Selected Experience Details

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Safety Analysis 

At NUS Corp., Mr. Miller developed models for containment pressurization and heat-up.  He also developed models to simulated  reactor behavior  during loss of coolant accidents (LOCAs) and for analyzing anticipated transient without scram (ATWS) events.  In addition, many other thermal-hydraulic related analyses, such transient and steady state fluid flow events, were performed. 

At the Idaho National Engineering Laboratory (INEL), Mr. Miller worked in the code verification and model development groups.  He developed input models for RELAP4 and FRAP (a fuel rod simulator) and supported code development efforts.  

At Black & Veatch, Mr. Miller was responsible for developing all computer programs used to perform safety analyses for the nuclear projects. 

At River Bend Station, Mr. Miller served in several different capacities that included the development of the ATWS model for boron injection, computer quality assurance for safety related programs, and the development of computer codes and programs used to perform engineering analyses on site.  Mr. Miller also served as a senior technical advisor and technical director on all nuclear fuel and thermal-hydraulic reactor design and analysis issues.

At Scientech, Mr. Miller provided senior technical engineering support and guidance on many technical issues and work assignments that were performed by in-house staff and by the Nuclear Regulator Commission’s research and regulatory staff.  

As the President and Principle Engineer of EDA, Incorporated, Mr. Miller provided technical consulting and Engineering on a number of issues related to reactor thermal hydraulic design and analysis for Commonwealth Edison corporate and field engineering and for the Boiling Water Owners Group (BWROG).

To develop the probabilistic risk analysis (PSA) for the power plant, precise system models of the plant were constructed using system specifications, plant drawings, plant walk downs, operator interviews and detailed analysis using CAFTA and MAAP.  The CAFTA and MAAP computer programs were used to develop accident scenarios, event trees and fault trees.  In addition to the nuclear steam supply systems (NSSSs) and balance of plant (BOP) systems, the containment systems were also simulated in the CAFTA analyses to predict the containment failure frequency.  The models were also modified to include fire risk.  From this information, maintenance and outage activities were prioritized and safety assessments were performed for special events and evolutions. The River Bend PSA was submitted to the NRC as the station’s individual plant examination (IPE). 

An example for using the IPE is presented in the following paragraph.  The power plant was to start-up from an outage when one of the two main transformers failed.  Since the transformers were not in the Technical Specification, they were not considered a safety threat to the plant.  The Senior Vice President, in trying to be thorough, still required an IPE to ensure that their start-up was safe.  The transformers were not in the IPE model so they were modeled and included in the IPE model and then risk change probabilities were determined.  As it turned out, having one transformer out when the plant started-up was risk significant and plant staff were informed of the finding.  The utility was losing $350,000 per day while the plant was down, and plant staff was anxious to start-up. It was imperative that a good explanation of the IPE process was presented.  Mr. Miller developed a presentation that explained in layman terminology the safety risk of starting the plant with only one transformer available.  There was a very robust interchange of information at the meeting, but by the end of the day it was concluded that it was risky to start-up the plant without the additional transformer. The plant stayed down for an additional three days so a new transformer could be installed.  This was an extremely beneficial use of the IPE since it kept plant staff from putting the plant at risk.

Mr. Miller was responsible for developing and implementing a plan to allow the BWRs to take the Shut Down Cooling (SDC) or Residual Heat Removal (RHR) heat exchangers out of services during the refueling of the reactor.  Fuel Pool Cooling was to be exclusively used to cool the fuel pool and reactor cavity.  Since the suction and discharge piping for fuel pool cooling was located at the top of the fuel pool, natural circulation through the reactor core had to be established to ensure adequate cooling of the fuel.  Detailed heat exchanger models were developed to perform with RELAP5 to calculate the natural circulation flow rate through the core and to establish when the SDC could be taken out of service.  Prior to the shutdown of the first unit, a testing plan was prepared that ensured proper measurement of the water temperature in different locations throughout the facility was achieved.  The validation of the analytical models was demonstrated by the successful use of natural circulation cooling in the outage.

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At NUS Corporation in Gaithersburg, Maryland, Mr. Miller was responsible for performing the first containment analysis work for the South Texas Project.  RELAP3 was the computer program of choice in performing subcompartment pressurization and heat-up analyses and CONTEMP-LT was the program used to simulate pressurization and heat up in the large dry containments.  Models were developed for each configuration.  Modifications to the input were necessary to satisfactorily model the compressibility of steam in RELAP3.  Mr. Miller also made modifications to CONTEMPT-LT source coding to include a subroutine that would simulate fires. Additionally, Mr. Miller used RELAP3B to simulate an anticipated transient without scram (ATWS) of the San Onofre Reactor.

Other example analyses that Mr. Miller performed were as follows.  Mr. Miller wrote a program that simulated the dynamic change in a tank with several different flow streams entering and exiting.  Mr. Miller also wrote a program that determined the concentration level of hydrogen gas and several gases when hydrogen is being released into the a large containment where a hydrogen recombiner was operating.

In 1977, Mr. Miller moved to Idaho Falls to take the position as Senior Engineer in Code Verification and Development at the Idaho National Engineering Laboratory (INEL).  His immediate supervisors were Tom Charlton and Dick Rice.  The laboratory was operating the Loss of Fluid Test (LOFT) facility, Semi-scale and monitoring the testing at the Two Loop Test Apparatus (TLTA) located at General Electric in San Jose, California.  His first assignment was to develop an input deck for a BWR/6 power plant using RELAP4 MOD5 and work with code development personnel to improve the jet pump modeling and the steam separator simulation.  The BWR/6 input deck was developed using original information from the Bingham Pump vendor and from General Electric, the power plant vendor.  This model simulated the BWR/6 power plants at River Bend Station and at Clinton Station.  To supplement the heat transfer models of RELAP4, FRAP-S was used to simulate the heat transfer in the 8 x 8-fuel assembly.  This work was performed for the NRC Research Staff.  In addition to this model development, several post and pre test predictions were performed for TLTA.  Several reports were written for the NRC that described the predictions.  Mr. Miller's responsibility was also to monitor the testing at TLTA, propose possible alternate testing and provide status reports for the NRC.  In addition, he helped develop scaling factors for the test and provided part of the testing plan for large break LOCA testing at TLTA.  While at INEL, some technical support was also given to the early development of RELAP5.

At Black & Veatch Consulting Engineers whom are located in Overland Park, Kansas, Mr. Miller was responsible for all safety related computer programs associated with thermal hydraulics.  Black & Veatch was the Architecture Engineer for the Black Fox nuclear power plant, which was being built in Oklahoma.  Mr. Miller was responsible for developing computer programs and qualifying them for use on the project. One example of usage and code development was in a case where transient hydrogen production was simulated and released into the containment.  A version of the MARCH code was modified and was used to simulate the BWR/6 reactor vessel during a degraded core event. Tests were reviewed for hydrogen production and results were used to validate the computer program.  To understand and respond to Utility and NRC questions on hydrogen combustion and migration a suite of computer programs was developed at Black & Veatch.  This code suite simulated the generation, burning and migration of hydrogen in the containment.  Mr. Miller was responsible for the development and testing of this code package.  A program development plan was written and a validation plan was also produced.  The code package consisted of primarily two programs, HYBRID and MIGRAINE.  The HYBRID computer program used module subroutines from three different programs, COMPARE, CONTEMPT and MARCH.  HYBRID could simulate pipe breaks and hydrogen burning in subcomparments and in Mark III containment/drywell.  It was a finite difference lump mass computation that could also track hydrogen migration.   MIGRAINE was a more sophisticated program, which used SOLA subroutines to track hydrogen migration in a three-dimensional containment structure.  Black & Veatch  participated in the hydrogen migration-testing program conducted at EPRI in 1981.  Black & Veatch provided pre test simulation results of the hydrogen migration test facility.  A presentation of Black & Veatch results was made at EPRI in San Jose, California.

To develop the nuclear fuel computer programs, an analysis and test program plan was developed.  Standard test problems such as the Peach Bottom Turbine Trip tests were placed in data banks for comparison to calculated results. Although the nuclear power plant was not a test facility, much information was obtainable from start-up testing and from the station scrams.  This information was retrieved and stored so comparisons to calculated results could be performed.  For all data stored, relevant statistical uncertainties were derived.  The nuclear fuel safety analysis test suite consisted of several well know computer programs.  These computer programs were SIMULATE 3, which was used to generate the three dimensional physics of the core, SIMTRAN, which was used to convert the 3D physics to 1D reactors kinetics, ESCORE, which was used to provide the nuclear fuel property characteristics, and RETRAN, which used input from SIMTRAN and ESCORE to simulate the reactor behavior during normal and abnormal events.  Simulation of the Peach Bottom turbine trip using RETRAN provided excellent agreement with the Peach Bottom test data.  In addition, comparison of the results from simulated plant transients to actual plant data provided good agreement.  A topical Report was written and presented to the NRC.

In 1984, Mr. Miller was appointed Project Engineer on the pressurizer safety valve qualification project, which was part of NUREG-0737 requirements.  Mr. Miller participated, on the Utilities’ behalf, in the safety valve testing at the Combustion Engineering (CE) test facility in Windsor, Connecticut.  Mr. Miller provided reviews and comments on all testing plans and witnessed several tests at the facility.  As part of his support to the Utility, Mr. Miller developed a RELAP5 model of the pressurizer, safety valve and discharge piping.  Modifications were made to RELAP5 that included a correction of the compressible gas subroutine and a correction to provide the capability to calculate forces on the piping system.  Comparing CE test results with the RELAP5 calculated results were part of the validation process.   Mr. Miller wrote a topical report on the testing and validation of the RELAP5 model and submitted it to the NRC.   All work was found to be acceptable.

Mr. Miller was responsible for buying and automating all the computers in Engineering Analysis at the RBS.  A computer plan was developed and verification and validation specifications for all software and hardware were written.  The plan consisted of developing a local area network (LAN) with many PC as clients.  Two RSIC 6000 workstations were used as servers and 6 other RSIC workstations along with approximately 30 PCs were connected to the LAN clients. This arrangement was used to support all Engineering Analysis activities on site.

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Computer programs, for which Mr. Miller prepared input decks, included ANISN, DOT III, MORSE, PDQ, NUMICE and KENO.  These computer programs were used for radiation shielding and criticality simulations.  At NUS, Mr. Miller was involved in the design of high density fuel racks for serveral different utilities.  Mr. Miller performed the following analyses: shielding calculations, the minimum fluid flow calculation through the fuel rack, the thermal hydraulic departure from nucleate boiling (DNB) calculation for the fuel, the heat load calculation using ORIGIN, and the criticality calculation.  A report was written for each design and submitted to the NRC for their review and acceptance.

At River Bend Station (RBS), several computer programs were used to perform radiation transport calculations. These computer programs included ANISN, CYLDOSE and SKYSHINE.  The off site dose was calculated using a Stone & Webster computer program.  Many analyses were provided for spent fuel shielding and radioactive pipe shielding.  In addition, criticality of the reactor and spent fuel pool during the fuel shuffle was analyzed.  Aging analyses were also performed to extend the life of qualified components.

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In 1985, Mr. Miller accepted a position at the River Bend Station working for Gulf States Utilities.  In addition to several other areas of responsibility, Mr. Miller was responsible for developing Engineering Analysis at the Station.  The Engineering Analysis Group was responsible for the following: the design and specification of the fuel for each reload, the development of the probabilistic safety assessment (PSA) model for the station, the development of radiation transport simulation and off-site dose calculation capabilities, loss of coolant accident (LOCA) simulation and thermal hydraulic simulation of the power station reactor and systems, and the development of a local area network (LAN) that tied all the thermal hydraulic and nuclear safety related computer programs and PCs to one server.

 Thermal-Hydraulic Analyses

A LOCA model of RBS was developed using RELAP5 MOD2.  The LOCA simulation results from this model was compared to General Electric calculations using SAFER/GESTER and good agreements between the results of the RELAP5 analyses and the results of the SAFER/GESTER analyses were achieved.  Unusual accident scenarios were also simulated using RELAP5 to provide input into the PSA model.  Other thermal hydraulic models of the reactor and associated systems were developed using RELAP5 MOD2.  These models were used to simulate many different scenarios that occurred or potentially could occur at the plant so questions by plant staff and oversight groups could be answered and solutions to problems could be provided.  Other thermal hydraulic analyses were performed to determine the qualification of equipment under unusual circumstances.  Other computer programs used to support these calculations were COMPARE, CONTAIN, LOCVS, GOTHIC and CONTEMPT.  To use these programs efficiently, an understanding of the thermal hydraulic principals, theories and applications of the computer programs was required.  Additionally, an understanding of the reactor, associated systems, and the containment system was required.

Most of Mr. Miller's direct experience related to a new reactor design was in the evaluation of the AP600 reactor design.  In evaluating this design, Mr. Miller was intimately familiar with the test models at OSU and SPEC-2.   The test hardware and systems must be evaluated to determine were abnormal thermal hydraulic behavior may occur.  Mr. Miller has developed these skills over the years by analyzing many different types of systems over a spectrum of pressures from 2500 psia to 0 psia and temperatures that included saturated, subcooled and superheated fluid conditions.  Mr. Miller has evaluated heat transfer for pressurized heat transfer surfaces at high velocities to natural circulation flow heat transfer surfaces at low velocities.  

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From 1985 to 1994, Mr. Miller supported River Bend Station staff in resolving many different plant problems.  He was located on site and the projects ranged from construction and design to analysis and licensing.

In 1994, Mr. Miller joined Scientech as a Senior Technical Advisor. His work consisted of several projects for commercial utilities and the remaining tasks dealt with supporting the NRC Research and Regulatory staffs.

In 1996, Mr. Miller started his consulting engineering firm and his largest client was Commonwealth Edison in Chicago, Illinois.  Mr. Miller began this work with Commonwealth Edison by evaluating an issue of the Auxiliary Feedwater Pumps (AFWPs) for Byron Units 1 & 2 and Braidwood Units 1 & 2.  Mr. Miller modeled the AFW system from the Condensate Storage Tank (CST) to the Steam Generators using the computer program RELAP5.  With this model, Mr. Miller was able to assess the root cause of the problem and propose a modification.  Once the modification was implemented, the issue was resolved.  This effort required him to walk down the systems and coordinate information from the site System Engineers.  Mr. Miller attended several meetings with plant staff to discuss results and to evaluate alternatives.

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Several projects were completed for the NRC.   One project consisted of performing a review for the NRC  staff on ASME code requirements of check valves for PWRs and BWRs.  Another project consisted of reviewing all power uprate submittals and Safety Evaluation Reports and identifying technical areas that were not adequately reviewed and discussed.  Mr. Miller was also the senior technical consultant for the NRC on two separate engineering inspections at nuclear power plants.  Reports were written and presentations were made for each of these projects.

Mr. Miller worked with NRC Research staff on the review of the APEX test facility at Oregon State University (OSU) and SPES-2 AP600 test facilities and on the review of Westinghouse licensing submittals.  Mr. Miller wrote a report comparing the RELAP5/MOD3.2 input decks of the INEL and ANSALDO.  Anomalies between the models and input decks were noted.  Cases were also run  using each of these input decks.  For another project, concerning the advanced reactors, Mr. Miller performed a reviewed of the OSU AP600 test facility and performed comparisons of calculated results using RELAP5/MOD3.2 and NOTRUMP.  A report detailing the calculations of OSU test SB11 was provided to the NRC research staff.  The APEX test SB-11 was analyzed using RELAP5 Mod3.2.  The APEX test facility at OSU was used to simulate thermal hydraulic phenomenon for the AP600 passive safety systems for loss of coolant accidents (LOCAs) and long term cooling.   Another test, SB-12, was modeled and simulated with RELAP5 Mod3.2.  The key experimental parameters measured were compared to RELAP5 simulation results.  A report was written and the results were presented to the NRC Research staff.

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At River Bend Station (RBS) teamwork and consensus building was imperative to successfully managing the onsite Engineering organization. Most work groups consisted of Maintenance, Operations, System Engineering and Engineering with oversight groups also participating. The oversight groups consisted of NRC Inspectors, Quality Assurance and Control Staff, the Facility Review Committee, and the Senior Oversight Committee.   All of these groups had diverse backgrounds and different reasons for participating the projects.  Usually when  a significant problem occurred, the primary groups, i.e., Operations, Maintenance, System Engineering and Engineering would hold a meeting to evaluate the problem and develop a root cause for the problem.  In many cases, Engineering would take the lead and Mr. Miller, as one the lead engineers, would provide leadership in identifying main personnel contributors for the meeting and organizing the meeting.  Since most of these groups had their own "bones to pick", it was really important to keep the meeting on track and to ensure the appropriate individuals provided critical information, e.g., system performance, collaborating test data and analysis, and maintenance and operation history.  It would generally take several meetings to develop a satisfactory root cause and propose possible design changes to fix the problem.  It was always essential that a consensus of the participants at the meeting was reached and all agreed on the root cause and the corrective actions (usually a design modification along with procedures and other document changes).  The meeting notes were drawn up and all participants concurred on the conclusions and recommendations.  Some of these projects were very costly and required more than a year to complete and it was crucial that all major groups in the company agreed with the proposed work schedule.

Another example, in which Mr. Miller was involved in regulatory affairs, occurred when Mr. Miller was asked to be the lead technical consultant for the NRC Engineering Inspections at selected plant sites.  This required a meeting with NRC representatives from Regions III and IV at a plant site and as a team, perform an Engineering Inspection of site engineering.  The inspection team met everyday for three weeks and discussed issues that surfaced during the day.  Plant staffs were also involved in these discussions.  It was crucial that the team concurred on most findings, which were presented to plant management at the end of the third week, and later to the NRC regulatory staff.  There were several times that the group disagreed on an issue, but, as a team, were able to resolve them successfully.

Mr. Miller, as an engineering leader at River Bend Station, was responsible for plant operation 24 hours a day and 7 days a week.  Many times, Mr. Miller would be called at 3 A.M. to report to the plant where a group of people gathered to solve a problem that could shut down the plant.  Some of these team members did not know each other very well, but it was necessary that these team members worked together and resolve the issues.

At River Bend Station, Mr. Miller was the emergency preparedness Technical Support Manager (TSM) for emergency drills.  This position required excellent communication among all groups associated with the drill.  These groups included the NRC, civilian authorities, company executives and the staff in the technical support center, included maintenance, operations, systems engineering and design engineering.  In these drills, decisions were quick and all communication had to be clear and succinct.

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In 1977, Mr. Miller moved to Idaho Falls to take the position as Senior Engineer in Code Verification and Development at the Idaho National Engineering Laboratory (INEL).  His immediate supervisors were Tom Charlton and Dick Rice.  The laboratory was operating the Loss of Fluid Test (LOFT) facility, Semi-scale and monitoring the testing at the Two Loop Test Apparatus (TLTA) at General Electric in San Jose, California.  His first assignment was to develop an input computer deck for a BWR/6 using RELAP4 MOD5 and work with code development to improve the jet pump modeling and the steam separator simulation.  The BWR/6 input deck was developed using original information from the Bingham Pump vendor and from General Electric in San Jose.  This model represented the BWR/6 plants at River Bend Station and at Clinton Station.  To supplement the heat transfer models of RELAP4, FRAP-S was used to simulate the heat transfer in the 8 x 8-fuel assembly.  This work was performed for the NRC Research Staff.  In addition to this model development, several post and pre test predictions were performed for TLTA.  Several reports were written for the NRC that described the predictions.  Mr. Miller's responsibility was also to monitor the testing at TLTA for the NRC, propose possible alternate testing and provide status reports.  In addition, he helped develop scaling factors for the test and provide part of the testing plan for large break LOCA testing at TLTA. 

Mr. Miller developed testing models and reviewed testing for Westinghouse's AP600 test program.  The OSU test facility was used to simulate thermal-hydraulic phenomenon for the AP600 passive safety systems for large and small break loss of coolant accidents and long term cooling.  During those evaluations, Mr. Miller developed RELAP5 models that simulated the testing performed in the APEX test facility at OSU.  The OSU experiments examined the passive safety systems response for small and large break LOCA transition into long term cooling.  The facility permitted a range of small-break LOCAs to be simulated at different locations on the primary system, such as the cold leg (CL), hot leg (HL), cold leg pressure balance line (PBL), and direct vessel injection (DVI) line. During the thermal hydraulic simulation using RELAP5, the thermal hydraulic behavior of the passive systems and the core interaction was observed.  For each of the tests, the test data was reviewed for consistency and reasonable behavior and then the RELAP5 calculated results were compared to the test data.  In many cases, there were differences between the test results and the calculated results.  When major differences occurred, both the experimental and the calculated results must be reviewed to determine abnormalities.  For example, if the measured water level in the core was much lower than the calculated water level, other measured parameters such as inventory and mass balances were checked for consistency with the water level to ensure that the measurement was correct.  For the analytical calculation, the inventory balance was also be checked along with flow rates into and out of the reactor vessel.  For the passive, low velocity, and natural circulation code assessments, it was found that in many instances, natural circulation and stratification were inherent in the tests.  Since most of the LOCA codes were based on high velocity flow, the code assessment was focused on the possible weaknesses of the computer program and new code development was proposed to improve those potential weaknesses.

In one project for a commercial utility, Mr. Miller was responsible for reviewing the testing results of a utility that tested a proposed water level backfill modification.  During their testing of the modification, some anomalies were observed and the utility required a review to be performed of the testing plan, uncertainties, scaling factors and test results to determine if there was a problem.  A thorough review was provided that also included a RELAP5 model of the test facility, which was used to  simulate the testing and determine the root cause of the anomalies.  It was found from this simulation that a small check valve in the water level instrumentation line caused the abnormalities and that this would not occur in the actual design modification.

On another project, Mr. Miller developed a safety valve dynamic model that was used to evaluate the operability of the pressurizer safety valves at a PWR plant.  For another project, Mr. Miller was the lead technical review engineer for the evaluation of three vendors that were selected to perform small break LOCA.  An inspection trip was taken to each of the vendor locations and a thorough review of their small break LOCA methods was conducted.   This review included the licensing of the methods, of the method development, of the testing that was conducted to validate the programs, of the error reporting, and of the unresolved technical issues associated with their computer programs. After these reviews were completed, a recommendation of a vendor, which met the needs of the utility, was made and a presentation was provided to the utility board members.

One of the projects at RBS was facilitating the testing of a critical component at the station.  A modification of the control rod drive (CRD) system in the plant was planned and a thorough test of the changes was required before the installation.  During the life of River Bend Station, cracking was observed in the CRD piping leading to the CRD accumulator tanks.  After a reactor scram, the CRD accumulator tanks are refilled from water taken from the condensate demineralizer water tank (CDWT).   The CRD pumps took the water from the CDWT, pressurized the water to 1500 psia and transported the water through piping and check valves into the accumulator tanks.  After a number of scrams, it was observed that the piping leading to the accumulator tanks was cracking.  A root cause evaluation determined that the ball check valves in the system piping were chattering during refill and that dynamic vibration of the chattering ball check valves caused the cracking of the associated pipes.  New check valves were designed, but before installation, testing of these valves in the system was required.  A test plan and testing program were developed that included design drawings, construction drawings, uncertainties, scaling factors and a test matrix.  The testing skid was constructed and testing was completed.

Mr. Miller has been associated with some very complicated thermal hydraulic problems that, in some cases, required testing to resolve.  In some of the thermal hydraulic analytical work Mr. Miller has been involved, Mr. Miller has successfully identified weaknesses of the test and of the computer programs used to simulate the tests.  Mr. Miller has worked with engineers and scientists to correct deficiencies and Mr. Miller has developed successful follow-up programs that lead to successful code development and verification.

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Evaluations of Nuclear Plant Safety Issues

At NUS Corp. in 1975, Mr. Miller developed models for containment pressurization and heat-up.  He also developed models to simulated reactor behavior during loss of coolant accidents (LOCAs) and for analyzing anticipated transient without scram (ATWS) events.  These were some of the initial analyses performed to address ATWS and containment pressurization concerns.

At NUS in 1976, Mr. Miller developed safety analysis and licensing submittals for some of the first high-density fuel rack designs.  Safety analyses were provided for heat generation analyses using ORIGIN, seismic calculations and criticality analyses.  Thermal hydraulic calculations were performed using basic equations for natural circulation, parallel flow and DNB correlations.  Licensing submittals were reviewed and approved by the NRC.

At River Bend Station (RBS), the probabilistic safety analysis (PSA) was used almost daily and Mr. Miller was the responsible manager and technical lead for its development and use.  Risk assessment was used in fire protection issues, miss-oriented bundle accidents, stability, HVAC issues, and licensing issues.  An outage risk scenario was developed to help outage planners take systems out of service with respect to risk.  During several of RBS's outages, risk factors were determined before the plant was shut down to help planners lay out the outage with respect to minimizing risk.

A comprehensive safety and risk assessment enrolled many tools from engineering analysis.  The mechanistic assessment must be conducted to determine the physical restraints and limitations of the system under investigation.  Mechanistic tools used were RELAP5, RETRAN, KYPIPE, NASTRAN and ADLPIPE.  Human factors must also be factored into the assessment.  This is done through the use of CAFTA using event trees and fault trees.  Sometimes it is important to determine the probability of a pipe failure just to determine how credible the event is.  Mr. Miller was the technical lead on many comprehensive safety and risk assessments using all the elements presented above.   Examples are given below.

The power plant was to start-up from an outage when one of the two main transformers failed.  Since the transformers were not in the Technical Specifications, they were not considered a safety threat to the plant.  The transformers were not in the PSA model so they were modeled and included in the PSA model and then risk change probabilities were determined.  Based on the new evaluation, having one transformer out when the plant started-up was risk significant and plant staff were informed of the finding.  It was concluded that it was risky to start-up the plant without the additional transformer.  The plant stayed down for an additional three days so a new transformer could be installed.  This was an extremely beneficial use of the PSA since it kept plant staff from putting the plant at risk.  This study, which was performed in 1988, was one of the initial PSA studies that were used to make a significant safety decision for the plant.  This study helped to provide a benchmark for the nuclear industry, which began using PSA for non-Technical Specification decision-making.

During a transient at the plant, high pressure core spray system (HPCS) injection was initiated.  Subsequent to this event, the motor operated valve (MOV), an isolation valve, was inadvertently opened when the reactor was at full pressure.  The other isolation valve, which is a check valve, should have closed against the back flow.  Based on test information taken during the event, it appeared that the check valve stuck open for a short period of time.  Since this system is designed as a cold-water high-pressure system, the hot water high-pressure situation represented a potential interfacing LOCA.  Since the MOV was opened inadvertently and the check valve stuck open for a period of time, it was necessary to determine how this changed the core damage frequency (CDF) of the plant.  Using mechanistic methods, it was determined how far the hot water proceeded into the HPCS.  Using this information, a structure analysis was performed to determine the impact on the structural integrity of the system.  This information along with other human factor information was integrated into the plant PSA to impact on CDF.  Since this was a thermal stress cycle to the HPCS that was not originally factored into the plant PSA, the plant PSA was also changed to include this condition.  This assessment was significant since it showed plant staff how important it is to keep check valves in good working order and that an event can change the CDF of the plant.

At River Bend Station, Mr. Miller served in several different capacities that included the development of the ATWS model for boron injection, computer quality assurance for safety related programs, and the development of computer codes and programs used to perform engineering analyses on site.  These evaluations were conducted to satisfy parts of the ATWS rule making, 10CFR50.62.  Through these evaluations, input for emergency operating procedures (EOP’s) and design specifications for high-density boron in the boron injection system were established and implemented.

At Scientech, Mr. Miller provided senior technical engineering support and guidance on many technical issues and work assignments that were performed by in-house staff and by the Nuclear Regulatory Commission’s (NRC’s) research and regulatory staff.  Several reports were written, which included AP600 testing evaluations, check valve analyses, and an evaluation of all power uprates performed up to 1996.

On another project, Mr. Miller developed a safety valve dynamic model that was used to evaluate the operability of the pressurizer safety valves at a PWR plant.  For another project, Mr. Miller was the lead technical review engineer for the evaluation of three vendors that were selected to perform small break LOCA.  An inspection trip was taken to each of the vendor locations and a thorough review of their small break LOCA methods was conducted.   This review included the licensing of the methods, of the method development, of the testing that was conducted to validate the programs, of the error reporting, and of the unresolved technical issues associated with their computer programs. After these reviews were completed, a recommendation of a vendor, which met the needs of the utility, was made and a presentation was provided to the utility board members.

Mr. Miller investigated the effect of water hammer on Fan cooler units (FCUs).  FCU are installed in PWRs to cool the containment during normal operation and, for some designs, to cope with the design-basis loss-of-coolant accident (LOCA) or main steamline break (MSLB) event simultaneous with a loss of offsite power (LOOP). The NRC issued GL-96-06, in part, to address the above concerns. The GL referenced NUREG/CR-5220, which provides a conservative approach for evaluating waterhammer events. Loads that were calculated based on EPRI methodology were compared to loads calculated by RELAP5.  The EPRI methodology includes an analytical model of the closure of an air and steam pocket, which could cushion the impact of the incoming water slug. The EPRI model hypothesizes that the pocket contains dissolved air that is released during boiling, and steam that is left uncondensed before the waterhammer event occurs.   The loads calculated using RELAP5 were slightly more conservative than the loads calculated using the EPRI methodology.  These loads were used to analyze structural integrity of the FCU piping for two power plants.  A report was issued to the NRC.

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Nuclear Licensing

Mr. Miller has been involved in many areas of nuclear regulatory oversight and has given effective communicative presentations and written documentation for many of the significant issues since 1974.  Mr. Miller is familiar with many NUREGs, Bulletins, Information Notices, SECY documents, ACRS documents, Generic Safety Issues, Standard Review Plans, Regulatory Guides, Branch Technical Positions and 10 CFR 50 documents that govern the nuclear industry.  He has addressed most of the major nuclear power plant licensing issues since the early 1970’s.  Some examples of these issues are provided in the following paragraphs.

At NUS in 1976, Mr. Miller developed analytical and licensing submittals for some of the first high-density fuel rack designs.  Licensing submittals were written for 10 fuel rack designs.   RG 1.13, "Spent Fuel Storage Facility Design Basis," and NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants was applied in these evaluations. Quality-assurance program documents for all rack designed were in compliance with 10 CFR 50, Appendix B and ANSI N18.2.  These submittals were reviewed and approved by the NRC.

At NUS, Mr. Miller developed licensing submittals for containment and subcompartment analyses for the South Texas Project.  Original subcompartment analyses were performed using RELAP3.  RELAP3 was validated against the standard subcompartment problems provided by the NRC.  Mr. Miller was involved in the first containment fire analysis.  These evaluations conformed to Standard Review Plan (SRP) Chapter 6 (NUREG–0800 and NUREG-75/087), 10CFR50 Appendix B & R and applicable codes and standards.

In 1983, Mr. Miller provided technical and licensing leadership for the compliance with 10CFR50.44 "Standards for combustible gas control system in light-water cooled power reactors”.  Igniter were designed and located to accommodate the hydrogen generated in the containment and limit the hydrogen concentration in containment below 10-volume % in compliance with 10CFR50.34(f) requirements.  Analyses were performed and licensing submittals were provided.   Mr. Miller developed the technical and licensing submittals for excess hydrogen management, which included the quantity and location of igniters in a BWR containment.

On behalf of a utility client, Mr. Miller was responsible for resolving NUREG 0737 issues with regard to safety and relief valve functionality, TASK II.D.1 - PERFORMANCE TESTING OF BOILING-WATER REACTOR AND PRESSUREIZER-WATER REACTOR RELIEF AND SAFETY VALVES (NUREG-0578, SECTION 2.1.2).   Mr. Miller observed testing at the safety valve test facility in Windsor, Connecticut.  He was responsible for the project that provided the validation of the loop seal safety valves to function as designed.  He developed a RELAP5 model of the test configurations and compared results from tests 917 and 908 to calculated results from RELAP5.  He also modeled a safety and relief valve of the power station and developed loads for ASME Class I piping.