Acronyms 

anticipated transient without scram (ATWS)

Auxiliary Feedwater Pumps (AFWPs)

Code of Federal Regulations (CFR)   

Commonwealth Edison (ComEd)

Condensate Storage Tank (CST)

Electric Power Research Institute (EPRI)

Final Safety Analysis Report (FSAR)

Idaho National Engineering Laboratory (INEL)

local leak rate testing (LLRT)

loss of coolant accidents (LOCAs)

Nuclear Steam Supply System (NSSS)

Pressurized Water Reactors (PWRs)

shortened version of RELAPSE (RELAP)

Reactor Leak & Power Safety Excursion code (RELAPSE)

three-dimensional (3-D)

 

To Find Almost Any Acronym for Science and Technology see EDA Master Acronym List.

 

Nuclear Safety Experience

 

The experience that Mr. Miller has attained for identifying requirements of safety analyses and experiments has been developed while working for several different groups.  Some of these projects included using the TRIGA reactor at Kansas State University for neutron activation analysis, testing and research projects at a fossil fuel plant, review of testing at TLTA in California, developing and validating computer programs for use in fire migration and mitigation in reactor containments, and testing and designing systems at River Bend Station (RBS).

At Kansas State University, Mr. Miller was a Graduate Research Assistance responsible for supporting test activities using the TRIGA research reactor.  He also assisted in the fuel reload of the TRIGA and the use of the reactor for activation analyses. 

While working for Arkansas Power & Light, Mr. Miller worked as an Instrument Repairman in supporting testing, maintenance and design changes for three fossil fuel stations. 

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.   

In 1977, Mr. Miller accepted the position as Senior Engineer in Code Verification and Development at the Idaho National Engineering Laboratory (INEL).  INEL was responsible for operating the Loss of Fluid Test (LOFT) facility and the Semi-scale test facility.  Additionally, they were responsible for monitoring the testing at the Two Loop Test Apparatus (TLTA) located at General Electric in San Jose, California.  Mr. Miller’s 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.  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 Nuclear Regulatory Commission (NRC).  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.   Also, he helped develop scaling factors for the test and provided part of the testing plan for large break LOCA testing at TLTA.  

In 1979, Mr. Miller accepted a position at Black & Veatch.  He was responsible for all safety related computer programs associated with thermal hydraulics.  One example of usage and code development was in a case where transient hydrogen production was simulated and released into the containment.  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.  Black & Veatch participated in the hydrogen migration-testing program conducted at EPRI in 1981.  Black & Veatch provided pre test simulation results of various tests in the hydrogen migration test facility, which compared favorably with test results.  

In 1984, Mr. Miller was appointed Project Engineer on the pressurizer safety valve qualification project, which was part of the 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.  

In 1985, Mr. Miller accepted a position at the RBS working for Gulf States Utilities (GSU).  Mr. Miller was originally responsible for design and safety analysis of the nuclear steam supply system (NSSS).  This required the modification and specification of testing requirements for many systems during the start-up and testing phase of the plant.  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 analysis (PSA) model for the station, the development of radiation transport simulation and off-site dose calculation capabilities, loss of coolant accident (LOCA) simulation, 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. 

Mr. Miller was responsible for developing LOCA models of RBS using RELAP5 MOD2 and TRAC-B.  The LOCA simulation results from these models were compared to General Electric calculations that used SAFER/GESTER and good agreements between the results of the RELAP5 and TRAC 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 models.  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.  Other thermal hydraulic analyses were performed to determine the qualification of equipment under unusual circumstances. 

During Mr. Miller’s tenure at RBS, cracking was observed in piping that supported the Reactor Control Rod System.  This pipe cracking occurred after a scram.  Mr. Miller directed a test program to determine the root cause of the pipe cracking and a modification was proposed.  Patent No. 5,327,930 entitled “Nuclear Reactor Locking Piston Drive System and Valve Assembly” was issued to Mr. Miller for these modifications.  

Mr. Miller worked at Scientech from 1994 through 1996.  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.               

Much of Mr. Miller’s thermal hydraulic experiences dealt with the commercial nuclear industry, but most of it dealt with unusual model situations.  For example, in one problem, Mr. Miller used a computer program to simulate flow at a very low velocity through heat exchangers and out into a pool.  Many of the same challenges are inherent in passive systems, such as low velocity flow and heat transfer, heat exchanger behavior in pool stratification.   Mr. Miller also performed several natural circulation calculations using computer programs where three-dimensional mixing was apparent in the facility.  Testing was also performed at the facility to determine probable three dimensional flow patterns.  

Mr. Miller has been associated with some very complicated thermal hydraulic problems that, in many cases, required testing to resolve.  Mr. Miller has successfully identified weaknesses of modeling methods by comparing to test data.  Mr. Miller has worked with engineers and scientists to correct deficiencies and he has developed successful follow-up programs that lead to successful code development and verification.  Many papers have been written that discuss these endeavors. (See Resume List of Publications)

  back to top

Independent Reviews and safety Analysis

   

At Black & Veatch, a Consulting Engineering Firm, Mr. Miller provided many independent reviews and analyses of nuclear safety related items.  Mr. Miller was responsible for all thermal hydraulic work performed for the design of a boiling water reactor (BWR), which was in the construction permit phase of design.  All work was documented and nuclear related quality assurance including 10CFR50 Appendix B was required for all computer programs and documentation.  For each major design, a system analysis was performed.  The system Analysis consisted of presenting three to four conceptual designs.  Any of these proposed designs was shown to perform the required task.  Pros and cons were developed for each of these alternatives and cost estimates were also developed.  After reviewing all of this information, a final design was recommended.  This approach was used on all design changes that were in Mr. Miller’s responsible charge.

In 1985, Mr. Miller joined Gulf States Utilities (GSU).  He was responsible, among other activities, for the independent review of NSSS modifications to the plant during start-up and testing and for the first eight years of operation.  He was also the manager and technical lead for fuel design and safety analysis.  He was assigned the responsibility of developing analytical methods for analyzing the fuel reload at the plant.  To develop the nuclear fuel computer programs for RBS nuclear fuel group, an analysis and test program plan was developed.  Mr. Miller was the technical and management lead on this project.  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.  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.  The development of the safety analysis methods for the utility was significant in that using these methods would have save millions of dollars in fuel reload cost each year and allow the utility staff to make critical decisions with respect to the fuel.  Performing the reload safety analysis for the fuel was a major step for the utility and it was a major policy change, which Mr. Miller convinced GSU management to proceed with the development of these methods.

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 RBS, 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.

At Scientech, Mr. Miller worked with NRC staff on the review of the APEX test facility at Oregon State University (OSU) and SPES-2 AP600 test facilities.  He also reviewed 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 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 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 staff.

 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.  This was significant independent review that provided the utility, which was in regulatory trouble, the alternative to perform LOCA analyses.

 

    back to top

Risk Analysis and Application

 

At RBS, the PSA was used almost everyday 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 one of RBS's major outages, risk factors were determined before the plant was shut down to help planners lay out the outage with respect to minimizing risk.

Mr. Miller was the lead engineer and supervisor in developing the PSA at RBS.  In the 1980's, Mr. Miller's group was one of the Nuclear Power Industry leaders in using PSA for decision-making.  To develop the 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 thermal hydraulic simulators.  The CAFTA and the T/H simulator 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 RBS PSA was submitted to the NRC as the station’s individual plant examination (IPE).

An example for using the PSA 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 Specifications, they were not considered a safety threat to the plant.  The Senior Vice President, in trying to be thorough, still required a PSA to ensure that the start-up was safe.  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.  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 were anxious to start-up.  It was imperative that a good explanation of the PSA 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 PSA since it kept plant staff from putting the plant at risk.  This study, which was performed in 1988, was one of initial PSA studies that were used to make a significant decision for the plant.  This was also a turning point for the nuclear industry, which started to use PSA for non-Technical Specification decision-making.

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.  With these tools a system can be simulated to determine thermal hydraulic and structural behavior of the system.  Human factors must also be factored into the assessment.  This is done through the use of CAFTA using event trees and fault trees.  Mechanistic tools can determine the failure or fault of the system and determine when it happens and CAFTA can combine these probabilities to determine the total probability of an end point condition such as pipe failure, containment failure or core damage.  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 a comprehensive safety and risk assessment using all the elements presented above.  During a transient at the plant, high pressure core spray 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.  It was shown that this single cycle of the HPCS system did not impact CDF significantly, but another similar cycle to the system would impact the containment damage frequency and the plant PSA was changed to reflect this new 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.