One of the duties of the RETS/REMP Steering Committee is to develop positions on various questions that arise with regard to the implementation of RETS and REMP. The steering committee has received a request for a position with regard to the sampling requirements of Table 4.11-2 of NUREG-1302. The position deals with the compensatory requirements for sampling of the gaseous effluent streams following startup, shutdown, and thermal power changes exceeding 15% in one hour. This presentation is designed to communicate to the workshop attendees the proposed position and to solicit comments in order to facilitate the development of the final position
During an NRC inspection of the HNP Radiological Environmental Monitoring Program (REMP) program, the NRC inspectors raised two concerns. The first concern was that the flow rates of some of the air samplers occasionally exceeded the vendor recommendation, resulting in a decrease in iodine collection efficiency, which was not taken into account in the analysis. 10 CFR 50.36a requires that procedures be in place for sampling and analysis of environmental samples. It is understood that these procedures will be technically adequate so as to ensure that accurate results will always be reported. The operation of air sampling pumps at elevated flow rates, coupled with the lack of procedural guidance to ensure that corrections are made for charcoal collection efficiency, presented a potential for inaccurate results to be reported, had any radioactive iodine been detected in the environmental air cartridges. The second concern was with the cumulative unavailability of the air samplers. Frequent environmental air sample pump malfunctions were occurring, resulting in decreased availability of the air sample pumps. This presentation will describe the causes and lessons learned from the inspection, as well as share some benchmark information from the industry.
This paper will present how Boiling Water Reactor (BWR) and Pressure Water Reactor (PWR) Nuclear Power Plants have successfully reduced their liquid curie discharge through the use of membrane based and media based wastewater processing systems. Effluent radioactive wastewater results, from several Nuclear Power Plants in the United States, using membrane and media based technologies for BWR's and PWR's will be presented and discussed. Membrane based technologies using reverse osmosis (RO) allow BWR's to achieve "zero" curies released to the environment and media based technologies using demineralization, allow PWR's to minimize their curies released and obtain "top quartile" performance ratings.
A brief description of the membrane and media based liquid waste processing technologies will be presented including typical equipment requirements and descriptions. A collection of empirical Nuclear Plant wastewater processing data such as gallons processed, water chemistry, influent curies and curies released will also be presented.
Information provided in this paper will specifically benefit two groups of nuclear facilities: (1) those operating under a zero or minimum liquid discharge scenario, and (2) those facilities that require less than minimum detectable activity (<MDA) concentrations of gamma emitting isotopes in the water discharged to the environment.
Elevated tritium levels at Perry have created challenges to plant effluents this past cycle. The primary cause for the increase can be attributed to control rod blade leakage and fuel cladding defects. In an effort to maintain a near zero discharge policy tritium was not diluted by water makeup. This allowed tritium to rise to a level that created challenges for plant effluents. The first challenge was that tritium was detected in the turbine building supply plenum drains. Air coming into the Turbine Building is cooled with chilled water. The condensate from these drains are then routed to storm drains during hot weather operation. The investigation for this event determined that a small amount of effluent from plant vents were recycled back into the turbine building supply plenums during the summer. The low levels of tritium in the effluents were then condensed in the drains creating a release of tritium to the environment. The second challenge was that the increase in tritium concentration required a significant reduction in radwaste discharge flow rate in order to ensure compliance with the ODCM methodology for maintaining effluent concentration within limits. The reduction in radwaste discharge flow could have had an impact on the outage water management plan if not properly controlled. This presentation will outline these challenges to communicate the need to have proper controls established for tritium in a BWR.
Hydrated vs. Dehydrated Silver
The USNRC issued an Event Description Bulletin 50-409 describing an off gas explosion at the LaCrosse NPP in 1986 that resulted from the use of a silver zeolite cartridge during an air sampling activity.
F&J SPECIALTY PRODUCTS, INC. undertook an investigation of the cause of the explosion and evaluated the differences in performance of hydrated vs. dehydrated silver zeolite adsorbent in radioiodine collection cartridges. This investigation yielded results that confirmed the acceptable material specifications for a product that would not cause an explosion when utilized in sampling an air stream containing a hydrogen gas concentration greater than 4% by volume in the presence of oxygen.
Ginna Station is a 490 net MW Westinghouse PWR located in western New York State on Lake Ontario. It has been in commercial operation since September, 1969. The Main Condenser was retubed in 1995, the Steam Generators were replaced in 1996, and the Reactor Vessel Head is scheduled for replacement in 2003. An application for License Renewal was submitted to the NRC in 2002. Ginna Station is owned by Rochester Gas and Electric Corporation, which is a subsidiary of Energy East Corporation.
Prior to implementation of standard technical specifications, in accordance with the guidance of NUREG-1431, ODCM requirements were taken from the technical specifications and placed into a chemistry administrative procedure. The ODCM existed as a chemistry administrative procedure until sugestions from the NRC led us to rewrite it to the guidance of NUREG-1301 and place it into Technical Specification supporting documentation along with our TRM (Technical Requirements Manual) and COLR (Core Operating Limits Report). In this presentation I will provide an historical timeline, some historical differences between our ODCM and NUREG-1301, some difficulties that resulted, and the resolution.
Issues described include:
- Weaknesses of the administrative procedure, including format and controls.
- Examples of events in which this weakness surfaced
- Benefits of the standard terminology of NUREG-1301
- Proper administrative "level" of the ODCM
- Improvements at Ginna Station as a result of the change to the ODCM
HVAC (heating, ventilating, and air-conditioning) is a common term that can include cooling, humidifying or dehumidifying, or otherwise conditioning air for comfort and health. HVAC also is used for odor control and the maintenance of acceptable concentrations of carbon dioxide.
Air-conditioning has come to include any process that modifies the air for a work or living space: heating or cooling, humidity control, and air cleaning. Historically, air-conditioning has been used in industry to improve or protect machinery, products, and processes. The conditioning of air for humans has become normal and expected. Improved human productivity, lower absenteeism, better health, and reduced housekeeping and maintenance almost always make air-conditioning cost effective.
Mechanical air-handling systems can range from simple to complex. All distribute air in a manner designed to meet ventilation, temperature, humidity, and air-quality requirements established by the user. Individual units may be installed in the space they serve, or central units can serve multiple areas.
The presentation will be an overview of HVAC design, construction, and operation with regard to Indoor Air Quality (IAQ).
Areas of importance will include;
Equipment selection as it applies to Control Room habitability
Operation of HVAC equipment;
Trouble shooting; reading a fan curve
Air flow from areas of lesser contamination to higher contamination
Challenge testing the occupied envelope; ASTM E-741 and Perfluorcarbon Tracer gas testing (PFTs)
IAQ case study outside of Nuclear
Selected sections of the presentation are highlighted below.
Equipment selection as it applies to Control Room habitability
Considerations in designing an air-handling system include volume flow rate, temperature, humidity, and air quality. Equipment selected must be properly sized and may include:
* outdoor air plenums or ducts
* supply fans and supply air systems
* heating and cooling coils
* humidity control equipment
* supply ducts
* distribution ducts, boxes, plenums, and registers
* return air plenums
* exhaust air provisions
* return fans
* controls and instrumentation
Operation of HVAC equipment;
WALK AROUND INSPECTION. NIOSH has determined that inadequate ventilation is the main problem in 52% of their IAQ investigations. Therefore, ventilation surveys should be initially conducted. During the walk around inspection, the investigator could determine the building characteristics, discuss with knowledgeable personnel the proper operation of the HVAC systems, verify information obtained from the employer and employee interviews perform ventilation-system testing, and, if appropriate, collect screening samples to identify potential causes of the problem.
IAQ case study outside of Nuclear
One of the most notorious examples of Sick Building Syndrome involved the Massachusetts Registry of Motor Vehicles new offices in Jamaica Plain (Boston Massachusetts): Opened 1994, so many registry employees fell ill that the branch finally had to be closed for investigation of its IAQ, the Registry of Motor Vehicles fled in 1995.
By the time the Massachusetts Department of Public Health learned that the building's fireproofing material had deteriorated, employee productivity had declined so badly that remediation of the building cost about $156,000, plus the price of a new air chiller. The project further deteriorated to the point where the Developer went bankrupt.
Northeastern University bought the property for $17 million. Renaissance Park, formerly known as Ruggles Center, the 165,000-square-foot building, which originally sold for $35 million.
The dispersion and deposition of radionuclides is highly dependant upon the local site characteristics and is also governed by physical, chemical and biological factors. The transport models of Regulatory Guide 1.109 and NUREG 0133 utilize simplistic environmental models, which do not always account for these factors. Under certain local conditions, the environmental concentration can far exceed those predicted by the transport models. The REMP is generally designed to monitor the pathways of highest potential dose to man as described in Reg. Guide 1.109. The effectiveness of the REMP in monitoring the impact of plant releases can easily be dependant on local conditions that are not considered in the regulatory guidance. A basic understanding of transport processes can be helpful in assuring that the REMP does indeed provide an accurate picture of the radiological impact to the plant environs.
RASCAL is the computer code that the NRC's Emergency Operations Center uses to calculate projected doses during a radiological emergency. The first Windows-based version of RASCAL, Version 3.0.0, was released in March 2001. The current version is RASCAL 3.0.3, which was released in June 2002. The three releases since the March 2001 introduction have been done to fix bugs, improve the clarity of the user interface, and make other minor improvements to the code. The release of a new RASCAL 3.0.4 is expected within a few months.
To calculate projected doses, RASCAL first calculates a source term that gives the activity release rate from the facility to the environment. For nuclear power plants, the source term can be based on plant conditions, for example, effluent monitor readings, containment radiation monitor readings, or the time that the reactor core is uncovered. The user must also enter information needed to estimate a leak rate from the plant. This information can be a leak rate from the containment, containment pressure and a hole size, or an effluent flow rate. RASCAL can also calculate source terms for fuel cycle facility accidents and radioactive materials accidents.
RASCAL then calculates atmospheric dispersion using a straight line Gaussian plume model near the plant and a Gaussian puff model further from the plant. Doses are most often calculated to 10 miles, but can be calculated as far as 50 miles. The user can enter meteorological data from any weather station within 70 miles of the plant and the code will generate a wind field that considers the effects of topography on the wind speeds and directions.
The result types include the usual TEDE and thyroid CDE used for comparison with the EPA protective action guides. The TEDE includes the deep dose from cloud shine, the 50-year dose commitment from inhalation, and 4-day ground shine dose. However, many other dose types are available, for example, any of the components of the TEDE, acute doses to the bones or lungs for estimating acute radiation effects, and ground deposition.
The models used in RASCAL are described in NUREG-1741, "RASCAL 3.0: Description of Models and Methods," 2002. A "RASCAL 3 Workbook," March 2003, which gives problems for students learning to use the code, is also available.
Most radiation exposure from power plants to the public is through liquid and gaseous effluents. It is impractical to make direct, accurate measurements of exposure to the general population. Sites, therefore, depend on complex combinations of sampling and calculations to ensure regulatory compliance. Computerized systems greatly simplify the processes required to estimate exposure and report results. This presentation describes a computer-based effluents management system (OpenEMS) designed to automate the steps required to meet regulatory commitments and reporting. The system has been thoroughly tested and validated for a variety of plant designs.
Indian Point Energy Center comprises 2 operating PWRs and one which is shutdown and defueled (unit 1). Because units 2 and 3 were owned and operated by different utilities, special (and complicated) arrangements were developed to complete the REMP requirements for the site. The RETS requirements were more independent and still somewhat dissimilar. Upon being purchased by Entergy, these programs are now combining. This presentation identifies the key issues: ODCM format and content, REMP distances and locations being identical in both ODCMs/lower tier documents, the use of site-specific dose factors, and decisions with regard to finite cloud correction and elevated, ground, or mixed mode releases. Although these issues may be perfectly resolved at each unit independently, the differences and the bases must be re-evaluated upon consolidation.
This presentation points out lessons learned at IPEC with regard to multiple system modifications or operator work-arounds that were adopted or allowed to continue without due regard to effluents. Modifications in liquid waste processing, FSB ventilation, and tank vent cross-connecting at IPEC resulted in discovering effluent challenges after-the-fact. A trend began to unfold that we are now identifying as insufficient questioning attitude. Some of our recent effluent challenges were a result of too much faith on previous modifications, with system functionality and operation unquestioned. Periodic review of original plant design, the FSAR, and other design basis documents appears to be crucial to successful management of plant effluents, especially when many rapid turn-overs are involved.
The Nuclear Regulatory Commission provides a standardized method of calculating radioiodine concentrations in reactor coolant based on the dose impact relative to that of iodine-131. This concept is referred to a Dose Equivalent I-131. Although the intent was to provide for assessment of normalized iodine concentrations based on dose impact, the methodology presented in Standard Technical Specifications allows for weighting of specific iodine nuclides based on dose conversion factors presented in three different source documents, TID-14844, Regulatory Guide 1.109, or ICRP-30. Due to differences in modeling assumptions in each of these source documents, the weighting factors can result in final DEI-131 results that can vary by more than a factor of two. Since normal operational iodine concentrations are typically well below DEI limits, such differences may not directly affect Technical Specification compliance. However, DEI-131 is being used as a NRC cornerstone performance indicator, and differences in selection of DEI weighting factors may defeat the intent of a standardized approach. This presentation will discuss the differences in bases of the three reference documents, and comparisons of DEI-131 calculated via all three methods.
The purpose of environmental monitoring is to show whether the total effective radiation dose to individual members of the public, exclusive if background radiation, complies with the dose limits set forth in federal regulation 10 CFR 20, Subpart D. This in turn requires (a) continuous sampling at a constant flowrate, so that the samples will represent a true time average; (b) unhindered movement of air to the collection media; and (c) accurate counting of the radioactivity on the collection media.
Unfortunately, there have never been any published performance standards for environmental monitoring stations, most of which have been purchased on the basis of what the salesman said. It has been the writer's experience that many of these stations were built with little concern for performance and will invariably produce low measurements of the airborne radioactivity in the event of a nuclear problem. The performance testing suggestions that follow are the result of nearly 50 years experience in building and perfecting environmental air sampling instruments, including the invention of the first constant flow pump in 1953.
The following performance testing equipment will be needed, which should be equipped with 3/8" quick-connective fittings for connection to the air sampler inlet, the filter holder outlet, and the charcoal cartridge holder outlet. The cost of this equipment is small compared with the peace of mind that it will bring:
· A good quality rotameter with ½" pipe connections, tripod base,10" glass tube, and 2.5 CFM shaped float;
· A 3/8" I.D. x 6 ft. long connection hose;
· A 3/8" bronze needle valve for simulating filter loads;
· A good quality vacuum gauge for measuring the simulated filter loads in the event that the air sampler being tested is not equipped with its own;
· A good quality leak test apparatus with toggle valve, bubble jar, and adjustable vacuum source.
Flow control systems are tested by connecting the flowmeter and load simulator valve to the air sampler inlet in place of the sample collectors. After the flow has been set at the desired value, the load simulator valve is closed slowly to simulate the effect of filter loading. With good flow control the inlet flow will remain constant while the filter load is increased from zero to over 10" of Hg, and will resume to its setting in off-on tests.
Leakage in the filter holder is tested by installing a die cut flow barrier in place of the filter media and connecting the filter holder to the leak test apparatus. With the test vacuum set at about 30" of water, the toggle valve is opened to evacuate the air in the filter holder through the bubble jar. If the filter holder is tight the bubbling will stop shortly. Leakage in the charcoal cartridge holder is tested in the same way, with the charcoal cartridge in place, and the inlet plugged with a suitable fitting.
Testing for shelter bias is best done with two identical air samplers operating near each other: one inside the shelter being tested, and the other outside protected by only an umbrella. The filters are weighed before the start of the test and again a week later to determine the amounts of particulate matter collected. If membrane filters are employed, the size of the collected particles also can be compared.
Filter bias depends on the kind of radioactivity being measured. If alpha and beta radiation emitters are important, only surface collection (membrane) filters will do, in order to avoid absorption within the filter media. For gamma radiation emitters any high efficiency filter media that does not shed fibers is acceptable.
All sources of error reduce the amount of sample collected and hence the amount of radiation dose to the public that is reported.
Gaseous effluent dose calculations rely upon an estimate of the dispersion of radioactive materials as a function of distance and direction from the release point. These calculations usually use dispersion (X/Q) and deposition (D/Q) factors derived from equations outlined in Regulatory Guide 1.111. In many cases, the highest concentrations of airborne effluents occur in sectors downwind from the site, due to the influence of predominant wind direction. Although some applications of X/Q and D/Q models assume flat terrain in the vicinity of the release point, more accurate values are obtained if terrain height is incorporated into the derivation of the factors. For plants where there is an appreciable deviation of topography from flat terrain, the effects of topography can result in relatively high concentrations in sectors that would normally be considered to be upwind sectors. This presentation will discuss the meteorological patterns of wind direction near Pilgrim Station, and the influence of local topography on the occurrence of maximum dispersion and deposition factors in upwind sectors.
Many nuclear power plants employ gamma-sensitive scintillation detectors to monitor noble gases in gaseous waste streams. In most situations, the monitors are operated in a gross counting mode, and gross count rate (cpm) is converted to effluent concentration (uCi/mL) based on a conversion factor. When multiplied by process flow rate, an estimate of release rate (uCi/sec) can be calculated for comparison to applicable limits. In many cases, the conversion factor is based on xenon-133, one of the few noble gases available with an appreciable half-life to allow for calibration efforts. However, the energy distribution and photon production rate of xenon-133 does not match that of typical offgas streams, potentially affecting the applicability of a single-nuclide conversion factor to a noble gas mixture. This presentation will examine the energy distribution of various noble gases that would be anticipated in gaseous waste streams, and the implications of energy distribution to calibration and derivation of conversion factors.
There are a variety of factors that can impact the quality and effectiveness of an effluent management program. Developing a strategy to balance the effects of mixed fission products, activation products, tritium, gases, on and off-site dose, liquid volumes and unprocessed liquid waste streams can present a severe challenge to the decision making processes. Additionally, across the industry station goals related to liquid effluent activity, on and off site exposure, and liquid release volume continue to be lowered. Conversely, the costs associated with processing liquid radwaste and disposing of the resultant solid waste continue to rise.
A team of industry experts from the liquid radioactive waste (LRW) processing, liquid and solid effluents, and chemistry disciplines were assembled and are developing an outline for evaluating a LRW effluents program.
The objective is to provide an industry document that provides guidance and considerations for use when developing a sensible long-term liquid processing effluents strategy.
This paper will present the background, content logic, and current status of this important utility project.
In recent years, many power plants have adopted a policy of zero liquid discharge. In many cases, this approach is based on an unfounded and often erroneous argument of effluent ALARA. However, in reality, the policy is most often driven by competitive comparison of perceived plant performance based on INPO indicators of liquid volume discharged, and liquid activity discharged. In the case of tritium, arguments can be made that such an approach can be non-ALARA, as the dose implications of liquid discharges on a per-Curie basis are much lower than when the same activity is discharged as a gaseous effluent. Since no commercially available means exists for tritium treatment and removal, plants will be faced with the dose consequences of increasing tritium inventory in reactor coolant. Zero liquid discharge policies may exacerbate problems associated with increasing tritium inventories, and offsite doses from gaseous discharges of tritium could potentially become limiting in the future. This presentation will discuss the characteristics of tritium that contribute to this problem, the dose consequences of liquid vs. gaseous releases, and potential strategies for reducing gaseous releases of tritium and resulting offsite doses.
North American Technical Center