ML20235N130

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Trip Rept of 671112-17 Visit to Paris,France to Represent ACRS at Third Meeting of Committee on Reactor Safety Technology of Enea & to Discuss Reactor Safety Questions W/ Appropriate French Representatives
ML20235N130
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Issue date: 12/07/1967
From: Palladino N
Advisory Committee on Reactor Safeguards
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Advisory Committee on Reactor Safeguards
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References
FOIA-87-40 ACRS-GENERAL, NUDOCS 8707170446
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n' gw ADVISv.<Y COMMITTEE ON REACTOR SA. f. GUARDS i UNITED STATES ATOMIC ENERGY COMMISSION wAswiuoron, o.c. soses December 7, 1967 MEMORANDUM r

To  : Members of ACRS From'  : N. J. Palladino, RS Chairman

Subject:

REPORT OF TRIP TO PARIS TO MEET WITH FRENCH EDA AND CEA REACTOR SAFETY REPRESENTATIVES AND TO ATTEN:) THE THIRD CREST MEETING s

This report summarizes my activities and cbservatiens as a repre:,entative of the ACRS during a trip to Paris, November 12 to 17, 1967.

Purpose of Trip:

The original purpose of the trip was to represent the ACRS as an observer at the third meeting of the Committee on Reactor Safety Technology (CREST) of the Euro--

pean Nuclear Energy Agency (ENEA) on Wednesday, Thursday, and Friday, November 15, 16, and 17, 1967 in Paris, France.

By going to Paris earlier, I was able to accomplish a second mission - that of dis-cussing reactor safety questions with appropriate French representatives on November 13 and 14, 1967.

Visit with EDF and LEA Representatives, November 13 and 14, 1967:

The meeting on November 13, 1967 was held at Saclay. The meeting on November 14, 1967 was held at CEA Headquarters, 29 rue de la Federation Paris 15. France. The U. S. representatives included Edson G. Case, Director of Reactor Standards of AEC, Joseph J. DiNanno, AEC Scientific Representative in Paris, and myself. French representatives included Messrs. Bourgeois, de Vathaire, Alami, Bacher, Dennielou, Costes, Meunier, Celery (phonetic spelling), Oullion, Havard, Burban and Delayre.

This meeting on November 13 concerned principal problems in safety of the HTGR reactor and the safety of prestressed concrete vessels and their liners. The meet-ing on November 14 was concerned with water-cooled reactors. The agendas for these meetings are given in Appendices IA and IB.

The following principal points were brought out during the discussions:

A. 11-13-67 (HTGR)

1. Marcoule reactors are develop-ental reactors for production of plutonium and electricity; they are known as G-2 (1959) (250 MWt) and G-3 (1960) .

These were horizontal reactors inside prestressed concrete pressure vessele 8707170446 070713 l PDR FDIA THOMASB7-40 PDR 1

They.used. push-chrough fuel. Each had burst detection systems which permitted 20-minute sampling of each channel. Used magnesium-zirconium alloy cans. They are operating very well. One fuel element melted because wrong thermocouple was used. to' ccntrol flow. Another fuel element. failure occurred because of flux distortion. Gas flow (002) l l enters core at 140 CC and leaves at 330 CC. I 1

1 2. EDF (Electricity de France)- builds power plants and CEA provides the fuel elements. The EDF nuclear power development program included 3 prototypes, EDF-1 (70 MWe); EDF-2 (350 MWe); and EDF-3 (480 MWe) .

EDF-1 and 2 utilized steel pressure vessels-10 meters in diameter; 6" thick, and 20 meters-high... When this program started in 1956, they..

had no experience with. prestressed concrete vessels.- The steel vessels were field erected; EDF-1 had a crack in the vessel next to a weld during construction which had to be repaired. Operating pressure in.

EDF-1 and 2 is 25 atmospheres.

These reactors had vertical channels. - EDF-2 and later reactors have on-line refueling. On EDF-2, the CO2 enters the core at 200 C and leaves at 360 C. j

3. "EDF-3 was another matter." Its power is higher. It was designed in 1959-60. Economics and experience with G-2 andG-3igicateduseof prestressed concrete pressure vessels. CO2 enters cere/240 C and leaves at 400 C. In other respects it was similar to EDF-2.
4. Problems were experienced with EDF-3.. The fc11owing were discussed:
a. Failure in piping cf failed-element detection system: EDF-3 uses a series of channel sampling tubes fer fuel element detection arranged in a cross matrix to identify the channel having the failed element.

The system involves 800 tubes each of which passes through the vessel wall via a bulb which is provided with a small inflow of unheated CO2 for make-up leakage flow. The tubing (made of- 304 S.S.) was fixed at two points and experienced severe temperature cycling every half hour when 400 C CO 2 was passed through the tubing as part of the sampling process. These tt.bes failed in 2 days of operation.

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I8 E 8 CO4 (O O *c, y l R xED g V cyggy, @g gg, j The tubes were replaced after changes were made to pemit greater flexibility of the piping; the low temperature gas in the bulb was also eliminated. l l

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b. Difficulty with detection in the failed-element detection system:

Detection is sensitive provided that the radioactive background is low. However, high moisture content gives problems. They found that problems arise at 1000 ppm; they desire 500 ppm. Hence they have had to operate at low power to dry out the system. Condensa-tion in low-temperature parts of the system also gave problems.

This system is tied into their scram circuits. If any three channels exceed pre-established limits during the half-hour scanning process, the reactor will scram automatically. Any gross monitoring signal in excess of limit will also cause an automatic scram. If one of the '

channels shows a leak it is placed on continuous monitoring.

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c. Heat exchanger tube failure: This plant uses 192 heat exchangers with an average of about 50 tubes per exchanger. These heat exchangers "gave trouble" on start up. A number of leaks were found. It takes weeks to find and isolate a leak. These tubes are of carbon steel with grooves formed on the exterior walls to improve heat transfer. It was found that periodic failure of the forming ~ tool led to inclusions in the tubes; this is believed to be the source of the difficulty. They have decided to replace the tubes, but they did start up with the old tubes in place, hoping that the big problem areas have been identified and corrected. On EDP-3, these heat exchangers are external to the reactor vessel. The activity of the gas inside the vessel was 250 mr/hr
d. Turbine difficulty: The turbine does not have the lube oil pump integrally tied to the turbine shaft because the shaf t is already j quite long without it. Instead an electrical tie is used. Two simul- i l taneous faults developed in this electrical tie, and, as a result, the shaft bearing got no oil, and the bearings failed.
e. Alternator problem: Overheating and melting of wires took place at hot spots. This has been a commen industry problem in France. These alternators were built under Westinghouse license.
5. Shutdown cooling requirements fcr EDF-3 were discussed. It was pointed out that natural circulation could rcmove 50 MW of shutdown heat. (I don't remember on what temperature condition cf the fuel this is based.) This was felt to be adequate for the core whose normal full power output is 1650 MWt because the 4 blowers are turbine driven with steam from the main heat exchangers, and they can provide for circulation of C02 until the decay heat reaches the 50 MW 1evel.
6. Design philosophy for subseq;ent EDF plants was then discussed.
a. Because of concern about the possible failure of large diameter coolant ducts (4 ft. diameter), it was decided to locate the heat exchangers for EDF-4 and subsequent HIGR's inside the PCRV as the British did.

However, the British located their heat exchangers around the outside

i of the core. This, the French noted, could be done for the British 280 MWe plant without having a PCRV of excessive diameter. Inasmuch as the French were looking at plants to produce 500 MWe, this arrange- ,

ment was not felt to be the best. Hence, the heat exchangers were l placed below the core. The chief problem with this arrangement is  ;

that of isolating leaky tubes; approximately 1/2000 of the tubes can i be isolated. I This apprcach is being taken in the following plants: Saint Laurent

  1. 1 (EDF-4) (500 MWe) and #2 (530 MWe), Burgey #1 (550 MWe) (new design) and Fessenheim (700 MWe) .

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b. The new reactors, beginning with EDF-4, require blower power to pre-vent shutdown melting of fuel cans. EDF-4 has four independent turbine j driven blowers each of which is supplied with steam from a separate l

oil-fired steam boiler if the nuclear steam source is lost. These oil-fired boilers operate continuously at reduced power to provide l steam to auxiliary components. l l

The philosophy behind this arrangement is that one boiler must be pro- <

vided to accomodate maintenance, and one must be available for supply- l ing the blower steam even if two others fail. However, the plant has i only 2 storage fuel tanks and 2 separate water feed systems for the four boilers.  ;

During normal operation, the steam for driving each of the boilers is ,

supplied from a separate bank of two primary-system heat exchangers; {

there are eight banks of heat exchangers.  ;

7. Mr. Costes (CEA) then discussed the design and general safety problems associated with prestressed concrete reactor vessels. The following i summarizes the notes taken:
a. Functions and requirements:
1. Must contain gas; leak tightness requirement leads to need for steel liner.
2. All concrete near liner must be in compression.
3. It is not realistic to have a "no cracking" requirement on the exterior surfaces.
4. They have no requirements on concrete porosity.
5. They assume linear distribution of pressure in cross-section of the concrete.
6. Criteria assume thin liners ($;1"). Anchorage of liner to concrete is not specified.
7. Safety lies in the steel of the PCRV. They limit the tendon stresses, but allow them to go to 80*/. of the rupture limit (Y . P . = 9 0*/.) . Steel should have 3%

thatgoodwirecharacteristica11%. elongation;itwasstated elongation. Suppliers will only guarantee 2 to 2i% elongation. Only drawn wire is used.

8. Steel shall resist corr'asion; grouting or circulating dry air around tendons is required.
b. In response to questions about the use of grease for corrosion pro-tection, it was stated that tests would be needed to prove the reliability of the grease.
c. In one (or some?) tendon at Marcoule, relaxation due to corrosion was experienced.
d. A program is planned te test if a PCRV will burst under pneumatic pressure because of tendon corrosion. It is believed that the PCRV car burst if all tendens have their cross-sections reduced.
e. Loads on concrete are normally limited to 1/3 of resistance for static loads end 0.55 for dynamic icads. Near penetrations, static tensile loads are permitted to go to 1.8 of resistance if steel is used to pick up the load,
f. Each contractor is required to build a 1/6 scale mock-up of a PCRV unless the design is essentially a duplicate of a previously tested design. The reasons for the mock-up were not exactly clear at this point. See Item 8 g for additional discussion. It was pointed out that it doer show up construction problems. The scale of mock-ups is dictated by relative size of vessel, tendons and openings. Tendon diameters are not scaled.
g. The calculated rupture for a PCRV must equal at least 2.5 design j pressure which is greater than the working pressure. Test must be j made at 1.1 working pressure.
h. In response to questions about the reliability c f closures for pene-trations in a PCRV, the Prench saw no problem. They do not view the failure of a penetratica closure as a safety item requiring the same ,

consideration that we give to failure of primary system piping. They l sketched the following diagram of a blower closure, but indicated that j i

they would prefer tc eliminate the bolts by use of segmented conical shear blocks. 's i i i i i ,ii,ii/

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8. Mr. Celery (EDF) discussed inspection and testing of PCRV's. He spoke only French, so my notes represent my impressions of the translations.

This is important because the ratio of French words to the translated English equivalents was at least 30 to 1. The following summarize my notes- J

a. Concrete quality is assumed by tests en samples poured from concrete l batches; these include creep (or shrinkage) and compression tests l which are made by the Generating Board. No creep quarantees are made by the supplier, i
b. Samples of all grouted cables (a few meters long) are cast in con-  !

crete beams at the time of pouring of the main vessel. These are tested periodically.

c. Tests of tendons include sample measurement of tensile strength, >

elongation, and Young's Modalus.

d. Reference was made to a paper by N. Bearjoint entitled " Examination of Pressure Vessels for Reactors" delivered at a conference in <

London, March 1967.

c. Extensive instrumentation is installed for evaluating characteristics ,

of the vessel during its lifetime.

These include: I I

1. Vibrating wire test guages on 311 wires to test relaxation or local deterioration.

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2. Twenty to thirty hygrometers are installed at pouring.

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3. Pendulums are used to note tilt or distortion. I l
4. Extensometer devices are placed on cracks in concrete.
5. A few ungrouted cables are always installed for testing relaxa-tion and general changes in vessel characteristics.
6. Some 260 thermocouple are installed on the concrete side of j the liner.

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i f. Pressure test is made on vessel prior to operation at 1.25 times work-ing pressure or 1.10 times design pressure; this is done cold.

j g. Tests on models (1/5 or 1/6 scale) are made as follows:

1. Load tests on tendons
2. Test of cold elastic action ,
3. Determination of elastic limit '
4. Rupture pressure (2.5 working pressure)

Tests on EDF-3 models included cold tests with and without liner and I

hot tests with liner.

h. Although there had been a feeling that PCTd's were well understood, they have had some surprises on testing of models for EDF-4 and Burgey. Two models ~were built for each, one for thermal tests and one for other tests.

B.- 11-14-67 (Water Reactors)

1. The French have been interested in water reactors for a long time. They have a swimming pool test reactor comparable- to Spert IV. They plan a loop for this core which can provide cooling up to 100 MWt for 10 minutes.
2. Several incidents on SENNA were reported.
c. Turbine blade failure on first stages of high pressure and of inter-mediate pressure turbiner.,
b. Movement of a control rod out of the core; high power level scram occurred.
c. Fire of masonite screen provided by Westinghouse when pressure vessel was at 2800 0. See sket'ch below. CONTA IN H ENT (C&osED) 9 A

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st.nss wool- MetTEb (esT. c,oo'c )

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h H nsou tTE Bunc ue p A F T E R, P. v. wA.5 Nevrtory AT l., 8 O'C, N ON ITOR The temperature of the primary system was being raised as part of the l program to measure the temperature coefficient.

When the masonite started to burn, the pressure vessel temperature-was decreased as rapidly as could be allowed (55 C/hr); six hours were required. Water could not be poured on the. fire for fear of damaging the vessel. The masonite burned completely.

Neutron meter cables burned. In 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, all indication of neutron level was lost. The glass wool at the mechanism housings melted; the estimated temperature was ~ 6000C.

After the incident, the studs were removed and given a visual and j micrographic examination. The surface of the vessel was also checked.  ;

No damage was found on the studs, head, or vessel. It took a month j to recover from the incident, i

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4 The French were disturbed by the fact that this same incident occurred on SELNI but they were not told about it by Westinghouse nor were any design changes made.

Mr. DiNunno suggested that this type of incident be reported by the Europeans so that it could be included in a new AEC newsletter which had been started, l

3. The French do not subscribe the British overpressure test concept by which they try to assure themselves that no untenable crack growth will occur over the next cycle of operation.
4. Interest was expressed in progress on the development of acoustic crack detection mer. hods. There was also interest in the possibility of using acoustics to identify broken parts in the bottom of a pressure vessel.
5. Other items of interest to the French which were discussed included:

1 l a. Iodine retention in suppressive pool of BWR's. The French are planning tests in this area.

b. Local overpressure within the containment during a loss of coolant accident. They are concerned that compartmentalization of the con-tainment has not reflected this concern. They point to che LINGE j (Germany) reactor design in which a heat exchanger is enclosed by {

a brick wall which could blow apart and generate missiles; it is l now being removed, i

c. Inconsistency between the containment design pressure and the collaps-ing pressure capability of components inside the containment. They point out that SENNA has pumps within the containment that are good only for 1,7 hars while the containment is designed for 4 barsr The adequacy of ion chambers to withstand containment pressure was also j questioned. Mr. Case pointed out that these have withstood repeated  !

containment testing at 25 psi on the N. S. Savannah. l

d. The reasons why valves can be relied upon to provide containment for BWR's, Would this practice be acceptable for PWR's which use heat exchangers outside the containment? i
e. The adequacy of calculations to predict blowdown forces. They pointed out that Prof, Gallagher of Illinois calculates forces j three times those predicted by Phillips Petroleum,
f. Adequacy of spray tests to confirm cooling and flow distribution tests.

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I C. 11-15,16,17-67.(Third CREST Meeting)

1. Location: OECD Headquarters, 91, Boulevard Elemans, Paris 16e, France.
2. Attendees -

See Appendix II i

3. Ag2nda - See Appendix III ]

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4. Copies of Papers Distributed - On File in ACRS Office
5. Ss tmary of Discussions and Activities The Third CREST Meeting involved four types of activities.
a. Summarizing of CREST Specialist Meetings.
b. Progress of work in several technical areas of interest to CREST.
c. Reports of Working Group on Safety Assessment of Water Reactors.
d. Planning of future activities.
e. Election of officers.

Each of these topics is discussed below. In addition, I have appended a brief personal commentary.

l 6. Summary of CREST Specialist Meetings

a. Meeting on Prestressed Concrete Pressure Vessels (Foulness, March 15, 1967). Oral presentation only was by Mr. Nicholson (UK).

PCRV's are of interest because of advantages that accrue to them.

There is great deal of interest in failure modes. The U.K. is trying to determine failure mode of a cracked prestressed beam; they are also checking acoustic recordings of loaded beams to determine micro cracks. Germans reported unique pressure vessels made of slabs pre-stressed by jacking.

Of interest was the report of UK tests made on a model of a hexagonal vessel with straight tendons longitudinally and along the roof. It was pressure cycled from zero to progressively higher pressures up to 575 psi. It failed prematurely and violently due to failure of anchors on vertical prestressed tendons. These anchors had been poorly manufactured; and, in fact, others like them had been recalled; unfortunately, these had been overlooked.

Another UK model tested, failed more predictably at more'than twice design load (1100 psi vs 500 psi). Again the vertical tendons l failed; another also failed in this way. Failure started with an i equitorial crack near the center, and the vessel bulged outward.

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b. Meeting on Reliability of Electronic Equipment and Systems for ,

Nuclear Reactor Safety (Brussels, April 11 and 12,1967). Oral I presentation only by Mr. Vinck (Euratom, Belgium) .

This meeting concentrated on failure-rate data collection. Data are required from several types of sources, operators, manufactures, 1 and regulatory bodies. l i

Data reported ranges from operational nuisances to safety data. l Methods of analysis range from simple to complicated, depending )

I on purpose for which the data are to be used. It is difficult to arrive at uniform dcta gathering methods in view of wide range of equipment involved and the. wide range of faults to be reported.

It appears smrth giving attention to the relevance of the data before trying to systematize data collection.

I It was concluded that, with respect to public safety, electronic l instruments do not constitute a major hazard; mechanical failures, such as_the rupture of yipes or other components, affect public j safety far more greatly. l 1

Mr. Nicholson added that many people desire to attend this type of

, meeting as observers. They should be permitted to do so but sit l in an audience so as to permit free discussion by persons on the pannel. He suggested three further specialist meetings in this i area. 1 l 1. Formulation and specification of reliability systems. l l

2. Extension of reliability to heavy electrical equipment.
3. Extent to which computers can be used in reliability  ;

j analysis.

Much discussion ensured. Mr. Iansiti (Italy) pointed out that elec-tronic components are mass produced. How do we get data on the more

! important mechanical components which are not mass produced?

c. Meetings on Fast Reactors at Aix and Karlsruhe. Oral presentation

, by Mr. Millot (France).

l l The fast reactor conference covered a number of topics including sodium boiling, burnout, and crystallization.

With regard to boiling, two types were discussed; nucleate and single boiling.

Few data were available on recondensation. Some data were available from Ispra and the U.K.

Burnout: It is difficult to define thermal flux. Several codes have been established based on water studies and single bubble boiling. At present it is difficult to use available data. There is much scatter in the data (20-8000 ). The boiling people decided to hold special meetings on this subject.

Kilby of the U.K. gave "a fine analysis on safety" using probability methods. This was considered premature by some. An entire session was devoted to safety. At present, it appears unnecessary to improve the Bethe-Tait method.

Fuel tests in loops: Results from TREAT were presented. The Belgians presented results of needle tests.

With regard to steel-cooled reactors, it is difficult to compare steel cooling with sodium cooling so far as safety is concerned.

A report was presented on the Fermi reactor.

General comments: Twenty-five out of thirty papers presented theo-retical work; only five reported experimental results. The wide ranges of scatter were among experimenters rather than within the work of a particular experimenter. There appears to be little difference between boiling water at 80 atmospheres and boiling sodium at 2 atmospheres. Nucleation centers are not as prevelant, however, as they are in water, l

d. Papers were passed out regarding two Euratom meetings. These were discusred. They have been deposited in the ACRS files.

l (1) " Steel Pressure Vessel Safety Problems" (Exchange of Experi-l mental Results and Theoretical Investigations) . Meeting at Stuttgart, September 26 and 27, 1967.

(2) "Two Phase Flow Dynamics", Summary of Symposium at Eindhoven University, September 4 to 7, 1967.

(3) "Information Meeting on Prestressed Concrete Vessels". Meeting in Brussels, November 7 and 8,1967.

7. Progress of Work in Several Technical Areas of Interest to CREST Papers were presented in the following areas and discussed. Copies of the papers have been placed on file in the ACRS offico,
a. " Progress in the U.K. on the Safety Assessment of AGR, Using the Probability Method" by J. H. Bowen.
b. " Progress Report No.1 - UKAEA 3afeguards Division" by staff of ,

people. Presented by J. H. Bowen.

This report treats experimental work on rapid heat transfer between hot spheres moving through coolant; fundamental work on boiling processes; analysis of accident involving 100 lbs of molten mass which fell into a trough of water with great energy conversion; interaction experiments; depressurization studies; spray cooling; test of re-enforced plastic to contain a pipe failure; fission product work; and vented containments.

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c. " Progress of Ispra Safety Activities in the Heat Transfer and Mechanical Fields" by M. Montagnani and H. Holtbecher.
d. " Studies Related to Depressurization of Water-Cooled Power Reactors Performed in 1967 in the Federal Republic of Germany" by H. Karwat.
c. " Current Research on Coolant Blowdown in the U.S.A." by D. L.

Morrison and T. R. Wilson.

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f. " Conclusions sur la Reunion sur les Transferts deChalver an cours de Transitories rapides, Cadarche, 26-29 Avril 1966" by J. P.

Millot.

g. " Analysis and Automated Handling of Technical Information at the  !

NSIC" by J. R. Buchanan and F. C. Hutton. (This paper was passed out but not presented or discussed.)

8. Reports of Working Group on Safety Assessment of Water-Cooled Reactors Mr. Vinck presented a very interesting progress report on the CREST Working Group which has been appointed to assess water-reactor safety.

This report was followed by specific topical progress reports on various areas as outlined in the main repert. This group has had a number of meetings and spent from October 2 to October 20, 1967 visiting a number of installations in the United States and discussing safety questions ,

and back-up information with various American experts. j Preliminary drafts of the various repcrts were made available on a " limited usage basis" to members at the Third CREST Meeting.

i The scope of activity of the working group was the subject of much dis-cussion. The approach of the working group is summarized in its report I as follows: l

1. To discuss the various topics which are commonly considered l in safety reviews of nuclear power plants of the water-cooled  ;

type and more specifically of the boiling-water and pressurized- )

water type and to identify the uncertainties encountered during l such reviews; I

2. To identify those technical problems which could more quantita- l tively be evaluated through further interpretation of running j theoretical and/or experimental programs; '
3. To identify specific technical problems which require further theoretical and/or experiments 1 investigation.

Mr. Iansiti commented that regulatory bodies need detailed information, including particular codes used in design, the basis for these, the extent to which they are supported by experiment and experience, accept-ance limits, and identification of problem areas.

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i Mr. Farmer pointed out that the topical reports attempted to summarize and discuss technical information of rapidly changing nature often with-out any indication of its implication or its usage in evaluating safety.

He also questioned the advisability of trying to identify areas in which further research is needed, particularly if the safety implications are not set forth. He pointed out that specific topics are the subject of specialist' meetings where the experts can exchange far more detailed-information.

Mr. Iansiti pointed out that specialists worry primarily about phenomens; safety evaluators must be concerned about their safety implications and.

usually aren't at specialist meetings.

Mr. Alonso of Spain reminded the CREST members that the problems of a l well-developed country are different from those of a less-developed I country and that the problems of both groups must be satisfied. For example, his country needs specific data on various phenomena as pell' as the implication of such data; they also need information on calculs-tional procedures, inspection techniques and standards, acceptance tasting and ctandards, applicability of codes, end operational specifi-cations.

Mr. Beck felt that the report should list the research work going on in cach country. I pointed out that the report should also give atten-tion to prevention of accidents and that examples of experience on various points should be cited to illustrate the importance of both the steps for prevention and the means for coping with circumstances.

On the next day, Mr. Vinck listed on the board a number of items which he felt this working group should consider, (1) For any accident, consider which are the points to look at.

(2) Existing methods of calculation critically listed.

(3) Reference figures against which the check has ta be made.

(4) Fields in which research is needed.

(5) Accident tree.

(6) Specific topics to be covered for water reactors (a) Primary circuit failures (blowdown, core heat-up, emergency cooling, hydrogen formation, etc.)

(b) Reactivity accident (c) Activity release, transport, deposition, removal (d) Spatial instability-(e) Selected topics in design analysis - prevention of accidents (f) Other accidents Mr. Birkhofer of Germany expressed the opinion that the committee should i

not try to set forth general guidelines for safety; this must be done by l each country.

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l Many other comments were made, but most people fel't it worthwhile

. proceeding with the efforts of the working group.

It was agreed, therefore,' that there would be a meeting 'of 10 to 12 (

interested.peeple, in about three months, to discuss the future direc-tion- of this work. In the meanwhile, individuals should write to the Secretariat with appropriate suggestions.and comments and expressions -!

of 1,nterest. Comments should be received by Christmas so that they

-can be circulated to expected' attendees. The workin6 group would also ,

attempt to complete the reports _which had been reported on et this meeting.

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9. Future' Activities .)

l In addition to the meeting on the direction of the working group on safety assessment of water-cooled reactors, as discussed above, the #

following activities are tentatively planned:

l a. CREST Specialist Meetings on:

(1) Reliability (Ispra, May or June,1968)

(7.) Calculational methods on blowdown and related' topics and heat transfer (Munich, Spring,1969)

(3) PCRV (Toulan, September,1968) '

(4) Depressurization and Fast Heat Transfer (October,1968)

b. CREST. 4th Annual Meeting (Late October or Early November,1968)
c. Other Meetings:

(1) European-American Committee on Reactor Physics to hold meeting on " Application of on-Line Computors .to Nuclear Reactors" (Norway, September,1968)

(2) IAEA is to have meeting of a working group on " Porting and Entrance of Nuclear Ships" in February, 1968 (10-15 people)

(3) IAEA will hold meeting on " Safety of Critical Facilities and Test Reactors" in May, 1968 (10 to 15 people) )

I D. General Commentary:

Both aspects of my trips were extremely worthwhile. Not only did I learn a  !

great deal of detailed information, but I.was reewakened to the many common ,

questions we still must answer in the safety field, particularly those re-lating to large reactors.

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I was struck with the intensity of interest in coping with safety problems  ;

I on the part of all countries. It would do us well to participate because  !

many of these countries have real needs, and if these are not met, they could j get in trouble. Any large accidents in Europe would have an impact in the United States.

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I believe the American industrial companies should participate as U.' S.

representatives in many of the specialist meetings as well as general safety meetings. .They could both cor.tribate and learn.

In working with the French and the' CHEST Working Group on the Assessmer.t of Water-Cooled Reactors, ptartic;.larly during their visit to the U. S.. I felt .

a bit. uncomfortable abeut. facing some of their questions about core melt-through, and coping with accidents and the implications of need for more work in these areas.

I would add a word of caution to anyone who attends such meetings and who must rely on simultaneous translation to follow the discussion; these are not always too precise; many significant phrases are left out or misinter-preted. It pays to ask for clarifiestion wher. this discussion is not clear. -

I would urge continued active participation of the U. S. in.the activities of CREST.

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I

A00 M FC/1 VISIT 07 i LD."D'.! 0.43 A 'D INISIO TALIAD1D h I 07 Ti!F. U.S. 430 :i s .. n. i aii .:r,,

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(DTA IL 11. OrcerA Flcor . Bui'tcip; 33) ihtcrc. Thch:r, Alnni, l Donnic1cu, (.o Vcthairo, 10:C) A.1.. Prineile .1 prom:rc cor. crning t1.2 J}3tn'.cr, Cost 3c tr.foty of tb U}0P. rs cier .~.

  • ) i: 0 3 P , 5 - 1.p nch .. y r:05i.d bp E T30ur;20ic- ,

. P.tO r.G ., F f(ty of Pactrecc:o Con::rct:

Verr.:On cid th:ir Ste;l Lin$r0

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Pr er ni':' ior, b:.- l', Co st?

. Function:, of the Vect:1 : Requira..

, r: W int h :.1 r:d I.ccident:1 0;;rctist .

. C.micul' ti: a : Crit:rin,12 tcir . cf .

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. Ce03it; ef Con:rei- .

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. Control: durin; Construt,ticn and Dcv21 ; :r.t ,

, Yesta on 1:cdnu cnd on Vecc:19 1

1 l.

. .T.r,~a rrky, 14. l'r. .e.:..:' :r 'J.0,67, (CEA Heedqusrtora - 29, rr2 (.: in Tid!ratica, }Sscrc. Bourgeois, DurMn;  ;

Pe.rit 15) Hovard, . Verd:au, D:1ryn;  ;

do Vedb:iro, ISurd:r,  ;

9:30 A.% . Diccuniera cn hier P.ac.etro Coctoc, Oullion, .j Donn101ou ,

u l

L _ - . _ _ . _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ ..__-_._..__.____._._._-_____.___m._ . _ _

Groupe de Trcvail DEP/GTSP/593 de S6rtt' den Filco 13 novembre 1967 j JO/dp I f.1Pl's: ; ((\ / F TB Vicit of El'. E. CASE and J. pALLADINO on 14.11.67 Discuccion cbout water cooled reactor Safety i

~~~--- ~

f Tentntive agenda J .. P.R

. E.7.m. ..L.. E. . S. S T I. ".,.,

c) What crc the princirc) requirencnts about the surface aspects of thc preccurc vcccc1 7 Accociated controlc ?

j i

b) At precert titc, vhtt cre the surycillance and inopoetion technique: I for the recoci cnd the internni structures 7 l

l l

c) Further dcycloptent of inopection techniques  ; poccibility of over l

precLurc tects. A.E.C. requirencnts for inspectability of the vecoci and the prinary circuit for the ucy reactors : influence of the i requironento on the design.

l d) Etterici probleuc  : requirement'c for the test-pieces irradiation i l

procrci.s J. r c cucplcs taken fron't

~ rolded zones

~

concs thornclly affected by the weldingc overlength of the vessel i c) Fluid fle, inducoc vibrctions of the internal structurec. Incidente l occured on several recctors and broken pieces' fell on the botton of the vescol. In quite c11 cases, these brchen picccs were discovered af ter several veche . What tbcut an acoustic cethod based on the

.../...

5

( 1 i

analycit of the noire spectrum response of the vescel to detect I these lost pieces ?

I I C.O. .;. 7 /3. 17.:.:.r_T.T.. S v s_?.r ..t: S_

c) T2 et cure cuppretsdon cyctens, can you connent a little the follo. 1 wing points : l

  • For cc3culation of this systen, does the fluid flow cominc out of the pricary cysten corrocpond.to a rupture on the biccent pipc or on a cca12cr one. I Tecin procran before and dur5nc operation, l j

l Effcetivenecc of the presoure supproccion pool for the r<-

tentien of ficrion products (Iodine) b) Purinc a b3c. Corn accident, transient overprecourco nay occur in local g rtr of the containncnt. Influence of thcoc_overprescu- j rco on the ecsign.

l l

IJI - J.If,}.};Q:gM2K:{f n) lio:. nach inportc.n c e it given, at project stage, to the demonctre.

tion of renietance of core and prccourc vencel internale againct biordcvn f o:?cc c.

l b) Enercency core coc3inc effectivenons '

  • with a non deforced core : influence of steau up-flou on tbc repartition of the spray flow.

+

with a deforned core : what vill be the rec 1 hect tranc-l fer mechanice : radiation ? conduction ?

c) In connection with the coolinc of the containenent, what precisc acsuuptionr do you take for the pcwer history in the euhininnent 1 (power ccnint out of the vecsc1 including heat produced by netal water reactionc).

.../...

i

1

(

l l

3-El) Evr cuch censidcretion ic presently civen to the inportanco of hyCrocen fornation, und the connected explocion hazard, due to radiolytic decomposition of emergency coolant water 7 c ) If o c cu el! consideration is presently given to d) the further development of calculation codels fo:s i t

accerning the consequences of core melting ff) t).e cinult tion of such conditions by scall expc -

rincnte i f) Oppori,ai;y of htvir; icole. tin; valien on the privnry loopn 7 l

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l l

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..._____._.J

O November 1967 i ),

COMcIITTEE CI; REACTOR Sid'ETY TECHNCLOGY COMITE DEU ';ECn.;1(UES DE SECURITE DES REACTEURS

{ /W' i 1 ' [i s  ??

Third Meetinc Troisilce R6 union 4 l

1 Paris, 15th-17th November 1967 i LIST OF PARTICIP/J;TS LISTE DES PidlTICIP!JiT3 1

l 1,US'Ht 1 A Dr. Heinz SCHMIDL AU2'r:1GnE Division for Heat Technology Osterreichische Studiengese11schaft fur' Atomenergie GmbH )

Reaktorzentrum Seibersdorf/U.0.

BELUIE M. Fernand LEO';ARD bcuilgUE Chef de Service Groupe Concun d' Exploitation du R6acteur 3R 2 Centre d' Etude de l'Energie Euc16 aire Mol CIJ:11DA Mr. Gusst LJ'2 r .

l Manager, Nuclear Safety Engineering Chalk River Euclear Loborntories ,

Chnik River Ontario EURATOM M. Sergio FINZI Commission des Communaut6s Europ6ennes Centre Ocmmun de Recherche B.P. No 1 Ispra, Varese

.italie M. G. GRASS Concission des Consunaut6s Europ6ennes Centre Commun de Recherche B.P. No 1 Ispra, Varese italie ,

1 M. Mario MONTAGLJiI EUE TCfsuite)

((cont'd) Commission des Communaut6s Europ6ennes Centre Commun de Recherche B.P. No 1 -

,I rmra_ , V are s e 4taile M. Willem F. VINCE Commission des Communautss Europ6ennes l 51, rue Belliard l l Bruxelles 1 l Belcique l

1 l FRAKCE M . R . BR ANDT S . E . G . li . I Electricity de France l 2, avenue de la Lib 6 ration 92 Clamnrt M. Didier CCSTES Groupe de SQret6 des Piles ]

D6partecent d' Etudes de Piles  !

Commissariat i l'Energie Atomicue l i

Centre d' Etudes Nuc16aires de Sacicy l B.P. lio 2 , j i

1 91 Gif-sur-Yvette i 1

1 M. Y DENUIELCU S.E.G.N.

I Electricit6 de France l 2, svende de la Lib 6 ration 92 Clamart M. J . FURET Adjoint au Chef de Service Electronique,R6acteurs Conmissnriat a l'Energie Atonique Centre d' Etudes Nuc16aires de Saclay B.P. No 2 91 Gif-sur-Yvette )

M. Aurtle MEUNIER Groupe de SGret6 des Piles D6partement d' Etudes de Piles Centre d' Etudes Nuc16aires de Saclay B.P. No 2 9] Gif-sur-Yvette M. Jean-Paul MILLCT

-Groupe de Travail de Sarets des Piles CABRI Commissariat a l'Energie Atomique Centre d' Etudes-Nuc16aires du Csaat u C;,e -

B . P. h,o 1 132Jainl-Paul-les-Dumaea

O

_3_

FRA::CE (cont'd)

~ M. J. OULLION (suite) Groupe de 50 ret 6 des Pilec D6partenent d' Etudes de Piles Commissariat h l'Energie Atomique Centre d' Etudes Nuc16aires de Saclay B.P. No 2 91 Gif-sur-Yvette M. Frangois de VATHAIRE Chef du Groupe de Sarets des Piles D6partement d' Etudes de Files

)

l Commissariat i l'Energie Atomicue Centre d' Etudes Nuc16aires de Saclay B.P. No 2 91 Gif-sur-Yvette GER : .. .:2 Dr. Adolf BIRKHOFER ALLE iG:iE Institut fur Mess- und Regelungstechnik Technische HochschuJe MUnchen Arcisstrasse 21 8 Mdneten 2 Dr. BUCHIER Bundesministerian fur Wissenschaftliche Forschung Heusse13cc Bonn l Dr. H. KARWAT Institut fur Mess- und Regelun5stechnik Technische Eochschule MUnchen Arcisstrasse 21 8 MUnchen 2 1TALX Dr. Alberto FERRELI l li% lie Safety and Control Division Comitato Nazionale per l'Energia Nucleare Via Belisario 15 Roma Dr. Enzo IANSITI Director of Safety Control Division Comitato Nazionale per l'Energia Nucleare Via Belisario 15 Roma Dr. Gianni PETRANGELI Safety Annlysis of Industrial Plants Safety and Control Division Comitato Nazionale per l'Energia Nucleare Via Belisario 15 Roma

U V 4_

JAPla Mr. Harumitsu IWAMOTO JnrOn First Secretary for Scientific Affairc Japanece Delegation to the 0.E.C.D.

5, av. Pierre 1er de Serbie 75 Paris 16e THE NETi!ERLAND3 Mr. Robert G. SCHDLVIECK l' I%Yu-y,A6 Reactor Division Reactor Centrum Nederland Petten KORWAY Mr. Kjell Petter LIEN LOdyLGE Institutt for Atomenergi -

P.O. Box 40 Kne13er SP_A IX D: . Agustin ALOKSO l ESFnUZE lie.75, Kuclear Safety Group Junta de Energia Kuclear Ciudad Universitaria Madrid 3 M. Francisco OLTRA Junta de Energia Kuclear Ciudad Universitaria Madrid 3 .

M. Eduardo RODRIGUEZ-MAYQUEZ Junta de Energia Nuclear Ciudad Universitaria Madrid 3 S'. Z D E N Dr. L. CARIBOM SUED 5- Aktiebolaget Atomenergi Studsvik hykbping Mr. Erik JAKSSOK Swedish Reactor Safety Comnittee c/o Kgl. Financ-departenentet Stockholm 2 i

UNITED KINGDOM

. . . . . . . . . _-- Mr. J.H. B0JEK l E0,1a u an- va i Safeguards Division Authority Health and Safety Branch U.K. Atomic Energy Authority Rislev, Warrington Lancs. 1 a, . -

_5_

U::5D K' .:GD0' ( c ont ' d) Er F rank R . FAIEER h0X4LiWUI.2 (auite) Head, Safeguards Division ,

Authority Health and SEfety Branch ,

U.K. Atomic Energy Authority )

Ric3ey, Warrington Lancs.

Mr. A . J . BOURHE i Control and Instrumentation Section  !

Authority Health and Safety Branch j U.K. otomic Energy Authority Ris2cy, Varrington

,uancs.

1 Dr. H.M. KICHOLSON l Safeguards Division Authority Hen 1th and Safety Branch U.K. Atomic Energy Authority Risiez, Varrington Lancs.

UNIrPED SADES Dr. Clifford BECK E M 3-65..ii - - Deputy Director of Regulation U.S. Atomic Energy Co: mission Washington, D.C. 20545 Mr. Edson G. CASE Director of the Regulatory Division of Recc;or Standards U.S. Atocic Energy Connission ~ l Washincton, D.C. 20545  !

1 Mr. Edwin H. DAVIDSON  !

Kuclear Safety Division of Reactor Development and Technology U.S. Atotic Energy Commission ,

Washington, D.C. 20545 '

Mr. Joseph J. DIKUHKO AEC Scientific Representative -

U.S. Enbassy 2, avenue Gabriel

_7_5 Paris 16e Dr. David L. MORRISON Associate Chief Chemical Physics Research Battelle Memorial Institute 505 King Avenue Colombus Ohio 45201 1

l O O. , ~ . . .

6-ual. r a o a . . ._

(cont,aj., .. . . _.

J.r. 2.unzi o o . .:n.ut ,D .1. 0 h1%'/..'. - L. . .. ., (~Eaite) (Chsirann, advisory Connittee of Reactor Safeguardc)

Den:. r;f Engineering Pennsylvania State ' University Penn.

h.r . m2. 4< . .v !1Lo O.. a atonic znergy ,.vivision Phillipe Petroleum Co.

P.O. Box 2C57 Idaho Pnlin w- .

1erano EA0i:P Cbce ccr Mr. ::arko XOCELE Institut fur I;eutronenphysik und Reaktortechnik Kernforschungszentrum Farlsruhe Postfach 947 75 Kar3 sruhe 1 A E i. Mr. F.LSSERA Health and Strety and Waste Disposal Divisic International Atomic Energy Agency Zaerntnerring 11 n ul v.. yienna .

austria ENE e Mr. L.W. 30ZER l j Head of Technical and Economics Division Europesn ::uclear Energy Agency, O.E.C.D.

1

,7 ef -

38, boulevard Suchet

\ '

.sj 75 Paris 16e i ,,r, 1-d i l nme. n,. SOu...dzo,.

.s  !

(, L t .\ Agence Europ6enne pour l'Energie liuc16 aire, d 0.0.D.E.

36, boulevard Suchet 75 Paris 165me  ;

5

v-o 1

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l l Novembe.t 1967

/ / ,o. N.'4,r. ,?)7 CD:GIO??E C:! FAGTO2 S'SE0Y TECEOLOGY Provisional 1.genda of the ThirS Ucotins

  • i l

.V.ed.ordrv Fth

1. Adoption of the Agenda of the Third Heeting. -

l

2. A6 option cf the Record of the Second Meeting.

3 Ecport of the CRESO Specialist Ecotins on the Safety of Prectrccced Concrete Precourc Eicholson  ;

Veccels (Fouirccc, 15th March 1957). Sub- ,

i cequent significant developments.

4. Ecporb of the CREST S socir.lict Eceting on the Reliability of Elcetronic Equipnent and Systc:_r for Nuclear Reactor Safety (Brussels, Vinck j

11th-12sh April 1957). Subsequent signifi-cant developncnts. ,

l 5 Procrocc reporc on rapid trancient studies Millot i eni chock structure interactions.

6. Dcprcscurication of unter reactors. Infor- .

nation on 1:ork carried out in USA, UK, Bouen 1 Germany, etc. Wilcon Thursdey Z th 7 Report of the Working Group on Safety Acccco-nont of Water Reactors. VincP'

8. Gac-cooled reactor accecanent using proba- Bowen*

bility approach. Procrecc and problens. Green French approach to cac-cooled reactors de Vathaire Friday 17th 9 Fact reactor ca'ety status. Outcone of the Air and Xaric mhe hectings.

K131 *

10. ProGrecc in exchcJGo of information on fast heat tror.nfer. Reconnendations for future ENEA -

activities. ,

fi i

l s

e

11. Bibliocrophy of reports on scfety research in Europe. E:.:chanse of information with NSIC.
12. Future cotivities of the Co' ittee.

a 13 Orcenict. tion of the Fourth CREST Meeting a) 1,Genda, dato and pince .

1 i

b) Election of Chairman and Scientific l_ Secrctcry.

M 0

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