ML19282A078
| ML19282A078 | |
| Person / Time | |
|---|---|
| Issue date: | 04/17/1979 |
| From: | Meyer J, Williams P Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7905100277 | |
| Download: ML19282A078 (14) | |
Text
!
I MEETING
SUMMARY
DISTRIBUTION Project Fi 616)
G. Kuzmycz NRC PDR J. Carter ARB Reading File H. Holz TIC J. Long NRR Reading File L. Lois E. G. Case IE (3)
R. S. Boyd H. Gearin R. C. DeYoung ACRS (16)
D. J. Skovholt J. Knight D. B. Vassallo D. Ross R. Denise R. Tedesco K. Kniel S. Hanauer T. P. Speis S. Pawlicki C. J. Heltemes F. Schauer R. W. Houston T. Novak L. Crocker Z. Rosztoczy R. Mattson V. Benaraya R. Clark W. Gammill P. M. Williams M. Tokar J. " eyer F. Rosa D. Bunch W. Haass R. Bosnak Applicant & Service List J. T. Larkins R. Hartfield R. D. Schamberger R. Foulds H. Cobb E. Delaney C. P. Tan C. Kelber F. B. Litton R. E. Ireland H. Berkow 4
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Project No. 616 (riASAP Review)
Applicant: Department of Energy REPORT OF MEETINGS ON GCFR AND HTGR
SUMMARY
On February 26 and 27 we met in San Diego with representatives of the General Atomic Company on the subjects of our NASAP reviews of the GCFR and HTGR. Both meetings were also attended by representatives of several other organizations which are identified in the Enclosure.
Copies of the viewgraphs presented at the meeting may be seen in offices of the authors and the NRC public document room.
The topics presented for the GCFR were (1) program description, (2) general description of the 300MW(e) plant, (3) licensing overview, (4) research and development programs, (5) environmental aspects, (6) fuel cycle physics considerations for a 1200 MW(e) plant, (7) alternate fuel cladding materials and (8) accomodations for core melt. Agenda items deferred because of lack of time were (1) alternate shutdown systems, (2) potential effects of up-flow design on accident analysis conclusions and (3) in-service inspection criteria and requirements.
Principal general observations from the GCFR meeting are:
1.
Substantial amounts of new safety design and research information since our August 1974 SER (Project No. 456) was issued is now available for the 300MW(e) demonstration plant.
?.
The design for a 1200MW(e) reference plant for the NASAP study can be only outlined at this time.
3.
If the "up-flow core cooling" design option is submitted as an amendment to the PSEID, conclusions and findings pertaining to emergency core cooling provisions, core design, and certain accident scenarios of the 1974 SER will have to be re-examined.
Topics presented for the HTGR were (1) design and program overview, (2) program and description of the HTGR-GT, (gas turbine), (3) licensing and environmental review, (4) fuel cycle for MEU fuel, (4) research and development programs, (5) primar y system integrity, and (6) fuel transient response.
Principal observations are:
1.
The HTGR-GT is in the early conceptual design stage, although many features are generic to both the gas turbine and steem cycle HTGRs.
2.
Studies of safety problems and accidents unique to the HTGR-GT have only been recently initiated.
Internal pressure equilibrium accidents can be more severe on primary system components than the design basis depressurization analyzed for earlier HTGRs.
7905100 D7 i
. 3.
General Atomic is (oopera t tnq Internallonally with West Germany in the design of the HTGR-GT, and domestically with IJnited Technologies Corporation (turbine - compressor unit) and Combustion Engineering (recouperator).
For both the GCFR and HTGR, the meetings provided opportunity for GA Comments and to amplify information presented in the PSEID reports.
questions are being prepared for transmittal to DOE in the near future based on this amplified information.
I.
GCFR PRESENTATIONS A.
Program Descriptic'1 - R. H. Simon Comparisons were made between the GCFR and Clinch River. A Bechtel study was cited which showed a capital cost advantage for the GCFR over '.MFBRs in general.
Development costs thus far of $140M were reported. Costs of $400M will be required for further development of components and fuel.
The R&D benefits from the LMFBR program and the HTGR program were estimated at 5500 M and $250M, respectively.
The GCFR research program has involved international cooperation. More recently, the international program has addressed design aspects with the Swiss and Germans favoring up-flow core cooling. Up-flow cooling has been under investigation by GA for about a year with a decision as to how to procede coming in April 1979. Fort St. Vrain experience has given GA confidence that substantial development problems will not exist in helium technology.
Funding by DOE in 1980 will be increased to $26M, about double the 1978 funding level.
The need for a CP in about 5 years is foreseen for the 300MW(e) development plant. The development plant would be followed by a 1200MW(e) demonstration plant, prier to building a first concercial unit.
Currently under investigation is a two-step plan to commercializa-tion which would combine the functions of the development and demon-stration plants.
B.
General Descript ion of the 300MW(e) Plant - L. J. Kabe Revisions to the 19/4 de'. lyn were presented.
These included (1) once through steam generator'.. (?) electric drive circulators with motors outside the l'CRV (3) radial rather than axial compressors and (4) housing for control rod drives in the upper PCRV plug (i.e., like fort St. Vrain).
The use of the once through steam generator will simplify design and control problems. Electric drives for the circulators will permit full scale pre-operational testing.
Water bearings will be used for shaf t seals that will see the full pressure of the PCRV. The orives will have pony motors qualified f or all accidents except rapid depressurization.
. A core auxiliary cooling system will provide for the rapid depressurization accident. Both the main and auxiliary cooling systems will be seismically qualified. The steam generator will be made from Crome-Moly steel and have no bi-metallic welds.
The GA staff saw no major problems in extrapolating these design features to a 1200MW(e) plant.
If the up-flow design is adopted there will be only minor changes in the steam generator design but the circulators will Le located at the bottom of the PCRV. Emergency core coolinq by natural convection is implied except for depressurization events. The up-f ow design also offers advantages of core support.
Descriptions of the control rod and refueling designs were not presented for the upflow core.
C.
Licensing Overview - C. R. Fisher General Atomic closely follcwed the Clinch River licensing review and has used this infonnation to anticipate the NRC intent with respect to the GCFR. As examples, GAC is now considering core disruptive accidents and residual heat removal, and has patterned its reactor siting source term and its containment configuration after Clinch River.
Fisher reviewed the licensing history of the GCFR and noted that since completton of the NRC's SER in August 1974, amendments to the Preliminary Safety Infonnation Document (PSID) were submitted on core An ACRS cooling in 1976 and on general design criteria in 1977.
information meeting was held in July 1977.
Although not part of the GCFR NASAP PSEID submittal, it is worth noting that the preapplication review plan of Helium Breeder Associates calls for amendments to their PSID addressing (1) general design criteria (July 1979), (2) core safety limits (December 1979), (3) core cooling design bases (April 1980) (4) core assembly development plan (December 1980) and (5) bases for accommodating core disassembly accidents (April 1981).
The anticipated licensing issues were identified by GAC under the (1) thermal margins, (2.) depressurization accidents, following headings:
(3) core-disruptive accidents, (4) shutdown system diversity, (5) core cooling systems (6) containment systems, (7) core design, (8) PCRV, and (9) in-service inspection.
., D.
R&D Program - R. A. Moore The GCFR research programs are organized under topics of (1) materials development and irradiations, (2) fuel assembly development, (3) reactor physics and shielding, (4) safety, and (5) components.
Funding for GCFR R&D will total $12M in 1979, with a substantial increase being planned for fiscal 1980. The major portion of the funds are expended at GA, ANL and ORNL together with other programs at LASL, PNL, INEL, and BNL.
Highlights of the R&D program of major interest to safety and licensing are:
1.
The Core Flow Test Loop (CFTL) at ORNL will be operational in 1961.
This loop will p6rmit testing of a bundle of 91 electrically heated, full length rods under accident and transient helium flow conditions.
2.
The first irradiations at the BR-2 loop in Mol, Belgium have been completed successfully - see Paragraph G below.
3.
A core mock-up critical is planned by ANL for 1983.
Earlier ZPR critical experiment work at ANL has been completed and reperted.
4.
An experiment at the Tower Shielding Facility at ORNL to measure doses in the grid plate has been successfully completed and will be reported shortly.
5.
A substantial number of fuel rod irradiations have been completed and others are underway or planned.
Programs include single rods, rod bundles, vented rods and fast flux irradiations. Data are being obtained from both international and domestic programs.
Planned facilities included an inpile loop at the FFTF.
6.
Prototype core assembly tests involving test loop operations are being considered for the Carmen 2 loop in Saclay, France, for a modified EBOR facility in Idaho and a new facility by KFA, (Julich,FRG).
7.
Physics and shielding studies for 1979 are being carried out at several locations. General Atomic: (1) critical assembly analysis and design, (2) recriticality analysis, (3) methods development, (4) shielding experiment analyses and (5) alternate fuel cycle studies.
ORNL: (1)GCFR alternate fuel cycle studies, (2) heterogeneous designs. and (3) GCFR shield experiments in Tower Shielding Facility. ANL: critical assembly analyses and planning.
EIR (Switzerland): steam entry benchmark and thorium blanket experiment.
. 8.
The safety studies for 1979 are being carried out at the following locations. General Atomic: (1) probabilistic methods development, (2) primary containment technology, (3) core accident technology, (4) reliability analysis, and (5) experimental support. ANL: (1) core accident technology and (2) direct electric heating experiments (DEH).
ANL/INEL: GRIST II.
LASL: continuation of the duct melting and fallaway tests.
International programs will study core retention technology, natural convection and review major safety issues for up-flow /down flow assessment.
9.
The GRIST II test will be an in-pile loop experiment in the TREAT reactor where fuel response will be observed in conditions simu-lating loss of flow without scram and reactivity insertion without scram.
Operation is planned for late 1984.
10.
The duct melting and fallaway test series at LASL uses electrically heated fuel rods. The first qualification test, a 37 rod bundle, evidenced localized duct melting which indicated the effects of natural convection. Additional tests are planned with a full size test in late 1979.
11.
Most of the component and systems development is at General Atomic.
This includes (1) circulator test facility design, (2) circulator service system design, (3) model testing, (4) bearing and seal test, (5) steam generator development, and (6) development planning.
ORNL is conducting PCRV and closure tests.
E.
Environmental Aspects of the GCFR - C. A. Perry Thermal wastes would be less than for LWRs while chemical and biocide wastes would be about the same.
Handling of gasecus radioactive wastes differs because of the vented fuel concept.
Noble gases could be retained at the plant site or released in accordance with Appendix I.
Liquid and solid radioactive wastes are judged less than for LWRs. General Atomic's overall conclusion is that lower releases to the public of all types of wastes is characteristic of the GCFR.
F.
Fuel Cycle Physics for 1200MW(e)/NASAP Design - C. J. Hamilton General Atomic examined seven fuel cycle cases in terms of breeding ratio and specific power and selected the following three cases for further core design studies: (1) plutonium core with uranium blankets, (2) plutonium core with thorium blankets and (3) low enriched U-233 with thorium blankets. The corresponding breeding ratios for these cases were 1.41, 1.27 and 1.07.
Various roles of the GCFR breeder
. in relationship with LWRs and advanced converter reactors were examined in terms of resource utilization and long-term reactor It was stated that (1) the long-tenn strategy to the year 2080.
energy potential of any fuel cycle which does not employ plutonium is extremely limited, (2) there is no assured fuel-supply scenario which eliminates the use of weapons grade material, and (3) some form of secured energy center will be required to reduce Drolifer-ation risk while assuring a reasonable supply of fissionable material.
G.
GCFR Fuel Bundle Test - R._J_.
Campana_
A bundle of 12 part length GCFR vented fuel rods were tested in the BR-2 reactor at Mol, Belgium. The inpile facility, identified as HELM (Helium Loop Mol), will continue to be utilized for two more The test bundle tests in addition to the test recently completed.
facility closely matches the GCFR reference fuel design with the principle exceptions of rod length (600mm vs 1130) and rod number Experiments confinned the vented rod design per bundle (12 vs 265).
and the heat transfer calculations including the effects of surface There was a small amount of fission gas in the coolant roughening.
Peak linear heat rates of up stream coming from an unknown source.
to 50 kw/m were experienced versus the GCFR reference value of 33 kmw/m.
H.
Alternate Fuel Cladding Materials - S. Lang Langer stated that 316 stainless steel cladding does not permit economic fuel burnups in a fast reactor due to swelling. A program Six is now underway to investigate alternate cladding materials.
advanced metals nave been selected for further study which include alloys of up to 45 percent nickel.
Ferrit.ic alloys are receiving greatest emphasis because high percentage, of nickel effect neutron Some alloys are now being studied which are compatible econorqy.
with a helium environment but not sodium. The objectives of the advanced alloy program is to increase core life time, decrease core metal fraction and increase the allowabie hot spot clad tempera-ture. Research activities are aimed at solving problems of fabri-cation and establishing performance by irradiation testing, and transient tests on cladding and fuel rods. This later activity might use the GRIST II loop.
I.
Accommodations for Core Melt - A. Torri General Atomic is assessing the technical feasibility of molten fuel containment within the PCRV and is developing detailed time dependent models to predict the behavior of fuel in a core catcher. To date feasibility in principle has been established and data needs have
. been identified. Several core retention system design concepts have been identified (including a high temperature crucible, a heavy metal bath, a borax bath, and a stainless steel bath). A preliminary 2-D analysis model is complete. Future work will assess the core retention concepts including an assessment for a large pl ant. Analyses will be made of molten fuel penetration into concrete and thermal behavior in the refueling region.
II. HTGR PRESENTATIONS A.
Design and Program Overview - A. J. Neylan The evolution of the HTGR design, the program status and the parti-cipation of national and international organizations was described.
The progression of the basic design of the HTGR from the 1975 Summit, Fulton, and GASSAR designs to the lead plant concept pre-sented in the NASAP PSEID for the HTGR of February 1979 was des-cribed. The changes given in the PSEID are (1) reduction of power to 7 w/cm3, (2) use of small, gray control 3
density from 8.4 w/cm rods to reduce temperature fluctuations during load changes, (3) increased core cavity height to reduce hot streaking problems, (4) use of a radial flow reheater and a modified upper steam generator closure, (5) uprating of core auxiliary cooling system together with design modification of the auxiliary heat exchangers, (6) change of the PCRV support to a ring design permitting simplified pipe runs in this vicinity, (7) use of a single-turbine electric generator, (8) use of MEU fuel, and (9) asymetric location of separated safety and non-safety related components in the PCRV.
With a lowered core power density the conversion ratio is increased because thorium can occupy space previously needed for cooling channels. Other potential advantages were cited as lower fuel cycle costs, lower U 03 8 requirements over the plant life, more con-servative fuel and core design, less primary circuit activity, lower core pressure drop with a corresponding increase in plant efficiency, and reduced emergency core tooling requirements. Disadvantages are higher initial core costs, increase plant costs (because of a larger PCRV) and increased refueling time.
Not described in the PSEID ist noted by Neylan were the character-istics of the 900 MW(e) HTGR-SC (steam cycle) considered by Gas Cooled Reactor Associates (GCRA).
In this plant electric drives for me circulators are used and reheat is provided by steam external to eqe :CRV. This concept is a " base case" with the major problems regarding design, safety and licensing well known.
. B.
Overview of Gas Turbine Program - C. O. Peinado The choice of the gas turbine program over the steam cycle program was recently reaffirmed. Spending in fiscal 1979 will be about
$33.5M with $3.8M specific to the direct cycle. The remainder is generic to the HTGR and includes both component development and fuel technology. The DOE program will emphasize support of the HTR gas turbine program in Germany. Currently, GA is developing a refer-ence design with alternatives (unspecified at present). Combustion Engineering is designing the recuperators and United Technology Corporation is designing the turbine-compressor. GA is working closely with the Germans on safety design criteria.
C.
Design Features of the Closed Cycle Gas Turbine Plant - C. R. Boland The main features of the closed-cycle gas turbine plant are:
(1) plant life of 40 years, (2) 3000 MW(e) core thermal rating, (3) 40 percent efficiency with dry cooling or 48 percent efficiency with an ammonia ottoming cycle, (4) design parameters to be optimized for minimum overall cost of power, (5) integrated configuration (i.e.(15626F) all primary system components are within the PCRV, (6) 850 C reactor outlet temperature, (7) multiple gas turbine loops, (8) non-intercooled cycle (German design provides for intercooling), and (9) a high degree of heat recuperation. The design philosophy includes making maximum use of existing HTGR and gas turbine tech-nology with state of the art technology for materials and fabrication methods. System simplification is a goal even at the expense of slight penalties in efficiency. Conservative design parameters will be used with stress levels in major primary system components commen-surate with a full plant operating life of 280,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. Three means of part load control considered are: helium inventory, core bypass, and atemperation.
Inventory control results in the highest part load efficiency but is the most complex.
The path between the core exit plenum and the inlet to the gas turbine is known as the hot duct and is subje'ct to special design considerations. Means will be provided to remove, inspect and/or replace this duct. An onsite maintenance facility will provide for turbine maintenance at six year intervals.
Unique safety and licensing aspects for the HTGR-GT were stated as:
(1) shaft seal failure, (2) internal pressure equilibrium accidents and (3) turbomachinery failures. A safety evaluation of the design has not yet been performed by GA.
_9_
D.
HTGR Licensing - C. R. Fisher Highlights of the Summit, Fulton and GASSAR reviews were cited.
These activities covered the period from April 30, 1973 when the PSAR for the Summit plant was submitted through the meeting on July 11,1977 with the ACRS (Subcommittee) on the interim SER (draft) on GASSAR.
GA has participated with GCRA in three meetings held in 1978 where reviews of several HTGR generic topics were pro-posed under the NRC's topical report program.
Environmental impacts for the HTGR-SC are described in the PSEID.
For the HTGR-GT the radiological and chemical effects from nornal operation will be similar to those of the HTGR-SC. The HTGR-GT will have a minimal effect on water consumption due to the utili-zation of dry cooling.
It was noted that Fort St. Vrain experience has been encouraging with respect to occupational exposure, since radiation levels have been lower than expected. Fisher also noted that the HTGR offered a potential for small land use due to the time dependent fission product release characteristic of the fuel.
E.
Core Design - A. M. Baxter Baxter stated that the impact of MEU fuel on core design will be mininal with effects similar for both the steam cycle and gas turbine designs. The decrease in fissile concentration results in a design with a smaller fuel rod diameter, thinner particle coatings, lower fissions per initial heavy metal atom and larger fuel kernels.
A significant increase in the negative temperature coefficient occurs because of a shift of the thermal flux spectra toward the Pu-240 resonance at lev. There is no significant change in the Doppler coefficient. An anti core fluctuation program is underway which includes (1) review of available Fort St. Vrain data, (2) close nonitorinq of the Fort St. Vrain program, (3) comparison of Fort St. Vrain and large HTGR designs, (4) anair ic studies with core flow models, and (5) an experimental prcr
,f
.c ening tests and flow tests.
F.
Fuel Cycles - R. F. Turner Nine HTGR fuel cycle cases were investigsted wF'th considered com-binations of the following:
(1) MEU, LEU, and TU fuals, (2) U-235, U-233 and Pu fissile makeup, (3) once through anu recycle refueling nodes, and (4) annual and semi-annual refueling frequ'ncy.
In all cases, thorium was the fertile material. Fo t'. cases were examined
. once through) to 0.92 (HEU/Th-recycle)ging from 0.54 (MTU/Th -
further with the conversion ratios ran The current version of the PSEID considers only MEU fuel. A supplement is being prepared that will consider the HEU cycle.
The generic safeguards advantages of HTGR fuel and fuel cycle were cited as follows:
(1) bulky, heavy elements, (2) delute fissile fuel, (3) particle type fuel facilities disposal of waste and plutonium, and complicates recovery of fissile material if diverted, (4) diversion of U-233 is detered by associated radioactivity, (5) low plutonium production, and (6) attractive for export.
G.
R&D Programs - B. E. Olsen, A. J. Neylan Research programs pertaining to our nine categories of HTGR safety issues were described with the exception of inservice inspection.
Highlights were:
1.
A topical report on core seismic response will be ready in September 1979.
2.
The ASME Section 3, Division 2 code connittee (graphite design criteria) has been reactivated. An ASME code section on structural graphite is planned for September 1980.
3.
The Phase II report of the ongoing AIPA study was cited as the current reference document where GA addresses the subject of low probability accidents.
4.
Tests related to the replaceable liner concept have been terminated.
5.
Some flow tests and thermal barrier tests critical to HTGR safety are carried in France under agreements with CEA dating back to 1973. GAC states that CEA is following NRC approved QA.
6.
The ORNL model test on the asymetric PCRV will be changed to support the PCRV design for the HTGR-GT.
7.
Hot streak tests are planned to be applicable to both the gas turbine and the fort St. Vrain upper plenum concern.
-l'-
8.
Fuel inspection tests and procedures will be developed to assure that production fuel is made to reference fuel specifications.
54{ ilm
. m Peter M. W iams Advanced Reactors Branch Division of Project Management Office of Nu ar Reactor Regulation Jam.s F. M yer Ad anced Reactor Branch Division of Project Management Office of Nuclear Reactor Regulation
g t/\\/.
GCFR PRESENTATION NASAP NRC hEVIEW OF PSEID February 26, 1979 Attendees D. G. Swanson Aerospace Corp.
R. O. Brittan ANL W. C. Lipinski ANL
- 0. Duttemer HBA A. P. Kelley HBA' J. K. Anderson HEDL Safety Assessment Office P. H. Willians NRC J. F. Heyer NRC D. T. Bradshaw TVA/0RNL N. C. Ostrander UCLA D. E. Doyack GA R. J. Campana GA C. R. Fisher GA R. J. Grenda GA C. J. Hamilton GA L. J. Kube GA S. Langer GA R. A. Moore GA C. A. Perry GA R. H. Simon GA H. J. Snyder GA A. Torri GA A. R. Veca GA 4
HTCR PRESENTATION NASAP-NRC REVIEW OF PSEID February 27, 1979 Attenciegi R. A. Mey..r At:ro: pace Corp.
D. G. Swcadoo A.rospace Corp.
R. O. Brittan ANL W. C. Lipinski ANL T. E. Collins DOE D. E. Davis GCRA J. K. Ande:rson HEDL P. M. Lvillmm s r >Rt J. F. licy:>r NRC
- 11. Tokar NRC D. T. Bradshaw TVA/0RNL K. Solocon Rand /UCLA W. Kastenberg UCLA A. M. Baxter GA M. Blumeyer GA C. R. Soland GA C. R. Fisher GA A. J. Neylan GA B. E. Olsen GA C. O. Peinado GA J. Read GA R. F. Turner GA i
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