ML20006C417
| ML20006C417 | |
| Person / Time | |
|---|---|
| Issue date: | 01/10/1990 |
| From: | Williams P NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | Joshua Wilson NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| References | |
| PROJECT-672A NUDOCS 9002070479 | |
| Download: ML20006C417 (39) | |
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MEMORANDUM FOR: Jerry N. Wilson, Acting Chief, Advanced Reactors and Generic Issues Branch, Division of Regulatory Applications, RES FROM Peter M. Williams, Advanced Reactors and Generic issues Branch, Division of Regulatory Applications, RES
SUBJECT:
FOREIGN TRAVEL TRIP REPORTS i
Enclosed is a trip report abstract and individual trip reports of Sydney J. Ball and styself on our travel to England, France, and the F.R.G.
j on the subject of gas-cooled reactor research and operations. The visits i
were made between November 11 and 17, 1989.
1 Q
lA] b.l, A s t1.r' Peter M. Williams i
Advanced Reactors and Generic Issues Branch
+
Division of Regulatory Applications Offic,e of Nuclear Regulatory Research v
Enclosure:
-As stated 1
i cc:
E. S. Beckjord, RES J. L. Caron, RES l
D. F. Ross, RES M. Colagrossi, RES T. P. Speis, RES L. Eeltracchi, RES R. W. Houston, RES J. H. Flack, RES j
J. L. M. Cortez, RES J. C. Glynn, RES B. M. Morris, RES R. E. Johnson, RES Z. R. Rosztoczy, RES T. E. Murley, NRR T. L. King, RES W. D. Travers, NRR R. K. Frahm, RES C. L. Miller, NRR i
l M. K. Dey, RES S. H. Weiss, NRR OL: P;TGormley;: RES;a T. J. Kenyon, NRR g'W. B. Hardin, RES P. B. Erickson, NRR E. D. Throm, RES R. D. Hauber, IP
.Q H. J. Faulkner, IP
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TRIP REPORT ABSTRACT YEAR 1989 DATE OF REPORT 01/08/90 OFFICIAL TRAVELERS:
TRAVEL TO:
P. M. Williams, RES Heysham, Bootle and Harwell, England S. J. Ball, ORNL Paris and Grenoble, France Julich and Hamm, FRG BEGINNING ON:
11/08/89 L
OFFICE: RES Division:
DRA UNTIL:
11/17/89 l
Branch:
ARGIC A
A A
A A
MEETING TITLE AND/0R AFFILIATION:
Heysham-2 Reactor, Nuclear Installations Inspectorate, Harwell r
il Laboratory, England:
CEA (Fontenay-aux Roses), CLN (Grenoble), France:
i
~ ' s' -
KFA-Julich, THIR, Hamm, FRG ORGANIZED BY:
Country or Organization RES ABSTRACT AND/0R
SUMMARY
OF MEETING RESULTS The travelers visited gas-cooled reactor facilities and laboratories in the United Kingdom, France and the Federal Republic of Germany between November 8 and 17, 1989.
The purpose of this trip was to achieve primary
_ source information to benefit and improve the quality of regulatory staff assessments and decisions pertaining to NRC's safety review of the Modular High Temperature Gas-Cooled Reactor (MHTGR).
In each country, discussions were held on reactor design features, equipment performance, research activities, safety issues, licensing criteria, containment, and the human factors aspects of operations and accident mitigation.
Separate trip reports were prepared by each traveler and are enclosed.
European gas-cooled reactor operations and research have established a strong technical background for the MHTGR.
Gas-cooled reactors are expected to be operated in the U.K.
into the next century and in France to 1994.
In the FRG, both the AVR and the THlR are expected to be decommissioned without further operation, although no irreversible steps have yet been taken.
HIR research will continue at Julich to support possible export products for which I
markets are foreseen in the USSR, Italy, China, and the U.S.
France may be pQ~
considering HTGRs as a long term replacement for its PWRs.
fa i
REPORT OF FOREIGN TRAVEL TO THE U.K., FRANCE AND THE FRG PERTAINING TO GAS COOLED REACTOR SAFETY AND TECHNOLOGY l
PETER M. WILLIAMS MHTGR PROJECT MANAGER ADVANCED REACTORS AND GENERIC ISSUES BRANCH DIVISION OF REGULATORY APPLICATIONS OFFICE OF NUCLEAR REGULTORY RESEARCH PURPOSE The traveler, in the company of RES contractor Sydney J. Ball of ORNL, visited gas cooled reactor facilities and laboratories in the United Kingdom, France and the Federal Republic of Germany between November 8 and 17,1989.
The purpose of this trip was to achieve primary source information to benefit and improve the quality of regulatory staff assessments and decisions pertaining to NRC's safety review of the Modular High Temperature Gas Cooled Reactor (MHTGR).
In each country, discussions were held on reactor design features, equipment performance, research activities, safety issues, licensing criteria,-containment, and the human factors aspects of operations and accident mitigation.
Mr. Ball concentrated on information relating to research and our review of the MHTGR reseorch program.
His report was prepared separately and is given in m
fd Encl *iure 1.
No classified or proprietary material was discussed.
'#C010NS AND PRINCIPAL PERSONS VISITED Date Location Persons 11/8/89 Heysham-2 Reactor John Burchell, Principal Physicist i
Heysham, England Tom White, Assistant Operations Manager l
Neil W. Davies, UKAEA, Thermal Reactor l
Collaboration Manager 11/9/89 Nuclear Installations Dr. Derek Goodison, HM Deputy Chief Inspectorate (NII),
Inspector Bootle, England 11/10/89 Harwell Laboratory Dr. John R. Askew, Programme Director l
UKAEA, Oxfordshire, Gas Cooled Reactors England John Wilson, Deputy Programme Director
[
11/13/89 Commissariat a Daniel Bastien, Coordinator for Gas l' Engerie Atomic (CEA),
Cooled Reactors, CEA.
I Fontenay-aux-Roses, Marc Natta, Chef de Service, Department France d' Analyses de Surete f')
l l
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' 4 O
i
. 11/14/89 Centre d' Etudes Jean-Francoil Veyrat, Chef of Service, Nra.leaires (CEN),
Siloe Reactor Grenoble, France G. Dupont, Project Manager, COMEDIE COMEDIE Loop.
11/15-KFA, Julich, FRG Dr. Erwin Balthesen Director, 16/89 Program Management for HTR Development Prof. Dr. Wolfgang Kr5ger, Director, Institute for Nuclear Safety Prof. Dr. Hubertus Nickel, Director.
Institute for Reactor Materials Helmut Holmers Chief Engineer, TUV Hannover, FRG Michael Will, Project Leader, Interatom, Bergisch Gladbach Dr. E. Teuchert, Institute for Reactor Development Dr. Klaus Kruger, Reactor Analysis and Experiments (AVR)
.i 11/17/89 THTR-300 Reactor Dr. Rudiger Bliumer, VEW Hamm, Hamni-Ventrop, Plant Manager FGR Dr. Ivan Kalinowski, HRG Manager of Inspection Dept, and Chief
&j Physicist W
Hr. Norbert R6hl, Chief of Production INTRODUCTION, BACKGROUND, AND
SUMMARY
The U.K. operates two types of gas cooled re' actors - the Magnox and the u
Advanced Gas Conled Reactor (AGR); France, the Magnox type only; and the FRG, the High Temperature Gas Cooled Reactor (called the HTR in Germany and the HTGR in the U.S.).
The Magnox reactors, the earliest type, are graphite moderated, cooled with carbon dioxide, and use natural uranium fuel clad in a magnesium-
{
aluminum alloy (Magnox).
The AGR is also graphite moderated and carbon dioxide l
cooled, but uses low enriched oxide fuel clad with stainless steel.
It delivers steam at modern conditions (1000'F) and generally operates under steady state thermal conditions approaching those of the HTGR.
HTGRs are graphite moderated, helium cooled, and use a coated psrticle type fuel in which a layer of silicon carbide provides the principal barrier to fission product release.
The HTGR fuel can withstand temperatures up to about 1600'C without i
failure but must be protected from the ingress of chemically reactive fluids such as water, steam, or air.
The design of the primary system must take into account the potential vulnerability of the metals in the system to the high temperature of the helium coolant.
The visit to U.K. included visits to the Heysham-2 reactor, the Nuclear Installations Inspectorate (NII) and to the Harwell Laboratory.
In France, at the Fonteney aux Roses facility, the circumstances of four fuel melting g'
experiences in the Magnox reactors were discussed and a visit was made to the i
1,
t
' g V
COMEDIE test loop at Grenoble.
Here experiments are to be performed under contract to DOE on fission product transport.
At a two day meeting at KFA-Julich, FRG, essentially all aspects of MHTGR safety were discussed including a review of the highly successful operation of the AVR, a 15 MWe pebble bed type HTGR.
In addition, a visit was made to the Thorium High Temperature Reactor (THTR) in Hamm-Ventrop, a 300 MWe plant now shut down because of a combination of safety and economic reasons.
In summary, European reactor operations and research have established a strong technical background for both the AGR and MHTGR power plants.
Gas cooled reactors are expected to be operated in the U.K. into the next century and in France to 1994.
However, no further gas-cooled reactor construction is anticipated in the foreseeable future in any country visited.
In the FRG, both the AVR and the THTR are expected to be decommissioned without further operation, although no irreversible steps have yet been taken.
HTR research will continue at Julich to support possible export products for which markets are foreseen in the USSR, Italy, China, and the U.S.
France may be considering HTGRs as a long term replacement for its PWRs.
A detailed report of discussions held at each vir.it location is given below with further details given in.
Er. closure 2 is a listing of referenced documentation obtained at each visit.
Highlights of the trip are:
1.
The AGR development and operating experience represents a technology q
resource that has not been fully appreciated in the development of V
HTGR's, particularly in the areas of steam generators, circulators, and instrumentation and control systems.
2.
The approach of the Nuclear Installations Inspectorate has worked well as evidenced by a continuing high level of safe plant operations in the U. K, 3.
The upflow design of the U. K. gas-cooled reactors, although complicating the design of the top-entry control rod drives, deserves credit for contributing to the zero occurrence of fuel melting or damage over their entire operating history.
4.
France has experienced four fuel melting events with its gas-cooled reactors.
Like current U.S. and German designs these reactors have downflow cooling.
In my opinion, the success of upflow design in U. K,
warrants serious consideration as a design alternative for advanced gas-cooled reactors other than the MHTGR type.
5.
The DOE funded fission product transport experiments at Grenoble, France should be discussed with DOE, ORNL and GA in the following areas:
(1) test section configuration, (2) composition of dust, (3) effects of impurities, and (4) potential errors from successive blow-downs over the same test section.
6.
The immediate future for nuclear energy and the HTR in the FRG is uncertain l /7 because of a combination of economic and safety concerns.
Dr. Balthesen of lV the federal research ministry has proposed the intriguing idea that the i
' O MHT(G)R be developed as an international initiative.
I observed that in addition to potential savings in cost and time, an internationally developed and approved advanced reactor might find public acceptance more easily attainable.
7.
I have requested in writing that a report prepared for the state of Lower Saxony, " Safety Assessment of the HTR Module" be made available to us.
This report could aid our containment decisionmaking and the preparation of the final SER for the MHTGR.
8.
The German severe accident safety philosophy for their HTR-Modul plant includes " emergency protective measures" (i.e., operator actions) and af ter initial venting, automatic use of a filter train for longer term building releases.
9.
The dominent accident sequence for the HTR-Modul is the water / steam ingress event based on the rupture of a single steam generation tube.
Concern exists because of both reactivity addition effects and corrosion of graphite structures and fuel elements.
The explosion of combustible gases was stated as not likely.
10.
The fission product source ter,m for the HTR Modul is taken to be wash-off rather than lift-off. The Gemans expressed a desire to collaborate with DOE in the Grenoble experiments.
g 11.
KFA has impressive data on UO -type fuel integrity up to 1600 C but none 2
on UC0 (the reference fuel kernel for the MHTGR).
The German fuel has lower burnup and enrichment than the U.S. fuel mainly because of water ingress concerns.
12.
In a meeting with Prof. Nickel (KFA), he stated he would send us a RSK (Reactor Safety Commission) report on containment and severe accidents as soon as it is available.
13.
The THTR-300 is now planning for decommissioning after operating for about three years.
This is due to a combination of safety concerns, economic concerns, the unavailability of a fuel supplier, and the anti-nuclear climate in the FRG.
Valuable information on the pebble bed concept was obtained and it is unfortunate that long term operating data will not be obtained.
UNITED KINGDOM Heysham-2 Reactor The Heysham-2 reactor and the Torness-2 reactor in Scotland are the most recently completed AGRs and representative of the generally successful AGR program that has been~ developing for over 20 years.
These AGRs are designed to meet a " tolerable" risk criteria of 10 8 per reactor year and contain safety features judged beyond those required for current generation LWRs in the U.S.
i
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a These " extra" safety features are:
1.
Completely passive decay heat removal, except for valve actions, is provided by naturally connective flows through the core and steam generators to a naturally convective, air-cooled ultimate heat sink 2.
Decay heat is routinely removed by forced convention, and can rely on any one of four redundant reactor cooling quadrants.
3.
A means of secondary reactor shutdown is the introduction of pressurized nitrogen gas into the reactor, which has a worth equal to the control rods.
Other characteristics of the AGRs are:
1.
Water ingress does not cause positive reactivity insertion as for the MHTGR.
2.
Experiences with circulators has been excellent.
Circulators use oil bearings with essentially no leakage to the primary system and are fully inspected biannually.
The only loss of forced circulation was a twenty minute occurrence at the Oldbury Magnox reactor, r-3.
No plans exist for life extension beyond 35 calendar years.
Graphite
(_g growth under irradiation and corrosion by C02 are the determining
)
factors.
4.
The control room was designed to assure operator awareness of responsibilities and is not fully automated.
A PRA meter called the Essential Systems Status Monitor makes reports of changes of safety level for changes in plant or equipment status.
An NRR team with a BNL contractor specialist and a representative from Pacific Gas and Electric visited Hayshem-2 on November 29 and 30, 1989 and will provide detailed information and comment on this device.
Major difficulties, faults, and lessons-learned are as follows:
1.
The performance of steam generators has been generally good and steam generators are not considered replaceable.
A single tube rupture event at a bimetallic weld occurred at the Hartlepool AGR in 1987 and is documented in Reference 1, " Final Report of Boiler Tube Leak."
2.
An erroneous control rod withdrawal incident occurred from faulty computer control at an earlier AGR.
Now operators must authorize all computer controlled withdrawal steps and, as stated above, the Heysham-2 control room was designed to promote increased operator awareness.
3.
The dropping of a full length fuel assembly outside or with an open
'T
! -(' l PCRV is taken as the design basis accident of greatest consequence.
Events occurring inside a closed PCRV such as limited fuel melting are of lesser concern.
An unirradiated fuel assembly was dropped at the Dungeness AGR and is a basis for the current licensing hold on refueling at Heysham-2.
l
' O 4.
Oxidation of graphite (up to 20 percent consumption) and degradation of metals by the C0 coolant has been a problem.
This has been 7
mitigated by the addition of controlled quantities of CO and methane, and careful contr<.1 of the level of water vapor at about 100 ppm.
' provides further information on the AGR's licensing basis and current licensing difficulties, and describes important aspects of AGR reactor physics, fluid flow, operation, maintenance, and coolant chemistry.
In summary, the U.K. has brougnt gas cooled reactor technology to a high level of safety and utility. AGR reactor experience represents a technology resource that has not been fully appreciated in the'U.S.
The Sizewell B PWR was accepted in England with great controversy but future PWRs have been deferred.
Scotland has not adopted the PWR and if a demand existed (which it doesn't),
would build an AGR.
Nuclear Installation Inspectorate (NII), Bootle The NII has a regulatory responsibility similar to the NRC.
It has a technical staff of about 150 persons, but depends mainly on licensees.and applicants to perform the safety analysis and present a " safety case." The NII looks at 3
uncertainties and the needs for sensitivity studies.
The NII is administratively independent of the Energy Ministry and reports through the Health and Safety Executive to the Health and Safety Commission, a body that has wide safety responsibilities and is not confined to the nuclear industry.
g NII's major differences from the NRC are that there are no resident inspectors, safety research is generally performed under contract to the reactor developer, i.e., U.K. Atomic Energy Authority (UKAEA), and safety offenders are brought to court rather than directly fined.
The Advisory Committee on the Safety of Nuclear Installations (ACSNI) advises the Commission and also the Secretaries of State for Energy and Scotland.
Reference 2, "HM Nuclear Installations Inspectorate, Safety Assessment Principles for Nuclear Power Reactors" describes the organization in detail. Originally, licensing decisions were made on a Maximum Consequences Approach (MCA) where events were deemed credible or incredible.
Accident doses were and are compared with Emergency References Levels (ERLs) of 10 rem whole body and 30 rem thyroid as determined for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> at the site boundary.
Since 1974, " bottom line" PRA has had a dominant role in safety analysis and has lead to the " Summary of Events" approach.
In this approach, which is described in detail in Reference 3, "The Tolerability of Risk from Nuclear Power Stations," the " tolerable" risk of a large uncontrolled release that could lead to a single " eventual" cancer death is taken as 10 8 per year.
Single contributing events must be shown to have a probebly of 10 7 and meet the ERLs.
Initiating faults estimated at a probability of less than 10 7 are not included in the-design bases.
These include (1) multiple failures of steam generator tubes, (2) graphite fire (although the Central Energy Generating Board (CEGB) is performing studies), (3) large-scale fuel melting, (4) the integrity of the prestressed concrete reactor vessel (PCRV) and its closures, and (5) fuel or control rod ejection.
The design basis reactor accident is a one hour depressurization that is calculated to fail one fuel pin although failure of 25 pins is used in the consequence analysis.
Only the Heysham-2 and Torness plants are designed for earthquakes.
Sabotage is not h
considered by the NII.
There is little concern about the liftoff of plated-out
o 7
/%
)
fission products during depressurization because the primary coolant is kept extremely clean.
A trip exists on Kr-8B t:tivity to detect any f ailed fuel and human entry into the PCRV is allowed for inspection.
Plant operation is controlled by NII approved " Operating Rules" which are general in nature. A lower level document, " Identified Operating Instructions,"
is more like NRC's Tech Specs, and the most detailed procedures are found in the
" Station Operating Instructions." Maintenance errors are of major conc.ern and different maintenance teams service redundant systems.
Human errors are included in risk studies.
A thirty minute delay time is required for human actions, including a manual reactor scram.
The licensee is responsible for computer code validation.
Every code has a verification statement against another code, and the Windscale AGR prototype and startup data from earlier plants provide bench mark data.
Currently, reactor physics tests are not performed over plant life because of available earlier data. There ate no guides for AGR code validativ similar to those recently developed for the Sizewell-B PWR, a copy of which is given as Reference 3. "The Inspectorate's Approach to the Assessment of Code Validation Submissions." In summary, the NII approach for the U.K. 's gas-cooled reactors has worked well as evidenced by a continuing high level of safe piant operations,
!!arwell Laboratory
' h, The UKAEA maintains facilities at several locations.
Harwell is known mostly as a center for basic research. We met with Dr. John Askew and John Wilson, who will be taking over Dr. Askew's role of Programme Director for Gas-Cooled Reactors when Dr. Askew retires in April 1990.
Highlights of' the discussion were:
(1) The UKAEA performs safety research sponsored Dy the Health and Saft:ty Executive which reflects the needs of the NII.
The gas-cooled reactor research budget totals about 50 million dollars, about evenly funded by the operating plants and directly from the government.
The research topics include graphite and fuel irradiations in Harwell's materials testing reactors, reactor physics, shielding, nuclear instrumentation, sof tware development, and post irradiation fuel examination.
About half of the gas-cooled reactor research is considered " industrial tasks" and not directly related to safety.
Reference 4, " Advanced Gas Cooled Reactor Research and Development" provides details.
(E) The development of AGR technology followed three independent design concepts, not an evolutionary progression.
The Heysham-2 type is judged the most successful.
If a follow-on to Heysham-2 were to be built, it would have a " partial" containment structure to protect against the refueling accident, but the PCRV is considered adequate protection for reactor accidents.
Although removal of all the stainless steel fuel clad could add about 105 positive reactivity, excessive temperatures cause oxidation of the stainless steel and it does not flow away from the core
]
by melting.
e y
1
' Y (3) The earlier Magnox reactors used steel vessels, but as larger sized cores developed the stoel vessel design gave way to the PCRVs.
Steel vessels i
remaining in use are monitored for leakage as they are designed under the leak-before-break criteria.
(4) No flow modeling tests have been performed recently for AGRs, but at each commissioning, measurements were made of channel flows, clad temperatures, and outlet temperatures. Variations in outlet gas temperatures are kept to 10'C by " gagging" (orificing) flow thus eliminating hot streaks.
There have been no laminar flow problems.
(5) Three areas of possible collaboration were explored.
A.
Reactor physics data may exist at the UKAEA Winfrith facility that may be of potential use to MHTGR development.
Dr. Askew has preptrc.
a proposal for the recovery and the assessment of these data.
DOE would ba the expected prime user, but it should also be of value to the NRC in its safety review.
B.
Data on fission product retention and depletion in the primary system were obtained by a Harwell investigator (Mr. Reg Faircloth) at Hinkley Point in 1984.
We plan to look into this work and its relevance to fission product transport models in MHTGR accident studies.
O C.
Existing and developing data on graphites, metals, concretes, and instrumentation performance should be explored for possible collaborative activities.
Discussions were also held on nuclear power costs and a combined cycle HTGR, These are reported in Enclosure 1.
Dr. Askew believes the upflow design of the AGRs should be retained in advanced HTGR designs and that small power sizes are the most promising.
FRANCE Comm'ssariat a L'Energie Atomique (CEA), Fountenay aux-Roses France has built seven CO2 cooled, graphite moderated reactors of the Magnox type and three remain in operation.
By 1994, when the last of these (Bugey) is shut down, an experience of about 160 operating years will have been attained.
An important difference between the French and English reactors is that the French use downflow cooling, similar to Fort St. Vrain and the THTR.
There have been four fuel melting incidents, two of them sufficiently serious to require extensive shutdowns (more than a year) for repairs. The direct causec included flow blockages, instrumentation failures, the close margin between the normal peak cladding temperature (515 C) and its melting point (660'C), and operator errors.
Design modifications were made to prevent their recurrence.
Reference 5, " Twenty-nine Years of French Experience in l
.w
-w
l Operating Gas-Cooled Reactors" gives. details of these events and also describes problems and successes with graphites, metals, steam generators and PCRVs.
As a result of this experience Mr. Bastien recommends that a new gas reactor design should use... "a PCRV containing a modular, interchangeable steam generator placed to allow removal of residual power by natural convection.
The fuel would be able to withstand the temperature permitted by graphite and be cooled by an inert gas such as helium."
France is known to have had a strong interest in the HTGRs before it committed to PWR's in the mid 1970's, as evidenced by the COMEDIE loop (see below).
Mr.
Bastien also mantioned that the HTGR is being considered as an option for the eventual replacement of its current PWRs.
Centre d' Etudes Nucleaires (CEN), Grenoble - COMEDIE Loop DOE funded experiments are being conducted with the COMEDIE loop to measure fission product plateout and liftoff phenomena for use in the MHTGR safety analysis.
This facility, Corrosion and Migration Etude des Depots pour Irradiation (COMEDIE) is locatcJ in and adjacent to the 35-MW Siloe test reactor, and is a renovation and improvement of a facility built in the early 1970's.
The loop can measure fission product transport at varying helium flow rates without the disturbance of physically transporting the apparatut away from the reactor, a problem encountered in the earlier experiments.
'est 73 parameters, within the capabilities of the loop, are established by QE, and Q
the French contractors are charged with performing the experiments wider a QA plan that includes provisions for the inspection of materials used, instrumentation calibrations, and operating procedures.
The QA program was stated to be similar to the program developed with GE for BWR fuel testing.
The experimental program is expected to cost 6 million dollars and take up to four years.
The in pile portion of the loop contains two sections.
The irradiation section uses purposely flawed Fort St. Vrain type fuel to produce fission products and a depositing section collects condensed fission products on the cooled inner surfaces of three tube bundles.
The out-of pile portion contains sets of filters to measure lif toff under varying helium shear force ratios.
This section also contains a cooler, purification train, and a re-heater, and returns helium to the in pile section.
A simplified description of the loop and loop operations is given in Reference 6, " Helium Loop COMEDIE."
Four blow-down (BD) tests are planned.
BD-0 is a checkout test and will not use irradiated fuel.
It will be performed and analyzed by June 1990 and will include a check on the ability to vary shear ratios in the bundles.
BD-1 will use irradiated fuel and pure helium; BD-2 will include dust (a mixture of carbon and iron oxide); and BD-3 will be steam.
Planning efforts are concentrating on BD-1 which is scheduled for early 1991.
Depressurization can be performed from 60 to 6 bars about 3 minutes.
Generation and plateout of fission products and analysis of the data are time consuming and account for most of the period between tests.
AG
M l'
' O Questions of the relevance of test parameters to MHTGR data needs were necessarily deferred to discussions with DOE and its GA and ORNL contractors.
For example, the experimenters were unaware that fission product depositions and removal for the MHTGR will be on the outside of tubes with cross flow rather than inside and lengthwise as in the test loop.
Other deferred questions include; how is the composition of the dust determined, what are the effects of impurities on liftoff, and will successive blow-downs over the same deposition surface result in erroneous data */ We also need to know the results from the previous COMEDIE experiments, what is being learned from the current MIT work, and what progress is being made on plateout and liftoff modeling.
FEDERAL REPUBLIC 0F GERMANY Glossary In order to aid the following reports a glossary of German acronyms is given below.
- Arbeitsgemeinschaft Versuchs Reactor, a 15 MWe pilot pebble bed HTR operating for 22 years.
BMFT
- Bundesministerum fur Forshung und Technologies, Federal Ministry for Research and Technology, g
BMU
- Bundesministerum fur Umgebung, Federal Ministry of the Environment.
GRS
- Gese11 shaft fur Reaktorsicherheit, Corporation for Reactor Safety.
HTR Modul
- A 200 MWT " side-by-side" modular HTGR similar to the MHTGR but uses pebble bed fuel.
HTR 500
- A 500 MWe HTR being designed by the high temperature reactor corporation (HRG).
KFA
- Kernfor'chungsanlaga, Nuclear Research Center (recently the word nuclear has been dropped from the title).
RSK
- Reaktorsicherheitskommission, a federal advisory group similar to the ACRS.
PTH
- Program Management for HTR Development, an organization of the BMFT.
THTR 300
- Thorium High Temperature Reactor, a 300 MWe demonstration HTR that operated in the state of North Rhine-Westf alia.
TUV
- Technische Uberwachungs - Vereine Technical Inspection Agency.
'N'
'- q.
.<)
program Management for HTR Development, KFA, Julich We met with Dr. Erwin Belthesen, who is Director of PTH and serves as the federal coordinator for HTR research and technology, and also with Mr. Helmut Helmers of the Hannover TUV and Mr. Michael Will of Interatom, the industry.
developer of the HTR - Modul.
Dr. Balthesen outlined the duties and relationships between the licensing authorities, KFA-Julich and industry.
The 3
principal licensing authority resides with each individual state, most of which have their own TUVs.
The TUVs, which are long established technical organizations, provide advice on nuclear and other safety matters and together founded GRS. The RSK advises the federal government on nuclear reactor safety but not normally the states.
Dr. Balthesen coordinates HTR development work outside KFA.
Environmental concerns are the province of the BMU.
The immediate future for nuclear energy and the HTR in the FRG is uncertain for a combination of c0 anomic and safety concern reasons.
The ready availability of both German coal and imported electricity from France, together with a major political party, the Social Democrats, embracing the post Cheranbyl nuclear safety concerns of the " Greens," leave nuclear energy as a long term option.
A
" critical mass" of 110 persons will remain oriented to the HTR at KFA-Julich in the future, but KFA, now simply FA (research center), is expanding in other areas.
The plan to perform a fission product transport experiment via blow-down of the AVR prior to its decommissioning has been terminated due to public objections and, as will be further discussed, the THTR-300 is also q
seeking a decommissioning license.
V A HTR-Modul licensing procedure in the state of Lower Saxony has been cancelled, although the technical review was close to permitting issuance of the licensing equivalent of an NRC construction permit.
A report, " Safety Assessment of the.
HTR Modul" prepared by Interatom, should be available to us by request to Dr. Balthecen and should be helpful in our development of the final SER for the MHTGR.
While this is a site independent report, it does not cover severe accidents-Dr. Balthe;.en gave us a recent IAEA consultants report, "Next Generation of Nuclear Power and the Role of the IAEA." Reference 7.
Briefly, this paper suggests an International Steering Committee that would establish " General Plant Requirements" for all nuclear power plants, and a Safety and Design Assessment Team (SADAT) that would review future designs.
Within the gas cooled reactor international community, Dr. Balthesen had advanced the idea that development of the MHTGR should become an IAEA initiative, and while support was received from many nations, support was not forthcoming from the U.S. DOE, and the U.K.
The NRC was not consulted on this matter.
KFA-Julich KFA is comprised of a number of " institutes." Those institutes that we met wit h were reactor safety (Prof. Dr. Wolfgang Kr6ger), reactor materials (Ptof. Dr. Hubertus Nickel) and reactor development (Dr. E. Teuchert).
Prof. Dr. Rudolf Schulten, the former director of the reactor development
,O institute has just retired, and Prof. Kr6ger is leaving KFA for the V
Paul Sherrer Institute. The institutes of reactor safety and reactor development will be merged and a new director will appointed.
' O:;
Prof. Kr6ger outlined the German safety philosophy and requirements.
PSA (i.e.,
PRA) is used to make risk estimates down to 10 8 per year for guidance purposes, but there are no quantitative safety goals.
The guiding principle of plant safety is defense-in-depth, with " emergency protective measures" to limit radionuclide releases to "as low as possible" levels for accidents beyond the design bases.
Normal annual dose limits to both the puolic and workers are 30 mr whole body and 90 mr thyroid.
Limits for " design-relevant" accidents are 5 rem whole body and 15 rem thyroid.
The RSK is preparing a statement on low probability accidents.
A goal is no public involvement in the event of an emergency.
Reference 8, "Probabilistic Safety Analysis of the HTR-Module" has determined that the dominant event sequences are initiated by leaks in the steam generator
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(water ingress) and leaks in the primary circuit (depressurization), and that transients (loss of forced cooling, loss of all cooling systems, and loss of feed water) have a minor significance.
Water ingress is the dominant concern principally because of r(activity addition effects, but also because of corrosion of graphite structures and fuel elements. The explosion of combustible gases was stated as not likely, although this had been a concern from earlier studies.
The DBA is taken as the failure of a single steam generator tube followed by reactor trip, shutdown of the circulator (by the three diverse signals of high moisture, high power and high pressure), and steam generator isolation.
Like the MHTGR, the location of the steam generator below core level in a separate vessel inhibits water and steam entry by natural g.
convection.
The gas purification system is sufficiently large that a single tube failure can be accommodated without lifting a pressure relief valve.
The PRA study considered steam generator manifold fracture and up to 20 tube failures can be accommodated, although the relief valve would lift. Both small leaks (less than 2 square centimeters) and large leaks (up to 33 square centimeters - a refueling tube) were considered in the analysis of depressurization events.
The source term and the containment philosophy have some similarities and differences from that of the MHTGR current design.
Like the MHTGR design,
" weak fuel" and fuel overheating are not taken as a concern by the designers.
The major source of fission products is taken as washoff rather than liftoff from steam generator tubes.
The Germans expressed interest in exploring collaboration with DOE in the Grenoble experiments.
The Germans have models for plateout and liftoff, but they state they have no reasonable model for dust.
Graphite fire is not considered a credible source term but studies, in the wake of Chernobyl, have been made showing that with unrestricted air flow through the core it would take at least seven hours to burn off enough graphite from the fuel balls to expose fuel particle surfaces.
The Germans are also considering a sic coating on the fuel ball surface as a further defense against graphite fire.
A confinement building is kept at negative pressure, and vented if an overpressure occurs followed by an automatic switchover to a filter train.
Containment leak tightness is not required, and maintenance and testing are not equivalent to PWRs.
Concrete is unlined and the use of an aluminum oxide type concrete for high temperature purposes was not considered practical.
More details on the containment-confinement design were requested, h
such as the design basis of the filter train, but these were not available.
13 -
A The operators' role in accident mitigation is taken into account in the safety analysis.
After 30 minutes, manual actions my be taken to:
(1) start the auxiliary circulator and emergency heat exchanger in the helium purification system, (2) close the primm y system relief valves, (3) close confinement relief values, (4) add water (fire brigade) to the water cooled reactor cavity cooling system, and (5) start diesels, which are not required to be available rapidly.
Descriptions of analytical and experimental work pertaining to core heat-up sequences are given in Reference o
?resentation of Thermodynamic Investigations into the Safety Res,ach and Development Work for the HTGR," and Reference 10 " Comparison of Theoretical and Experimental Studies of Af terheat Removal by Natural Convection / Circulation from an HTR."
Calculations were made for both the HTR-Modul and the THTR by the THERMIX-CRAY-20 code, which has been compared successfully with the experiments from the LUNA-HTR loop, as discussed in Reference 10.
A plan to use the LUNA loop for prismatic fuel experiments has-been considered by DOE, but has not been funded nor has a QA program been developed.
Reference ll, " German Research and Development for HTR Fuel," describes past research and the current design.
The Germans have been developing coated particle fuel for almost thirty years, both by irradiation testing and in operating reactors (AVR and THTR).
Their TRISO fuel has overall similarity to
,q the MHTGR fuel, but also important differences; the principal bases for which V
are the lower compaction density in the spherical elements versus the GA fuel stick design and the need for low enrichment (8% vs 19.9%) and burnup (8.2% vs 26% FIMA) to reduce the reactivity additions in the event of water ingress.
In addition, the fuel kernel is pure U02 rather than a mixture of UO and UC.
2 2
The Germans commented that the addition of UC to the MHTGR fuel is to minimize 2
the amoeba failure mode with the higher U.S. burnup and it also was said to facilitate the molding of the fuel stick. The as-manufactured quality and irradiation performance specifications are roughly the same for the German as U.S. fuel particle designs. The practical temperature limitation for the German fuel is 1600 C, above which the inner surface of the sic layer begins to be attacked by fission products.
This phenomena is described in detail in Reference 12, " Fission Product Release Profiles from Spherical HTR Fuel Elements at Accident Temperatures," and Reference 13, " Passive Safety Characteristics of Fuel for Modular HTR and Fuel Performance Modeling under Accident Conditions." The fuel development program on UC0 fuel is slow because it is not well funded and the ORNL HFIR reactor has been unavailable.
The irradiation program at Petten, which includes UC0 fuel, is international, with hydrolysis studies being paid for by the U.S.
Our visit to the Reactor Materials Institute concentrated on graphite development and high temperature metals.
The Germans design for no replacement of reflector graphite, so extensive work is performed on creep behavior at various temperatures and loads.
Thermal conductivity is not sensitive to graphite grade but to temperature.
It quickly drops to about one-third its initial value with irradiation.
Graphite specific heat is well known.
Recent
(~')
work on high temperature metals is included in Reference 14, " Proceedings of the v
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' ol Workshop on Structural Design Criteria for HTR."
The AVR experienced only a single pin hole leak in its-steam generator.
However, the experience at the Hunterston AGR reactor has supported the decision to eliminate bimatallic welds from the HTR-500.
In a meeting with Prof. Nickel, he stated that he was a member of the RSK and would send us its report on HTR containment when available.
Prof. Nickel i
stated he personally supports the GA containment concept.
He further stated he had met recently with members of the ACRS and is arranging a visit for them to Julich in the spring.
The main topic of discussion with Dr. E. Teuchert and W. Scherer of the Reactor Design Institute was the rea::tivity addition from a water / steam ingress accident.
The Germans consider this accident as a major safety concern, a choice reinforced by an AVR event where 26 tonnes drained unobserved from the overhead steam generator through the shutdown core of the AVR to the lower plenum. region.
The reactivity addition occurs because of a decrease in neutron leakage, but passes through a maximum as more water is added.
The effect is much greater in a cold reactor because of control rod shadowing.
The maximum leads to the "Chcrnobyl Syndrome," a hypothetical event postulated by the
" Greens," in which a HTR initially containing a large quantity of water would
" steam-out" water on startup leading to a-reactivity increase, or " positive coupling." The water effect was studied 10 years ago at the University of Gratz without temperature variation and additional studies are planned at the g
Paul Sherrer Institute in Switzerland, again without temperature variation.
To date, good agreement between experiments and calculations have been attained, but data with temperature variations are needed.
Modeling of rates and amounts of water / steam ingress has been underway for two years and continues.
Investigation topics included postulated initiating events, equipment failures, power peaking effects, and how steam gets out of the core.
Current work is concentrating on developing models and codes.
At present not enough studies have been made to determine data needs.
Reference 15, " Engineering and Licensing Progress of the HTR-Module," discusses in some detail progress and status of the water / steam ingress modeling.
During our visit to the AVR, we first discussed its highly successful operating history, which is summarized in Reference 16, " Experience Gained with the AVR Experimental Nucli Power Station." This history includes the demonstration of an ATWS event (
ss of forced circulation) in which the reactor shut itself down by its negative temperature coefficient and xenon buildup.
Recently its safe response to a LOCA was also demonstrated.
The second topic of discussion was dust in the primary system which was found to be mostly of small carbon i
particles less than one micron.
It was estimated that less than half the plated-out activity resides on the dust and most is on the steam generator.
Half of the dust is said to be everywhere and half is on horizontal " dead" surfaces.
This is indicated by the rapid increase in dust for increasing circulator speed.
Further information on dust may be available from Prof. Vonder Decken, head of a different laboratory.
O
' THTR-300, Hamm-Uentrop The THTR was first synchronized to the grid in November 1985, and has been shutdown since September 1988.
Because of a combination of safety concerns, financial problems, and the demise of its fuel manufacturer (HOBEG) the THTR is now planning for decommissioning, although a decommissioning license has not yet been issued and no irreversible steps have been taken in plant dismantlement.
A movie of the THTR's construction followed by a tour led to the discussion of the following points:
(1) there have been no leaks in the PCRV liner tubes, (2) the exterior of the PCRV is enclosed in a metal shroud that is kept at negative pressure and vented through filters to the stack (3) the leakage rate of helium is one third of the inventory per year (4) roller bearings allow shifts of the reactor internals relative to the PCRV, (5) the uses of " keys" allow the installation of graphite internals without restraint, (6) each closure is double flanged, (7) there was excellent agreement with reactor physics, fluid flow, and heat transfer calculations at startup, (8) fission product retention and load following have been as expected,.and (9) the personnel dose has been low.
Reference 17, "THTR 300 MW Nuclear Power Plant Hamm-Uentrop," describes the THTR in some detail; Reference 18, "The Start Up and Initial Operation of THTR 300" presents the early operational history; and Reference 19, "THTR Commissioning and Operating Experience," includes q
description of major problems.
These problems are:
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1.
Eight thousand out of 625,000 fuel balls were broken as a result of control rod insertions into the pebble bed.
Testing to establish rapid insertion capabilities was the main cause.
Later, operators became experienced with control rod maneuvers and many fewer balls were broken.
2.
Reduction to 40% power was necessary for on-line fuel ball removal.
Excessive counter current helium flow in the fuel discharge pipe necessitated frequent power reduction to permit gravity exit of the balls.
A subsequent plant modification that bypassed some of the counter flow permitted on-line removal at full power.
3.
During the summer of 1988, power reduction was repeatedly required to keep L
the-ambient air temperatures within permissible limits in those parts of the reactor hall that contained components of the steam and feed water circuit.
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4.
Bolt failure occurred on a few of the thermal barrier cover plates in the i
hot helium ducts from the outlet plenum to the steam generators.
Thirty-five of 2600 bolts were found broken during the September 1988 inspection, mostly at the plate centers, by a mechanism thought to be thermally induced bowing of the plates.
The THTR staff is convinced that operation could continue with a monitoring program to assess the impact of l
any further bolt failures.
Replacement of the bolts was judged extremely difficult.
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. O 5.
The rapid ball motion through the center of the core was not predicted and flow, heat transport, and reactor physics calculations _became less accurate than at the beginning of core life.
Many factors are believed to be causes for this unpredicted phenomena.
In spite of the difficulties identified, the THTR can be said to (1) confirm the pebble bed concept for use in higher power designs (such as the HTR-500), (2) provide an important data base for code validation (e.g., steam generator transients), (3) demonstrate the viability of many design choices (e.g., use of downflow boiling, oil lubricated circulator bearings, and ammonia gas lubricant for in-core control rod motion), and (4) demonstrate good load following capability (ramps of 8% per minute) and (5) d-monstrate low graphite corrosion (well less than 1% at H O levels at 1 ppm).
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FOREIGN TRIP REPORT
- ORNL/FTR 3477 DATE:
December 4,1989 sua n cTi Report of Foreign Travel of S. J. Ball, ORNL Manager of NRC-Sponsored HTGR Safety Studies, Instrumentation and Controls Division T 0:
A. W. Trivelpiece F Rou:
S. J. Ball PURPOSE:
To visit nuclear installations in England, France, and West Germany to obtain primary source information needec for reevaluating DOE's rewarch program plan for the modular 11TGR in the areas of primarv system components, reactor operations, fuel performance, reactor physics, heat transfer and fluid flow, fission product transport, safety analysis, and licensing criteria.
SIT 11S
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VISITED:
11/8/89 Heysham-2 Reactor Dr. Neil W. Davies, Heysham, England Thermal Reactor
. Coordination Manager,UKAEA 11/9/89 NuclearInstallations Dr. Derek Goodison, Inspectorate (NII),
Branch Chief Bootle, En' gland i
11/10/89 UKAEA Laboratories, Dr. John R. Askew, l
Harwell, England Director, Gas-Cooled.
I Reactor Progmm I1/13/89 Commissariat Alinergie Mr. Daniel Bastien, Atomique (CEA),Fontenay Coordinator for Gas-aux Roses, France Cooled Reactors 11/14/89 Centre dItudes Mr. Jean Francois Nucleaires (CEN),
Veyrat, Chief of Grenoble, France Service 11/15 16/89 KFA, Julich, FRG Dr. Erwin Balthesen, Director, HTR 1
Development 11/17/89 THTR-300 Reactor Site Dr. Rudiger O
Hamm. FRG Eitumer, vEw.
p Plant Manager l:
I UCN-2 3 8 ) A (3
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q ABSTRACT The traveler was asked by the U.S. Nuclear Regulatory Commission (NRC/RES) to travel with Dr. Peter M. Williams, NRC MHMR Project Manager to assist in obtaining information fmm researchers and licensing authorities in the United Kingdom and Western Europe relevant to the NRCs ongoing evaluation of the DOE Modular HTGR (MHTGR) research program plan and licensabihty concems. De NRC-sponsored ORNL prognm for HER safety reviews, of which the traveler is manager, has made significant use of foreign resources in conducting safety research, developmg independent safety analysis capabilities, and assisting NRC in preparation of safety analysis reports. De additional information derived on this trip from detailed discussions with researchers and licensing authorities, laboratory and reactor site tours, and literature received will be very valuable in carrying out the NRC program. Specific information and insights were obtained in the areas of primary system component xrformance, reactor operations, control and safety.
system design and performance, fue. performance and fission product transport, safety analysis, heat transfer and fluid flow, reactor physics, advanced designs, and licensing criteria and methods.
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p H'IBR RESEARCH AND LICENSING TRIP TO ENGLAND, FRANCE, AND j
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Hevsham-2 AGR t
he site visit to the Heysham 2 Advanced Gas Reactor (AGR) provided useful information on AGR operation, analysis, licensing, and design features and problems, much of which L
is applicable to HTGRs. The AGRs are a good exam)le of how the evolution of a design can result in a much-improved, smoother operation..Recently, however, a serious licensing problem has supped up. Previously, British reactors were licensed on the basis of " deterministic calculations of maximum credible accidents," while for the newest AGRs, probabilistic risk assessments (PRAs) are used in the AGRs, the predominant risks are from refueling accidents, and the sum of the faults must be <104/y. Heysham-2 is now coasting down in power, unable to refuel, because the calculated probability of dropping a fuel assembly dunng on line refueling is too large. De problem surfaced in a safety review, wherein it was decreed that smce two " independent" digital safety systems used the same type of processor hardware and the same pu pi-uing language (although designed by ditterent groups), the probability of failure is a tactor of 10 higher than claimed. The Heysham case is substantially weakened by the fact that an assembly was dropped during a refueling at another AGR. To license or not to license, based on low-probability PRA numbers, appears to be a very risky business due to the large uncertainties in the q
calculations and the vulnerability to criticism.
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.In the AGRs, the large graphite moderator blocks surrounding the (clad) fuel assemblies are cooled to nearly the inlet gas temperature by a major (60%) sidestream flow that later joins with the rest to cool the fuel. Hence, the bulk moderator temperature is relatively independent of power, so even though the moderator temperature coefficient of reactivity is positive and a factor of 8 larger than the (negative) fuel coefficient, the power coefficient stays negative throughout the fuel life. For fast transients, the fuel coefficient dominates and effectively ternunates the spectrum of " allowed" reactivity transients. Total-loss-of-flow accidents are kept to an acceptably low pmbability,by having four independent cooling system quadrants, any one of which could provide adequate shutdown cooling.
Heysham-2 has an interesting on line computing system, which calculates the varying risk of fuel damage as a function of post trip cooling equipment condition and availability. For configurations in which the base estimate for risk mcreases by a factor of 10, routine maintenance is allowed, while for factors of 100, emergency maintenance and 36-h c
shutdowns are mandated, m
ne 88 control rods are positioned automatically to keep assembly outlet temperaturer to within 10 C of the average, with manual assistance from operator control of the assembly inlet flow orifices (or " gags"). Individual region power control is limited by calculated pellet-clad interaction failures that would result from the design-basis depressurization accident. Heysham-2 had tried a more automated plant control system (for startup, power maneuvers) but went back to the more traditional one because of problems with unreliability.
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1 el Coolant chemistry for the AGR primary system (carbon dioxide) is quite different from that of the H'IGRs (helium). The normal moisture levelis -120 ppm (vs <1 ppm) and they make sure it stays >10 p )m to avoid dry atmosphere friction problems, and have a high-level trip at 350 ppm. Methaneis added to inhibit corrosion, which is coasiderable (20%
graphite weight and strength loss is expected over the plant lifetime). De moisture 1
detectors (fogged mimrs) are satisfactory. Primary coolant leakage is quite high
(~1%/ day) through particulate filters, with provisions im diverting leaking coolant through charcoal filters in an accident. De primary system is normally very clean, as they are able to do hands-on maintenance of the steam generators and circulators. All AGRs recently added a tdp on high circulating activity.
Nuclent Installations Insnectorate (NII)
Plant licensing is donc quite differently in the United Kingdom, but like the USNRC, the NII maintains a highly competent technical staff. De nil sets policy, reviews safety-related designs and operational problems, and grants operating licenses. It does not do independent research or safety analyses, but does advise on safety research done by UKAEA. Oversight of plant operations is limited to much higher levels (no resident inspectors, for example). One of NII's most effective administrative tools is apparently prosecution of a utility in the courts, since plant personnel tend to see themselves as personally involved in such lawsuits.
The U.K. equivalent of U.S. tech specs for a plant are a much broader set of " Operating Rules," which are subject to NII approval. At the next level are the " Identified Operating a
Instructions," which are more like the U.S. tech specs, but which are not subject to nil W
approval; and likewise for the most detailed documents, known as the " Station Operating Instructions."
Plant and safety system designs accommodate the operator action guidelines, which mandate that operators must not take action for the first 5 min following a scram, and must not be required to take action in the first half hour: The nil is generally more concerned with maintenance errors than with operator errors, and is developing rules for diversity of maintenance personnel (and procedures?) to avoid common mode failure problems.
In the area of severe accidents, several items ofinterest were noted: (1) NII staff pointed out that the cladding (reactivity) worth is ~$15, so it is essential that it stays intact to avoid prompt critical accidents. It was later noted at Harwell that experiments do not show the clad " falling r.way" when the fuel bundle is heated to very high temperatures; (2) graphite fires are excluded due to the low probability of multiple failures in the vessel; (3) water ingress accidents do not result in any significant positive reactivity insertion. A writeup was obtained on the one and only major AGR water ingress, a steam generator tube rupture at Hartlepool 1 in March 1987 that resulted in an ingress of several tons of water, and (4) finally--on the Magnox reactors at least-NII allows leak before-break assumptions for the vessels based on inspection criteria.
I obtained a writeup on NII policy and guidelines for accident code validation and verification, which were developed during the 2.5-year-long Sizewell-B (PWR) public hearings. This policy could be useful to the NRC in drawing up a similar guide for use by e
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U.S. licensees. Other useful documentation received included the U.K. regulatory policy on source terms and siting, policy statements (repons) on risks and safety assessment principles, and independent safety evaluations commissioned by NII of the U.S. and FRG advanced HTGR und LWR designs.
i UKAEA Harwell Laboratories We met with Dr. John Askew and the program's Deputy Director, John Wilson, who will -
take over as Director when Askew leaves next
' g. 'Ihey commented on the very recent change in England's power plant privitization to exclude all nuclear plants. Repons on BBC had stated that, consit ering costs of decommissioning, the nuclear power cost would
- be three times that of coal plant power. Askew sam UKAEA estimates that included complete dismantling and disposal after d==nmissioning (which he thought unnecessary) put nuclear costs only about 10 to 15% higher. England had not plenned to build any more AGRs, and recently announced that no more PWRs will be staned after Sizewell-B.
Two areas of possible collaboration were discussed. First, Askew agreed to write up a proposal for a contract to retrieve and analyze the data from Winfrith critical experiments that would be relevant to MHMR needs. Data from heated assembly and simulated steam ingress tests are included. Second, from discussions it appeared that Harwell's primary investigator for fission product (FP) ex planning or evaluating the proposed e@periments, Mr. Fancloth, could be usef
.entsin the MHER R&D program.
Dr. Askew also described an interesting reactor design that he has proposed to IAEA: a i
l pJ small combined cycle HTGR in which the use of a steam generator m place of a (direct cycle) recuperator would result in reduced pressure drop and capital cost, plus higher efficiency.
Commissariat h I' nergie Atomiaue Discussions at CEA included detailed descriptions by Mr. Bastien and other CEA staff members of the French experience with gas-cooled reactor operation, which is considerable
(>l50 reactor-years). Cunently, only three of the Magnox plants are still operating, and the last of these is to be shut down in 1994. To date, four accidents have resulted in significant fuel damage, and the lessons learned from these were discussed. The CEA expen on plant decommissioning and dismantling, Mr. Bemard Giraudel, noted that dismantling a ple.nt is much easier for PCRVs than for steel vessels (he has done both).
This news should be of interest to Fon St. Vrain and THTR 300 owners.
J Centre d'$tudes Nucleaires i
The purpose of the CEN visit was to review the pmposed DOE-sponsored fission product (FP) transpon experiments in the COMEDIE in-pile loop at the Siloe reactor. Siloe is a 35-MW pool type reactor used mainly for physics research and materials testing. The L
COM EDIE loop is being modified to accuruuodate depressurization tests to study FP L
liftoff from simulated steam generator (SG) tubes.
l CEN personnel noted that they were providing a service to DOE wherein they are given the loop design, test, measurement, and chemical analysis requirements by DOE and do not get involved in the interpretation of the resulting data. For example, they were not aware of O
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1 the geometrical differences between the MHTGR and test SG tubes (helium crosulow outside the tubes for the MRTGR vs inside for the test). Mr. Veyrat and Mr. Dupont (the COMEDIE loop project manager) have had considerable expedence with FP experiments.
They characterned the loop as a miniature " chemical plant" where impurity concentrations could have significant (and complex) effects on FP transpon behavior. Hence the.
"models" used to design the test conditions will be very important, and certainly crucial to the interpretation of the results. hey also stressed that detailed planning for the tests is
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still in progress, with the first real (with fps) blowdown test scheduled for early 1991.
1 Subsequent tests will look at effects of dust and moisture. They are designing a dust density probe using a laser system to measure opacity.
Kemforschtmosantnee (KFAL Julich f
Dr. Balthesen arranged our meetings at KFA and the subsequent trip to the THTR. He is responsible for managing the HTR development programs in the FRG, and is very L
knowledgeable about most aspects of HTR activities. He works for BMFT, which is the FRG Ministry for Research and Technology._
In the FRG, the state is the licensing entity for all reactors in that state. He mitustry of a state can (and does) get help from independent exprts, such as the TOVs, which have been established in 7 of the 11 states. R)Vs are a so called on for generic licensing studies. For example, TOV-Hannover wrote a safety evaluation repon for the HT 1-MODUL at Interatom's request (and in this case, BMFT supponed its completion when the contract with Interatom was cancelled). He Federal involvement in licensmg is through L~
BMU (formerly BMI), which is the environmental ministry that supervises state authorities. Licensing gmund rules in the FRG are consistent to the extent that licensing is done in steps (as in the U.S.), and while PRAs are considered, they are not used as a basis for licensing. State and federal couns also play a major role in licensing. For example, a blowdown test was planned at AVR, but a citizen's complaint resulted in a judge's decision that blocked the test.
KFA [which just recently had the Kem (nuclear) pan of its name removed] previously devoted 50% ofits effons to gas reactors, while the total now is 10%. KFA is divided up
' into a number of " institutes;" two of the principal ones we interacted with (Nuclear Safety Research and Reactor Development) are about to merge.
In the meeting with the Institute of Nuclear Safety Research, we heard presentations (and received repons) on the current status of KFA's work in accident selection and analysis, source terms, containment, the role of operators in accident mitigation, exgrimental confirmation of heat transfer and fluid flow analyses, and FP transpon ca' culations and expenments.
For the HTR MODUL design, the main reason they have restricted the U 235 loading is to p
mitigate the positive reactivity effect from water ingress from a postulated single-SG-tube-break accident. Water ingress accidents are the major contributors to risk. The calculations for this design basis accident assume a reactor scram, blower trip, and SG isolation and result in acceptable consequences. There are three independent ways to trip on this accident, each with three independent channels: high moisture (800 ppm, capacitance probes), high pressure, and high power (140%). KFA maintains that generauon of a O
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bumable or explosive gas mixture is not likely in the case of a single SG tube break. KFA has also looked at failure of water-side isolation valves and inleakage of steam from other '
modules. For these accidents, the major source for the release is from washoff of fps from the SG tubes, where they assume 2% of the tubes are wetted and all of that 2% gets off. The normal relief path is unfiltered, but credit is taken for the operator switching to the filtered one (after 30 nun). De design includes two helium purification plants and takes credit for the operator's ability to switch to the spare (after 30 min). The loss-of-forced-t convection accident relies on an active (vs passive) cavity moling system, but long term t
outages can be postulated (without damage) to allow for repairs on emergency diesels, reconnection of the outside grid, and/or pump repair.
KFA has done a lot of work on code verification. Its LUNA loop was described as a test facility for HTR thermal fluid flow codes; currently, possibilities of prismatic cort tests are being discussed with DOE.
Most of the KFA fuel testing has used uranium oxide (vs uranium carbide in the U.S.
design), with their lower enrichment (8% vs 20%) and lower design bumup. KFA staff l
said that while the carbide fuel has some advantages for fertile loadings (which they don't i
have) and is less susceptible to the amoeba effect (which they haven't seen at their power densities), the oxide fuel is easier to manufacture to the required quality..In their fuel tests, i
they have seen significant coating failures and deterioration at 1700-1800 C. ney have -
also seen more significant failure rates at bumu?s somewhat higher than their design L
burnup (but less than ours). At 1600 C, the ho' dup of fps in the graphite contributes significantiv to a reduction in the release. The relatively fast diffusion of silver through the coatings at tower temperatures is of no consequence to the overall risk. The KFA lab for e
testing fuelis very impressive, l
KFA staff have also been working on models for effective retention of releases into the reactor building. They noted that the amount of activity attached to dust is hard to calculate, that reasonable models for the dust /FP transport phenomena don't exist, and that J
their planned depressurization and dust release c' periment at the AVR was cancelled (via x
- the citizen lawsuit). They also claim that there are big uncertainties in their FP washoff and
" steam-off' models. They would be interested in collaborating in the DOE COMEDIE loop -
Cxperirnents.
KFA has classified the massive air ingress accident as being of too low a probability to be included in its risk study but is looking at it anyway (Chemobyl!) and have done some very interesting parametric studies. Micmchip manufacturers are assisting with the development of a process for coating the peb ales with a thin layer of silicon carbide. In i
some air furnace tests to date, those without the coating disintegrated while those with the coating looked " undisturbed."
KFA also has a very active pmgram in the metals (including graphite) institute and has an active collaboranon program with DOE /ORNL (Phil Rittenhouse, Ray Kennedy, and Tim l
Burchell). It is notable that the HTR MODUL design uses only Incoloy 800H for their SG tubes to avoid thermal stress problems at the bimetallic welds.
In our talks at the Institute for Reactor Development,it was noted that sources for their physics data for the water ingress accidents included those from Austrian (10 years old) f and Swiss (new) experiments. They quoted a surprisingly low error (5%) estimate for the A
reactivity vs wa:er ingress models. They referred to a "Chemobyl Syndrome" effect that a
O professor in the Green Party has brought up. The curve for reactivity vs water in the HTR-MODUL core peaks at about $4.5 with 1000 kg of water, so he has postulated that if one starts out with more water and quickly empties it out, enough reactivity could be inserted to go prompt critical. Reasonable mechanisms for effecting this have not been postulated. I obtained copies of repons on their accident analysis code *ITNTE (in German).
During our visit to the AVR, most discussions were about AVR operating history and dust data experiments. AVR personnel were surprised at how small the dust panicles were (the peak in the size distribution curve is at <1pm). For more information on the dust expenments, we were referred to Prof. VonderDecken (at KFA). He AVR is currently awaiting its decommissioning license, and no further operation is planned.
Several other miscellaneous topics were discussed:
1.
A lot of concern was expressed in FRG (and France) about how the 1992 Common Market normalization process would be implemented. Most seemed to think that the problems are too far from solution to be solved by then. In FRG, there is concern that they will not be able to compete with cheap Frtnch (nuclear) power, even at home in coal country. Apparently, France's current and near-term planned capacity is great enough to supply a lot of FRG's needs; 2.
It was mentioned that the primary concern at KFA about the U.S. MHTGR-NPR program was that FRG public support for their HTR would be seriously eroded if it were shown that a modular HTR was capable of being used as a weapons producer, and g
3.
The FRG reactor safety committee analogous to the U.S. ACRS completed a study of the HTR-MODUL and concurred with the no-containment building design. Prof.
Nickel, who is on that committee, will forward us the repon when it becomes available (January 1990).
THTR-300 (HKG) ne THTR has becn shut down since November 1988 and is awaiting a decommissioning license. It is a very sad situation, since THTR obviously had a great potential for a long, productive, " safe" life but operated only about 2 years. Its demise was mainly political. It is located in a primarily " Green" coal-country state, where the local authorities felt that they were mislead about the usefulness of THTR as a potential process heat reactor (which could make use oflocal coal). [An FRG study showed that the best (only) potential uses of HTR process heat were for aluminum manufacture and refineries.) Also, the THTR fuel supplier was shut down due to a scandal, the estimated time and money needed to build up an alternative was excessive, and apparently neither the state nor the federal government was willing to guarantee the needed fmancial support.
During THTR operation, some design deficiencies were uncovered, most of which were corrected. A problem in the pebble discharge cin:uit design prevented refueling at powers (flows) higher than 40%. Several other factors contributed to undesirable in-core pebble flows: a miscalculation of the pebble friction at high temperatures, control rods
" shielding" the outer ring of pebbles from flowing to the center, and an improper design of the angled core floor that impeded discharge flow. Also, an arbitrary high-temperature limit set for an euxiliary room limited power output on hot days. After the last shutdown, 8
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7 personnel discovered several (33 of 2600) damaged Incoloy 800 bolt heads on the cover plates in the hot lower alenum ducting. He failures were at the points of highest.
temperature gradient (c ifferences of 100 C were measured acmss the ducts). They claimed that while this was not a safety problem, it needed monitoring, and an on line means of detecting a detached plate was described. (I would guess that repairs would have been -
possible using remote welding techniques.) ^
ne THTR nuclear and thermal pre-operation prMaas were very good. While their initial critical loading predictions were very close, they didn't have any on line reactivity.
calculation, pmbably because of the large uncenainties in fuel distribution (fuel burnup and temperature vs position).
The THTR got good service from its capacitance probe moisture monitors (capacitance),
and staff members said that if the moisture level stayed at or below I ppm there would be no corrosion problems. The primary leak rate was 1/3 of the inventory per year.
The plant had excellent maneuvering capability, and could (by test) sustain a turbine trip without needing to shut down the reactor. He staff conducted a variety of dynamics tests for code validation.
Here are several problems with the THTR decommissioning, including uncenainties in the cott configuration and fuel loading distributions upon emptying the core, he fixed t
detector locations may also make reliable monitoring of enticahty a problem as well. I made several " helpful" suggestions, which I plan to follow up.
O The THTR has a wealth of o wrating data (on plant computer tapes), some of which may u
be ofinterest to the U.S. MF TGR Program. The THTR staff was interested in pursuing the possibility of a subcontract to retrieve and analyze some of the data. In particular, they have data that could be useful in code validation for coolant mixing analyses for the outlet plenum. Steam generator and other component operational data may also be useful.
Summarv of Sienificant Fir' dines and Recommendations 1.
The new U.K. policy of licensing reactors on the basis of PRA calculations has -
gotten them into a bmd with the AGR refueling safety case, perhaps unnecessarily. The 5
United States should be very wary of adopting such a policy.
2.
The U.K. NII is developing rules for diversity of maintenance personnel to avoid common-mode failure problems with maintenance, which is their major safety concem.
l
- 3. He Nil policy statement for accident code validation and verification could be useful to the NRC in drawing up a similar guide for licensees.
L 4.
The U.K. problem with the exclusion of nuclear plants from the privitization sale was due primarily to the uncertainties of the (large) costs for decommissioning the plants. The United States should work toward establishing reasonable policies and cost est mates for L
various decommissioning options. Collaboration with Mr. Giraudel of CEA would be useful (and would be particularly interesting for Fort St. Vrain decommissioning).
5.
We should pursue the subcontract with John Askew to retrieve the U.K. physics data ofinterest to the MHTGR program.
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6.
NRC should follow the design and planning for the COMEDIE loop experiments. It appears to me that additional collaboration with groups experienced in fission product (FP) transpon would be useful to everyone.
7.
FRG is still putting a lot of effort into HTRs in spite of its recent setbacks, most notably the *IMTR shutdown and the lack of a near-term expectation for advanced HTR sales in FRG. The FRG R&D cffort does not seem to be too dependent on U.S. work, but they are very interested in U.S. policies, public acceptance, and licensing criteria.
8.
FRG considers water and steam ingress accidents to be the maj'or contributors to risk in the HTR-MODUL design, and has modified the fuel design (fissile loading) to mitigate the positive reactivity insertion resulting from water / steam ingress. FRG is continuing an aggressive theoretical and experimental program to resolve the remaining problems. We should look for parallels to their "Chernobyl Syndrome" in the U.S. design, i
i 9.
Unlike the completely passive U.S. cavity cooling system design (the air-cooled RCCS), the HTR MODUL design relies on redundant power supplies and pumps, with margin to allow for equipment repair or replacement periods.
-i
- 10. FRG tests on its fuel are of much interest to our fuel performance evaluations, but the significant differences between the U.S. and FRG fuel designs preclude " direct" use of their data. 'Ihe two major items ofinterest were the marked deterioration of the kernel's protective coating with burnup, and the observation that the FRG (oxide) fuel is easier to manufacture to the required quality than is the U.S. (carbide) fuel.
gl
- 11. KFA noted that the current models for FP transport via dust and washoff are not reliable. KFA would be interested in collaborating in the COMEDIE loop tests.
- 12. FRG's development work on a silicon carbide coating for their pebble fuel to mitigate i
oxidation attack should be looked into both for requcing concerns about air ingress accidents for the U.S. design and for possible use as an additiond FP barrier in a fuel stick.
- 13. The FRG safety committee analogous to the ACRS " approved" the no-containment I
building design for the HTR-MODUL. The report on this should be of both technical and political interest to NRC deliberations.
- 14. A wealth of data and experience is tied up in the THTR design and operation that is I
pertinent to MHTGR concerns, and the THTR staff is interested in pursuing collaborative work. This should certainly be pursued.
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APPENDIX 1 h erarv
- November 6-7,1989 Travel from Oak Ridge, Tennessee, to Liverpool, England '
November 8,1989 Visit Heysham 2 Reactor Site, Heysham, England November 9,1989 Meeting with Nuclear Installations Inspectorate, Bootle, o.
England November 10,1989 Meeting with AGR Program Personnel, Harwell, England November 11 12,1989 Travel to Paris, France, for weekend November 13,1989 Meeting with CEA Gas-Cooled Reactor Personnel, Fontenay-aux Roses, France November 14,1989 Meeting with CEN personnel, tour of Siloe reactor and COMEDIE loop, Grenoble, France November 1516,1989 Meetings with HTR Program Personnel at KFA, Julich, Federal Republic of Germany
)
~
November 17,1989 Meeting with THTR-300 Reactor Plant personnel and plant tour, Hamm, Federal Republic of Germany November 18,1989 Travel from Frankfurt, Federal Republic of Germany, to Oak Ridge, Tennessee L
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t APPENDIX 2 Persons Contacted
- Hevsham-II Reactor Neil W. Davies-
'lhermal Reactor Collabomtion Manager, UKAEA-Risley
[
John Birchall Principal Physicist k
Tom White Assistant Operations Manager NuclearInstallation Insocetor Derek Goodison Branch Chief lan Tate.
Siting Criteria Jim Mmray Accident Selection and Containment Malcolm MacPhail Souwe Temis Bill Whiteley Accident Mitigation
~ UKAEA Harwell John R. Askew Director, Gas-Cooled Reactor Programs John Wilson Deputy Director, Gas-Cooled Reactor Programs CEA-Fontenav-aux-Roses
- Daniel Bastien Coordinator for Gas-Cooled Reactors Marc Natta Chief of Service 1^
l Bernard Giraudel Group Leader for GCR Decommissioning Gerard Chevalier Department of Mechanics and Them1odynamics y
1 12
A I, '
., 4 i
CEN-Grenoble Jean Francois Veyrat Chief of Service i
G. Dupont COMEDIEleop Manager Ted Beresovski U.S. DOE Consultant KFA Jillich Erwin Balthesen Director,HTR Development MichaelWill Interatom-GmbH Helmu:Helmers Chief Engineer, Hannover e.V.T0v Wolfgang Kroger Dimetor, Institute for Nuclear Safety Research Wemer Katscher Accident Consequences Dr. Moorman Fission Product Behavior, Source Terms Dr. Wolters Accident Analysis V.
Dr. W. Rehm Thermal Hydraulics Mr. Hennings Reliability Heinz Nabielek Accident Analysis, Fuel Performance Hubertus Nickel Dinctor, Institute for Reactor Materials 1
Mr. Haag Graphite Behavior Dr. Breitbach Materials Testing Bernd Thiele Institute for Reactor Materials
- Dr. E. Teuchert Institute for Reactor Development Dr. W. Scherer HTR Accident Codes Dr. Klaus Kruger Reactor Analysis and Experiments (AVR) l.
Mr. Pott Hot Cells-Fuel Heatup Tests o
i 13
I
' ~.
e THTR-300 Reacu 1
~ Rudiger Buumer-Plant Manager Ivan Kalinowski ChiefPhysicist Norbert R6hl
. Chief of Production i
Erwin Balthesen.
Director, HTR Development (KFA) eL f
i r
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l.
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Y APPENDIX 3 L
i d
ljterarme Acanired i
I-1.
Kershaw, Tony (CEGB) and Durston, John (NNC, United Kingdom).
"Heysham 2 AGR Station Nears Start-up." In: Modern Power Systems: Incorporating Energy Internatsonal; Barnwood, U.K.;
Central Electricity Generating Board; April 1988.
2.
London Health and Safety Commission. " Health and Safety Executive:
The Tolerability of Risk From Nuclear Power Stations." lendon, England; Her Majesty's Stationery Office; 1988.
3.
Health and Safety Executive. "HM Nuclear Installations Inspectorate Safety Assessment Principles for Nuclear Power Reactors." In:
Nuclear Sqfety; landon, England; Her Majesty's Stationery Office; 1982 (Amended January 1989).
4.
Handouts for NRC/ORNL 9/11/89 meeting on NII Siting Policy: _
Gronow, W. S. " Application of Safety and Siting Policy to Nuclear Plants in the United Kingdom"; SM 117-/21. From: Environmental Contamination by Radioactive Materials; IAEA; 1968.
Charlesworth, F. R. and Gronow, W. S. "A Summary of Experience in the l
Practical Application of Siting Policy in the United Kmgdom"; SM-89/41.
From: Containment and Siting ofNuclear Power Plants; 1AEA; 1967.
1 Gronow, W. S. and Gausden, R,"Licensin and Regulatory Control of-l Thermal Power Reactors in the United Kin m."IAEA SM-169/23.
"Parlimentary Debates (HANSARD)." Fifth Series, Volume 758. From:
House of Commons Oficial Report, Session 1967-68; London, England.
Her Majesty's Stationery Office.
Ryder, E. A. (Nuclear Installations). " Proof of Evidence of E. A. Ryder to the Hinkley Point 'C' Public Inquiry." HSE (NII)2. Health & Safety Executive.
-5.
"The Inspectorate's Approach to the Assessment of Code Validation Submissions." Sizwell-B Inquiry. NII; March 1983. (8 pages) 6.
" Final Report of Boiler Tube 1.cak on Reactor 1 on 17. March 1987."
Hartlepool Station.14 April 1987. (5 pages) 7.
" Compilation of AGR Trips."(6 pages)
O 15
u 1
8.
Spanton, J. H., Brighton, P. W. M., McLaughlin, D. Stevenson, and G.
J
" Inherent Safety in Small and Medium Sized Power Reactors:
Review of Some Currently Proposed Small and Medium-Sized Reactor Designs." UKAEA Hcalth and Safety Studies Committee;
. May 1988.
9.
Hunt, Colin. "PRA: Modular HTR Systems." Reactor Physics Division; AEE Winfrith. GA-MHTGR 350.
L a
10.
Stevenson, G. W. (UKAEA Safety & Reliability Directorme), Mannell,
. S. J. (UKAEA Safety & Reliability Directorate), Hunt, C.
(Winfrith Atomic Energy Establishment), Brighton, P. W. M.
(UKAEA Safety & Rehability Directorate). " Ultimate Safety in the Interatom HTR-Modul." September 1988. (105 pages) 11.
Hunt, Colin (Winfrith Atomic Energy Establishment). " Review of the Safety Case for the Interatom HTR-Module." Reactor Physics Division; AEE Winfrith; August 1988 (3 pages) t 1
12.
MacPhail, Malcolm. "UK Regulatory Position on Source Tenns." NII(UK).
9 November 1989. (11 pages) 13.
"The Elements of Success: United Kingdom Atomic Energy Authority Annual Report and Accounts 1988-89." AEA Technology; 1989.
(-
(88 pages) g
'14.
Wilson, John (UKAEA). " Advanced Gas Cooled Reactor Research and Development." In: Atom; Number 386; The United Kingdom Atomic j.
Energy Authority; December 1988.
l 15.
Allen, F. R. (AEA Technology). " Safety Aspects of Advanced LWR Designs."
j L
(7 pages) 16.
Bastien, Daniel (CEA). " Twenty-nine Years of French Experience in l
Operating Gas-Cooled Reactors." For: Technical Committee Meeting in Design Requirements, Operation and Maintenance of Gas-Cooled 4
Reactors; 21-23 September,1988; San Diego, Calif. CEC and l
OECD/NEA; September 1988, 17.
Mehner, A. W. (NUKEM GmbH), Abassin, J. J. (Hanau FRG; CEA), Sciers, P. (CEN Grenoble, France). " Characterization of HTR Fuel Element j
Samples by Activation Testing." (10 pages) 18.
" Helium Loop 'Comedid." In: Commissariat i L'dnergie Atomique.
Centre D'dtudes Nucidaires; Grenoble.
19.
Rehm, W. " Presentation of Thermodynamic Investigations into the Safety Research and Development Work for the HTR." Institute of Nuclear Safety Research. For: HTR-Seminar Between FRG and USSR'in Julich, FRG; October 23-27,1989.
O 16 L
.s
- g., q
- s l
i U,~
20.
Nabielek, Heinz (KFA Julich). " German Research and Development for
- HTR Fuel." November 1989.
21.
Rehm, W. Onstitute of Nuclear Safety Research), Jahn, W. (Institute of Nuclear Safety Research), Barthels, H. (Institute of Reactor Components). " Comparison of'Iheoretical and Experiment Studies of Afterheat Removal by Natural Convection /Cuculation from an
- HTR." For: Fourth Imernational Meeting on Nuclear Reactor Thermal-Hydraulics; October 1013,1989; Karlsruhe, FRG.
22.
Gerwin, Helmut (KFA Julich GmbH). "'Ihe Two-dimensional Reactor Dynamics Program TINTE - Jul 2167, Nov. 87." Volumes 1 and 2.
23.
Vollmer, Heinz, Theymann, Walter, Ivens, Gunther. " Experience Gained with the AVR Experimental Nuclear Power Station." From: Brown Boveri Review, Volume 74, No.1/1987. Brown, Boveri & Cie (BBC) and Hochtemperatur-Reaktorbau GmbH (HRB), Federal Republic of Germany,1987. Publication No. D KW I167 87 E.
(10 pages) 24.
Scherer, W. and Gerwin, H. Onstitut fur Reaktorentwicklung),
Kindt, T. (Interatom GmbH, Friedrich Ebert Strabe), and Patscher, W. (Hochtemperaturreaktorbau GmbH). " Analysis of Reactivity and Temperature Transient Experiments at the AVR
. High-Temperature Reactor." In: Nuc. Sci. Eng. 97 58 63 (1987).-
25.
BHumer, R. and Kalinowski, I. (Hochtemperatur-Kernkraftwerk GmbH).
"THTR Commissioning and Ope, rating Experience." (25 pages) 26.
Wohler, J. (Hochtemperatur-Kernkraftwerk GmbH), BHumer, R. and Schwarz, D. (Vereinigte Elektrizitdtswerke Westfalen AG). 'The Start Up and Initial Operation of THTR 300: A Milestone ont he Way Toward an Energy System Based on Coal and Nuclear Energy."
(34 pages)
L i
27.
Bhumer, R. " Selected Subjects on the Operation of the THTR 300." In:
VGB Kraftwerkstechnik 69, Number 2, February 1989, pp.141-147.
28.
Scherer, W., Druke, V., and Gerwin, H. (Institut fur Reaktorentwicklung),
L and Presser, W. (Hochtem peraturreaktorbau GmbH). " Adaption of the Inverse Kinetic Method to Reactivity Measurements in the Thorium High-Temperature Reactor-300 " In: Nuc. Sci. Eng. 97, pp.96-103 (1987).
LO l-17
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o 1j DISTRIBUTION 1.
David B. Waller, Assistant Secretary for Intemational Affairs and Energy _
Emergencies (IE-1), DOE Washington, DC 20545 '
2.
Director, Division of Intemational Programs, Office of Nuclear Energy Programs, DOE Washington, DC 20545 3.
Director, Division of Intemational Security Affairs, DOE Washington, DC 20545 4
Eric S. Beckjord, Director, Office of Nuclear Regulatory Research, U.S.
Nuclear Regulatory Commission, Washington, DC 20555 '
i 5.
A. Bryan Siebert, Director, Office of Classification and Technology Policy (DP-323.2), DOE, Washington,DC 20545 6.-
Bill M. Morris, Director, Division of Regulatory Applications, RES, U.S.
NuclearRegulatory Commission, Washington,DC 20555 7.
R. L. Egli, Assistant Manager, Energy Research and Development, DOFJORO, Oak Ridge,TN 37831 8.
Thomas L. King, Chief, Advanced Reactors and Generic Issues Branch, DRA/RES, U.S. Nuclear Regulatory Commission, Washington, DC 20555 1
9.
Peter M. Williams, MH'It3R Project Manager, DRA/RES, U.S. Nuclear Regulatory Commission, Washington,DC 20555 o
10.
D. J. Cook, Director, Safeguards and Security Division, DOE /ORO, Oak Ridge, TN 37831.
I 11.
Jerry N. Wilson, DRA/RES, U.S. Nuclear Regulatory Commission, q
Washington,DC 20555 V
12.
R.Korynta, DOE /ORO, Oak Ridge,TN 37831 13-14. ' Office of Scientific and Technical Information, P.O. Box 62, Oak Ridge, TN 37831 s
15 18.
A. W. Trivelpiece 19.
S. J. Ball 20.
R. S. Booth 21.
E. R. Bowers 22.
C. R. Brittain 23.
N. E. Clapp 5
i
- 24. - J. C. Cleveland 25.
J. C. Conklin 26.
B. G. Eads 27.
D. N. Fry 28.
F. J. Homan 29.
H. Jones 30.
J. E. Jones, Jr.
31.
R. A. Kisner 32.
H. E. Knee 33.
T. S. Kress 34.
W. C. Kuykendall 35.
D. L. Moses 36.
J. A. Mullens 37.
J. K. Munro 38.
L. C. Oakes 39.
P. J. Otaduy p ]
l /
40.
C. E. Pugh 41.
J. P. Sanders 19
/
..... ]
3
~
O (42.
R. L. Shepard 43.
L. H. Thacker 44.
R. E. Uhrig
- 45. - J. D. White 46.
R. P. Wichner 47.
R. T. Wood 48.
A.Zucker 49.
1&C Division Publications Office 50.
Laboratory Protection Division 51'-52.
Labomtory Records Department 53.
Laboratory Records Depanment RC 54.
ORNL Patent Section
- 55.. ORNL Public Relations Office O
1 i
l O'
~
=
j p
o LV-Referenced Documants s
1.
" Final Report of Boiler. Tube Leak on Reactor 1 on 17 March 1987."
Hartlepool Station.
14 April 1987.
2.
London Health and Safety Commission.
" Health and Safety Executive:.
The Tolerability of Risk From Nuclear Power Stations."
London, England; Her Majesty's Stationery' Office; 1988.
3.
"The Inspectorate's Approach to the Assessment of Code Validation Submissions." Sizwell-B Inquiry. NII; March 1983.
4.
Wilson, John (UKAEA).
" Advanced Gas Cooled Reactor Research and Development." Atom; Number 386; The United Kingdom Atomic Energy Authority; December 1988.
5.
Bastien, Daniel (CEA).
" Twenty-nine Years of French Experience in Operating Gas-Cooled Reactors." For:
Technical Committee Meeting in Design Requirements, Operation and Maintenance of Gas-Cooled Reactars; 21-23 September, 1988; San Diego, Calif. CEC and OECD/NEA; September h-
- 1988, O
6.
" Helium Loop 'Comedie'." Commissariat a L'energie Atomique.
Centre D' etudes Nucleaires; Grenoble.
7.
Balthesen, Erwin.
"Next Generation of Nuclear Power and the Role of the
'l
'IAEA'," Consultants working group, IAEA, headquarters, Vienna. 30 October -
2 November 1989.
-8.
KFA-Julich, "Probabilistic Safety Analysis of the HTR-Modul," information distributed at meeting on November 15, 1989.
9.
Rehm..W. " Presentation of Thermodynamic Investigations into the Safety Research and Development Work for the HTR."
Institute of Nuclear Safety Research.
HTR-Seminar Between FRG and USSR in Julich, FRG; October 23-27, 1989.
10.
Rehm, W. (Institute of Nuclear Safety Research), Jahn, W. (Institute of Nuclear' Safety Research), Barthels, H. (Institute of Reactor Components).
" Comparison of Theoretical and Experiment Studies of Afterheat Removal by Natural Convection / Circulation from an HTR."
For:
Fourth International Meeting on Nuclear Reactor Thermal-Hydraulics; i
October 10-13, 1989; Karlsruhe, FRG.
11.
Nabielek, Heinz (KFA-Julich).
" German Research and Development for HTR Fuel."
November 1989.
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v x
4
(.
-12.
Schenk, Werner, Pitzer, Dieter and Nabielek, " Fission Product Release i
Profiles from Spherical HTR Fuel Elements at Accident Temperatures,"
.h:
Jul-KFA-Julich, September 3988.
13.
Verfondern, K., Schenk, W. and Nabielek, H. " Passive Safety Charateristics of Fuel for Modular HTR and Fuel Performance Modeling Under-Accident Conditions, KFA-Julich Internal. Report KFA-ISF-1B-9/89, October.1989.
14.
Steitbach, G., Schubert, F. and Nickel, " Proceedings of the Workshop on
' Structural Design Criteria for HTR," KFA-Julich, ISSN 0344-5798, April 1989.
15.
I-A. Weisbrodt, " Engineering and t.icensing Progress of the HTR-Module,"
Siemans AG, UB KWU Interatom GmbH, GCRA 10th International Conference, San Diego, Calif., 18 September 1988.
- 16.. Vollmer, Heinz, Theynann, Walter, Ivens, Gunther.
" Experience Gained with the AVR Experimental Nuclear Power Station." Brown Boveri Review, Volume 74, No. 1/1987.
Brown, Boveri & Cie (BBC) and Hochtemperatur-Rea ctorbau GmbH (HRB), Federal Republic of Germany,1987.
Publication t
No. DKW 116787E.
17.
Hochtemperatur-Kernkraftwerk GmbH, "THTR 300 MW Nuclear Power Plant, Hamm, FRG.
18.
Wohler, J. - (Hochtemperatur-Kernkraftwerk GmbH), B8umer R. and Schwarz, D. (Vereinigte Elektrizitatswerke Westfalen AG).
"The Start Up and Wq
~
Initial Operation of THTR 300:
A Milestone on the Way Toward an Energy System Based on Coal and Nuclear Energy."
19.
B8umer, R. and Kalinowski, I. (Hochtemperatur-Kernkraftwerk GmbH).
"THTR Commissioning and Operating Experience."-
O 2
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