Key Words

Decommissioning

Shielding

Radwaste
 

Abstract

During the decommissioning of the Big Rock Point Nuclear Power Station many large components are dismantled, cut into smaller pieces and shipped for burial. These components often exhibit high radiation levels requiring shielding and special handling precautions. The activity of several components, and associated piping, valves, and exposure rates of the shipping container were evaluated and shielding recommendations made to reduce exposure to personnel working in the vicinity of the container. Thicknesses of lead and water boxes required to reduce exposures to reasonable levels was calculated.
 

Introduction

Big Rock Point Nuclear Power (BRP) was shut down in August 1997, after 35 years of operation and is currently being decommissioned. The BRP staff cut and removed piping and valves from several locations, to be shipped for burial. The list of the components and piping, and their measured exposure rates are provide in Table 1. A B-25 model container was stationed in the equipment hatch vicinity (west side of the containment sphere) for placing the removed material prior to shipment. The container is made of 0.0625 inch steel with dimensions of 6 ft wide 4 ft high, and 4 ft depth.
 

Methodology

Due to high radiation levels of the components the container had to be shielded in accordance with BRP policies and procedures for reducing exposures. The Radiation Protection & Environmental Services (RP&ES) staff placed the higher radiation level components in the middle of the container surrounded by the lowest reading components. This took advantage of the self-shielding provided by the components with lower radiation levels to reduce the amount of shielding needed for the container.

For this evaluation, a gamma spectroscopy analysis was obtained on October 1, 1998. This analysis was performed after the completion of the decontamination for decommissioning (DFD). The spectroscopy equipment was placed outside of the steam drum entrance area for obtaining samples during chemical decontamination and continued after the completion of the process. Table 1 provides the sample analysis and the isotopic ratios following chemical decontamination.
 

Table 1. Gamma Spectroscopy result from Steam Drum
 

	Activity (Gamma Spec)	Percent Fraction
	Ci
Mn-54	7.05e-07	4.42%
Co-60	1.27e-05	79.61%
Cr-51	1.96e-06	12.26%
Zn-65	4.02e-07	2.52%
Co-58	1.91e-07	1.20%
Total	1.60e-05	100.00%

 

The MicroShield^® code was used to model the equipment geometry in order to determine the total activity on these components. Models were based on several scenarios, such as length, and radius of the component, for determination of the activity. Calculated results were obtained based on Table 1 ratios to determine the activity. To be conservative the highest activity piping was used in the shielding calculations. Credit was taken for self shielding of materials within the box by assuming that contents of the box were 50% air and 50% steel. Although this should provide fairly realistic average dose rates external to shields, there maybe local hot spots (where streaming occurs, or a particularly hot item is near a surface) which could require additional localized shielding. Table 2 provides the activity for each component. The pre-decon exposure rates of Table 2 are exposure rates in contact with the specified equipment prior to chemical decontamination and illustrate the effectiveness of the decontamination process.
 
 
 

Table 2. Activity of each component
 

	Calculated Exposure Rate	Mn-54	Co-60	Cr-51	Zn-65	Co-58	Total	Pre-decon Exposure Rate
	mR/hr	Ci	Ci	Ci	Ci	Ci	Ci	mR/hr
Isotope		7.05E-07	1.27E-05	1.96E-06	4.02E-07	1.91E-07	1.60E-05
Fraction		4.42E-02	7.96E-01	1.23E-01	2.52E-02	1.20E-02
Pipe 3' long , 8" diameter, 0.1" thick	4.68E-01
Feed-water line		1.51E-02	2.72E-01	4.18E-02	8.60E-03	4.08E-03	3.41E-01	1.00E+04
Feed-water line		2.41E-02	4.34E-01	6.69E-02	1.38E-02	6.53E-03	5.46E-01	1.60E+04
Pipe 3' long, 4" diameter, 0.1" thick	8.60E-01
Condensate return line		2.46E-02	4.43E-01	6.82E-02	1.40E-02	6.66E-03	5.57E-01	3.00E+04
Condensate return line		3.28E-03	5.91E-02	9.10E-03	1.87E-03	8.88E-04	7.43E-02	4.00E+03
Pipe 3' long, 2" diameter, 0.1" thick	4.09E+00
Instrument line		1.72E-04	3.11E-03	4.78E-04	9.84E-05	4.67E-05	3.90E-03	1.00E+03
V-247		1.72E-04	3.11E-03	4.78E-04	9.84E-05	4.67E-05	3.90E-03	1.00E+03
V-249		1.03E-04	1.86E-03	2.87E-04	5.90E-05	2.80E-05	2.34E-03	6.00E+02
V-262		1.03E-04	1.86E-03	2.87E-04	5.90E-05	2.80E-05	2.34E-03	6.00E+02
V-Vns-159		1.03E-04	1.86E-03	2.87E-04	5.90E-05	2.80E-05	2.34E-03	6.00E+02
Poison line cluster (CV-4050)		6.04E-04	1.09E-02	1.67E-03	3.44E-04	1.63E-04	1.37E-02	3.50E+03
Pipe 10' long, 2" diameter, 0.1" thick	1.31E-01
Mo-N007A		3.23E-03	5.82E-02	8.96E-03	1.84E-03	8.74E-04	7.31E-02	6.00E+02
Yarways		1.88E-02	3.40E-01	5.23E-02	1.08E-02	5.10E-03	4.27E-01	3.50E+03
RWCU sample sink drain line behind#1 RCP		4.31E-02	7.76E-01	1.19E-01	2.46E-02	1.17E-02	9.75E-01	8.00E+03
RCP 1, 2 sample lines and transmitters		5.38E-03	9.70E-02	1.49E-02	3.07E-03	1.46E-03	1.22E-01	1.00E+03
NC #113 and 114 filters		8.08E-03	1.46E-01	2.24E-02	4.61E-03	2.19E-03	1.83E-01	1.50E+03
Mo-IA60A&B with 3/4" lines		1.08E-02	1.94E-01	2.99E-02	6.14E-03	2.91E-03	2.44E-01	2.00E+03
Jumper between the loops @the drain-lines		4.31E-02	7.76E-01	1.19E-01	2.46E-02	1.17E-02	9.75E-01	8.00E+03
Total Activity (Ci)		2.01E-01	3.62E+00	5.57E-01	1.15E-01	5.43E-02	4.54E+00

Recommendations
 

The Radiation Protection & Environmental Services (RP&ES) staff cut and removed the components listed in Table 2. These components were stored in B-25 container(s), placed inside the containment sphere in the vicinity of the equipment hatch. The containers were then transferred to the radwaste area to be processed for shipping. To maintain personnel exposures ALARA it was necessary to place shielding around the containers to reduce the general area radiation levels.
 

The components inside the container caused the general area radiation levels to increase. To reduce the area exposure levels and reduce worker exposure, the container was shielded in accordance with BRP's ALARA policy and practice. To determine the shielding, MicroShield^® was used to model the box, as it was describe in the introduction of this article. The loading activity was taken from Table 1 for this evaluation, also it was assumed the total activity is within one box. Table 3 provides the shielding requirements using available materials to reduce the exposure in that area.

Table 3 - Shielding Requirements for One Container
 

Lead	Water	Distance from side of the container	Exposure Rate	Dose Equivalent (ICRP-51)
inch	inch	inch	mR/hr	mrem/hr
0	0	0	5.17E+02	4.60E+02
1	15	1	1.20E+01	1.00E+01
2	N/A	1	2.90E+01	2.60E+01

 

Conclusions

During the decommissioning of a facility such as a nuclear power plant there will be many challenges the staff will face to keep exposures As Low As Reasonably Achievable. This article has presented a solution to one facet of the basic time, distance and shielding concepts of dose reduction.
 

The Authors

Rahim Ghanooni is a Senior Health Physicist at HPCI Consulting. He has a BS degree in Physics from Northeastern University in Boston. Mr. Ghanooni has been a technical reviewer since 1989, and has written several articles for RPM. He is a Certified Hazardous Material Manager (CHMM), and specializes in decommissioning, shielding calculation, fuel handling, dry cask exposure evaluation, mixed/ hazardous waste, internal/external dosimetry, and instrumentation. He is currently a member of the ANSI N-48 Committee (Radioactive Waste Management). At the time of this article, he was assigned to Big Rock Point for decommissioning.

Robert English is Health Physics Consultant for Consumers Energy at the Big Rock Point Nuclear Plant, Charlevoix, Michigan. He received a Ph.D. from the University of Michigan, and is certified (CHP) by the American Board of Health Physics. Dr. English has been with Consumers Energy for 25 years, and has served in capacities such as Midland Plant Health Physicist and Corporate Health Physicist. At Big Rock Point , he has been responsible for projects such as development of the decommissioning environmental report, development of the plant's defueled emergency plan, and implementation of the plant's radiological scoping study using In-Situ gamma spectroscopy.

Trademarks

Code MicroShield V.5.03 Grove Engineering
1700 Rockville Pike
Suite 525
Rockville, MD 20852

References

• ICRP 51, Data For Use in Protection Against External Radiation, 3/88