ML20054C656
| ML20054C656 | |
| Person / Time | |
|---|---|
| Site: | 05000142 |
| Issue date: | 04/08/1982 |
| From: | Hawley S Battelle Memorial Institute, PACIFIC NORTHWEST NATION, NRC COMMISSION (OCM) |
| To: | |
| Shared Package | |
| ML20054C649 | List: |
| References | |
| NUDOCS 8204210469 | |
| Download: ML20054C656 (10) | |
Text
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
Docket No. 50-142 THE REGENTS OF THE
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UNIVERSITY OF CALIFORNIA
) (Proposed Renewal of Facility License)
)
(UCLA Research Reactor)
)
AFFIDAVIT OF SEAN C. HAWLEY 1, Sean C. Hawley, do hereby depose and state:
1)
I am a research scientist employed by Battelle, Pacific Northwest Laboratory in the Health Physics Technology Section of the Radiological Sciences Department. A statement of my professional qualifications is attached to this affidavit.
2)
I have read " Interrogatories To S. C. Hawley, R. L. Kathren and M. A. Robkin As To ' Analysis of Credible Accidents for Argonaut Reactors' NUREG/CR-2079 PNL-3691".
The responses that follow explain or interpret the research or results from NUREG/CR-2079, of which I was a principal author and which is based on research at the Pacific Northwest Laboratory operated by Battelle Memorial Institute. My responses are numbered to correspond to selected interrogatories.
7.
The action or effect of a positive (graphite) temperature coefficient was considered in our analysis of the inadvertent transient accident.
A positive temperature coefficient associated with the graphite moderator / reflector in the Argonaut Type design would be a delayed temperature coefficient.
Therefore, the effect of such a coefficient 0204210469 820419 PDR ADOCK 05000
would not be significant during an inadvertent transient since the accident timescale is on the order of a few seconds or less.
If long term (i.e., several hours) operation of the reactor produces a gradual decrease in the negative temperature coefficient of the water moderator / coolant (presumably due to a positive temperature coeffi-cient of the graphite), the net value of the coefficient would still be negative as long as the (absolute) magnitude of the negative coefficient was greater than the positive coefficient.
If the two coefficients were equal or even if the positive coefficient were some-what larger than the negative coefficient, accumulation of fission product poisons (e.g., xenon poisoning) would counteract positive reactivity gains from the increased temperature.
8.
Not applicable; see answer to interrogatory number 7.
9.
In NUREG/CR-2079, the approach to the analysis of the inadvertent transient used a conservative value of the prompt neutron lifetime, which produces a shorter period for a given excess reactivity inser tion and ultimately a larger energy release. Another main factor in our approach was to use the fact that the fuel elements, being solids, have a relatively constant value for heat capacity.
Furthermore, it was assumed that all heat generated in the fuel elements would remain in the fuel elements, again a most conservative assumption.
The expected fuel temperature was calculated assuming no heat transfer and independently estimated from SPERT data and a generic method (i.e.,
j using an unspecified shutdown mechanism) was used to calculate the energy release.
It should be noted that some of the studies (ANL, 1961 ATL) mentioned in the intervenor's question predict a much higher reactivity insertion (4,75% to 4.5% ak/k) could be safely accommodated by an Argonaut or Argonaut type reactor.
1 19.
If the graphite in the moderator / reflector has a measurable positive temperature coefficient, then the negative temperature coefficient of l
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the water moderator / coolant would be reduced by that amount produced by the temperature rise in the graphite.
20.
The intervenor's question does nct specify a particular temperature coefficient (e.g., fuel, water or graphite).
To a first approxima-tion, temperature coefficients are constant with temperature, and so the values would be similar, if not identical, over a range of tem-peratures.
40.
The 12 MWs energy release, when proportioned among 240 fuel plates in the core..will produce 0.075 MWs/ fuel plate for.the.centermost fuel plates. The average value :would be 0.050 MWs/ fuel plate.
In the referenced work [Ivins, R. O. ' 1963.
"A Study. of the Reaction of
' Aluminum / Uranium Alloy Fuel Plates with Water Initiated by a
' Destructive Reactor Transient."
Trans. Amer. Nuc. Soc.
6(1):101-102]
the energy per gram of fuel plate at which deformation was observed but only negligible reaction occurred was 174 cal /g.
Energy conver-sion factors (see for example table of conversion factors in CRC Handbook of Chemistry and Physics, 57th ed.) will enable cal to be converted to MWs.
A nominal mass for a fuel plate was taken to be about 215 grams, recognizing that differences in design and manufac-turing tolerances will create some variation.
Therefore 174 cal kW-hr 1 MW 3600s 5
1000 kW
- hr
__ 9 8.6 x 10 cal 215 g 0.156 MWs
- fuel plate
- fuel plate Therefore, the calculated maximum energy release (12 MWs) would not produce core disruption leading to cladding failure.
41.
As mentioned on page 15 of NUREG/CR-2079, 60 C is the nominal operat-ing temperature of the water moderator / coolant.
This value was
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assigned to the fuel plates to minimize the temperature increase required to reach the nominal melting point of the fuel (640*C) and thus be conservative.
The value of 60'C was derived from the average of values for the primary coolant outlet temperature for Argonaut reactors. References include recent (i.e.,1979 or 1980) SARs from UCLA, VPI, and the University of Florida.
46.
The cladding of the fuel plates is generally specified as aluminum.
However, the cladding is probably not pure aluminum but contains other, albeit minor, constituents. Thus, the exact properties of the cladding,, including ductility, could be expected to vary from those of pure aluninum.
Furthermore the properties would be expected to vary
/according' to the. condition, form and treatments given the metal.
Aluminum.is ductile, to some degree, at any temperature below its melting point until some temperature when brittleness will supersede ductility.
Therefore, there is no specific temperature at which aluminum becomes ductile.
To discover a specific temperature, tem-perature range or temperature threshold at which the cladding fails because " internal pressure" or expansion of solid or liquid components inside the cladding stretch the cladding beyond its limit of ductility would require further research.
47.
There is no specific temperature (other than perhaps absolute zero) at which volumetric expansion of the fuel begins.
As long as the fuel material has a positive coefficient of expansion, any increase in tem-perature will cause the material to expand to some degree.
l 60.
Since the exact number and configuration of beam tubes will be site l
specific, the compensation between positive reactivity generated by flooded experimental facilities and negative reactivity generated by water in the interstitial spaces among the graphite blocks was assumed to be equal for the purposes of our report. The maximum estimated (UCLA SAR 1960) reactivity gains from beam tube changes (+0.18% ak/k)
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is less than the reactivity changes based on the experiences at the University of Washington in which moisture was confirmed to be in the spaces among the graphite blocks. The reactivity loss was about -
0.3% ak/k (letter to Mr. D. J. Skovholt from Dr. A. L. Babb, dated 12-16-68).
Therefore, our assumption of equal compensation is con-servative.
61.
As reported in the University of Washington Safety Analysis Report 2350 is obtained with the (1960), the minimum critical mass of 1.9 kg slabs together and control blade slots eliminated.
71.
For the purposes of MUREG/CR-2079, " shock wave phenomenon" would be
- the result of formation of steam or steam bubbles -from the' surface of hot fuel plates.
94.
The intervenor's question does not specify what SL-1 data. However, basically the nature of the SL-1 reactor was too dissimilar to Argo-naut type reactors. Furthermore the (unplanned) destructive event did not permit the opportunity to gather extensive data on tie reactor behavior, as did occur with the test program for the SPERT reactor.
102.
The answer to this question would require a detailed analysis of more SPERT data than is presently available to me.
The value in question (590 C) was obtained by simply equating the ratio of SPERT peak tem-perature: energy release with the calculated energy release from the Argonaut inadvertent transient analysis.
To assign er ror bars and determine a 95* confidence level would require uncertainties be known for all the components and factors.
I do not know these at the pre-sent time.
The value of 1500 C (obtained from Miller R. W., A. Sola and R. K. McCardell.
1964.
Report of the SPERT I Destructive Test Program on an Aluminum, Plate-Type, Water Moderated Reactor.
m
100-16285, Phillips Petroleum Co., NRTS, ARCO, Idaho) was not reported with error bars or assigned an uncertainty.
A(1).
A current resume and a statement of professional qualifications for S. C. Hawley are attached to this affidavit.
A(2).
I am not now, nor have I been within the last five years, an employee of any of the five current licensees of Argonaut reactors.
A(3).
I have not received, either now or within the last five years, a pay-
. check from any= of. the,five current Argonaut reactor. licensees.
A(4).
I have personal acquaintance with one current and one past staff of the University of Washington Argonaut reactor (Mr. P. Miller, pro-fessional acquaintance and Mr. W. E. Wilson, a former employer).
Ho w-ever, I am unaware of both the specific background experience of all my acquaintances and the current and past staff of all five current Argonaut licensees.
Therefore, if specific individuals are identi-fied, I will try to describe the nature of my acquaintance with each.
A(5).
I have gained personal knowledge of the general configuration of Argonaut reactors through a visit to the University of Washington Argonaut reactor facility.
A(6).
I have never been employed at an Argonaut reactor.
A(7).
I have never operated an Argonaut reactor.
A(8).
While I did not personally perform all of the calculations or research used in this report, I have nonetheless reviewed the document and find nothing illogical or unreasonable.
Thus, I endorse the report as written, with the errata dated July 21, 1981, and the correction of M
the typographical errors on page 39, line 10, fifth word ("present,"
rather than " prevent") and on page 43, line 3, which should read
"...Si nce the open..." rather than "... Since opening..." and on page 19, lines 13 and 16, which should read ".. 54 C..." rather than
". 74 C...".
A(9).
I do not have any reservations about the report with the errata cor-rections.
A(10).
I believe the abstract and summary of the report (with errata correc-
.tions) reflect.the content of the report.
A(11).
1 do not have any reservations cabout.the' abstract -and summary (with errata corrections) reflecting the content of the report.
A( 12).
There are almost an infinite number of accident scenarios which some one could consider credible.
However, the majority of these would require site specific conditions or be of less magnitude than the generic-type accidents illustrated in NUREG/CR-2079.
A(13).
Not applicable.
A(14).
I believe we did not analyze all destruct modes for Argonaut reactors because sabotage was specifically excluded.
A(15).
Based on the work performed to produce NUREG/CR-2079, the answer to this interrogatory is no.
3)
I hereby certify that the preceding information based on the research conducted in connection with NUREG/CR-2079 is true and correct to the best of my knowledge and belief.
Z $ r,_ (l.N b-
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Sean C. Hawley d
Subscribed and sworn before me on this 6i day of April 1982.
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< Notary Public
- / S~~ / k I k My commission 2 expires:
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SEAN C. HAWLEY Professional Qualifications My name is Sean C. Hawley.
I am a research scientist employed by the Radiological Sciences Department at Battelle, Pacific Northwest Laboratory, Richland, Washington.
I provide support to senior staff in external contacts with sponsors and technical experts and occasionally direct the activities of small groups.
I occasionally interact directly with sponsors and scientists external to my group and usually publish as a junior author.
I received a Bachelor of Arts, Degree.in Chemistry from Reed. College, Portland, Oregon in 1978.
In addition, I have completed 10 credit hours of graduate level studies in Radiological; Sc.iences at Washington State University and the
. University of Washington (Joint Center for Graduate Studies, Richland,
. Washington).
I have about eight years of experience working in areas related to research reactors.
I received my first Senior Operator's Permit in 1973 for the Reed College Reactor Facility.
I was employed there as a Senior Reactor Operator, Assistant Health Physicist, Reactor Supervisor and Training Supervisor.
I received my second Senior Operator's Permit in 1979 for the Washington State j
University Reactor.
I was employed there as Reactor Supervisor.
l I am a member of the American Chemical Society, and the Columbia Chapter of the Health Physics Society.
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SEAN C. HAWLEY, Research Scientist, Health Physics Technology Section, Radiological Sciences Department Education B.
A., Chemistry, Reed College 1978 Washington State University 1979-1980 Joint Center for Graduate Studies 1980-Present (University of Washington)
Experience Nr. Hawley has eight years of experience working with research reactors and one year of experience evaluating emergency preparedness at nuclear power plants.
e Aeactor Operation and Supervision. ' Mr. Hawley received his first Senior
! Operator's Permit in 1973 for the Reed College Reactor Facility.
In his six years.there, he was a senior reactor operator, assistant health physicist, reactor supervisor, and training supervisor.
In 1979, he received his second Senior Operator's Permit for the Washington State University reactor.
As reactor supervisor, he was responsible for the safe operation and maintenance of the reactor, and also advised and instructed researchers in the methodology of neutron activation analysis, e Accident Analysis and Energency Preparedness.
Since joining Battelle, Mr. Hawley has been analyzing credible accidents for research reactors and participating in emergency prepardness appraisals and exercise obser.
vations at nuclear power plants.
He is also reviewing emergency plans for other NRC licensed facilities.
Mr. Hawley is a member of the Anerican Chemical Society and the Health Physics Society.
Publications Hawl ey, S. C., R. L. Kathren and M. A. Robkin.
1981.
Analysis of Credible Accidents for Argonaut Reactors.
NUREG/CR-2079, PNL-3691, Paci fic Northwest Tiboratory, Richland, Washington.
Ha wl ey, S. C., a nd R. L. Ka th ren.
1982.
Credible Accident Analyses for TRIGA and TRIGA-Fueled Reactors.
NUREG/CR-2387, PNL-4028, Pacific Northwest Laboratory, Richland, Washington, s