ML20053A862
| ML20053A862 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 05/24/1982 |
| From: | Madden F TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| To: | |
| Shared Package | |
| ML20053A827 | List: |
| References | |
| NUDOCS 8205270366 | |
| Download: ML20053A862 (18) | |
Text
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING! BOARD 3 -
In the Matter of
)
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TEXAS UTILITIES GENERATING
)
Docket Nos. 50-445 and COMPANY, -et al.
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50-446
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(Comanche Peak Steam Electric
)
(Application for Station, Units 1 and 2)
)
Operating Licenses)
TESTIMONY OF FRED W. MADDEN, JR.
REGARDING BOARD QUESTION ONE RELATED TO HYDROGEN GENERATION 01.
Please state your name, residence and educational and professional qualifications.
A1.
My name is Fred W.
Madden, Jr. I reside in Cleburne, Texas.
A statement of my educational and professional qualifications is attached hereto as Attachment 1.
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O2.
What is your current position?
A2.
I am the Lead Nuclear Engineer, Technical Support Group for Texas Utilities Services, Inc. ("TUSI").
As such, one of my responsibilities is to perform engineering and technical evaluations of plant systems related to hydrogen generation and control.
03.
What is the purpose of your testimony?
A3.
The purpose of my testimony is to describe the method of handling hydrogen gas that may be generated in the CPSES containment.
To facilitate understanding of this 18205270 34 6 O
, of this matter, my testimony is divided into two sections.
The first section describes the relevant hydrogen genera-tion mechanicms at CPSES and summarizes two analyses set forth in the CPSES Final Safety Analysis Report
("FSAR") which calculate the quantity of hydrogen that the CPSES hydrogen control systems must be designed to handle.
The second section decribes the systems de-signed to handle this amount of hydrogen.
Additional discussion of this subject is set forth in Sections 6.2.5 and 6.2.5A of the CPSES FSAR (Applicants' Exhibit 3).
I.
HYDROGEN GAS GENERATION 04.
What are the methods by which hydrogen may be generated in the CPSES containment?
lj A4.
Significant quantities of hydrogen can be generated in the CPSES containment by only four methods:
(1) a zir-coniu 2-water reaction, (2) release of the free hydrogen contained in the primary coolant system, (3) radiolysis l
of water and (4) corrosion of susceptible construction materials in containment.
FSAR $6.2.5 at p. 6.2-79 and 96.2.5A at p.
6.2-103 (Applicants' Exhibit 3).
l 05.
How are these hydrogen generation mechanisms evaluated?
AS.
In the FSAR, each of these hydrogen generation mechanisms is analyzed and combined using two independent methodolo-gies to provide the total quantity and concentration of r
, hydrogen as a function of time necessary to be considered in the design of the combustible gas control equipment at CPSES.
The two methodologies used are a Westinghouse model (discussed in FSAR $6.2.5A) and an NRC model (discussed in Regulatory Guide 1.7,
" Control of Combustible Gas Concentrations in Containment Following a Loss of Coolant Accident").
06.
What are the results of those analyses?
AG.
The results of the Westinghouse model analysis and the NRC model analysis are set forth in FSAR Figures 6.2.5A-6, 6.2.5A-7, 6.2.5A-8, and 6.2.5A-9 (Applicants' Exhibit 3).
Based on the Westinghouse and NRC models (both assume no hydrogen control equipment), hydrogen concentrations of
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8 volume percent (the concentration necessary to sustain a hydrogen deflagration throughout the containment would not be present until after approximately 100 and 75 days, respectively, had elapsed since onset of a hypothetical design basis accident ("DBA").
The two analyses, extending only to 100 days after initiation of an assumed DBA, never reach the point at which hydrogen concentrations would be in the detonable range (18-59 volume percent).
07.
What do you mean by hydrogen deflagration?
A7.
Deflagration is the propogation of a slow flame through-out a flammable mixture.
In the temperature and pressure conditions relevant here, the lower deflagra-tion limit (referred to as lower flammable limit) 0
, of hydrogen in air is 4.0% by volume for upward propo-gation.
Ignition of such concentrations would result in a very thin and momentary upward flame traveling to the top of containment or to some intermediate point obstructing further upward movement.
There is no detectable pressure rise associated with such a deflagration.
Lower deflagration limits for horizon-tal and downward propogation are about 6.5 and 8 volume percent, respectively.
08.
What do these analyses assume with respect to hydrogen distribution within the containment?
A8.
Hydrogen inside the CPSES containment is assumed to be uniformly distributed.
This assumption is supported by the outstanding mixing characteristics of hydrogen and the configuration and systems in the CPSES contain-ment.
Specifically, hydrogen mixes readily with other gases, and once mixed will not separate in the contain-l ment environment.
Mixing is promoted by convective currents created by temperature gradients in contain-ment, containment sprays, subcompartment vents and drains, and jet-stream entrainment from the assumed l
break in the primary coolant system giving rise to 1
hydrogen generation.
This assumption is discussed in FSAR $6.2.5.3.2 (Applicants' Exhibit 3).
09.
Please describe the generation of hydrogen by a j
zirconium-water reaction.
, A9.
The production of hydrogen by the reaction of water and the zirconium cladding around the fuel is described by the following exothermic chemical equation:
Zr + 2H O -+ ZrO2 + 2H2 + Heat 2
This reaction, however, proceeds in significant quantities only in the presence of very high temperatures.
Such temperature can only be achieved during a hypothetical loss of coolant accident coupled with loss of emergency cooling water from the emergency core cooling system
("ECCS").
In such a situation, the core may be exposed and excessively high temperatures may be present.
This reaction is described in more detail in FSAR % 6.2.5.3.1, 6.2.5A.1 and 6.2.5A.2.1 (Applicants' Exhibit 3).
010. What limits are imposed on ECCS design with respect to
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the zirconium-water reaction?
A10 The ECCS, a safety grade system with redundant trains and power sources, is designed to assure compliance with NRC regulations limiting zirconium-water reaction following a DBA to that associated with the reaction of 1% by weight l
of the total quantity of zirconium in the core.
j50.46(b)(3).
This is also discussed in FSAR t6.2.5A.1 (Applicants' Exhibit 3).
ECCS calculations, however, have shown that in the event of a DBA less than 0.3% of the zirconium will react.
For the hydrogen generation analyses the Westinghouse and NRC models conservatively i
assume a 2% and 5%, respectively, zirconium reaction.
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, 011. What changes have been made in ECCS operation since the Three Mile Island accident?
All. During the Three Mile Island accident a loss of coolant accident followed by operator interference with the ECCS resulted in an exposed core and excessive hydro-gen production due to a zirconium-water reaction.
Subsequent to this accident Commission directives required the development of procedures to assure that such premature operator interference with ECCS opera-tion will not occur.
To comply, procedures at CPSES wil-1 require that in the event of an ECCS initiation, operators will not terminate ECCS operation absent positive indications that the core is completely covered.
Core subcooling monitors will be installed to
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augment existing equipment and procedures, thus providing such positive indications.
This is discussed in FSAR Volume XIV, %II.F.2, Response to the NRC Action Plan Develope 6 as a Result of the TMI-2 Accident.
In addition, operators receive significant class room and simulator training in this area.
This training is discussed in the same section of the FSAR at { {II. A.2, II.B.4 and II.F.2 (Applicants' Exhibit 3).
j Q12. Please describe the generation of hydrogen by the release of free hydrogen in the primary coolant system.
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A12. The hydrogen generation analyses set forth in the FSAR assume that the maximum equilibrium quantity of hydrogen in the reactor coolant system during normal operations is immediately released into containment following a LOCA.
Such quantities include hydrogen dissolved in the primary coolant and hydrogen trapped in the pressurizer gas space.
This reaction is described in FSAR ((
6.2.5.3.1, 6.2.5A.1, and 6.2.5A.2.2 (Applicants' Exhibit 3).
013. Please describe the generation of hydrogen by water radiolysis.
A13. Water radiolysis is a complex process in which water, in the presence of radiation, is broken down into hydrogen and oxygen in accordance with the following equation.
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H 0 ;$t H2 + 1/2 O2 2
The FSAR analyses consider the only two major sources of water for radiolysis that would be present following a DBA, i.e.,
the reactor coolant inventory in the reactor coolant system and the reactor containment sump water.
Significantly, the radiolysis process is relatively slow, and is retarded by increasing concentrations of hydrogen which force a reverse reaction (i.e., combining hydrogen and oxygen to produce water).
While the Westinghouse model takes credit for reduced yield of hydrogen due to such reverse reactions, the NRC model does not.
This reaction is described in FSAR ((6.2.5.3.1, 6.2.5A.1, 6.2.5A.2.4, and 6.2.5A.3 (Applicants' Exhibit 3).
. 014. Please describe the generation of hydrogen by corrosion of susceptible construction materials.
A14. Oxidation of metals in aqueous solutions results in the generation of hydrogen gas as one of the corrosion products.
Extensive corrosion testing has been conducted to determine the behavior of the various metals used in the containment during accident conditions.
Metals tested include zircaloy, inconel, aluminum alloys, cupronickel alloys, carbon steel, galvanized carbon steel, and copper.
The results of the corrosion tests have shown that only aluminum and zine will corrode at a rate that will signifi-cantly add to the hydrogen accumulation in the containment atmosphere.
The corrosion of aluminum and zine is described by
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the following two equations:
2 Al + 3 H O
> Al O3+
"2 2
I Zn + 2H O g > Zn(OH)2 + H j
2 2
Based on the corrosion rates and the aluminum and zine inventory in the containment, the contribution of aluminum and zine corrosion to hydrogen accumulation in the containment following the design basis accident was i
calculated and factored into the FSAR hydrogen generation aralyses.
To be conservative, no credit was taken for l
protective shielding effects of insulation or enclosures l
from the spray, and complete and continuous immersion was assumed.
This reaction is described in FSAR $$6.2.5.3.1, 6.2.5A.1, and 6.2.5A.2.3 (Applicants' Exhibit 3).
, II.
HYDROGEN GAS CONTROL 015. What measures are taken at Comanche Peak to handle the amounts of hydroge. postulated to be generated?
A15. To safely handle the amount of hydrogen assumed to be generated by the four above referenced methods, re-dundant, electrical hydrogen recombiners and a backup hydrogen purge system are provided in accordance with NRC Regulatory Guides 1.7, 1.22, 1.26, and 1.29; General Design Criteria 41, 42, 43, and 50; and Branch Technical Positions CSB 6-2 and APCSB 9.2. This system is described in FSAR %6.2.5 at p. 6.2-79 (Applicants' Exhibit 3).
l 016. Please describe the design and operation of the electric hydrogen recombiners.
A16. Two redundant, electric hydrogen recombiners are provided in containment as the primary hydrogen control system.
Each recombiner has sufficient capacity to assure that containment 7:ydrogen concentration levels do not exceed 4 volume percent based on the conservative hydrogen release model set forth in Regulatory Guide 1.7.
The recombiners are safety related and designed to sustain all normal loads as well as accident loads including a safe shutdown earthquake (SSE) and pressure-temperature transients from a design basis LOCA.
Each recombiner is powered from a separate safeguards bus.
There is no interdependency between this system and the other engineered safety features systems.
In operation, O
, hydrogen is removed from the containment atmosphere by heating in the recombiner to a temperature sufficient to cause recombination of hydrogen with the containment oxygen.
The operation of the recombiners is discussed in FSAR 596.2.5.1.2, 6.2.5.3.3 and 6.2.5.4.1 (Applicants' Exhibit 3).
FSAR Figure 6.2.5-3 (Applicants' Exhibit. 3) illustrates containment hydrogen concentration as a function of time assuming operation of one recombiner started 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after initiation of a DBA.
The Figure shows that even for the conservative NRC model, hydro-gen concentration does not exceed approximately 2 volume percent, far below even the lower flammable limit for upward flame propogation.
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Q17. Please describe the design and operation of the hydrogen purge system for Comanche Peak.
A17. The hydrogen purge system, serving both CPSES containments, functions as a backup for the electric hydrogen recombiners.
Like the recombiners, the purge l
system has the process capacity to maintain hydrogen 1
concentration in the containment below 4 volume per-t cent based on the conservative hydrogen generation model set forth in Regulatory Guide 1.7.
The hydrogen I
l purge system for each containment consists of two 700 standard cubic feet per minute ("scfm') blowers t
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, for air supply, isolation valves, two atmospheric cleanup systems, and two exhaust fans.
The blowers are capable of transporting 700 scfm of fresh, fil-tered air to the containment.
The exhaust fan draws air from either containment, as required, and passes the air through high efficiency particulate and iodine filters before discharge through the plant discharge duct at 1.evels that assure compliance with 10 CFR Part 100 guideline values.
Two trains are provided for each containment, each capable of exhausting the design airflow of 700 scfm.
The system components are designed for SSE loads and maximum temperature and pressure transients from a DBA.
018. What is your conclusion regarding the design of the
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hydrogen control systems at Comanche Peak?
018. The design of the CPSES hydrogen control systems provide a high level of assurance that in the event of an accident leading to hydrogen generation, levels of hydrogen gas in the containments will be maintained below 4 volume percent, in accordance with the conservative hydrogen generation model set forth in Regulatory Guide l
1.7.
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ATTACHMENT Jl FRED W. MADDEN
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STATEENT OF EDUCATIONAL AND PROFESSIONAL QUALIFICATIONS POSITION:
Lead Nuclear Engineer, Technical Support FORMAL EDUCATION:
196B-1972, B.S. Engineering Physics, Texas Tech University 1972-1974, M.S. Nuclear Engineering,/
Purdue University EXPERIENCE:
1981 - Present Texas Utilities Servic~es,. Inc., Comanche Peak Steam Electric Station, Glen Rose, Texas, Lead Nuclear Engineer, Technical i
Support Group.
Activities include design and engineering of TMI-related plant modi-l fications; engineering resolution of li-censing issues; and development of analyt-ical capabilities.
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1980 - 1981 Texas Ut'ilities Services Inc., Dall'as, l
Texas, Licensing Engineer.
Activities
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included preparation of licensing infor-s mation such as FSAR, responses'to NRC questions, an'd interrogatories; and review and interpretat' ion of regulatory criteria.
1976 - 1980 Brown t, Root, Inc., Houston, Texas, Senior
, Licensing Engineer.
Activities included
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preparation and coordination of licensing information 'such as SAR, environmental s
reports and NRC questions; review and interpretation.of regulatory. criteria.
Coordinator of project design review tenzn following TMI accident...
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1974 - 1976 Bechtel Power Corporation,. Los Angeles, California, Engineer on. Nuclear Analysis c
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staff.
Activities include accident 1
analysis calculations; nuclear fuel cycle.
l analyses; radiation dose c'alculations;
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and shielding design and an,alysis.
Other I
project activities include. system design; l
preparation of specifications and bid...
l evaluation.
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PROFESSIONAL:
Registe:ded Professional Engineer
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(Texas and California), American Nuclear. Society, Tau Beta Pi, Phi Kappa Phi, Sigma Pi Sigma.
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P BOB C.
SCOTT STATEMENT OF EDUCATIONAL AND PROFESSIONAL QUALIFICATICNS POSITION:
Principle Quality Assurance Specialist FORMAL EDUCATION:
1961-1962, Management, University of Houston 1970-1971, Management, San Jacinto College EXPERIENCE:
1981 - Present Ebasco Services Incorporated, Principle Quality Assurance Specialist assigned to the Comanche Peak Steam Electric Station.
Responsibilities include supervision of quality Engineering personnel; review, comment and/or approval of quality Procedures / Instructions; control and distribution of Quality Procedures and Instruc-tions for Non-ASME functions; basic indoctrin-ation and required technical training for Quality Control personnel; compliance of site procurement activities to established CPSES QA requirements; and, review, processing and tracking of non-conformance reports.
197; - 1981 General Dynamics Corporation as a Senior Quality Assurance Engineer.
Assigned to the Procurement Quality Assurance Department to supervise and instruct Quality Assurance / Control Engineering Staff.
I 1977 -1979 Brown & Root, Inc. as Site Quality Assurance Manager for CPSES.
Responsible for establishing, implementing and assuring compliance of the CPSES Quality Assurance / Control Programs; managed an organization of quality engineers / technicians /
inspectors responsible for identifying quality problems, recommending or providing solutions and verifying implementation of solutions; and responsible for the control, processing, delivery and installation of nonconforming items or unsatisfactory conditions until acceptable dispositions were accomplished.
l 1976 - 1977 Brown & Root, Inc. as a Quality Assurance Super-visor for CPSES.
Responsible for Records Control,' Receiving Inspection, Nonconformance/
Corrective Action, Training, Procurement, System Turnover Verification, and Calibration; developed l
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. a Quality Assurance Instruction Program to train personnel in Quality requirements and practices.
Appointed as Quality Instruction / Consultant by the Corporate Personnel Training and Development Department to administer quality training sessions to Construction and Inspection personnel.
1974 - 1975 Brown & Root, Inc. as a Quality Assurance Audit Section Manager.
EstT.blished and implemented a comprehensive Quality Assurance Internal / External Audit Program for three (3) nuclear power plant projects.
This program included planning, scheduling, staffing, personnel training and certification, procedural preparation and audit techniques for all phases of power plant con-struction, including design, procurement and component installation.
1969 - 1974 ILC Industries as a Quality / Reliability Supervisor and a Quality / Reliability Foreman.
1968 - 1969 E.
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DuPont Company as an Assistant Manufac-turing Engineer.
1963 - 1968 Goodyear Tire & Rubber Company as a Manufacturing Technician.
1961 - 1963 Southwestern Pipe Inc. as a Quality Control Inspector.
PROFESSIONAL AFFILIATIONS:
American Society of Quality Control American Management Society
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DAVID H. WADE
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STATEMENT OF EDUCATIONAL AND PROFESSIONAL QUALIFICATIONS POSITION:
Senior Licensing Engineer FORMAL EDUCATION:
1971, BS Mechanical Engineering, University of Texas, Arlington REGISTRATION:
Professional Engineer, State of Texas
- 47622 EXPERIENCE:
1982 - Present Texas Utilities Services Inc. as Senior Licensing Engineer 1981 - 1982 Texas Utilities Services Inc. as Pro-ject Mechanical Engineering Depart-ment Head.
Supervised Comanche Peak Mechanical Engineering efforts.
1980 - 1981 Texas Utilities Services Inc. as Area Supervisor.
Supervised Mechanical Field Engineering Activities at Comanche Peak.
1978 - 1980 Texas Utilities Services Inc. as Design Engineering Supervisor for 1
the Field Support Design Group at i
Comanche Peak.
Responsible for resolutions of field interference problems.
1975 - 1978 Texas Utilities Services Inc. as Com-anche Peak Mechanical Engineer.
Respon-sible for specifi;ation and procurement of piping, valves, supports and in-line components.
1973 - 1975 Dallas Power and Light Company as Associate Engineer in the Engineering Dept.
Responsible for design and engineering of power plant systems and modifications to existing facilities.
1972 - 1973 Dallas Power and Light Company as Junior Engineer in the Plant Dept.
Responsible for plant start-up, testing, maintenance and technical assistance to operations.
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Educational and Professional Qualifications MICHAEL J.
HITCHLER WESTINGHOUSE ELECTRIC CORPORATION Carnegie Mellon University, M.S. Mechanical Engineering 1977 Lowell Technological Institute, B.S. Mechanical / Nuclear Engineering 1974 Mr. Hitchler is the Manager of the Probabilistic Risk Assess-ment Group at Westinghouse Nuclear Center.
He has current lead responsibility for a probabilistic risk study of the Sizewell B (British National Nuclear Corporation) Nuclear Station, which includes development of a risk baseline and an assessment of potential design alternatives.
He has recently worked on the Zion and Indian Points risk studies, contributing extensively in the following areas:
plant and containment event tree con-struction, systems success criteria for fault tree develop-ment, external (seismic, wind, fire, etc.) event analysis and review of the ramification sections.
Previously, Mr. Hitchler was involved in the development and implementation of strategic programs to enhance and apply risk assessment technology for use in nuclear power plant design e
and licensing.
This included development, quantification and NRC defense of event trees for use in reviewing emergency and abnormal operating procedures as part of the Westinghouse Owners' Group response to Post TMI issues.
He has assisted in the development and review of Auxiliary Feedwater System Reliability Studies for three nuclear plants.
Prior to this, his responsibilities included performing accident analyses for accidents used in licensing documents.
He has served as a Westinghouse interface with the NRC, architect engineers and utilities for issues concerning reactor protection system design requirements.
Specific areas of specialization include core and systems response to transients initiated in the pri-mary system, development of methodology for safety analysis of reload cores, and simulation of actual plant transients for computer verification purposes.
Included was the lead respon-sibility for the transfer of the above technology to various utility customers.
This included the structuring of classroom as well as on-the-job training for a number of utility personnel.
ll Mr. Hitchler is a member of the American Nuclear Society and I
the American Society of Mechanical Engineers.
He has served on two ANS Standards committees and contributed to several AIF
1
. and IEEE committees on Development of Risk Criteria and Utili-zation of PRA Approach to Licensing.
He is currently a member
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of the PRA Methodology Procedures Handbook Committee.
He is author or co-author of three reference articles, several papers and numerous project reports.
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Ed2cational and Professional Qualifications KENNETH RUBIN WESTINGHOUSE ELECTRIC CORPORATION My name is Kenneth Rubin.
My business address is Westinghouse Electric Corporation, P.O.
Box 355, Pittsburgh, Pennsylvania 15230.
I am employed by Westinghouse as an engineer in the Mechanical Equipment and Systems Licensing group, within the Nuclear Safety Department of the Nuclear Technology Division.
I am currently attending the University of Pittsburgh and will shortly receive a Bachelor of Science Degree in Applied Mathematics.
My current responsibilities include development and implemen-tation of improved radiological consequence analysis codes, evaluation of containment spray systems for fission product removal capability, corrosive effects on materials of construc-tion, and general post-accident hydrogen production.
I have performed numerous accident analyses for use in safety analysis reports and in support of operating plants and have served as a Westinghouse interface with the NRC, utilities and A/E's for issues concerning post-accident radioactivity releases and hydrogen production and accumulation.
I have provided technical support to the Westinghouse Equipment Quali-fication Program regarding containment spray chemical environ-ment.
I hold a U.S.
patent in the area of containment spray system testing.
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