ML19338F210
| ML19338F210 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 10/02/1980 |
| From: | Boger B, Peter Hearn, Jensen W, Voglewede J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19338F186 | List: |
| References | |
| ISSUANCES-SP, NUDOCS 8010070639 | |
| Download: ML19338F210 (21) | |
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UNITED STATES OF AMERICA NUCLEAR REGULATORY C0fHISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of METROPOLITAN EDISON COMPANY
)
Docket No. 50-289 (Restart)
(Three Mile Island Nuclear Station, Unit No.1)
NRC STAFF TESTIMONY OF WALTON L. JENSEN, JR.,
JOHN C. V0GLEWEDE, BRUCE A. 80GER, AND PETER L. HEARN RELATIVE TO INSTRUMENT RANGES (tGNP GONItNTION Id)
Ql) Mr. Jensen, please state your position with the NRC.
A)
I am an employee of the U. S. Nuclear Regulatory Comission assigned to the Reactor Systems Branch, Division of Systems Integration, Office of Nuclear Reactor Regulation. From June through December 1979, I was assigned to the Analysis Group of the Bulletins and Orders Task Force, Office of Nuclear Reactor Regulation.
Q2) Have you. prepared a-statement of professional qualifications?...
A)
Yes. A copy of this statement'is attached to this testimony.
Q3) Please state the na'ture of the responsibilities that you have had with respect to the Three Mile Island Nuclear Station - Unit 1.
A)
The accident at Three Mile Island Unit 2 (TMI-2) on March 28, 1979, involved a feedwater transient coupled wit..: the equivalent'of a small 80.1007,0 [
break in the reactor coolant system, though the accident's ultimate severity resulted from a number of interacting elements including lack of complete understanding of system response, misleading instrument readings and inadequate operator training and procedures.
Because of the resulting severity of. ensuing events and the potential generic applicability of the accident to other reactors, the NRC staff initiated prompt action to:
(1) assure that other reactor licensees, particularly those plants such as THI-1 which have a similar design to THI-2, took the necessary actions to substantially reduce the likelihood of future TMI-2-type events from occurring, and (2) initiate comprehensive investig::tions into the potential generic implications of this accident on other operating plants.-
To accomplish some of this work, the Bulletins and Dr.1ers Task Force i
(B&OTF) was established within the Office of Nuclear Reactor Regulation (NRR) in early May 1979.
The B&OTF was responsible for reviewing and 1
directing the_THI-2-related staff activities associated with loss of feedwater transient and small break loss-of-coolant accidents (LOCAs) for all operating plants to assure their continued safe operation 4
I was assigned to the Task Force in June 1979.
I participated in the preparation of' NURE'b0565," Generic' Evaluation 'of: Small ' Break :1.ossdof
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Coolant Accident Behavior in Babcock & Wilcox Designed 177-FA Operating Plants."
Following my assignment to the Reactor Systems Branch, I participated in j
the evaluation of potential feedwater transients at operating B&W plants and participated in the final preparation of the staff Safety Evaluation on the Three Mile Island 1 restart.
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Q4) Mr. Voglewede, please state your position with the NRC.
A)
I am an employee of the Nuclear Regulatory Comission assigned to the Core Performance Branch, Division of Systems Integration Office of Nuclear Reactor Regulation.
Q5) Have you prepared a statement of ;our professional qualifications?
A)
Yes. A copy is attached to tais testimony.
Q6) Please state the nature of the responsibilities that you have had with respect to Three Mile Island Nuclear Station, Units 1 and 2.
A)
From June to October 1979, I was assigned to the Three Mile Island Unit 2 Lessons Learned Task Force as a representative of the Core Perfonnance Branch.
Prior to this time, I had no responsibilities directly related to the review or
. ming of either Unit at TMI.
From October 1979 to January 1980, I was reassigned to the Three Mile Island Unit 1 Restart Review Task Force and was involved in the staff review of the Restart Report. In January 1980, I resumed my duties with the Core Performance Branch of the Division of Systems Safety.
I have continued to work in this capacity although the Core Performance Branch is now part of the NRC's Division of Systems Integration.
4 Q7) Mr. Boger, please state your position with the'NRC.
A)
I am a Reactor Engineer assigned to the Operator Licensing Branch, Office of Nuclear Reactor Regulation.
Q8) Have you prepared a statement of professional qualifications?
A)
Yes. A copy of this statement is attached to this testimony.
Q9) Please state the nature of the responsibilities that you have had with respect to the Three Mile Island Nuclear Stations.
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1 A) During November 1978 and April 1980; I administered operator license examinations on Unit One.
During November 1978, March 1979, and March 1980, I administered operator license examinations on Unit two.
I was a member of the TMI-2 emergency responses team and assisted in the preparation of emergency and contingency procedures during the period of March and April of 1979.
From July 1979 to the present, I have been a member of the TMI Technical Support Staff and have conducted audit examinations on post-accident installed equipment on TMI-2.
I also participated in the review of train-ing and procedures in conjunction with the TMI-l restart programs. This has included preparation of SER inputs and testimony.
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1 010) Mr. Hearn, please state your position with the NRC.
A)
I am an employee of the Nuclear Regulatory Conunission assigned to the Containment Systems Branch. I have been a member of this branch since April 1979.
Qll)
Have you prepared a statement of professional qualifications?
A)
Yes. A copy of this statement is attached to this testimony.
Q12)
Please state the nature of the responsibilities that you have had with respect to Three Mile Island, Units 1 and 2.
A)
During the Operating Licensing Review of TMI, Unit 2, I was responsible for the review and technical evaluation of auxiliary and power conversion systems.
Q13) Please state the purpose of this testimony.
A)
The purpose of this testimony is to respond to that portion of ECNP Contention 1(d) dealing with the ranges of core cooling system and containment isolation system instrisnentation as limited by the board. NRC testimony concerning radiation effluent monitors is contained in the NRC testimony of Phillip G. Stoddart in response to ANGRY (Contention V(D))and partial response to ECNP Contention 1(d).
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That portion of Contention ECNP 1(d) dealing with the ranges of the core cooling sys:sm and containment isolation system instrumentation read: as follows:
"The TMI-2 accident showed that many monitoring instruments were of insuii-cient indicating range to properly warn control room operators of ambient conditions. For example, the " hot-leg" thermo-couples went off-scale at 620*F and stayed off-scale for over 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> f about.13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br /> for reactor coolant loop (b)(5gr reactor coolant loop A and A higher temperature limit would have provided important information to the reactor operators. This situation is unchanged at TMI-1. All monitoring instruments for TMI-l must be calibrated to provide full and accurate readings of the complete range of possible conditions under both nomal and worst-case conditions.
This contention and ECNP Contention 1(c) were accepted by the board with the following limitations:
The scope of these contentions was limited by the board to signals sent to the control room from the core cooling and containment isolation signals.
j (First Special Prehearing Conference Order, December 18,1979). Reference to worst case and worst possible accidents were not accepted by the Board.
The board said that the response to Contention 1(c) should address the adequacy of Class IE control room instruentation following a feedwater transient and small break LOCA and the response to Contention 1(d) should address the ranges of instrumentation in connection with 1(c).
(Memorandum and Order on Licensee's Hotion for Sanctions Against ECNP, June 12,1980).
014) How do you intend to address this contention?
A)
We will identify the instrumentation used by the operator to perfom necessary functions and monitor important variables following a feedwater transient and small break LOCA without regard to whether the instrumentation is labeled Class IE and will demonstrate that the ranges of instruments are adequate.
(See further discussion in NRC staff response to ECNP 1(c).
Mr. Voglewede will sponsor the response to question 18 of this testimony dealing with ranges of reactor system instrumentation; Mr. Boger will sponsor the responses to questions 22, 23, 24, 28 and 29 of this testimony dealing with operator actions; Mr. Hearn will sponsor the responses to Questions 26 and 27 dealing with containment isolation; and Mr. Jensen will sponsor remainder of this testimony.
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Q15) What actions must the operator perfom following a small-break loss-of-coolant accident regardless of whether preceeded by a feedwater transient to assure adequate core cooling?
A)
The actions required of the operator to assure adequate core cooling are described in the Small Break Emergency Procedures for TMI-l (EP 1202-6).
Basically the functions of the operator are to: (1) monitor core cooling, (2) trip the reactor coolant pumps, (3) adjust HPI flow to permit reactor system depressurization maintaining 50'F subcooling in the primary system, and (4) increase the steam generator level to 95% of the operating range using the EFW system.
Once stable conditions are established in the primary system, the operator is to cool down the plant by controlling secondary system pressure.
(Operator actions in connection with contain-ment isolation are discussed in the answer to question 26 below).
Ql6) Following a small-break LOCA, how will the operator at TMI-l monitor core cooling?
A) Core cooling is indicated by the pressure and temperature of the coolant sureounding the core.
Its condition is indicated by approximately 50 core exit themocouples, four cold leg temperature sensors, four hot leg temperature sei. sors, wide and narrow range pressure sensors in each hot leg, and by a subcooling metec.
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A pressure measurement is requirea to determine the boiling temperature of the core water. A temperature measurement is required to detennine the difference between the coolant temperature and the boiling temperature.
If the temperature of the coolant is at or below the boiling temperature, the core will be adequately cooled. If the core is not adequately cooled the temperature of the coolant will be above the boiling temperature. The relationship of the coolant temperature to the boiling temperature will be monitored by the operator from (1) the subcooling meter; (2) a chart maintained by the operator utilizing the hot leg temperature sensors and the reactor system pressure; and (3) by a chart maintained by the operator using the highest reading core-exit thermocouple and reactor system pressure.
Q17) Why does a coolant temperature that is equal to or below the boiling temperature indicate that the core is adequately cooled?
A) The analyses of small-break loss-of-coolant accidents discussed in the NRC response to UCS contention 8 demonstrated that so long as the core is covered by water or a mixture of steam and water it will be adequately cooled.
If the core becomes uncovered heat transfer to the steam will raise the temperature of the steam in and above the core to above the boiling temperature. Therefore, temperature measurements taken above the core which are at the boiling temperature (saturated) or below the boiling temperature (subcooled) indicate that the core is covered and adequately cooled. Temperature measurements taken above the core which are above the boiling temperature (superheated) indicate 8
that the core is not covered.
Q18) What are the ranges of the instrume.ntation used to monitor core cooling?
A)
The range of the core exit thar.nocouples is between anproximately 2500'F and that of the reference junction which is below 100 F.
The method of readout for the core exit thermocouples has not yet been determined but will j
be required to includetemperature indication in excess of 2000*F to the operator.
The hot leg temperature indication has been modified to extend the readout range from (520'F to 620'F) to (120*F to 950"F). The primary system hot leg pressure indication has a range of 0 to 2500 psig. The range of the subcooling mete? display is 100*F superheat to 400'F subcooling.
Q19) Are these ranges sufficient to indicate the condition of the coolant surrounding the reactor core to the control room operator following a feedwater transient and small break LOCA?
A)
Yes. Possible boiling temperatures within the core could range from 669'F at the safety valve pressure setpoint to 212*F at atmospheric pressure and the ranges of the temperature sensors will bracket these possible boiling temperatures so that the operator can determine if the coolant in and above the core is subcooled, saturated or superheated.
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Q20) What is the significance of 50 F subcooling?
A)
Steam cannot exist in the reactor coolant at temperatures below its boiling temperature. Since temperature measuring devices are located at the core exit and hot legs above the core, subcooling at the core exit and hot
" legs indicate that the core is covered with water. The 50 F subcooling 0
requirement for manually throttling HPI flow provides a margin in the procedures.
Q21) Is the range of instrumentation that is available to the operator at TMI-1 adequate to monitor 50 subcooling for the purpose of controling HPI flow?
A)
Yes. At the safety valve setpoint of 2500 psig, 50 F subcooling would be 7
0 achieved at a coolant temperature of 619 F.
At atmospheric pressure of 14.7 psia, a.subcoolino of 50'F would be achieved at a coolant temperature of 162*F. These temperatures are within the ranges of the instrumentation available to the operator.
Q22) Following a small-break LOCA, what information must be available to the operator to indicate that the reactor coolant pumps must be tripped?
A) The reactor coolant pumps are required to be manually tripped when RCS pressure decreases to the ESF initiation setpoint of 1600 psig. Accordingly, the operator must be able to monitor RCS pressure and/or be alerted when RCS pressure has decreased to the ESF initiation setpoint. At TMI-1, RCS pressure is continuously indicated and recorded.
In addition, ESF initiation is annunciated on the alarm panel, thereby, providing a visual and audible indication to the operator.
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Q23) Are other annunciators expected to alam in the event of a small-break LOCA?
A) Yes.
Q24) How does the operator distinguish the correct annJnCiator?
A) Facility procedures and operator training on these procedures direct the operator to monitor RCS pressure during a small-break LOCA. After noting a continuing decrease in RCS pressure, the operitor will anticipate ESF actuation and thereby be able to distinguish the correct annunciator.
In addition, the operator should note each alarming annunciator prior to silencing the alarm.
Q25) Following a small-break LOCA, what infomation will be available to the operator for controlling auxiliary feedwater to increase the steam generator level to the 95% of the operating range?
A) Each steam generator is provided with " start-up", " operate" and " wide" range indication in the control room. The " start-up" and " operate" range indications are each provided with redundant inputs. The " start-up" indication provides level measurement to the operator from 0 to 250 inches of water, the " operate" range is from 100 to 394 inches of water and the " wide" rary indication is from 0 to 600 inches of water. Ninety-five percent of the operate range corresponds to 380 inches of water which would be indicated by both the " operate" and " wide" range indicators.
Thus, three level sensors will be caoable of monitoring the steam generator water level at 95% of the operate range for each steam generator.
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Q26) What is the function of the operator regarding containment isolation following a feedwater transient and small-break LOCA?
A) There is no operator action necessary to isolate the containment following a small-break LOCA. The containment would be isolated automatically following a reactor trip and iritiation of the safety injection system.
The operator only need verify that the containment isolation valves are closed. A check list of these valves will be provided before restart in the small-break LOCA procedure EP-1202-6B.
i Q27) Identify the valves for which the operator must verify closure following a feedwater transient and a small break LOCA.
A)
These valves are:
RP Sump WDL-V534 WDL-V535 RC Drain Tank WDG-V3 WDL-V303 WDC-V4 WDL-V304 RCS Sample CA-V1 CA-V3 CA-V2 CA-V13 RB Purge AH-V1A AH-V1C AH-V1B AH-VID Core Flood Tank CF-V2A CF-V2B CF-V19A CF-V193 CF-V20A I,7-V20B 12
Domin. Water CA-V189 OTSG Sample CA-V4A CA-V5A CA-V4B CA-V5B Let Down Cooler MU-V2A MU-V2B Containment Air Sample CM-V1 CM-V3 CM-V2 CM-V4 Reactor Coolant Makeup System MU-V18 Reactor Building Normal Air Coolers RB-V2A RV-V7 Q28) How does the operator verify that these valves are closed?
A)
Indication of valve closure is provided to the operator by indicator lights for each valve in the control room.
1 Q29) What action does the operator take if a valve closure is not indicated?
A)
The operator will attempt to close the valve using the specific manual valve actuation switch in the control room.
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WALTON L. JENSEN, JR.
PROFESSIONAL QUALIFICATIONS I am a Senior Nuclear Engineer in the Reactor Systems Branch of the Nuclear Regulatory Commission.
In this position I am responsible for the technical analysis and evaluation of the public health and safety aspects of reactor-systems.
From June 1979 to December 1979, I was assigned to the Bulletins and Orders Task Force of the Nuclear Regulatory Commission.
I participated in the preparation of NUREG-0565, " Generic Evaluation of Small Break Loss-of-Coolant Acr'aent Behavior in Babcock & Wilcox Designed 177-FA Operating Plants."
From 1972 to 1976, I was assigned to the Containment Systems Branch of the NRC/AEC, and from 1976 to 1979, I was assigned to the Analysis Branch of the NRC.
In these positions I was responsible for the development,and evaluation of computer programs and techniques to calculate the reactor system and containment system response to postulated loss-of-coolant accidents.
From 1967 to 1972, I was employed by the Babcock and Wilcox Company at Lynchburg, Virginia.
There I was lead engineer for the development of loss-of-coolant computer programs and the qualification of these programs by comparison with experimental data.
From 1963 to 1967, I was employed by the Atomic Enwrgy Commission in the Division of Reactor Licensing.
I assisted in the safety reviews of large power reactors, and I led the reviews of several small research reactor:.
I received an M.S. degree in Nuclear Engineering at the Catholic University of 1
America in 1968 and a B.S. degree in Nuclear Engineering at ;iississippi State University in 1963.
I am a graduate of the Oak Ridge School for Reactor Technology, 1963-1964.
I am a member of the American Nuclear Society.
4 I am the author of three scientific papers dealing with the response of B&W reactors to Loss-of-Coolant Accidents and have authored one scientific paper dealing with containment analysis.
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Professional Qualifications i
l John C. Voglewede
/
Core Performance Branch Division of Systems Safety U.S. Nuclear Regulatory Comission My name is John C. Voglewede.
I am employed as a Reactory Engineer with the Core Performance Branch, Division of Systems Safety, U.S. Nuclear Regulatory Commission, j
3 Washington, D.C.
The responsibilities of this position include the review of nuclear 1
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, fabrications characterization, and criticality control of nuclear ceramics.
I am also familiar with the regulatory requirements associated with nuclear 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 e.xperiments. I developed computer models for the analysis of these experiments and waJ also involved with property specification and model development for the laberatory fue's performance codes. As principal investigator in a mechanical properties program, I stuited high-temperature creep and densification behavior of oxide nuclear fuels.
I was.'esponsible for the nuclear criticality and operational control of a plutonium mechaniel 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 responsibilitia 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.
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l JOHN C. V0GLEWEDE PUBLICATIONS 1.
R. O. Meyer and J. C. Vog1ewede, S'emperature Gradient Vacuum Purnace for Diffusion Studies to 2000*C, Rev. Sci.1nst. M(7), 993-995 (July 1971).
2.
A. A. Solomon, J. L. Routbort, and J. C. Voglevede, Pis:fon-induced Creep of UOp and its Significance to Puel-element Performance, ANL-7857 (September 1971).
3.
J. L. Routbort, N. A. Javed, and J. C. Voglevede, Thermal Creep of Niesd-o=ide Puel PeIIsts, Am. Ceram. Soc. Bull. 51, 389 (April 1972). ABSTRACT 4.
J. L. Routbort, N. A. Javed, and J. C. Voglevede, Compressive Creep of Nimed-oxide Puel PeIIsts, J. Nucl. Mater. 4_4,(3), 247-259 (September 1972).
5.
J. L. Routbort and J. C. Voglevede, Creep of Niced-oxide Puel PeIIets at Righ Stress, Am. Ceram. Soc. 3u11. 52(4), 352 (April 1973). ABSTRACT 2
l 6.
J. L. Routbort and J. C. Voglevede. Correlation of 0..ide Puel Creep uith Microstructure and the Influence on PueI-eZement Performance, Am. Ceram.
Soc. Bull. E(4), 398 (April 1973). ABSTRACT 7.
J. L. Routbort and J. C. Voglevede, FinaZ-stage Densification of Nimed-oxide Puel, Am. Ceram. Soc. Bull. g(9), 721-722 (September 1973). ABSTRACT 8.
J.T.A. Roberts, J. L. Routbort, J. C. Voglevede, and A. A. Solomon, Development of a Nechanical Model of In-reactor Fuel Behavior: Status Report, ANL-8028 (July 1973).
9.
J. L. Routbort and J. C. Voglevede, Creep of Nixed-om *.de Puel PeIIsta at High Stress, J. Am. Ceram. Soc. 56,(6), 330-333 (June 1973).
10.
J.T. A. Roberts and J. C. Voglevede, Application of Deformation Naps to the Study of In-reactor Behavior of 0:ide Puels, J. Am. Ceram. Soc. 5_6_(9),
472-475 (September 1973).
11.
J. L. Routbort and J. C. Vog1ewede, Final-stage Densification of Nimed-oxide Puel, Am. Ceram. Soc. Bull. M(4), 363 (April 1974). ABSTRACT 12.
J. C. Voglevede, 2'hermal Densification of Nixed-o:ide Puel, Am. Ceram.
53,(8), 619 (August 1974). ABSTRACT Soc. Bull.
3 13.
J. C. Voglevede, Performance Analysis of Cache Nemory, M.S. Thesis, Illinois 1nstitute of Technology (May 1976).
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14.
C. R. Kennedy, F. L. Yaggee, J. C. Voglevede, D. S. Kupperman, B. J. Wrona, W. A. Ellingson, E. Johanson, and A. C. Evans, Cracking and Realing Behavior of UOg as Related to PeIIet-Cladding Mechanical Interaction: Interim Report July 1976, Argonne National Laboratory Report ANL-76-110 (October 1976).
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15.
C. R. Kennedy, J. C. Voglevede, and T. L. Yaggee, Cracking and Crack EcaIird of UO,, PeIIsts in Simulated IFR Pouer Cyclec, Am. Ceram. Soc. Bull. g (9),
821 (september 1976). ABSTRACT
.16.
C. R. Kennedy and J. C. Voglevede. Relocation Phenomena in UO2 PeIIets Subjected to Simulated IFR Paper Cycles, Am. Ceram. Soc. Bull. g(3), 342 (March 1977). ABSTRACT 17.
J. C. Voglevede, Application of Puei Properties Data to Out-of-Reactor Simulation Studies, Am. Ceram. Soc. Bull. g(3), 342 (March 1977). ABSTRACT 18.
B. J. Wrona, J. C. Voglevede, and T. M. Calvin. Effects of PeIIets Density and Asial Restraint on Pailure Phreshold, Trans. Am. Nue1. Soc. g, 376-377 (June 1977). ABSTRACT 19.
R. O. Meyer, C. E. Beyer, and J. C. Voglevede, Pission Gas Release from Puel at Righ Burnup, U.S. Nuclear Regulatory Commission Report NUREG-0418, (March 1978).
20.
R. O. Meyer, C. E. Beyer, and J. C. Voglevede, Pission Cas Release from Puel at Eigh Burnup, Nuclear Safety g(6), 699-708 (November-December 1978).
21.
J. L. Routbort, J. C. Voglewede, and D. S. Wilkinson, Final-Stage Densification of Niced Ocide Puels, J. Nucl. Mater. 8_0,(2), 348-355 (May 1979).
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e PROFESSIONAL QUALIFICATIONS LIST BRUCE A. BOGER F4acation June 1971 Received BSNE - Lhiversity of Virginia June 1972 Received isE - University of Virginia Work Exnerience June 1972 to Virginia Electric and Power Canoany June 1977 Surry Nuclear Power Station Assistant Endw - Performed startup testing on Unit No. 2.
E W, - Assisted the Supervisor-Engineering Services:
trained for and received a Senior Reactor Operator License.
Supervisor - Engineering Services - Directed the activities of the onsite engineering staff.
June 1977 to Virginia Electric and Power Company September 1977 Richcond, Virginia Supervisor - Nuclear Engine @ Services - Directed the activities of the offsite endn-W staff in support of Surry Power Station.
October 1977 to U. S. Nuclear Regulatory en mdesion Praent Bethesda, Maryland Reactor Engineer in the Operator Licensing Branch - Admin-ister licensing exa=inations to nuclear power plant and research reactor personnel.
Professional Affiliations Registered Professional Engineer - State of Virg h Member - American Nuclear Society
I.
PROFESSIONAL QUALIFICATICNS Peter C. Hearn I am a Senior Containment Systems Engineer in the Containment S of the Office of Nuclear Reactor Regulation, U. S. Nuclear Regulator sion.
In this position, which I have held since April 1979, I an responsib for the review and technical evaluation of containment related aspect s for PWR applications for both construction pemits and operating licenses Among the plants for which I have this responsibility are San Onofre Nuclear Po a-tion, Units 2 and 3; Joseph M. Farley Nuclear Power Station
, Unit 2; Virgil C.
Summer Nuclear Station; Commanche Peak, Units 1 and 2; Midland Plan 3
. Units 1 and 2; North Anna, Unit 2; and Offshore Power Systems.
From June 1968 to February 1973 I was employed as an engineer in th l
e Power Systems Analysis Branch. of the Naval. Ships Systems Command
, Washincton, D c-iiy reschn~sTbiTities included feasibility and performance analysis power plants including heat balances, vibrational and hydrodynamic stud' From February 1973 to April 1979 I was a Senior Auxiliary Syste i
ms Engineer in the Auxiliary Systems Branch of the Office of Nuclear Reacto r Regulation.
U. S. Nuclear Regulatory Comission.
In this position, I was responsible for the review and technical evaluation of auxiliary and power conver i s on systems for application for both construction permits and operati ng licenses.
My academic training includes a Bachelor of Science in Mechani cal Engineering from the Polytechnic Institute of Brooklyn in 1968 followed by a Mast er of Mechanical Engineering from the University of Maryland in J974
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