2014 RETS-REMP & Groundwater Workshop, June 23-27, 2014, Westin Savannah, Savannah, GA
Sponsored by Nuclear Energy Institute


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Abstracts from 2000



Topic Name Company
Ultra-Filtration Test Plan John Kerrigan Callaway
Managing Plant Effluent During A Failed Fuel Incident Bob Artz Oyster Creek
Coolant Activity Limits and Survey Results on Performance Indicator for RCS I-131 Steve Sandike Indian Point 3
Radiochemical Challenges of Two Year Fuel - Follow-up Steve Sandike Indian Point 3
Assessment of the Molecular Separations Paul Snead Carolina P&L/EPRI
Independent Spent Fuel Storage Installation Environmental Monitoring - One Year Later Doug Wahl Exelon
Information Exchange at the International Level Greg Barley and Ken Sejkora Progress Energy
Industry Effluent Trends and Comparisons Mark Strum DE&S/EPRI
Background Radiation Study for Decommissioning Mark Strum DE&S/EPRI
Meteorological Instrumentation Performance John Chamberlain Vermont Yankee
Reactor Oversight Process Steve Klementowicz NRR
NRC "Ask the NRC" session    
Panel Discussion, topic to be determined # #
Gaseous Effluent Radiation Monitor Setpoints Tom Murphy Arizona Public Service
Gas Monitor Reading/Grab Sample Comparison Tom Murphy Arizona Public Service
Assessing Dose to Site Boundary at VY during NMC and HWC Pat LaFrate Vermont Yankee
Training for State Personnel on Environmental Sample Collection Steve Skibniowsky Vermont Yankee
Changes to the Millstone Effluent Control Program due to Unit 1 shutdown Claude A. Flory Millstone
The Millstone Effluent Control Program Claude A. Flory Millstone
Lesson learned from the Lab Transition   Teledyne Brown Engineering
Evaluation of an Agricultural Dose Pathway Jim Key Key Solution, Inc.
Containment Release Criteria John Doroski Dominion Nuclear Connecticut / Millstone Station
Lessons Learned - NRC Audits John Doroski Dominion Nuclear Connecticut / Millstone Station
Trend Analysis of 1999 and 2000 U.S. Nuclear Power Plant Gaseous and Liquid Effluent Releases Jason Harris and David Miller University of Illinois
Simultaneous Measurement and Separation Method on Beta Emitter in NPPs Goung Jin Lee and Hee Geun Kim Korea Electric Power Research Institute
Carbon-14 Monitoring in the United States Ken Sejkora Entergy Nuclear/Pilgrim Station
Comparison of Reg. Guide 1.109 and ICRP-30 Dose Conversion Factors Ken Sejkora Entergy Nuclear/Pilgrim Station
Status of the Accreditation Program for Suppliers of NIST-traceable Radioactivity Standards Daniel Montgomery Analytics
Energy Characteristics of Gamma Skyshine Radiation in the Vicinity of Boiling Water Reactors David Keefer, Ken Sejkora, and Pat Lafrate Duke Engineering and Services/Entergy/Vermont Yankee

Abstracts from 2000



Ken Sejkora

Entergy Nuclear/Pilgrim Station


Effluent release points are typically monitored with radiation-sensitive process radiation monitors (PRMs). In most cases, effluent monitor response (counts/second, counts/minute, mR/hr) is converted to a corresponding release rate (mCi/sec, Ci/day) through use of a conversion factor derived from measured monitor response to a known/measured release rate. Such applications involve ‘relative’ monitor response. In other cases, such as high-range or accident monitors, the actual monitor response (mR/hr, R/hr) is used with nomagrams or emergency conversion factors to estimate releases under emergency conditions. In these situations, the true, or ‘absolute’ monitor response is of interest.  Such monitors are periodically recalibrated on frequencies ranging from once per quarter to once per fuel cycle using a combination of electronic and radiological calibrations. An important consideration in such calibrations is linear monitor response, which is often used as an acceptance criterion. This presentation will discuss some problems encountered with PRM calibrations, including techniques to determine if a monitor’s response is indeed linear. In some cases, a monitor may exhibit a linear response, without providing an accurate true, or ‘absolute’, indication of radiation level. If such monitoring involves further application of a conversion factor to derive release rates, this lack of ‘absolute’ or true monitor response can be corrected for. In situations requiring an accurate ‘absolute’ response, additional steps may be necessary to ensure acceptable linear monitor calibration.   Required range of recalibration will also be addressed. In many cases, PRMs are used to estimate daily release rates from readings where the monitor is operating in the lower end of its response range. However, alarm/trip setpoints are typically established in higher regions of the monitor response, which may be two to three orders of magnitude (100X to 1000X) greater than normal operations. Such cases can confound calibration efforts, where the desire is to cover both normal/routine operation, as well as alarm/trip functions.



Nick Panzarino

Duke Engineering & Services Environmental Laboratory


 During October, 1999, the DOE’s Environmental Measurements Laboratory and the EPA’s Office of Radiation and Indoor Air, refereed an intercomparison of spectrometers in Grand Junction, Colorado. Our laboratory, the Duke Engineering & Services Environmental Laboratory, was one of the participants in the intercomparison.   Previous intercomparisons were limited to the infinite half-space associated with surface soil measurements. The Grand Junction intercomparison featured in situ measurements of the Walker Field Large-Area Calibration Pads.  In this presentation, we will describe the calibration pads, the protocol used in the intercomparison, the technique used to optimize the efficiency of our spectrometers for the large-area pads, and our preliminary results.


Dr. Jens Hovgaard and Bob Grasty

Exploranium G.S. Ltd.

6108 Edwards Blvd., Mississauga, Ont, Canada L5T 2V7


A new system has been developed for monitoring noble gases (Xe-133, Xe-135, Ar-41) at the fence-line around the Pickering Nuclear Generation Station (PNGS), near Toronto, Canada. The system differs from the traditional method of assessing the dose to the critical group based on meteorological models. Instead, the new system monitors the dose directly at the fence line. This new procedure has reduced the reported doses by 1 to 2 orders of magnitude. Typical annual Minimum Detectable Levels (MDLs) in air Kerma are 1.8 nGy (Xe-133), 4.2 nGy (Xe-135), and 11.0 nGy (Ar-41).  The complete system comprises seven NaI(Tl) detector units installed at the PNGS boundary. Each detector system is a stand-alone unit with internal data storage capability and automatic gain stabilization. Spectral data from 512 spectral channels are recorded every five minutes. The detectors are calibrated for Air Kerma, Ambient Dose Equivalent, Effective Dose (Adult), and Effective Dose (Juvenile).  Spectral data from the seven systems can be downloaded via telephone line to one or more central computers on a real-time basis but are normally polled each day. The data are stored on the central computers in an SQL database. Each month, the noble gas air Kerma rates and dose rates are calculated using wind information provided by PNGS. Automatic monthly reports are created with tables of the monthly air Kerma, effective dose, and MDLs from the noble gases. The reports also include plots of the air Kerma rates during the month.



Douglas Wahl

PECO Energy


Currently, the Peach Bottom Atomic Power Station easily demonstrates that it is in compliance with the requirements of 10CFR72.104 and 40CFR190 by stating in its annual dose report that doses from plant releases did not exceed two times the Appendix I limits. Per the ODCM, such a statement constitutes a 40CFR190 assessment. However, demonstrating compliance may become more difficult in the future, when the nearest real individual is expected to receive 18 mrem direct radiation from dry cask storage. This paper discusses: (1) changes to the Peach Bottom radiological environmental monitoring program to monitor dry cask fuel storage; and, (2) development of a decision criteria for interpreting direct radiation measurements to demonstrate 10CFR72.104 and 40CFR190 compliance.


Barbara Williams

Vermont Yankee Nuclear Power Station


During a NRC inspection, conducted in September 1999 questions were raised about Vermont Yankee’s program for complying with the requirements of 10CFR50.75(g). Enhancements to the 10 CFR 50.75(g) file to make the information more readily available. We have maps indicating areas of contamination keyed to a matrix that summarizes the events and contamination levels. There is a separate file for each area containing detailed information and copies of historical documents. The file is maintained by the Chemistry Department in a locked fireproof cabinet. Vermont Yankee has administratively established a threshold of 1 millirem for determining the significance of contamination. This threshold is defined as the dose to the total body and any body organ of a hypothetically exposed individual (a member of the general public or a non-occupationally exposed worker) from the potential pathways of exposure to the contamination. Although several items have been included in the decommissioning cost estimate; to date none of the areas listed in the 10 CFR 50.75(g) file has a dose exceeding this threshold.  Vermont Yankee has been and continues to be in compliance with the requirements of 10CFR50.75(g).



David W. Miller

North American Technical Center

Regional Director, ISOE NEA/IAEA


Jason Harris

NATC Research Assistant

University of Illinois at Urbana-Champaign


Gaseous and liquid effluent databases for US nuclear power plants have been under development by the North American Technical Center (NATC) since 1998. The North American Technical was established in 1995 by US, Canadian and Mexican utility radiation protection managers and is located in the Nuclear, Plasma and Radiological Engineering Department, University of Illinois at Urbana-Champaign, Urbana, Illinois, USA. Starting with the 1994, US effluent data, the NATC staff has been collecting and tabulating US nuclear power plant effluents previously performed by the Brookhaven National Laboratory under a US Nuclear Regulatory Commission research grant. The Chairman of the US Nuclear Regulatory Commission designated the North American Technical Center as the lead scientific organization to collect, tabulate, analyze and distribute the US effluent database in 1995.  This paper describes the program developed to implement the task assigned regarding US effluent databases. The latest UNSCEAR Report on Occupational Doses to Man will be discussed. Also, the formation of a North American Expert Group to advise the NATC on the effluent analysis project will be discussed.


Douglas Wahl

PECO Energy


An effluent benchmarking survey was conducted. A total of 21 plants participated. Results of the survey are presented.


Mark W. Harvey

North Atlantic Energy Services Corporation


Recent regulatory changes have resulted in an increased awareness in self-improvement initiatives. To effectively improve, a utility must be able to self identify areas for improvement. This is accomplished through effective internal and independent oversight. This discussion will address approaches to both internal (self-assessment) and independent oversight that have been successful at Seabrook Station. I will discuss methods to select topics for self-assessment, team makeup, assessment methodologies, coordination with Oversight organizations, and independent assessment. Reviewing the NRC evaluation of our plant’s ability to recognize and correct deficiencies and make program improvements will summarize the discussion.



Steve Skibniowsky and Barbara Williams

Vermont Yankee Nuclear Power Station


Challenge: To improve plant personnel response to off-normal releases of radioactive materials from plant systems. Off-normal releases of radioactive materials from plant systems are infrequent. It is, however, important to effectively mitigate the degree and impact of these releases when they do occur both from a regulatory compliance as well as from a "good neighbor" standpoint. Events have occurred where Operations personnel response was adequate but could have been improved upon with some additional training concerning evaluation methods used by RETS/REMP coordinators when faced with off-normal radioactive releases. This presentation states the problem and presents methods that can be used to improve operator knowledge in the areas of release mitigation and assessment relating to off-normal radioactive releases.

Steve Skibniowsky has been involved with RETS/REMP issues since 1973 with his first assignment at Vermont Yankee as a Chemistry supervisor. He has, since that time, provided plant support in many roles including Chemistry Department Manager, and is currently the Chemistry Instructor at Vermont Yankee. As Chemistry Instructor, he has provided effluents assessment training not only to Chemistry personnel but also to station operators and State of Vermont emergency response personnel.


Mark W. Harvey

North Atlantic Energy Services Corporation


Chemistry professionals develop skill sets that are highly valued by other departments.  
For this reason, Chemistry staff personnel frequently take other positions within the 
company.  This results in many short term issues which leaves Chemistry Managers scurrying 
address the holes left by personnel reassignment/promotion.  This discussion will address 
the long-range gains that can be accomplished as personnel are rotated or assigned to new 
positions.  I will discuss "seeding" the organization with personnel with a Chemistry 
background and the benefits that can be accomplished.  The message here is short-term loss 
results in long term gains.  Benefits can be readily seen as Chemistry personnel have moved 
to positions in Operations, Oversight, Training, Outage Management and Corrective Action.


Results of Phase I of the EPRI sponsored Tailored Collaboration Agreement

Sean Bushart, EPRI;

Louis Furlong & Gabriel Collins, MSI;

Paul Snead, CP&L


No process has been demonstrated in the US for the removal of tritium from light water reactor liquids. While tritium release from US nuclear plants is well within regulatory limits, several plant operators have expressed interest in tritium removal due to unique situations.  A new media based process by Molecular Separations, Inc. is presently being developed for the removal of tritium from liquid waste streams. Initial testing has been performed using both surrogates and process water from CP&L’s Harris Plant. The data obtained in this initial testing shows the process capable of tritium reduction in the range of 60 to 90% in one pass. However, there are a number of details relating to this technology requiring evaluation to determine the practicality of this process. As several U.S. and international utilities have expressed a strong interest in the evaluation of this technology, EPRI has initiated a project directed at providing the essential information needed to evaluate the merits of its application at a utility site. Phase I of the EPRI project involves pilot plant scale testing to evaluate the costs and efficiency of this process in LLW processing. Should the results warrant, a Phase II will demonstrate the processes in a full-scale test at a nuclear plant.This paper presents the scope of work to date in Phase I of this project. Results and work in progress will include:

  1. The quantity of tritium removed from nuclear plant liquids.
  2. The effect of common nuclear plant ionic species found in the primary coolant system on the removal media’s tritium removal performance in the processing of radioactive liquid waste streams.
  3. Key characteristics that have been defined in the testing related to the removal media process. This will include data such as optimum temperatures, flows, removal efficiency; media quantity and durability; and the media regeneration effects such as time and temperature profiles and regeneration cycles versus removal efficiency.

A short discussion of the EPRI and utility involvement in this project will be presented. This will include details of the phase II pilot testing proposed for the process water from a U.S. or international utility site to be determined. Additional opportunities for utility involvement in phase I & phase II testing will also be discussed.


John White

U.S Nuclear Regulatory Commission

Region 1


Pat Donnachie

AmerGen Energy


In 1992, a private citizen contested the Post-Defueling Monitored Storage (PDMS) plans for TMI Unit 2, the reactor damaged in 1979 accident. In September 1992 he reached an agreement with GPU Nuclear and the Nuclear Regulatory Commission to drop his opposition to the storage plans in return for several conditional activities. These activities included:

  1. The recognition of this individual's formation of a new Independent Monitoring Group, simply know as the Group.
  2. Sending all PDMS reports to the Group.
  3. Publishing all tritium monitoring results in the local media.
  4. Installation of a temperature monitoring device in the TMI-2 reactor.
  5. The monitoring of unexpected radiation levels in the TMI-2 reactor.
  6. Providing funds for robotics research.
  7. Commit to not storing any non-TMI radioactive waste at TMI.
  8. TMI-2 will be decommissioned and not refurbished as an operating facility.
  9. Disposing of accident generated water in accordance with NRC license specifications.
  10. Funding a citizens program for radiation monitoring.
  11. Funding the training of these citizens.
  12. Funding a third party to do air monitoring around TMI for a period of 7 years.

On December 31, 1999 this agreement ended. This presentation will reflect on costs of this agreement, lessons learned, benefits and radiological findings.


Greg Barley

Calvert Cliffs Nuclear Power Plant


In August 1999, tritium was detected in the effluent water of the Calvert Cliffs Nuclear Power Plant’s sewage treatment plant. While the presence of gamma emitters in sewage sludge is not uncommon in the industry, the presence of tritium in this pathway was unusual. The disbelief of the analytical results led to delays in important actions. Ultimately, some shipments of sewage sludge that contained tritium were unconditionally released from the site. While the dose consequence from this event was minimal, subsequent NRC and licensee review of controls revealed several areas for improvement. These lessons learned are important to prevention of similar events at Calvert Cliffs and elsewhere.


David Smith and Lori Glander

Consolidated Edison of New York

Indian Point 2 Nuclear Generating Station


A steam generator tube leak occurred at the Consolidated Edison Indian Point Unit 2 Nuclear Power Plant on February 15, 2000. During post ALERT recovery, a multi-disciplined committee was formed to identify and quantify related environmental releases. This presentation includes a chronology of the event, a summary of identified potential pathways, assumptions used in pathway modeling, and the results of bounding dose calculations for these pathways.


John White

U.S Nuclear Regulatory Commission

Region 1


John White, Jason Jang, Laurie Peluso

U.S Nuclear Regulatory Commission

Region 1


Panel Members

Pat Donnachie, Jr./AmerGen;

Barrie Gorman/New York Power Authority;

Al Hogan/Teledyne Brown Engineering;

Al Kaupa/BGE;

Nick Panzarino/Duke Engineering & Services Env. Lab

Moderator: Jo-Ann Pelczar, Duke Engineering & Services


Nick Panzarino

Duke Engineering & Services Environmental Laboratory


Federal regulations require the control of releases of radioactive materials from nuclear facilities and the measurement of radioactive materials in the effluents and environment outside of nuclear power plant facilities. Licensees must conduct surveys, including measurements of levels of radiation of concentrations of radioactive materials as necessary to demonstrate compliance with the regulations in 10 CFR 20.

Analysis of effluent samples, or other similar samples, split with one or more independent laboratories is an important part of quality assurance programs because it provides a means to detect errors that might not be detected by intralaboratory measurements alone. To help assure the quality of radiological monitoring measurements in client company power plants, the Duke Engineering & Services Environmental Laboratory administers an In-Plant Quality Assurance (IPQA) Program. This program is needed for the following reasons:

    1. To Identify potential deficiencies in the sample preparation and measurement process to those responsible for these operations so that corrective action can be taken;
    2. To obtain some measure of confidence in the results of monitoring programs in order to assure the regulatory agencies and the public that results are valid; and,
    3. To provide QA samples to participants.

In this presentation, we shall discuss the conduct of the IPQA program. The types of samples analyzed and how the program has benefited participants will also be discussed.


Panel Members

Bill Eakin, Northeast Nuclear – Millstone

Mark Strum, Duke Engineering & Services

Debbie Voland, Vermont Yankee

Jason Jang, NRC


Jerry F. Kinsman, Sr. Engineering Technologist

Northeast Nuclear Energy Company - Millstone Station


The current renewed industry effort is focused on reducing radioactivity in effluent discharges, specifically Liquid Fission and Activation (F&A) Curies. Millstone Station has a VISION statement to be "BEST of the BEST". One of the supporting Key Performance Indicators (KPI) for this VISION is to have less than 0.1 Curies liquid F&A Curies released for the year 2000. This is more than the current INPO top quartile for PWR.

Should it be liquid Curies or should it be public dose? That is the question!

A review of Millstone Units 2 & 3 liquid and airborne releases for the 1990 - 1999 time period shows a different perspective. Data from the Annual Radioactive Effluent and Dose Reports is charted. Review of the liquid Max. Individual/Max. Organ doses shows that, in some cases, most of the annual dose is a result of effluents from just one or two quarters. In the airborne case, review of annual Max. Individual Thyroid Airborne doses show that, in one case, the annual dose was due a discharge over a short period of time. Additional review of the Population doses show similar dose effects. The annual Max. Organ liquid Population doses are due to effluents from again just one or two quarters. The airborne doses have a similar, noticeable occurrence from the single I-131 event.

Further analysis of the supporting data from LADTAP computer runs is charted. A small percentage of Curies results in a large percentage of the Max. Individual/Max. Organ Invertebrate doses from liquids is shown to be due to one isotope, Ag-110m (activated silver). At Millstone Units, it has generally been understood that a major contributor to dose has been due to Ag-110m. This compares to the majority of Curies that come from Tritium.

Further questioning led to the possible source of the silver. The Reactor Vessel O-Ring is a silver-coated component which leaves residue on both the RV flange and the head O-ring grooves. The method of silver residue removal is considered to be a likely source. In the early 1990’s the silver residue was manually removed by dry scrubbing of the RV flange and the head O-ring grooves. In order to reduce worker doses due to this activity, it was decided to perform the RV flange cleaning with the reactor cavity flooded, utilizing SCUBA divers. In the 2000 Unit 2 refueling outage, an underwater RV flange-cleaning machine was utilized. The 2 head O-ring grooves are still manually scrubbed. A video of the 2000 RV flange cleaning operation has been made and reviewed.

Consider these issues -

Does the silver residue from the RV flange cleaning result in the Ag-110m in the liquid effluents?
Can the silver residue be captured at the source, during the cleaning operation?
How much money should Millstone stations, and the Industry, spend to remove liquid curies (0.1 Curies) in this unregulated environment?
Or can a more cost-effective way of reducing Public Dose be achieved?
Why is Silver dose high?
Site specific factors per RG 1.109
Tech. Spec.: REMP program verifies the adequacy and accuracy of effluent monitoring program
INPO influence to reduce Curies regardless of dose


Ken Sejkora

Entergy Nuclear/Pilgrim Station


Situations can arise when airborne releases are monitored from facilities which have low process and sample flow rates and limited operational run times. Such situations can result in very low sample volumes which can cause problems in achieving required gaseous lower limits of detection (LLDs) with counting criteria established for higher process and sample flow rates. When combined with limited run times at low process flow rates, a higher LLD can be argued with respect to equivalent dose impact to normal gaseous discharge pathways.

This presentation will discuss a bounding calculation approach for deriving higher LLDs to be applied in situations of limited process flow rates and operational run times. The technique involves calculation of nuclide-specific ‘Minimum Equivalent Dose’ (MED) values for a normal, high-volume gaseous release point recognized in the ODCM. The MED values are determined by coupling ODCM-required LLDs with nominal process flow rates, and calculating the corresponding offsite dose. New LLDs can then be derived for application to situations of low process flow and/or limited operational run times which would yield an equivalent MED. This bounding approach will be illustrated for airborne tritium, cobalt-60, iodine-131, and cesium-137.

Different methods of monitoring tritium in gaseous effluents will also be discussed. Since most methods of airborne tritium monitoring involve capturing and analyzing water in a liquid form, the derivation of a corresponding liquid LLD for tritium is applicable for determining whether or not the required gaseous effluent LLD is achieved. Using conservative assumptions for temperature and humidity, an LLD of 1E-2 mCi/g in condensed water will ensure meeting the required airborne tritium LLD of 1E-6 mCi/mL.



Daniel P. Dunn, Supervising Engineer-HVAC

Susquehanna SES


Troublesome problems throughout the Susquehanna Reactor Building ventilation systems were not recognized as being caused by a common problem, i.e., low airflow. Brainstorming the problem with every knowledgeable person involved centered the focus on the exhaust stack air monitor.

The exhaust stack air monitor at Susquehanna is common to all exhaust in each unit. The monitor is called the SPING which is an acronym for Stream Particulate Iodine Noble Gas monitor. The monitor utilizes an isokinetic probe assembly that, in turn, relies on two air straighteners to equalize flow across the entire duct. The straighteners, mounted inside the duct, rest in a very restricted area where access to the ductwork is physically challenging.

The Unit 1 11th refueling outage allowed us the opportunity to get in and clean the air straighteners because all ventilation could be shut down. Prior to cleaning, several modifications were installed to allow access to the components. First, a hatch was installed in the reactor building roof to allow direct access from the roof. This eliminated the need for an LCO each time the door was opened. Next, power was added to allow lights and vacuum cleaners to be used. Then ladders were added to allow outside access to the reactor building roof from the control structure roof. Finally, access ladders and platforms were installed inside the "skin" area to allow immediate access to the new duct access door locations.

As soon as ventilation was secured, work began on access doors for the duct itself. Two doors were needed along with scaffold supports within the ducts. Installation of these took 36 hours. Ventilation was restored to reduce radon buildup within the building. One week later, ventilation was secured for the second time and actual cleaning was commenced. Long-bristled vacuum cleaner brushes were utilized on the face of both straighteners. No bottlebrushes were used because there was little evidence of fouling inside the channels. In addition, the total number of openings (88,000) drove us to the conclusion that we would probably achieve 95% of our goal by face cleaning alone. Actual vacuum work took approximately 4 hours. No cleaning was done to the top of the straighteners.

The work crew attempted to clean the area below the straighteners where debris could be seen lying on turning vanes. All obvious debris was removed using extensions to the vacuum cleaners. No access to the lower area was available. Scaffold was removed and the access doors restored within 6 hours. Ventilation was restored and airflow at the SPING was recorded. Success was achieved with a 40% increase in airflow noted.

Cleaning will be completed on Unit 2 during the up-coming 10th refueling outage. All pre-outage modifications on unit 2 are already installed. Similar success is anticipated for this effort also.


David Montt and Chris Shelton

Connecticut Yankee – Haddam Neck Plant

Duke Engineering & Services Resources


A new effluent release point was identified at a decommissioned plant as a result of monitoring "clean" systems for compliance with NRC Information Bulletin 80-10. A new and independent spent fuel pool cooling (SFPC) system was filled with demineralized water, tested, and placed in service late in 1998. By mid 1999, while analyzing routine SFP and SFPC samples (collected for the IEB 80-10 monitoring program), trace amounts of Co-60 and Cs-137 were detected. As part of the program, a safety evaluation was completed to permit continued operation of the system, and an investigation initiated to identify the suspected source of the activity. The demin water storage tank had been contaminated, drained and flushed years earlier, and was the source of water for the SFPC system. Extremely low level activity (well below LLD) in the demin storage tank was concentrated to above detectable levels in the fuel pool cooling system. The design of the system is such that, when operating, a continuous release of radioactivity in the exhaust vapor occurs.

During the last 2 weeks of 1999, the Spent Fuel Island Modification was completed and implemented. Part of the modification directed exhaust from the SFP out of a new release point: the Spent Fuel Building Vent. The predominant radionuclide anticipated in this stream was tritium (though all are monitored for). The vent is in close proximity to the SFPC system intake. In late January, tritium was detected in the SFPC system. Tritium from the fuel building ventilation exhaust appears to be "inhaled," concentrated, and "exhaled" by the new fuel pool cooling system. Although the potential for contamination in this system was considered and anticipated, the sources of the contamination discovered had either originated from the least likely or unanticipated locations.



Jim Key

Key Solutions, Inc.


Environmental data routinely includes some observations which are below the limit of detection. If this censored data were randomly distributed it could simply be ignored and the remaining data analyzed in the normal fashion. A problem with censored environmental data arises from the fact that the missing data is not distributed randomly but tends to be missing at one end of the distribution. Ignoring this in the statistical treatment of the data results in biased estimates of mean and standard deviation.

Various methods for dealing with censored data sets are offered and compared. In cases where the data is heavily censored (greater than 50% of the data below limit of detection) a simple graphical method is presented.


Steve Sandike

NYPA – Indian Point 3


Discussing the shift in ratios and matrices in counting room results, the introduction of Hafnium isotopes to RCS from new assemblies, a partially failed neutron source generating Te-123m, and increased Ternary Fission.



R. Brad Harvey

Mark S. Strum

Duke Engineering & Services


Long-term average atmospheric dispersion and deposition factors based on historic meteorological data are typically generated for use within an Offsite Dose Calculation Manual (ODCM) to demonstrate compliance with Technical Specification offsite dose and concentration requirements. However, subsequent comparisons of the maximum offsite doses resulting from airborne effluents calculated via ODCM Method I (which utilized atmospheric dispersion factors based on default meteorology) and Method II (which utilized atmospheric dispersion factors based on concurrent meteorology) have sometimes revealed that doses estimated via Method II are higher than the doses estimated via Method I. This occurs even though Method I is typically perceived as the more "conservative" method. This phenomenon results from the fact that the default long-term average dispersion factors utilized by ODCM Method I do not always produce higher results for short-duration releases. This presentation will describe the generation of a revised set of ODCM Method I atmospheric dispersion factors that are a function of release duration. This revised set of atmospheric dispersion factors increases the probability that the maximum offsite doses calculated via Method I will be more conservative than the doses estimated via Method II.


North American Technical Center
Last Updated 06/12/2014

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