February 3, 2015 Mail Control No. License No. Docket No. 585630 07-30728-01 030-35986 Licensing Assistance Team U.S. Nuclear Regulatory Commission Region I 2100 Renaissance Blvd King of Prussia, PA 19406-2713 Subject: Mail Control No. 585630 - Amendment to License No. 07-30728-01 Dear Mr. Dennis Lawyer, lncyte Corporation 1801 Augustine Cut-off Wilmington, Delaware 19803 Tel 302.498.6776 Fax 302.425.2109 Web www.incyte.com ... !-·-!.<· en :r.:i :::.:;:: This is to provide our response to NRC's e-mail dated January 15, 2015, requesting additional information to lncyte Corporation's December 18, 2014 request for removal of the DuPont Experimental Station, buildings E336 and E400, Route 141 & Henry Clay Road, Wilmington, Delaware 19880 location from NRC Material License No. 07-30728-01. Following is each of the items for clarification {1.a. through b. and 2.a through 1.) followed by our response (shown in bold font). 1. Prior to termination of a license, 10 CFR 30.35{g), 30.36{k)(4) and 30.51 require that you submit certain records to the NRC. a.. for unsealed materials with half-lives greater than 120 days, records for disposal made pursuant to 10 CFR 20.2002 (alternate disposal procedures, including burial authorized prior to January 28, 1981), 20.2003 (disposals to the sanitary sewerage system), 20.2004 (incineration of wastes), 20.2005 (disposal of specific wastes including liquid scintillation cocktail and animal tissue), and 20.2103{b)(4), evaluations of effluent releases. b. records important for decommissioning as described in 30.35(g), 40.36(f) and 70.25(g). Examples of such records include but are not limited to: records of contamination, identifying the radionuclides, quantities and concentrations; as-built drawings and modifications of structures and equipment in restricted areas and locations of inaccessible contamination such as buried pipes; a single list, updated at least every 2 years, of areas to which access is limited for the purpose of radiation protection (restricted areas); and records related to the provision of financial assurance. lncyte Corporation is retaining records for all of the above records as required for the Buildings E336 and E400, Route 141 & Henry Clay Road, Wilmington, Delaware facility. 1 ... t" !""'\') 'Nr .. r • .;....1.,, .......... 1.
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February 3, 2015
Mail Control No. License No. Docket No.
585630 07-30728-01 030-35986
Licensing Assistance Team U.S. Nuclear Regulatory Commission Region I 2100 Renaissance Blvd King of Prussia, PA 19406-2713
Subject: Mail Control No. 585630 - Amendment to License No. 07-30728-01
Tel 302.498.6776 Fax 302.425.2109 Web www.incyte.com
... !-·-!.<·
en :r.:i :::.:;::
This is to provide our response to NRC's e-mail dated January 15, 2015, requesting additional information to lncyte Corporation's December 18, 2014 request for removal of the DuPont Experimental Station, buildings E336 and E400, Route 141 & Henry Clay Road, Wilmington, Delaware 19880 location from NRC Material License No. 07-30728-01. Following is each of the items for clarification {1.a. through b. and 2.a through 1.) followed by our response (shown in bold font).
1. Prior to termination of a license, 10 CFR 30.35{g), 30.36{k)(4) and 30.51 require that you submit certain records to the NRC.
a.. for unsealed materials with half-lives greater than 120 days, records for disposal made pursuant to 10 CFR 20.2002 (alternate disposal procedures, including burial authorized prior to January 28, 1981), 20.2003 (disposals to the sanitary sewerage system), 20.2004 (incineration of wastes), 20.2005 (disposal of specific wastes including liquid scintillation cocktail and animal tissue), and 20.2103{b)(4), evaluations of effluent releases.
b. records important for decommissioning as described in 30.35(g), 40.36(f) and 70.25(g). Examples of such records include but are not limited to: records of contamination, identifying the radionuclides, quantities and concentrations; as-built drawings and modifications of structures and equipment in restricted areas and locations of inaccessible contamination such as buried pipes; a single list, updated at least every 2 years, of areas to which access is limited for the purpose of radiation protection (restricted areas); and records related to the provision of financial assurance.
lncyte Corporation is retaining records for all of the above records as required for the Buildings E336 and E400, Route 141 & Henry Clay Road, Wilmington, Delaware facility.
2. The DuPont Experimental Station Building 336 and 400 Radiological Final Survey Report had several items that did not appear to be accurate or was missing information needed to validate and confirm results. Please clarify the following and resubmit the report as needed.
Attached please find RSO lnc.'s, clarification to (2.a. through I.) for the NRC's January 15, 2015 e-mail and a revised report for resubmission to the NRC.
If you require additional information please contact Mark Czerwinski at 302-498-6827 or by e-mail at [email protected].
Mark Czerwinski lncyte Corporation 1801 Augustine Cut-off Wilmington, DE 19803-4404
Radiation Service Organization
Subj: lncyte Dupont Experimental Station Bldg. 336 and 400 Final Survey Report Re: Response to NRC's email of January 15, 2015
Dear Mr. Czerwinski:
This is to provide our responses to the NRC's email of January 15, 2015. Following is each of the items for clarification (2.a. through I.) followed by our response (shown in bold font). Also we have provided a revised report (attachment) for resubmission to the NRC.
2. The DuPont Experimental Station Building 336 and 400 Radiological Final Survey Report had several items that did not appear to be accurate or was missing information needed to validate and confirm results. Please clarify the following and resubmit the report as needed.
a. The report describes P-33 as a potential contaminate in some of the laboratories, specifically Building E336 labs 238, 266, Dock Cabinet, and Building 400 Lab 1440. The report in section 2.3 states that only H-3 and C-14 were considered potential contaminates but then states that P-33 was identified as a potential contaminate. Please explain what the distinction is made in this section and what implications it had on the survey design.
Section 2.3 has been re-written to accurately describe the potential radionuclides. p.33 was a potential contaminate. P-33 Is a beta emitter with a beta energy higher than that of C-14 (294 keV max), and a DCGL higher than that of C-14. P-33 did not impact the survey design because the C-14 DCGL was more restrictive, and the detection methods used are the same as that for C-14.
We also added S-35 as a potential contaminate, as our own clarification of past use shows a one-time use about 1 year prior to the survey. S-35 is a beta emitter with a beta energy (167 keV) nearly identical to that of C-14 and a DCGL also higher than that of C-14. Adding S-35 as a potential contaminate had no impact on the survey design for the same reasons provided above.
b. It is unclear how you determined the DCGLw for P-33. Please provide the specific reference or Dando calculation that supports this value.
We re-calculated the DCGL using Dando Version 2.1 O and obtained a value of 4.09E+07 (a copy of the DandD summary report is attached). We have corrected the report to this value. While it is slightly lower than the value in the report, it is relatively close and the difference had no effect on the survey design (e.g.: calculations of the number of survey locations needed in a survey unit is the same as before and the same as for H-3 and C-14).
c. In section 2.10, step 1 discusses P-33 as radioactive material used in the facility but then it infers that it could not exceed the OCGL. No justification is made on this statement. Step 2 did not discuss P-33.
Washington DC (301) 953·2482 www.rsoinc.com Baltimore (410) 792-7444
RSO, Inc.
We have re-written Step 1 through Step 7 to more accurately discuss and include P-33 and S-35. We also added this information: The amounts of radioactivity used at any 1 time were relatively small (milllcurle or sub-millicurie amounts) and there were no spills or incidents resulting in widespread contamination.
d. There is no discussion of how P-33 was surveyed or if the results for the data given support release of the facility for P-33.
We have added statements and information in Section 2.9 regarding the applicability of the survey methods for each radionuclide, including P-33 and S-35.
e. Section 2.4 states that laboratory rooms where relatively high activity concentration of radioactivity were used were treated as Class 2 areas. There was no discussion of which labs were class 2 and if any were class 3. If any laboratory was considered Class 2, based on the survey diagrams given, there was no apparent systematic square grid with random start surveys performed as stated in section 2.9.
We have removed that sentence regarding treating any of the laboratory rooms as Class 2 areas as they are all treated as Class 3 areas.
f. Section 2.7, Table 4 is titled, "List of radionuclides listed on license (with data) and recent use." Please define what recent use means. In particular, were some laboratories not included that had tritium and carbon 14 use longer than 4 years ago?
We have removed the reference to recent use and rephrased the description for Table 4. All laboratory rooms that used tritium and C-14 were included.
g. Section 2.9 describes detector surveys. For scan surveys, it does not describe the percentage of area that was scanned. Normally this is different for Class 2 and 3 areas, but this did not appear to be described either. Very few static surveys were performed on walls. Were walls scanned?
Yes, scan survey were conducted on 5 to 10% of the walls and 50 to 90% of the bench tops and floor areas.
h. For the table described as Survey Data, subtitled Survey Meter Information:
• The serial number of the meter is the same as the serial number of the probe. Calibration data shows that the meter number and probe number is different. This makes it unclear as to how efficiency of the detector was determined if the probe has the same serial number.
• Meter 3 states that it used Ludlum model 43-37 probes, but the detector size is 126 square centimeters. Model 43-37 should be 584 square centimeters.
The correct information has been entered in the revised report.
• The page gives formulas in the bottom section. These values are sometimes not given or not clear in the above data chart. For example, the human factor efficiency that is used is not given. Counting interval is not given. Source efficiency in formula is stated as just efficiency, 2pi efficiency is displayed. Background Counts is given as Background (c) in table. T counting time in minutes is different for background and survey counting and thus needs to be factored. Source efficiency is not given. Based on these factors this license reviewer could not duplicate the numbers calculated in the survey data. Also carbon-14 is a low energy beta and a source efficiency of half that for P-33 is normally used. Please provide all data used and make the data in the table consistent with the variables listed on the page. It is suggest to send a detailed example calculation.
Washington DC (301) 953-2482 WW\\ .rsoinc.com Baltimore (410) 792-7444
RSO, Inc.
We have updated our spreadsheet and using a "source efficiency" of 0.25 for C-14 and provided a detailed sample calculation page in the results and a description of the formulas and calculations in the body of the report.
i. Building 336, Lab 228 survey data shows the Gross high (cpm) being lower than the Gross Average (cpm) which seems to be an inaccuracy. Please correct or explain.
The correct information has been entered In the revised report.
j. Surveys do not appear to include Building E336 Dock cabinet or 282 Freezer Farm which was described in section 2.7. There is a survey for the Deck Area which did not seem to correlate with any description in section 2.7.
The area identified as Dock Cabinet In Section 2.7 is the same as Dock Area as shown on the floor diagram and Is Identified as Dock Area on the survey data report. The area Identified as 282 Freezer Farm In Section 2.7 is the area identified as room 282 on the survey data report.
k. The calculations for carbon 14 did not seem to include a factor for the low energy self-absorption of measuring low beta energy on surfaces. Please describe what factors were used and how this was accounted.
We have updated our spreadsheet and using a "source efficiency" of 0.25 for C-14 and provided a detailed sample calculation. This also changed the data shown in the data sheets and we have changed the data used and shown in the body of the report.
I. Calibration data was supplied for four instruments. Only two were used for these surveys. Why was the additional detector calibration data submitted? Is the survey data inaccurate to which detectors were used?
The additional survey meter calibration reports have been removed as they were not used for the Final Survey. The survey data Is accurate as to which detectors were used.
Please contact me or Korressa Williams if you have any questions or need any additional information.
Sin/ly,
i-~v1~ Gregory D. Smith, CHP RSO, Inc.
Washington DC (30 I) 953-2482 """' .rsoinc.com Baltimore (410) 792-7444
DandD Building Occupancy Scenario
DandD Building Occupancy Scenario
DandD Version: 2.1.0 Run Date/Time: 1/21/2015 11 :36:54 AM Site Name: lncyte Description: P-33 FileName:C:\Program Files\DandD2\docwnents\P-33 DandD Bldg Screening Value.med
Options:
Implicit progeny doses NOT included with explidt parent doses Nuclide concentrations are distributed among all progeny Number of simulations: 12800 Seed for Random Generation: 8718721 Averages used for behavioral type parameters
External Pathway is ON Inhalation Pathway is ON Secondary Ingestion Pathway is ON
Initial Activities:
Area of Nudidc:
Contamination (m2)
Distribution
33P ~LIMITED K'.ONSTANT( dpm/I 00 cm .. 2)
lu.iification ff'r .. , -·Screening Value Value 4.09E+07
Site Specific Parameters:
General Parameters:
Correlation Coefficients:
Summary Results:
90.00% of the 12800 calculated TEDE values are< 2.48E+Ol mrem/year. The 95 % Confidence Interval for the 0.9 quantile value ofTEDE is 2.46E+Ol to 2.50E+01 mrem/year
RSO, Inc. • lncyte Corporation • Dupont Experimental Station Buildings 336 and 400 Radiological Final Survey Report
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1.0 INTRODUCTION AND RADIOLOGICAL DECOMMISSIONING ACTIONS
1.1 Introduction
lncyte Corporation maintains a Nuclear Regulatory Commission issued byproduct material license of limited scope (07-30728-01). The use of licensed material at the Dupont Experimental Station, Building E336 and E400 in Wilmington, DE has ended and radiological final survey was needed to support removal of this location as an authorized place of use.
1.2 Background-Historical Site Assessment
lncyte began using radioactive material in buildings E336 and E400 in 2003 with regular use continuing through until November 2014. The radionuclides authorized for use are listed in Table 1. H-3 and C-14 have been used; however, of the radioactive materials with half-lives of less than 120 days only P-33 was used in the last 4 years. This use was with non-volatile forms in processes that did not create airborne radioactivity.
Table 1. List of radionuclides listed on license (with data) and recent use.
Authorized Use within 4
Half Life Principle Emission Energy years of the Radionuclides
on-site survey
H-3 12.3 yrs beta 18.6 keV (max)
C-14 5730 yrs beta 156 keV (max)
P-32 14.29 days beta 1.7 MeV (max)
P-33 28 days beta 249 keV (max)
S-35 87.3 days beta 167 keV (max)
1-125 60.1 days photons 27.5-36 keV
(144%) Note 1 last use was 10/6/2005, elapsed time is 3313 days, effectively 100% radioactive decay Note 2 last use was 9/10/2009, elapsed time is 1878 days, effectively 100% radioactive decay
Yes
Yes
No1
Yes
Yes3
No2
Note 3 1 use (ie: benchtop assay) since 11 /03/09 and that was on 3/12/13 in E336/Lab 238, elapsed time is 609 days, effectively 99.2% radioactive decay
The license applications and amendments to this license identified multiple rooms and areas where radioactive materials were to be used or stored.
The laboratory rooms shown in Table 4 were identified as areas authorized for the use or storage of licensed material.
Routine contamination surveys were performed by the licensee during this period of use. No large spills were known to have occurred. Small spills that occurred were cleaned by lncyte personnel.
1.3 Radiological Decommissioning
The use of radioactive material has ended, radiological decommissioning actions taken/completed, and a Radiological Final Survey was performed to allow the release of the building for unrestricted use.
Prior to the final survey all remaining radioactive material was either transferred to lncyte's other authorized location of use, or disposed of as radioactive waste.
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2.0 RADIOLOGICAL SURVEY APPROACH
2.1 Survey Design Basis
This radiological survey was designed in consideration of the guidance provided by the Nuclear Regulatory Commission (NRC) regarding Final Radiological Surveys. In particular the guidance provided by the NRC in NUREG 1757 for what is termed Group 2 facilities (see following excerpt) was used. Group 2 includes facilities that "would not have contaminated work areas at the levels above the decommissioning screening criteria".
From NUREG 1757 v1 Chapter 7:
Group 2 facilities may have residual radiological contamination present in building surfaces and soils. However, licensees are able to demonstrate that their facilities meet the provisions of 10 CFR 20.1402 ("Radiological Criteria for Unrestricted Use") by applying the screening approach dose analysis described in Chapter 6.
Additionally, licensees in Group 2 typically possess historical records of material receipt, use, and disposal, such that quantifying past radiological material possession and use may be developed with a high degree of confidence. Furthermore, these licensees have radiological survey records that characterize the residual radiological contamination levels present within the facilities and at their sites. That is, they are able to demonstrate residual radiological contamination levels without more sophisticated survey procedures (greater than those used for operational surveys) or dose modeling. These licensees do not need to use site-specific parameters or establish site-specific DCGLs in order to demonstrate acceptability for release of their sites."
Derived Concentration Guideline Levels (DCGLs) are radionuclide-specific concentration limits used by the licensee during decommissioning to achieve the regulatory dose standard that permits the release of the property and termination of the license. The DCGL applicable to the average concentration over a survey unit is called the DCGLw. The DCGL applicable to limited areas of elevated concentrations within a survey unit is called the DCGLEMC·
2.2 Decommissioning Criteria
The Radiological Criteria for Unrestricted Use - NRC (10 CFR Part 20)
"A (The) site will be considered acceptable for unrestricted use ifthe residual radioactivity that is distinguishable from background radiation results in a TEDE to an average member of the critical group that does not exceed 25 mrem per year, including that from ground water sources of drinking water, and that the residual radioactivity has been reduced to levels that are as low as reasonably achievable (ALARA)."
2.3 Potential Radionuclide Contaminates and Screening Values
At the time of the survey, the facility was licensed for use of relatively long half-life radionuclides, H-3 (T112 = 12.3 y) and C-14 (T112 = 5730 y) and relatively short half-life (T112
of less than 120 days) radionuclides, S-35, P-33, P-32 and 1-125.
Based on the Historical Site Assessment, H-3 and C-14 and P-33 (T112 = 28 d) were identified as the potential radionuclide contaminates.
The NRC has established Screening Values derived using scenarios and default values for assumptions that result in a Derived Concentration Guideline Limit (DCGL). These values have been derived for common beta-gamma emitting radionuclides for building surface contamination as published in the Federal Register (63 FR 64132, November 18, 1998)
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RSO, Inc. • lncyte Corporation • X-Station Buildings E336 and E400 Radiological Final Survey Report
and also shown in Appendix B Table 8.1 of NUREG 1757 and by definition is the DCGLw. These are values, which can also be derived using the default parameters and the computer code DandD, for the concentration (dpm/100 cm2
) equivalent to 25 mrem/y.
For beta-gamma emitters the DCGLw is typically higher than the facility operational contamination limits. The DCGLw for the potential radionuclides are shown in Table 2.
1 bl 2 DCGL f . I a e w or po ent1a contaminates.
Radionuclide Surface Contamination (dpm/100 cm2)
H-3 *1.2 x 108
C-14 *3.7x106
S-35 *1.3 x 107
P-33 **4.09 x 107
*from NUREG 1757 ** calculated used DandD Ver. 2.1.0
2.4 Performance of Radiological Surveys
The radiological surveys were conducted using guidance provided by the NRC in NUREG-1575, EPA 402-R-97-016, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM): Revision 1, August 2000.2.5 Survey Design: Area Classification
Impacted Areas
• Impacted areas are areas that may have residual radioactivity from the licensed activities.
• Non-impacted areas are areas without residual radioactivity from licensed activities.
• NRC guidance provides that Final Status Survey (FSS) radiation surveys do not need to be conducted in non-impacted areas.
The impacted area for this license was considered to be the identified laboratory rooms where radioactive material was authorized for use, the radioactive waste storage area, and the connecting corridors.
Classes
Impacted areas can be classified into one of the three classes, listed below, based on expected levels of residual radioactivity.
• Class 1 Areas are impacted areas that, prior to remediation, are expected to have concentrations of residual radioactivity that exceed the DCGLw (DCGLw is defined in Section 2.2 of MARSSIM);
• Class 2 Areas are impacted areas that, prior to remediation, are not likely to have concentrations of residual radioactivity that exceed the DCGLw.;
• Class 3 Areas are impacted areas that have a low probability of containing residual radioactivity.
All areas were treated as Class 3 with no or very limited areas of residual contamination expected.
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RSO, Inc. • lncyte Corporation • X-Station Buildings E336 and E400 Radiological Final Survey Report
2.6 Decommissioning Guideline Levels
The decommissioning guideline levels used for the surveys in are shown in Table 3.
li bl 3 s rf t · d r f r · r d t h F 1 s a e u ace con aminat1on gu1 e mes or 1cense term1na ion use ort e 1na urvey.
Maximum Operational Limit Operational Limit
Removable Total Radionuclide
DCGL (dry wipe method) (scan with survey (dpm/100 cm2)
(dpm/100 cm2) meter)
H-3 1.2 x 108 200 N/A
Not detectable
C-14 3.7 x 106 200 3 x bkg (estimated to be about 2500
dpm)
1.3 x 107 3 x bkg (estimated
S-35 200 to be about 2500 dpm)
4.09 x 107 3 x bkg (estimated
P-33 200 to be about 2500 dpm)
2.7 Survey Units (Areas)
Table 4. lists laboratory rooms or areas where radioactive material was authorized for use (or storage), lab area and the radionuclide(s) used. If the radionuclide used has a T 112 of less than 120 days, "use" is defined as within the 4 years prior to the Final Survey.
Table 4. List of radionuclides listed on license (with data) and use. Laboratory or Lab area
Building location in SQ. ft. C-14 H-3 1-125 P-32 P-33 S-35 1 E336 238 476.1 x x x x 2 E336 266 299.72 x x x 3 E336 268 126.81 x x 4 E336 Dock Cabinet 25 x x x 5 282 Freezer
E336 Farm 25 x x
6 E400 3212 469 x x 7 E400 3214 705 x x 8 E400 3226 705 x x 9 E400 3238 705 x x 10 E400 3403 225 x x 11 E400 3407 705 x x 12 E400 3419 705 x x 13 E400 3474 -600 x x 14 3rd Floor
E400 Corridor freezer 25 x x 15 E400 1440 x x x
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RSO, Inc. • lncyte Corporation • X-Station Buildings E336 and E400 Radiological Final Survey Report
Floor plans of each building showing the locations of the laboratory rooms/areas are shown below:
l •
BUILDING E336, SECOND FLOOR H
BUILDING E336, BASEMENT H
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RSO, Inc. • lncyte Corporation • X-Station Buildings E336 and E400 Radiological Final Survey Report
.. • • e 9 •
BUILDING E400, THIRD FLOOR H
BUILDING E400, ARST FLOOR H
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RSO, Inc.• lncyte Corporation• X-Station Buildings E336 and E400 Radiological Final Survey Report
2.8 Survey Number of Samples
Using MARSSIM's guidance for determination of the number of samples needed for a survey unit when the DCGL is large (as is the case for this survey), the relative shift is also large (>2.5), and using equal values of 0.05 for Type I and Type II errors, results in a number of data points needed of 12.
A minimum of 12 samples were collected (per survey unit) using an informal rectangular grid, with a random start point, and additional sample locations selected by the survey team. A scan (floor monitor or hand-held survey meter), direct (static) measurement, and wipe test was performed at each survey location except where noted. Additional sample locations were chosen by the discretion of the surveyor in each survey area including floors, bench tops, cabinets, doors, and sinks to provide additional survey data.
2.9 Survey Methods
The survey methods for each potential radionuclide
Wipe Test for Radionuclide Static/Direct Scan Removable
Contamination H-3 Not practical with survey Not practical with Yes liquid scintillation
Beta Scan Survey Surface scanning speeds were 2 detector widths per second. To optimize detection of elevated radiation levels (1.5 to 3 times background) during scanning, audible speakers were used in addition to noting the fluctuations in the analog meter reading. Floor scans were performed using a Ludlum Floor Monitor (Model 2221 survey meter coupled to a Ludlum 43-37 gas flow proportional detector (thin window coated mylar of 0.8 mg/cm2 with an area of 584 cm2
). Scans of other surfaces were performed using a Ludlum Model 2221 with a Ludlum 43-68 probe (gas proportional detector, thin window of 0.8 mg/cm2 with an area of 126 cm2
). Scan survey were conducted on 5 to 10% of the walls and 50 to 90% of the bench tops and floor areas.
Static (Direct) Measurements of Surfaces Static radiation measurements for beta/gamma surface contamination were performed at random (informal square grid with a random start) and biased (sinks, floor drains) locations using a Ludlum Model 2221 with a Ludlum 43-68 detector (gas flow proportional, thin window of 0.8 mg/cm2 with an area of 126 cm2
). Measurements were conducted by integrating a 1-minute count time with the probe in direct contact with the surface.
Removable Contamination A wipe test for removable contamination was performed at each survey location. The wipe
test consisted of wiping a minimum of 100 cm2
of the surface with a dry paper, using moderate pressure and measuring the amount of radioactive material on the test material using liquid scintillation counting (RSO, Inc. Packard TriCarb 3100).
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Quality Assurance Survey meters used to perform the Final Survey had been calibrated within 12 months of their use using radioactive standards traceable to NIST. Also, performance checks were completed on each survey meter at the beginning of the survey.
The laboratory instruments used by RSO, Inc. to analyze the wipe tests were maintained under RSO's laboratory quality assurance program which includes a service agreement with the manufacturer, daily quality control performance charts and background and standard samples counted with every sample batch.
Personnel Qualifications All personnel had levels of training and experience commensurate with their assigned tasks. For those individuals involved in taking radiological measurements and samples, special instruction was provided when necessary on equipment, special techniques, and practices relating to survey activities.
Laboratory Services Wipes or swabs were screened for gross gamma activity and further were analyzed for gross beta/gamma activity. All wipes for the final survey were analyzed by RSO, Inc. personnel.
2.10 Data Quality Objectives (MARSSIM)
The survey planning used the Data Quality Objectives (DQO) Process to ensure that the survey results are of sufficient quality and quantity to support the final decision. The use of the DQO Process assures that the type, quantity, and quality of environmental data used in decision making will be appropriate for the intended application. The DQO Process consists of seven steps, as shown below. The output from each step influences the choices that will be made later in the Process.
Rev 1-30-2015
1. State the problem: Radioactive materials (H-3, C-14, S-35 and P-33) primarily in the form of liquids was used in this facility. The use of these materials was strictly limited to work benches and chemical fume hoods in the impacted rooms. It is unlikely that the use of radioactive materials caused residual contamination at levels exceeding the activity DCGL (Derived Concentration Guideline) for H-3, C-14, S-35 and/or P-33. The amounts of radioactivity used at any 1 time was relatively small (millicurie or sub-millicurie amounts) and there were no spills or incidents resulting in widespread contamination ..
2. Identify the decision: Determine if residual radioactivity on structure surfaces of the laboratories and other areas where H-3, C-14, S-35 and P-33 were used with site-specific surface activity DCGLs derived as unrestricted release criteria to comply with dose limits prescribed in 10 CFR 20, Subpart E.
3. Identify inputs to the decision: Radiological survey data was collected for impacted structure surfaces.
4. Define the study boundaries: Historical analysis had identified in Table 4 as the impacted use areas.
5. Develop a decision rule: Given that sufficient data has been collected, if the mean concentration in the facility is less than the DCGL, then the facility is determined to be in compliance with the release criterion. Compliance with applicable DCGLs is demonstrated using the Sign and/or Wilcoxon Rank Sum (WRS) Tests to disprove the null hypothesis that the survey unit being evaluated exhibits contamination at concentrations exceeding the applicable DCGL.
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RSO, Inc.• lncyte Corporation• X-Station Buildings E336 and E400 Radiological Final Survey Report
6. Specify limits on decision errors: MARSSIM's guidance for determination of the number of samples needed for a survey unit when the DCGL is large, the relative shift is large (>2.5), using equal values of 0.05 for Type I and Type II errors, results in a number of data points needed of 12.
7. Optimize the design for collecting data: H-3, C-14, S-35 are low energy beta emitters and P-33 is a "medium" energy beta emitter. Survey equipment and sampling techniques were chosen that were appropriate and sensitive for the detection of the potential contaminates.
3.0 SURVEY INSTRUMENTATION
3.1 Description of Field and Laboratory Instrumentation and Sensitivity (also see Survey Data Sheets)
Field Instrument Used -
Ludlum Floor Monitor: consists of a cart mounted Ludlum Model 2221 with a Ludlum 43-37 probe (gas proportional detector, thin window of 0.8 mg/cm2 with an area of 583 cm2
)
Ludlum Model 2221 with a Ludlum 43-68 probe (gas proportional detector, thin window of 0.8 mg/cm2 with an area of 126 cm2
).
Laboratory Instrument Used -
Packard Tricarb Liquid Scintillation Counter for analysis of wipe tests.
3.2 Description of Instrumentation 1 bl 5 s t d t a e urvey me ers use d t d' I . I o con uc ra 10 0Q1ca surveys.
Survey Meter Detector Detector Probe
Use Model Type Area/Size
Floor Monitor Gas Flow
Ludlum Model Ludlum Proportional 584 cm2 Scans of Floors
2221 Scaler/Rate 43-37 meter
Detector
Ludlum Model Ludlum
Gas Flow Scans of Surfaces 2221 Scaler/Rate Proportional 126 cm2 Direct/Static
meter 43-68
Detector Measurements
3.3 Instrument Calibration and Efficiency Data
The calibration and efficiency data for the survey meters that were used during the Final Survey FS are summarized in Table 6.
1 bl 6 s /' 'b d ff . d a e urvey meter instrument call ration an e 1c1ency ata.
Meter w/ Probe Detector Model Radionuclide Efficiency (47t)
Ludlum Model 2221 Ludlum 43-68 C-14/S-35/P-33 20% cpm per dpm
Scaler/Ratemeter
Ludlum Model 2221 Ludlum 43-37 C-14/S-35/P-33 20% cpm per dpm
Scaler/Ratemeter
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Packard 3H -50% cpm per dpm Liquid Scintillation LS 14C -85% cpm per dpm
Counter 32P -95% cpm per dpm
4.3 Minimum Detectable Concentration for Scanning Technique
Beta Scans:
The minimum detectable concentration for the beta scans was calculated using the suggested method in NUREG-1507 and in Abelquist 2001 (See Equation 9.11 ).
Equation 1
Where:
Scan MDC =estimated minimum activity (dpm/100 cm2) that can be detected
MDCR
p = E; =
during a scan,
= Minimum detectable count rate, see Table 6.6 MARSSIM
surveyor efficiency considered to be 0.25
2n efficiency (c/d)
Es = surface efficiency Note: E; estimated assuming the 2n efficiency was approximately 2 times the 4n efficiency
Note: Es assumed to be 0.5
Table 7. Example of calculation of the MDC for Scanning (scan MDC).
.. *The C-14 instrument efficiency was used for the MDA and act1v1ty calculations as S-35 1s considered to be the same as that of C-14 given the beta energies are nearly equal and the instrument efficiency for P-33 is greater, given the higher beta energy and as demonstrated by the efficiency for Tc-99 {beta energy of 294 keV) shown on the instrument calibration certificates.
4.4 Static Measurement Data Reduction
Determinations of the total surface activity were based on static measurements with the detector in direct contact with the surface. For each analysis gross counts were converted into area activity concentration using the following method of data reduction:
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Where:
A = c = T = Rs = E = a =
Equation 2
A= E*( a ) 100cm2
total activity ( dpm/100 cm2),
integrated gross counts (counts),
count time (min),
background count rate (cpm),
total efficiency (c/d) *source efficiency
detector area (normalized to 100 cm2).
4.5 Minimum Detectable Concentration for Static Measurements
Using the equation shown below the minimum detectable activity for the static measurements was calculated using the following equation for instances in which the background and sample are counted for the same time intervals:
Equation 3
3+4.65/R; Static MDC=---------
K * (detector area) * T. 100cm 2 s+s
Using the equation shown below the minimum detectable activity for the static measurements was calculated using the following equation for instances in which the background and sample are counted for different time intervals:
Rev 1-30-2015
Equation 4 .--~~~~~~~~
3+3.29 Rs *Ts+sO+Ts+~) Static MDC= B
K *(detectorarea) * T 1 O<lsm2 s+s
Where:
Static MDC = activity (dpm/100 cm2),
C = integrated gross counts (counts),
Ts+s =
Ts =
Rs = K =
sample count time
background count time
background count rate (cpm)
proportionality constants e.g.: total efficiency
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RSO, Inc.• lncyte Corporation• X-Station Buildings E336 and E400 Radiological Final Survey Report
T bl 8 E a e f I I . f h MDC f S f M xamp e o ca cu at1on o t e or ta 1c easuremen s.
Detector Probe Efficiency Background Background
Survey Meter Model Area/Size
(cpm per Count Rate and Sample dpm) (cpm) Count Time
Ludlum 2221 Ludlum
1 minute (bkg) Scaler/ 126 cm2 0.21 207 1 minute
Ratemeter 43-68
(sample
Ludlum 2221 Ludlum
1 minute (bkg) Scaler/ 584 cm2 0.18 873 1 minute
Ratemeter 43-37
(sample)
Static MDC (dpm/100
cm2)
528
267
Note: Calculation uses equal count times of 1 minute for background and sample.
3.3 Laboratory Instrumentation Sensitivity
Laboratory Instrument Used - Packard Tricarb 3100 liquid scintillation counter.
The minimum detectable activity for H-3 on a wipe test was estimated to be less than 40 dpm for a 1-minute count time, 1-minute background count time, efficiency of 0.4 cpm/dpm and a background count-rate of 6 cpm.
The minimum detectable activity for C-14 on a wipe test was estimated to be less than 30 dpm for a 1-minute count time, 1-minute background count time, efficiency of 0.8 cpm/dpm and a background count-rate of <15 cpm.
The minimum detectable activity for P-33 on a wipe test was estimated to be less than 30 dpm for a 1-minute count time, 1-minute background count time, efficiency of 0.8 cpm/dpm and a background count-rate of 10 cpm.
4.0 SURVEY RESULTS
The radiological final survey showed residual contamination was less the DCGLs for the survey. Also all measurement results were less than the detection limits for the survey method except for several small areas of contamination. The licensee performed decontamination of these areas and they were resurveyed to complete the Final Survey.
4.1 Survey Results
Attachment A contains the survey results by survey unit. Results include: survey unit drawing annotated with survey locations for wipe tests and direct measurements, and instrument scan results (raw data shown).
Attachment B contains the LSC analysis data print-outs reports.
Attachment C contains the survey meter calibration reports.
4.2 Beta Scans and Direct Measurements-Summary
Final Survey: Small areas (<0.5 m2) of contamination were found during the beta scans.
The levels of contamination found were small fractions of the DCGLw for C-14. lncyte personnel performed decontamination of these areas.
Decontamination included scrubbing with a professional grade cleaner and as needed stripping down the surface material. As observed in Table 9., the techniques applied for removing the contamination significantly reduced the concentration levels in many of the locations. In Table 9., the first column shows the initial sample number before
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RSO, Inc. • lncyte Corporation • X-Station Buildings E336 and E400 Radiological Final Survey Report
decontamination, followed by the resulting measurements. The 5th column shows the corresponding sample number assigned after decontamination had been performed. Sample numbers 263/478, 257/476, 258/477, 393/479,394/480 were re-surveyed and wipe tested. It was determined that the five floor tiles had embedded contamination and were removed. Table 10. lists the pre and post concentration levels for these locations.
"J; bl 9 D A a e econtam1nat1on ttempts
Direct Direct 1 Post
DPM/100 Sam Decon Sample# Area Description 1 min. cm2 (C-14) pie#
16 Corridor, Floor,lile over Concrete 833 10447 46 239 734
21 Corridor, Floor,lile over Concrete 607 7031 51 193 379
4.4 Removable Contamination-Summary
Over 500 wipe tests were collected and were analyzed using liquid scintillation counting. None of the wipe test samples showed any removable activity above 200 dpm.
5.0 CONCLUSIONS
The Radiological Final Survey of the affected areas demonstrates that the surfaces were less than the DCGLw (25 mrem) for surface contamination and no contamination was detected that was greater than the facility operational limits for contamination.
The subject rooms/areas appear to meet the requirements for unrestricted use.
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6.0 REFERENCES
6.1 USNRC, Regulatory Guide 1.86., Termination of Operating Licenses for Nuclear Reactors, June 197 4.
6.2 USN RC, "Guidelines for Decontamination of Facilities and Equipment Prior to Release for Unaffected Use or Termination of Licenses for Byproduct, Source, or Special Nuclear Material", May 1987.
6.3 NUREG 1757, USNRC, "Decommissioning Process for Materials Licensees", Final September 2003.
6.4 NUREG-1575, EPA402-R-97-016, and Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM): Final, August 2000.
6.5 10CFR Part 20 § 20.1402 Radiological criteria for unrestricted use.
6.6 Abelquist, Decommissioning Health Physics, A Handbook for MARSSIM Users, IOP Publishing Ltd 2001, Philadelphia PA.
7 .0 ATTACHMENTS
Attachment A
Attachment B
Attachment C
Rev 1-30-2015
Radiological Survey Results
Wipe Test LSC Analysis Report Print-Out
Survey Meter Calibration Reports
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RADIOLOGICAL SURVEY RESULTS
Rev 1-30-2015 17
Survey Meter Information
Site: Inc e Building: Building 336
Meter1 Meter2 Meter3 Date: 11/10/2014 Not In Service 11/10/2014
B = Background Counts T = Counting Time In Minutes c1 =Total Detector Efficiency in Counts/Disintegration
A = Physical Probe Area in cm' C =Other Constants and Factors When Needed Ehr = Human Factor Efficiency c, = Source Efficiency s, = 1.38.SQRT(B,)
i = Counting Interval
18
Example of Survey Meter Calculations of MDCR, Scan MDC, and Direct MDC
Meter 1 Scanning Sensitivity Date: 11/10/2014 MARSSIM Value Description
Make: Ludlum MDCR Calculation Model: 2221 d' 1.38 detectabilitv value associated with the desired performance
SN: 161591 i 1 interval Probe Make: Ludlum b; 3.443 bi=background counts in interval=B*(1/60)
Probe Model: 43-68 s, 2.56 si=d'*SQRT(bi)
Probe SN: 118227 MDCR(net 154 MDCR(net)=Si*(60/i Probe Area (cm2
): 126 MDCR (qross) 360 MDCR(qross)=Si*(60/i)+B
Next Cal. Date: 11/7/2015 Scan MDC Calculation Backqround Surface Material Laminate p 0.5 Survevor Efficiencv Ehf or Efficiencv human factor) BackQround(c) - Time(Min)): 2066 10 Source Efficiency 0.25
CS Isotope - Activitv(mCi): C-14 0.159 Meter efficiency from Cal 0.21 CS Source(cpm) 8443 e; 0.105 Meter Efficiency from Cal * 2(2Pi factor) * Source Efficiency
L,. Ld (Counts) 33 70 e, 1 Surface Efficiency (decreases for porous surfaces) Direct MDC, Scan MDC
Door Metal <200 <200 ... , -<-2-00_..__ - -----+B-e_n_c_h_T_o_p_L_a_m-in-a-te-+------:;:250 <2oo <200
1Shelf Laminate <200 <200 <200 Shelf Laminate <200 - I <2oo <200 Bench Top Laminate <200 t--~~OQ- <200 Door Metal -- <200 I <200 +-, _<_2_0_0--+
I Bench To <200 _ r- <?:QQ .,__<_20_0_ .. ___ _
1 Bench To Laminate ~1g§-+-~§~ :~~~ I Bench To Laminate <200 I <200 <200 1Drawer Metal - <26o 1 =-~?o()__ <200 I Cabinet Under Hood Metal <200 I <200 ,.__<_2_0_0--+
Floor Tile Over Concrete <200 Tile Over Concrete <200 Tile Over Concrete <200 Tile Over Concrete ·<200·
""~"-"~"--··-"--
Tile Over Concrete <200 Tile Over Concrete <200 Tile Over Concrete <200 <200 Tile Over Concrete <200 <200 Tile Over Concrete ···:;:250·· <266 Tile Over Concrete <200 <200
Metal <200 <200 Vacuum Filter Metal -·:;:2·55····- <200
Laminate <200 ··~-··-··---····-
Laminate <200 Laminate <200 Laminate <200
Bench To Laminate <200
··------·-··--····-·- Bench Top Laminate ----<2oo· Bench Top Laminate <200 Sink Drain Metal --<2oo· Sink Metal <200
<200 <200 <200
<200 . <200 ··1-:"'.""".,....,:-----:""."'""-,---t····· <200 t. <200
<200 r <200
Under Sink Metal Sink Drain Metal
I Sink Basin Metal .Under Sink Metal <200
Building: _4_00 __ Lab/Room: 3407
Finish Date: 11/11/14 --------Surveyor: Matthew Mueller
<200 . L_::~Q.o <200 I <200 t---:=,..-+· :· 'Sink Drain Metal
157 Sink Basin Metal 158 Under Sink Metal 159 JUnder Fume Hood Metal
Rev 1-30-2015
<200 - t·. <200 ... }
<200 f <200 <200 <200
<200 <200 <200
173 151 156
·f·· -133 I 224 -299 t 224 +- 224 -262 i
J 150 J. I 150 i-
150
-307 -307 ~367
33
Site: lncyte
Start Date: _1c..;1;..;/1;..;1;..;/1'-4 _______ _
Surveyor: Korreesa Williams
Area Surve Results Wi e Tests
Wipe Number
p p p
Description (dpm/100 (dpm/100 (dpm/100
cm2} cm2
) cm2)
H-3 C-14 High E
'Door Metal <200 Fume Hood Base Metal <200 Fume Hood Base Metal <200 Fume Hood Base Metal <200 Fume Hood Wall Metal ---<266--Fume Hood Back Wall Metal <200
'"!Fume Hood Back Wall Metal "~""""'<200 v-·
Fume Hood Right Wall Metal ···"~<2Qo" Fume Hood Cover Metal <200 Fume Hood Lip Metal <200
RSO, Inc. RSO Job No. Rl 1154 P.O. Box 1450 Laurel, MD 20725 (301) 953-2482 Certificate of Calibration
ISSUED TO: RSO, Inc. 5204 Minnick Road Laurel, MD 20707
CONTACT: Dave Wellner PHONE: (301) 953-2482 PO NO: RSO 299
INSTRUMENT: LUDLUM MODEL: 2221
TYPE: RATEMETER
SN: 161591
RSO, Inc. certifies that on 11/07/2014 the above described instrument was calibrated using a radioactive source to determine the efficiency for a specific radionuclide(s) and using electronically generated pulse for the linearity. Pulsed using Ludlum 500-2, SIN 159Jl0.
Notes
The results are tabulated below. Calibration is traceable to NIST.
THE SUGGESTED RECALIBRATION DATE FOR THIS INSTRUMENT IS 11/07/2015 Calibrated By: ~ /, \.P ± /.U'p~
-~'"'-""""-' Reviewed By: V't"~ ~ racy Austin.
Cal Date: 11/0712014
Maryland License MD-33-021-01
Rev 1-30-2015
16004
83
RSO, Inc. RSO Job No.R10942 P.O. Box 1450 Laure~ MD 20725 (301) ~53-2482 Certificate of Calibration
ISSUED TO: RSO, Ine. 5204 Minnick Road Laurel, MD 20707
CONTACT: Dave Wellner PHONE: (301) 953-2482 PO NO: RSO 299
INSTRUMENT: LUDLUM MODEL: 2221
TYPE: RATEMETER
SN: 89650
RS01 Inc. certifies that on 06/1112014 the above described instrument was calibrated using a radioactive source to determine the efficiency for a specific radionuclide(s) and using electronically generated pulse for the linearity. Pulsed using Ludlum 500-2, S/N 159110.
Notes
MODEL
The results are tabulated below. Calibration is traceable to NIST.