NUREG-0401, Response to Jf Doherty Tenth Set of Interrogatories.Contains Clarification Re NRC Safety Evaluation Rept,Suppl 2,Section 6.2.1(5),in Ref to GE Ongoing Test Program & Explanations Re NUREG-0401 on BWR Monitors.W/Affidavit & Certificate of
| ML19296B535 | |
| Person / Time | |
|---|---|
| Site: | Allens Creek File:Houston Lighting and Power Company icon.png |
| Issue date: | 02/12/1980 |
| From: | Moon C, Sohinki S Office of Nuclear Reactor Regulation, NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| To: | Doherty J DOHERTY, J.F. |
| References | |
| NUDOCS 8002210010 | |
| Download: ML19296B535 (23) | |
Text
I 02/12/80 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
HOUSTON LIGHTI'.G & POWER COMPANY
)
Docket No. 50-466
)
(Allens Creek Nuclear Generating
)
Station, Unit 1)
)
NRC STAFF RESPONSE TO JOHN F. DOHERTY'S TENTH SET OF INTERROGATORIES The NRC Staff responds as follows to the tenth set of interrogatories pro-pounded by John F. Doherty in this proceeding.
10-1.
NUREG-0687, Supp.1, on page 5 states, " Safety relief valve flow data was not found for any size valve flowing steam-water in the supercritical state.
This was surprising since many facilities have been designed and operated with steam-water in the super-critical state for many years." Does staff take the position that it is not necessary for such data to be collected to license the ACNGS facility? (D-17)
,,monse The Staff is unable to locate a report identified as NUREG-0687, Supp.1.
If HUREG/CR-0687 is the intended number, see response to your interrogatory 9-1.
10-2.
According to the Government Accounting Office Report, "The NRC needs to Aggressively Monitor and Independantly [ sic] Evaluate Nuclear Powerplant Construction", page 21, as of Sept. 7,1978, there was no rule or regulation "... to protect individuals from reprisals by employers or others if those individuals tell NRC of poor construction activities as a nuclear power plant con-struction site."
(T-AA4 on Applicant competence to construct ACNGS) a.
Is there now such a rule?
b.
If not is there a rulemaking proceeding in progress?
80 9210 \\D h
. Response S-Section 210 of the Energy Reorganization Act of 1 - ~ ;, as amended, entitled " Employee Protection."
10-3.
Item #021.01 on Page 10 21.01 cf the PSAR lists 3 items of additional information required by Staff of Applicant. Have these items been supplied yet? If not, which are incomplete?
(D-17)
Response
a.
In Section 6.2.1(5) Supplement No. 2 to the Safety Evaluation Report we stated:
As a result of test data that have been obtained thus far in the Mark III testing program, the applicant has made a number of modifications to the design and orientation of structures located above the suppression pool.
In addition, the applicant has committed to evaluate the effects of additional test data,that will be cbtained
'from further testing and redesign or reinforce structures as necessary. During the operating license stage of review, we will verify that the final design of the containment accounts fo" pool swell loads calculated in accordance with the results of our review of the Mark III testing program.
In the two years since the Applicant provided the response to Item 021.01 (Amendment No. 42 4/3/78) the General Electric Company has continued to update information from its ongoing test program. The Applicant's revised information provided by Amendment No. 54 is believed to include its reaction to some of General Electric's updated information but not all that has or will be submitted to the Staff by the General Electric Company. The Applicant has not identified any deviations and has provided a commitment to identify
. deviations that may arise. The Staff has not accepted the Applicant's position that a benchmark case may not be required.
b.
Supplement No. 2 to the Safety Evaluation Report for Allens Creek -
NUREG-515, March 1979 - page 6-5.
Preliminary Safety Analysis Report, Allens Creek - pages 1021.01-1, 1021.01-2, 1021.01-3 and 1021.01-4.
c.
Same as (b).
d.
The response to Interrogatory 6.1 continues to be applicable.
Since the Applicant has made acceptable commitments no other work by the e.
Staff is required prior to a decision or issuance of a construction permit.
Prior to that decisica, the Staff has a continuing responsibility to advise the Atomic Safety and Licensing Board of any changes in the Staff bases (as reported in the Safety Evaluation Report and Su?plements) for concluding that the application satisfies the Commission's requirements for issuance of a construction permit. The results of General Electric Company's ongoing test program will be reviewed by the Staff on a continuing basis to establish acceptance criteria for Allens Creek at the operating license stage of review.
f.
See (d).
10-4.
In NRC tems (jargon, lingo) does a Power Excursion Accident (PEA) include Anticipated Transient Without Scram (ATWS) accident?
(D-17)
Response
Table 15-1 of Standard Format and Content of Safety Analysis Reports for a.
Nuclear Power Plants, LWR Edition," Regulatory Guide 1.70 Revision 3, November 1978, lists " Representative Initiating Events to be Analyzed in Sections 15.x.x of the PSAR." Item 4, " Reactivity and Power Distribution Anomalies," does not include Ancicipated Transient Without Scram (ATWS) accidents. These are tabulated separately under Item 8, " Anticipated Transients Without Scram."
b.
Not applicable except as already included under (a).
thru f.
10-5. 'What is Staff's current position with regard to the adequacy of the Dutt-Baker Correction factor in predicting fission gas release from fuel rods?
(D-20)
Response
a.
The Dutt and Baker correlationSS was developed to describe the release S
.S. Dutt, D.C. Bullington, R.B. Baker, and L.A. Pember, "A Correlated D
Fission Gas Release Model for Fast Reactor Fuels," Trans. Am. Nucl. Soc.
15, 198 (1972).
S
.S. Dutt and R.B. Baker, "SIEX:
D A Correlated Core for the Prediction of Liquid Metal Fast Breeder Reactor (LMFBR) Fuel Thermal Perfomance,"
Westinghouse Hanford Report, HEDL-TME 74-55, June 1975.
. of fission gas from EBR-Il fuel at high burnup.
From the Dutt and Baker correlation, a correction method was developed by the NRC Staff for commercial LWR fuel applications beyond 20,000 mwd /MtU. This correction method is currently required in many safety reviews. The derivation and application of this correction method are the subject of a Staff report 2/
Although we continue to believe that this correction method is adequate when combined with previously approved methods for calculating fission gas release above 20,000 mwd /fitU, other methods for calculating the release at high burnup can also be used after they have been reviewed and approved by the Staff.
b.
The documents are listed as referenced in 10-5(a). Other basic references can be found in NOREG-0418.
Fission gas release reports are numerous and have appeared over a period c.
of many years. The pertinent references are noted above.
d.
The authors of NUREG-0418 support the answer to the question.
R.0. Meyer and J.C. Vogelwede are with the USNRC.
C.E. Beyer is with the Hanford Engineering Development Laboratory, Richland, Washington.
e.
The Staff is active in the development of the ANS-5.4 Standard, " Method for Calculating the Release of Fission Products from 0xide Fuels."
2/ R.O. Meyer, C.E. Beyer, and J.C. Voglewede, " Fission Gas Release from Fuel at High Burnup," U.S. Nuclear Regulatory Commission Report NUREG-418, March 1978.
. R.0. Meyer is a member of the ANS-5.4 Committee. Other work in this area involves the review of modified fuel performance models as submitted by the various vendors (B&W, CE, E von, GE and W) for approval, f.
R.0. Meyer or J.C. Voglewede of the Core Performance Branch will be available to testify on this matter.
10-6.
In Applicant's letter (E.A. Turner to D. Vassallo of NRC),
Attachment C, Page 3, with regard to I.E.Bulletin 79-08 (Events at TMI, relevant to BWRs) Applicant responds to the Position (2.1.2 " Performance Testing for BWR and PWR Relife [ sic] and Safety Valves" by saying it will support industry efforts in determining the accident and operating conditions to which ACNGS valves will be subjected to.
In view of the fact that NUREG/CR-0687 Supp.1 " Study of Safety Relief Valve Operation Under ATWS Conditions" concluded (Pg. 5) there was no " Safety relief valve flow data...
for any size valve flowing steam-water in the supercritical state":
a.
Is Applicant being considered to test its SRVs in the conditions quoted from NUREG/CR-0687, Supp.1, above?
(Doherty Contention 17) b.
Is Applicant's commitment here only to determine the-demands on SRVs under anticipated transient events?
c.
Who will conduct such research?
Response
See response to 9-1.
10-7. Your response to Interrogatory #7 of Set #6 quotes 55.2.6 of GESSAR SER. My NUREG-0152 does not show this statement.
Please check your reference.
Is the quote from NUREG-01247
. Response The intended reference is the GESSAR-238 Nuclear Island Standard Design, a.
NUREG-75/110, Section 5.2.6, page 5-7.
10-8.
Relevant to Contention #17, what is the highest temperature in tests mentioned in your response to question #14 of my Sixth Set of Interrogatories dealing with the quencher design?
Respo _e a.
Approximately 212 F.
b.
" Test Results Einployed by G.E. For BUR Containment and Vertical Vent Loads," NEDE-21078-P Clcss III, Company Prop.-ietary, General Electric Company, October 1975. Appendix III pages 1-28.
c.
None.
d.
The response to Interrogatory 6-1 continues to be applicable, e.
No.
f.
See (d).
10-9.
Relevant to Doherty Contention #44 with regard to the danger of LOCA from cracked ' pipes being subjected to water-hammer forces, and in particular detecting pipe cracks, would staff agree that acoustic emission test data analysis and extrapolation to real structures has not been completely proof tested in operating nuclear plants?
' Response a.
- Yes, b.
Since proof testing of acoustic emission test data analysis and extra-thru f.
polation to real structures is not a Staff requirement in this Allens Creek construction permit stage of review, (b) through (f) are not applicable.
10-10. How far behind schedule is Staff in the rulemaking for ATWS mentioned on line 20 and 21 of Page 8 of the 225th General Meeting of ACRS (1/4/79) provided in reply to Question #1 of my First Set of Inter-rogatories?
Response
See response to 9-1.
10-11. The response by staff to my question #8 (reply part "d")is worded futuristically. Has there been any change in the role of costs in safety decisions; that is in the publicly stated policy?
Response
See response to 9-1.
10-12.
If the answer to 10-11 is "no", has the current policy been affirmed or strengthened by the Commission at any particular meeting?
Response
See response to 9-1.
_g.
10-13. Pertinent to Doherty Contention #25, how might the Main Steam-Line Radiation Monitor and Off-Gas System Radiation Monitor fail to detect in the case of the postulated BWR flow blockage accident as stated on page 23 of NUREG-0401, " Fuel Failure Detection in Operating Reactors"?
(Note: The NUREG gave no explaination.
[ sic] What is there about the accident that makes it difficult for these monitors to do their jobs?)
Response
a.
In NUREG-0401, the Staff concluded that BWR activity monitors, along with other primary system sensors, are adequate to give an early warning and allow a timely response to degrading fuel conditions. A possible exception to this adequacy was stated to be associated with a postulated BWR flow blockage accident, which might proceed undetected in its early stages.
The inadequacy of the BWR monitors in the early stages is only a postulated concern and has as a basis, several factors:
(1) blockage confined within a channeled BWR fuel bundle might not lead to any immediate unusual primary system indications (pressure and temperature); (2) approximately 2 minutes would be required for fission products leaving damaged fuel to reach the sensitive off-gas radiation monitors; and (3) if the blockage were complete (and the bundle must be almost totally blocked to produce severe t: mage),
fission products might remain bottled up in the damaged fuel bundle and not reach the radiation monitor.
Therefore the early stages of damage might go undetected for about 2 minutes or even longer if fission products were not being released to the primary system.
. b.
The references for this subject are listed in NUREG-0401.
c.
Other references may exist but were not reviewed for the answer.
d.
Members of the Core Performance Branch, Accident Analysis Branch Radiological Assessment Branch, Environmental Evaleation Branch and the Fuel Behavior Research Branch are involved in this area of review.
e.
The Fuel Behavior Research Branch is supporting programs in this area.
The fic, blockage experiments (see our reply to Interrogatory 8-19) are scheduled for after 1983. As part of the P3F program, fission product detection systems are currently under study. The results of this study will better dsfine the capabilities of the monitoring systems but may not necessarily reduce the time delay as noted above.
2 f.
M.D. Houston of the Core Perfennance Branch will be available to testify on this matter.
10-14.
Pertaining)ta Doherty Contention #41, (adequacy of core water level indicators have there been any additional a.
Investigation reports since that of May 29, 19797 b.
Any other documents with regard to t' e May 2,1979 Oyster Creek incident other than the Amendments t, de i:! ants operating license of July 23, 1979 (#39) and V.ty 30 1979 (#36)?
Response
a.
Direct Response
" Additional Information for NRC Staff Generic Report on BWRs,"
a.
NED0-24708, General Electric Company - August,1979.
" Generic Evaluation of Feedwater Transients and Small-Break Loss-of-Coolant Accidents in GE-Designed Operating Plants and Wr-larm Operating License Applications," NUREG-0626, January 1980 (being printed for availability from NTIS within a few weeks).
b.
None.
b.
None except as listed under (a).
c.
- None, d.
Question only refers to reports and documents.
e.
None in response to the question.
f.
See (d).
10-15. Will Applicant be required to keep at least one recirculation loop open during all times except when the reactor vessel head is removed?
(This requirement was imposed on Oyster Creek following the May 2 1979 incident, except they were limited to two being in operation since that plant has four recirculation pumps.)
D-41.
. Response a.
No. Natural circulation characteristics are shown in Figure 4.2.9 of the Allens Creek PSAR. Since Allens Creek uses jet pumps and a different internals configuration the Staff at this time has not identified a need to keep at least one recirculation loop open during all times except when the reactor vessel head is removed.
b.
Not applicable because Allens Creek and Oyster Creek designs are different thru f.
10-16. Pertaining to Caherty Contention 640 (site adequate in view of the higher than expected gas releasts from TMI) NUREG-0581, " Summary of NRC LWR Safety Research", page i, statas:
Fission product release measurements made at ORNL, indicate the amount of Cesium and Iodine escaping from a defected PWR rod during-a LOCA with success-ful ECCS operation will be on or two order of magnitude less than gap release assumptions for these species used in licensing work, a.
Did this experimental observation have any hint or suggestion of confirmation at the TMI incident?
b.
In answering (a) please try to give an indication for both source terms.
Response
a.
The inquiry presented in Interrogatory 10-16 has been, for the most part, addressed previously in our response to Interrogatory 8-9.
The statement concerning ORNL release measurements specifically refers to the fission product emission from a perforated (defective) fuel rod during a LOCA with successful ECCS operation.
In TMI-2, fuel rod damage is certainly
. more severe than simply perforation and the upper one-third of the core is assumed to be rubblized. Thus, there should be no correlation between the ORNL statement and the release experienced at TMI-2.
Cesium and iodine are fission products formed during the thermal fission of U-235.
For every 100 uranium atoms fissioned, twenty atoms of cesium and one atom of iodine are fonned.
For typical BWR fuel irradiat' i to 25,000 mwd /tU, the approximate activities of cesium and iodine from the calculated composition given by the ORIGEN computer code are as follows:
Nuclide Curies /MT of Ilranium CS-134 126,000 CS-135 0.268 CS-136 48,800 CS-137 83,200 I-129 0.0233
~
I-131 963,000 (Allens Creek contains about 148 MT of Uranium) b.
The assessment of the TMI-2 core damage was presented in Appendix A of NUREG-0557 " Evaluation of Long-Term Post-Accident Core Cooling of Three Mile Island Unit 2."
The nuclide activities were taken from Table G.6 given in NUREG-0404, Volume 2, " Draft Generic Environmental Impact State-ment on Handling and Storage of Spent Light Water Power Reactor Fuel."
. c.
Numerous references can be found in this general area. We did not review any others to answer this question.
d.
Members of the Core Performance Branch and the Fuel Behavior Research Branch have been involved in this review and will support the answer.
e.
As part of the PBF and LOFT programs supported by the Fuel Behavior Research Branch, information concerning fission product release from defective fuel rods will be forthcoming. Further information concerning the true status of the dan =ge to the core at TMI-2 will be determined as the clean-up activity continues.
f.
M.D. Houston of the Core Performance Branch will be available to testify on this matter.
10-17. Relevant to Doherty Contention #17, Page 52 of "The Three Mile Island Canmission Report, says:
"For example, at TMI-2, the PORV was not a ' Safety-related' item because it had a block valve behind it. On the other hand, the block valve was not ' safety-related' because it had a PORV in front of it."
a.
Are the ACNGS PORVs " safety related"?
b.
Are there block valves behind the ACNGS power operated relief valves?
c.
If the answer to (b) is "yes", are the block valves " safety related" or are they excused from this requirement?
Response
a.
Direct Answer a.
The Allens Creek Safety Relief Valves (SRV's instead of PORV's) are safety related.
b.
There are no block valves behind the SRV's.
b.
Allens Creek SER Supplement No. 2, NUREG-0515, pages 5-3 and 7-6.
c.
GESSAR SER, NUREG-0152, pages 5-7 and 7-26.
d.
Answer is based on information already documented.
e.
None planned.
f.
None planned.
Professional Qualifications John C. Voglewede g
Core Performance Branch Division of Systems Safety U.S. Nuclear Regulatory Commission My name is John C. Yoglewede.
I am employed as a Reactory Engineer with the Core Performance Branch, Division of Systems Safety, U.S. Nuclear Regulatory Commission, Washington, D.C.
The responsibilities of this position include the review of nuclear fuel design and performance data and the related analyses as used in support of power plant licensing submittals.
My general technical background is that of a nuclear fuels engineer with experience in high-temperature materials, steady-state and transient fuel performance modeling, and scientific application of data processing equipment.
I am familiar with the mechanical properties, testing, fabrication, characterization, and criticality control of nuclear ceramics.
I am also familiar with the regulatory requirements associated with nuc1 car fuel performance.
I hold the degree of Bachelor of Science in Physics (1969) from St. Procopius College and the degree of Master of Science in Computer Science (1976) from Illinois Institute of Technology.
From 1965 to 1969, I was an undergraduate student at St Procopius College (Illinois Benedictine College) at Lisle, Illinois.
From 1969 to 1977, I was employed as a Scientific Associate with the Ceramics / Fuel Properties Group in the Materials Science Division at Argonne National Laboratory.
During this period. I worked with high-speed data acquisition and control systems in order to study the transient behavior of nuclear fuels in'out-of-reactor simulation experiments.
I developed computer models for the analysis of these experiments and was also involved with property specification and model development for the laboratory fuel performance codes. As principal investigator in a mechanical properties program, I studied high -temperature creep and densification behavior of oxide nuclear fuels.
I was responsible for the nuclear criticality and operational control of a plutonium mechanical testing facility.
In February 1977, I began working for the Core Performance Branch, Division of Systems Safety, U.S. Nuclear Regulatory Commission, as a Reactory Engineer. The responsibilities of this position include the review of nuclear design and performance data, which are submitted as part of an applicant's Safety Analysis Report. The specific areas of review are the nuclear and fuel systems design as well as the thermal and hydraulic design of the reactor core (Chapter 4 of the Standard Format). My major responsibility has been the review of analytical methods for fuel thermal performance predictions, which are developed and described by each reactor vendor and subsequently referenced in nuclear power plant licensing submittals. These analytical methods are normally implemented in the form of computer codes.
In June of this year, I was reassigned to the Three Mile Island Unit 2 Lessons Learned Task Force as the representative of the Core Performance Branch.
I served in this capacity until October 1979, at which time I was transferred to the Three Mile Island
\\
Unit 1 Restart Review Task' Force.
I am an active member of the American Nuclear Society. A list of my professional publications is attached.
JOHN C. V0GLEk'EDE PUBLICATIONS 1.
R. O. Meyer and J. C. Vogleuede, Temperature Gradient Vacu:en Furnace for Diffusion Studies to 2000*C, Rev. Sci. Inst. 42_(7), 993-995 (July 1971).
2.
A. A. Solomon, J. L. Routbort, and J. C. Voglewede, Ficsion-induced Creep of UOp and its Significance to Fuc2-ele ~:ent Fcrformance, ANL-7857 (September 1971).
3.
J. L. Routbort, N. A. Javed, and J. C. Voglewede, Thermal Creep of Nired-oxids Puel FeZZets, Am. Ceram. Soc. Bull. 51, 389 (April 1972).
ABSTRACT 4.
J. L. Routbort, N. A. Javed, and J. C. Voglewede, Compressive Crccp of Nixed-oxide Fuel FeIZcts, J. Nucl. Mater. 4_4,(3), 24 7-259 (September 1972).
5.
J. L. Routbort and J. C. Voglewede, Creep of Nimed-o ide Fuel FcIIets at High Stress, Am. Ceram. Soc. Bull. 52(4), 352 (April 1973).
ABSTRACT 6.
J. L. Routhort and J. C. Voglewede, Correlation of Oxide Fuel Creep vith Microstructure and the Influence on Fuel-element Fcrformance, Am. Ceram.
Soc. Bull. g(4), 398 (April 1973). ABSTRACT 7.
J. L. Routbort and J. C. Voglevede, FinaZ-stage Densification of Nixed-(
oxide Fuel, Am. Ceram. Soc. Bull. 5_2(9), 721-722 (September 1973). ABSTRACT 8.
J.T.A. Roberts, J. L. Routbort, J. C. Voglevede, and A. A. Solomon, Development of a Mechanical Nodel of In-reactor Fuct Behavior: Status Report, ANL-8028 (July 1973).
9.
J. L. Routbort and J. C. Voglevede, Creep of Nixed-oxide Fuel FeIIcts at Hi 'h Stress, J. An. Ceram. Soc. 56(6), 330-333 (June 1973).
t 10.
J.T. A. Roberts and J. C. Voglewede, AppZication of Deformation Naps to the Study of In-reactor Behavior of Oxide Fucis, J. Am. Ceram. Soc. 56(9),
472-475 (September 1973).
11.
J. L. Routbort and J. C. Voglevede, Final-stage Densification of Nimed-oxide Fuel, Am. Ceram. Soc. Bull 53,(4), 363 (April 1974).
ABSTRACT
~
12.
J. C. Vogleuede, Thermal Densification of Nixed-oxide Fuel, Am. Ceran.
53(8), 619 (August 1974). ABSTRACT Soc. Bull.
3 13.
J. C. Voglewede, Performance Analysis of Cache Memory, M.S. Thesis, Illinois Institute of Technology (May 1976).
14.
C. R. Kennedy, F. L. Yaggee, J. C. Voglewede, D. S. Kuppermen, B. J. Wrona, W. A. Ellingson, E. Johanson, r.nd A. G. Evans, Cracking and ReaZing Behavior of UOg as Related to Feltet-Cladding Nechanical Interaction: Interim Report
(
July 1976, Argonne National Laboratory Report ANL-76-110 (October 1976).
15.
C. R. Kennedy, J. C. Voglewede, and F. L. Yaggee, Cracking and Crack EcaZing of UOn Pellets in Simulated Ik'R Po:Jer Cyc7cs, Am. Ceram. Soc. Bull. 55,(9),
821 ($eptember 1976). ABSTRACT l es.
C. R. Kennedy and J. C. Vog1euede, Relocation Phcnomena in UOg Pellets Subjected to Simulated Ik'R Pouer Cycles, Am. Ceram. Soc. Bull. 5_6(3), 342 (March 1977).
ABSTRACT 17.
J. C. Voglewede, Application of Puel Properties Data to Out-of-Reactor Simulation Studics, Am. Ceram. Soc. Bull.,5_6(3), 342 (March 1977).
ABSTRACT 6
18.
B. J. Wrona, J. C. Voglewede, and T. M. Calvin, Effcets of Pc1Icts Density and A=ial Restraint on Failurc Threchold, Trans. Am. Nucl. Soc. 26, 376-377 (June 1977).
ABSTRACT 19.
R. O. Meyer, C. E. Beyer, and J. C. Voglewede, Fission Gas ReIcase from FucI at High Burnup, U.S. Nuclear Regulatory Commission Report NUREG-0418, (March 1978).
20.
R. O. Meyer, C. E. Beyer, and J. C. Voglewede, Fission Cas RcZease from Puel at Eigh Burnup, Nuclear Safety 19(6), 699-708 (November-December 1978).
21.
J. L. Routbort, J. C. Voglewede, and D. S. Wilkinson, Final-Stage Densification of A'ized 0 ide PucIs, J. Nucl. Mater. 80(2), 348-355 (May 1979).,
(
A e w
N m
M. Dean Houston Professional Qualifications In 1953, I received a Bachelor of Science degree in Ceramic Engineering from Iowa State College.
In 1957, I received a Master of Science degree in Ceramic Technology from Pennsylvania State University.
I joined the nuclear industry in 1961 as a Senior Ceramic Engineer at NUMEC (now Babcock and Wilcox) with responsibilities for the develop-ment of uraniu=-plutonium oxide fuel and fuel fabrication processes for
. Light Water Reactors (LWR) and Fast Dreeder Reactors (LMFBR).
I con-tinued with research and development programs involving the mixed oxide fuel at Westinghouse Atomic Power (with Battelle-Northwest), 1964-1966, and at the Battelle Columbus Laboratory, 1966-1972.
These studies were 4
concerned with the chemical, physical and electrical prcperties of reactor T,
fuel material and cladding, both out-of-reactor and in-rcsetor.
Since June 1972, I have been a Reactor Engineer with the U.S. Nuclear Regulatory Commission.
I have been a member of the Reactor Fuels Section of the Core Performance Branch of Reactor Safety since November,1973.
In my present position, I_am responsible for reviewing reactor fuel decigns, fuel performance ridels and the ' fuel behavior research projects that are directed by Reactor Safety Research (RSR).
i Ralph 0. Meyer Prc.fercional Out'iricatie-?
In 1960 I received a E.S. in physics from the University of Kentucky and was cade a member of Phi Beta Kappa.
In 1966 I re-ceived a Ph.D. fro: the University of North Carolina (Chapel Hill) with c thesis subject in the field of solid state physics.
Following graduation, diffusion studies related to the thesis topic were continued while I..as a Research Associate in physics at the University of Arisona.
In 1968 I was employed as an Assistant Metallurgist in the rcactor development program of the Materials Science Division at Argonne National Laboratory, Illinois.
At
~
Argonne diffusion techniques were applied to study the properties of nuclear reactor fuels.
This research included studies of gaseous fission product ci ration, segregation of fissile fuel caterial, and 5
restructuring of oxide fuel elements.
More than 20 technical journal papers and topical reports were published on this fundamental and applied research.
In 1973, I joined USNRC as a Reactor Engineer in the Reactor Fuels Section of the Core Performance Branch. In addition to other duties related to the performance of nuclear fuel, I was the principal reviewer of fuel densification analyses. Since 1976, I have been the Section Leader of the Reactor Fuels Section and have a continuing responsibilit, for the review of fuel densification, fission gas release and overall fuel performance.
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
HOUST0fl LIGHTING & POWER COMPANY Docket No. 50-466
)
(Allens Creek Nuclear Generating
)
Station, Unit 1)
)
AFFIDAVIT OF CALVIN W. MOON I hereby depose and say under oath that the foregoing NRC Staff responses to interrogatories propounded by John F. Doherty were prepared by me or under my supervision.
I certify that the answers given are true and correct to the best of my knowledge, information and belief.
) MYY Neizr#
Calvin W. Moon Subscribed and sworn to before me this / ? 'd day of Fu,:t - y 1980.
/
,i'.b.ac. n!
5.1...
Etary Public My Commission expires:
.: 'd <
/. M.C.?
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
HOUSTON LIGHTING & POWER COMPANY Docket No. 50-466 (Allens Creek Nuclear Generating
)
Station, Unit 1)
)
CERTIFICATE OF SERVICE I hereby certify that copies of "NRC STAFF RESPONSE TO JOHN F. DOHERTY'S TENTH SET OF INTERROGATORIES" and " AFFIDAVIT OF CALVIN C. MOON" in the above-captioned proceeding have been served on the following by deposit in the United States mail, first class, or, as indicated by an asterisk, through deposit in the Nuclear Regulatory Commissicn's internal mail system, this 12th day of February,1980:
Sheldon J. Wolfe, Esq., Chairman
- Richard Lowerre, Esq.
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Atomic Safety and Licensing Board Panel Asst. Attorney General for the U.S. Nuclear Regulatory Commission State of Texas Washington, DC 20555 P.O. Box 12548 Capitol Station Dr. E. Leonard Cheatum Austin, Texas 78711 Route 3, Box 350A Watkinsville, Georgia 30677 Hon. Jerry Sliva, Mayor City of Wallis, Texas 77485 Mr. Gustave A. Linenberger
- Atomic Safety and Licensing Board Panel Hon. John R. Mikeska U.S. Nuclear Regulatory Commission Austin County Judge Washington, DC 20S55 P.O. Box 310 Bellville, Texas 77418 R. Gordon Gooch, Esq.
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Texas Public Interest Margaret Bishop Research Group, Inc.
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Houston, Texas 77043 8302 Albacore Houston, Texas 77074 Glen Van Slyke 1739 Marshall Brenda A. McCorkle Houston, Texas 77098 6140 Darnell Houston, Texas 770/4 J. Morgan Bishop 11418 Oak Spring Mr. Wayne Rentfro Houston, Texas 77043 P.O. Box 1335 Rosenberg, Texas 77471 Stephen A. Doggett, Esq.
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,1118 Montrose 1014 Montrose Blvd.
Houston, Texas 77019 Houston, Texas 77019 Robin Griffith Leotis Johnston 1034 Sally Ann 1407 Scenic Ridge Rosenberg, Texas 77471 Houston, Texas 77043 Elinore P. Cummings Atomic Safety and Licensing
- 926 Horace Mann Appeal Board Rosenberg, Texas 77471 U.S. Nuclear Regulatory Comission Washington, DC 20555 Mrs. Connie Wilson 11427 Oak Spring Atomic Safety and Licensing
- Houston, Texas 77043 Board Panel U.S. Nuclear Regulatory Commission Washington, DC 20555 Docketing and Service Section
- Office of the Secretary Carolina Conn U.S. Nuclear Regulatory Comission 1414 Scenic Ridge Washington, DC 20555 Houston, Texas 77043 Mr. William J. Schuessler Mr. Robert Alexander 5810 0arnell 10925 Briar Forest #1056 Houston, Texas 77074 Houston, TX 77042 U
. Sohinki Stephen [for NRC Staff Counsel