Decommissioning UTR-10

29 November 2000

Duke Engineering and Services (DE&S), describe the decommissioning of the research reactor at Iowa State University.

The UTR-10 research reactor at Iowa University was housed in a building located on the west edge of the main campus that was deeded to the University in 1946 by the US Department of Agriculture. The building housed the reactor, enclosed in a concrete biological shield, the process pit, the fuel storage pit, and a five-ton bridge crane.

The reactor was installed in 1959 by the Advanced Technology Laboratories division of the American Radiator and Standard Sanitary Corporation. It was an Argonaut type reactor which used uranium enriched to 19.75% U-235 in a graphite reflected, water moderated core. In 1991 the fuel was changed from its original high-enriched uranium to low-enriched uranium.

The reactor was controlled with four window-shade type Boral control rods. Heat from fission was removed from the primary coolant by a 34,000Btu/hr shell-and-tube heat exchanger that utilised city water as a heat sink. It would automatically shut down if there was a loss of AC power or if safety parameters were exceeded. The reactor was equipped with a number of experiment features including beam ports, thermal column, shield tank, internal reflector, rabbit tube, and radiation cavity.

Initial reactor criticality was on 31 December 1959. Final reactor criticality followed on 8 May 1998 and reactor operations officially ceased on 15 May 1998. Radiological characterisation surveys and sampling efforts began in July 1998 and finished in September 1998. A decommissioning plan was submitted to the NRC in January of 1999 and approved on 8 May 2000 without comment.

Decommissioning of the UTR-10 began in June 2000 and was concluded on 4 August 2000.

First stage of decommissioning

Decommissioning began with facility modifications that would create a larger workspace in the reactor area and facilitate removal of the reactor components. The north stairs, platform and stairs to the reactor top were removed using flame torches. None of these items was contaminated.

As pieces were sectioned they were surveyed, unconditionally released and removed for disposal as demolition waste or scrap metal for recycling. Access to the reactor top during the balance of the project proceeded with the use of temporary ladders.

Each reactor closure was removed and staged for survey and remediation. The radioactive portions of each closure were removed using chipping hammers, drilling/splitting, and torch cutting in preparation for unconditional release criteria. Upon removing the activated portions, the closures were moved or oriented to a low background area for final survey. Following unconditional release, each closure was loaded for disposal as demolition waste. These items included the thermal column door, shutdown closure, operating closure and experimental plugs with activation through the lower 30cm.

The tank was removed to the point where it had entered the concrete wall, along with the associated piping for the reactor water. Embedded piping was left in place or removed as necessary. All reactor water equipment was surveyed, unconditionally released and disposed of as recyclable metals or demolition waste.

Graphite was removed, beginning with that located at the south end of the bioshield. As each stringer was removed, it was numbered and segregated for survey. As the removal operations continued, each piece was surveyed in a low background area. Sections of graphite containing detectable radioactive material were sectioned and disposed of as radioactive waste. The remaining graphite was disposed of as demolition waste. Core mechanical components, activated support structures and internals were removed from the core, sectioned as necessary, and packaged as radioactive waste.

The water tank at the north end of the reactor was scored horizontally and vertically with a diamond blade concrete saw. It was then rubbleised using a hydraulic ram. As the tank walls were outside the reactor activation zone, this material was surveyed, unconditionally released and disposed of as demolition waste.

The planned sequence of dismantlement was to remove all reactor components and graphite and leave the concrete bio-shield to be demolished last.

Following the removal of the reactor components and graphite, a more definite activation profile was determined, using a combination of radiological surveys and concrete coring.

Contamination levels inside the reactor cavity were not high enough to require the use of containment during the removal of graphite and core components for packaging. As only non-transferable activated material was present, few dust control measures were required beyond the application of water. Continuous air samples were taken in the work area at the south end of the reactor room and on the roof, where large ventilation fans were placed to exhaust fumes for the heavy equipment and tools.

When necessary, additional HEPA ventilated containment was provided as an added dust control measure and to prevent cross contamination of offices, occupied areas and non-contaminated areas of the reactor room. Excess water from misting activities was collected and re-used as necessary. The remaining water was sampled and unconditionally released.

Bio shield removal

The demolition of the bio-shield began once the area was staged to accommodate the dismantling.

A pattern of vertical holes (1.5-3cm deep) was drilled through the bioshield, to provide a clear demarcation of the interior activation zone. The holes were then filled with an expansive grout and allowed to cure for approximately 24 hours. The grout created a series of fracture lines on the external surface as well as within the non-activated portions of the bioshield.

The bioshield was demolished using a remotely-operated demolition device called a Brokk. Care was taken to remove only the non-activated portions of the shield delineated by the expansive grout.

During this stage of demolition, misting and ventilation was again employed to control dust. Air samples and daily surveys of the work area were used to confirm that cross contamination did not occur. Non-radioactive portions of the bioshield were disposed of as demolition waste.

Upon approaching the activation zone, work was stopped and a HEPA ventilated containment constructed around the remaining portion of the bioshield.

After removal of the bioshield, portions of the underlying floor area that had been activated were removed and like the remaining bioshield material, disposed of as radioactive waste.

Directly underneath the reactor centreline, a small pit containing activated soil was identified. This area was sampled, a screening derived concentration guideline level (DCGL) determined and the area excavated until no detectable radioactivity remained. The removed material was disposed of as radioactive waste.

Final status survey

The NRC rule on licence termination, 10 CFR 20.1402, provides radiological criteria for release of a site for unrestricted use. Acceptability criterion for unrestricted use is a maximum total effective dose equivalent (TEDE) of 25mrem/year from residual radioactivity above background. Application of as low as reasonably achievable (ALARA) is also a requirement.

DE&S employed an alternate method described in MARSSIM section 2.6.1, which was to perform a direct comparison of each measurement result to the DCGLW, to demonstrate compliance with release criteria.

Specifically, MARSSIM states “At some sites a simple comparison of each measurement result to the DCGL, to demonstrate that all the measurement results are below the release criterion, may be more effective than statistical tests for the overall demonstration of compliance with the regulation provided an adequate number of measurements are performed”. A degree of conservatism is built into this method, given that an elevated measurement criterion (DCGLEMC) will not be utilised. DE&S believes that this type of conservatism is warranted, especially at a facility where a low potential for residual contamination exists. The overall result in implementing this ALARA method is reduced risk to a building occupant following licence termination.

Derived concentration guideline levels (DCGLS)

The DCGLs used to screen individual measurement values were based on the results of radioanalytical data as input into decontamination and decommissioning (D&D). ISU site-specific DCGLs were established by adjusting the generic limits to account for hard-to-detect radionuclides (HTDN) that cannot be measured with typical survey instrumentation.

A DCGL was developed to address the contaminated soil that lay beneath the former location of the reactor housing (CBA06). D&D v.1 was used to develop screening values for each of the five nuclides identified in the analysis of soil samples taken during the remediation phase of the project. The samples were obtained from a localised area of activated soil at the core centreline. The only input to the D &D code was the nuclide soil concentration (1pCi/g). All other input parameters were left at their default values. In addition, the resident farmer scenario was applied. Given the use of default parameter values and the application of the resident farmer scenario, DE&S believes the screening values to be conservative.

As the affected area was small, and continuation of activation products under the concrete slab was ruled out by survey, all remaining soil was removed and packaged as radioactive waste. To ensure that no licensed radioactive material remained, a composite sample from the perimeter of the excavation and a sample at the reactor centreline were taken and analysed to the appropriate environmental LLDs. Analytical results of these samples indicated that no licensed radioactive material was present.

A data quality objectives (DQO) process was implemented in the final status survey plan using a graded approach as recommended in MARSSIM. The final area classification of the ISU NEL facility was subdivided into three classes:

•Class 1. Areas that had, prior to remediation, known contaminationin excess of the DCGL, based on characterisation survey.

•Class 2. Areas that had, prior to remediation, a potential for radioactive contamination or known contamination, but are not expected to exceed the DCGL.

•Class 3 Affected areas that are not expected to contain any residual radioactivity, or are expected to contain levels of residual radioactivity at a small fraction of the DCGL.

As the characterisation report had shown that only Class 1 areas contained contamination that could be confirmed, these areas received the highest degree of effort during the final survey.

A survey documentation package was prepared for each survey unit. Each documentation package included:

•General work instructions.

•Survey location designations, results and comments regarding conditions encountered.

•A map of the survey area.

Prior to acceptance of a survey area for the final status survey (FSS), several conditions had to be met. Decommissioning activities that could ontaminate the survey area were completed. All tools and equipment not required to perform the survey were removed, and housekeeping and area cleanup were undertaken, including decontamination.

Final operational radiological surveys ensured that no additional remediation was necessary. These comprised:

•Scan surveys to ensure that surface contamination was within the FSS total surface contamination limits.

•Smear surveys to ensure that the removable surface contamination was within the FSS removable surface contamination limits.

An investigation was performed to confirm the initial FSS measurement, and if verified to exceed action levels, to define the bounds of the elevated activity and extent of remediation, and reclassify and re-survey.

In the FSS for survey areas where there is a potential for contamination, the following control measures were employed:

•Installation of barriers to control access to surveyed areas.

•Installation of postings requiring personnel to perform contamination monitoring prior to surveyed area access.

•Locking of entrances to surveyed areas of the facility.

All personnel directly involved with the performance of the final status survey and data quality review received appropriate training, including:


•Final status survey procedures.

•Equipment operation.

Quality assurance

Field surveys and the collection of radiological data in the field require additional quality provisions under MARSSIM. The quality assurance (QA) programme specifies the policies, organisation, objectives, and QA/QC procedures used by DE&S for site characterisation and final survey activities. The primary goal of the QA programme was to identify and specify the implementation of surveying, sampling, and analytical methodologies, which limited the introduction of errors into the analytical data.

The specific purpose of the QAPP was to ensure that the samples were collected, analysed and reported in a consistent manner and that the quality of the resulting data could be independently evaluated.

To perform the final status survey, both field survey instrumentation and analytical laboratory equipment were selected based on: the necessary minimum detectable concentrations (MDC); and stability and reliability under environmental conditions. DE&S utilised the appropriate instrumentation to perform the final status survey of the NEL site. All field survey instrumentation was calibrated, operated and controlled in accordance with applicable DE&S procedures.

Minimum detectable concentration (MDC) values for field and laboratory counting instrumentation were determined using the following equation, taken from Equation 3-10 in Draft NUREG-1507:


MDC = the minimum activity concentration on a surface or material volume that can be statistically detected above background with a 95% probability,

ts = sample counting time interval,

Rb = background counting rate,

tb = background counting time interval,

E = counting efficiency, and

A = area of the detector, or the area sampled for smear samples.

Background activity and radiation levels for the final status survey were determined from appropriate background reference areas and by obtaining local area measurements within the survey unit. Background levels were established for each type of instrument to be used for total surface contamination measurements and exposure rate measurements. Background levels were subtracted from total radiation or radioactivity levels to determine the net residual contamination from licensed operations.

Measurement frequencies, or the physical spacing of samples and measurements, were selected to allow for a concentrated survey effort in those areas most likely to be contaminated or activated, taking into account the type and size of the survey unit. Measurement locations distributed throughout a survey unit were documented in accordance with the Final Status Survey Plan. Locations were clearly marked to facilitate re-survey.

Survey methods

Scan surveys were performed to screen surface areas for the presence of any locations of elevated contamination above the release limits and to detect localised areas above the maximum release limit. The scanning methods (instruments and survey technique) for surface contamination measurements were designed to detect less than 75% of the mean total surface contamination limits. If an area of elevated contamination was identified during the scan of a survey unit, the location was marked and included as a part of the total surface contamination measurements.

Surface contamination measurements were taken at discrete measurement locations and at frequencies that were based on the classification of the survey unit. As no alpha contaminated areas were identified during characterisation, only beta total surface contamination measurements were taken at each location.

Exposure rate measurements were obtained at discrete locations, based on the classification of the survey unit. They were taken at a distance of 1m from the surface and could only be obtained in accessible survey units, normally on floors.

Smear samples for a given survey location were not collected until the scan survey and TSC measurements for that location were completed.

ISU site-specific DCGI's
Bulk material composite sample 10 CFR 61 analytical results summary
ISU screening values for soil

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