Berlex Laboratories, Inc., Deficiency Response Ltr. dtd 03 ... · the study. Figure 3 shows the...

DAQ Inc. Health and Safety Consultants RADIOLOGICAL ENVIRONMENTAL OCCUPATIONAL

March 11, 2005

Steve Hammann US Nuclear Regulatory Commission, Region I 475 Allendale Road King of Prussia, PA 19406-1415

Subject: Berlex Laboratories, Inc. Final Status Survey: Control No. 136163

Dear Mr. Hammann,

On behalf of Mr. Santo Guillermain of Berlex, I am forwarding you the Final Status Survey Report for Berlex Laboratories, Inc., in Wayne, New Jersey, License No. 29- 30621-01. Please contact Mr. Guillermain or me if you have any questions or comments.

Regards,

0 u w ; l . L Dennis M. Quinn, CHP

3 Shadow Lane Hopewell Junction, NY 12533

[email protected] Phone: 845-223-1960 Fax: 845-223-1099 / 36 id3

' 7 ' - n ! r l , ? l t l IZATERIALS-G;? 1 , , . . , . , I l , 1

mailto:[email protected]

FINAL STATUS SURVEY

NRC License No. 29-30621-01

Prepared For: Berlex Laboratories, Inc.

300 Fairfield Road Wayne, New Jersey 07470

Prepared by:

DAQ, Inc. Health and Safety Consultants RADIOLOGICAL ENVIRONMENTAL OCCUPATIONAL

DAQ, Inc. Health and Safety Consultants RADIOLOGICAL ENVIRONMENTAL OCCUPATIONAL

March 2005

.

Final Status Survey March 2005 Berlex Laboratories. Inc .

TABLE OF CONTENTS

TABLE OF CONTENTS .............................................................................................. i

ACRONYMS AND ABBREVIATIONS ..................................................................... i1

1.0 INTRODUCTION ............................................................................................. 1

1.1 HISTORICAL REVIEW AND RADIONUCLIDES OF CONCERN .................................. 1 1.2 RADIOACTIVE MATERIAL LABORATORY ACTIVITIES ......................................... 2 1.3 DECOMMISSIONING ACTIVITIES ........................................................................ 2

DATA QUALITY OBJECTIVES ..................................................................... 3

2.1 STEP 1: STATE THE PROBLEM ........................................................................... 3 2.2 STEP 2: IDENTIFY THE DECISION ...................................................................... 3

2.2. I Principal Study Question ............................................................................. 3 2.2.2 Decision Statements ..................................................................................... 3

2.0

2.3 STEP 3: IDENTIFY INPUTS TO THE DECISION ...................................................... 3 2.4 STEP 4: DEFINE THE STUDY BOUNDARIES ........................................................ 4 2.5 STEP 5 : STATE THE DECISION RULES ................................................................ 4

2.5. I Decision Rules ............................................................................................. 4 2.6 STEP 6: DEFINE ACCEPTABLE DECISION ERRORS .............................................. 4

SURVEY DESIGN AND METHODOLOGY .................................................. 5

3.1 DETERMINE IMPACTED OR NON-IMPACTED ....................................................... 6 3.2 SURVEY UNIT BREAKDOWN .............................................................................. 6 3.3 RELEASE CRITERIA ........................................................................................... 6 3.4 DETERMINE THE NUMBER OF DISCRETE SAMPLE LOCATIONS ............................. 7

3.0

3.4.1 Relative Shift ................................................................................................ 7 3.4.2 Survey Unit Grid Spacing ............................................................................ 7

3.5.1 Sur$ace Scan Surveys ................................................................................... 9 3.5.2 Fixed-point Measurements .......................................................................... 9 3.5.3 Smear Sample Collection and Analysis ........................................................ 9 3.5.4 Systems Surveys ......................................................................................... I O

4.0 SURVEY RESULTS ........................................................................................ 10

3.5 SURVEY METHODS AND INSTR~MENTA~ON ...................................................... 9

5.0 REFERENCES ................................................................................................ 14

APPENDIX A: Survey Drawings and Photos

APPENDIX B: Survey Instrumentation Quality Assurance

Page i

Final Status Survey March 2005 Berlex Laboratories, Inc.

ACRONYMS AND ABBREVIATIONS

cm2 CPm A DCGLw dPm DQO FSS hr HSA LBGR LSC m2 MARSSIM MDC mrem NIST NRC QA QC ROC

SOR su Yr

0

square centimeters counts per minute delta Derived Concentration Guideline Level (average) disintegrations per minute Data Quality Objective Final status survey hour Historical Site Assessment Lower Bound of the Grey Region liquid scintillation counter square meters Multi- Agency Radiation Survey and Site Investigation Manual Minimum Detectable Concentration millirem National Institute of Standards and Technology U.S. Nuclear Regulatory Commission Quality Assurance Quality Control Radionuclides of Concern sigma Sum of the Ratios Survey Unit year

Page ii

Final Status Survey Berlex Laboratories. Inc.

March 2005

1 .O INTRODUCTION

Berlex Laboratories, Inc. (Berlex) has decided to stop all use of radioactive materials at Berlex’s Wayne, New Jersey facility. This requires termination of the Berlex Radioactive Materials License with the U. S. Nuclear Regulatory Commission (NRC). The current license is NRC License No. 29-30621-01, Amendment No. 6. The NRC provides guidance regarding decommissioning leading to termination of a license in NUREG- 1757, Vol. 1 Consolidated NMSS Decommissioning Guidance (NRC 2002). All references to license termination guidance in this document refer to NUREG- 1757, Vol. 1, unless otherwise specified.

NUREG-1757 identifies 7 types of facilities based on the potential amount of residual contamination, the location of the contamination, and the complexity of the activities needed to decommission the site. Berlex is considered a Group 1 facility because of its use of short-lived radionuclides and sealed sources. No long-lived unsealed radioactive material has been used. Residual contamination, if any, is expected to be less than the screening criteria (Appendix B of the NUREG).

The NUREG does not require a Decommissioning Plan for a Group 1 facility, and the radioactive material license does not specifically require a Decommissioning Plan.

1.1 Historical Review and Radionuclides of Concern

The first date of radioactive material use at Berlex was after receipt of the radioactive materials license fiom the NRC on March 29, 2001. The most current amended (Amendment No. 6) license allows for the possession of the following isotopes:

1. Molybdenum-99 10 Curies

2. Technetium-99m 10 Curies

3. Barium-133 250 microcuries (sealed source)

4. Cesium- 137 250 microcuries (sealed source)

Samarium-153 was briefly on the license for storage and transportation, but no Samarium-153 was ever present on the site, and it has since been removed fiom the license.

Molybdenum-99 (Mo-99) has a 66 hour half-life, and it decays to either Technetium-99 (Tc-99, 11%, 0.293 MeV beta) or Tc-99m (89%, 0.140 MeV). While Tc-99 is a product of the decay of Mo-99, it is produced in very low activity amounts. Although Tc-99 is included as a radionuclide of concern, due to the low amount of radioactivity produced in the decay process, it is not expected that any Tc-99 will be detectable. Table 1 provides information on the radionuclides of concern.

Page 1

Final Status Survey March 2005

Various up to 0.78 MeV, - 30%abundance 1 M e 9 9 I 66 hours I

1.2

0.44 MeV, and 1.21 MeV

1.3

Berlex Laboratories, Inc.

Table 1. Radionuclides of Concern

I Tc-99m I 6 hours 1 0.140 MeV None I I Tc-99 I 213,000 yr I None I 0.293 MeV

Radioactive Material Laboratory Activities

The Radioactive Material Laboratory was utilized to support the quality control testing of two pharmaceutical products which, when administered to patients, were labeled with Tc- 99m. Tc-99m was eluted from a Mo-99mc-99m generator. On a frequency of every one to two weeks, a calibrated Mo-99/Tc-99m generator (3665 millicuries) was received at the facility. Generally, the generator was calibrated by the supplier on a Friday and received by Berlex on the following Monday, thereby containing a decay-corrected activity of approximately 1700 millicuries. Mallinckrodt Medical, Inc supplied the generators, and the depleted generator was returned to Mallinckrodt at the end of its useful life. This return was generally accomplished every week, or every other week, although the time frame was sometimes longer depending upon whether or not a new generator was to be received the following Monday. On one occasion, Tc-99m solution was delivered to the facility for that day's use.

The sealed sources (Ba-133 and Cs-137) were used in the Radioactive Materials Laboratory for the dose calibrator and the Caprac well counter.

The last Mo-99mc-99m generator received was calibrated as 3665 millicuries on October 8, 2004, and received as 172 1 millicuries on October 1 1, 2004. This generator's last use was October 14, 2004, and it was shipped back to Mallinckrodt on December 15, 2004. The remaining activity, even if left in the laboratory, would have decayed to less than one microcurie by January 1,2005.

Decommissioning Activities

The NRC was notified of the intention to close out the radioactive materials program at this Berlex site on December 15, 2004. The New Jersey department of Environmental Protection, Radiation Protection Program was notified of the close out intention on January 10, 2005. Cleanup of the laboratory began on January 14, 2005, and the sealed sources were shipped offsite. On March 2, 2005, a final status survey was performed, and the results are provided in a later section of this document.

Page 2


2.0 DATA QUALITY OBJECTIVES

2.1 Step 1: State the Problem

Radioactive material was used within one laboratory area at Berlex Laboratories. The objective of Final Status Survey (FSS) activities is to obtain data of sufficient quality and quantity to support unrestricted release of the portion of the Berlex facility use for radioactive materials, and to support license termination by the US Nuclear Regulatory Commission (NRC).

2.2 Step 2: Identify the Decision

2.2.1 Principal Study Question

Do the radionuclides of concern (ROC) concentrations in the Berlex Laboratory exceed applicable levels for unrestricted release?

2.2.2 Decision Statements

The following statements assume that ROC concentrations inside buildings exceed release levels. If ROC concentrations inside laboratories do not exceed the derived concentration guideline limit (DCGL,), the laboratories will satisfy the release criterion.

a. Determine whether survey unit (SU) ROC concentrations inside buildings exceed background concentrations by more than the applicable release criteria.

b. If survey unit ROC concentrations inside buildings exceed background by more than the applicable release criteria, then affected SUs must be remediated to levels satisfying the release criteria.

2.3 Step 3: Identify Inputs to the Decision

This section lists data needs, describes the sources of that data, and discusses the means of obtaining the required data. The following site characteristics must be determined in order to resolve applicable decision statements:

Concentrations of residual radioactive material in the survey units will be determined by means of:

Direct surface radioactivity measurements,

Removable activity concentration measurements (liquid scintillation counter results),

Gamma Exposure Rate Surveys

Page 3


2.4 Step 4: Define the Study Boundaries

The population of interest is the concentrations of ROCs on building and laboratory surfaces, including floors, benchtops, hoods, lab drawers, and sinks. The population of interest study is horizontally and vertically limited to impacted areas located inside the Radioactive Material Laboratory (RAM Lab). On facility drawings, the RAM Lab is also designated as Isolation Lab No. 2, and D-112. See Figure 3 for information on surrounding laboratories and hallways.

The pathway from the generator delivery/warehouse area to the RAM Lab and the pathway from the RAM Lab to the used generator pickup location will be considered as part of the study boundaries (See Figure 2). For a period of time, one room was used to store the used generators awaiting shipment, and this room will be considered as part of the study. Figure 3 shows the pathway from the RAM Lab to the used generator pickup location.

2.5 Step 5: State the Decision Rules

2.5.1 Decision Rules

2.5.1.1 Surface Radioactivity Scan Surveys

If areas of elevated radioactivity are identified during scan surveys, the radioisotope will be identified, if possible, and contaminated areas will be decontaminated as appropriate and re-surveyed. Smear samples will also be collected and analyzed.

2.5.1.2 Residual Radioactivity

If residual radioactivity is found in an isolated area of elevated activity, in addition to residual radioactivity distributed relatively uniformly across the survey unit, the unity rule, also called the Sum of the Ratios (SOR), will be used to ensure that the total dose is within the decommissioning guidance (NRC, 2000). When multiple contaminants are present on a site, site radiological conditions are evaluated using the SOR and a DCGL, of SOR=1 .O. The SOR is calculated as follows:

Concentration of Radionuclides - Where: Cl,2,3.. -

DCGL1,2,3.. - - DCGL for that Radionuclide

2.6 Step 6: Define Acceptable Decision Errors

NRC guidance provides a discussion regarding decision errors (NRC 2000). This

Page 4


discussion includes the concept that acceptable error rates, which balance the need to make appropriate decisions with the financial costs of achieving high degrees of certainty.

Errors can be made when making site remediation decisions. The use of statistical methods allows for controlling the probability of making decision errors. When designing a statistical test, acceptable error rates for incorrectly determining that a site meets or does not meet the applicable decommissioning criteria must be specified. In determining these error rates, consideration should be given to the number of sample data points that are necessary to achieve them. Lower error rates require more measurements, but result in statistical tests of greater power and higher levels of confidence in the decisions. In setting error rates, it is important to balance the consequences of making a decision error against the cost of achieving greater certainty.

Acceptability decisions are often made based on acceptance criteria. If the mean and median concentrations of a contaminant are less than the associated acceptance criteria, for example, the results can usually be accepted. In cases where data results are not so clear, statistically based decisions are necessary. Statistical acceptability decisions, however, are always subject to error. Two possible error types are associated with such decisions.

The first type of decision error, called a Type I error, occurs when the null hypothesis is rejected when it is actually true. The probability of a Type I error is usually denoted by a. Considered in light of the null hypothesis used for this investigation, this error could result in higher potential doses to future site occupants than prescribed by the dose-based criterion. The maximum Type I error rate is 0.05.

The second type of decision error, called a Type I1 error, occurs when the null hypothesis is not rejected when it is actually false. The probability of a Type I1 error is usually denoted by p. The power of a statistical test is defined as the probability of rejecting the null hypothesis when it is false. It is numerically equal to 1-p where p is the Type I1 error rate. Consequences of Type I1 errors include unnecessary remediation expense and project delays.

For the purposes of this FSS, the acceptable error rate for both Type I and Type I1 errors is five percent (Le., a = p = 0.05). These acceptable error rates were only used to develop the number of fixed survey points and the number of contamination samples necessary for the primary survey unit. For scans, fixed point measurements, and removable contamination checks, any result above the release limit will be evaluated and decontaminated as necessary. Therefore, statistical analysis of sample results will not be performed.

3.0 SURVEY DESIGN AND METHODOLOGY

Page 5

3.1


Determine Impacted or Non-Impacted.

Based on the review of radioisotope use, the areas considered to be impacted are the RAM Lab, the pathway from the generator delivery/warehouse area to the RAM Lab and the pathway from the RAM Lab to the used generator pickup location. The room that had been used to store the used generators awaiting shipment is considered part this path and is considered to be impacted. See Figures 2 and 3.

Impacted areas were then classified based on contamination potential as per guidance in the MARSSIM sections 2.2,4.4, 5.5.2, and 5.5.3. (NRC 2000) Namely,

0 Class 1: The area had been contaminated above the release criteria, and it is possible to find radioactivity above the release criteria; Class 2: The area had radioactive material use, but it is unlikely to have radioactivity above the release criteria; Class 3: The area had some use of radioactive material, but it is very unlikely to have radioactivity above the release criteria.

Based on the potential for contamination in 2005, the RAM Lab is considered a Class 2, and all other impacted areas are considered Class 3.

3.2 Survey Unit Breakdown

There are two survey units for this FSS; one survey unit is the RAM Lab, and the second survey unit is all other pathways described in section 3.1.

3.3 Release Criteria

The screening criteria of NUREG 1757, Vol. 1 Appendix B are applicable; however, since Mo-99 and Tc-99m have such short half-lives, they are not listed in the NUREG. Tc-99, while unlikely to detect due to its low production, is listed for completeness. However, because that value is fairly high, and will be very easy to detect (1,300,000 d p d l 00 cm’), a more conservative value is selected for this FSS (see Table 2).

Since specific guidance is not provided for short-lived radionuclides, Table 2 presents conservatively low contamination values that are the Derived Concentration Guideline Levels (DCGL). If these values are exceeded during the survey, then the area will be remediated, or an evaluation will be performed to ensure that the residual radioactivity is within 25 mrem per year.

Page 6


March 2005

Table 2. DCGLs for Berlex Final Status Survey

Tc-99, T~-99m, Mo-991 5,000 15,000 1,000

3.4 Determine the Number of Discrete Sample Locations

3.4.1 Relative Shift

The relative shift describes the relationship of site residual radionuclide concentrations to the DCGL, and is calculated using the following equation, found in Section 5.5.2.2 of MARSSIM (NRC 2000):

A - DCGL, - LBGR 0 0 - _

where:

DCGL, = the derived concentration guideline level (i.e., release limit)

LBGR = concentration at the lower bound of the gray region. The LBGR effectively becomes the survey’s action level. For conservatism, the LBGR will be set to 0.5 times the DCGL, for this FSS.

0 = An estimate of the standard deviation of the concentration of residual radioactivity in the survey unit (which includes real spatial variability in the concentration as well as the precision of the measurement system). 0 is estimated as 0.3 times the DCGL,.

Therefore, a relative shift for the Berlex FSS is calculated as 0.5 / 0.3 = 1.6. Section 2.6 establishes the acceptable decision errors a=p=0.05. Based on these acceptable decision errors, the minimum number of measurement locations in the SU was calculated to be 17, including the MARSSIM recommended 20% additional samples to protect against the possibility of lost or unusable data (Table 5.3 or Table 5.5 from the MARSSIM document. For practical purposes, the number was raised to 20 in the Berlex RAM Lab.

3.4.2 Survey Unit Grid Spacing

Page 7


Grid spacing and placement of fixed-point measurement locations within the RAM Lab survey unit will be based on a relative coordinate system. The starting point is selected and the data points are located within the survey unit using a rectangular grid. In this survey the starting point was biased to ensure that most of the benchtops and hoods were part of the fixed-point surveys. Spacing of the rectangular grid is 5 feet by 7 feet. This can be seen in Figure 1 where the systematic sample locations are shown.

For surveys outside the RAM Lab, the fixed-point measurements were biased along the travel route and at the loading dock and storage area. No specific grid spacing was used.

Page 8


March 2005

3.5 Survey Methods and Instrumentation

3.5.1 Surface Scan Surveys

RAM Lab floor, bench, and wall surfaces were surveyed utilizing a Ludlum Model 2360 meter with a Ludlum 43-68 gas proportional detector (100 cm’), and a Ludlum Model 12 meter with a gamma scintillation detector (1” x 1” Sodium Iodide, Ludlum Model 44-2).

The minimum number of fixed-point measurements per SU is in accordance with the SU’s classification, 20 samples in the RAM Lab, and 20 samples outside the RAM Lab. If background response varies greatly during scan survey activities, instrument background measurements may be performed at representative measurement locations. During this survey, it was determined that background measurements were higher at the floor level.

If a survey results in a sustained instrument response above the ambient level, it will be considered an action level, and the following additional data will also be collected:

0 A 1-minute static beta-gamma measurement at the location; and

A ‘Biased’ smear sample to be analyzed for gross alpha-beta activity.

3.5.2 Fixed-point Measurements

Fixed-point measurements are defined as static counts performed with a portable instrument at locations of suspected contamination. Fixed-point measurements should be at least 1-minute in duration with the instrument placed in scaler mode. If the instrument does not have scaler capability, a count rate indication is acceptable provided a 1-minute static reading is taken with the instrument’s meter set to “slow response.” The highest indicated count rate shall be recorded as the result of the fixed-point measurement.

The fixed-point (static) survey action level will be the DCGL plus the average background specific to the material type being evaluated.

3.5.3 Smear Sample Collection and Analysis

Selected building surfaces will have smear samples taken to assess the presence of removable contamination. Smears will be taken at every location that a fixed-point measurement is taken and at any other locations deemed necessary based on visual inspections and professional judgment. The amount of removable radioactive material per 100 cmz of surface area will be determined by wiping that area with a dry filter, applying moderate pressure, and counting the smears.

Page 9


March 2005

3.5.4 Systems Surveys

The only service system being investigated under this FSS are the ventilation hoods and the sink drains. Surveys will include smears for removable contamination and fixed- point measurements in accessible areas, where practicable. Scans of sinks and sink drain traps will be performed using a gas proportional detector and a gamma scintillation probe. Release criteria for these areas are the same as those being used for building surfaces; see Table 2.

4.0 SURVEY RESULTS

4.1 Instrumentation data

The instruments used in this survey have been discussed in section 3.5, and the instrumentation quality assurance checks are shown in Appendix B, along with the liquid scintillation data from the analysis contractor.

4.2 Results Summary

The survey was performed on March 2, 2005 by Dennis M. Quinn, CHP. Liquid Scintillation results were performed by CoPhysics Corporation, and the results are shown in Appendix B. The results of the survey are presented in Tables 3,4, and 5.

Both systematic and biased surveys in the RAM Lab do not indicate any scan readings above background for either the gas proportional detector or the gamma scintillator. 100 percent of benchtops, floors, and walls to 6 feet in height were scanned. Particular attention was paid to areas most likely to have been contaminated in the past, including benchtops near the generators, sinks, sink drains, and hood surfaces. Likewise, the contamination results indicated background levels of radioactivity. Results of the systematic surveys are shown Table 3, and the biased locations are shown in Table 4.

For areas outside the RAM Lab, survey results (Table 5 ) indicate that gamma scintillation scans, gamma scintillation fixed point readings, and removable activity measurements are all at background levels.

In summary, systematic surveys and biased surveys within the RAM Lab, and biased surveys outside of the RAM Lab indicate near background levels of radioactivity, and results are less than the Derived Concentration Guideline Levels discussed in Table 2.

Page 10

Final Status Survey Berlex Laboratories, Inc.

I 3

7

7

March 2005

Table 3. RAM Lab Fixed Point Measurements - Systematic Survey

*Removable Contam.

dpm/100cm2

2

0

A

4

8

A

3

1

10

4

6

7

6

Removable contamination is a smear counted on a Liquid Scintillator. The highest of gross beta, tritium, or carbon-14 is used as the dpmll00 cm2

Page 1 1


March 2005

Table 4. RAM Lab Biased Measurements

* Removable contamination is a smear counted on a Liquid Scintillator. The highest of gross beta, tritium, or carbon-I4 is used as the dpm/100 crn2

back round

back round

back round**

background I 4 I I

background I 2 1

** Nal background generally 1200 - 1500 cpm, but less near lead, higher near floor

Page 12


2000

1900

1900

1700

1650

1800

1600

1600

March 2005

3

4

5

1

8

4

1

5

Table 5. Survey Results Outside RAM Lab

33

34

35

Location

~

Foyer between labs and warehouse

Floor in walkway of warehouse


Description

Outside Lab D-1 1 1 - floor

Outside RAM Lab in hallway floor

1600

1200

1700

2200

2300

2100

21 00

2100

2050

1800

1900

8

6

2

6

7

9

4

3

1

5

7

36

37

38

39

40

41

42

36

37

38

39

40

41

42

43

44

45


Floor just inside rollup door between warehouse areas

Floor on outside of rollup door between warehouse areas

Floor - near location where generators were temporarily stored upon arrival

Floor - near location where generators were temporarily stored upon anival

Concrete next to truck bay, in warehouse


Floor just inside rollup door between warehouse areas

Floor on outside of rollup door between warehouse areas

Floor - near location where generators were temporarily stored upon arrival

Floor - near location where generators were temporarily stored upon anival

Concrete next to truck bay, in warehouse

Metal in truck bay, in warehouse

In room next to building entrance, where used generator had been stored, prior to shipping


Foyer outside security office

Metal in truck bay, in warehouse



Foyer outside security office

Steps near building entrance and security office

Floor in hallway from building entrance to RAM Lab

Floor in hallway from building entrance to RAM Lab

Hallway outside RAM Lab

I Hallway outside RAM Lab

*Removable

Removable contamination is a smear counted on a Liquid Scintillator. The highest of gross beta, tritium, or carbon-14 is used as the dpm/100 cm2

Page 13


5.0 REFERENCES

(Berlex, 2004) NRC License No. 29-3062 1-01, Amendment No. 6.

(NRC, 2000)

(NRC, 2002)

NUREG- 1575, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM), U.S. Nuclear Regulatory Commission, August 2000.

NUREG- 1757, Consolidated NMSS Decommissioning Guidance, U.S. Nuclear Regulatory Commission, August 2002.

Page 14


APPENDIX A

Survey Drawings and Photos

Page 15

4

cn 0

0

c

a3

a

E

m 4

4

4 4 i 4

Figure 2. Path from Generator Receipt to RAM Lab

Final Status Survey Plan Berlex Laboratories, Inc.

APPENDIX B

Survey Instrumentation Quality Assurance

Appendix B

Final Status Survey Plan Berlex Laboratories. Inc.

75.000 cpm (contact) prior to survey

background = 1450 cpm

75,000 cpm (contact) after the survev 3:lO PM

10: 12 AM

Instrumentation used for Berlex Survey, 3-2-05

3/2/2005

3/2/2005

Ludlum Model 12

ISer. No. 83334 lA83334 I

Cs-137, 1.35 uCi. Ludlum 44-2 Nal 12/72 NEN

9056 cprn (1 minute count under spacer) background =230 cpm (1 minute count)

8265 cpm (1 minute count under spacer) background = 206 cpm ( 1 minute count)

ISer. No. 141322 IPRl60117 I

10:17 AM 3/2/2005

2:02 PM 3/2/2005

Ludlum Model 2360

background = 1400 cpm

C-14, 0.148 uCi, Ludlum 43-68 10/2/79 NEN

Appendix B

03/05/2005 14:2l FAX 845 783 7101 CoPhvsics Core -B h M 1 S

Client: DAQ, Inc. Am: Dennis Quinn 3 Shadow Lane Hopewell Junction, NY 12533

SAMPLE DESCRIPTION

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31 32 33

Wipe Wae Wipe Wipe Wipe Wipe Wipe Wipe Wipe Wipe Wipe Wipe Wipe Wipe Wipe Wipe Wipe Wipe Wipe Wipe Wipe Wipe Wipe Wipe Wipe wipe Wipe Wipe Wipe Wipe Wipe Wipe Wipe

RADlOANALY TICAL RESULTS

Lab # 464

Gross Alpha fdpmlwipe)

-0.2 -0.2 0.8 1.2 -0.2 4.2 0.5 0.1 0.8 0.1 -0.2 0.8 -0.2 0.1 0.1 0.8 0.1 -0.2 0.5 0.1 0.8 0.8 -0.2 1.2 0.8 -0.2 0.5 0.1 1.2 0.1 0.5 0.1 0.5

03 03 0.7 0.9 0.3 0.3 0.6 0.4 0.7 0.4 0.3 0.7 0.3 0.4 0.4 0.7 0.4 0.3 0.6 0.4 0.7 07 0.3 0.9 0.7 0.3 0.6 0.4 0.9 0.4 0.6 0.4 0.6

Gross Beta (d W w i p e )

2 5 2 0 + _ 2 4 + 3 4 2 3 8 5 3 5 2 3 4 + 3 3 t 3 7 ?. 3 7 2 3 3 2 3 1 + 2 9 2 3 2 + 2

7 1 : 3

8 2 3 2 t 2 8 + 3 8 ~ 3 4 2 3 6 2 3 3 + 2 6 ~ 3 2 2 2 6 + 3 3 + 3 6 2 3 4 + 3 3 _ * 2 4 + 3 5 5 3

6 2 3

6 + 3

igjUlJ1

Project. Berlex survey Type: Wipe Tests Date Collected: 32/05 Oak Analyzed: 3m05 Report Date: 3/5/05

Tritium (dpm/wipe)

0 + 1 0 2 1 O t l 0 + . 1 0 + 1 4 2 2 1 2 2 0 2 1 6 + 3 6 2 3 2 z 2 O t l 10 * 4 4 + 3 2 5 2 4 5 3 0 2 1 7 2 3 0 2 1 4 + 2 5 2 3 0 2. 1 5 5 3 0 + 1 0 - 1 1 2 1 3 2 2 0 + 1 3 + 3 0 2 1 0 + 1 O _ t l 3 , 2

Carbon-14 (dpdwipe)

0 5 1 0 5 1 4 + 2 o z 1 0 5 1 5 2 2 4 5 2 0 + 1 0 + 1 o z 1 0 2 1 0 2 1 0 2 1 0 + _ 1 0 2 1 4 2 2 0 2 1 0 2 1 O t l 3 t l 0 + 1 0 2 1 3 z 1 0 + 1 4 + 2 0 2 1 0 2 1 0 + 1 0 + 1 0 5 1 0 2 1 0 5 1 0 2 1

Page 1 of 2

SAMPLE DESCRIPTION

34 Wipe 35 Wipe 36 Wipe 37 Wipe 38 Wipe 39 Wipe 40 Wipe 41 Wipe 42 Wipe 43 Wipe 44 Wipe 45 Wipe 46 Wipe 47 Wipe 48 Wipe 49 Wipe 50 Wipe 51 Wipe 52 Wipe

Gross Alpha

0.1 2 0.4 0.5 $ 0.6 0.8 5 0.7 0.1 2 0.4 0.1 2 0.4 0.1 2 0.4 0.8 ? 0.7

0.1 2 0.4

0.1 : 0.4 0.5 5 0.6 0.1 2 0.4 0.5 0.6 0.5 5 0.6

0.5 2 0.6 0.1 2 0.4

(dpmhipe)

-0.2 2 0.3

-0.2 0.3

-0.2 2 0.3

-0.2 t 0.3

Gross Beta (dpmlwipe) 1 2 2 8 + 3 4 + 3 1 2 2 5 2 3 6 2 3 a t 3 6 2 3 2 5 2 6 2 3 7 + 4 9 2 4 4 5 3 3 2 3 1 5 2 5 2 3

4 + 3 2 + 2

6 % 3

Tntium (dpmlwipe)

0 5 1 7 - + 4 0 5 1 0 1 1 1 0 + 1 7 5 4 4 5 3

O f 1 2 + 2 0 2 1 5 2 3 2 5 2 0 - + 1 O f 1 4 5 3 7 5 3 0 2 1 0 + . 1

6 + 3

MDAs 2.3 dpm 8 dprn 5 dprn

Method: pulse shape, liquid scintillation counting Instrument: Wallac Model 1415 LSC, Serial #4150043 Uncertainlies are 1 -sigma counting and quench cwrection errors. MOA - Minimum Detectabk Activity Standards are traceable to the National Institute of Standards and Technology. Radioactive Materials License: NYS 2691-3949. expires 7/31/07.

Carbon-14 (dpm/wiw) 0 2 1 0 2 1 0 + 1 0 + 1 5 2 2 3 2 1 0 2 1 0 + 1 I t 1 0 2 1 0 + 1 0 2 1 0 + 1 O t l O Z 1 0 2 1 0 + 1 0 + 1 3 + 1

2 dpm

Theodore E. Rahon, Ph.0.. CHP President

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