ML19210C732
| ML19210C732 | |
| Person / Time | |
|---|---|
| Issue date: | 09/27/1979 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1649, NUDOCS 7911200016 | |
| Download: ML19210C732 (27) | |
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4p, t MINUTES CF TH.I ACRS SUSC2MITTEE MEETING CN
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DdIAGENCY CCRE COCLIE SYSTEMS JUNE 19-20, 1979 WASHING 7.N, D.C.
The ACRS Subcommittee on.rCCS held a meeting en June 19-20, 1979 in Recm 1046, 1717 H Street, N.W., Washington, D.C.
.Be purpose of this meetf g was to review the small break analysis perfor ned with the B&W model sub-sequent to the Three Mile Island 2 accident and the NRC Staff review of the B&W w rk.
te second day of the meeting was devoted to a review of the proposed FY 1981 tudget for ECCS related Safety Research and the proposed Supplerantal to de 1980 budget. te notiet for the meeting appeared in the Federal Register en Monday, June 4, 1979. A copy of the notice is included as Attachment A.
A list of meeting attendees and a meeting schedule are included as Attachments 3 and C.
No written statements or requests for time to make cral statements -ere received frca members of the public.
Executive Sessien - Ocen Dr. Plesset, Subcommittee Chairman, opened the meeting at 3:40 and indicated that it was being cenducted in accordance with the Federal Advisory Committee Act and the Gover. ment in the Sunshine Act.
Cr.
A. Sates was de Cesignated Federal Employee for the meeting.
Curing a short executive session the Subccmmittee members and censultants discussed the topics to be reviewed at de meeting.
Dr. Plesset indicated d at de Research Sudget should be examined to determine de appropriateness of de items being funded in FY 1981 and requested in de supplemental 1980 budget. Or. Catten indicated dat he had scme questiens on the B&W :mdels used in de steam generator and in the hot leg as dey relate to hea:
transfer, bubble growth, and natural circulatien.
1371 224 7911200 0 1 6
ECCS Meeting June 19-20, 1979 Meeting with the NRC Staff Dr. Ros toczy, NRC, indicated that subsequent to the Three Mile Island accident the Staff met with B&W, Westinghouse, and CE a netter of times to review small break calculations. The area of concentration has been on very small breaks and natural circulation cooling. Also reviewed were the vendor guidelines for emergency procedures and operator retraining programs. In respense to a question, Dr. Ros toczy indicated t.%t all of the Appendix K conservatisrs apply to the small break analysis, the 1.2 ANS decay heat value is the most significant of the conservatisms.
Dr. Ros toc:y indicated that there were some~ differences in opinion regarding the advantages and disadvantages of tripping the reactor coolant pumps following initiation of de LDCA. Discussions are continuing en this point.
The other change made to the operating reactors has been the provision for reactor trip prior to actuation of the relief valves for the varicus transients which produce over pressure events. It is not believed that this will cause any problems with system operation, 'd and CE already run their plants this way, S&W plants have reversed their PCRV and Trip setpoints.
Dr. Zudars questioned the wisdom of replacing 148 PCRV actuations with 148 reactor trips; the mechanical effects on the primary system ray be more undesirable for the reactor trips. Dr. Ros: toc y indicated that efforts were being rade to reduce the nurter of reactor trips also.
Mr. Michelson suggested that ene alternative to de PCRV set point changes might be the automatic closure of de down stream block valve follcwing actuation of the FCRV.
Mr. B. Sheron reported on the NRC review of the B&W calculation of small breaks and the capability of natural circulation heat removal (see Attachment D).
The concerns arose from a report prepared by Mr. C. Michelson, T7A, on very, small breaks in B&W reactors where repressurizatien of the primary system may occur if natural circulation cooling is lest or the steam generators are not available. B&W respended to these concerns and others in a topical 1371 225
o ECCS Meetirg June 19-20, 1979 Report to the NRC on May-7,1979.
In the B&W plants very small breaks may not provide adequate enthalpy flux at the break to cope with core de-cay heat. If the sytem depressurizes far enough to produce a steam bubble in the primary system, natural circulation cooling to the steam generators will be lost. te system will then repressurize until the enthalpy flux at the break provides adequate heat removal, or natural recirculation is restored. Similar effects may be produced in U-tube steam generators.
Mr. R. Jones, 3&W, indicated that the higher elevation of cold water in the raised locp plant provides a greater potential to reestaclish de natural circulaticn. Se lowered icop plants go through one cycle of repressurization and water solid natural circulation, the following depressurization ends up in a reflux boiling mcde with condensation of steam in the steam generators when the secondary side water level exceeds de primary side water level. We raised loop plants go through three cycles of repressurization. Se calculations are sensitive to how the hot leg and steam generator are modeled. Se bubble rise, the separation of steem and water, and the interface levels in be steam generator are very important in calculating heat transfer.
The actual heat transfer coefficients in the steam generator are not as important as are 2e water and steam elevation and interface areas.
Dr. Catten indicated some concern over be models used for de 3&W Steam generators.
Dr. Ros:tec:y indicated that he believed they were adequate.
Dr. 7udans suggested that de several instances of small breaks in actual reactors provided a goed source of information for bencrnarxing calculations of small breaks.
Dr. Resztec:y indicated that de 34I-2 accident was probably be only event dat was adequate for doing detailed calculations, other small LTA's have occurred at low power or during start-up. Also not all of the desirable infocnation is available.
1371 226
ECCS Meeting June 19-20, 1979 Drs. Catton and Theofanous indicated that they believed that some thought and hand calculations like those contained in the Michelsen report would be more adequate dan large codes for analyzing some small break situations.
Dr. Catton indicated that with the big codes one may locse sight of important effects which are buried in the mass of computer calculations. Small changes in the model and nedalization might have significant effects; if time is devoted to thinking abcut the processes and doing some hand calculations insights unavailable frem computer codes may be developed.
A number of questions arcse as to de use of a volume balance for the calculations for B&W small breaks. Both mass and enthalpy balances are needed to determine losses from the primary sytem and possible core uncovery.
Mr. Sheren indicated that de calculations shcw dat a condensing surface is reestablished in de steam generator prior to any core uncoveq. At the Davis-Besse plant the high head safety injection pumps shut off (:ero discharge) at 1600 PSIA. Under these conditions repressurizatien could prevent additional HPIS water frem being added. Se make uc pumps can provide some water above 1600 PSIA but they are not safety grade. Auxiliary feedwater is required in ceder to depressuri:e 2e sytem belew 1600 PSIA and allcw operation of the FPIS pumps.
The Staff agreed that pressurizer level indication is not a good indication of system water inventory for certain pressuri:er breaks. Sey expect to see de pressuri:er drain for cold leg breaks.
Mr. Michelsen and others cautiened the Staff that detailed analysis of de pressuri er heat 1csses are necessary to determine if it will drain for all small breaks.
Small changes in temperature are important; if the upper head of de vessel or the hot leg are several degrees hotter than the pressuri:er it will' not drain due to de loco seal.
1371 227
ECCS Meeting June 19-20, 1979 Mr. Sheren reported that they reviewed the possible system effects of the operator isolating a break during depressurization. Se Staff have cen-cluded that the system is adequate for this type of situation provided that make up flow or HPIS is provided to depressurize the system ard auxi-liary feedwater is available. Various opertor actions could cause problems.
The procedures should provide adequate guidance en what to do. Negative guidance (i.e. prohibitions of certain actien) are generally not included in procedures.
The Staff indicated that they have locked at tests of ECC injections in LCET and Semiscale for water harrer-pressure escillation type effects and have not seen any indication of potential damaging pressures. REIAP calculations shew oscillations which have been removed with centrol of the water packing problems that the code had.
A review of sources of non-cendensable gases indicates that the primary system would have to depressurize to abcut 400 PSIA before enough hydorgen and nitrogen would ccme out of solution to fill the hot leg U bends. Abcut 190 cubic feet of gas are involved in order to fill the U-bends. Be CE and W plants require similar volumes to fill the U-tubes in the steam generators. For very small breaks in the B&W plants the Staff dces not sae the accumulators emptying and there shculd be no nitrogen injection.
There should not be problems with neri-condensables in the hot legs. Bis is provided that the core is not uncovered (as in 7.I-2) and there is no metal-water reaction.
Mr. Michelsen cautioned that under scme circumstances cperator actions to isolate a break could produce undersirable censequences such as injection of non-condensable gases, repressurization of the primary system, cere uncovery, or interruption of natural circulation.
1371 228
ECCS Meeting June 19-20, 1979 Mr. R. Audette, NRC, reviewed the results of the small break calculations performed by B&W following the BI-2 accident. We recent calculatiens 2
2 were preformed for break si::es between.07 ft and.0005 ft with various combinations of auxiliary feedwater, no auxillary feedwater, HPIS flew with one or two pe=ps going and different initiation times, and with and 'without the primary coolant p.nnps running. Results show that for 2
breaks si::es below 0.02 ft repressurization will occur.
In one case, with no auxiliary feedwater, with 1.2 ANS decay heat, a stuck open PCRV,1 HPI pump, and the RCPs off the core was shown to uncover; if 1.0 ANS was used it did not uncover. We Staff concluded that if auxiliarf feedwater was provided at 20 minutes core uncovery would be prevented for both 2
raised and levered Icep plants for breaks smaller dan.02 ft. Cther conclusions are shown in Attachment E-1.
Ouestions revealed that the pump heat added to the primarf system was not acceented for in de analysis. Bis may have an effect as it is comparable to the decay heat after extended periods of time.
Mr. N. Lauben reviewed the audit calculation the NRC Staff had performed for the B&W small break LCCA. Tso cases were calculated by EC&G using RELAP 4/
2 MCD 7.
D e first case was a 0.01 ft break with Auxiliarf Feedwater (AN) delayed 20 minutes, the second case assumed normal AEW and one HPI pump.
In de first case the HPI actuation point is reached and repressurization occurs, the second case produces voids in the candy cane of the hot leg and less of natural circulation occurs. De Staff concluded that T6.IAP and Craft produce differences in key varibles cat are calculated and that additional study is needed. Core uncovery was not calculated with either code.
Mr. B. Wilson, NRC Staff, reviewed the new S&W Guidelines given to de utilities for the operator procedures for coping with a small break (Attachment F). Two items were discused, the first was the development of de precedures by B&W and de Staff review of dem, the second are related to the utility implementa-tien of de procedures for their operators.
1371 229
ECCS Meeting June 19-20, 1979 h e guidelines consisted of symptems and indications of small breaks,-
immediate action to be carried out, precautions, and followup actions.
Cne major item was to provide criteria undershich FPI flow could be terminated. Bis was based upon maintaining hot and cold leg temperatures 50 F below saturation or having the ISIS operating at 1000 gpm for 20 minutes. Recommendations for action to follow given reactor coolant pump operation or tripping and availability of auxiliary feedwater were also made. A number of censultants recommended that the sir:ulator training of operators be reviewed to observe de mistakes made by operators, tese than could be used to augment de training program arx! peccedures as Sings one does not do.
Mr. Wilson indicated dat de staff is looking at de human factor in control rocm engineering and dere may be future NRC recuirements or re-commendations in this areas in de future.
Mr. 3. Wilson reviewed the NRC Cbjectives used when reviewing utility procedures. Se intent is to assure conformance with vendor guidelines and workability for the Operators. W e problem areas identified when reviewing precedures included lack of adequate knowledge of small break phencmenen, utility exceptions to the guidelines, and de adaptien of the new precedures to existing precedures.
A number of questiens arcse as to how an cperator determines tether these is a break and how big it is. Se anx:unt of make up flew and fiPIS flow as well as system de-pressurization are important criteria in 2e determination.
Mr. 3. Bogar reviewed the operator training that has been carried cut since TMI for B&W plants.
It has included review of the "MI-2 accident, 24I-2 simulater training, formal classreem training, written exams, an NRC audit of the trainina, folicw uo training and requalificaticn training.
1371 230
ECCS Meeting June 19-20, 1979 Facility changes and procedural changes have been incorporated in the training as well as review of the auxiliary / emergency feedwater system operation.
Subecmittee members and censultants suggested that when grading the tests consideration should be given to the sericusness of the mistake when grading - not knowing an answer may not be as serious as doing the wrong thing - especially when procedures can be looked up. Also discussed was the information available to the operator and how they are able to decide the exact plant situation so as to respond properly.
During a sumary session of the meeting various censultants expressed their epinion in the need for additional work. Several felt that additional thought is needed to assure that there are not other accident sequences and operator mistakes that could lead to T4I-type situations.
24I-2 is being thcughly studied, but other possible accidents should also be looked at just as thoroughly.
The meeting was recessed at 5:45 p.m. to recenvene the folicwing day.
1371 231
ECCS Mtg June 19-20, 1979 Sumary of the Meeting June 20, 1979 T0cic Presentation Transcrict Paces 1.
Cpening Ccmments M. Plesset 262-265 2.
Introduction T. Murley 265-305 3.
Separate Effects Research 305-375 a.
SEMISCALE A. Serki:
305-331 b.
Slowdown /Reflood Heat Transfer A. Serki:
331-349 c.
20/3D Program G. Bennett 349-362 d.
tdel Develegrant Experiments and Technical Supcort A. Serki:
362-375 4.
LCFT Pesearch D. McPhersen 376-396 5.
Analysis CevelcErant Branch S. Fabic 396-441 6.
Closing Coments M. Plesset 441-462 3371 232
ECCS Mtg June 19-20,1979 ECCS Meeting Wednesday, June 20, 1979 Ocenino Comments - M. Plesset (Transcript pages 262-265)
The House Ccemittee preparing the 52dget for the NRC has proposed three changes to increase the utility of the ACES report.
1.
te ACRS should prepare its report in accordance with a schedule that permits it to be used by the NRC in preparation of the fiscal year 1981 authorizatien request.
2.
te ACRS should prepare a clear statement of research priorities including specification of projects to be added or dropped.
3.
te ACRS should include a discussion of the specific manner in which the NRC's reactor safety research projects are expected to affect the NRC's reactor regulations.
Introduction - T. Murley, NRC (Transcript pages 265-305)
S e FY 80 supplement for safety research budget was presea.ted. Bis budget (S29.3 million) is for:
1.
better understanding of transient and small LOCA accidents (13.4 million) 2.
enhanced operator capability (3.9 million) 3.
plant respense under accident conditions (5.1 millien) 4.
p:st mortem examination and plant recevery at NI (2.7 million) 5.
improved risk assessment (3.1 million), and 6.
i.: proved reactor safety (1.7 millien).
(A breakdown of each of de above six categories is presented in de meeting handcuts.)
Se budget for the LOCA-ECCS program is scheduled to rise and then peak in 1981,as scenarios that go beycnd design basis accidents are censidered.
Current plans are to upgrade SEMISCALE by adding a secondary system. It is tentatively planned to upgrade the Two-locp test apparatus (TLTA) facility to study EWR transients and small LCC)s.
1371 233
0FFICIAL USE ONLY ECOS Mtg June 19-20, 1979
~. f s planned that a data bank be established for each operating reactor that would include applicable computer codes which could be employed for calculations following an accident or event at a plant. B is probably will 1:e done at National Labs.
An important area is enhanced opeator capability.
De diagnostic system used at F.alden (Norway) ceuid be installed at' LCFT. B is diagnostic system is being considered for the Groven-Rhinefeld reactor in Bavaria.
The auxiliary feedwater pumps are not considered engineered safety features (ESFs). A study of ESFs should consider heat removal without the avail-
~
ability of the turbine condenser after the reactor scram.
Dr. Catton requested furder infor.ation on the fundi.g for the better understanding of transients and small break LOCA events.
Iters discussed in addition to the detailed breakdown of this area (previded in de attach-ments) were: re-examination of RmN, secordary system treatment for TRAC, study of the CE code IRT, addition of a noncondensable gas model to varicus codes, incorporation of CCBRA into TRAC, odification of the SSC code (for liquid metal systems) to handle water reactors, and the creation of irrne-diately operable ccmputer cede decks for each operating reactor. Calculatiers performed en postulated event trees shculd provide a greater understanding of transients and small break LOCA events.
A review of all intercennected systems shculd be included in 2e i. proved reactor safety area of bis budget. Bis interaction step should include an examination of environmental ascects and a determination of the adequacy of instrumentation between intercennected systers.
1371 2M
ECCS Mtg June 19-20, 1979 Dr. Murley presented the Pl 1981 budget (please see handouts).
Dr. Murley will provide a list of priorities for this budget and the supplement before mid, July.
The German computerized monitorirq system is not really a great advance in the state-of-the-art since this type of plant monitoring is performed in refining cperations. We nuclear industry, however, is generations behind in terms of control and display diagnostics. Bis system could be imple-mented for LCET. Be Germans have been working on it since about 1971.
The need for larger scale test results was pointed out. We use of commercial power reactors for specific tests (e.g. small break LCCAs) could be accomplished without hazard to the plant. Special accident following instrumentation could be installed in commercial reactors to aid the study of abnormal transients.
Secarate Effects Research - (Transcript pages 305-375)
SEMISCALE - A. Serki:, NRC (Transcript pages 305-331) te budget for the SEMISCALE is 6.7 millien (?? 80) and 10.5 million (requested for PI 81). We upgrading of de facility includes the addition of ence-through and U-tube steam generators estimated at 3.5 million.
ECC bypass studies are being phased cut and will be essentially ecepleted by Fi 1981. Technical support is being redirected into instrzentation and diagnostics.
SEMISCALE has been set up for upper head injection (UHI) experiments. Be facility will be modified with additional insulation since it has been gstulated that excessive heat from the dcwncemer walls has led to voiding asso71ated with core back-ficw and high core steaming rates. B is modifi-catiin will correct surface area to volume ratio effects. A test rm at this facility considering the relationship of pressurizer level and core level has been accomplished. Bis Subccatmittee will get copies of this report.
137i 235
ECOS Mtg June 19-20, 1979 The Proposed schedule for the SDtISCALE includes:
1.
small break testing through August 2.
rerun test S-06-7 to test the hot wall effect with the new insulation installed 3.
feedwater transients during late fall-early winter.
All break effects (e.g. pressure, 2D-3D) cannot be studied at any one facility. Sey must be integrated using codes and engineering judge-ment. Subcommittee members questioned the relation of SDtISCALE to a full sized CG.
Questioned raised concerning pressure effects, 2D-30 effects, externally mounted thermoccupies on fuel pins, scali.q, and two-phase pheonmena are under study. me shift from the large break to the small break LCCA require some redirecting of the facili-ties.
Mr. Ebersole pointed out ptential problem areas (vortex formation in an accumulator, activating valves under dynamic heads, discharge of nitrogen into a primary loop from the accumulators, and UHI).
Mr. Michelson pinted out situations where large breaks are charged to small breaks as in the case wherein isolation valves fail to close fully. Some WRs have loop isolation valves.
Slowdown /Reflood Heat Transfer - A. Serki::, NEC (Transcript pages 331-349)
Each of the LOCA/ECCS tests facilities have their own capabilities.
SDiISCALE is a WR facility whereas TL"'A is. a BWR facilief.
FLECHT-SEASET has a testing capability of up to 2300 F and was designed to pri:rarily handle flow blockage effects during refleed. Preliminary analysis of steam cooling tests at FLECHT show a 50% enhancement over current models.
1371 236
ECCS Mtg June 19-20, 1979 Several current Oak Ridge articles have addressed the concerns that electrical heaters cannot really simulate the ther:asi hydraulic boun-dary conditions or simulate fuel pins.
Mr. Serki: has the information on these articles.
The AWS program is related to LOCA analysis. A EWR reflooded with cool, clean water recuires the insertion of control rods. Be rod drive control insert and exhause tubes are in direct line of the mu::le blast from LOCAs inside the drywell. tere may be a mechanism for tube darage that will degrade rod insertion capability.
It is difficult *w model nuclear feedback effects with electrically heated bundles.
2D/3D Procram - G. Bennett, NRC (Transcript pages 349-362)
The 2D/3D program is an international cooperative program (Japan, Federal mpublic of Germany, USA). Me program analytes steam binding that may occur during a PAR LOCA, flow distribution effects within the core, flow hydro-dynamics in the downcomer and upper plenum during core uncovering, and natural circulation.
Se facilities involved are the Cerran upper plenum test facility (full-scale vessel), the Japanese cylindrical test facility (2000 electrical rods, full height core, 4 loops), the Japanese slab core facility (full radial capability), and instrumentation and analytical work (via TRAC) at W Alamos. 2e German FKL facility (full height core - 3 loop cabe.bility) is not formally part of the program but has been used to tent instrumentation for the other facilities. tese facilities are in various stages of' construction and testing. Se program is currently on schedule.
New instr =entation (e.g. level detectors and improved pressure, temperature instr =entation) to be developed in this program shculd be usable in operat-ing !!aRs instead of only in research.
1371 237
ECCS Mtg June 19-20, 1979 21s program has considered the possibility of the study of an evapora-tive-reflux-condensation steam generator if natural circulation is determined insufficient in U-tube steam generator systeers.
Three dimensicnal effects such as countercurrent flew, cocurrent flow, cross flow, and the chimmey effect may be expected in this program (The chimmey effect occurs when steam is generated in the central region of the core, rises up the hot channels., and den falls back along the core peripherf).
Model Develoceent Ex eriments and Technical Succort - A. Serki::, NRC (Transcript pages 362-375)
The researchers and institutions involved in developing, benchmarking, and verifying heat transfer correlations were listed. Various examples of technical support (e.g. advanced instr =enation) for the 2D/3D project were presented.
Additional kncwledge is needed in the following areas; de relationships and characteristics of :rass, volume, and energy transport through orfices and relief valves, the perfonnance of boilers at reduced pressure and reduced heat transfer surf aces, and de determination of heat sinks for the break and secondary system.
LOFI' Research - D. McPherson, NRC (Transcript pages 376-396)
Se LOFT program has completed all non-nuclear tests in the L-1 series, two nuclear loss-of-coolant experiments in de large break series, and an isother:ral small break icss-of-coolant experiment. A new series of experiments will focus on system respense to off-normal conditons.
An assessment of conventional process information and a comparisen to de special research instrumentation will be conducted.
1371 238
ECCS Mtg June 19-20, 1979 A two-phase flow calibration facility will be completed at LCET in 1980.
It will have the capacity to calibrate flows in full si:e pipes.
The development of audio detection devices (e.g. loose parts monitoring, onset of boiling) was addressed since the Subecmmittee members didn't recognize it as part of the NRC instrumentation develcanent program at the LCET facility. Se development of these instruments is net currently ongoing in the 3D program. De 3D program will be looking at accustic monitoring.
Preliminary studies at I.CET on noise in ion chambers and self-powered neutron detectors indicate that two-phase flew in the downcemer can be monitored.
Analvsis Develcoment Branch - S. Fabic, NRC (Transcript pages 396-441)
A faster version of TFAC (TRAC-PFl) applicable to EWRs will be available in late 1979.
Since the complex codes are unwieldly for extensive mapping of great varieties of accidents, two suggestions have been offered - the development of hybrids and very fast digital routines (intelligent shortcuts using microprocesso rs).
Se cedes presently in use and in develegnent do not provide for operator action. Some preprogram eptions are available which allcw us to centrel, for example, wtien a pump will step or valve will open.
The aspirator in the main feed line inlet which brings the water into the downcemer of S&W steam generators is not modeled in any of our or B&W's codes. Bis may be i :pertant in natural circulation calculations. Se auxiliary feedwater in B&W steam geneaters enters the generator and is 1371 239-
ECCS Mtg June 19-20, 1979 apread acrcss the tubes by a perforated baffle. Bis heat transfer mechanism is not treated in either REI.AP or TRAC and is probably the reason our ccmparisons have not been good.
Comparisons with the reconstruction of the TMI accident and ccmputer cede evaluation have generally not been good. A few problem areas have been identified:
1.
importance of de steam geneator as a part of the ther-al hydraulic system 2.
two-phase flew 3.
flew through relief valves 4.
the mass of auxiliary feedwater into the steam generators 5.
hemogeneous medel assumption, and 6.
model of the "cardy-cane" loep configuration.
Key indicators (e.g., events, quench times, time to empty the pressurizer) measured in LOCA tests and calculated by TFAC were in reasenable agreement.
This relationship is used to assess de LOCA codes.
Closinc Ccements - M. Plesset (Transcript pages 441-462)
The SD4ISCALE f acility simulated a PCRV NI transient in a Westinghouse U-tube steam generator design. Se surge line, however, was piped to match de S&W 10cp seal design. Se pressuri:er remained full during the s:all break LOCA (break ac te top of the prescuri:er). B is indi:ates that the pressurizer may give false indications of liquid inventory in all
?ds. De small size of the pressurizer se -la surge line and volume raise questiens as to the meaning of results obtained at SD4ISCALE for small break LOCAs.
Me.tt:ers of de Subecmmittee felt that 2e activities presented at this meeting in respense to IMI-2 were not organized. Risk assessment ceuld be employed to order the priorities for experiments in order to organi:e the efforts at dese facilities.
1371 240
9 ECCS Mtg June 19-20, 1979 Notes:
(1) For additional details a complete transcript of the meetirq is available in the NRC Public Decrant Room, 1717 H St., N.W.,
Washington, D.C. 20555, or frcm Ace-Federal Reporters, Inc.,
444 North Capitol Street, N.W., Washington, D.C. 20001.
(2) Materials provided to the Subcommittee at this meeting are on file in the ACRS Office. In general, these materials include:
a.
Budget su:maries for W 1980 and W 1981 of all projects discussed.
b.
Program outlines and a schedule of testing for the facilities discussed.
c.
ECCS, LOCA, ATdS Computer Code availability.
d.
ECCS Computer Code Applications.
e.
NRC Sponsored Calculations of the 24I-2 Accident Scenario, f.
Nodali::ation models of reactor systems -sed in TBAC.
137i 241
Federal Rer,ister / Vol. 44. No.108 / Enday June 4.1979 / Notices 3 21)S 4 Dated: May :9.19r9.
HUCt. EAR REGULATORY John C. flayle.
COMMISSION Advisory Committee Management Officer.
Advisory Committee on Reactor
- n o,., % m 3 m a w,s
..g osu.seo coot nu-a Saf eguards Subcommittee on Emergency Core Cocitng Systems (ECCS); Meeting The ACRS Subcommittee on Emergency Core Cooling Systems will hold an open meeting on [une 19-20.
1979 in Room 1046.1717 H St N.W.,
Washington, DC.,20555, to review ECCS models for small breaks in Babcock and Wilcox reactor systems.The Subcommittee will also review the proposed FY-at NRC budget figures for ECCS.related research activines. Notice of this meeting was published in the Federal Register on. fay :4.1979.
N In accordance with the procedures outlined in the Federal Register on October 4.1978 (43 FR 439:51, oral or written statements may be presented by members of the public recordings will be permitted only dunng those portions of the meeting when a transcnpt is being kept, and questions may be asked only by members of the Subccmmittee. Its consultants. and Staf! Persons desiring to make oral statecer.:S should notify
, the Designated Federal F=;!cyee as far in advance as practicable so that appropriate arrangements can be made to allow the necessary time dunng the meeting for such statements.
Tbe agenda for subicet meeting shall be ns follows: Tuesdc*,*. fune 19 and Wednesday. June :0.1979. 3:.10 c.m. until the conclusion of business ecc' dcy.
The Subcommittee may meet in Executise Session. with any ofits consultants who may be present to explore and exchange their preliminary
. opinions regarding matters wmch snould be considered denng the meetmg and to formulate a report and recommendations to the full Ccemittee.
At the conclusion of the Executive Session. the Subcommtuee will hear presentations by and hold discussions with representatives of Dabcock and Wilcox. the NRC Staff. and their consultants. pertinent to the above topics.
Further information regarding topics to be discussed, whether the meeting has been cancelled or rescneduled. the Chairman's ruling on requests for the opportunity to present oral statements and the time allotted therefore can be obtamed by a perpaiti telephone call to l.7 -[- l 2h2 the Designated Federal Emplovce for
.)
this meeting. Dr. Andrew t. U.stes (telephone 200-434-3 071 between 8:15 a.nL and 5.00 p.m EDT.
ATTACHMENT A
ACRS SUBCOMMITTEE MEETING ON EMERGENCY CORE COOLING SYSTEMS JUNE 19-20,1979 WASHINGTON, D.C.
ATTENDEES LIST NRC ACRS R. Audette M. Plesset, Chainnan B. Sheron J. Ebersole
- 2. Rosztoczy I. Catton P. Norian K. Gar 11d M. Lauben W. Lipinski W. Lyon C. Michelson R. F. Smith R. Shumway R. Anderson H. Sullivan R. Lee T. Theofanous A. L. M. Hon T. Wu L. Thompson F. Zaloudek Y. Hsu Z. Zudans G. L. Bennett A. Bates, Staff
- G. D. McPherson L. Shotkin
- Designated Federal Employee W. H. Beach R. Hoskins BNL W. D. Beckner G. Rhee K. R. Perkins A. Sertz B&W_
KEPC0 R. C. Jones K. Ota J. J. Cudlin K. Noda E. R. Kane BBR PICKARD, LOWES & GARRICK K. O. Layer M. Schwartz EPRI MPR R. H. Leyse D. M. Chapin DUKE POWER CREARE G. Swindlehurst J. Block ACE-FEDERAL WYLE LABS L. Weinschel R. Cumings 1371 "743 ATTACHMENT B
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TENTATIVE ME' TING SCHEDUII XPS ECCS SUBCCNMITTEE JUNE 19-20, 1979 Tuesday, June 19, 1979 - B&W Small Break Analysis 8:30 - 8:45 I.
Executive Session - M. Plesset 8:45 - 9:00 II. NRC Staff Introduction - Z. Poszteczy 9:00 - 10:45 III. EC Staff Review of Michelson Concerns - B. Sheron 10:45 - 12:30 IV.
Small Break Analyis - R. Audette 12:30 - 1:30
- Lunch -
Review of New B&W Guidelines for Small Breaks - B. Wilson 1:30 - 2:30 V.
2:30 - 3:30 VI.
NRC Methods for Review of LOCA Procedures at B&W Plants -
B. Wilson 3:30 - 4:30 VII. Audit of Cperator Training - r. Bogar 4: 30 - 5:00 VIII. Additional Questions and Adjourn Wednesday, June 20, 1979 - NRC Water Reactor LOCA/ECCS Research 8:30 - 8:45 I.
Executive Session - Opening Coments - M. Plesset 8:45 - 9:00 II.
Introduction - T. Murley 9:00 - 11:00 III. Separate Effects Research Branch a.
Semiscale - A. Serkiz b.
Blowdown /Reflood Heat Transfer - A. Serki:
c.
2D/3D Program - G. Bennett d.
ECC Bypass - A. Serkiz Model Developent Experiments - A. Serkiz/S. Fabic e.
f.
Technical Support - A. Serkiz/G. Bennett 11:00 - 12:30 IV.
LCET Research Branch - D. McPherson 12:30 - 1:30
- Lunch -
1:30 - 3:00 V.
Analysis Developent Branch - S. Fabic a.
Systems Codes b.
Compenent Codes c.
Code Assessment d.
Code Sensitivity 3:00 - 3:30 VI. Closing Coments - Adjourn 157 \\ 2W
D-)
CONCERNS 1.
ACCEPTABILITY OF INTERMITTANT HATURAL CIRCULATION 2.
TIME DELAY IN TRANSITIONING FROM NATURAL CIRCULATION TO POOL BOILING 3.
PRESSURIZER LEVEL WAS NOT CORRECT INDICATION OF WATER LEVEL IN CORE 4.
CONSEQUENCES OF SMALL BREAK ISOLATION /REPRESSURIZATION 5.
PRESSURE BOUNDARY DAMAGE DUE TO BUBBLE COLLAPSE 6.
BREAK ENERGY NOT REPRESENTATIVE OF CORE EXIT ENERGY 7.
EFFECT OF NON-CONDENSIBLE GASES (FROM CE SYSTEM 80 REP 1371 245
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SUMMARY
NO DISAGREEMENT ON PHENOMENA DESCRIBED BY C MICHELSON, CONCERNS UNDERSCORED IMPORTANCE OF NATURAL CIRCULATION FOR DECAY HEAT REMOVAL DURING SMALL BREAKS, B&W HAS PERFORMED DETAILED ANALYSES TO AJDRESS CONCERNS, RESULTS SHOW PHENOMENA OCCUR, BUTTHAT DECAY HEAT REMOVAL IS NOT UNACCEPTABLY IMPACTED.
1371 246 e
B&W SMALL BREAK GENERIC STUDY BREAK AFW llPI EC PUMPS LONG-TERM COOLING
.07 FT 0FF 2
0FF 390 SEC.
2 650
.02 0FF 1730
.01 1 a 20 f11N.
.01 0FF 2 a 20 MIN.
27711 LOFW 2
1 IIPI ON 1000 0FF 1000 PORV PORY (ANS*1.2) 0FF PORV (ANS*1.0) 0FF 11700
~
11900
.01 2
11975 g
.01 (ASYM) 1 5000 ft N!
.005 2
-~
N 6000 C
.01 (DB-1) 2
4E-l CONCLUSIONS 1.
AFW AT 20 MINUTES PROVIDE CORE COVERING FQR LOWEREP AND LOOP PLNTS FOR BREAKS SMALLER THAN,02 FT4 2.
HPI ONLY AT 20 f.iNUTES PROVIDE CORE COVERING FOR LOWERED LO 2
FOR BREAKS SMALLER THAN.02 FT,
3, 1 HPI TRAIN PROVIDES CORE COVERING FOR STUCK PORV IN LOWER AND RAISED LOOPS.
4.
HOT LEG BREAKS B0UNDED BY RESULTS FOR COLD LEG BREAKS DU ACTION OF VENT VALVES, 5.
SINGLE STEAM GENERATOR OPERATION IS ADEQUATE TO MAINTAIN COVERING FOR SMALL BREAKS.
i 371 248 o
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IF Tile IIPI SYSTEM llAS BEEN ACTUATED BECAUSE OF A LOW PRESSURE CO IT MUST ret 1AIN IN OPERATION UNTIL ONE OF Tile FOLLO Tite LPI SYSTEM IS IN OPERATION AND FLOWING AT A RATE IN EXCESS 3.
OF 1000 GPM IN EACil LINE AND Tile SITUATION llAS BEEN STABLE FOR 20 MINUTES.
OR ALL ll0T AND COLD LEG TEMPERATURES ARE AT LEAST 50 BELOW Tile 2.
IF TifE SATURATION TEMPERATURE FOR THE EXISTING RCS PRESSURE.
SUBC00 LING CANNOT BE MAINTAINED, THE IIPI SIIALL BE REACTIVATED.
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