the zircaloy embrittlement study at ANL which will provide an assessment of conservatisms in Appendix K and aid in the development of refined fuel damage criteria. Tests are conducted by subjecting cladding spet imens to a simulated LOCA by heating and pressurizing until rupture, then cooled and quenched by bottom flooding.
Results are shown on Attachments MM-1 and MM-2.
Powers felt this program will be useful in ti.cir generic study to obtain a more realistic DNB failure criterior..
3.6 Closing Statements In response to a question from Dr. Isbin concerning the Edgar Series tests, Dr. Picklesimar said that a CSNI presentation published in 1976 and dis-cussed at the June 30, 1978 workshop meeting in Springfield, England indi-cated that the Edgar Series used the same type of techniques used by Hindle.
He said as far as he could find out there is no information in these tests which is really applicable to a LOCA in a PWR.
Dr. Isbin thanked the NRC Staff for their participation in the meeting.
The meeting was adjourned at 2:05 p.m.
NOTE:
For additional details, a complete transcript of the meeting is available in the NRC Public _ Document Room,1717 H St., NW, Washington, DC 20555, or from Ace-Federal Reporters, Inc., 444 North Capital St., NW, Washington, DC.
/
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E
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.,I
(.
before the meetir% This application stat ement s.
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review by the full committee.
U.S. Nuclear Regulatory Commissici Room 300s Washington, D.C.
Further information regarding Washington, D.C. 20SS5, Attention: Di Signed at Washington, D.C., this 7th toples to be discussed, whether the rector, Division of Technical Informa-day of August 1978.
meeting has been canceled or rcsche-tion and Document Control. Tele.
)
duled, the Chairman's ruling on rc.
phone requests cannot be accommo.
IsAntr. 7. SAwntt.t.*
quests for the opportunity to present dated. Regulatory guides are not copy.
I Director, Nctio'ncl oral statements and the time allotted righted, and Commission approval is i
Commistfon /or menpoteer Policy.
therefore can be obtained by a prepaid not required to reproduce them.
IFR Doc. 'T8-22448' Filed 8-10-78; 8:45'am) telephone call to the designated Fed-(5 ESE. 552(an 7
eral employee for this meeting. Dr.
Andrew L. Bates, telephone 202-034-Dated at Rockville, Md., this 3d day 17570-01]
3267 between 8:15 a.m. ar.d 5 p.m.,
of August 1973.
I'
- d l*
'For The Nuclear Regulatory Co=.
NUCLEA!! Rf!GtJLATOP.Y y
Dated: August 8,1078.
Inission.
COMMISSION on C. Hm.
Roam E ?. Mom, ADVl50P.Y ColUA!TTTE CN PEACTOR 5AFE.
stduisory Commf ttee Director, O// ice OUARDS, SU0CCWARThE CII E.V.E!tGDICY Mananmenf O/Accr.
oman&rds Deretop.ent Coitt COOL!rtG 3MtO45 (LCCS)
(FR Doc. 78-22405 Filed 8-10-78; 8:45 aml
[FR Doc. 78-22405 Filed 6-10-78; 8:45 a r.1 M nllog The ACRS Suiscemmittee on Emer- [7590-01).
F7590-01]
gency Core Cooling will hold an open mectirig on Augtest 28-20-30,1978 in REGUIATO'tY CUIDE IDocket Nos.SIW 50-556 AND STN 50-55%
Room 1046,171711 Street NW., Wash-I'8"*"" *"4 A'"II*MU'Y PUntlC SERVICE COMPANY OF CKLAHOMA ington, D.C. 20555, to discuss the status of a variety of programs related The Nuclear Regulatory Commission
^""**Y d " 3"* D'd'"" d #'
to ECCS-LOCA rerearch programs has issued two revised cuides in its Atonde sow and Uundng and and lleensing activitics. This meetinc Regulator 1 Guide Series. This series was previou*.ly trulounced as sched-has been developed to describe and Pursuant to the National Envirom uled to be held on August 22-23, 1978 make availabic to the public methods mental Policy Act of 1969 and the (43 FR 26102 and W31, June 1G nnd acceptable to the NRC staff of imple-United States Nuc! car Regulatory July 17, respectively).
menting specific parts of the Commis. Commisalon's regulations in 10 CFH In ac.:ordance with. the procedures sion's regulations and, in come cases, Part 51, notice is hereby given that i outilned in the Pcmuert. RtctsTER on to delinente techniques used by the Partla! Initial Decision AuthorirIM October 31,1977 (42 FR 50072), oral or staff in evaluating specific prob! cms or Lim! Led Work Authorization date:
Written stntemchts may be presented postuinted accidents and to provide July 24.1978. by the Atomic Safety by members of the public, recordings guidance to applicants concerning cer-and Lleensing Donrd in the abovc<a?
will be permitted only during those Lain of the h1 formation needed by the tiened proceedings is available for 11 l'
portions of the meettn;: when a tran-staff in its review of applications for spection by the public in the Comrp script is being kept, nnd questions may permits and licenses, sion's Public Document Room at 171*
be esked only members of the subcom-lleculatory Guide 1.84, Revision 13, H Street NW., Washincton. D.C., n~4 mittee, its coruuttants, and staff, l'er.
" Design and Fabricatiou Code Case at the TuSa City County Library, N' sons desiring to make oral statements Acecutability-ASME Section III Divi. Civic Center Tulsa, Okla.
4 Based on the record developed in th public hearinc in the above.captione.{
should notify the designated Federal sion 1," and Reculatory Guide 1.85 employec ns int in advance as prnetica. ; llevision 13. "Materints Code Case Ac.
'ble so' that npproprinto nrrancements ; ccptability-ASME Section ll! Divi-matter, the Partial Initial Dee!A'S
., can be made to allow the necessary alon 1," lis't' those code cases that arc modified in certain respects the co*
FEDERAL RfCtSTER, VOt. 43, No.15A TRIDAY, AUGUST 11, W8
-h ATTACHMENT'A 1
k_-___
___.-.-._____i'.______.,_.1______.
'Ibntative Schedule for ICPS ECCS Subcommittee ' Meet 2ng August 28, 29, 30, 1978 Monday - August 28, 1978 I.
Executive Session - Opening Comments
'H. Isbin 8:45 am II.
Summary of IOCA-ECCS Pesearch L. S. Tong 9:00 am A.
Matrix of Efforts B.
Completed Tasks III.
Blowdown and Paflood Heat Transfer E. H. Davidson 9:25 am Summary IV.
NRC/EPRI/GE Blowdown ECCS Program G. Sozzi 10:00 am (GE)
Break 10:15 - 10:30 am V.
PWR Blowdown Heat Transfer R. A. Hedrick 10:30 am (ORNL)
VI.
NRC/EPRI/_W FLECHT-SFASET L. Hochreiter 11:00 am VII. NRC Inhouse Heat Reflood Studyrts L. Thompson 11:30 am VIII. Instrumentation and M'odel Development Y. Y. Hsu 12:00 an Summary i
Discussion 12:00 - 12:30 pn Innch 12:30 - 1:30 txa IX.
Instrumentation and Model Development j
Prograns A.
Ichigh University Program J. Chen 1:30 pm 4
B.
ANL Program P. Kehler 2:00 pn ATTACHMENT B g
f-r 4e
.n/
2CCS Tentative Schedule August.28-30, 1978 C.
Suny-Stonybrook Program R. Lee 2:30 pm D.
RPI Program R. Lahey 3:00 pm Break 3:30 - 3:45 pm E.
Instrumentation Progams at:
Y. Y. Hsu 3:45 pn 1)
Sandia Program 2)
LASL Program F.
University of Houston Program A. Dukler 4:30 pm G.
ANL Program P. Lottes 5:00 pn H.
Modeling Programs at:
Y. Y. Hsu 5:30 pm 1)
BNL 2)
Northwestern Discussion and Adjourn 6:00 pm
'Itesday - August 29, 1978 I.
'3ubcommittee - Opening Comments H. Isbin 8:45 am II.
Summary of Selected Code Development S. Fabic 9:00 am Programs III.
OJBRA Code D. Trent 9:30 am (PNL)
IV.
'IHOR Code O. C. Jones 10:00 am (BNL)
Creak 10:30 - 10:45 cm V.
UHI Modeling M. Berman 10:45 am (Sardia)
VI.
Statistical Anu_ysis M. Berman 11:15 am (Sandia)
M. Buckner 11:45 am VII. hPAP-EM (SRL)
Inrx6 l
. h k;
ECCS Tentative' Schedule August 28-30, 1978 VIII. IMR Containment Modeling and' Analysis S. Fabic 1:30 pm.
[
Summary
~
A.
Studies at MIT A. A. Sonin 2:00 pm B.
Studies at UCLA Y. K. Dhir 2:30 pm Break 3:00 pm C.
Studies at LLL E. McCaully 3:15 pn I
D.
PELE-IC W. McMaster 3:45 pn (LLL)
Discussion and Adjourn 4:15 - 5:00 pm i
Wednesday - August 30,1578 I.
Executive Session - Opening Comments
- I. Isbin 8:45 am j
II.
Status of NRR Programs D. Ross 9:00 - 12:30 pm l
A.
Reactor Systems Branch T. Novak (1)
Failure Modes and Effects (2)
Component Reliability Studies
]
(3)
Recent. Events B.
Analysis Branch Z. Rosztoczy (1)
Plans for Audit of Vendor Code Quality Assurance Programs (2)
Appendix K Behavior (3)
GE CDEN Code (4)
Subcooled Ioads (5)
Transient Analysis Work at BNL C.
Core Performance Branch K. Kniel (1)
Zircaloy Swelling & Embrittlement
'(2)
GAPCON Code 4
Discussion and. Adjourn
.12:30 - 1:00 pm e
3' h &
t 4
ACES SUBCOMMI'I'IEE MEETI1U ON EMEIGENCY CORE COOLING SYSTEMS WLSHI1GION, DC AUGUST 28-30, 1978 A'ITENIW4CE LIST AUGUST 28, 1978 i
ACRS NRC STAFF H. Isbin, Chairman Y. Hsu M. Plesset, Co-Chairman R. Budnitz H. Etherington L. Tong I. Catton, ACRS Consultant N. Zuber R. Shumway, ACRS Consultant W. Paulson T. Theofanous, ACRS Consultant F. Zaloudek, ACRS Consultant A. Bates, Designated Federal Employee ARGONNE NATIOR\\L LAB.
G. Quittschreiber, ACRS Staff P. Lottes SUNY - S'IONYBROOK WESTINGHOUSE ELECTRIC CORP.
R. Leel J. Srinivasan H. Massieur L. Hochreiter GENERAL ELECTRIC CO.
HARWELL G. Sozzi P. Hutchinson OAK RIDGE NATIONAL IAB.
UNIVERSITY OF HOUS' ION J. White A. Dukler ATTACHMENT C i
l.
g, 4 -
g a
i-ACRS SUBCOMMI'ITEE MEETING ON MEFGENCY CORE COOLING SYST t
~ ' '
hASHINGION, DC l
AUGUST 28-30, 1978
.ATIENDANCE LIST AUGUST 29, 1978 NRC STAFF _
ACRS L. Thompson H. Isbin, Chairman L. Tong M.-Plesset, Co-Chairman S. Fabic H. Etherington N. Zuber I. Catton, ACRS Consultant R. Shumway, ACRS Consultant T. Theofanous, ACRS Consultant SAVANNAH RIVER IAB.
F. Zaloudek, ACRS Consultant A. Bates, Designated Federal Employee S. Duraiswamy, ACES Staff M. Gregory OAK RIDGE NATIONAL LAB.
PACIFIC NORTHWEST LAB.
J. hhite D. Trent i
M. Thurgood UNIVERSITY OF CALIFORNIA f
V. Ohir BROOKHAVEN NATIONAL IAB.
i :
s.
O. Jones,-Jr.
DEPA_RINENT OF ENERGY W. Wulff W. Kato F. Watson Ab WESTINGHOUSE ELECTRIC CORP.
L. Hochreiter 1
i L'-
{
f.
1-a
--+
ACRS SUBCOMMITTEE MECTING ON DIERGENCY CORE COOLING SYSTEMS M\\SHINGION, DC -
AUGUST 28-30, 1978 1
i ATTENDANCE LIST AUGUST 30, 1978 ACRS NRC STAFF H. Isbin, Chairman J. Olshinski M. Plesset, Co-Chairman B. Mills H. Etherington J. Watt I
I. Catton, ACRS Consultant M. Rubin R. Shumway, ACRS Consultant' T. Novak T. Theofanous, ACRS Consultant D. Ross F. Zaloudek, ACRS Consultant M. Picklesimer A. Bates, Designated Federal Employee P. Norian S. Duraiswamy, ACRS Staff E. Throm i
G. Quittschreiber, ACRS Staff N. Wright F. Odar D. Powprs WESTINGHOUSE ELECTRIC CORP.
W. Hodges j
D. Paddleford i
BROOKHAVEN NATIONAL IAB.
BABCOCK & WIILOX D. Dia:rond l
l R. Borsum
[
k i
m i
d i
1 i
i i
1
6 DOCUMENTS PROVIDED TO THE EMERGENCY CORE COOLING SYSTEMS SUBCOeNITTEE FOR THE AUGUST 28, 29, AND 30,1978 SUBCOMMITTEE MEETING 1.
Presentation Schedule (Attachaunt B) 2.
Copies of viewgraphs (Attachments E-1 through MM-2). A complete set of. attachments is available in the ACRS Office copy of these minutes.
9 ATTACHMENT D
]
LOCA ECC EXPERIMENTAL PROGRAM MATRIX t
I Separate Etlects Model Size Integrat
.g Linear Scale Systam Blowdown Heat Reflood Heat ECC Bypass Pump and Steam Upper Plenum Transfer Transfer Birfn & Refill Generator De entrainment
]
Jspanese SCTF
- CE/EPRI FRG UPTF FRG UPTF Nearly Full Analysis
.CE/EPRI BDHT W FLECHT-SEASET 1/4 & ;;S sca;e Japan SCTF g
GE TLTA-1
- PKL-340 rod bundle
[BWR CCFL/Reflood]
Intermediate INEL LOFT INELCUFT INEL LOFT INEL LOFT (CE LOFT Pump)
Jspan CCTF Scale Japan CCTF Japanese CCTF Japan CCTF INEL LOFT S.G.
BCL 2/15 Model (WCL Semiscale Pump)
FLECHT SE ASET pan S
'.NE l.
iNEL Semiscale t
e Small C ae /5 del Scale W LEC T SET GE TLTA-2 S.G.
ANL Freon Loop Basic
- (W tube BDitT)
MIT Freon DARTMOUTH
- Creare/EPRI H*'* ii Model (MIT Freon Loop)
AECL Model Dev.
S PR M Study (Hannc,ver Freon)
RPl Model Test Lehigh Model Test l
\\
i i
( ) Completed Program. [ ] Proposed Program Non-NRC Funding 8 00 /1 1
~
7..
i l: ',.
7
,.f i
's t &
3, UNION CARBIDE RNL PLANS FOR ORNL-PWR BL'OWDOWN HEAT -TR ANSFER_
i SEPARATE-EFFECTS PROGRAM TESTS ASSUMED TO -
DATE BE CONDUCTED DESCRIPTION OF TEST SsRIES INITIATED WITH BUNDLE NO.
- CHOPPED COSINE POWER PROFILE, TIME 5/27/76 1
l TO CHF, REACTOR. BASE CASE
}
- INACTIVE FUEL ROD SIMULATOR TESTS 11/5476 1
1 (TIME TO CHF)
- SPECIAL SEPARATE-EFFECTS TESTS 1/13/77 1
(TIME TO CHF) e. POWER DECAY TESTS 2/79 1 OR 2
- FUEL AND CONTROL ROD SIMULATOR DEPENDS ON 2 OR 3 TESTS (TIME TO CHF)
BUNDLE NO.
o' FLAT POWER PROFILE, TRANSIENT 8/79 3
PRE-AND POST-CHF HEAT TRANSFER TESTS
- NORMAL OP$ RATING CONDITIONS
- OFF-NORMAL OPERATING CONDITIONS
- DATA BASE GENERATION-h, LOCAL FLUID CONDITIONS TRANSlENT SWITCHING CRITERIA
- SEVEN ROD ZlRCALOY CLUSTER,-
6/80 3
j HIGH-TEMPERATURE TESTS.
1 o
3.'
i j.
4 f
f p
kr
z: :~
N. :.
Q CARBIDE v ""
LONG RANGE PLANS FOR ORNL-PWR BLOWDOWN HEAT TRANSFER SEPARATE-EFFECTS PROGRAM R..
4
~
TESTS ASSUMED TO DATE BE CONDUCTED DESCRIPTION OF TEST SERIES INITIATED WITH BUNDLE NO.
- BOTTOM POWER SKEW TIME TO CllF TESTS 11/80 4
- BOTTOM POWER SKEW, TRANSIENT POST-CHF 6/81 4
_.I HEAT TR ANSFER TESTS
- BOTTOM POWER SKEW,7-ROD ZlRCALOY 1/82 4
CLUSTER HIGil-TEMPERATURE TESTS
- TOP POWER SKEW TIME TO CliF TESTS 6/82 5
llEAT TRANSF.ER TESTS 3[b3 5
- TOP POWER SKEW,7-ROD ZlRCALOY CLUSTER HIGH-TEMPERATURE TESTS
- SMALL BREAK LOCA STUDIES 3/84 6
i.
j; 10
.g CARBIDE RNL.
EXPERIMENTAL DATA o CALIBRATED AND. CONVERTED ONLY o SCOPE AND RANGE OF PARAhlETERS FOR PWR BLOWDOWN o APPLICABLE TO LWR SAFETY IN GENERAL (BWR, STEAM GENERATORS) o LARGE ~1/2 MILLION DATA POINTS / TEST o EXPERIMENTAL DESIGN AND ACQUISITION ALLOWS VERIFICATION OF THEORETICAL PREDICTIONS o QUICK LOOK, DAT REPORTS, AND DATA BANK F-3 1
,c.
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i ORNL INTERPRETATION I
i o
RECORDED INSTRUMENT RESPONSES e
PHENOMENOLOGICAL SEQUENCE TIME TO CHF L
TRANSIENT HEAT TRANSFER SWITCHING LOGIC COUPLING OF ENERGY TRANSFER AND TRANSPORT o
(INDEPENDENT) RELAP VERIFICATION-NOT SEMISCALE GEOMETRY SCALE HEAT SOURCE VERIFIABLE HEAT TRANSFER A
o DATA EVALUATION REPORTS FA
7 -..
- b.,
.n. n..
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- 4. s:
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n.
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ANAiYSIS SYSTEM
~
.,.________r r.._ _ _
DATA ACQUISITION PROCEDURES CALIB RATION NUMERICAL TECHNIQUES DIGITAL REDUCTION ENGINEERING PARAMETERS SIMU LATO RS t4 o
HEAT TRANSFER COEFFICIENTS o
LOCAL FLUID CONDITIONS 4
o ORTCAL, ORINC, PINSIM, ETC.
g Iq o
DATA EVALUATION REPORTS
~
o ORNL REPORTS Q
.. *lI 31
'-^ ~
--~ ~ ~
1:
CARBIDE RNL
SUMMARY
OF MAJOR ACCOMPLISHMENTS 1.
. TEST SERIES 1,2,3A, AND 3B HAVE BEEN COMPLETED 1
QUICK LOOK REPORTS m
DATA REPORTS DATA EVALUATION REPORTS i
2.
LOCAL FLUID CONDITIONS HAVE~BEEN CALCUi_ATED FOR TEST SERIES 1
[
RELAP 4/5/2 COBRA 4 3.
LOCAL h VALUES HAVE BEEN CALCULATED; COM-
~
PARISONS HAVE BEEN MADE WITH CURRENTLY USED:
(a)
STEADY-STATE CORRELATIONS (b)
TRANSIENT SWITCHING CRITERI A i
4.
BUNDLE 3 DESIGNS H' AVE BEEN COMPLETED; MODIFICATIONS TO THE THTF FOR IMPROVED CALCULATIONS OF LOCAL FLUID CONDITIONS HAVE BEEN DESIGNED l
f
.I
t
'p42, CARBIDE
"" TIME TO CRITICAL -HEAT FLUX,, THTF TEST 105,.
SHOW. SIGNIFICANT DELAYS' COMPARED TO 0.1 SEC REQUIREMENT IN EVALUATION MODEL-NO. OF T/C AVERAGE STANDARD NO.OF NO. OF T/C TO CHF TIME D EVI ATION.
LEVEL T/C
'TO CHF
(<-2 sec)
TO CHF
' OF TIME-f D
18 17 17 6.6301 D-01 6.9285D-02 E
19 18 18 7.3274 D-01 1.4336D-01 U
.F 18 18 18 6.2981 D-01 1.3969D-01
'19 5.6387D-01 7.7565D-02 G
19 19 H
42 41 41' 7.2245D-01 1.4203D-01
~
I 42 37 19 1.0263D 5.2733D-01 J
39 39 7
9.1669D-01 4.4515D-01 K
20 14 0
0.0 0.0 L
41 7
0 0.0 0.0 Y
M 4
0 0
0.0 0.0 i 4 N
3 0
0 0.0 0.0 D
0 19 0
0 0.0 0.0 p.
8
...c 1
i FLECHT DATA AND REFLUX PREDICTIONS a
FOR CLADDING TEMPERATURE VERSUS TIME M
~
1250 g
i
.g i
g j.,:l * '.'
9 7,
.p"),
DATA N
REFLUX PREDICTION 1000 -[
N
(
...i>
o s
?
N N
w N
cc s
D s
H
's 750 VAPOR FILM (mm).
N
\\.06 w
0.03, *.
0 g
\\
w H
a 2
500 h
FLECHT TEST 9782 3~
FLOODING RATE 2.5 cm/sec O
PEAK POWER 4.1 kW/m DECAY CURVE
'A' 250 PRESSURE 0.39 MPa 0
INLET SUBCOOLING 77 C I
I I
I 0
60 120 180 240 300 360 O
TIME (sec)
If
&-)
9
- n. 1 -
- +,,
W
~,
FLECHT DATA AND REFLUX PREDICTIONS FOR CLADDING TEf1PERATURE RISE VERSUS INLET VELOCITY I
I I
I I
I i
l l
l
[
INITI AL PEAK AXIAL POWER SHAPE POWER RANGE (kW/m) 500 A COSifJE {
2.3-2.4 A SKEWED J
@ COSINE 2.9-3.3 INITIAL PEAK TEMPERATURE = 720-870 C 400 FOR V <10 cm/sec PRESSURE = 0.28 MPa DECAY POWER = ANS + 20%
jg g
e TEMPERATURE RISE AT PEAK AA og POWER LOCATION y 300 e
e E
R A
2
^
y 200 5
0 iA REFLUX A
3.1 kW/m 100 2.3 kW/m O
A O
g0 i
i i
i 1
2 3
4 5
6 7
8 9 10 20 30 40 50 INLET VELOCITY (cm/sec) 4, 0F 9
PWR REFLOOD RESEARCH INFORMATION LETTER
SUMMARY
OF FLECHT RESULTS COMPARISON OF FLECHT RESULTS WITH DATA FROM OTHER FACILITIES
- SEMISCALE - 40 R0D PWR, 1.7 M HEATED LENGTH
- KWU - 340 R0D PWR, 2.9 M HEATED LENGTH
- KFK - 25 RODS, ZIRCALOY AND STAINLESS STEEL CLADDING, 3.9 M HEATED LENGTH ANALYTICAL PREDICTIONS
- C0ilVECTION-CONTROLLING MODELS FOR THERMAL HYDRAULIC BEHAVIOR
- CONDUCTION-CONTROLLING MODELS FOR QUENCH FR0i(T VELOCITY
- SEMI-EMPIRICAL MODELS FOR QUEHCH FRONT VELOCITY G75 9 0F 9 6
L_;
/
--t u..
o tn BLOWDOWN Af>D REFLOOD HEAT TRANSFER l.
Leung, J. C. M., " Critical Heat Flux Under Transient Conditions:
A Literature Survey", Argonne National Laboratory, ANL-78-39, NUREG/CR-0056, June 1978.
2.
Cha, Y. S. and Henry, R. E., " Effects of Dissolved Gas During Blowdown Tests", ANL, Accepted for 1978 Symposium on Fluid Transients and Acoustics.in the Power Industry, San Francisco, California.
3.
Henry, R. E. and Leung, J. C. fl., "High Pressure Freon Blowdown Studies in a.2.75 m Nonuniform Heated Tube", ANL, to be published.
4.
Henry, R. E. and Leung, J. C. M., " Nucleation and Transient Critical Heat Flux Under LOCA Conditions", Argonne National Laboratory, to be published.
5.
Henry, R. E. and Leung, J. C. M., "A Mechanism for Transient Critical Heat Flux", Proceedings of Topical Meeting on Thermal Reactor Safety, Sun Valley, Idaho, July 31-August 4, 1977.
6.
Henry, R. E., and Leung, J. C., " Transient Critical Heat Flux in a 0.91 m Long Uniformly Heated Test Section During Blowdown of High Pressure Freon", Argonne National Laboratory, ANL/ RAS / LWR 77-1, February 1977.
7.
Henry, R. E., "A Comparison of Two-Phase Critical Flow Models j
with the Marvikea Test Results", to be published in the fourth quarter of 1978.
j 8.
Cha, Y.
S., " Thermodynamic Theory of Cavitation Nuclei in Dilute Liquid-Gas Solution", Argonne National Laboratory, ANL-CT-78-33, June 1978.
l 9.
Fung., K. K., " Post-CHF Heat' Transfer During Steady State and Transient Conditi.ons", ANL-78-55, NUREG/CR-0195, to be published.
10.
Chawla, R. C. and Ishii, M., " Multichannel Drif t-Flux Model in Constitutive Relation for Transverse Drift-Velocity", Arconne National Laboratory, ANL/ RAS / LWR 78-2, March 1978.
l 11.
Ishii, M., "One Dimensional Drif t Flux liodel and Constitutive Equations for Relative Motion Between Phases in Various Two-Phase Regimes", ANL-77-47, 1977.
]
I
,#l
=*
H <
Blowdown and Reflood I
g Heat Transfer 12.
Jones, O. C., and Zuber, N., " Evaporation in Variable Pressure Fields', Argonne National Laboratory, ANL/ RAS / LWR 76-2, April 1976.
- 13. Cheng, S. C., Heng, K. T. and Ng, W., "A Technioue to Construct a Boiling Curve from Quenching Data Considering Heat Loss", Int. J.
Multiphase Flow, 3, pp. 495-499, 1977.
14 Ragheb, H.
S., "Developinent of Electric Probes to Detect Phase Change at a Heat Surface," University of Ottawa, Master's Thesis, November 1977.
15.
Cheng, S. 'C., et al., " Transition Boiling Heat Transfer In Forced Vertical Flow", Final Report, June 1977-June 1978, University of Ottawa, June 1978.
16.
Cheng, S. C., et al., " Transition Boiling Heat Transfer In Forced Vertical Flow, Final Report, June 1976-June 1977", University of Ottawa, June 1977.
17.
Groeneveld, D. C., "Effect of a Heat Flux Spike on the Downstream Dryout Behavior", Journal of Heat Transfer, pp. 121-125, May 1974.
18.
Fung, K. K. and Groeneveld, D. C., " Heat Transfer Experiments in the Unstable Post-CHF Region", Soviet-Canadian Symposium on Thermal Physics, Moscow, USSR, October 20-22, 1976.
19.
Groeneveld, D. C.,"Fost Dryout Heat Transfer:
Physical Mechanism and a Survey of Prediction Methods", Nuclear Engineering and Design.
20.
Delorme, G. G. J. and Groeneveld, D. C., " Prediction of Thermal Pon-Equilibrium in the Post-Dryout Regime", Nuclear Engineering and Design, 36, pp. 17-26, 1976.-
l 6
21.
Gard,i,ner, S. R. M. and Groerieveld, D. C., " Post-CHF Heat Transfer Under Forced Convective Conditions", Thermal and Hydraulic Aspects of Nuclear Reactor Safety, 1, pp 43-73, 1977.
22.
Chen, J.
C., Sundaram R. K. and Ozkaynak, R.
T.,
"A Phenomenological Correlation for Post-CHF Heat Transfer", Lehigh University, NUREG-0237, June 1977.
~,
23.
Lehigh University, " Analysis and Correlation of Post-CHF Heat Transfer", Quarterly Reports; October-December 1977, LU-NUREG-PR/74, January-March 1978, LU-NUREG-PR781, April-June 1978, LU-NUREG-PR782.
N Es ns Blowdown and Reflood Heat Transfer.
24.
Munno, F.
J.,
Sheaks, O. J. and Smoker, R' R., "An Experimental Investigation of Heat Transfer in the Post-Critical Flux Regime" Final Report, University of Maryland, April 1978.
25.
Bengston, S.
J., " Analysis of Uncertainty In Steady State Flow Film Boiling Data", Letter Report, EG&G, July 1977.
26.
Varacalle, D.
J., " Status of Local Conditions Methods Devleopment of the Heat Transfer Data Bank", Letter Report EG&G, January 1977.
27.
Anderson, B.
S., " Comparison of a Developmental Transient Critical Heat Flux Correlation (TCHF1) To Semiscale Experimental Data",
EG&G, PG-R-77-03, February 1977.
28.
Schulz, G.
L., et al., "NRC/RSR Data Bank Program Description",
Letter Report, EG&G, November 1976, 29.
Richlen, S.
L., " Status of Rod Bundle Screening and Correlation i
Comparisons", Letter Report, EG&G, October 1976.
30.
Schulz, G.
L., et al., "NRC/RSR Data Base System Design and Implementation Proposal", Letter Report, EG&G, September 1976.
31.
Richlen, S. L., "Effect of Axial Conduction During Reflood",
Letter Report, EG&G, June 1978.
32.
Richlen, S.
L., "Chen Non-Equilibrium Correlation Evaluation",
EG&G Idaho, CDAP-TR-004, January 1978.
33.
Nakovick, N.
C., " Flow Configurations In A Simulated Small Break Loss-uf-Coolant accident", Raster's Thesis, Massachusetts Institute of. Technology, January 1978.
34.
Griffith, P. and'Kaufman, J., M., " Post-Critical Heat Flux Heat Transfer to Water in a Vertical Tube", Massachusetts Institute of Technology, NUREG-0164, January 1976.
35.
Smith, T.
A., " Heat Transfer and Carryover of Low P'ressure Water in a Heated Vertical Tube", Massachusetts Institute of Technology, NUREG-0105, August 1976.
36.
Kirchner,W. L., "Reflood Heat Transfer in a Light Water Reactor",
Massachusetts Institute of Technology, NUREG 0106, August 1976.
37.
Cluss, E. M., Jr., " Post-Critical Heat Flux Heat Transfer in a Vertical Tube Including Spacer Grid Effects", Massachusetts Institute of'Technoicgy, NUREG/CR-0337, June 1978.
A
~~
~
e-4 ed
- u. '
C)
Blowdown and Reflood Heat Transfer oo 38.
Gonzalez-Rivas, J., " Carryover Predictions from the Upper Plenum of a Pressurized Water Reactor Based on a First Order Void Fraction Analysis", Massachusetts Institute of Technology, NUREG/CR-0338, May 1978.
39.
Azzopardi, B. J. and Smith, R. V,
" Summary of Reported Droplet Size Distribution Data In Dispersed Two-Phase Flow", Wichita State, August 1978.
Tong, L.
S., " Heat Transfer In Reactor Safety", in Proceedings of the 40.
Sixth International Heat Transfer Conference, Toronto, Canada, August 1978.
41.
Tong, L. S., et al., " Heat Transfer Phenomena and Predictive "ethods Important in Reactor Safety Studies", to be published as NUREG Report, Fourth Quarter FY 1978.
42.
Hsu, Y. Y. and Thompson, L.
B.,
Reflood Research Information Letter, Systems Engineering Branch, NRC, Fourth Quarter _FY 1978.
43.
Y. Y. Hsu, Proceedings of the Von Karman Institute for Fluid Dynamics Lecture Series, Brussels, Belgium, April 1978:
"Two-Phase Flow Problems in PWRs", " Boiling Heat Transfer Equations", " Condensation Heat Transfer" and " Codes and Two-Phase Flow Heat Transfer Calculations".
e a'
~
m*
u
-s cn INTERFACIAL TRANSPORT PHENOMENA AND
~ INSTRUMENTATION DEVELOPMENT 1.
Chen, L. Y., Drew, D. A. and Lahey, R.
T., " Virtual Mass Effect In Two-Phas'e Flow" Rensselaer Polytechic Institute, NUREG/CR-0020 Master's Thesis, January 1978.
Krycuk, G., Lahey, R. T. and Saba, N., "An Experimental Technique 2.
for the Determination of Steam / Air Fraction", Rensselaer Polytechic Institute, NUREG/CR-0021, 27th ANS Proceedings, December 1977.
3.
Lahey, R. T., " Critical Power In Boiling Water Nuclear Reactor
~
Fuel Bundles", Proceedings of Topical Meeting on Thermal Reactor Safety, Sun Valley, Idaho, July 31-August 4, 1977, CONF-770708.
Rensselaer Polytechic Institute, Two-Phase Flow Phenomena, Quarterly 4.
Reports; June-August 1977, NUREG/CR-0023, September-November 1977, NUREG/.CR-0035, December 1977-February 1973, NUREG/CR-0233.
5.
- Kehler, P., "Two-Phase Flow Measurement by Pulsed Neutron Activation Techniques", ANL-NUREG-CT-78-17, January 1978.
6.
- Kehler, P., " Measurement of the Emergency Core Coolant Bypass Flow cn the LOFT Reactor", NUREG/CR-0208, ANL-CT-78-37, July 1978.
7.
- Kehler, P., " Feasibility of Neutron Activated Tracer Techniques for Measurement of Flow Distributions. in 30 Upper Plenum Tests",
Argonne National Laboratory, August 1978.
Lee.S. L. and Srinivasen, "Me'asurement of Local Particle Size and 8.
Velocity Probability Density Distribution in Two-Phase Suspension Flows by Laser-Do' pler Technique", International Journal of Multiphase p
Flows, 4, pp. 141-155, 1978.
e
~
tr a
M uo Interfacial Transport Phenomena and S
Instrumentation Development 9.
Lee, S.
L., and Srinivasen, J., "tieasurement of Turbulent Dilute Two-Phase Dispersed Flow in a Vertical Retangular Channel by Laser Doppler Anemometry", accepted for 1978 ASME Winter Meeting, San Francisco, California, December 10-15, 1978.
10.
Lee, S. L. and Srinivasen, J., "An Experimental Investigation of Dilute Two-Phase Dispersed Flow Using LDA Technique", in Proceedings of 26th Heat Transfer and Fluid Mechanics Institute, June 1978.
"nscillatino Flow of Lee, S. L., Einaz, S. and Digiozanni, T. R.,ler Anemometry", in 11.
a Laminar Suspension Measured by Lhser Dopp Proceedings of the Sixth Canadian Congress of Applied Mechanics, Vancouver, 8.
C., Canada, May 30-June 3, 1977.
12.
Azbel, D. S., Lee, S. L. and Lee, T. S., " Acoustic Resonance Theory for the Rupture of Film Cap of a Gas Bubble At a Horizontal Gas-Liquid Interface", in the Proceedings of the 1978 International Conference on Momentum, Heat and Mass Transfer in Two-Phase Energy and Chemical Systems, Dubrovnik, Yugoslavia, September 4-9, 1978.
- 13. Srinivasen, J., " Development of LDA Technique for the Measurement of Two-Phase Dispersed Flow", Ph.D. Thesis, SUNY Stonybrook, flay 1978,
- 14. Bankof f, S. G., Thankin, R. S. and Yuen, M. C., " Condensation Rates In Steam-Water Mixing", Northwestern University, Quarterly Report June-September 1977, NUREG/CR-0024, March 1978.
- 15. Simoneau, R.
J., " Release of Dissolved Nitrogen from Water During Depressurization", NASA Technical Memorandum, NASA TM-73822, April 1978.
- 16. Oak Ridge National Laboratory, JAdvanced Instrumentation for Reflood Stud.its, Quarterly Reports, October-December 1977, NUREG/CR-0213, January-March 1978,.NUREG/CR-0249.
- 17. Oak Ridge National Laboratory, Advsnced Two-Phase Flow Instrumentation Quarterly Reports, January-March 1977, ORNL/NUREG/TM-119, April-June 1977, ORNL/NUREG/TM-140, July-September 1977, ORNL/NUREG/TM-183, October-December 1977, NUREG/CR-0242.
- 18. Jones, Jr., 0. C. and Saha, P., "Non-Equilibrium Aspects of Water Reactor Safety", BNL-NUREG 23143, July 1977.
- 19. Leonhardt, W. J., 'et al., "Brookhaven Heat Transfer Facility No.1:
Design Report", BNL, to be published.
"1 u.o e -i l
Interfacial Transport Phenomena and Instrumentation Development 4
- 20. Saha, P., "A Review of Two-Phase Steam-Water. Critical Flow Models With Emphasis on Thermal Non-Equilibrium", BNL, to be published.
- 21. Abuaf, N., Jones, 0. C. and Zimmer, G. A., " Response Characteristics of Optical Probes", BNL-NUREG-50791, March 1978.
- 22. Abuaf, N., Jones, 0. C. and Zimmer, G.
A., " Void Fraction and Interface Velocity Measurements with An R-F Probe", ASME Conference, May 1978.
- 23. Jones, O. C., " Liquid-Deficient Cooling in Dispersed Flows:
A Non-Equilibrium Relaxation Model', BNL-NUREG-50639, Part 3 December 1977.
Abuaf, N., Jones, O. C. and Zimmer, G. A., " Optical Probe for 24 Local Void Fraction and Interf ace Velocity Measurements", BNL-NUREG-50791, March'1978.
- 25. Estrada, H. and Sheppard, J. D.,'"Some Aspects of Interpreting Two-Phase Flow Measurements In Instrumented Piping Spool Pieces",
j ORNL-NUREG-0280, June 1977.
- 26. Hsu, Y. Y. (Editor), " Proceedings of the Two Phase Flow Instrumentation Review Group Meeting", NUREG-0375, January 1978.
l O.
S.
b fI L_ 9N-e
INSTRUMENTATION Improve New Devices Modeling '
Calibration 3-D Instrumentation Existing Device t.
- 0RtlL Adv.' Spool Auburn Meter Cross Calibration Support 3-D R/D Piece AP-device Fluidic Correlation Methods l0RNL local probes for Eads 3-D tests
- LASL
. Storz lens (Low T.P. &
. llelmick -
Software)
Bt1L Local probes Global &
Jones for Vel. & void local modeling RPI IR, RF, Modeling Lahey-2-probe X-ray SUNY-SB Laser Doppler
~
Lee (for drop site
& vel.)
Attl Neutron beam Calibration Kehler activation
'Sandia Pulse neutron i
generator Q
T Lehigh Local super Film probe R/D Chen heat
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COBRA-TF MODEL DESCRIPTION
[
Lo TFIELD EQUATIONS-
~ MASS-VAPOR, LIQUID, ENTRAINED LIQUID-ENERGY-VAPOR, LIQUID MOMENTUM-VAPOR, LIQUID, ENTRAINED LIQUID o ~ CONSTITUTIVE IEQUATiONS.
REGIME MAP VAPOR GENERATION RATE
~
INTERFACIAL DRAG WALL SHEAR
~
ENTRAINMENT RATE WALL HEAT TRANSFER
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STATE-LIQUID AND VAPOR
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THOR FEATURES
.i 1.
LUMPED-PARAMETER AND DISCRETE PARAMETER DESCRIP ARE COMBINED.
2.'
EULERIAN ME5H AND MOVING FLOW REGIME BOUNDARkES SERVEAbBOUNDRIESOFCOMPUTATI0tALZONES PROVIDEBOUNDARYCONDITIbNSWHICHARECONTINU0US IN IIME; 1
PROVIDETRCKIN5CAPABIblTY((EVELS;ETC.),
REQUIRE LOGIC.
3.
INDIYiDUALCOMPONENTMODEbhNd SPECIALIZING ON EACH PERIOD OF TRANSIENTS.
TAKING ADVANTAGE OF INDIVIDUAL CHARACTERISTICS, LEADING TO DIFFERENT SSUMPTION5FOREACHCOMPONENT.
INDIVIbuALCOMPONENTN0DULES, i
SEPARATELYEXECUTEDINTHOROPERATINGSYSTEti, INDIVIDUAbDEVELOPMENTbVERIFIbATION.
BROOKHAVEN NATIONAL LABORATORYg3 g)]
A5500ATED UNIVERSITIES, INC.O 15 2 k
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THOR FEATURES (CONT.')
4.
EFF1CIkNTCOMPUTINGPROCEDURE, l
i PARTIONING OF HYDRAULIC NETWORK INTO CHAINS
-l BLOCK ELIMINATION AtlD GAUSSI All ELIMINATION i
5.
USE OF LARGE CORE MEMORY WITHOUT OVERLAY:
THOR CODE INSTRUCTIONS IN SQ{
i DATA STORAGE ARRAY IN LCM..
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BROOKHAVEN NAll0NAL LABORATORY 33 33 ]
A5500ATED UNIVERSITIES, UK.G 13 3 i
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V il l P R O G R A ll RES FY78/79 A,
PROVIDE TECHillCAL ASSISTAi!CE TO LICENSIllG B.
VERIFY TilEORY Af1D APPLICATION OF SLIP AflD QUEf1CH F10DELS BY:
- 1. INVESTIGATIi4G SIMPLE TEST PROBLEMS 1
2,,
COMPARING AGAlilST RELAP4/ MOD 6 AilD AGAINST SATAN ON SIMPLE AflD UHI PROBLEMS
- 3. COMPARING AGAlilST EXPERIMENTAL DATA C.
If1VESTIGATE THE f1EED FOR NEW OR REVISED MODELS.
IF NECESSARY, DEVELOP, AND VERIFY (AS IN B) MODELS Ill TliE FOLLOWING AREAS:
- 1. LICENSING MODELS'
- 2. WATER PACK
- 3. BUBBLE RISE
- 4. REFILL AND REFLOOD e
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l' S=T A T I S T I C A L LOCA FY78/79 4
1 i
GOAL - ESTIMATE PCT PROBADILITY DISTRIBUTI0il FOR BLOWDOWii 1
l AND REFLOOD FOR ZI0il PWR USING RELAPl1/ MOD 6.
j PROGRAM:
l.
IMPLEMENT MOD 6/UPD3 ON SAf1DIA 7600, 2.
CHECK BL0bD0ili'l DIALS, 3.
SENSITIVITY STUDIES INCLUDIllG INITIALIZATION.
11.
AUTOMATE BD INPUT, 5.
MAKE RUNS, BUILD BD RESPONSE SURFACE.
6.
GENERATE REFLOOD INPUT USING " BRIDGE "
7, REPEAT 2-5 FOR REFLOOD, 8.
GENERATE PCT PROBABILITY DISTRIBUTI0tl,
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G 0
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RELAP4 NODALIZATlGN FOR BE/EM STUDY
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. Parameter Rango Value 4
1.
DLti:nY = cubcooled discharge 0.7 + 1.2 0.9 coefficient'-
2 2.
DL11C:1 a saturated discharac'
-0.25 + 1.0 0.
coefficient 3.
SLIP = slip correlation dial
-1.
+ 1.
O.
4.
DLTP = 2-phace form loss dial 0.4 +.l.G 1.0 4,
DLTPF:1 =.2-phase fanning.
friction loss dial Thece dials are assumed to be
" " ' ~ ~
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equal, and a ningle variable l
5 DCl!F = critical heat flux dial 0.3 + 3.0 1.0 J
G.
D11TCG = Condic-Dongston dial 0.5 + 2.0 1.0 7.
Dl!TC7 = free convection and 0.6 + 1.5 1.0 radiation dial 8.
D!!TC8 = Dittus-Boolter dial 0.5 + 2.0 1.0 9.
D!!TC9 e !!su and Bromlcy-Pomeranz 0.7 + 1.5 1.0 dial
]
10.
DL2LK = flow blockage dial 0.4 + 1.6 1.0 multiplier 11.
DLMWR.= multiplier of Cathcart-0.25 + 1.15 1.0 Pawel reaction rates i
12.
DLPWR = power icvel multiplier 0.94 + 1.06 1.0 13.
DLCPR = increment to be added
-5. + 10. psia O.
to containment pressure
~
14.
DLPUMP = dici for 2-phase pump
-1.
+ 1.
O.
head multiplier 15.
ECCTMP = temperature of accumulator
- 40. + 140 F 900F 0
and safety injection system water 1G.
DLACC = accumulater pressurc 593.2 + 693.2 psia 643.2 psia 17.
TLP = time i$ lifo O + 440 months 226 months j
18 PPU :C = penking f actor uncertainty
.84 + 1.16 1.0 multiplice j
i 19.
DLECON = thermal conductivity dial
.6 + 1.3 1.0 multiplier 20.-
DLCAP = additivo uncertainty in dl.5 mils 0.
r: dial gan site 3
NOO = 0 + frech tuuA
' =
l'+ once burned fuel s.
- 21. 'DLDF.C = decay heat multiplier
.06 + 1.0 0,
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ADVA!!TAGES OF IMPLEMEi1 TAT 10N UNDER JOSHUA o
J03 SUBMISSION FROM CRT TERMINALS o
It!?UT A!!D OUTPUT DATA STORED O!1 DISK AS PERMANENT AUD SEMI-PED 20ENT
- DATA, RESPECTIVELY
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o DATA CATALOGUED IN USER'S LANGUAGE VIA ALPHANUMERIC RECORD ilAMES o
DATA DISPLAYED Of1 If! FORMATIVE TEMPLATES o
DATA ENTERED DIRECTLY INTO DATA BASES o
DATA BASE READlLY INSPECTED AND MODIFIEN I
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IfiPROVED liiPUT FORiiAT o
HAfiED TEi'. PLATED INPUT RECORDS USED INSTEAD OF CARD IMPUT o
PART USED TO DEFINE COLLECTION OF CONNECTED
- VOLUilES, jut;CTI0ilS, SLABS AND OTliER PARTS I
o NAMED COMP 0NEilT USED INSTEAD OF NUMBERED COMPONENTS s
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BASIC EXPERIMENTS f
l WAU_ EAT TRANSFER ECOTRA(GRD0BLE) l
& CC00ENSATI0fl DISPERSEDDROPLETFLG1 REBECA(CADARADE)
CHARACTERISTICS EPRI(DREXED JETIFPACT BATTELLE-FRANKFURT MC POOLSWELLaPACT DEVELOPENTAL SCIBJCES (EPRI)
A;R VENTING-UCLA MIT STEAMVBITING UCLA MIT BREAKFLOW L
f 1
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INTEGRAL TESTS ESI M
COTS BATTELLE/FPANKFURT V4 SCALE BEACON DRYC0iEAltiETT C0FPAE MJLTI-C0FPARTIBIT C00flDFT/4 HDR EAR FULL-SCALE BEACON DRY C0iHAltIOTT COMPT /4 MJLTI-C& PARI N FARVIKENI,II 1/2 SCALE
- BEACON PESSUE SUPPESSICN C0fffEFFT-LI VERTICAL VENTS JAERI V6, FULL-SCALE PESSUPE SUPPESSION VERTICAL VENTS GKSS 1/2 SCALE, FULL-SCAE PESSUE SUPPESSION VERTICAL VENTS PSTF 1/3,1/J 3, FULL-SCALE CON 19FT-LT PESSUE SUPPESSIW ELE-IC HORIZONTAL VENTS
- Willi ESPECT TO NWINAL FULL SCALE PARK II VENT DIAETER i
l h.
w
d 1.
FUNDAMENTAL BASIS FOR SMALL-SCALE MODELING OF POOL SWELL TYPE j
PHENOMENA 1
y SCALING LAWS,2 I
SYSTEMATIC TESTS OF LAWS IN REDUCED-SCALE SYSTEMS I
PHENOMEN0 LOGICAL INVESTIGATION IN REDUCED-SCALE S COMPARISON BETWEEN REDUCED-SCALE SIMULATIONS AND CODE PREDICTIONS:
OUR OWN (SIMPLE) 6
- OTHERS, E.G., NICHOLS & HIRT AT LASL
- QUANTIFICATION OF EFFECTS WHICH MAY CAUSE DEPARTURES FROM SCALING LAWS:
AIR BUBBLES IN POOL '
EXCESSIVE VAPOR PRESSURE '
BREAKTHROUGH
' VENT VOLUMES AND ORIFICING 1.
Anderson,Huber,&Sonin,NUREG/CR-003(1978).
2.
Sonin & Huber, J. Heat Transfer, in print (1978).
3.
Anderson, Huber, & Sonin, J. Heat Transfer, in print (1978).
4.
Work in progress at MIT.
S..Snyder, B.Sc. Thesis, M.E. Dept., MIT, January 1978.
e 6..
Nichols & Hirt, ANS Transactions, 28,, 416 (1978).
7.
Ruggieri, M.S. Thesis, M.E. Dept., MIT, in preparation.
I i
4
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s
(f51)
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2.
FLUID-STRUCTURE INTERACTIONS AND THEIR EFFECT ON POOL 3
BOUNDARY HYDRODYNAMIC LOADS (LOCA, SRV)
SYSTEMATIC"SMALL-SCALE EXPERIMENTS TO GENERATE CONTROLLED F.S.I. DATA FOR CODE CHECKING DESCRIPTION IN SIMPLEST POSSIBLE TERMS DEVELOPMENT AND TESTING OF RATIONAL SCHEME FOR USING 1
HYDRODYNAMIC DATA FROM RIGID-WALL SMALL-SCALE MODELS AS INPUT INTO A SIMPLIFIED, LINEARIZED COMPUTATION FOR THE PERTURBATION CAUSED BY F.S.I.
HENCE, TO DERIVE FLEXIBLE WALL DESIGN LOADS FROM RIGID-WALL SMALL-SCALE SIMULATIONS OF THE HYDRODYNAMICS WORK IS IN PROGRESS l
l r
J IN T
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^
1 1
3.
STEAM CONDENSATION CHUGGING
FIRST-CUT" THEORETICAL MODEL FOR CHUGGING, BASED ON POSTULATED RATE EQUATIONS FOR CON-DENSATION: CRITERIA FOR SMALL-SCALE MODELING.I FUNDAMENTAL EXPERIMENTS ON CHUGGING CONDENSA-TION MECHANISM AND RATE EQUATIONS.
GOAL: SEMI-EMPIRICAL CORRELATION EQUATIONS FOR (TRANSIENT)
CONDENSATION RATE GIORK IN PROGRESS).
FORMULATION OF IMPROVED THEORETICAL MODEL AND IMPROVED SCALING CRITERIA.
1
}
- 1. K0WALCHUK & SONIN, NUREG/CR-0221 (1978).
i h
a:
~
4.
INVESTIGATION OF DISTORTED-GEOMETRY TESTING D-G-T s TESTING IN A SYSTEM WHERE FLOW-WISE LENGTHS (MAINLY VERTICAL) ARE FULL-SCALE, BUT ALL CROSS-SECTIONAL AREAS ARE REDUCED IN PROPOR-TION.
ASSUMPTION: MEASURED LOADS = FULL-SCALE LOADS OUR WORK (IN PROGRESS):
EXPERIMENTAL INVESTIGATION OF HOW WELL THE CONCEPT WORKS, AND WHAT ITS LIMITATIONS ARE TESTS ARE BEING CONDUCTED IN REDUCED-SCALE SYSTEMS SIMULATING POOL SWELL IN MARK 3-LIKE GEOMETRY, WITH AREA RATIOS 1,1//5, AND 1/3 f
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=e w
c
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4 TAYLOR ItlSTABILITIES DURIflG VENT CLEARIflG 1
N s
s 1
s P
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Y s
s,
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7-d s
s
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ACCELERATI0fl ACTS OPPOSITE TO GRAVITY If a p >g AND A
<d, DISTURBANCE W!LL GROW.
c LINEAR THEORY MINIMUM UNSTABLE ilAVELENr,TH A =2nYo/Ca(Pe - pad c
MOST UNSTABLE ~ WAVELENGTH Ad" Ac GROWTH RATE OF MOST iJNSTABLE WAVELENGTH C
- m = O.62 (Pw+Pa).
'J Volume of 4.76 cm I.D. L r
Tank and Valve = 6305 cmjne between l
r B
r 1
a m.
q 16.5 cru Volume of 4.76 cm I.D. Line between Valve and Test Chamber = 2080 cm3 N'
91.4 cm 39.4 cm Ane<nometer
> W
~
Pressure Transducer 22.7 cm N
-133.4 cm Orffice or No Orifice -
- -g:-
h2.7 4cm Pressure Transducer Reservoir Tank Volume =
l 12.7 3
Filled with
.A -- -kC"T-0.255 m p
(- g < f Argon. Helium, Test Chamber:
or Air Sealed Cylinder, 45.7 cm Diameter y
cm g
i Filled with Air, Argon, or Helium 121.9 cm J
[
. Ab, 60
- h. -. / g%
I cm cm
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25 cm
.'8 3
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Pressure Transducer DN Side Pressure Transducer
-t 22 f cm.9 Side N}i M
7 Pressure Transducer /
Cen ter Schematic Diagrcm of the Experimental Apparatus.
~
rc 60 Main Line Flow Rate No Orifice Q
Upstream Reservoir Pressure = 177.2 kPa
~
Ambient Test Chamber Pressure = 74.3 kPa (l) 1 he skc mo<
E 20
( N A ^ ;- #j# * ^
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130 Upstream Press' e
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0 20 40 60 80 100 I?0 140 160 180
.'n0 Time (msec)
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Dubble Fomation and Collapse for Various Pool Temperatures.
l.
POOL TEMPERATURE 139.7 0F 159.3 0F 126.5 0F No. of Bubbles Time Between Chugs, esec.
No. of Bubbles Time Between Chugs, esec.
No. of Bubbles Time 'u-tween Chugs, asec.
Between Chugs Between Chu2s Between Chugs Chug #1
~
4 Dubbles 411.9 3
343.
1 25' Chug #2 4 Bubbles 517.3 2
284 3
476.
4 Bubbles 517.8 2
297 1/2 183.3 s
Chug #4 2 Bubbles 245.8 4,,
649.
2 259.5 Chug 85' 1 Bubble 331.7 5
560.
5 909.5 Chug 86 5 Bubbles 634.3 10 1210.
1 230.9 Chug 87 4 Bubbles 562.
8 879.
3 480.9 Chug #b 4 Bubbles 503.3 9
- 1036, 1
295.2 Chug 29 3 Bubbles
$45.8
__Chur slo 31 Bubbles'for 4274.9 43 Bubbles 17 Dubbles Total 2"97.4 9 Chugs for 8 Chugs 5258.
for 8 Chugs N
Detailed in Step 3.
Detailed in Step 2 (Fig.1).
1 0
m
-~w
m.
f Chugging Characteristics At Various Pool Temperatures POOL TEMPERATURE 126.5 0F 139.7 F
159.3 0F
-1 Average Chugging Frequency sec 2.85 2.1 1.52 Average Bubble Frequency sec~
14.24 13.97 12.1
\\."
Bubble Growth Time (msec) 101.4 77.5 94.3 Bubble Collapse Time (msec) 25.7 7.5 8.6
. Duration of Chug (ms) (Average) 165.7 % 260.7 msec. 203 N 283 msec 169.0 % 202.8 msec.
- Max. licight of Chug -(Downcomer Diameter) 1.75 -
2.6 2.4 1.4 Max. Bubble Frequency sec'I 15.15 21.3 16.7 Min. Bubble Frequency sec"I 7.43 10 9.6 Max. Chug Duration (msec) 260.7 283 169 Min. Chug Duration (msec) 165.7-203 202.8 N
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SOLA SOLUTION ALGORITHM M
e Conservation of momenturn Navier-Stokes equations 0 Conservation of mass incompressible assumption e Solution strategy Solve momentum equation for trial velocities iterate on pressure field until V
V 4; e.
This yields P"+1,u"+1,v"+1 4
N
p GENERAL TREtIDS IN DSS: REACTOR SAFETY
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i VE!! DORS ARE PROPOSli!G APPE!! DIX K MODE TO GAIT! ADVAilTAGE WITli RESPECT TO 2200 F.
STAFF STILL CARRYIf!G BURDE!! ON APPENDIX K
- CHANGE, LITTLE OR l10 EMPHASIS 0?! IMPROVED SYSTEMS SY VENDOR
~l N0 GENERAL EMPHASIS BY VENDORS ON BE CODE-(SOME STAFF I?lTEREST GUICKEllING ON RELIABILITY STUDIES FUEL DESIGt!S.RELATIVELY STATIC - GE GOING
- FUEL, ALL VE!!DDRS DEVELOPING ADVNICED FUEL DE FOR HI'-BlJR'!UP F G. RELEASE, AND TO GAIN ADVANTAGE FR li1 PROVED MODELING),
PEAKIrlG FACTOR DEVELO.PMENTS RATHER QUIET, GENERAL INTEREST IN ACCELERATING OUR INVOLV SYSTEf! TRAflSIElT ANALYSIS, LITTLE PROGRESS 0?! GENERIC A ITEMS A-3, 4, 5 (SG TUB AtlD A-22 (MSLB) ),
MODEST PROGRESS Oil A-1 (WATER HAMI'ER) (
REVIEW),
GOOD PROGRESS ON A-2 (SU3C00 LED LOADS),
COMPLETED RHR (A-31),
ATNS MAKING PROGRESS (A-9).
BWR CORE SPRAY MAKING GOOD PROGRESS (A-16),
s PLAtlNING CONTINUATIO!! 0F TA INVOLVEMEN (FUELS, PHYSICS, THERMAL-HYDRAULICS),
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REClRCU LATION TESTS PLANT TEST-l.
Trojan in-Plant Salem In-Plant Diablo Canyon in-Plant Beaver Valley in-Plant North Anna In-Plant and 1/3 Scale Three Mile Island 2
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1/3 Scale p
Davis Besse 1/2 Sca'le Farley Unit 1 Full Scale Farley Unit 2 Full Scale (One Sump)
Arkansas Unit 2 Full Scale (One Sump)
Sequoyah 1/4 Scale McGuire 1/3 Scale D.C. Cook 1/2.5 Scale
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c EG&G FAILURE MODES AND LIKELIHOOD Relative Failure Component Failure Mode Likelihood
- 1. Pumps A. Failure To Start 1
B. Failure To Pressurize 1
C. External Leakage 3
- 2. Valves A. Failure To Actuate Power Operated 1
B. Inadvertent Actuation Valve Dependent 1,2,3 C. External Leak 3
- 3. Lines A. Blockage Dependent On Size & Route 1,2,3 B. External Leakage 3
- 4. Screens And Filters A. Blockage 1
B. Rupture 1
- 5. Tanks And Accumulators A. External Leakage 3
[d
- 6. Orifices A. Blockage 1
B. External Leakage 3
b
e FMEA Studies FY-75 A. GESSAR (GE) ECCS Single Failure Analysis FY-76 A.
CESSAR (CE) ECCS Single Failure Analysis B.
RESAPs 41 (W) ECCS Single Failure Analysis C. Trojan (W) ECCS Single Failure Analysis D. Pebble Springs (BEtW) ECCS Single Failure Analysis E.
GESSAR 4GE) RHR Single Failure Analysis 5
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PROBABILtTY PROBABILITY REVISIONS TO "9
with revisions OPER ATING PR' CEDURES CONF GU TION 4
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f la 9.5x10-6 4.7 x 10- 7 Leak Test Every Two Years i
2.4 x 10- 7 Leak Test Every Year i
1h 1.8x 10-4 4.2x 10- 9 Lock Valves Closed and Leak Test Every Two Years 1c 1.0 x 10-- 5 2.5 x 10 -7 Leak Test Every Year 1d 3.0x 10-- 6 7.4 x 10- 9 Leak Test Every Two Years le 2.8x10-4 4.7 x 10'- 7 Leak Test Every Two years 2.4 x 10- 7 Leak Test Every Year 1f 9.Gx10-6 4.7x 10-7 Leak Test Every Two Years 2.4 x'10-- 7 Leak Test Every Year
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RHR 4
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.I Cold Leg 4
2 Check Vcives 3.8x10-5 9.6 x 10 -7 Hot Leg 6
2 Check, Closed Motor 1.7x 10-3 1.4 x 10- 6 Upper Head Injection 4
2 Check 3.8x10-5 9.6x10- 7 Total Probability 1.8x10-3 4.3x10- 6
- Leak Testing Frequency of Once Per Year I
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LOW SAFETY HEAD INJECTION PUMP P
BEFORE MODIFICATIONS:
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Maximum Bearing Wear Run Time 235 MILS 3 HRS.
AFTER MODIFICATIONS:
Maximum Bearing Wear Run Time 3 MILS 550 HRS.
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m OUTSIDE RECIRCULATION SPRAY PUMP BEFORE MODIFICATIONS:
Maximum Bearing Wear Run Time 100 MILS 47 H RS.
AFTER MODIFICATIONS:
Maximum Bearing Wear Run Time 3 MILS 450 HRS.
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ALL GE #1SWERS SEPTEEER 1,1978 STAFF REVIB1 A. ODYN CO E OCTOBER 1,1978 B, ODYN COE QUALIFICATION CCT0EER1,1978 C. AUDIT CALCULATIONS SEPTEBER20,1978 D. STATISTICAL STUDY 0-1 SEPTEBER 20.,1978 l
E NISWERS TO STATISTICAL STUDY CCTOBER 15,1978 GE TOPICAL REPORTS OCTOBER 15,1978 STAFF PE/IB4 CCFFLETED AND SER ISSUED NOVEIBER 30,1978
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CRAFT 2 Computer Program BAW-10132P March 1977 i
l Topical Report Evaluation in DRAFT i
l Scheduled for September 15, 19 78 i
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HINDLE'S TEST CONDITIONS PROMOTED UNIFORM CLADDING TEMPERATURES THAT WILL NOT EXIST lil-REACTOR FOR THE FOLLOWING REASONS.
(1)
IN HINDLE'S CREEP-RUPTURE TESTS A ZERO HEAT FLUX ACROSS THE PELLET-TO-CLADDING GAP EXISTED.
NUCLEAR HEATED FUEL RODS WILL ALWAYS HAVE A fl0N-ZERO HEAT FLUX ACROSS THE GAP.
(2)
RADIATION HEAT TRANSFER TO COLD WALLS IN HINDLE'S SINGLE R0D TESTS STABILIZED BALLOONING PROCESS.
(31 0FF-CENTER FUEL PELLETS, WHICH PROMOTE CIRCUMFEREilTIAL 1
AND AXIAL VARIATIONS IN GAP CONDUCTANCE, PRODUCES TEMPERATURE NONUNIFORMITIES THAT LEAD TO UNSTABLE 4
AND LOCALIZED RUPTURES, (10 FLECHT TEMPERATURE MEASUREMENTS SHOMED VARIATIONS MORE THAN SUFFICIENT TO LOCALIZE RUPTURES.
(5)
PELLET ENRICHMENT, DENSITY, AND GEOMETRICAL VARIATIONS CREATE LOCAL TEMPERATURE VARIATIONS THAT ARE ADEGUATE TO CAUSE LOCALIZED RUPTURES.
I (6)
GERMAN IN-PILE RUPTURED FUEL RODS EXHIBIT NO " SAUSAGE" BALLOONS.
(7)
ORNL AND GERMAN USE OF INTERNAL-C0tlDUCTION HEATERS TO CREEP RUPTURE SPECIMENS RESULTED IN NO " SAUSAGE" BALLOONS.
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HISTORICAL DEVELOPMENT OF FUEL PERFORMANCE CODES NRR CODES RES CODES l
GAPCON-THERMAL-1 FRAP-S1 (1973)
(1975)
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(1976) l GAPCON-THERMAL-3 FRAP-S3 (1977)
(1977)
NRR/RES COOPERATIVE AGREEMENT (1977)
FRAPCON-1 I
(AUGUST 1978) 6 a
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HAS THE ENDORSEMENT OF BOTH INEL AND PNL (3)
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