ML20235N416
| ML20235N416 | |
| Person / Time | |
|---|---|
| Issue date: | 04/18/1972 |
| From: | Fraley R Advisory Committee on Reactor Safeguards |
| To: | Kaufman P US ATOMIC ENERGY COMMISSION (AEC) |
| Shared Package | |
| ML20235M427 | List: |
| References | |
| FOIA-87-40 ACRS-GENERAL, NUDOCS 8707200062 | |
| Download: ML20235N416 (22) | |
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- t P. D. Kaufman Headquarters Services TRANSI.ATION OF GERMAN TECHNICAL DOCUMElftS
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Reference is made to the conversation of Mr. Rendrix of this office with you on April 7, 1972 concerning our need for translations.
We are enclosing one copy each of the following for translation as soon as can be arranged:
- 1. Untersuchung der Virguange bei der Druckenthastung k- Wassergehuhlter Reactoren-Versuche mit der 11,.2-o-crob-behalter ohne Einbauten Band 1 July 1971 (about 100 pages)
- 2. (As above) Band 11 July 1971 (about 120 pages) l
- 3. Burnoutsessungen in Rahmen sicherheitslechnischer Untersuchungen - Abschlubbericht Teil 1 July 1970 (about 85 pages)
- 4. (As above) Teti 11 July 1970 (about 50 pages)
- 5. Ein Blow-Down Programs fur Siedewasserreaktoren mit j Wasserstrahlpumpen 3RR 89 September 1971 (about 75 pages
- 6. statusbericht uber Prot ene der heute ublichen Containmentkonzepte =l 91 September 1971 (about 90 pages)
- 7. Langzeitrechenmodell sur Ep.'assung der thermohydraulischen Vorgange im Primarkreislau. eines Reakt@rs bei einem Kuhlaittelverlusterfall - MRR 92 November 1971
.: 8.
- Optimierung (about 40 pages) von indirekten Meisleitern - IRR 85a May 1571, , ;' * )
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- 9. Versuchsanlage sur Wiederauffullung und Notkuhlung des >.
N j 72 2 070713 Reaktorkerns - stabbunde1versuche (Niederdruckversuche) ..* * -- S ^
ET RT 34 June 1970 (about 24 pages)
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- 11. Statusbericht - uber die Forschung auf dem Kernreakter Druckbehalterwesen Februmey 1971 (about 105 pages)
- 12. Notkuhlpregrama - Niederdruckversuche Versuche sur Wiederauffullung und Notkuhlung des Raaktorkerns i leichtwassergekuhiter Leistungeraaktoren nach caU L R$ H september 1971 (about 12 pages) s /r
- 13. Versuche Zur Wiederauffullung Und Notkuhlung Des l , Reaktorkans Leichtwassergekuhlter Leistungsreaktoren g
) Nach Grosstem Anzunehmenden Unis11 (Bruch Des l Frimarkuhlsystems) (about 30 pages) l
- 14. Notld1programa - Hochdruckteil der Notkuh1versuche fut DWR and SWR January - June 1971 (about 172 pages)
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- 1. Conference with Commissioner Uchida }
Dr. Uchida is a member of the Nuclear Safety Commission and he was most anxious to discuss with me the proposed NRC rule on general siting criteria. He gave me a very rough draft of the Safety Commissions' l
" Comments on Advanced Hotice of Rulemaking: Revision of Reactor Siting j Criteria". They take the proposed siting criteria very seriously and feel very strongly that these criteria will give rise to many difficulties for the nuclear power program in Japan. It should be pointed out that the RSK had the same concern for the program in West Germany. These concerns were expretsed very forcefully during their recent visit last September in l
Washington, D. C.
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A copy of the rough draft given me by Commissioner Uchida is 1
attached as Appendix I. -
Dr. Uchida also expressed particular interest in Class 9 accident studies and related backfitting requirements. He was also aware of the l
l discussions in the U. S. on the hydrogen burn problem as related to contain-l ment integrity.
1
- 2. The Incident at OHI Number 1 There are two UHI, ice-condenser plants, at OHI. These plants 1
I were built under the supervision of Westinghouse for the Kansai Electric Power Company. While the incident of concern occurred in July 1979, a representative of MITI brough it up as a fairly important question regarding quality control in the United States, j gj9 - /# ' I ACllSOFRCECDPY _ _ _
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- F' The incident was described to me by Mr. N:buaki Mori cf the Publie' 1.-
Utilities Department, Agency of Natural Resources and Energy, MITL The ,
l reactor was tripped by a spurious signal from the RCP trip circuit. Then a spurious large pressure difference between steam lines initiated ECCS pump flow. This spurious signal was the result of the failure of the pressure 1
l sensor on the steam line; this sensor was housed in a bronze tube which failed. By specification the tube was supposed to be stainless steel. The pressure signal gave a spurious indication of low primary system pressure.
I l' Since the system pressure was high, ECCS flow was supplied only from two l high pressure charging pumps. Fortunately, the operator intervened and shut these pumps off before the system went solid. Further material on OHI number 1 and the incident is given in Appendicies II and III.
- 3. Meeting with Dr. Nozawa, Manager. Division of Reactor Safety, JAERI.
1 For our meeting, Dr. Nozawa made available the research engineers in his division who are responsible for the major facilities in the reactor safety research program. These engineers give the impression of being very able, and t1ey all have an excellent understanding of the limitations, .j l
as well as the potential capabilities, of their major facilities. I mention this understanding because most of these facilities are involved in the' 3-D International Program. Comments have been made on this International Program in the ACRS reviews of the United States reactor safety research program. The Japanese would, I believe, agree with the sense of most of these comments. As a general comment let me make the observation that J
these major facilities, which I shall describe, are usually very well-built, suffer from an inadequate complement of research engineer staff. At this m_ ',s:, we discussed again the possibility of NRC Engineers coming i to Japan for a. -wended stinf (say 6 months to a, year) to participate in the
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1 safety research progrcm. The JAERI p;cpb cro cnthusinstic cbout such a possibility. I am confident that it would contribute importantly to reactor safety research.
A. The Slab Core Test Facility The facility described in Appendix IV 4 is nearing completion and was designed to study two-dimensional effects in refill and reflood following a LOCA. There are 2,048 rods in the vessel which are of the same size as in W 15 x 15 fuel, but they are grouped in 16 x 16 assemblies. The facility has 10 Mw power available for electrical }
heating of the rods which are full-scale in length. The downcomer width is scaled and is 250 mm. The Japanese are aware of the difficulty with heat sinks at the walls in a scaled facility and the outside *valls are insulated to reduce these effects. The low temperature of the slab walls give an ,
incorrect heat sink which is only partly overcome by the honeycomb insula-tion. Other errors in the model pointed out by Mr. Adachi included: No t l
l heat source from steam generators; pump simulator has no driving force; bypass modelling is incorrect, the initial condition is steady state in the model while in the prototype it is in a transient; the containment volume is small so that it is not prototypical. Mr. Adachi emphasized that the tests cannot be directly translated to full scale. It was refreshing for me to j hear such a frank evaluation of the test facility. Further, the indication was that the hope that TRAC can bridge the modelling defects is clearly at this time optimistic. j 4
I Tests in SCTF will begin in April 1981 and the first core will simu-l late the W type and will examine the effects, in two-dimensions, of balloon-
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ing and flow blockage. The second core will simulate the German KWU core, l
and the third core will simulate a W core unblocked and will also include a
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vent valve installation to simulate the B&W type. The results from SCTF will be very interesting. ,
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' As a fin 1 comment on this subject, I suggact that tha Inst paga cf
.r .' Appendix .IV and the next to the last page be compared by the reader. Again s
I find this comparison interesting as an expression of an expert attitude in l.
Japan on the TRAC " bridge" (see also the ACRS Safety Research Review).
B. The Cylindrical Core Test Facility This facility is described in ,
l: Appendix V and. like SCTF, is S low pressure facility which was designed 1
originally to' study the refill and reflood phases of a LOCA. One might note .
that the rodr are grouped in 8 x 8 bundles, rather than in 17 x 17 bundles; this was done to get more upper plenum structure in the vessel. The 32 bundles give us 2,048 rods as in SCTF with the same electric heating capacity. JAERI has made an extensive study of the effect of exterr.a1 thermocouple and found a large effect, and consequently have gone to thermocouple embedded in the clad (again see the ACRS Safety Research Budget Review).
Figure 1 of this appendix gives a very good view of the facility arrangement as compared with a PWR. The other figures included there
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are of interest including Figure 5 which compares CCTF data.with FLECHT data.
C.ROSAIV This facility is described in Appendix VI and is a new facility now under design and development. The present schedule calls for construction to bh completed by the end of FY 83. The budget cost her the facility is about130 M. It is also now designated as the Large Scale Test Facility and it is the only facility which is considered to be'an integral test facility in contrast with the other facilities which are considered to be i separate effects tert facinties. The ROSA IV program also includes a Two Phase Test Facility.which I shall discuss below.
The LSTF will operate at full prototype pressure,16MPa, and temperature, 598 K. It vill have 800 heater rods, a full-height core and m._._ . _ . _ __ _ _ _ _ _
a volume ratio to e PWR of 1/48. It will hava two activa full-hsight ctsam generators and a full-height pressurizer. Clearly, this facility is admirably
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suited to the study of small break LOCAs, and it is designed to cover the The range of breaks up to 10 percent of the area of the largest PWR pipe.
facility is also well-suited to the study of natural circulation heat transfer under a wide range of conditions. It should be an outstanding integral test facility well-adapted to our current views of important problems related to LOCAs.
The program, as has been mentioned, includes a Two Phase Test Facility which is directed toward code assessment. The group in JAERI with whom I met have not been overly impressed with TRAC. They are much more favorable inclined toward RELAP5, as has been the ACRS in its Safety Research Reviews. The TPTF will allow separate effects tests for improvements in a 2V, ZT code.
D. Mark II Containment Response Test The last discussion with the JAERI research personnel was on the full-scale Mark H Containment response tests. The presentation was led by Dr. Shiba who has been running these tests, and who will present a paper on his findings at the November,1980, Light Water Reactor Safety Conference in Germantown.
The facility is described in Appendix VU.
So far in FY 80, fourteen blowdown tests have been run and the tests should be completed before the end of FY 81. General results, only, were reported by Dr. Shiba. He indicated that his measured loads are all bounded by the load definition report in Nureg 0487. All of the loads after vent clearing were below those in that report. He also finds that his measured diaphragm pressures are s'maller than the values reported in the EPRI data.
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6 Further work in the facility will be concentrated on the chugging- .
problem. He made a comment on the question of the simultaneity of the chugs to the effect that his data indicate a definite time interval between chugs of the order of 50 msec. This finding is of importance for the deter-mination of total chug loads.
- 4. Visit to' Energy Research laboratory of Hitachi, Ltd.
This Energy Research Laboratory is at Hitachi City and, while Hitachi City is a some distance from Tokyo, my visit there was very worthwhile.
The Hitachi Company is a licensee for the construction c.f boiling water reactors.
The Toshiba Company is airo a licensee for the construe-tion of boiling water reactors. The licensee for PWRs is Mitsubishi which should be visited some time. The JAERI engineer who accompanied me to Hitachi City mentioned that Mitsubishi has done some work on ice-condenser PWRs.
The research on BWRs at Hitachi is rather sensit'ive. .JAERI is not involved in it and the financial support for the program comes from Tokyo Electric Company and a few other Japanese utilities. Ihave been told that Tokyo Electric is the largest private electric utility in the world, but Ihave no figures. In any case the utilities involved in the work at Hitachi regard the safety research as their private program, and Hitachi of course respects this view. However, I must add that both the high management and the - . .
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l research engineers were very forthcoming and certainly gave me the im- l pression of not holding anything back., My only regret concerning this visit l
is that it was too short since it would have been profitable to have a more extended visit.
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'. The information the Hitachi personnel gave me should be regarded as proprietary for obvious reasons and my comments should also have limited distribution.
A. Two Bundle Blowdown Loop The ACRS has recommended in its Safety Research Budget Review that there is a real need to replace the present TLTA facility for LOCA testing of BWRs. The present TLTA has a significant flaw in the height relationships. It is rather remarkable to learn that the utility group in Japan has funded a TLTA type of test facility at Hitachi. The Japanese facility has full-scale height and is much superior to anything available to the NRC. The Japanese facility is full beight up to the steam separator; il e dome height is reduced but this reduction is not considered significant.
Some details of this facility are given in Appendix VIII. It will take 2 full-size 8 x 8 fuel bundles; these are, of course, electrically heated with 10 Mw of power available. The designers of the facility have made en effort to improve the simulation by electrically heated rods with indirect heating. The nichrome resistance wire is packed with electric insulation material. The facility has 2 recirculation pumps,,and 2 jet pump s. The ECS system has 1 HPCS, 2 LPCS, and 3 LPCI.
The facility was completed at a cost of about $13 M and is now in the shakedovm study period. The IDCA tests will begin in November with the DBA.
B. Full Scale Spray Test Facility The spray test facility has been in recent use to study the behavior of BWR spray cooling. The spray facility is full size and corresponds to the GE 251 core. The spray arrangement can be used to get spray behavior over the full 360* with air upflow. With steam upflow, the area coverage is limited to a 60 sector because of the amount of steam flow available.
. l This spray distribution study includes a fair cmrunt ei antlysio l The model .
which began with the development of a one nozzle flow model.
used one-nozzle observations to develop a code which would describe the j
Next an " interference model" was developed observed spray distribution.
which was compared with experiment. A multi-nozzle code was then devised The code seemed to give which was compared with their measurements.
This research should be given further review within reasonable results. ,
the limitations of the proprietary nature of the code.
Finally, I should add that the Hitachi research engineers in their i study of the DBA in the Two-Bundle Blowdown Loop rely, not on TRAC, Since, so far as I am aware, no BWR but on a BWR version of RELAP5. !
version of RELAP5 has been released, the Hitachi people went to a private l I
There group (Intermountain Technology, Idaho Falls) to get this version.
was a small remark regarding the high cost of getting this BWR RELAP5, l
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JAERI FULL-SCALE MARK- CONTAINMENT RESPONSE IEST i POOL SWELL STEAM CONDENSATION OSCILLATION PRESSURE AND TEMPERATURE RESPONSES l FY-1977 PLANNING AND IEST FACILITY DESIGN FY-1978 TEST FACILIYT CONSTRUCTION FY-1979 BLOWDOWN TEST (COMPLETED)- IM FY-1980 BLOWDOWN IEST AND FACILITY MODIFICATION c4 pp' FY-1981 BLOWDOWN IEST AND IEST FACILITY REMOV'AL ,
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MAJOR ACTIVITIES IN FY-1980 MODIFICATION OF MEASUREMENT SYSTEM l CONTAINMENT STIFFENING ' CONTAINMENT OSCILLATION MODE TEST CHUGGING TEST e I b
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PAse 7 . i Ditaensions of Test Facility and Reference BWR Typical 1.100MWe Full-scale Mark H
- cRT Facility . BWR (TOKAl-2)
- Primary system 7.0/286 7.3/289 coolantpressure(MPa)/ temperature (*C) 14 /0.5* 150/9.6 ,
coolant volume (m3)/ mass (ton), steam
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water ! 1(Blowdownline) 2 /4 Number of recirculation lines / main ste m lines 610/660 Max 240mm ID(ra) of. recirculation line/ main steam lute 380 21 Total coolant enthalpy(MJ) ntainment ' ' 500/150 385/171(104WW) Design pressure (KPa)/ temperature (*C) 5400(including Drywell volume (m3) 329*(5918)' 25 Wetwell air space /wetwell pool volumes {m3) 4 91)/ 71) 4100/3400 4-7 106 Number of vent pipes 14.0/0.59 13.5/0.59 Vent pipe lenght(r.}/:3(r) 3.4* 3.4 1 Nominal vent pipe submergence (m) .~ 1.91' - '31.55 Total vent pipe flow area (m3) 24.81(447T 450 Wetwell pool surface area'.m2) Ratios
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Transmittal of Geneva Trip Report Mr. R. E. Simonds, Director Contracts and Support Division
- Idaho Operations Of fice U. S. Atomic Energy Cortission j Idaho Falls, Idaho 83401 -
l
Dear Mr. Simonds :
Reference:
R. E. Simonds to C. M. Rice, " Foreign Travel - G. F. Brockett ," ! August 18, 1971. Enclosed are thirteen copics of a trip report on G. E Brockett's participa- ] ' tion in the Fourth United Nations Conference on the Peaceful Uses of Atomic Energy, Geneva, Switzerland, and subsequent visits to the German water reactor safety research facilities. We regret that a significant volume of reference material, which was mailed in Europe, hr.s not yet arrived and thus the enclosed report was unable to specifically reference these very recent reports on results of exocrimental work which is relevant to the U.S. water reactor safety program. Your attention is directed to the conclusions and recommendations section, pages 28 and 29, of the report uhich in sucmary fora cites examples where the accelerated pace of the safety research activities in Europe has brought the status of their safety programs to a point where the U.S. could benefit f rom an enhanced interchange of information among the parti.cipating technical l personnel. ) Very truly yours,
./
jlm h f/QOy%y'j"f C. M. Rice 1 .4 g 3 General Manager f Enclosure cc: Director, Contracts and Support Division .J , Director, DRDT g l Assistant Director for Nuclear Safety, DRDT g l Assistant Director for Project Management, DRDT LOFT Project Manager, DRDT W h LOTT Program Manager, DRDT P' i C,s Chief, Thermal Reactor Safety Branch, DRDT Chief, Water Project Branch, DRDT C h Director, Division of Reactor Standards t.":d O A Director, LOFT Project Division, ID Director, Nuclear Technology Division, ID Manager iUlSP0Q,(
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interoffice Correspondence 1 %sd1 !1 Novecher 17, 1971 , I Geneva Participation- I Trip Report B ro c-92-71 -{ J. C. Haire j HQ-Anne:: 3-9 , i Please find attached the report covering my trip to the Fourth' United l Nations Conference on the Peaceful Uses of Atomic Energy, Genev.a. Switzerland, and describing discussions resulting from other. official contacts and visits. l ( As a participating author for the paper, " Status of United States Atomic l Energy Commission Water-Reactor Safety Research and Development Programs," ) the subject matter-centained within the report is principally directed toward reactor safety and power reactor performance session's, although observations are of fered regarding other sessions I attended. My-reflections f rom discussions with others at the conference frora different countries who were working in reactor safety are also reported. In additior., I visited German f acilities working on reactor safety and licensing methods. The visits in Germany are discussed in considerable depth since they provide the basis for recommending greatly increased information interchange between the U.S. and European reactor safety research activities. The escalation in European reactor safety research is, or will shortly be, providing information in areas not being investigated in the U.S. and so should be of high interest in this country. Additionally, some significant dif ferences between U.S. and German siting philoc,,ophy are cited. The trip was stimulating to this author. I considered it _ highly worthwhile both in clarifying the role of the NRTS reactor safety work and the overall U.S. approach to reactor safety. The delay in submitting this report is due in part to the absence of certain materials shipped from Europe which still had not arrived as of this writing. The materials en route contain additional f actual information and references ~ related to the conclusions and recommendations which follow. Regretably they cannot be cited in detail at this time 7
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TRI? REFORT ON ATTEE?ANCE TO 7:!*2TH 'J ITI ::ATIT;S CONTEP.ENCE CN THE FEACEFUL USES OT- ATOMIC ENEEGT GENEVA, S'CTZEELAND AND VISIT 3 TO GEEAN '..'ATER REACTOR SA2ETY EESEnRCH FACILITIES' September 16, 17, 18 IN TE0 DUO!! ^... I participated'in this conference as coauthor and presenter of a
. paper' entitled, ?3tatus of l' nit.ed States Atomic Energy Commission ~4ater-Reactor Safety Research and Development.?rograms,+ by G. T. 3rockett, E. 3. Case, _. C. Eaire, S. O. Johnson and M. A. Rosen. The principal subjects of this report are reactor safety and power reactor performance, althou;h a number of related are as will be mentioned.
The report is divided into the followin'; sections : 313hli; hts of the Geneva .:onference Discussions with T.eseerchers in Reactor Safety from Various Co'r.tries Tisit to Scr:an 'Jater Reactor Safety Research Facilities Conclusions and Reconnendations EI SLIG.i!3 37 GENE'lA C0"TERENCE
";ene ral The opening session of the Fourth Geneva Conference was indeed impressive and inspiring. Recollections in the talks of the notables, many of whom participate in three previous conferences, illuminated. the significant -
accor:plish::ents of the seven yer.rs sir cc the Third Geneva Conference. The-doninant ccacern expressed in the opening discussions was for the development and implementation of clean, safe nuclear power consistent with' protection for the environment. However, nowhere was there as great a concern for the environmental issue as with U.S. delegates.- Many of the opening speakers pleaded that all' nuclear activities relating to the use of atomic energy for weapons cease. Also prominent was the recognized need' for new and ef fective techniques for informing the public concerning nuclear power and for puttine: then into practice.
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ATTAC11 MENT Broc-92-71 017ICIAL v.93 c.17 Page 2 i j Evident in many opening remarks was the tendency toward developing i and extending nuclear energy as an end rather than a means. Only one ! speaker, Professor I. I. Rabi, expressed concern for the need to develop l j safe, clean, reliable power while at the same time considering the need - l for conservation of all resources. In tany ways the concerns in the l opening session were summarized by the remark of the representative of the United Nations Secretary General in which he said "what science l has done for mankind must now be tempered by those scientists to protect mankind." I With regard to organization of the conference, the ability to proceed through a planned schedule, acquire information, obtain materials and l services was good, particularly when considering the international aspects l with the attendant language and custom difficulties. Instructional material was adequate and the assistance supplied by the staff at the U. S. Iussion was friendly and competent. The largest single difficulty was the lack of a ecs: age bocrd or other ennmunication schemes at the conference proper l for caking contact with individuals both from and outside the U. S. It was ainost impossible until late in the first week of the conference to determine in which hotel a group or individual was staying. Finally on . I the first day of the second week a message hoard was installed at Palais de Nations. In terms of the environment for the presentation of papers, translations were good and the sessions usually held to schedule. Unfortunately the time allotment for discussions in the technical sessions was too short to promote a lively dialogue. In the more fortal philosophical sessions ' l presented in the general assetbly hall, the tire allotment for questions I was much longer but the climate was too formal for lively dialogue. The result was some ' artificial or fill-in discussions. ; Oater Reactor Safety There were three principal categories of safety considerations which dominated the three reactor safety sessions. They were safety philosophy, saf ety experience, and safety concerns. With regard to philosophy, there vas a sharp contrast betveen the positions of many foreign participants and the U.S. The U.S. position connated heavy reliance upon the excellence l _ _ _ _ _ _ _ - - - _ - ___a
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OFFICILL DE: C LT I i of engineering practices to provide for total plant reliability. Foreign participants were genuinely concerned with "what if" questions such as the effect of external forces (aircraft) and acts of God. These considerations ul tinately f ocused upon the questions of containment penetrations, primary vessel failures, and core meltdown. Germany in particular, and also Sweden and other 3 scandinavian countries, acknowledged that they must contend with net ropolitan siting now. Further, nany of the nuclear power sites in Europe are chosen on rivers to provide an economical sink for waste heat. European ; rivers are so polluted from chemicci plants and other industrial wastes i that the fich have Jong gone; cicanin; up rivers ultimately becomes an ! international problen as rivers form international borders in many cases. Further, nost of the rivers have ci:12s along them, Thus the sitin; of reactors on rivers creates circumstances in which distance will not be l l ava12able to attenuate a nassive release of fission products before the 1 ra il ca cti"i t) rea:hes a city or international border. As a result, the
- uropeans are very concerned with primary vessel rupture and its prevention. i 1
A Swedish paper proposed an external concrete vessel in conjunction with a l 1 thin totallic vecsel for EPR constru: tion. Vessel fallute for this configuration ] is presumed to alucys produce a 'sec11" leak situation which perndts an ] l crderly plant shutdovn. I 1 a l There was significant acknowle dgment and reference by non-U.S. speakers to the lead role of the U.S. in the producing of codes,. standards, Q. A. -3 i inspecticn techniques, etc., essential to reactor safety. These speakers j l observed, however, that it was time f or the European countries to speed l up their own development of systems to provide reactor safety assurance beyer.d that of the U.S. to suit their own particular metropolitan siting needs. For example, Cercan limits for fission product release during operatier. are core s tringent than current U.S. limits. In addition, there is strong European interest in utilizing vaste heat to heat cities and to supply certain industries with process steam. A Russian paper noted that we have teached the physical limit of light-water reactor size because of the problens of transporting reactor vessels. To provide increased reactor output, the author noted that the power density must be increased and the flux flattened with the attendant req ui renen t for increased consideration to reactor safety.
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l Another area where there was difference in approach in the U.S. and f creign countries is the applicatica of a probabilistic approach and the use l of reliability analysis techniques in determining the adequacy of plant safety. The U. S . considers this approach less realistic .since both reactor , syrtc: cenponents cnd reactor system designs are still evolving and thus suf fi cient J on;: tcrn experience data are not availabic. European groups, cr. the other hand, are serious about peoling coeponent performance data and applying it as best one can. Many of the long established inspection organizations in Europe have been collecting related data on non-nuclear ) i sys ten for years. They consider qualified use of this experience justifiable l until more specific and relevant reliability data is available.
':ith regard to reactor safety experience, the world wide record is renarkable. Failed fuel cladding, coolant envelope leaks, and corrosion problens vere in evidence in first generation plants. However, safe shutdown or lir.ited operation without saf ety dif ficulties was always possible.
Several papers were presented reviewing reactor malfunction which resulted 1 In the l f ron flow ir.duced vibrations of main reactor vessc2 internals. l Italian Carigliano Power Station, the sparger ring injecting soluble f poison into the vessel severed. Poiscr injection into the primary pipin; with at tc.. dant uncertainties in the adequacy of poison nixing within the vessel was necessary. In the Trino Vercellese IWR reactor flow vibration l caused loosening of the thermal shield support structures. 1.oosened bolts vere trcnsportv1 into the bovl of the steam generator where they induced prinary to secor.dary stean generator leaks. Leakage was the first index of the reactor system problen. Similar internals failure were reported in ar.other reactor of this type in France. legr.rding the third category, reactor saf ety concerns, the European connunity was outspcken in their concern for the ef fects of containment Transient pressure loads which result f rom a loss-of-coolant accident. differences between the various compartments within the containment could I produce loads suf ficient to damage internal structur'es within the containment and subsequently endanger the piping and control systen associated eith operation of energency core cooling equipnent. A Norwegian paper preposed underground siting primarily to preclude damage to the containment by
ATTACHMEST i B rec 71 l Pare 5 U#Il0lll U- IE I I external forces such as chemical explosions or aircraf t. The Japanese understandably are concerned about the seismic question and the criteria for antiscistic design. They have lovered the equipment design threshold i to less than 0.2 g acceleration fron an earthquake until design of antiseismic j I structures and damping schetes has been verified. Foreign participants nere net as concerned about the emergency core cooling performance question as were U.S. participants. Part of the difference, as will be discussed Ij lat e r , results from the much lower cladding temperatures calculated by European groups for their INR and BWR reactor's. The Europeans calculate ] I much hi;her core flow for the cold leg break. However, their approach appears to have deficiencies in system nodalization and in geometric chnracterizatinn vhich vill influence core flow analysis, Future 3NF 's , to be built in Europe by Germany, will elininate external circulating I loops by incorporation- internal circulating pumps within the reactor vessel. This cc,fijaation nearly eliminates the pipe rupture question for the B22 1 and thus r:d ;ces the denand on energency core cooling systems. a rr c r .m a e t gr : e rf ormar ce
?) ant r. ailabi2ity f actors throughout the world stood out as testimen" j l to the total accomplishments in the nuclear industry in the seven years l
l since the Third Geneva Conference. :lant availabilities in the najority { I of the reported plants were ranged to 70% and above for significant periods ) of time. The major reported causes of down time were related to malperfornance l of non- nuclear components within the system, which is indeed a paradox. From a nuclear standpoint, the expected burnup for many cores was not achieved due to various medes of fuel cladding failure and to corrosion problems. Tuc1 cladding f ailures vere by f ar the most plaguing difficulties acknowled;ed in the papers. In spite of these, however the plant operators became resourcef ul in 3 earning how to live with a limited amount- of f ailures and still not exceed a radioactive discharge limit or produce radiation levels ! which would prohibit refueling or other maintenance activities. 1 Exposition The exposition was both awesome .and disappointing. Overall the exposition was a significant monument to Hadison Avenue technolo;ies, with ! the displays lavish and highly sy6bolic of extremely abstract art forms. 1 i
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- Page 6 The U.S. fusion display and the space exhibits were commendable. The on-line t l
moon measureraent data display and the direct information retrieval system l frca OFSL were "one up" relative to other exhibits. However, the impact on ne of the U.S. cxhibit generally was that it lacked substance especially j whcn cc pared tc other, less spacious, exhibits of other countries. The i Russian exhibit was slightly more technical than that of the United l States, but was equally disappointing in explicit guidance. It was i difficult to find someone within the Russian exhibit to answer questions. j The nost technically comprehensive were the German and British exhibits, while the most ably staffed vere the Japanese, Swedish, and British f exhibits. In general the stall countries provided good displays en current technological hardware; whereas, the large countries provided la >ish, artf ul symbolic representations en their current research activities . ! To the representatives of governments, utilities, and other non-nuclear specialists the displays may have been informative and appropriate from I an impressionistic *1cipoint. Tros c technical vicwpoint, the dicplays were generally c disappointment. DISCUSSIONS 'JIT:4 nESEARCEEES IN REACMR SATETY FROM VARIOUS COUNTRIES _ Contacts we have developed in Idaho with various safety researchers in Europe produced several ad hoc meetings at Geneva where there were useful information exchanges. United Kin;drn i j A discussion was held with E. Gilby of the UKAEA at Risley, England, regarding their work in water reactor safety. Of specific interest was j the status of the deprcssurization studies by A. R. Edwards. Edwards has ccaputcrized his dspressurication efflux model to incorporate a bubbic nunber and an arbitrary constant. He has applied the model to the initial surge of water level in the Battelle-Frankfurt tests and has established a remarkable fit to the measured temperature and pressure data. The result i is derivation of the superheat required to support the initial flashing process. This todel has also been reasonably well confirmed by decompression l l experiments carried on in England during the last two years which have not as yet been reported. Gilby explained that their incorporation of automatic data l processing techniques using digital computers has been a slow tedious process. I Only recently has the data from the tests been reduced to a state suitable for
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fer publishing. He expects a report on this work (carried out at Foulness) to be available in early spring. The experimental work. covers depressurization f rom ! pipes with exit diameters ranging up to 8 inchen. My discussica.; ,ei th Gilby, R. Tarme r (Direct or UKASA Safety Studies) and others indicated a rencued interest in light-water reactors for power j generatin; purposes in . England. Apparently England has a desire to develop a cenpetiti- e position for marketin; light-uater cooled-and-moderated power reactor systens. Eccent discussions with R. E. Duffey of the Central : 1 Electric Generating Board (CIGB) also confirmed the desire of this only l UK utility to acquire experience in reactors other than gas cooled and hea;/ water Eeretofore their interest in water safety problems had j been nativated primarily by the English pressure tube D 20 noderated 1 rea ct ors . I 1 l Italv l Discussions were held with~Dr. Enzo Iansiti, Division Director for Eacuri ty n,d rnnt rn19, rNr:, and Mr. Giovanni retrangeli also from C:EX l on the heat transfer program in chich transient CEF is being studied l l l in a dual-ecolant-channel dif ferential-power f acility. This progran is l being, or has now been, terminated. Althou;h the configuration was nearer te pressure tube geometry, the fluid conditions were useful for j determining the relationship between transient and steady-state CHF beh cior. T..cir results indicate that CHF conditions under transient conditions differ little from CHF conditions observed in steady-state or departure f rom steady-state tests. They are considering other j heat transfer studies in geometries more closely related to the bundle , reon e*.rv o f current licht-water reactors, but are very limited en funds. Thei: heat transfer information is presented on tape and should be available threagh the Ispra Library. Sweden 1 talked with Dr. L. Carlbom, Chief of the Safety Department of the Atomic Er.ergy Ccapany, concerning Sweden's approach to reactor safety as well as the disposition of the forthcoming blowdown tests in the l'arviken reactor and containment. Regarding safety philosophy, the Swedish are very I __m_m.___m__ _ m_ _ - _ -
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. B rc c-9 2- 71 , CFrycyjg py; g;;7 rate 8 conce rned about pressure vessel failure and therefore requirements for a core cat che r. Their concern for this subject was distasteful to some from the U.S. , however, I pursued the motivation for their apparent strong interest in the " ultimate safeguard." Carlbom's response indicated, not unlike that of the Germcns, that there is no " remote siting" in Eurcpe. Furthc r, with the potential for using waste heat fr'om reacters, c1cse in metro;311 tan siting is attractive, therefore, for heating cities and providing industrial steam. They are continuing to give at tent.icn to the n.citdewn q uest ion .
The Marviken reactor, which was originally built as a direct cycle natural circulation reactor with heavy water moderation, never started up. It is now to be used in a program of full scale safety experiments whiph ,:111 include the following studies (1) containment response to various size ruptures in the primary piping system. (2) iodine transport within the centaintent during and af ter blowdown, (3) mechanical and electrical corponent 'cehavior during and fol3 owing blowdown, and (4) containment leakage behavior durin: and following blowdown. (It is worthwhile to note here that a lively exchange developed during one of the Geneva reactor performance sessions between the governmental and commerical organizations connected with the thrviken reactor in which the commerical representative accused the gover ncnt of a " lack of follou threugh in demonstrating and confirming this reactor concept." The government representative countered with the corment that "had the plant been built to codes and standards similar to those enf er:e a in the United States, the plant could be working today.") With the L]cwdown program at Marviken, the Swedish AEC is trying to l exploit this f acility to contribute in some areas where they feel there is a lach ef information at large scale. The cost of this program is current ly es t inat ed to be appro:<imately $1,000,000 with the program continuing over a year interval . They are soliciting support from other European agencies and have contacted the U.S. with no apparent indication of interest. The reactor system configuration has many of the characteristics of the W r,. In the pregram blowdowns for break sizes up to 25 centimeters at dif fe rent elevations and locations will be conducted. The quantity of air carried fro the dry well to the wet well will be determined. The 3vedes are interested in making good measurements of the fluid behavior l __ --- - --- - - - - - - 1
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- l and containment structural response. A complete cet of reports en the
, program was provided. This program should be reviewed in some detail because it appears that a small U.S. contribution could produce useful information and could Icad to a core rapid test program. Tests in the facility must be completed in one year so that the plant can be converted to an oil fired steam supply sys ten.
F ran ce In response to c query on loss-of-coolant accident ntudies in yrance, . i I 7rench cent acts indic;te, that a rather in-depth program was being planned . at the Grenoble f acility by the heat transfer section of the Department of Safety. Subsequent discussions with S. Fabrega and M. Courtaud in the ] U.S. confirmed the French interest in developing in-house capability in water reactor safety. Th2y ultinately plan what appears to be a sophisticated TLECHT progran but this may not become a reality for tuo to three years. The French, cuch 31ke the Germans , feel strongly that the basic processes of abnornal ; reactor behavier should be understood in the laboratory before engineerin; j sca3e, inte; rated errerirents are carried out. Consequently, they are ) studyin; tube type heat transfer to deternine the conditions which control O:F. They eventually intend to build a 25 rod bundle capable of operatir.; up to 2000 psi under ?7R and EUR fluid conditions. Their current safety evaluation ecdes are similar to those of the U.S. in that their blowdoun code, CEF2ES (derived from the German BRUOO , protidcs the hydraulic inf ormation vnich is then used to derive their core thermal code, FLICA, l dhi:h is sinilar to CO3RA. Their codes are develeped at Saclay in an organization administrative 1y rcrote from the experimental institute j in Creno.iei Termany Discussions were hcid with various nembers of industry, governtent, and regulatory groups includin, . Birkhoffer, Ec11crman, Sicpic, Voyt, ar.d Karvat. Specific discussions relating to the safety research work and the Gernan safety assessnent approach will be addressed in the next s e ctior. . i Connents at this point, however, are appropriate for identifying key l differences between U.I. and German practices in emergency core cooling - l l design as applied to B'JF.'s and TWR's. ; i l l s-
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The most significant difference in ECC design for the most current TIS's is the discharge of the accumulators into both the hot and cold legs of each icop. When queried about their rationale, a Siemens representative simply indicated that they were concerned with the steam binding problem for a co.!d lei; break so that any water tchich could be delivered to the hot Ic;; was t ound to do r. ore good than water delivered to the cold leg only. Tne res; case to my question concerning the potential for " pipe plurg:ing" uith he t 3 cg injcetion.uith the attendant probability of vorsening the stean bir. ding situation indicated that the German's had not given much consideratit.n to this particular situation. Also significant was the Ger .an intent tc reduce the injecticn pressure at the accumulators to 200 psi or so to reduce the loss of ECC inventory which occurs under large break tonditier.s '11t.h the more conventional 600 psi system. For sma'1 1er breaks where there is lack of ECC if the accumulators do not initiate until 203 psi, the Gen.an's are incorporating a high pcmtre injection systcm in 723 designs which vill deliver at 5 times the rate ef U.S. designs. , In discussing with the German's the adequacy of choked flou calculations usin; ;:ondy or liv.e nethods, they agreed that for systems with relatively small pipes, the Moody method significantly underpredicts discharge early in the blowdown and overpredicts later in the blowdown. They offered that an in-depth study o.f World War Il Geman water rocket data has indicated that f or very large pipes, such as would exist in current LPWR designs, the : cod: method.with a constant corraction factor, should be more applicable than for dischar;es from smaller pipes. The Germans feel a significant innovation has been it.corporated ir.to the EUF. configuration which improves both reactor performance and safety. The elir.ination, af ter the 67 product line, of any major pipes penetratin; the bot tom plenum region, pemitted all breaks in a BUR to appear as hot Icg breaks and thus provided a slight edge for the BWR over the P'.!p. trith regard to energency core cooling delivery capabilities. The current German version of the bMR's has further extended the assurance of protection against the loss-of-coolant accident in two ways. First, the incorporation of internal circulating pumps (in place of the jet pumps) eliminated the
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I l potential for recire line breaks. The only pipes which could now contribute 4 to a J ess-of- coolant accident are the four relatively sna11 sparr.er line
' rake up pipes and one or two stean outlet pipes Since the stean outlet pipes are fitted with flov limiters proxinate to the pipe connection at f
the vesse], the b3 0udovn tine for all pipe breaks vill becone relatively 1 lon: f or all break locations unless the connection of the steam outlet pipe nt the vessel becone severed. The shortest bloudoun time for all but this last situation is calculated to be in excess of two minutes. Conparison of this blevdoun duration to an approximate break size in a U.S. M;F. vould provide for good cure cooling during blowdown and a nentaxir.g denand on the e.eriency core cooling systens. The AI? staff is currently cel alatine the S.e1 cladding temperatures for this internal pump configurative-to not exceed 1700-1400'T under the vorst circumstances. Further, this neu reactor systen configuration elininates the need for emergency core coeling spray headers above the core. The sparger ring is non divided into four secters. T.ach sector is supplied f rom a totally independent sys tcr su:h th t during norrh1 nperation the four sectors provide the necessary nakeup water for turbine steam generation. During the loss-of-coolant accident. however, the sparger sectors sinply deluge the annulus between the core shroud and the vessel with emergency coolant. As the internal circulating pumps are more characteristic of an axial fic or screv punp the pressure drops from the shroud annulus to the bottor plenum is very loc ..s a re s ult , the energency core cooling for this cor. cept c: a ; ; is botto:a flooditg for break conditions which will aluays be characterized as snall hot leg breaks. The stean binding problen cannet : exist and the whole sequence of events appears to promise an energency core cooling approach with a great deal of nargin relying on only the assur.ption that the sparger nakeup punts will continue to be driven by diesels, sona of v1.ich are always in operation,
'J:3 I T -'? O E P W : t?!.TER ?I!.CTGR SAFE ~' PISEA'1"H FACILIT!EE In t roduct ion The visit to the German reactor vendors, safety research facilities, j l
and regulatory institutions was especially beneficial as a result of three sonexhat l l unrelated situations. First, the Germans have placed seven or so participants l l 1 l
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l Face 12 in the ::p.TS safety pro;;rcm for training over the last ten years. This, ! through corresponded.ce and subsequent visits on their. part, has provided the SRTS relatively good insight into the approach and progress of the German water reactor safety effort. Second, the CPIST specialists meetint; l of June 1969 on reactor systen blowdown analysis and experiments demonstrated that the Cernans were playing ene of the nore active roles in Europe towards ; developin; their orn capability both experimentally and analytically in water rcactor rafety. This situation provided then uith a keen awareness of where they were in regard to the U.S. , Japanese, and British and since that time they have further endeavored to get on top of the principal issues in water reactor safety. Third, work had been started in 1969 by AEG-!cleiunken, the licensee for GE IWF. technology, and by Siemens, the
. licensee for Westin!; house FWR technology, on bundle heat ' transfer programs dealing with both the blowdown heat transfer and emergency core cooling hect t rans f cr ques tions. Additionally, the Battelle Memorial Institute at Frcnkfurt was extending their blowdown work in 1969 to include blowdown f ron a vessel o 30 tect in lengtn.
l The schedule for the German visit thus included a visit to the two ! I water reactor vendors, AEG cnd Siemens, the Battelle Memorial Institute, ; and the institute of Reactor Safety, IRS. This tour of facilities was arranged by H. Steple, Chief of Nuclear Reactor Technology and Safety, Federal Finistry of ::calth, Education and Science. Dr. Siegler, Deputy to Mr. Sieple, was designated as guide and interpreter for all the tours and discussions. The following is a brief report of the discussions at each of the organizations which are now producing information which will both confita and complement existing programs. ! Discussions with AEC and Sienens l AE3 and Siemens, although competitive to a degree, are tied together in corporate structure through an organization which when translated appears as Oraf t Worker; Union and becorees the fabrication and construction organization for both the AEG-SUR's and Siemens-?UR's. As such, there is little difficulty in aspects of proprietary information between the two vendors. The discussions on water reactor safety tests with these two vendors
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0"F:3'? r - g7 7 Broc 71 Fa::e 13 l were held at AEG both for its convenience in location to Frankfurt as well as the experimental facilities in reactor safety research there are more i extensive than at Siemens, I I At tendees at this meeting were the following: Dr.. Siegler, Federal Ministry of Education and Science Dr. Ulrych, Siemens Dr. Shad, AEG Dr. Lochman , AEG Dr. Eicken, AEG l The :ooperative program between the Federal Minictry and the vendor industry in blowc:wn and (car;cncy cooling heat transfer consists of the following contributions f ren the Federal Minis try :
- 1. 7 million AEG Blowdown heat transfer 0.5 million Eier.cns - Energency core cooling heat transfer
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0.e elllion TU Xh'achen LEA Analytical code de"eloprent For the experimental programs each of the vendors are supplying an additional 25E or 30%. The AEG program is primarily directed at blowdown heat transfer under both BWR and ?WR fluid conditions. The Sierens program is directed at ECC heat transfer for both flooding and splay conditions. The following will first discuss the AEG blowdown hcat transfer and will be followed by discussicn of the Siccens ECC work. The AEG studies are in three stages of increasing sophistication in characterizing the blowdown heat transfer probica. The first stage, l recently conpleted but not yet reported, is a heat transfer study using sin;1c tubes directly heated and properly power profiled to establish a hsse of data to deternine if c pirical heat transfer for reactor safety analysis cannot indeed be discerned from tubes rather than the more difficult apprcath using rod bundles. The second stage will be a four rod bundle under similar heat sourec, plenum geometry and fluid conditions tc determine the dif ference between tube heat transfer and few-rod heat transfer. The third stage willir. corporate a 36 or 49 rod bundle usinc filament heaters. 1
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1 The progran purpose is to provide information in each of the three i areas of pre-CHF, eveat of C"F, and pos t-C2F heat transfer. The principal focus for all of the /IG heat transfer work is not for pure correlation f ie.clopment. They are very concerned that one should not continue carryir; l l cut uniquely purpcsed heat transfer progrcns forever. As long as bundle j
. eat t ra:afer is ex; ected to be the priScipal heat source configuration l In reactor stear supply systems for the next 20 years then experimental programs should te comprehensively planned to incorporate each and every parameter in geoentry and fluid conditions which will influence and provide an analytical basis for long tcrn design and safety assessnent ,
requirements. In this regard they are earnest ir exploring methods of experimental heat transf er wnich vill ultir.ately provide an approach to determining the basic physical processes that influence system transient heat trancier in orcer to develop describing analysis tools rather then l torrelaticas. { I l :.c teet pl.11vr oihy is to usoid -hc complexity cf duplicating er representing the "real' reactor system plushing configuration. They set up the proper entrance and exit resistance and plenua reonetry cor.fi;ura!.io s J so that the '. eat source 'ithin the vessel is "volune scaled ' to a reactor ] i core and vessel. Ther., depending upon the type of system pipe rupture to j te characterized, they rupture tvo discs in controlled sequence to cause
- the desired flui-d conditions of pressure, temperature, and mass fact te occur within th: core region. The first test phase using a single i
tube has no spacer sinulation, hevever, the second phase four rod i confi;uration will have spacers. The rod bundle heaters are under considerable study to determine ho - best to represent the Cp and diffusi-ity of a nuclear pin under transient ecnditions. They have developed schemes for separating the startup syste:... which generates the proper initial fluid conditions in the test section, from the test section within milliseconds using auxiliary rupture disks to actuate special val es. , l l i i i l _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ._ ____________ _ ________________ _ _ _ ___ __J
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3 re c 71 Iare 15 The test sectien geometry is 3 meters in length with thermocouple ter.perature neasurenents at 11 elcvations. Differential pressure acrcss ; the bundle, mass flow with orifice and turbine flow meters, fluid temperature, and absolute pressure cens titute their experimental measurements. Direct hentin is applied to the tube vall from a constant current source controlled to corro:t for thermal coef ficient of resistivity.
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Ti.e AE3 tcst fscility is arrtnged in a highly ficxible manner for - undertakin; a variety of reactor systen perf'ormance tests. A 5 M?.* S CF. pouer supply can accommodate power for 60 ? at 83 Kamp or 240 V at 21 Ec2p with 10 rsee on-of f control rate. Any intermediate voltage-current relationship and lorger power le zt1 change rates can be easily accommodated. 7he po:ct ir available at many phces in the labo ra to r/' fron an aluminur l bus bar w?.th a cross section of s 6 in. x 8 in. Concepts of " minimum-aren' ! betr2:n s:pp3; e.d return bu:.scs have been incorporated to minimize magnetic field interference uith low level instrumentation The Sier.cns program on ECC bcat transfer has begun using a sincle sheath heater within a '. tube arrantenent . The test section is 3 raters in Icagth. The other leg :f the U-tube is intended to represent the do . . corer a,uulus and is maintained at a controlled pressure with respect t o the top of the heater / opt plenum). The E.7C is injected part way doun the passive 3 eg of the U- tube. They have carried out a detailed l set of parametric studies which c: pare very f avorably with the TLECHT heat trans fer data uhen the TIECH! conditions were reproduced. The si;nificant and neu aspect of their current results is an oscillatory behatior which exists for an extended period of tine during the early portion of f:codi : Thic . cork ic primarily reported to the Tederal Ministry in renthly reports. One report uns, ho"ever, provided to me and is still in transit f ror Scrnany. This progra, is now being expanded to a 385 pin bundle 3 meters in length usin; filanent heated fuel rod simulators with step power distribution no: unlike ::_.cn? heaters . The program in intended to extend the base of the FLEC37 Nork '!ider ranges in back pressure including atmospheric prcssure co: ditions will be incorporated. The U-tube arrangement will be
OIIIOI/.! UTS C;; T ATTACl:T!T Broc-92-71 page 16 1
- aintained wj!h pressure dif ferential control between top of the passive The bundle test will incorporate three
- e; and the top of the heated leg.
dif ferent internal arrangements in erder to establish the influence and
- nnser,uences of cross flor, if any. Tests vill be run with a normal
- c. pen lattice configuration. Similar tests will be run with several For thin nerl crane septa separating thics annular ring: ef feal red 9 the extreme situstion the septa vill be maintained closed to fully prevent the potential for fluid to migrate outward in 'a radially profiled power distribution between annuli. Othet cest series will provide for various ar.ounts of cross f'Jow by perforating the septa with various increasing fracticns of epcn area. At this point they are currently evaluating their heaters in pre;ararion for assenkling the test bundle.
Battelle I.erorial Institute First, a discussion The visit to BMI-Frankfurt included three activities. and tour relating to the blowdown and containment response programs at Batt elle-Frcnkfurt; second, a discussion with the German analysis develepnent group from the Technical University MGnchen, LEA (LRA related to group on reactor analysis); and third, en informal discussion with a group comprised of renbers from the various German reactor safety activities including the Federal Ministry, the Reactor Safety Commission, Battelle-Frankfurt, f
,LRA, AEG, and Siemens.
El oudoun and Cent ainment Resoonse Prograns. Attendees included: j Dr. Rudiger, Battelle Institute Dr. Sinon, AEG Dr. Ziegler, Federal tunistry The bIcwdoun progrr.n at Battelle over the last four years has progressed , from the early glass vessel tests used for cinematography and conducted at two atmospheres and a volume of approximately I cubic foot, to current tests at near PWR aed LWR fluid conditions in a vessel near 100 cubic feet. The test progran has principally addressed blowdown phenomena questions in LWR's; eg, level swell, phase separation, and long L/D influence on leak e
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ATTACit'J.':T Broc-92-71 Fage 17 1 flo.' rat e . In cddition, information applicable to WR technology has ] been acquired; eg, flow choking and thrust forces. Tests since the 1969 CREST meeting have been carried out in a. 4 foot long vessel with exit nozzles up to 6 inches in diameter and 20 feet long. Information on histories of water level, pressure, differential pressure alcng the nozzle, differential pressure across the core cockup, and thrust forces are available in reports which were acquired. This test vessel eventually cracked from blowdown thermal stress. A new vessel 33 feet in length and 2 feet in diameter was fabricated. The new vessel has nozzles at many elevations and is arranged for dynanic weight neasurcnents. Internals which can simulate the core shroud of a 3'.G or a core barrel annulus of a WR are incorporated. To acconplish j l the ectly Nater level surge" and the later " level swell" they have provided j i for forced temperature stratification of up to.16 C within the vessel. Evtendv.* t"nmarison of fluid temperature and pressure data have shoen j that during the early part of the surge and swell cycle, the fluid conditions are in a netastable or nonecuilibrium state. The extent and duration of superheat is greater than was observed in the U.S. semiscale program. The data s!.cw that the extent of superheating required to drive the energy f rom the liquid f raction to the vapor fraction is about 8 C. Tests using a flared nozzle on their blowdown pipe have apparently held the choking plcne stable in one position and thus improved their ability ll to explore the downstream behavfer of the effluent. The downstrean process is of concern to them in creating the potential for impingement loads en other structures and ECC plumbing. Arrangements were discussed to permit us to acquire the high pressure glass vessel movies of the 3WR level swell behavior. Their future progran j will include test:ing using the 33 foot long vessel with a circulating 1000 I nockup but will not include core heat. j The containment. response progran is to begin in the spring of 1972 and is intended to provide extensive information for the question of transient IJ loads across compartment structures internal to the containment for both !
ATTACHMENT o E roc-92-71 C027c7f,7, p~_
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The effluent from the 33 foot vessel vill proceed into a simulated vessel within the model containment and thence to the various containment compartments for a very vide variety of breek location representations including both . breaks at the junction of major pipes with the vessel and vessel ruptures. For the 3'.:R configuration the dry well will be simulated, however, the vet vell vill not be simulated. Pressure suppression vill be the subject of specia3 studies at the Gisthoch Institute where they will mockup a full scale representation of the Otto Hahn (Nuc1 car Ship) containment l con fi gurat ion. Analysi r, Develertent Progran: Attendees included: Dr. Karvat, LRA Technical University, E0hchen Mr. Brosche, LRA Technical L51versity, M0hchen Mr. Wolfort, LRA Technical University, MGachen Mr. Hofer, LRA Technical University, MUnchen Dr. Ziegler, Federal Ministry Analysis development for independent safety assessment of German reactors is concentrated at the Technical University of MUnchen, LRA under ! Dr. Rarvat. Although the analysis development program is rapidly expanding to include core heatup analysis, containment response codes, and structural response codes.their principal effort has been in the development, application, j and extension of the ERUCE codes ( sitilar to our LELAP code). Originally the BRUCH codes were intended to provide calculational techniques for I 1 determining hydraulic forcer during blowdern, including reaction forces ; on the primary systen support and as well forces on containment rtructures. l Partially because of concern they saw evidence in the U.S. about the core thernsi response and the emergency core cooling question, the BRUCH codes were extended ) to incorporate considerably more detail to provide for calculation of fluid states ! and flows within the system which would permit a determination of the core thermal rcaponse. The analysis development approach at LRA is to develop the various versions of the 3RUCH codes and then let the industry and other safety assessment organizations apply the codes. Feedb ack f rom the users is used to determine the capability, limitations, and confidence in the codes.
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codes in that BRUCH codes carry out a situitaneous solution to the energy and monentus equations _ in which pressure is calculated as an ludependent variable. The IILA? and FLASH type codes eniculate internal energy as i an independent variabic.- The Germans clain that one benefit in their approach co: pared to 12 LAP is the time steps can be larger without producing instabilities, at least in the longer tern part of the blowdown. Incorporated in the code is an autonatic tine step change which is actuated by the rate of change of certain key variables. They incorporate ecuations of state ! l for fluid conditions rather than steam tables and do not consider this a di f ficult preb1cn since one only needs to use a very restricted area of the steam tables. The najor disadvantage of the 3RUCH code is its fixed number of 15 nodes er cont rol volures and the fixed geometry it must characterize. This i situatica precludes characterizing the variety of geometries representing ! curren: desinns of different vendors. They have attempted to predict Scriscale lest S;8 using DRrCH, but because of the dif ficulty in characterizing; the pecnetry, they have only been abic to proceed through the first 200 til11 seconds of the test. ! In terms of additional refinements to the BRUCH code, they are planning te incorporate a flexible representation of the systers by
'previding f or variable degrees of noding much as is done in RELAP and U.S.
codes today. In addition, they p ,, to incorporate a combination of implicit and explicit techniques in order to further shorten the total calculational time. They intend to incorporate provisions for calculating the core the rms: response behavior in 3RUCH in order to prevcnt, as now exists, decoupling of the core thermal and blowdown processes. I cautioned then to at least provide a printout of the important blo.7down variables and the core flow such that they can more readily trace errors and dif ficulties than has been possible in the approach taken by a U.S. vendors. Their core thermal model ir. BRUCH is an average channel and averace pin representation. The code provides four axial regions and permits up to ten radial nedes withir the co're. They will use the core thermal codel in ERUCH to develop a reflood code and are particularly interested
ATThCHZp.NT - b/8ICU L UE i . LY Broc-92-71
- Page 20 in the radial distribution of coolant during the reflood processes. They have a code development program under consideration for predicting parallel channel behavior. They also have a scheme within the core thermal calculation part of the code to permit an empirically ' derived. delay before CHF is permitted, but cuce it is perrdtted no rcwettin; can occur. They are concerne'd about the influence of the stear generator and pump behavior but as yet have not done a great dea) in characterizing these components.
Their coatsinment response Enalysis is centered around a code called ZOCO. The 2000-1 code incorporates 14 pressure nodes within the containment and provides for a variety of configurations for the L/D connections between conpa rtr.ent s within the containment. They use thermodynamic equilibrium but do test for choking conditions at constrictions between compartments. In this particular code the principal aim is to determine the loads with heat transfer neglected. 20Co-5 is under development and will provide for a variable ntrber of nodes and is tc work che very early damage producing peried of the containment response. They are picaning for a ZOCO-4 and -5 l version which is naintaincd as cn internal code at the moment but is to be a very sophisticated code and to be used for predictions in conjunction l with the Battelle containment test program at Frankf urt. The code will l also accommodate nonequilibrium ccnditions between the water and the steam and as well permit the gas f raction to become superheated. This code has teen tested with the Humboldt Bay data and compares very favorably. The code has been applied to varicus c::perimente carried out on compressible gas flow through ducts, again with good agreement. l l 2000-5 is to become a standard containment response code to be used l in Ge rr.ar.y. The code description is currently being written and will be fi rs t released for use by Gernan vendors. Ultimately it is to be released to the Ispra Library. A separate discussion indicated potential for acquiring an early unofficial version of the code for evaluation. I They are clso developing a code called DRABSYS-T for calculating the pressure buildup in a BWR type pressure suppression system. The code was under development before CONTEMPT-PS and is still under development. Currently it tar.cs the form of a three volume node characterization; one for the wet well, one for the vent, and one for the dry well vessel. l e
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ATTACHMENT Etoc-92-H OIIICIAL U; c:;7 Page 21 There is a new version of DRABSYS started in 1968 which will incorporate parallel and separately calculated vent pipe behavior and in addition will incorporate nodes for certain compartmentation configurations with doors which at certain pressures can open and change the local pressure conditions. Infernal Discussion on ICC Analysis Approach with Various Represente-tive f ro- the Cerran '.?ater Reactor Safety Community. The following were in attendance: NAME _0S';!f IZATION RESPONSIBILITY WITHIN ORGANIZATION Ziegler Minist ry of Science Coordinator, Reactor Safety Research and Education Work l 1 Mayinger Tcchnical University Section leader and Chairman of the Reactor i of Hannover Safety Commission l 1 HFrtner TU MUnchen, LRA System Analysis, Design Criteria Farber IR$ Group Leader, lieat Transfer l Rohde IPS Group Leader, Fluid Flow Uolfert TU M0nchen, LRA Unit Leader, Blowdown Code Development RUdige r Bat telle-Frank f urt Project Leader, Experimental Study of , Blowdown and Containment Response l Schad AEC Unit Leader, Experimental Research, Heat l l Transfer Under Blowdown l Frenkel /.EG Group Leader Emergency Core Cooling Analysis l Kech AEG Unit Leader, Loss-of-Coolant Safety Analysis j l Winkler Sienens Section leader, Reactor Thermodynamics and Safety Analysis Blank Siemens Reactor Safety Experiments During the first part of this session the emergency core cooling work j in the U.S. was briefly described in which the recent completion of the I FLECHT work was sur.marized. In addition the Semiscale Tests, for which they had reviewed the monthly reports in-depth, vere ~ described and discussed. The forthcoming Semiscale Tests, the LOFT ?rogram and the Analysis Development l l Program were also described. In exchange they provided a candid revelation to their apprcach to accommodate the assessment of water reactor safety system adequacy. The candor was refreshing and remarkable particularly in view of varied responsibilities of the attendees representing every ) facet of industry, research and regulation.
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w..s,. E ro c-9 2- 71 Page 22 The following are the significant points of interest during this exchange. (1) Structural loads during bicwdown are still of significant concern. Currently they are tcsting the vendor codes with the BAM code developed by V. T Berta at the NRTS. (2) They too are concerned about properly characterizing the plumbing ! configuration of re actor primary systens and its components, but as yet have not developed anything beyond the approach used earlier in FLASH l and RELA? in accommodating elevation and gravity, annulus geometry, pump cavitation, fic,w split at the annulus, etc. (3) They too are not including a realistic representation of the secondary side of Ph'R's but feel it is important. ( 1 (4) The system flow, as cciculated by BRUCH, generates pulsations in the primary loop which thpy have not been able to unravel. I pointed out that r,ost of the U.S. codes were calculating relatively smooth system flows in the piping and that oscillatory behavior requires the coupling of both inertance terms and compliance terms which simply do not exit during the two-phase portion of blowdown unless the code was capable of calculating " organ pipe whistles." The frequencies, however, were acknowledged to be lower and therefore must be numerical instabilities. 4 (5) BRUCH calculates no differences in blowdown behavior with or without pressurizer. It should be noted, however, that their surge line is 20 meters l long and orificed. They accommodated in the design of the system the ability to handle the four or five atmospheres of overpressure for loss of stean Joad and therefore are net cencerned about steam generator response tire. 1 (6) They e.re willing to accommodate a calculated level risk in the risk / benefit considerations relating to their safety research philosophy. They have a strong desire for " basic understanding' before they do integral 1 e l l l ______-_-__-_--l__-____E
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+ systen tcsts. Thcy are, however, keenly interested in exchanging their information on basic studies for our information on integral effects.
(7) As centioned earlier, they are still referencing and strongly interested in a comparative probability approach toward determining uhere they assign reactor research funds. For exampic, if they establish a 1 in 108 for core meltthrough for loss of emergency core cooling, they consider that with a reasonably effective core catcher, containment melt-through would be 1 to 10 12 (8) They are constantly aware that metr'opolitan siting is here and nne and c1though they have explicit f aith in the inspection and compliance or;cnizations, they are dominantly concerned with the vessel f ailu're problen. In forthcoming licensing reviews the vendors have been asked to addreas the question again of vessel reliability and consequence of failure. The vendors are peeved and are required to look at a second vessel enclosing the primary vessel. (9) They intend to exploit the incipient flaw detection schemes to on-li .c capability anc have a basic research prograr.. Scing en in l this area at Battelle Frankfurt. (10) The ECC delivery problem for cold leg breaks in PWR's was not considered nor understood until the A.:C monthly began to reveal the results of the semiscale tests. They had carried out their own evaluation
, of the dif ficulty and concurred with the interpretation currently ackr.owledged by the Interim Criteria of the U.S.
Discussions with IFS on Reactor Safety Assessment Attendees include: Mr. Sieple, Federal Ministry Dr. Ziegler, Federal Ministry Mr. Schwarzer, IRS Mr. Farber, IRS Mr. Rhode, IRS Mr. Jahns , IRS for the discussions at IRS, as well as for all other visits, I disqualified myself as be,ing a spokesman for U.S. licensing practices or philosophy. The Gernans chose to volunteer their safety assessment approach and practices. The following are observations from those discussions. Figure 1 is a schematic
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-~..4 ba page 25 of the arrangement of the various bodies involved in the safety . assessment process. The applicant / vendor relation is much like that existing in the U.S. The applicant / vendor co==unicates primarily with the state licensing authority and the TUV organization. The TUV organization is an inspection and compliance group which is the coordinator for the total licensing procedure. The TUV has in addition all non-nuclear licensing responsibilities ; eg, elevators, public transportation, chair lif ts , etc.
They develop snd maintain many of their own codes, criteria, standards, and quality assurance procedures. There are eleven states each with its own licensing and inspection commission. The state agency has no technical capability so mus t rely upon both the Tederal Ministry and IRS 'through TUV to support their own approval procedures in technical matters. The Federal ISnistry maintains a strong role in impicmenting procedures, criteria, aad directions for state authority actions. The R$ is an independent ' review body comprised of university professers e'd other qualiff ed experts outside the nuclear industry per se. Although the R5K serves a function not unlike the U.S. ACRS it has additional responsibilities.= in establishing appropriateness and priority of the distribution of funds for analysis development and supporting experimental l work. The IR5 is a captive organization to the Federal Ministry for supplyin5 any needed analysis to the TUV organization, to the rederal "inistry and to the RSK. In addition the 17.5 initiates and monitors both the analysis development and the experimental work in support of the licensing process. The analysis development for water reactor safety is carried out primarily at TU MS~nchen, LEA, while the experimental work is carried out by the tvo vater reactor vcadors and Battelle-Frankfurt. The Federal 1Hnistry must ultimately sign of f on ea:L license prior to state approval. Points of interest during these discussions are as follows: (1) Reactor safety issues are given steering committee treatment.
.4 typical steering committee may be comprised of a member from RSK, IRS, the Federal 1Hnistry, LRA, and the particular reactor vendor involved.
1 l l ,
I ATTAC!D:D;T OTFTCT : - B ro r-9 2- 71 l
. Page 26 The steering committee wrestles with the issue until there is common i
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understanding of its structure and approach to its resolution. Committees ] have been established and acet several times per month on (a) blowdown, . j which includcs five subprojects on such as the Marviken containment tests l J in Sueden, ice condensers and pressure suppression fer the ship reactor, ) Otto Eahn, (b) core meltdcen, (c) eraergency core cooling design and. l performance, (d) containment response and structural adequacies. (2) Reactor safety questions can be . raised by anyone in the chain l 1 of the groups shown in Figure 1, however, there appears to be no require- l nents in the system for any particular group to search out and establish reactor safety questiens. The IRS group had so:ae difficulty in-answering this question. i (3) Applicants are billed % 1% of the cost of the plant for the I total TU' licensing process. i 1 (4) Although there is no definite organization for reactor safety criteria and standards development, a reactor safety criteria document 1 I has been generated and issued by a cooperative effort of several institutions
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and a copy was acquired. l ! (5) With the exception of the TUV licensing / compliance requirements j there is no Technical Specification document ec,uivalent to that used in the U.S. scheme. l l (6) Although IRS does not have a criteria based reference system I for safety assesstent, they do have a strong pressure by the Federal Flnis try to pursue consequence understanding which they feel ultimately provokes adequate design requirements for accident mitigating safeguards and their describing analysis. (7) As centioned earlier, the Germans are keenly concerned with the issue which surround metropolitan siting. In addition to the considerably greater than zero probability for primary vessel failure and the subsequent core meltdown question, they are concerned for for accident initiation by such as exploding barges on adjacent rivers, 1 l
ATTAGMC T hTIC , 7. .... .. B ro c 71 Fage ?7 containment penetration by aircraft, and sabotage. Because of the international circumstances in that part of the world, sabotage takes on a unique concern. For example, they are currently attempting to develop a criteria requiring at 1 cast three disgruntled empicyees to take concerted action in order to initiate a loss-of-coolant accident. This requires that emergency diesels, for example, be independent in their fuel supply and located remotely f ron each cther around the plant site. Further, schemes for , continuous interrogation of all operating personnel within the_ plant must be made available to the operator at the console and in addition to at lesst one other supervisor area. The German's consider the sabotagc problem to be at least an order of magnitude more probable than a loss of coolant from system failure or acts of God. (B) The IR5 does not deal with Technical Specification Documents , as is part of the U.S. Regulatory process. The TUV with certain " operating limits" from the Federal Ministry administers and monitors the Technical Specification aspect. (9) 'Ibey have no cladding temperature limit in their criteria as they j calculate maximum temperature well below 2000 F. In probing, however, and being supplied data from a typical multilcop FWR calculation, the l core flow history, although similar in time history to U.S. results, is of greater anplitude generally. The resulting core heat transfer history, also supplied to me, was roughly five times greater for a major f raction of the blowdown compared to similar U.S. calculation. Their cladding temperature position is thus not surprising. (10) Until Germeny can assure calculational adequacy for accident description and attendant safety system performance evaluation, they will impose greater e: phasis on: redundancy and single f ailure considerations, inspectaoility, testability, maintainability, and coolant leak rate.
I At t a cht.?n t broc-92-71 Pa;p 23 Orr;c, 6-~ C LY i CO::CLUSIC':S A.:D 2:00M':d:DATIONS 1 l In sunnary the feregoing conclusions and reconnendations nay be ; i catec,erized as fellows: l Genov: Atcr.s Conf erence i ihe conf arcnce nust be termed a success in terns of indicating attention to and adninictration of programs 2videncing global and natio.a. n;eds regarding cnerg/, enrichnznt, environmental protecticn, fast r2 actor develepnent, radiation effects on man, reactor operatint
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exp e rien ce, and reactor safety. ' In contrast with previcus Geneva l conferences this conference did not constitute a' forum for detailed ] dialogus in special:ted aspects of nuclear energy. In fact the general , level of organizational representation precluded such dialogue. i
}atcrFxactersafetyResaarch 9 I
The U.S. vas ac.<nouledged in all discussicas as leading in safety re s u a .x :: ... : Je vals;..u.; cf critcrin end st:ndarde.^ .. a result, the everall U.3. posturc cn safety has spawned rigorous design and licensing crit:ri r th roughout Europe. Many Europcan agencies are now pursuing I their or prograns in areas relating to metropolitan sitin; at a pace which ! cxpact will reach or overtahe the U.S. state-of-understanding;. The U.S. should be able to evidence a broader cognizance in the felloring areas or procrans: I 1
- Marviken Containment Test in Sweden
- Batta11e BWR Bleudoun Tests in Frankfurt
- Pat ta11e Ccat ainnent-Cenpartm'.nt Loads Tcsts in Trankfurt
- latt slie Flau Imissions Study in Frankfurt
- A20 31cedern Heat Transfer Tests in Frankfurt 1
- L?.A Less-of-Ceolant Accident Analysis Development in Muchen ]
- Siemens ECC Heat Transfer Tests
- Otto Hahn Pressure Suppression Tests
- UKAfA Depressurization Studies at Foulness h L'at:r p.; actor Saf ety Philosophy L
Eurcia in cen2ral is facing t. metropolitan sitine no.:. They are adanant ! in e.ssuring thens:1ves that the containn:nt is preserv:d as the ultinate l l J
_ _ - _ - __= _ - _ _
.' o ' DTTICII.L IC2 CULT Attachnent l Broc-92-71 I
?qe 29 ;
barrier to radioactivity release from an accident. To secure this assurance considerabic attention is being given to the following:
- Cause, consequence, and prevention of reactor primary vessel f ailure j
- Course, control, and confinenant of molten core if ECC or vessel j should fail i
- Ri;;id centrols on single failure design, quality assurance, reactor )
nal-operation, diagnostics, testability, inspectability, and maintainability i 1
- ipplying accident probabilistic approach to safety systen design, l safety assecsment cnd allocation of research funds ,
Design for sabetage protection There are recognized philosophical differences between the U.S. and Europe in these matters; however, it would be imprudent for the U.S. I not to nalntain cognizance or tne fundamental requirements and underlying rationale which supports their dpproach in each of these areas in order to opticisc icas range ple.ns in this country which will address continually changing requirements within the U.S. j l i _ _ _ _ . . _ . _ _ _ . . _ _ _ _ _ _ . _ _ _ . _ _ _ . _ _ . _ _ _ _ _ .}}