ML20056F279
| ML20056F279 | |
| Person / Time | |
|---|---|
| Issue date: | 12/19/1988 |
| From: | Scarborough J NRC COMMISSION (OCM) |
| To: | NRC COMMISSION (OCM) |
| Shared Package | |
| ML20056F270 | List: |
| References | |
| NUDOCS 9308260286 | |
| Download: ML20056F279 (12) | |
Text
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,f UNITED STATES f"
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NUCLEAR REGULATORY COMMISSION
.E WASHINGTON, D.f'. 20555 g
b, j[
Decamber 19, 1988 OFFICE OF THE COMMISSIONE R MEMORANDUM FOR:
istr ut
. _[fN carbor~ough g -
FROM:
ac l
Technical Assistant to Commissioner Rogers
SUBJECT:
COMMISSIONER ROGERS VISIT TO THE FRG AND VIENNA, November 3-12, 1988 The attached trip report (Attachment 2) summarizes for your information the major facilities visited and discussions held during Commissioner Roger's visit to the FRG and IAEA Vienna from November 4-12, 1988.
Commissioner Rogers has the following questions (Attachment 1) on issues addressed in the report.
I will contact Messrs.
Beckjord, Murley, and Thompson or alternatively Mr. Stello to establish a process for resolving these questions.
Attachments:
As stated Distribution:
Mr.
D.
Rathbun Mr.
C.
Ader Mr.
S.
Burns Mr. J.
Gray Mr. H.
Denton, GPA Mr. V.
Stello, EDO Mr. S.
Chilk, SECY Dr. T. Murley, NRR Mr.
E.
Beckjord, RES Mr.
H.
Thompson, NMSS Mr. J.
Shea, IP 9308260286 930520 PDR COMMS NRCC CORRESPONDENCE PDR..,_
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Man-machine interaction studies and potential vulnerabilities, Further German risk assessment studies related to containment loading.
E.
References (1)
P. Weiss, M.
- Sawitzki, F. Winkler, "UPTF, A Fu-j ll-Scale PWR Loss-of-Coolant Accident Experiment Program," Atomkern-Enerale Kerntechnik, Vol. 49 (1986).
(2)
P.A. Weiss, R.J.
Hertlein, "UPTF Test Results -
First 3 Separate Effects Tests, Proceedinas of Ath WRSRIM, USNRC, October 27-31, 1986.
(3)
P.S.
Damerelkl, N.E.
- Ehrich, K.A. Wolfe, "Use of UPTF Data to Evaluate Scaling of Downcomer (ECC Bypass) and Hot Leg Two-Phase Flow Phenomena,"
Proceedinas of 15th WRSRIM, USNRC October 26-29, 1987.
3.
LWR SEVERE ACCIDENT RESEARCH, KARLSRUHE NUCLEAR RESEARCH CENTER The following day (November 4) we drove a short distance to the Karlsruhe Nuclear Research Center (KFZ) for discussions with Dr. Hans H. Hennies, Director and Member of the Board.
We were briefed for approximately two and a half hours by Dr. Hennies on KFZ's overall nuclear research program with particular emphasis on their severe accident safety research program.
A.
Kev Manacement Contacts Dr. Hans H. Hennies, Director KFZ Dr. Hermann Rininsland, Chemical Engineering Department, KFZ Prof. Kessler, KFZ Dr. Koerting, KFZ Mr. Wilhelm, KFZ B.
Facility Description and Purpose The principal programs at KFZ Karlsruhe are described in a brochure entitled Main Activities. Karlsruhe Nuclear Research Center in the trip report file.
. They include fast breeder technology, uranium enrichment (Becker nozzle) process, nuclear fusion, reprocessing and waste management, nuclear waste repository, environment and safety, solid state and materials research, nuclear and particle physics, microtechnology, handling technology, and other research activities.
Current emphasis in nuclear fission technology is on:
l Severe LWR accident core melt concrete j
interactions with the large scale BETA experiments, j
Detailed fuel assembly thermal degradation under severe overpower /undercooling transients with the large scale CORA experiments, Time dependent aerosol production, transport, and deposition within containment, and j
Containment aerosol filter development under accident conditions.
C.
Kev Research Results The initial condition of a Siemens-KWU PWR severe accident core melt (oxidic and metallic) is shown in Figure 3.
The core melt-concrete interaction has been simulated at KFZ by the BETA test facility, shown schematically in Figure 4.
A 10 MWe radiofrequency induction coil surrounding the concrete crucible deposits the requisite time-dependent thermal energy in the core thermite melt to simulate fission product decay heat for the 5 to 6 minute test run.
Detailed simulations of the core melt and concrete oblation under different initial conditions (temperatures), oxidic and metallic ccmpositions, concretes (silicates and carbonates), and crucible geometries have been performed with off-gases and aerosols carefully monitored.
Detailed three-dimensional analysis of material and l
energy flows throughout the concrete oblation of each test has been performed with the WECHSL code.
WECHSL j
invokes at least four different models for heat transfer from melt to concrete during different phases of the oblation and melt cooldown to the solidus temperature.
In addition to crucible spatial e
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' temperature measurements, off-gas analysis provides H 0, CO, and CO2 (release of CH4 molar fluxes of H2, 2
or other hydrocarbons is negligible).
The aerosol production (source term) throughout the BETA experiment is quantitatively measured.
(Detailed fuel assembly melting under prototypic overpower and under cooling core conditions is simulated in the separate CORA facility.)
Detailed results of BETA and CORA experiments and WECHSL analyses are presented in Appendix B.
For conditions prevailing in current FRG PWRs, WECHSL calculations normalized to BETA experiments indicate that for the core melt-concrete interaction phase, concrete erosion cannot be stopped by long-term sump water ingression into the (dry) reactor cavity and subsequent surface cooling of the interaction region.
l Under worst case assumptions Hennies indicated that
)
local penetration of the 6 m deep concrete basemat cannot be excluded after some five to six days.
HEPA Filter Development Related development emphasis has been placed on a high efficiency particulate aerosol (HEPA) filter which would function reliably under severe accident conditions of high containment temperature, pressure, and moisture.
A modular HEPA unit consisting of an approximate one square meter filter " pillow" of 2 to 30 micron diameter stainless steel fiber " fleece" has been developed as a highly efficient (DF of 104) aerosol filter under the environmental extremes (up to 250 degrees C) of a severe accident, pressure vessel breech, and core-concrete interaction.
Performance of these filters has been reported at the 18th through 20th DOE-NRC Nuclear Airborne Waste Management and Air Cleaning Conferences.
The new filters have been installed in 4 PWR and 2 BWR German nuclear power plants since June 1988.
Observations of H.H. Hennies During the course of the discussions Dr. Hennies offered a number of observations concerning LWR nuclear safety initiatives as follows:
Standardized Advanced LWRs have already been developed and deployed in the FRG - the latest Siemens-KWU " Convoy" PWRs and current BWRs.
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%J
%J These evolutionary, fourth generation ALWRs contain many of the design features of the large APWR and ABWR which have been developed by Westinghouse and General Electric with Japanese industry and utility participation.
Dr. Hennies believes the newest class of German i
ALWRs is " safe enough," even though reliance is placed on active engineered safeguards systems for beyond design basis accidents.
PSA studies show that core melt probabilities from internally initiated events are below 10 6/ reactor year.
Development of passively-actuated engineered safeguards systems in his opinion is unnecessary with highly reliable robust active safeguards systems.
Improvements can be made in the below reactor vessel concrete basemat design to better cool and contain the molten core resulting from a severe accident.
The next generation Siemens-KWU LWRs will incorporate improved basemat cooling as a potential passive safeguards feature.
The current Siemens-KWU Convoy PWR containment can withstand the internal pressure buildup for at least 5 days from adiabatic flashing of the primary coolant inventory and WECHSL-computed gas generation and aerosols from core melt-concrete interactions.
The WECHSL-computed gas generation and aerosol source term, based on extensive BETA and CORA experiment benchmarking, is the significant contributor to containment loading.
During the 5 day period before containment venting is necessary as shown.in Figure 6, more than 90 percent of the iodine, tellurim, and biologically significant rare earth aerosols deposit on the extensive concrete surface area within the containment (aerosol deposition has been confirmed by other extensive tests under representative conditions).
The deposition results in off-site doses at the plant boundary after containment venting which are within FRG regulations for normal operation.
Assurance of containment integrity for up to an approximate 5 day period following a severe accident, based upon the KFZ BETA-WECHSL and
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. I CORA research and development program, is considered by Hennies a major step forward in closure of the German severe accident research program.
Hennies noted that the concrete aerosol deposition surface area within containment is much greater in the Siemens-KWU PWR than in typical U.S. PWR designs.
a Dr. Hennies believes that economic incentives in Germany, where economies of scale are well documented from their construction experience, favor continued emphasis on large LWRs since available reactor sites are limited.
(The FRG, a nation having roughly the land area of the State of Oregon, has 11 PWR and 7 BWR operating power stations.)
Hennies does not see a role for downsized U.S.
advanced modular reactors for electricity production; the German modular HTGR on the other hand may have a role in German process industries.
Though our discussions did not focus on fuel cycle research, Hennies is a proponent of mixed oxide thermal recycle in German LWRs as a 30-40 year " transition regime" for fast breeders and a plutonium economy.
D.
Future Plans and Procrams KFZ future plans are generally described in the Karl-sruhe brochure cited earlier.
E.
References (1)
H. Alsmeyer, " Melt Concrete Interaction During Severe LWR Accidents," Karlsruhe Nuclear Research Center, to be published.
(2)
B.
Kuczera, " LWR Severe Accident Research Activities and Trends at KFK," undated paper.
(3)
V. Ruedinger, C.I. Rickets, J.G. Wilhelm, "The Realization of Commercial High Strength HEPA Filters," KFK.*
(4)
H.G.
Dillmann, H. Posler, J.G. Wilhelm, "Off-Gas Cleaning Devices for Containment Venting System," KFK.*
i A
p V
Q
. i l
(5)
C.I. Rickets, V. Ruedinger, J.G. Wilhelm, "The l
Flow Resistance of HEPA Filters in l
Supersaturated Airstreams," KFK.*
(* Proceedings of 20th DOE /NRC Nuclear Air Cleaning Conference.)
4.
MATERIALS TESTING INSTITUTE (MATERIALPRUEFUNGSANSTALT, MPA), STUTTGART After tours of research facilities and luncheon at the Karlsruhe Research Center we drove to our appointment with Professor Kussmaul at the Materials Testing Institute, i
University of Stuttgart the afternoon of November 4.
A I
history of the 100 years of testing, research, and teaching of the MPA Stuttgart is contained in a brochure in the trip file.
A member of the German Parliament, Dr.
Lauf, was on hand to meet Commissioner Rogers at the beginning of our MPA visit.
He stressed the severe negative impact on German nuclear programs of any nuclear plant problems in other countries includina the U.S.
He also urged the harmonization of international nuclear regulatory codes and practices as a necessary step in gaining international public acceptance of commercial nuclear power.
A.
Kev Manacement Contacts The MPA was separated administratively from the University where it was_ located in 1980 and assigned l
to the Ministry of Trade and Transport.
Since 1976 Professor Dr. Ing. Karl Kussmall has been director of l
the Institute and lecturer on the subjects Materials l
Testing, Materials Science, and Strength of Materials within the University's Mechanical Engineering curriculum with Professor Dr.-Ing. H. Dietmann.
B.
Facility Purpose and Description Present MPA activities comprise general material and component investigations and material development as well as welding techniques, forged nozzle-forged shell " joint" technology, high-pressure testing, Fracture mechanics, corrosion, tribology (wear),
failure analysis, reactor materials safety research, l
plant engineering, and calculation / investigation of large components and systems under operational and accident conditions.
Such investigations require
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. 1 APPENDIX B RESULTS OF BETA AND CORA EXPERIMENTS AND WECHSL ANALYSES OF CORE MELT-CONCRETE INTERACTIONS i
The CORA facility electrically heats. full length LWR fuel bundles to melting and failure, providing valuable data on bundle melting (individual fuel pin " candling"), melt relocation, and fuel rod fragmentation due to rapid water j
reflooding.
One interesting finding is that Inconel grid spacers initiate Zircaloy liquefaction at temperatures well below the liquidus temperature of Zircaloy.
CORA is providing data for validation of computer codes to be used in assessing the effect of accident management strategies involving injection of water into a damaged core, including the influence of quenching on clad oxidation and hydrogen production processes in damaged bundles.
Results from the BETA experiments include:
Greater downward heat transfer (and oblation) into the concrete than predicted by earlier modeling (high heat transfer results in rapid temperature reduction of the core melt),
Dispersion of the lower metallic melt into the upper oxidic melt under high gas fluxes, i
Low aerosol releases with the exception of an initial aerosol peak during and immediately after pouring of the
- melt, Fast preferential oxidation of chromium followed by iron oxidation reducing the freezing temperature of the molten concrete to about 1200 degrees C after some interaction 1
time.
The most important conclusions from the BETA and CORA experiments and WECHSL analyses are:
The slow pressure buildup would not fail the large PWR containment before 4 to 5 days, and overpressurization could be prevented by controlled venting.
(Hydrogen production may require ignition or recombination to prevent deflagration.)
Release of fission products into the containment after initial core melt-concrete interaction is very small (0.1
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- 32 g/s); filtered venting of the containment would retain these aerosols to an acceptable level.
Basemat penetration by the core melt to a depth of approximately 6 meters as shown in Figure 5 must be expected, and cooling at the underside of the basemat, either by engineered heat sinks (e.g., a sump water annulus extending further into the base concrete), may not necessarily stop the sideward progression of the melt into the basemat.
Development of improved containment concepts with passive accident survivability offer a design challenge for the future.
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