Safety Assessment
·
Developed Design Basis Events in support of the
nuclear waste vitrification facility PSAR at the River Protection Project
–Waste Treatment Plant, RPP-WTP, for the Hanford National Laboratory.
·
Supported preparation of an Upgraded Safety
Analysis Report for a small DOE Nuclear Reactor. This required extensive knowledge of the
contents and requirements of regulatory documents such as DOE Order 5480.23,
DOE-STD-3009-94 and ANS-15.21-1996.
·
Performance, documentation, and QA review of
Chapter 15 transient analyses and large break LOCA analyses using the SPC PWR
licensing codes for Siemens Power Corporation.
These tasks frequently included the Safety Analysis Reports sent to the
utility clients concerning the analyses performed.
·
Completed several licensing-related tasks,
including blowdown, reflood, heatup, and fuel rod calculations using a PWR
LOCA/ECCS Evaluation Model package of codes for the Siemens Nuclear Power
Corporation (formerly Advanced Nuclear Fuels and Exxon Nuclear Company). The work produced updates to USARs for the
applicable plants.
·
Performed reanalysis of USAR Chapter 15 reactor
system transients to support core power uprate and fuel reload at the Wolf
Creek Nuclear Generating Station. This
required knowledge of Reg. Guide 1.70 "Standard Format And Content Of
Safety Analysis Reports For Nuclear Power Plants, LWR Edition" and
NUREG-0800 "Standard Review Plan For The Review Of Safety Analysis Reports
for Nuclear Power Plants, LWR Edition".
·
Technical assessment of the Seabrook Station
UFSAR Chapter 15 to verify the UFSAR has been maintained as an accurate
representation of the design basis of the plant. This was part of the response to the Oct. 9,
1996 NRC letter requesting information pursuant to 10 CFR 50.54(f) regarding
the adequacy and availability of Seabrook Station’s design bases. Potential Impacts (PIs) were prepared from a
review of source documents supporting each analysis in Chapter 15 of the
UFSAR. The information obtained was
tabulated in an ACCESS database. An
integrated assessment was then performed, and the results documented in an
Engineering report.
·
Primary technical contributions at RFETS were
CSEs associated with Solution Stabilization in
Buildings 771, 371, and 774. The
Building 771 CSEs included those for draining Tanks 83, 84, 85, 467, 551, 952,
1001, 1002, 1010, 1011, 1012, and 1081.
As part of these tasks, had responsibility for review of all associated
documents which governed the performance of specific operational activities, such
as Unreviewed Safety Question Documents (USQDs) and the formal procedures for
draining specific tanks.
·
Other important technical contributions included
assistance on the following CSEs:
Ø
Removal of resins from columns inside Gloveboxes
MT-4 and 29 of Building 771
Ø
The Carrier Precipitation Process of Building
774
Ø
The PROVE Vacuum System in Building 371
Ø
The Caustic Waste System of Building 371
Ø
The tap and drain (TAD) process for draining
tanks in Building 371.
·
Independent technical review of a natural
circulation cooldown analysis report to evaluate conformance of this design
basis analysis with the intent of NUREG-800 Reactor Safety Branch Technical
Position RSB 5-1, "Design Requirements of the Residual Heat Removal
System."
Probabilistic Risk Assessment
·
Responsible for performing several Probabilistic
Risk Analyses (PRA) for facilities and processes at the Rocky Flats
Environmental Technology Site (RFETS). These assessments provided the basis for
the development of operational safety limits for specific operational
evolutions. Specific to these assessments, developed and documented Criticality
Safety Evaluations (CSEs) whose span included operational procedures and
processes within several buildings and processes related to the Waste
Stabilization Program. Was a key member
of a CSE team providing expertise in Criticality Analysis, PRA, and Human
Reliability Analysis. The team prepared the formal CSE which included necessary
Nuclear Material Safety Limits (NMSLs, sometimes referred to as Criticality
Safety Operating Limits). The
Probabilistic Risk Analyses performed were key to ensuring the CSEs clearly
demonstrated compliance with double-contingency criteria for all credible
criticality accident scenarios.
·
During the performance of each of the
criticality safety evaluations, closely worked with operations personnel in:
Ø
Conducting joint walkdowns
Ø
Discussing each operational evolution prior to
the development of the scenarios for each CSE
Ø
Talking through and reviewing operational
procedures used during the evolutions of concern, to be sure everyone had a
realistic understanding of the tasks
Ø
Resolution of all operational concerns
associated with the final CSE
·
As a member of the team developing CSEs, became
intimately familiar with the application of the computer codes and techniques
used in the assessments. This included event tree processes used to formulate
scenarios, the HRA models used to quantify and assess the importance of human
vulnerabilities, and the KENO code used to establish limits on the amounts of
special nuclear materials. This assured
the adequacy of the contingency measures, which must be in place to assure
adequate levels of safety.
·
Participated in a Level 2 risk assessment for
the Perry Plant, responsible for review of the application of EVNTRE and NUCAP+
computer codes and overall verification of the Level II PRA.
·
Over the years, have developed, used, and/or
reviewed many plant-specific models. These include Trojan, Palisades, Haddam
Neck, Zion, H.B.Robinson, South Texas, St. Lucie, Turkey
Point, Wolf Creek, D.C. Cook, Kewaunee, Shearon Harris, WNP-2, Hatch-2, Perry-
1, River Bend, N-Reactor, the Semiscale, LOFT, EBR-II, PBF, and some foreign
reactors.
·
Provided severe accident analyses to evaluate
success criteria for various mitigating systems in support of PRA.
Nuclear Power Plant Thermal Hydraulic Analysis
·
More than twenty years experience in using and
applying nuclear plant thermal-hydraulic codes (including RELAP4, RELAP5,
RETRAN, CONTEMPT, TRAC-BF1, and the Siemens' PWR LOCA licensing codes) for both
code assessment and licensing calculations.
·
Provided model development, steady state
initialization, and transient benchmarking of RELAP5 models of BWR reactor
systems.
·
Performed thermal hydraulic analyses, primarily with
RETRAN, RELAP5, CONTEMPT and TRAC, to support licensing and operational
analytical needs for utility, government, and fuel vendor clients.
Responsibilities were also expanded to include the performance of probabilistic
risk assessments and safety analyses for clients in both private and public
sectors. Completed eighteen month assignment at the Rocky Flats Environmental
Technology Site (RFETS) as part of a PRA/Human Factors/Criticality Assessment
Team charged with establishing safe operational limits for individual processes
and activities which were part of the RFETS remediation process.
·
Utilized T-H experience in the performance of
BWR thermal hydraulic benchmarking analyses with the RELAP5 computer code.
·
Participation in two projects which developed
RELAP5 models of BWR/4 reactor systems for best estimate transient analyses
using plant FSAR, USAR, Technical Specifications, drawings, and other
information. These projects included
documentation of the system models, steady state initialization, check-out of
the control system models, operational transient analyses, and a large break
LOCA analysis.
·
Completed and documented a synthesis of data and
analysis concerning natural circulation cooling in U.S. Pressurized Water
Reactors during off-normal operation and accident transients, for EG&G
Idaho, Inc.
·
Assessment of the ability of the TRAC-BF1
thermal-hydraulic nuclear system code to calculate BWR Instability, for
EG&G Idaho, Inc.
·
Performed analyses of power reduction during a
BWR ATWS transient. This study used
space-time (1-D) kinetics and was performed to evaluate both proposed plant
operator procedures and the calculational capability of the computer code.
·
Performed plant-specific analysis to support
modification of Technical Specifications for a nuclear plant. Also developed
improved RETRAN models of the steam generators for all of the utility's nuclear
plants.
·
Assisted Florida Power and Light by performing
four analyses that were used as acceptance tests for the plant simulator. The transients were loss of feedwater ATWS, a
small-break LOCA, a main steam line break, and a steam generator tube
rupture. The RETRAN-02 models included
plant-specific best-estimate assumptions.
·
Performed several analyses of PWR small break
LOCAs, with the objective to develop small break analysis methodology.
·
Developed models for the Loss-of-Fluid Test
(LOFT) facility anticipated transient tests, performing both pre-test and
post-test analysis, with comparison to data.
·
Served as
EI interface with utilities that used the RETRAN computer program during the
code development stage. Performed
analyses to support the verification and qualification of the RETRAN code and
coauthored Volume 3 (User's Manual) and Volume 4 (Applications) of RETRAN
- A Program for One-Dimensional Transient Thermal-Hydraulic Analysis of Complex
Fluid Flow Systems, EPRI CCM-5, December 1978.
·
Assisted in development of a general
multi-compartment containment analysis program designed for analyzing
pressure-temperature transients in reactor containments, with specific
application to the Hanford N-Reactor.
·
Performed the RELAP analyses used in EI reports
EI-76-36, "Conceptual Design Study for Water Drain-Steam Vent System for
the Experimental Breeder Reactor II Steam Generator System", and EI-77-5,
"Conceptual Design of the Heat Rejection System for EBR-II Safety Research
Modification". The analyses were
used to determine optimum valve sizes and locations, piping flow areas and
piping locations, and maximum stresses on the piping.
·
Directed the analysis support of the Aerojet
Nuclear (Idaho Nuclear) - Nuclear Safety Analysis LOFT Program Planning Branch
for Experimental Test Prediction (ETP) and Experimental Operating Specification
(EOS) analyses. This included supervision
of technical employees, development of RELAP4 models of the LOFT system, the
analysis, and the documentation. Other
accomplishments included:
Ø
Assisting in the development of the NRC's Water
Reactor Evaluation Model (WREM).
Ø
Developing a two-phase frictional pressure drop
correlation for use in the RELAP computer codes, and which was used later in
the RETRAN code. This was accomplished
by substantially modifying the Baroczy correlation.
Quality Assurance and Auditor/Investigator
·
Served on the team performing a detailed audit
of EI's Quality Assurance/Engineering Procedural conformance. This was particularly important to the
company's development and sale of Emergency Response Facility Data Acquisition
Systems (ERFDAS). Was one of two staff
members who wrote the Quality Assurance Procedures Manual for use by the
Software and Consulting Group of EI.
This specifically included QA Procedures for development of the
department’s Calculation Workbooks.
·
Reviewed proposed BWR Plant Startup and
Preoperational Test Procedures for a client utility, and recommended changes
prior to plant startup.
·
Performed and documented thermal-hydraulic
analysis for the revised pressurizer PORV system for the Palisades Nuclear
Plant. Included both low temperature
overpressurization (LTOP) and once through cooling (OTC) assumptions. The studies evaluated the PORV and piping
system flow rate capacity, and resulting hydrodynamic forces on the piping
network.
·
Investigated the environmental information
available for the Hanford N Reactor, in order to produce the report, "Environmental
and Seismic Conditions for Equipment Qualification at the N Reactor". This was related to the proposed startup of
the N-Reactor.
·
Supported plant-specific audit evaluations of
the containment functional design capability for the Systematic Evaluation
Program (SEP). The work included
developing appropriate methodology for postulated secondary and primary system
breaks, and performing the analysis with RELAP and CONTEMPT codes. One final product was a Safety Evaluation
Report (SER) on "Containment Pressure and Heat Removal Capability, SEP
Topic VI-3, and Mass and Energy Release for Possible Pipe Break Inside
Containment, SEP Topic VI-2.D, for the Palisades Nuclear Power Plant".
Procedures/Proposal Development
·
Responsible for preparation and review of
numerous reports and proposals generated by the Safety Analysis
Department. This required extensive
knowledge of the contents and requirements of many regulatory documents,
typified by:
Ø
DOE
Orders 5480.22, 5480.23, 5480.24, and 6430.1A
Ø
ANSI/ANS-8
Series (e.g. 8.1, 8.7, 8.15, 8.19)
Ø
USNRC's
NUREGs for "Standard Technical Specifications For Power Reactors"
Ø
Reg.
Guide 1.70 "Standard Format And Content Of Safety Analysis Reports For
Nuclear Power Plants, LWR Edition"
Ø
NUREG-0800 "Standard Review Plan For The
Review Of Safety Analysis Reports for Nuclear Power Plants, LWR Edition"
Ø
SROM-5480.5-1, “Nonreactor Nuclear Facilities
Safety Analysis Report Format and Content Guide”.
Training
·
Instructor for RETRAN training workshops at EI
and utility offices. This intensive
instruction included presentation of RETRAN code theory, discussion of required
input, and assistance in setting up and solving problems.
Nuclear Criticality Safety
Evaluation, "Draining of Tanks D1010, D1011, D1012, and D1081",
(coauthor), Rocky Flats Plant, NMSL 940096; YSK-004, (was in progress).
Nuclear Criticality Safety
Evaluation, "Draining of Tank D-551 in Support of the Liquid Stabilization
Program", (coauthor), Rocky Flats Plant, NMSL 950043; CAA-002, (was in
progress).
Nuclear Criticality Safety
Evaluation, "Draining of Tank D952", (coauthor), Rocky Flats Plant,
NMSL 950033; YSK-003, Sept. 1995.
Nuclear Criticality Safety
Evaluation, "Removal of Resin from Columns in Gloveboxes 29 and MT-4",
(coauthor), Rocky Flats Plant, NMSL 940095; SJW-12, 1995.
Nuclear Criticality Safety
Evaluation, "Draining Tanks 83, 84, and 85 for Waste Stabilization",
(coauthor), Rocky Flats Plant, NMSL 940062; SJW-13, April 1995.
Nuclear Criticality Safety
Evaluation, "Draining of Tank D-467", (coauthor), Rocky Flats Plant,
NMSL 940037; MFS-2, August 1994.
Nuclear Criticality Safety
Evaluation, "Draining of Tanks D-1001 and D-1002", (coauthor), Rocky
Flats Plant, NMSL 940034; MFS-3, July 1994.
"Natural Circulation Cooling in U.S. Pressurized Water
Reactors",
(coauthor), NUREG/CR-5769, EGG-2653, January 1992.
"Assessment of TRAC-BF1 Using Frigg Test Data for Steady-State and
Nodalization Studies", (coauthor), EGG-EAST-8854, December 1989.
"Analysis of Thermal Hydraulic Instability with TRAC-BF1: Assessment Using FRIGG-2 Data", 1989 Stability Symposium,
EG&G Idaho, Inc., Idaho Falls,
ID, August 1989.
"Investigation of Decreasing Reactor Coolant Inventory as a
Mechanism to Reduce Power During a Boiling Water Reactor Anticipated Transient
Without Scram", (coauthor), Nuclear Technology, Vol. 70, No. 1, July 1985.
"Final Report-Enrico
Fermi Unit 2 Containment Minimum Pressure Analysis", (coauthor) Energy
Incorporated, April 1985.
"Modeling Considerations for a Hot Leg SBLOCA", (coauthor), Third
International RETRAN Conference Proceedings, April 1984.
"Methodology for the Calculation of Boron Concentration in Boiling
Water Reactors", (coauthor), Third International RETRAN Conference Proceedings, April
1984.
"RETRAN-02 Calculations of Operational Transients in the
Loss-of-Fluid Test Facility", (coauthor), Nuclear Technology, Vol. 61, No. 2, May
1983.
"RETRAN-01 - A Program for One-Dimensional Transient
Thermal-Hydraulic Analysis of Complex Fluid Flow Systems - Revision of CCM-5,
Volume 4: Applications", (coauthor), EPRI NP-2175,
December 1981.
"RELAP4/CONTEMPT
Methodology for the Palisades Nuclear Power Plant", (coauthor), EI-81-22,
Energy Incorporated, June 1981.
"Analysis of a Loss-of-Feedwater Transient in LOFT with
RETRAN", (coauthor),
First International RETRAN Conference Proceedings, September 1980.
"RETRAN - A Program for One-Dimensional Transient
Thermal-Hydraulic Analysis of Complex Fluid Flow Systems, Volume 3: User's Manual" and "Volume 4: Applications", (coauthor), EPRI CCM-5,
December 1978.
"Conceptual Design of
the Heat Rejection System for EBR-II Safety Research Modification",
(coauthor), EI-77-5, Energy Incorporated, April 1977.
"CONTEMPT-EI/M, A
Multi-Compartment Code for the N-Reactor Confinement", (coauthor), EI-76-42,
Energy Incorporated, December 1976.
"Comparison of RELAP4 Predictions with Standard Problems 1, 2, and
3", (coauthor),
EPRI NP-205, November 1976.
"Analysis of Selected Mod-1 Semiscale Blowdown Heat Transfer
Tests", (coauthor),
EPRI NP-206, October 1976.
"Conceptual Design
Study for Water Drain-Steam Vent System for the Experimental Breeder Reactor II
Steam Generator System", (coauthor), EI-76-36, Energy Incorporated,
September 1976.
"LOFT Nonnuclear Planning Analyses, RELAP3 MOD058, Period March 1973-June
1974", (coauthor),
LTR 20-25, TID-27596, November 1974.
"Nodal Sensitivity in Modeling a Large PWR System for a
Loss-of-Coolant Accident", (coauthor), Conference on New Developments in
Reactor Mathematics and Applications, CONF-710302, March 1971.
