ML20052C830
| ML20052C830 | |
| Person / Time | |
|---|---|
| Issue date: | 01/04/1982 |
| From: | Catton I Advisory Committee on Reactor Safeguards |
| To: | Boehnert P Advisory Committee on Reactor Safeguards |
| References | |
| RTR-NUREG-CP-0024, RTR-NUREG-CP-24 ACRS-CT-1407, NUDOCS 8205050605 | |
| Download: ML20052C830 (7) | |
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9th Water Reactor Safety Research Infomation Meeting 9 48 My attendance at the 9th Water Reactor Safety Research Infomation Wetting was highlighted by being given a car belonging to the Chaiman of the Commission by mistake.
It was an Oldsmobile diesel that did not run very well.
2 The lessons of TMI were very evident in that very little is left of large break LOCA studies. Operational transients and small breaks are the theme.
The code development program looks like it is loosing momentum as most of the job is done.
The following paragraphs contain some notes on the sessions I attended. I can go into more detail on any of the research programs mentioned if needed.
INTEGRAL SYSTEMS EXPERIMENTS i
LOFT. LOFT tests L3-5 and L3-6 (Pumps on/off) lead to some interesting resul ts.
It was found that pump current allows one to clearly distinguish 1
what type of accident is underway. The LOFT personnel believe that a pump on/off criteria can be based on pump current. They recomend that vessel liquid level measurement not be required when the pumps are on.
Dr. Linebarger supports the use of manual trip providing the operators are properly trained even though there is only 1.3 minutes for action. He believes that the penalties associated with " pumps off" when they should be on are great enough (no reason given) to cover the increased risk of a delayed trip (t 1.3 min.).
It was noted by Dr. Zudans that computer software could advise the operator very quickly and the operator could then take manual action if he chose to do so.
Multiple failure LOFW transient results (L9-1/L3-3) demonstrated that predictions grossly overestimate mass loss. Even so, realistic calculations are needed for operator training, procedure development, technical specifications and development of simulators.
It is not possible to tell what is conservative if actions based on predictions are necessary. Overprediction of mass loss may lead to a more risky action.
(Of course, an inventory based view, rather than a calculational view would avoid this problem). It was disappointing to see the best available calculational tools do so poorly. The problem was in the secondary side modeling. Heat transfer to the secondary side was under predicted and small heat transfer effects cause large differences in the loss of water from the system. The fine details of themal modeling are very important in slow transients. This means a lot of nodes. Older codes could 8205050605 820104 ggs DE3ICnTD ORIGIRL Certified By M
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Catton to:BSehnert 1/4/82 page 2 not handle such detail and even RELAP5 needed simplifications. Care is needed in what these simplifications are as improper simplifications lead to large differences in the predicted results. One wonders how much faith to place in the vendor codes and predicted reactor system behavior.
LOFT experiment L6-7 (simulated loss of lead compounded by ar, uncontrolled secondary side steam leakage) data support the conclusion i
drawn from LOFT L3-6 data (pumps on/off) that pump current can be used to differentiate between transients that result in fluid density decreases (LOCAs) and those which result in fluid density increases (cooldown transients). The sensitivity of the LOFT pump current to coolant density variation is encouraging. Whether or not full scale PWR pumps are as l
sensitive should be determined. Sensitivity to coolant density variation could be of value in helping the operator decide whether or not he should 1
turn off the pumps and also give some anticipatory evidence of an impending ICC situation.
Further, with pump RPM, estimates of cctual mass flor are possible.
RELAP5 predictions compared quite favorably with the data from L6-7 The L9-2 transient (pump trip at the end of L6-7) which was to provide data on the effects of voids and themal stratification on natural circulation, was not as well predicted. This was believed to be due to poor boundary conditions rather than code inadequacies.
t Intermediate break experiments (L5-1 and L8-2) yielded results that were relatively unexciting.
RELAp5 predicted the observed phenomena very well. There were no surprises. The LOFT ECC systems were very effective 3
at mitigating the consequences. The implication being that PWR ECC systems designed for large break LOCAs will successfully terminate an intermediate break LOCA.
Conclusions from the LOFT tests relevant to licensing were presented i
by Ralph Landry of RES.
He noted that the RETRAN code (EPRI) was found to 1
be inadequate in its LOFT application. This will make EPRI unhappy. The importance of details in thermal modeling were emphasized. The LOFT 1
experiments have shown that a highly trained Technical Support Center (TSC) staff is very helpful.
An evaluation of the LOFT instrumentation, which is similar to PWR process instrumentation, showed that it was adequate i
and that it would be helpful if fuel rod temperatures and density in the l
piping were available, j
l The results of a lenghty study of external thermocouples show that external thermal couples cause rapid quench but that gap conductance tends i
to isolate the fuel so that fuel temperature is relatively unaffected.
This is to be contrasted with a Semiscale type solid rod where inner temperatures are highly impacted.
SEMISCALE. Results of the Semiscale MOD-2A UHI test series were presented. Overall, UHI was found to have little impact on the transient l
behavior other than to provide increased core liquid iventory over that observed in baseline experiments. A greater amount of inventory depletion i
Catton to Boehnert 1/4/82 page 3 was observed for some UHI tests. More water helps but the lower ECC accumulator set point used with UHI counters some of the benefits by allowing greater coolant depletion before the system is turned around.
Natural circulation tests demonstrated that core cooling by reflux boiling was adequate for inventories greater than 50%. The secondary side was found to give adequate cooling as long as the level was above 8%. There was, however, some pressure rise for secondary side level below 50%.
An injection of 10% of the steam generator volume of N2 changed the reflux boiling by causing carryover to cease.
Larger amounts of N2 causrd the system pressure to rise but had little effect on the reflux rate C r cooling.
ROSA-III BWR LOCA/ECCS.
ROSA III is a 1/424 volumetrically scaled single channel half length facility.
Five experiments have been conducted with breaks in the recirculation line varying from 2% to 100% with the HPCS locked out and the pressure control system assumed to be inoperative af ter blowdown.
Over the entire range of break size, no new phenomena were observed. Only the time scale was different. The maximum PCT measured was 922'K for the 50% break.
ROSA-IV pWR SBLOCA.
ROSA IV simulates a 3423 MWt PWR with 17 x 17 fuel with a volumetric scale ratio of 1/48 and is full height. The maximum core power if 10MW which provides the equivalent of 14% of full power. The facility has two loops, a simulated secondary side,1080 simulated fuel pins and is capable of full temperature at full pressure.
It will be some time before experimentation begins.
HITACHI BWR INTEGRAL FACILITY. The facility consists of two full sized electrically heated bundles capable of SMW each (full power). There are two recirculation loops with two jet pumps. The height to the separators is full scale. The height of the steam dome and the lower plenum are less than full scale. The upper tie plate is an actual tie plate from a BWR/5-251 to insure that CCFL is properly simulated. Some results were presented for simuletion of a DBA.
It was found that there was little i
difference between bundles during blowdown even when they were differentially heated (6/4).
Initiation of heat up was later in the higher power bundle due to higher water entrainment. Because the higher power bundle had better cooling it did not always have the highest PCT. The highest observed l
PCT was 861'K.
LOBI. The test facility and experimental program were described with particular emphasis on the test facility design. A total of 90 tests have been run during the past two years. The tests have formed the basis for an extensive pre-and post-test prediction evaluation. The codes (TRAC, DURFAN and RELAP4) were found to have only limited ability to describe the complex phenomena.
It was of interest to note that the so-called
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Catton to Boehnert 1/4/82 page 4 advanced codes did not a ptiori give better results than simpler lumped parameter codes like RELAP4 The large differences between results of different participants using the same code indicate the importance of user experience in the application of the code. The quality of the predicted result depends strongly on the code input data including the nodalization of the system, the selection of code options and the establishing of proper initial and boundary conditions. This requires an experienced code user who is cognizant of the capabilities, limitations and idiosyncracies of the code he is using.
This only comes with a great deal of use of a given code.
PKL. The facility is a Europena Semiscale without full pressure capabTTTty.
It represents a KWU 1300 MW, 4-Loop PWR at a volumetric scale of 1/134 capable of 40 bars with full h6fght components and a fairly complete secondary side. The core consists of 340 electrically heated rods.
It has three loops, one of double capacity, each having an active steam generator.
Energy and mass flow through the system were shown for LBLOCAs.
Such plots are very informative and should be required of all such facilities In particular, the energy transport during a SBLOCA clearly shows the role played by the steam generator and the break flow in achieving mitigation.
It shows where the energy to be removed comes from and where it goes.
One could see immediately that the largest share of energy to be removed is energy stored in the system (fluid and structure). A new facility is being constructed PKL II.
Results from it will yield more detail rather than anything new.
SEPARATE EFFECTS NRC's BWR Safety Research. The information presented on NRC's BWR Safety Research was basically the same as that presented at the BDHT Review Group Meeting. The reader is referred to my report on that meeting.
(Report to Paul Boehnert dated 9 Nov 1981).
i A paper by Dr. Lee of KAERI investigated precursory cooling during reflood of a hot vertical rod. His experimental apparatus was a single rod of the ORNL type in a pyrex tube. He foun that the measured heat transfer coefficient ranged from 60 to 300 W/Mg'C.Precursory cooling was found to inc of about W/Mgease with decreasing subcooling and to typically have a value
'C which is quite a bit higher than the usual steam cooling.
ORNL THTF bundle heat transfer studies now include in-bundle void I
fraction measurements. The void measurements show a rather sharp transition from a = 0.5 to a = 1.
Drift flux models do not predict this and as a result typically overpredict froth height (overpredict the portion of the core being cooled). Comparison of heat transfer coefficients for high temperature t
l steam cooling with data show Dougall-Rohsenow to be excessively high.
Catton to Boehnert 1/4/82 Pe9e 5 a
Groeneveld5-7 to be pretty good and Dittus-Boelter to be excessively high.
Some EM models use the Dougall-Rohsenhow or Dittus-Boelter correlations and should be changed.
Ft.ECTH-SEASET seems to be alive and well. The W well instrumented steam generator is now being used to study single and two phase natural circulation including refluxing with N2 effects. They are looking at the impact of secondary side perturbations on various stable cooling modes.
Flow blockage studies continue with both forced and gravity flow and coplanor as well as non-coplanor blockage.
l The B&W Post-CHF data analysis was not very exciting.
It was shown I
that one needs to include droplets in the flow and thermal entrance effects l
to obtain good predictions.
Some interesting studies of the basic physics of two phase flows are l
underway at ANL. The focus is on interfacial transport mechanisms and interfacial motion (interfacial means the steam-water interface). This leads to the need to describe interfacial area concentation (or density).
Some nice measurements of interfacial area per unit volume and the relationship to void fraction were reported.
RPI work on parallel channel effects in BWRs during a LOCA were reported. The results confirm the Hitachi resu'its showing good cooling in the high power bundle.
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Direct contact condensation of steam flowing over cold water is being studied by Dr. Bankoff at NWU.
l ADVANCED INSTRUMENTATION An improved method of measuring clad temperature has been developed for use at the LOFT facility. The fuel pellets are grooved for the thermocouple leads and near the thermocouple junction the leads are imbedded i
in a slot on the inner surface of the clad.
Five fuel rods will have these l
thermocouples for February 1982 testing (better late than never). The new turbine meters measure velocities with reasonable accuracy down to 0.3 M/sec (natural circulation).
l ANL work on two-phase flow measurements using pulsed neutron activation (PNA) techniques continues.
It was shown that mass flow rates can be measured to +2 or 3% in small pipes. PNA techniques are most accurate at low density where other methods give trouble. To date I am unaware of any substantive application of PNA methods.
ORNL has studied the use of an ultrasonic ribbon for reactor liquid level detection. Signals are bounced off a notch in a ribbon. Transit time is a function of temperature and the surrounding fluid density.
It was not clear what a two phase mixture would do the echo. The device yields both density and temperature profiles.
It looks promising.
The vessel penetration may, however, give difficulties.
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Catton to Boehnert 1/4/82 page 6 4
A non-invasive liquid level and density gauge for reactors is being developed by Prof. Jester of Penn State. The approach is to use the fast neutrons leaking through the reactor vessel wall into the air gap surrounding the vessel by placing a series of fission detectors in existing instrument wells external to the pressure vessel. Some simple experiments show promise.
If combined with another indicator to eliminate ambiguity, it could be very effective.
A method of determining non-condensible gas concentrations using wet and dry bulb temperature measurements was investigated by Prof. Griffith j
at MIT.
His results were rather inconclusive.
Two phase performance charactersdtics of the LOBI pump were given by Dr. Piplies of ISPRA. He presented a nice collection of data. A strong degradation is shown for void fractions greater than 0.25.
The homologous curves work very well even for two phase flow.
Studies of containment emergency sumps by SANDIA in response to USI A-43 had a nice conclusion: really no problem.
ANALYSIS DEVELOPMENT RELAP5 has progressed to the point where it is a very good tool.
Techniques to allow the code to run in a quasi-steady or steady state mode will soon be incorporated to give much faster running capability.
Radiation heat transfer and variable area flow passages are being added as a first step towards degraded core calculational capability.
TRAC-BWR code work is concentrating on BWR hardware representation.
It is a cooperative effort between NRC and GE. The development seems to l
be moving right along.
Certain aspects of BWR modeling that require multi-dimensional representation are available in TRAC. Lots of fancy user conveniences are being incorporated into the code. The documentation is the best of any of the code programs.
TRAC-PFI looks good enough to be put to use.
It's not clear what further developmental effort will yield. LASL has done a good job.
Modeling of H2 migration in LWR containments is being done at LASL.
It appears as if this has the makings of another code development program.
A code is being set up for three components: air, H2 and steam.
Diffusion terms are to be in the transport equations. This means a turbulent transport model is needed.
Condensation both at boundaries as well as in the containment volume itself must be considered. This problem is not unlike obtaining a cumulus cloud development d*scription. It is a much more difficult problem than LASL is anticipating.
s Catton to Boehnert 1/4/82 page 7 COBRA / TRAC and COBRA-TF are still developing.
It is interesting to note how COBRA has evolved from a sub-channel code to a containment code (not unlike COMMIX at ANL). The COBRA / TRAC has a good reflood model although top down quench still needs some work. The steam generator l
model is also weak and in need of some improvement. COBRA-TF is being used as the basis for a containment code (see previous paragraph).
Preliminary results look okay.
l An interesting example of FSI was studic
- by Dr. Lee of KAERI. He studied fuel assembly failure due to impact of flow through gaps between baf fle plates on the flow. Some Korena PWRs found fuel clad completely l
broken. Through similitude and analysis of non-linear oscillations, Dr. Lee arrived at criteria for gap sparing and fuel pin location to t
eliminate the problem.
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