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i JAN 2 21932 Mr. J. LaGrone, Manager San Francisco Operations Office U. S. Department of Energy 1333 Broadway - Wells Fargo Building Oakland, C44ffoonia 94612
Dear Mr. LaGrone:
FY 1982 NUCLEAR REGULATORY RESEARCH ORDER NO. 60-82-118 FOR LAWRENCE LIVERMORE
,ATIONAL LABORATORY Please authorize Lawrence Livennore National Laboratory to execute the prograil described in the enclosed NRC Order.
If this meets with your approval, it is requested that acceptance be indicated on the enclosed fonn and the original be returned to the NRC Controller and a-copy to this office.
Sincerely, 1
- 0. E. Bassett, Director Division of Accident Evaluation Office of Nuclear Regulatory Research
Enclosures:
- 1. NRC Order DISTR l 2. Program Brief Subj Circ cc w/encleeures: Chron R. W. Carber, DOE /NSC Riggs:rdg
- D. D. Mayhew, DOE /ASEV:DFS A. Pug 11se, CON B. Bownan, LLNL J. Mate, CON DAE RES:D Wylbur 0702050512 070120 Fin File p 702 PDR
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i NRC some 173 u s NucLt An macutAtom, CouwissioN omDa R Nuust a 12 761 60-82-118 STANDARD ORDER FOR DOE WORK gy pp ISSUED TO (DOE OH.ce) ISSUE D S Y - (NRC Of fice) ACCOUNTING CITATION APPROPRI ATION SYMBOL San Francisco Operations Office Division of Accident Evaluation m 0200.602 34 R NUMBE R PERFORMING ORG ANIZATION AND LOCATION 601911 -
FIN NUMBE a Lawrence Livermore National Laboratory A0116-2 WORK PERIOD THl$ ORDER FIN TITLE FlxtD O ESTIMATED $ _
BWR Containment Pool Dynamics FROM. TO 10/1/R1 4/10/R7 OBLIG ATION AVAILABILITY PROVIDED BY:
A THiS ORDE R S 15,000 B TOTAL Of ORO(R$ PL ACE D PRIOR TO THl$ DATE WITH THE PERF ORMING ORG ANIZ ATION Ta% *N'M S^"C M'o'",$"'EATION SvMeOL AND THE FIRST FOuR DIGITS Os THE 4,160,000 C TOTAL ORDERS TO DATE (TOTAL A & BI $ 4.175,000 0 AMOVNT INCLUDED IN "C" APPLIC A8LE TO THE " FIN NUMBE R" CITED IN THis ORDE R. $ 15,000 FINANCI AL FLExlBILITY.
, O FUNDS WILL NOT BE REPROGRAMMED BETWEEN FINS LINE D CONSTITUTES A LIMITATION ON OBLIGATIONS
( AUTHORIZED Q FUNDS MAY BE REPROGR AMMED NOT 70 E XCEED 210% OF FIN LEVEL UP TO SSOK. LINE C CONSTITUTES A LIMITATION ON 08LIGAT60NS AUTHORIZED STANDARD TERMS AND CONDITIONS PROVIDED DOE ARE CONSIDERED PART OF THl5 ORDER UNLESS OTHERWISE NOTED.
ATTACH ME NTS THE FOLLOWING ATTACHMENTS ARE HEREBY SECURITY.
MADE A PART OF THIS ORDER; Q WORK ON THIS ORDER 15 NOT CLAS$1FIED.
O STATEMENT OF WOR K O WORK ON THl$ ORDER INVOLVES CLAS$1 PIED C ADDITION AL TERMS AND CONDITIONS INFORMATION NRC FORM 18715 ATTACHED.
O OTHER REMARKS crA ISSUING AU.INORITY ~
ACCEPTING ORGANIZATION SIG N{f u p' \ D Q T M 88GN Af unt O. . Bassett, Director fif tt Division of Accident Evaluation TITLE Office of Nuclear Regulatory Research NRCFORM 173(2 78)
4 .
FY 82 REVISED PROGRAM BRIEF PROGRAM: AE TITLE: BWR CONTAINMENT P0OL DYNAMICS FIN NO: A0ll6 CONTRACTOR: LLL SITE: LIVERMORE STATE: CALIFORNIA NRC TECHNICAL MONITOR: T. LEE PRINCIPAL INVESTIGATOR: 8. B0WMAN OBJECTIVE: LIAIS0N FOR FULL SCALE STEAM CONDENSATION TESTS BUDGET ACTIVITY:
FY 1982 SCOPE: 15K THE FOLLOWING TASKS. ARE TO BE PERFORMED TO CONCLUDE THIS PROGRAM:
- l. SHIP INSTRUMENTS / EQUIPMENT, SALVAGED FROM THE 1/5 - SCALE MARK I CONTAINMENT TEST, TO MIT AND INEL PER NRC INSTRUCTION.
- 2. COMPLETE DATA REDUCTION AND REPORTS FOR THE FIRST SERIES OF MODAL TESTS CONDUCTED ON JAERI-CRT FACILITY.
( 3. TRANSMIT TO NRC TAPES, LISTINGS AND MANUALS OF THE DATA REDUCING ROUTIN,ES DEVELOPED FOR GKSS AND TPC PROJECTS.
, 4. TRANSMIT'TO NRC ANY UNDELIVERED DOCUMENTS, TAPES / FILMS AND OTHER INFORMATION RECEIVED FROM GKSS, JAERI AND TPC.
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NRC Translation 1044D-LANGUAGE OF FOREIGN DOCUMENT: Japanese TRANSLATED TITLE OF DOCUMENT: Test bata Verifying the Relia-bility of the Containment Pressure Control System UNTRANSLATED TITLES:
AUTHOR (S): Isao Takeshita, Nobuo Yamamoto,
, Michio Curoki, Yutaka Kukida, Ken Namatame, Masayoshi Shiba TRANSLATED NAME AND ADDRESS Japan Atomic Energy Research OF CORPORATE AUTHOR: Institute l DATE OF ORIGINAL FOREIGN DOCUMENT: January 1982 FOREIGN DOCUMENT ID NUMBER (S): JAERI memo 9928 1-NUMBER OF PAGES IN TRANSLATION:
202 pages I
DATE TRANSLATED FOR NRC: April 1982 L
TRANSLATED BY: Techtran 3 NAME:
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ADDRESS:
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JAERI -memo 9928 Test Data Verifying the Reliability of the Containment Pressure Suppression System Report 23 (Test 3191)
Tokai Institute, Faculty of Safety Engineering, First Department of Safety Engineering Isao Takeshita, Nobuo Yamamoto, Michio Kuroki*, Yutaka Kukida, Ken Namatame, Masayoshi Shiba (Received January, 1982)
Tests verifying the reliability of the containment pressure suppression system were conducted on thermohydraulic phenomena developing in the pressure suppression system of the Mark II containment vessel for BWR during LOCA, and the objective was the acquisition of data confirming the reliability of the containment vessel. Sealing of the test facility is 1/18, and the wetwell section was a 20 sector of the wetwell of the actual reactor.
This report is the report on the data of Test 3191 conducted on December 4, 1981. The test was a test of initial discharge following reinforcement of the shell, pedestal and pool bottom to raise the rigidity of the pressure boundary of the pressure sup-pression pool. The air within the drywell was purged (prepurged) to wetwell gas space using steam prior to the discharge. These test conditions were identical with those in test 3104 which was con- ducted prior to reinforcement. The results of comparison of the data on strain and acceleration in the pressure boundary of the wetwell reveal rigidification of the pressure boundary by rein-forcement.
This report is designated research for the Japan Atomic Energy Research Institute by the Science and Technology Agency based on the special expenditure directive to pro-mote the development of new power sources: This summarizes the test data of Test 3191 of the 1980 " Testing to verify the reliability of pressure control system of containment vessels".
Ilitachi Engineering K.K.
This document has been prepared for exclusive use of JAERI personnel. It may not be quoted from, referred to or repro-duced in any way.
2
- 1. Introduction Tests verifying the reliability of the containment pressure suppression system have been conducted simulating thermohydraulic phenomena during a loss of coolant accident in the Mark II contain-ment used in the newest type of boiling water reactor in Japan, and the objective was the acquisition of data confirming the reliabil-ity of the containment vessel.
The planning of these tests was entrusted to JAERI by the Min-istry of Science and Technology as part of the special expenditure directive to promote the development of new power sources. It was conducted for five years beginning in 1977. The Faculty of Safety Engineering, First Department of Safety Engineering inside the Japan Atomic Energy Research Institute was responsible for develop-ment of the test project and for analysis of the test results, while the Safety Testing Technical Department was responsible for operation and administration of the test facility. The test facil-ity was completed in March, 1979 on the premises of the Tokai institute of Japan Atomic Energy Research Institute, and tests were conducted subsequently at the rate of approximately one a month, and a total of 23 discharge tests have been completed to date.
The effect of FSI (fluid structure interaction), due to vibra-tion of the pedestal and shell of comparatively low rigidity within the pressur,e boundary of the wetwell, was confirmed to be included in the test data of the tests conducted thus far (1). Tests such as hammering tests on the shell-pedestal, etc., steady oscillation tests (2) or tests based on single vent pipe have been conducted to quantify the FSI effect. In addition, attempt has been made to mitigate the FSI effect by increasing the rigidity of the facility itself at the same time these tests and test data are analyzed (3).
Reinforcement of the facility was completed before this test (Test 3191). Reinforcement was accomplished by lining the shell and pedestal among the pressure boundaries of the wetwell with 0.3 m thick concrete on the inner side of the pool to a height of 7.5 m to cover the water level of the pool during testing. At the pool bottom, epoxy resin (Japan Jikko K.K. , Anchor Epo) was injected between the steel plates on the bottom and the concrete foundation,
! thereby raising rigidity by plugging the gaps.
This test report is a report on the data of Test 3191 that is
( the first test conducted after the above described reinforcement of the wetwell (December 4, 1981). Test 3191 is a test of water
, discharge from a discharge nozzle of 74 mm bore, and the air within i
the drywell prior to discharge was purged to wetwell gas space.
The test conditions were virtually identical with those of Test 3104 which was conducted prior to reinforcement.
. . 3 The acceleration and strain in this test at the shell, pedestal and pool bottom, which were the reinforced parts of the wetwell, were below the levels in Test 3104, and rigidification of the pressure boundaries due to reinforcement was thought to have been achieved. Quantitative evaluation of the reduction in the effect due to FSI included in the pressure data within the pool must await the results of frequency analysis.
- 2. Test Facility and Test Instrumentation Figure 2.1 illustrates an outline of the test facility. The test facility is composed primarily of the test containment, pressure vessel and discharge piping. The test containment wetwell section is a simulation of a 20 sector of the wetwell of the 1100 MWe class Mark II containment vgssel. The pool water volume has been reduced approximately 14 m due to reinforcement of the wetwell and pedestal. The volumes of the drywell and primary system are approximately 1/18 those of the corresponding portions of the actual reactor. These various dimensions are. illustrated in table 2.1.
Table 2.2 illustrates the items measured in the test facility while tables 2.5 to 2.9 illustrate the measurement channel lists and Figures 2.2 to 2.17 illustrate the measurement points. In these tests, some of the water level gauges, strain gauges, etc.
which were installed on the shell and pedestal walls were removed due to reinforcement, some were re-installed on the walls following reinforcement. In addition, thermocouples for measuring the water temperature on the drywell follor and for measuring the temperature on the drywell wall were newly installed. Grooves were dug into the surfaces of the pads, fixed on the drywell walls in order to measure the wall temperatures, and the thermocouples were embedded in electrically conducting resin (Fujijura Kasei: Rotite D-753).
In addition, strain gauges were added at the lower bracing, and the strain gauges installed at the same bracing and at vent pipe VP-5 all measured strain by the two gauge method. Table 2.17 illustrates all modifications prior to the present test; the items in the facility which were modified as well as the measurement systems which were newly installed, moved or removed.
The data are recorded by twin system recording systems as before. Specifically, signals of conparatively slow change (temperature, water level meter output, etc.) are recorded online by a mini-computer while signals of comparatively fast change (most pressure, strain, acceleration etc.) are recorded by PCM (pulse code modulation) method, and the data are then reedited via a small computer following the test. Table 2.3 illustrates the various specification of the data recording equipment, and four tracks were used in the same PCM data recording system as in the previous test.
The number of channels per track was 39 while the data recording rate was 911.11 WD/CH/s.
. . 4 Tables 2.10 to 2.14 illustrate the actuation status states of the measurement devices in these tests; table 2.15 illustrates the calibrated test results of the pressure transducer, differential a pressure transducer, etc. while table 2.16 illustrates the existing values of measurement ranges.
Failures in the test equipment as well as those in the measurement system which occurred in the previous test were corrected before the present test. There were five failures in the pressure gauges, one failure in the accelerometer, 15 failures (primarily noise) involving the strain gauge and two failures in the water level gauge in the cur- rent tests, but there were not major failures for the measurement system as a whole.
- 3. Test Conditions and Test Results Test 3191 is a water discharge test from a discharge nozzle of 74 mm bore, and in the test air within the drywell was purged to wetwell gas space prior to blowdown; the so-called prepurging.
Specifically, chugging was induced for prolonged periods of time by rupturing a comparatively small bore
- in these tests, and the air fraction in the vent stream was reduced by prepurging. Thus, conservative conditions pertaining to the amplitude of pressure oscillation were achieved. The test conditions were virtually identical with those in Test 3104 conducted prior to reinforcement of the wetwell pressure boundary discussed in chapter 2.
Evaluation of reduction in the FSI effect due to reinforcement was possible by comparing both test results. Table 3.1 summarizes the
' test conditions under which the tests were conducted, including both tests illustrated in table 3.1.
The temperature and pressure of the pressure vessel were raised over two days, December 2 and 3. Nitrogen gas was injected into the discharge piping as before in order to ensure rupture of the rupture disk. In conjunction with this, prepurging was conducted using saturated steam (pressure 600 kPa) supplied from a steam boiler. The amount of water which accumulated within the drywell during the 36 minutes of prepurging, excluding the time of
, temporary shutdown of th boiler during the prepurge, was measured l to be approximately 1.6 m 3 by the newly installed water gauge. In t
addition, the wall temperature rose virtually linearly for approxi-mately 20 minutes after the start of prepurging, and at the comple-
[ tion of prepurging, the temperature had approached the saturated
- The ration between the rupture bore area (total surface area of critical area) during a double-ended rupture acci-dent of the recirculation system piping is estimated to be approximately 0.0101 in the Mark II plant in Japan. In contrast, the surface area ratio in these tests was approx-imately 0.0022.
.. . 5 temperature (130*C). More than 90% of the air within the drywell shifted to the wetwell gas section through this prepurging, and the temperature of the drywell gas section was raised to approximately 100*C. Furthermore, the wetwell water temperature rose approxi-mately 6 C.
During prepurging discharge of the primary system steam took place and the safety valve of the primary system was actuated for approximately one minute. Thus, the primary system cooling water holdup was reduced, but the test was not affected, and the discharge test was begun following completion of prepurging.
Table 3.2 compares the target values with the attained values of the initial conditions in different parts of the test facility during the test. Table 3.3 illustrates the state of data recording while table 3.4 illustrates changes in the typical physical quanti-ties before and after the tests. In addition, figure 3.1 illus-trates the facility operational records before and after tests while figures 3.2 and 3.4 illustrate the distributions of the initial temperatures within the pressure vessel, discharge pipes, containment vessel (values af ter prepurging and immediately before the start of discharge) and of the temperatures after completion of the tests (350 seconds after the start of discharge).
Quantitative evaluation of the test results and analysis are deferred to a subsequent report. Only general discussion of the evolution of major phenomena are presented here.
The water level within the pressure vessel and the primary system pressure at the start of discharge were somewhat lower than those in test 3104 due to actuation of the safety valve during pre-purging. The temporarily reduced pressure in the discharge piping and in the primary system pressure vessel immediately after the start of discharge was approximately 300 kPa. The water level within the pressure vessel reaches the inlet of the discharge pipe approximately 100 seconds after the start of discharge, and the quality of the discharge fluid increased subsequently.
This test was one in which the rupture bore was comparatively small and prepurging approaching 100% was implemented; consequent-ly the rise in the wetwell water level due to pool swell identical to Test 3104 was not observed.
The temperature of the gas section and walls within the drywell virtually coincided with the saturation temperature, corresponding to pressure, from immediately after the start of discharge and gradually increased, but the gas section shifted to the superheated state after approximately 170 seconds. The i diaphragm floor temperature rose rapidly from approximately 60 C at i the end of prepurging, but a value somewhat lower than the satura-L tion temperature was exhibited throughout the blowdown period. The amount of water accumulated within the drywell was calculated to be approximately 7,750 kg under atmospheric pressure after the completion of testing. This corresponds to approximately 49% of
Lawrsnca Livermore National Laboratory
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NUCLEAR SYSTEMS SAFETY PROGRAM TF82-220/0329a January 18, 1982 Mr. L. D. Watson E.G.& G. Idaho, Inc.
INEL CF601 Scoville, Idaho 83415
Subject:
Transfer of Salvaged Transducers / Equipment
Reference:
U.S. NRC FIN #A0ll6
Dear Mr. Watson:
As authorized and directed by U.S. NRC, we are shipping to you under separate cover, tranducers and equipment salvaged from our U.S. NRC sponsored 1/5-scale MKI BWR Pressure Suppression Experiment program.
In the enclosure we list the transducers by type, range, and serial number. Enclosed with the shipment are supporting manufacturers specification and calibration data.
! In a separate shipment, we are sending an unused pressure vessel designed for use as a 150lb/sec. flash boiler in the above BWR test program.
Yours truly, l.S Edward W. M8Cauley Thermo Fluid Mechanics Group Nuclear Test Engineering Division Dic/yd
! cc (w/ encl. )
T.M. Lee (NRC-DAE) 1 i
k,EyI %)cvwryErr1%cyx L Un'ssayof Caitvm ' PO B3M3 Lwmcre.CMforrM 94550 Teleahone(4B)422-1100
- Twx910-386-8339 UCLLL LWR 1
page 1 of 4 List of Salvaged Instrumentation (reference: UCRL-52314 and UCRL-52367)
Item Transducer Type Model No Operating Range Quantity 1 Pressure Senso Metric, #601087 0-100 psia 34a 2 Pressure Precise Sensor, #70114 0-75 psia 12b 3 Pressure Precise Sensor, 0-1000 psia 4
- 113-3-1000-A-34-5-P 4 Pressure Precise Sensor, 0-1000 psia 3
- 111-2-011-6-10-P-70106-34 (water cooled) 5 Pressure Kaman, fKP-1910 0-1000 psia 2 6 Differential Pressure Validyne #P24 I 2500 kPa 3 7 Differential Pressure Validyne #P24 I25psid 1 8 Differential Pressure Validyne #P24 I 100 kPa 3
+
9 Differential Pressure Validyne #P24 - 10 psid 3 10 Differential Pressure Validyne #P24 + 25 kPa 3
h a includes 13 unpotted units b includes 4 unpotted units m
page 2 of 4 -
List of Salvaged Instrumentation, cont.
Item Transducer Type Model No. Operating Range Quantity 11 Flow Meter Annubar, #754-316-SS - 2 Size 16.2 12 Flow Meter Annubar, #734-316-55 - 1 Size 5.047 13 Load Cell Interface #1232 t 100,000 lb 3 14 Load Cell Interface, #1220 150,000 lb 2 15 Load Cell BLH, #U3L2 I25,000 lb 10 16 Load Cell BLH, #U3L2 Il0,000 lb 2 17 Linear Potentiometer New England, #100 LIB-102-268 0-0.5 inch 8 i
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page 3 of 4 Transducer Serial Numbers Item # 1 34 each S/N (Potted w/RTV) S/N (unpotted) 5343, 5344, 5345, 5351 5341, 5377, 5399, 5407 5354, 5361, 5363, 5365 5408, 5410, 5411, 5412 5368, 5369, 5376, 5383 5413, 5414, 5415, 5416 5384, 5387, 5391, 5395 5417 5396, 5397, 5403, 5404 63 Item # 2 - 12 each S/N (Potted w/RTV) S/N (unpotted) 3286, 3288, 3289, 3290 3276, 3294, 3296, 3298 3291, 3292, 3293, 3295 (w/RTV boots)
Item No 3 - 4 each S/N 3255, 3266, 3524, 3525 Item No 4 - 3 each S/N DPT-127, DPT-127-1, DPT-127-2 Item No 5 - 2 each (w/ signal conditioner)
S/N 6335-0-01, 6335-0-02 .
I Item No 6 - 3 each S/N 24189, 24190, 24191 l
l Item No 7 - 1 each l
f S/N 25990 l'
Item No 8 - 3 each S/N 24196, 24197, 24198 l
[ _ _ .__ _ _ . ._.
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page 4 of 4 Transducer Serial Numbers, Cont.
Item No 9 - 3 each S/N 25991, 25992, 25993 Item No 10 - 3 each S/N 24193, 24194, 24195 Item No 11 - 2 each S/N 129853, 129854 Item 12 - 1 each S/N 132575 i
Item 13 - 3 each l S/N 7423, 74248, 7598A i Item 14 - 2 each l
l S/N 7782, 7783 Item 15 - 10 each -
S/N 31581, 31582, 31583, 31585, 31586, 31587, 31588, 31589, 31590, 31591 Item 16 - 2 each S/N 35411, 35412 Item 17 - each S/N None
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LAWRENCE LIVERMORE LABORATORY l
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N'JCLEAR SYSTEMS SAFETY PROGRAM March 2, 1982 TF 82-039/0385a Mr. E. Aust GKSS Institut Fuer Anlangentechnik .
Forschungszentrum Geesthacht OnbH.
Postfadi 1160 - Reaktorstrasse 7-9 D-2054 Geesthacht-Tesperhude Federal Republic of Germany
SUBJECT:
Proprietary Data Tapes-GKSS/PSS Program REFEREtCE: U.S. NRC/LLNL Liaison to GKSS Multivent Research Program FIN #A0116
Dear Mr. Aust:
Now that our liaison to your program has ended we are returning to you the proprietary data tapes whidi you so kindly lent to us. A summary list of these items is enclosed for your reference.
I take this opportunity to again thank you for your kindness and cooperation during the years of our association. It would be apareciated if you could continue to retain me on your distribution list for your research reports. Best regards.
Yours truly, Y t Edward W. McCdbley, Ph.D. ,'1T.E.
Thermo Fluid Mechanics Group Nuclear Test Engineering Division ENMc/ca t
cc: B. Bowman, LLNL ,
j T. Lee, NRC I
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Enclosure 1 TF 82-039 Summary List - GKSS/PSS Data Tapes Item Tape No. Name Item Tape No. Name 1 1 PROGA 6 18 LLLM2A 2 PRlGRA 19 LLLM2B 3 PSl006 20 LLLM2C 4 PSUDO7 21 LLLM2D 5 PStJ7A 7 22 LLW3A 2 6 LLLVIA 23 LLLM3B 7 LLLV1B 24 LLLM3C 3 8 LLLV2A 25 LLW3D 9 LLLV2B 26 LLLM3E 4 10 LLLV3A 27 LLLM3F 11 LLLV3B 28 LLLM3G S 12 DASM1A 29 LLLM3H 13 DASMlB 8 30 LLLM4A 14 LLLM1A 31 LLLM4B 15 LLWlB 32 LLLMAC 16 LLLMlC 33 LLLM40 17 LLWlD 34 LLLM4E 35 LLW4F '
36 LLLM4G
l_AWRENCE l_lVERMORE LABORATORY l
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NUCLEAR SYSTEMS SAFETY PROGRAM March 2,1982 TF 82-038/0384a Mr. M. Taylor Nutech International ,
6835 Via Del Oro San Jose, California 95119
SUBJECT:
TPC-SRV Test Data Tapes REFEREtCE: U.S. NRC/LLNL Liaison to Foreign Containment Research FIN #A0ll6
Dear Mr. Taylor:
Enclosed are the 10 data tapes for TPC-SRV main tests MT-1, -2, -10, -11, and -12. Also enclosed are the 4 data tapes for TPC-SRV shake-down tests
- SD.-1, -2, and -3.
Thank you for your extensive support and cooperation during our period of liaison to the 1981 SRV test program at Kuosheng #1. Please extend my sincere thanks and best wishes to your colleagues both at Nutech and Wyle Labs who were all so helpful to our efforts.
Yours truly, l .0 nst Edward W. McC.auley, Ph.D.Q.E.
l Thermo Fluid Mechanics Group l Nuclear Test Engineering Division i
! EWMc/ca cc: 8. Bowman, LLNL l 8. P-H. Chu, TPC T. M. Lee, NRC P. C. Liu, TPC/APD '
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May 14, 1982 Dr. T. Lee Division of Reactor Safety Research Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Eashington D.C. 20555 U.S.A.
Dear Dr. Lee:
On behalf of Mr. M. Shiba, Chief of Reactor Safety Lab. 1,JAERI, I am enclosing herewith copies of fourteen papers submitted by the JAERI CRT Group (with one exception) for oral presentation at domestic and foreign academic meetings.
Enclosures I through 12 are preprints for presentation at the Japan Atomic Energy Society (JAES) and the Japan Society of Mechanical Engineers (JSME) meetings. These are publically open although they are considered to be preliminary in nature. And we are afraid we have not sent to you all of these priprints. Contents of these priprints are as follows:
Encl. Top Meeting No. Author Presented at Contents 1 Namatame JAES '79 Fall Test facility and blowdown analysis 2 Kukita JAES '79 Fall Preliminary test results 3 Kukita JAES '80 Anual Chugging regime classification 4 Kukita JAES '80 Fall Pool swell test results 5 Takeshita JAES '81 Anual Continued blowdown analysis 6 Kukita JAES '81 Anual Chug desynchronization effects
- 7 Takeshita JAES '81 Fall Containment shaker-test results 8 Kukita JAES '81 Fall Single vent test results 9 Namatame JAES '82 Anual Chug desynch.-time-window analysis 10 Takeshita JAES '82 Anual Vent steam mass flux and air content analysis 11 Kukita JAES '32 Anual Evaluation of max. bounding condensation loads 12 Kukita JSME '82 Ntnl. Characterization of steam condensation loads l Enclosure 13 was presented by Mr. M. Utamura of Energy Research Lab., Hitachi, Ltd. at the same meeting as Enclosures 9 through II. His evaluation of the Mark ll plant load is based on the JAERI data, though it is not explicitly stated in the preprint.
i l Enclosure 14 is for my presentation at the oncoming ANS Anual Meeting held on June 7-9 in Los Angeles. (I am mailing a copy of this summary to Dr.
Lehner of BNL at his request.)
During my trip in June I wish to make a quick survey on the U.S. researches on containment integrity during severe accidents for some of my colleagues who are interested in this problem: I will attend a workshop on containment l Integrity hosted by Sandia National Laboratories and held on June 7-9 in Washington D.C., and asking for permission for my visit to SNL and the Hanford Engineering Development Laboratory to see the test facilities.
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Already I have sent a TELEX to Dr. C.J Anderson of NRR asking discussion with his staff at NRC~on June 4. Hopefully, I will be able to take to my trip draft of Evaluation report on Steam Condensation Loads (part 2) which is in print now, and results of dynamic-response calibration of the cavity-type transducers using the NRC/LLNL' transducers as the standard. These will be provided to NRC.
Sincerly,
& Wh'fC~
Yutaya kukita Reactor Safety Laboratory 1 Division of Reactor Safety Japan Atomic Energy Research Institute r
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November 18, 1982 Dr. E.W. McCauley '
burence Livermore National Laboratory University of California P.O. Box 808. Livermore, CA 94505
,U.S.A.
Dear Dr. McCauley The draft copies of your report describing the llNL modal test on the JAERI CRT facility have been duly received. Although we have made only a quick review of them, we are pleased to find general agreement between your and our results as to the system eigenfrequency and the damping factors. This viil be quite encouraging to us in evaluating significance of the fluid-structure interaction effects on the suppression pool hydrodynamic loads measured in the CRT facility.
Some of the figures in your report will be included in our project final report as supplementary and cupporting data, although we do not have intention to publish all the contents of your report.
Also, we understand that the financial and technical contribution made by the U.S. NRC and 11RL for this task shall be. strictly kept confidential as has been agreed. The same shall be true for the previous task concerned with installation of flush-diaphragm pressure transducers into the CRT facility. We vould very much appreciate your continued and careful considerations in this regard.
We are pleased to have had a beneficial co-operation with NRC concerning the Mark II containment hydrodynamic load issues, and to have contributed to the NRC's load evaluation program. We would like to express our sincere appreciation for your efforts made during years of your liason activities for NRC, that supported our co-operation with NRC. We particularly appreciate your understanding the administrative restrictions we have, and for sharing difficulties steming from these restrictions.
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Sincerely yours.
-o_Qa M bdd Masayoshi Shiba, Chief ,
3 Reactor Safety Laboratory I d Division of Reactor Safety Japan Atomic energy Research Institute MS/yk S
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,_ , _/ January 31, 1984 Mr. M. Shiba Reactor Safety Division Japan Atomic Energy Research Institute Tokai-mura, Naka-gun . .
Ib.araki-ben 319-11 Japan
SUBJECT:
Dynamic Testing of JAERI-CRT Facility Report
Reference:
- 1. U.S. NRC/LLNL Liaison to Foreign Containment Research
- 2. Ltr: E. McCauley to M. Shiba, October 28, 1982
- 3. Ltr: M. Shiba to E. McCauley, November 18, 1982
Dear Mr. Shiba:
I extend my greetings to you after such a long time. My assignment in West Gennany is ended and I am once again at LLNL. I have enjoyed reading the -
several published evaluation results from the CRT tests. The paper by Dr.
Kukita, et al, on the noncondensible gas effects is particularly interesting.
From the address given on that paper, it would appear Mr. Namatame has left JAERI. I am sure this is a loss for your Institute but a gain for his new company.
As we had agreed, the LLNL report on the CRT facility dynamic testing would not be published until your project final report had been released. This work is now apparently complete.
I am therefore writing to obtain your permission to release our report on the ;
subject dynamic modal testing results. Also, as you mentioned in reference 3, some of these results were to be incorporated in the CRT project final report. Would you please send me 2 copies of that report?
Please give my warmest regards to Dr. Nozawa, Dr. Kukita, and Mr. Takeshita.,
Sincerely, 9 i Edward W. McCauley, Ph.D., P.E.
g EM:ka cc: L. Cleland/G. Cummings 4;Mthe (NRC)pr C. Meier /
H.J. Weaver y 1278R/
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