ML031490485

IR 05000160-02-201, on 10/21/2002 Through 10/23/2002, for Georgia Institute of Technology, Atlanta, Ga

ML031490485
Person / Time
Site:	Neely Research Reactor
Issue date:	06/24/2003
From:	Madden P NRC/NRR/DRIP/RORP
To:	Hertel N Neely Research Reactor
Holmes S, NRC/NRR/DRIP/RORP, 415-8583
References
IR-02-201
Download: ML031490485 (30)

See also: IR 05000160/2002201

Text

June 24, 2003

Dr. Nolan Hertel, Director

Neely Nuclear Research Center

Georgia Institute of Technology

900 Atlantic Drive

Atlanta, GA 30332-0425

SUBJECT:

NRC INSPECTION REPORT NO. 50-160/2002-201

Dear Dr. Hertel:

The inspection effort involved the coordination of the confirmatory radiological survey activities

performed by our contractor, Oak Ridge Institute for Science and Education, of your research

reactor on October 21-23, 2002. In addition, various aspects of your reactor operations,

decommissioning, and radiation protection programs were inspected, including selective

examinations of procedures and representative records, interviews with personnel, and

observations of the facility.

Based on the results of this inspection, it has been determined that: 1) the decommissioning of

the 5 MWt Research Reactor has been performed in accordance with the approved

Decommissioning Plan; 2) the terminal radiation survey and associated documentation from the

licensee demonstrated that residual radioactive material at the facility and site is less than the

NRC-approved guideline limits; and 3) since the licensee has met their NRC-approved guideline

limits, the facility and site meet the criteria for license termination set forth in 10 CFR Part 20.1401(b)(2).

No safety concern or noncompliance with Nuclear Regulatory Commission (NRC) requirements

was identified. No response to this letter is required.

Dr. N. Hertel

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In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter and its

enclosure will be available electronically for public inspection in the NRC Public Document

Room or from the Publicly Available Records (PARS) component of NRCs document system

(ADAMS). ADAMS is accessible from the NRC Web site at (the Public Electronic Reading

Room) http://www.nrc.gov/NRC/ADAMS/index.html. Should you have any questions

concerning this inspection, please contact Mr. Stephen Holmes at 301-415-8583.

Sincerely,

/RA by Daniel E. Hughes, Acting for/

Patrick M. Madden, Section Chief

Research and Test Reactors Section

Operating Reactor Improvements Program

Division of Regulatory Improvement Programs

Office of Nuclear Reactor Regulation

Docket No. 50-160

License No. R-97

Enclosures: 1. NRC Inspection Report No. 50-160/2002-201

2. Confirmatory Survey Plan for the Georgia Tech Research Reactor dated

October 9, 2002

3. Confirmatory Survey of the Georgia Tech Research Reactor, dated

February 2003

cc w/enclosures: Please see next page

Georgia Institute of Technology

Docket No. 50-160

cc:

Mr. Charles H. Badger

Office of Planning and Budget

Room 608

270 Washington Street, S.W.

Atlanta, GA 30334

Mayor of City of Atlanta

55 Trinity Avenue, S.W.

Suite 2400

Atlanta, GA 30335

Dr. William Vernetson

Director of Nuclear Facilities

Department of Nuclear Engineering

Sciences

University of Florida

202 Nuclear Sciences Center

Gainesville, FL 32611

Joe D. Tanner, Commissioner

Department of Natural Resources

47 Trinity Avenue, S.W.

Atlanta, GA 30334

Dr. Rodney Ice, MORS

Neely Nuclear Research Center

Georgia Institute of Technology

900 Atlantic Drive

Atlanta, GA 30332-0425

Ms. Pamela Blockey-OBrien

D23 Golden Valley

Douglasville, GA 30134

Mr. E.F. Cobb

Southern Nuclear Company

42 Iverness Center

Birmingham, AL 35242

Dr. G. Wayne Clough, President

Georgia Institute of Technology

Carnegie Building

Atlanta, GA 30332-0325

Ms. Glen Carrol

139 Kings Highway

Decatur, GA 30030

Charles Bechhoefer, Chairman

Atomic Safety and

Licensing Board Panel

U.S. NRC, MS: T3-F23

Washington, DC 20555-0001

Mr. James C. Hardeman, Jr.

Manager, Environmental

Radiation Program

Environmental Protection Division

Dept. of Natural Resources

State of Georgia

4244 International Parkway

Suite 114

Atlanta, GA 30354

Dr. Jean-Lou Chameau, Dean

College of Engineering

Georgia Institute of Technology

225 North Avenue

Atlanta, GA 30332-0425

Dr. Peter S. Lam

Atomic Safety and Licensing Board Panel

U.S. NRC, MS: T3-F23

Washington, DC 20555-0001

Dr. J. Narl Davidson, Interim Dean

Chair, Technical and Safety Review

Committee

Georgia Institute of Technology

225 North Avenue

Atlanta, GA 3033-0360

Dr. Charles Liotta, Vice Provost

of Research and Dean of

Graduate Studies

Georgia Institute of Technology

225 North Avenue

Atlanta, GA 30332

Dr. N. Hertel

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June 24, 2003

In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter and its

enclosure will be available electronically for public inspection in the NRC Public Document

Room or from the Publicly Available Records (PARS) component of NRCs document system

(ADAMS). ADAMS is accessible from the NRC Web site at (the Public Electronic Reading

Room) http://www.nrc.gov/NRC/ADAMS/index.html. Should you have any questions

concerning this inspection, please contact Mr. Stephen Holmes at 301-415-8583.

Sincerely,

/RA by Daniel E. Hughes, Acting for/

Patrick M. Madden, Section Chief

Research and Test Reactors Section

Operating Reactor Improvements Program

Division of Regulatory Improvement Programs

Office of Nuclear Reactor Regulation

Docket No. 50-160

License No. R-97

Enclosures: 1. NRC Inspection Report No. 50-160/2002-201

2. Confirmatory Survey Plan for the Georgia Tech Research Reactor dated

October 9, 2002

3. Confirmatory Survey of the Georgia Tech Research Reactor, dated

February 2003

cc w/enclosures: Please see next page

DISTRIBUTION:

PUBLIC

RORP/R&TR r/f

TDragoun

PDoyle

WEresian

PIsaac

SHolmes

CBassett

MMendonca

FGillespie

WBeckner

EHylton

AAdams

BDavis (Ltr.only O5-A4)

ACCESSION NO.: ML031490485

TEMPLATE #: NRR-106

OFFICE

RORP:LA

RORP:RI

RORP:SC

NAME

EHylton:rdr

SHolmes

PMadden

DATE

06/ 04 /2003

06/ 04 /2003

06/ 24 /2003

C = COVER

E = COVER & ENCLOSURE

N = NO COPY

OFFICIAL RECORD COPY

U. S. NUCLEAR REGULATORY COMMISSION

Docket No:

50-160

License No:

R-97

Report No:

50-160/2002-201

Licensee:

Georgia Institute of Technology

Facility:

Georgia Institute of Technology Research Reactor (GTRR)

Location:

900 Atlantic Drive

Atlanta, GA 30332

Dates:

October 21-23, 2002

Inspector:

Stephen W. Holmes

Approved by:

Patrick M. Madden, Section Chief

Research and Test Reactors Section

Operating Reactor Improvements Program

Division of Regulatory Improvement Programs

Office of Nuclear Reactor Regulation

EXECUTIVE SUMMARY

Georgia Institute of Technology Research Reactor

Report No: 50-160/2002-201

This routine, announced inspection involved the confirmatory radiological survey and the on-site

review of selected activities being performed at the Georgia Institute of Technology Research

Reactor. In addition, the activities audited during this inspection included: organization and

staffing; review and audit functions; procedures; removal of materials; decommissioning

activities; release criteria; confirmatory final survey; maintenance and surveillance; and

radiation protection program. The inspector was assisted by the NRCs contractor, Oak Ridge

Institute for Science and Education Environmental Survey and Site Assessment Program.

Organization and Staffing

The organizational structure and their corresponding functions were consistent with

Technical Specification Section 5.0, Amendment No. 14, dated July 22, 1999, and the

Decommissioning Plan for the Georgia Institute of Technology Research Reactor facility

dated June 1998.

Review and Audit Functions

The audits conducted by the Technical Safety Review Committee and Georgia Institute

of Technology Research Reactor staff were in accordance with the requirements

specified in Technical Specification Section 5.2. and Decommissioning Plan Section 2.4.

Procedures

The procedural control and implementation program was acceptably maintained and

met Technical Specifications and Decommissioning Plan requirements.

Removal of Materials

Fuel and radioactive and non-radioactive waste was removed from the site in

accordance with the Georgia Institute of Technology Research Reactor

Decommissioning Plan requirements, and Department of Transportation and Nuclear

Regulatory Commission regulations.

Decommissioning Activities

Decommissioning activities were performed as required by Decommissioning Plan

Section 2.3 and licensee procedures.

Release Criteria

Duratec used the appropriate guideline and screening values, as required by the

NRC-approved Decommissioning Plan, in performing the final survey.

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Confirmatory Final Survey

The elevated surface activity and exposure readings in the basement compressor room

were due to naturally occurring radioactive material.

Based on the results of the licensees final status survey and Nuclear Regulatory

Commissions confirmatory measurements, Georgia Institute of Technology has

adequately demonstrated that the Georgia Institute of Technology Research Reactor

facility satisfies the criteria for release for unrestricted use.

Maintenance and Surveillance

The maintenance program was implemented as required by Georgia Institute of

Technology procedures.

The licensee's program for surveillance and limiting conditions for operation

confirmations satisfied Technical Specification and Decommissioning Plan

requirements.

The licensee's design change procedures were in place and were implemented as

required by licensee procedures.

Radiation Protection Program

The radiation protection program satisfied the requirements of 10 CFR 19.12 and

10 CFR Part 20.1101.

.

Radiological postings satisfied regulatory requirements.

Surveys were performed and documented as required by 10 CFR 20.1501(a), Technical

Specifications, and licensee procedures.

The personnel dosimetry program was acceptably implemented and doses were in

conformance with licensee and 10 CFR Part 20 limits.

Portable survey meters, radiation monitoring, and counting lab instruments were

maintained according to Technical Specifications, industry/equipment manufacturer

standards, and licensee and contractor procedures.

The evaluation and administration of the respiratory program were adequately

performed according to Decommissioning Plan and Nuclear Regulatory Commission

requirements.

The program for monitoring, storage, and release of effluents was acceptable.

Report Details

Summary of Plant Status

Georgia Institute of Technology (GIT), in Atlanta Georgia, has completed decommissioning its

5 MWt Research Reactor (GTRR) and associated systems. The reactor was located within the

Neely Nuclear Research Center (NNRC) on GITs main campus. The reactor was designed for

several different research applications including experiments using high intensity neutron

beams, gamma ray beams, and an uniform thermal neutron flux through a large sized beam.

Although it was originally designed for 1 MWt output, it was upgraded to produce 5 MWt in

1974. The GTRR was built in the early 1960's as a research and training reactor. Operating

under the Nuclear Regulatory Commission (NRC) License No. R-97, it went critical for the first

time on December 31, 1964.

On November 17, 1995, all operations at the reactor ceased. GIT contracted NES, Inc. to

perform the initial characterization survey and to provide a decommissioning plan for the GTRR.

In October 1997, NES performed a characterization survey of the GTRR, based upon the GIT

Decommissioning Project - Radiological Characterization Plan. Results of the characterization

survey were provided in NES Georgia Institute of Technology Research Reactor

Decommissioning Project Characterization Report issued May 1998. GIT requested the NRC,

by letters dated July 1, 1998, February 8, 1999, and May 28, 1999, to grant them the

authorization to decommission the reactor according to their submitted decommissioning plan.

On July 22, 1999, the NRC issued Amendment No. 14 to the reactor licence that approved

GITs Decommissioning Plan. GIT contracted with IT Corporation (IT) to decommission the

GTRR facility. IT, through its subcontractor GTS Duratec (Duratec), started decommissioning

operations December 1999. Final waste shipment was made August 2001.

The Final Status Survey Report for the GTRR facility was completed and issued June 2002.

According to the report, all contaminated systems and components had been removed from the

site. Potentially contaminated structural surfaces identified during characterization surveys had

been removed and/or remediated such that the residual radioactivity is less than NRC

Regulatory Guide 1.86 limits.

The NRC requested Oak Ridge Institute for Science and Educations (ORISE) Environmental

Survey and Site Assessment Program (ESSAP) to perform a confirmatory survey of the GTRR

facility. On October 21-23, 2002, the ESSAP team, accompanied by an NRC inspector,

conducted this survey.

1.

ORGANIZATIONAL STRUCTURE AND FUNCTIONS

a. Inspection Scope (Inspection Procedures (IP) 69001 and 40755)

The inspector reviewed selected aspects of:

organization and staffing

qualifications

management responsibilities

administrative controls

decommissioning activity records

GTRR Decommissioning Plan (DP) dated June 1998

-2-

Technical Specifications (TS), Amendment No. 14, dated July 22, 1999

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b. Observations and Findings

The general organizational structure and staffing had not changed since the last

inspection. The organizational structure and staffing at the facility were as reported in

the Annual Report and as required by TS Section 5.1 and Figure 5.1. Review of

records verified that management responsibilities were administered as required by

TS Sections 5.2 thru 5.6 and applicable procedures.

The decommissioning of the reactor required GTRR management to assume

additional project management responsibilities. Through record reviews and

interviews with the reactor manager, radiation safety officer (RSO), and Duratec

project manager, the inspector confirmed that both GTRR management and the

decommissioning project organization structures were as required by DP Section 2.4

and Figure 2.2.

c.

Conclusions

The organizational staff and their corresponding functions and responsibilities were

consistent with TS Section 5.0, Amendment No. 14, dated July 22, 1999, and the DP

for the GTRR facility dated June 1998

2.

REVIEW AND AUDIT FUNCTIONS

a. Inspection Scope (IPs 69001 and 40755)

The inspector reviewed selected aspects of:

Technical Safety Review Committee (TSRC) meeting minutes

GTRR staff safety review records

TSRC and GTRR staff audit records

responses to safety reviews and audits

personnel qualifications

GTRR DP dated June 1998

TS, Amendment No. 14, dated July 22, 1999

b. Observations and Findings

DP Section 2.4 states that the TSRC: 1) will review and approve all plans, policies and

procedures to be performed under the GTRR Decommissioning Project, 2) will review

and audit the decontamination and decommissioning project operations and activities,

3) members will be appointed by the President of Georgia Tech, and 4) will keep a

written record of the meetings and will report directly to the President.

During inspections in 2000 and 2002, the inspector reviewed the qualifications of the

TSRC members and confirmed that they met the requirements specified in TS Section 5.2 and DP Section 2.4. The results of the 2000 inspections were

documented in NRC Inspection Report (IR) No. 50-160/2000-201 dated March 15,

2000, NRC IR No. 50-160/2000-202 dated August 31, 2000, and NRC IR

-4-

No. 50-160/2000-203 dated December 1, 2000. The inspector noted that the TSRC

met more often than the required semiannual frequency and that a quorum was

present each time. The inspector reviewed the minutes of the TSRC and determined

that they provided guidance, direction, operations oversight, and 10 CFR 50.59

request reviews as required by the DP and TS.

TSRC meeting minutes and audit records and GTRR staff audit records showed that

safety reviews and audits were conducted as required by TS Section 5.2(d). The

content of the audits and safety reviews were consistent with the TS. These reviews

provided appropriate guidance, direction, and oversight to ensure satisfactory

decommissioning of the reactor.

By examining the TSRCs review of the DP and their audits of the operations and

training programs, the inspector determined that the safety reviews, audits, and

associated findings were satisfactory and that the licensee took the appropriate

corrective actions in response to the findings.

The inspector reviewed selected decommissioning and facility change approvals.

Records and observations showed that changes at the facility were acceptably

reviewed in accordance with 10 CFR 50.59 and applicable licensee administrative

controls. None of the changes constituted an unreviewed safety question or required

a change to the TS. The inspector determined that TSRC 10 CFR 50.59 request

reviews were adequately performed.

c.

Conclusions

The audits conducted by the TSRC and GTRR staffs were in accordance with the

requirements specified in TS Section 5.2 and DP Section 2.4. TSRC 10 CFR 50.59

request reviews were adequately performed.

3.

PROCEDURES

a. Inspection Scope (IPs 69001 and 40755)

The inspector reviewed selected aspects of:

administrative controls

records for changes and temporary changes

DP dated June 1998

TS, Amendment No. 14, dated July 22, 1999

decommissioning procedures

logs and records

b. Observations and Findings

During decommissioning activities, the inspector confirmed that written health physics

(HP) and decommissioning procedures were available for those tasks and items

required by TS Section 5.3 and the DP Sections 2.3.1.1. and 3.1.2.2. The procedures

-5-

were routinely updated and then approved by the TSRC while minor modifications to

the procedures were approved by the facility director.

Decommissioning procedures and operating plans reviewed and approved by the

TSRC included those dealing with:

-

Initial Radiological Survey Plan and Procedures

-

Health and Safety Plan and Procedures

-

Waste Management Plan and Procedures

-

Management Plan

-

Quality Assurance Plan and Procedures

-

Radiation Protection Plan and Procedures

-

Decommissioning Work Plan

-

Final Radiological Survey Plan

Through review of the 2000 training records and interviews with staff, the inspector

determined that the training of staff and contractor personnel concerning procedures

was adequate. During the inspectors tours of the facility, it was observed that

personnel performing radiation surveys, conducting instrument checks, issuing

dosimetry, and performing the decommissioning work were doing so in accordance

with applicable procedures.

c.

Conclusions

Based on the procedures and records reviewed and observations of personnel during

the inspections in 2000, it was determined that the procedural control and

implementation program was acceptably maintained and met TS and DP

requirements.

4.

REMOVAL OF MATERIALS

a. Inspection Scope (IPs 69001, 86740, and 85102)

The inspector reviewed selected aspects of:

transportation records

disposal records

NRC Forms 741 and 742

DP dated June 1998

b. Observations and Findings

From 1964 through 1995, the licensee operated a heavy water moderated and cooled

research reactor at the NNRC. The reactor was shut down on November 17, 1995, in

preparation for the summer Olympic Games in Atlanta, GA, and was never restarted.

As noted in a previous NRC IR No. 50-160/1996-01, the irradiated fuel was shipped to

the Savannah River Site on February 18, 1996. The licensee had previously shipped

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the unirradiated fuel to the Oak Ridge National Laboratory site in Tennessee on

January 31, 1996. The inspector confirmed that, as noted by DP Section 1.5, all fuel

had been removed from NNRC prior to decommissioning.

Fifty-six (56) total radioactive waste shipments were made during the GTRR

decommissioning. The final waste shipment occurred on August 3, 2001. Radioactive

waste was sent to one of four consignees: 1 Duratek Inc.; 2 CNSI Barnwell; 3

Envirocare of Utah; and 4 Westinghouse Savannah River Site. During 2000, the

inspector confirmed through records review, interviews with licensee staff, and actual

observation, that radioactive waste was disposed of as required by DP Section 3.2

and in accordance with Department of Transportation and NRC regulations.

c.

Conclusions

As a result of the records review and on-site observations made during

decommissioning tours, it was confirmed that the fuel and radioactive waste were

removed from the site in accordance with the GTRR DP requirements, and

Department of Transportation and NRC regulations.

5.

DECOMMISSIONING ACTIVITIES

a. Inspection Scope (IPs 69001 and 40755)

The inspector reviewed selected aspects of:

operational logs and records

decommissioning procedures

decommissioning logs and records

DP dated June 1998

the facility during tours

b. Observations and Findings

As noted above, the reactor was permanently shut down on November 17, 1995. All

irradiated reactor fuel was removed from the site on February 18, 1996. On July 22,

1999, following a request by the licensee and a review by the NRC, Amendment No.

14 to Facility License No. R-97 was issued which authorized decommissioning of the

GTRR. The licensees contractor started its decommissioning of the facility in January

2000. (Actual decommissioning of the facility was completed in May 2001, although

the contractors final survey of the facility continued for several months afterwards.)

Decommissioning activities focused on the dismantling and removal of the reactor

proper, its support structures, auxiliary equipment and components, and the biological

shield. The inspector examined the following selected tasks as directly described in

DP Section 2.3, Decommissioning Activities and Tasks:

Reactor Complex

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Vertical Beam Ports - The vertical beam ports will be removed - including the

thimbles, thimble plugs, sample tubes, and liners. The lead will be removed from

the plugs and sent to a mixed waste processor. The other items will be

segmented as necessary, packaged, and disposed of as radioactive waste.

Shim Safety Rods and Drives - The four shim safety rods will be disconnected

from the drives, removed through the top shield, cut in half, and disposed of as

mixed waste. The shim safety rod drives will be disconnected, removed,

segmented, and disposed of as radioactive waste.

Horizontal Beam Gates - The ten horizontal beam gate drive motors will be

disconnected and removed. The gates will be separated from the shafts and cut

open. The lead inside will be removed and disposed of as mixed waste, and the

remainder disposed of as radioactive waste.

Spent Fuel Storage Holes - The spent fuel storage hole plugs will be removed and

disposed of as radioactive waste. The hole liners will be core drilled out and each

liner will be cut in half, packaged, and disposed of as radioactive waste.

Piping and Instrumentation - This task involved the removal of miscellaneous

piping and ventilation in and around the reactor complex. The materials will be

disposed of as radioactive waste.

Lead Cover Plate - The lead cover plate will be removed in two distinct pieces - the

inner plate and outer plate. The 24 lead and steel port plugs will be removed from

the inner plate and cut open with an abrasive saw. The lead will be removed and

disposed of as mixed waste, and the steel will be disposed of as radioactive waste.

Upper Top Shield - The upper top shield will also be removed in two distinct pieces

- the inner shield plug and outer shield plug. The 24 concrete and steel inner port

plugs and eight concrete and steel outer port plugs will be removed and disposed

of as radioactive waste. The inner concrete and steel upper top shield will be

removed and disposed of as radioactive waste. The outer concrete and steel

upper shield plug will be removed and disposed of as radioactive waste.

Lower Shield Plug - The 31 lead, concrete, and steel port plugs will be removed

from the lower top shield plug and cut open with an abrasive saw. The lead will be

removed and disposed of as mixed waste. The remaining concrete and steel will

be disposed of as radioactive waste.

Fuel Spray Manifold - The fuel spray manifold pipe will be cut free within the

reactor, utilizing long-handled tools, and transferred to the contamination control

envelope. The manifold will be further segmented and disposed of as radioactive

waste.

Reactor Vessel - A remote operated robotic arm will be installed in the reactor

vessel to facilitate segmentation. Using an abrasive saw connected to the robotic

arm, the horizontal beam ports and through tubes will be cut free and lifted out.

-8-

The bottom pipes will be core bored and removed. The reactor vessel will be cut

into sections using an abrasive saw mounted on the robotic arm. Lifting holes will

first be drilled into each section with a drill attached to the robotic arm, and each

section rigged. Each section will be lifted out with the overhead crane, transferred

to the packaging area and disposed of as radioactive waste.

Graphite Retaining Sleeve - The graphite retaining sleeve will be removed in a

similar fashion as the vessel. Each section will be disposed of as radioactive

waste.

Graphite Removal - The 4-inch by 4-inch graphite stringers will be removed using

long-handled tools from either the top of the biological shield or through the

thermal column. The graphite will be packaged and disposed of as radioactive

waste.

Horizontal Beam Ports - The beam port and through tube plugs will be removed

and disposed as radioactive waste. Lead will first be removed from the through

tube plugs by cutting the top off the plugs with an abrasive saw. The lead will be

disposed of as mixed waste.

Boral Removal - The 1/4-inch boral sheet staked to the inside of the steel tank will

be removed in a similar fashion as the vessel. Each section will be disposed of as

radioactive waste.

Inner Steel Tank - The inner steel tank will follow a similar removal scenario to that

described for the boral removal. The tank will be cut into sections using an

abrasive saw mounted on the robotic arm. Lifting holes will first be drilled into

each section, and each section will then be rigged. After cutting, the section will be

transferred to the packaging area using the overhead crane. Each section will be

disposed of as radioactive waste.

Lead Thermal Shield - The lead thermal shield was formed by pouring molten lead

into the space between the inner and outer steel tanks. With the inner tank and

cooling coils removed, the lead will be pried free of the outer tank in easily handled

pieces with long-handled tools. The pieces will be lowered into a basket and

transferred to a waste container. The lead will be disposed of as mixed waste.

Outer Steel Tank - The outer steel tank will be removed using the same methods

as the removal of the inner steel tank. The tank may have to be pried free of the

concrete prior to removal. Each section will be disposed of as radioactive waste.

Thermal Column Shutter and Shielding - In order to remove the thermal column

shutter and shields, the two thermal column door plugs will be removed first,

segmented with an abrasive saw and the lead removed. The steel cover plate will

then be removed, segmented and packaged. The exposed lead shield will then be

removed and packaged for processing. The concrete and steel blocks will also be

removed and packaged. Segmenting of these blocks is not required. The

concrete, steel and lead doors will be removed, segmented and packaged. Any

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remaining lead will then be removed and packaged for disposal. The concrete and

steel will be disposed of as radioactive waste and the lead as mixed waste.

Biomedical Irradiation Facility Shutter and Shielding - In order to remove the

biomedical irradiation facility shutter, the aluminum cover plate will be removed first

and segmented. The exposed lead bricks will then be removed and packaged.

The movable shield plugs and doors will also be removed. The outer bismuth

shield, the water tank, and the inner bismuth plug will be removed and packaged.

Due to the package restrictions, segmenting of these items will have to be

performed. The materials will be disposed as radioactive waste.

Fission Chambers - The fission chambers will be removed and packaged for

disposal. The remaining U-235 will be packaged and shipped to an appropriate

site.

Biological Shield

Activated Concrete - Due to the relatively small amount of activated concrete and

the limited access, the concrete will be removed with a bobcat/jackhammer. The

waste will be packaged and disposed of as radioactive waste.

Bottom Shield - As above, due to the relatively small amount of activated concrete

and the limited access the concrete will be removed with a bobcat/jackhammer.

The waste will be packaged and disposed of as radioactive waste.

During the inspections in 2000, the inspector observed various of these activities as

they were being conducted including: piping and instrumentation, upper top shield,

graphite removal, lead thermal shield, fission chambers, and activated concrete. In

order to verify that all the above tasks had been performed in accordance with the DP,

the inspector also reviewed the related licensee and contractor records and surveys,

and toured the facility. The inspector determined that the above tasks had been

completed in accordance with final approved DP.

c.

Conclusions

Based on the observations made during the inspection, decommissioning activities

have been performed as required by DP Section 2.3 and licensee procedures.

6.

RELEASE CRITERIA

a. Inspection Scope (IPs 69001 and 40755)

The inspector reviewed selected aspects of:

DP dated June 1998

Georgia Institute of Technology Research Reactor Decommissioning Project

Characterization Report, issued May 1998

Final Status Survey Report for the GTRR facility issued June 2002

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b. Observations and Findings

The primary contaminants of concern for the GTRR are beta-gamma emittersfission

and activation productsresulting from reactor operation. The NRC-approved

guidelines for release for unrestricted use for building surfaces were based on those

for beta-gamma emitters contained in NRC Regulatory Guide 1.86 (NRC 1974). These

guidelines are:

5,000 -- dpm/100 cm2, averaged over a 1 m2 area

15,000 -- dpm/100 cm2, maximum in a 100 cm2 area

1,000 -- dpm/100 cm2, removable.

However, due to the presence of the hard-to-detect-radionuclides H-3 and Fe-55, the

above guidelines were modified to account for the contributing activity of these

radionuclides. The modified guidelines are (Shaw 2002):

2,400 -- dpm/100 cm2 average activity in a 1 m2 area

7,200 -- dpm/100 cm2 maximum activity in a 100 cm2 area

313 -- dpm/100 cm2 removable activity

GITs final survey plan (GTS 2000) stated that radionuclide concentrations in soil for

the contaminants of concern would meet the NRC published (Federal Register Vol. 64

page 68396, December 7, 1999) screening values for selected radionuclides in

surface soils. The screening values for the GTRR radionuclides of interest are

summarized below.

Radionuclide

Guideline Value (pCi/g)

H-3

110

Fe-55

10,000

Pu-239/240

2.3

U-233/234

13.0

U-238

14.0

Ni-59

5,500

Cs-134

5.7

Cs-137

11.0

Co-60

3.8

Eu-152

8.7

Eu-154

8.0

Mn-54

15.0

Ag-110m

3.9

Zn-65

6.2

Sr-90

1.7

C-14

12.0

Ni-63

2,100

Tc-99

19.0

-11-

The inspector observed and interviewed Duratec, ITs representative.

The inspector determined that Duratec used the appropriate guideline and screening

values as calculated in the Characterization Report and specified in the approved DP.

c. Conclusions

Duratec used the appropriate guideline and screening values as required by the DP, in

performing the final survey.

7.

CONFIRMATORY FINAL SURVEY

a. Inspection Scope (IPs 69001 and 40755)

The inspector reviewed selected aspects of:

DP dated June 1998

Georgia Institute of Technology Research Reactor Decommissioning Project

Characterization Report, issued May 1998

Final Status Survey Report for the GTRR facility issued June 2002

b. Observations and Findings

(1) Overview

DP Section 4.0, Proposed Final Radiation Survey Plan, describes the final

radiation survey to be conducted of the facility prior to license termination. This

survey is required in order to ensure that the area satisfies the unrestricted

release criteria for radioactive material according to NUREG/CR- 5849. (DP

Section 4.1) Additionally, DP Section 4.2.3 specifies, As stated in

NUREG/CR-5849, proper documentation of every aspect of the final survey is

necessary for future reference to the decommissioning survey. An accurate

mapping of the reactor containment building and surrounding areas within this

decommissioning project will be maintained for future review and verification by a

regulatory inspector.

Although the licensee is responsible for performing and documentation the

decommissioning and final status survey (Final Status Survey Report for the

GTRR facility issued June 2002), the NRC verifies the licensees performance

through inspections during decommissioning and a confirmatory final survey at the

end.

As part of this confirmatory process ESSAP reviewed and evaluated GITs final

survey plan and report (GTS 2000 and Shaw 2002). The documents were

reviewed for general thoroughness, accuracy, and consistency. Data were

evaluated to assure that areas exceeding guidelines were identified and had

undergone remediation. Final status survey results were compared with guidelines

to ensure that the data had been interpreted correctly. Comments were provided

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to the NRC, documenting the review of the final survey plan and the final survey

report.

The procedures, methods, and data submitted by GIT were considered to be

appropriate and adequately documented the radiological status of the GTRR.

ESSAP confirmed that the licensee modified the gross activity guidelines to

account for hard-to-detect radionuclides. This data was reviewed by ESSAP to

evaluate its appropriateness of use and determined it to be satisfactory.

ESSAP performed confirmatory surveys of the GTRR during the period October 21

to 23, 2002. The surveys were performed in accordance with the site-specific

survey plan submitted to and approved by the NRC and the ORISE/ESSAP Survey

Procedures and Quality Assurance Manuals (ORISE 2002a, 2000a, and 2002b).

ESSAP surveys, their individual findings, and overall results are described in the

sections following.

(2) Surface Scans

Surface scans for beta and gamma radiation were performed over approximately

100 percent of the floor surfaces in the basement and on the first floor and 50

percent of the floor surfaces on the second floor. Surface scans for beta radiation

were performed over approximately 50 percent of the lower walls in the basement,

excluding the Stairwell General Area, 10 percent on the first floor, and 5 percent

on the second floor. Surface scans for beta radiation were also performed in the

vessel tunnel over approximately 50 percent of the surface.

Particular attention was given to remediated and adjacent surfaces, cracks and

joints in the floors and walls, and other locations where residual radioactive

material may have accumulated. Surface scans were not performed on any upper

wall or ceiling surfaces, in the Helium Rupture Disk Chamber, or in the Reactor

Building Ventilation Hold-Up Duct areas. Scans were performed using gas

proportional and NaI scintillation detectors coupled to ratemeters or ratemeter-

scalers with audible indicators. Locations of elevated direct radiation were noted

for further investigation.

ESSAP identified two areas of elevated beta surface radiation. One area was

found on a scabbled portion of the wall in the Bismuth Leak area. Another area

was found on the floor of the processor equipment room. The concrete block walls

in the air compressor room were also noted as being uniformly elevated. Scans of

the remaining surfaces did not identify any additional locations of elevated beta or

gamma radiation.

Surface scans of outdoor locations including soil areas, paved areas, and gravel

surfaces were performed over approximately 50 to 100 percent of the accessible

areas using a sodium iodide scintillation detector coupled to a ratemeter.

-13-

Gamma surface scans were within the range of ambient background levels except

for an area adjacent to the NNRC that was determined to be caused by radiation

shine from the hot cell facility and storage vault.

(3) Surface Activity Measurements

Construction material-specific backgrounds were determined in areas of similar

construction, but without a history of radioactive material use. Ambient gamma

backgrounds were determined in areas where direct beta measurements were

performed; these background measurements were used to correct gross beta

surface activity measurements.

Direct measurements for total beta activity were performed at 35 locations, chosen

randomly and based on surface scan results. Additional measurements to

determine the average activity level in one area were also performed. Dry smears

were collected at each direct measurement location for determining removable

gross alpha and gross beta activity. Wet smears were collected from areas

adjacent to direct measurement locations to determine the H-3 and C-14 activity.

Direct measurements were performed using gas proportional detectors coupled to

ratemeter-scalers.

ESSAP identified an activity of 9,700 dpm/100 cm2 over approximately 0.5 m2 in

the elevated area identified in the Bismuth Leak area, with an average activity of

1700 dpm/100 cm2 over the contiguous one square meter area. The elevated

area identified in the process equipment room was limited to approximately

100 cm2 with an activity of 4,100 dpm/100 cm2. An activity range of 2,700 to

5,100 dpm/100 cm2 was determined for the concrete block in the air compressor

room, which GIT claimed resulted from naturally occurring radioactive material in

the blocks. Confirmatory scans on the interior and exterior of the room found the

radiation levels to be evenly distributed throughout the blocks, confirming the

activity was from the material used to make them. Removable activity levels

ranged from 0 to 3 dpm/100 cm2 for gross alpha and from -5 to 45 dpm/100 cm2

for gross beta. H-3 removable activity levels ranged from 3 to 466 dpm/100 cm2.

C-14 removable activity levels ranged from -2 to 86 dpm/100 cm2.

(4) Exposure Rate Measurements

ESSAP obtained background exposure rate measurements from various locations

within the NNRC, having similar construction as the GTRR. The NNRC has a site

history of radiological material usage; however, there are no other buildings similar

in construction to the GTRR and NNRC on the GIT campus. Exposure rate

measurements, using a microrem meter at one meter above the floor, were

performed in the center of selected areas or rooms within the GTRR.

Average interior building exposure rates ranged from 9 to 25
R/h. Background

exposure rates performed in the NNRC ranged from 18 to 20
R/h.

-14-

Exterior exposure rate measurements, using a microrem meter at one meter

above the surface, were performed at five random locations from the reactor yard

area surrounding the GTRR.

Average exterior exposure rates ranged from 14 to 18
R/h. Background

exposure rates performed at various intersections on the GIT campus ranged from

12 to 20
R/h.

(5) Sampling

ESSAP collected surface soil (0-15 cm) samples at each exposure rate

measurement location.

Analysis of the soil samples by gamma spectroscopy for gamma-emitting mixed

fission and activation products identified Cs-137 at typical fall out concentrations.

Radionuclide concentrations for Co-60 and Cs-137, which are the predominant

radionuclides of concern at research reactor facilities ranged from -0.02 to 0.03

pCi/g for Co-60 and -0.02 to 0.21 pCi/g for Cs-137. All other radionuclides of

concern were reported as less than the respective minimum detectable

concentration of the procedure, which ranged from 0.03 to 0.11 pCi/g.

(6) ESSAP Results

Compliance for residual surface activity was shown using the GIT calibration

methodology approved by the NRC. Since ESSAPs calibration method differs,

this required adjusting the ESSAP-calculated surface activity by the ratio of the

efficiencies for the GIT and ESSAP methods. The correction factor was

approximately 2.3. All corrected ESSAP confirmatory surface activity

measurements, including the identified elevated areas, met guidelines and did not

require further remediation. Additional investigation by the inspector verified that

the concrete block in the air compressor room was made from material with a high

composition of naturally occurring radioactive material.

Except for the air compressor room in the basement, all exposure rate

measurements were less than 5
R/h above background for each survey unit.

Confirmatory surface soil samples were less than the screening values listed in the

GIT final survey plan (GTS 2000).

c.

Conclusions

Based on the above observations, surveys, evaluations, and analyses, the inspector

concluded that: 1) the elevated surface activity and exposure readings in the

basement compressor room were due to naturally occurring radioactive material; and

2) based on the results of the licensees final status survey and ESSAPs confirmatory

measurements, GIT has adequately demonstrated that the GTRR facility satisfies the

criteria for release for unrestricted use.

-15-

8.

MAINTENANCE AND SURVEILLANCE

a. Inspection Scope (IP 40755)

The inspector reviewed selected aspects of:

maintenance procedures

equipment maintenance records

surveillance and calibration procedures

surveillance, calibration, and test data sheets and records

reactor periodic checks, tests, verification, and decommissioning activities

facility design and DP changes and records

NNRC Procedure 4200, 10 CFR 50.59 Review Program for Changes and Tests

During Decommissioning, Revision 01, dated November 1, 1999

TS, Amendment No. 14, dated July 22, 1999

-16-

b. Observations and Findings

(1) General Maintenance

During decommissioning general maintenance was focused on the support

services and equipment and not on any reactor systems. The inspector reviewed

maintenance records, interviewed staff and observed minor maintenance

performed on the various systems in operation. Based on the inspectors

interviews and observations, general maintenance was acceptable for an

industrial site.

(2) Surveillance

The inspector reviewed records of the TS Section 3 required surveillance

verifications performed during 2000. The results of the surveillances for the

radiation monitoring system and the ventilation system were within prescribed TS

limits and procedure parameters, and in close agreement with the previous

surveillance results.

(3) Change Control

TS or DP related 10 CFR 50.59 changes required review by the TSRC in

accordance with TS Section 5.2.

The inspector reviewed various TSRC approved change packages for changing

the method of accomplishing certain decommissioning activities. The inspector

determined that the changes had been evaluated, reviewed, and approved as

required by NNRC Procedure 4200, 10 CFR 50.59 Review Program for Changes

and Tests During Decommissioning, Revision 01, dated November 1, 1999. The

reviews were technically complete and adequately documented. Additionally, the

inspector concluded that TSRC 10 CFR 50.59 reviews and approvals were

focused on safety, and met licensee program requirements.

c.

Conclusions

The licensee's program for surveillance and limiting conditions for operation

verification satisfied TS and DP requirements. The licensee's maintenance and

design change programs were in place and were being implemented as required by

licensee procedures.

9.

RADIATION PROTECTION

a. Inspection Scope (IPs 69001 and 40755)

The inspector reviewed selected aspects of the radiation protection program (RPP):

Radiation Protection Training

radiological signs and posting

-17-

facility and equipment during tours

routine surveys and monitoring

survey and monitoring procedures

dosimetry records

maintenance and calibration of radiation monitoring equipment

periodic checks, quality control, and test source certification records

NNRC Radiation Protection Program (RPP)

event/incident records

b. Observations and Findings

(1) Radiation Protection Program

Although individual procedures had been revised and some added, the RPP had

not functionally changed since the last inspection. The licensee reviewed the RPP

at least annually in accordance with 10 CFR 20.1101(c). This review and

oversight was provided by the TSRC as required by TS Section 5.2.d(9) and DP

Section 2.4.3.

The inspectors review of procedure change records, revisions, and radiation work

permits (RWP), confirmed that the RSO, individually and as a TSRC member,

reviewed and approved RWPs, and advised the Director and TSRC on matters

regarding radiological safety as required by TS Section 5.1.b, DP Section 2.4.1,

and the RPP.

Through record reviews and interviews with GTRR and Duratec staffs, the

inspector confirmed that the RPP was applied to all activities during the

decommissioning project, as required by DP Section 3.1 and GTRR procedures.

(2) Radiation Protection Postings

The inspector observed that caution signs, postings and controls to radiation and

contaminated areas at the NNRC were acceptable for the hazards involved and

were implemented as required by 10 CFR Part 20, Subpart J. The inspector

observed licensee and contractor personnel and verified that they complied with

the indicated precautions for access to such areas. The inspector confirmed that

current copies of NRC Form-3 and notices to workers were posted in appropriate

areas in the facility as required by 10 CFR Part 19.11.

(3) Radiation Protection Surveys

The inspector audited the GTRR daily, monthly, quarterly, and other periodic

contamination and radiation surveys, including airborne activity sampling,

performed from 2000 to 2003. The surveys were performed and documented as

required by DP Section 3.0, and GTRR survey procedures. HP surveys required

for special decommissioning activities, such as RWPs, were also performed and

documented as required. Results were evaluated and corrective actions taken and

documented when readings/results exceeded set action levels.

-18-

-19-

(4) Dosimetry

The inspector confirmed that dosimetry was issued to staff, contractors, and

visitors as outlined in licensee procedures. The licensees dosimetry issuing

criteria specified that dosimetry should be issued to individuals who might receive

a dose equivalent exceeding 10 percent of the annual limits specified in 10 CFR Part 20.1201(a). This criteria meet the requirements of 10 CFR 20.1502 for

individual monitoring. Training records showed that personnel were acceptably

trained in radiation protection practices. During the inspection the inspector

observed that workers and staff wore their dosimetry as required.

The licensee used a National Voluntary Laboratory Accreditation Program-

accredited vendor to process personnel thermoluminescent dosimetry. Dosimetry

results were reviewed by the RSO and doses above the facilitys ALARA limits

were investigated as required. The inspectors review of the licensees radiological

exposure records from 2000 to 2003 verified that occupational doses were within

10 CFR Part 20 limitations.

(5) Radiation Monitoring Equipment

The calibration and periodic checks of the portable survey meters, radiation

monitoring, air sampling, and counting lab instruments were performed by facility

staff or by certified contractors. The inspector confirmed that the licensees

calibration procedures and annual, quarterly, semiannual and monthly calibration,

test, and check frequencies satisfied TS Section 4.3.3, DP Section 3.1, and

10 CFR 20.1501(b) requirements, and the American National Standards Institute

N323 Radiation Protection Instrumentation Test and Calibration or the

instruments manufacturers' recommendations. The inspector verified that the

calibration and check sources used were traceable to the National Institute of

Standards and Technology and that the sources geometry and energies matched

those used in actual detection/analyses.

The inspector also reviewed Duratec instrument calibrations. Their calibration and

periodic checks of the portable survey meters, radiation monitoring, air sampling,

and counting lab instruments were performed by their staffs or by certified

contractors. The inspector confirmed that calibration procedures and annual,

semiannual quarterly, monthly, and daily calibrations, tests, and check frequencies

satisfied Duratec HPS procedures. Calibrations also met 10 CFR Part 20.1501(b)

requirements, and the American National Standards Institute N323 Radiation

Protection Instrumentation Test and Calibration or the instruments manufacturers'

recommendations. The inspector verified that the calibration and check sources

used were traceable to the National Institute of Standards and Technology and

that the sources geometry and energies matched those used in actual

detection/analyses.

The inspector reviewed the calibration lists and confirmed that calibrations for the

radiation monitoring and counting lab equipment in use had been performed and

that all portable instruments in use were calibrated.

-20-

All instruments checked by the inspector had current calibrations appropriate for

the types and energies of radiation they were used to detect and/or measure.

(6) Respiratory Protection

DP Section 3.1.6 states that the Respiratory Protection Program will be

implemented by the decommissioning contractor in compliance with ANSI Z-88.2,

US NRC Regulatory Guide 8.15, 10 CFR 20.1701 through 20.1704, and OSHA

requirements.

While conducting inspections during decommissioning activities at the facility, the

inspector reviewed the respiratory protection program in use by contractor

personnel. The inspector noted that the licensee and contractor had established a

respiratory protection program as required by DP Section 3.1.6 and were using

tested and certified NIOSH/MSHA equipment as required. Records and

observation showed that air sampling was being conducted, surveys and

bioassays were completed as required, testing of respirators was being done, fit

testing of individuals was performed, and individuals were required to pass a

physical in order to qualify to use a respirator. The respiratory protection program

was in compliance with 10 CFR 20.1703 and the DP.

(4) Effluents

The program for the monitoring and storage of radioactive liquid, gases, and solids

was acceptable. Radioactive effluents were monitored and released when within

established limits as outlined in licensee procedures and the regulations. The

principles of As Low As Reasonably Achievable (ALARA) were acceptably

implemented to minimize radioactive releases. Monitoring equipment was

maintained and calibrated as required. Records were current and acceptably

maintained.

c.

Conclusions

Based on the observations made and records audited, it was determined that,

because: 1) surveys were completed and documented as required by

10 CFR 20.1501(a) and licensee procedures, 2) postings met regulatory requirements,

3) the personnel dosimetry program was acceptably implemented and doses were in

conformance with licensee and 10 CFR Part 20 limits, 4) portable survey meters,

radiation monitoring, and counting lab instruments were maintained and calibrated as

required, 5) the evaluation and administration of the respiratory program were

adequately performed, and 6) the program for monitoring, storage, and release of

effluents was acceptable, the RPP implemented by the licensee satisfied NRC and DP

requirements.

5.

EXIT MEETING SUMMARY

The inspector presented the inspection results to members of licensee management at

the conclusion of the inspection on October 23, 2002. The licensee acknowledged the

findings presented and did not identify as proprietary any of the material provided to or

reviewed by the inspector during the inspection.

PARTIAL LIST OF PERSONS CONTACTED

• T. Bauer

Project Leader, ESSAP

• T. Brown

Field Staff, ESSAP

• R. Eby

Executive Engineer, (Vice President Energy, Environment, and Systems)

CH2M HILL

• N. Hertel

Director, Neely Nuclear Research Center

• R. Ice

Manager, Office of Radiation Safety

P. Jones

Project Manager, GTS Duratek Field Services

G. Kalinauskas

Senior Project Engineer, IT Corporation

R. Morton

Field Staff, ESSAP

• Attended exit meeting.

The inspector also contacted other supervisory, technical and administrative staff personnel

as well.

INSPECTION PROCEDURE (IP) USED

IP 69001

Class II Non-Power Reactors

IP 40755

Class III Non-power Reactors

IP 85102

Material Control and Accounting - Reactors

IP 86740

Inspection of Transportation Activities

ITEMS OPENED AND CLOSED

Open

None

Closed

None

PARTIAL LIST OF ACRONYMS USED

Duratec

GTS Duratec

DP

Georgia Institute of Technology Research Reactor Decommissioning Plan dated

June 1998

ESSAP

Environmental Survey and Site Assessment Program

GIT

Georgia Institute of Technology

GTRR

Georgia Institute of Technology Research Reactor

HP

Health Physics

IT

IT Corporation

NNRC

Neely Nuclear Research Center

NRC

Nuclear Regulatory Commission

ORISE

Oak Ridge Institute for Science and Education

RWP

Radiation Work Permits

RPP

Radiation Protection Program

RSO

Radiation Safety Officer

TS

Technical Specifications

-22-

TSRC

Technical Safety Review Committee