ML090960651

Decommissioning Plan for the Leslie C. Wilbur Nuclear Reactor Facility at Worcester Polytechnic Institute

ML090960651
Person / Time
Site:	05000134
Issue date:	03/31/2009
From:	Curley M Worcester Polytechnic Institute
To:	Document Control Desk, NRC/FSME
References
W19-1579-003, Rev 0
Download: ML090960651 (116)

Text

S 100 Institute Road Worcester, MA 01609-2280, USA 508-831-5754, Fax 508-831-5064 www.wpi.edu Financial Services March 31, 2009 Document Control Desk U.S. Nuclear Regulatory Commission 11555 Rockville Pike Rockville, MD 20852 Attn. Mr. Ted Carter

Dear Ted,

Please accept the enclosed Decommissioning Planfor the Leslie C. Wilbur Nuclear Reactor Facility at Worcester Polytechnic Institute. Should you have any questions regarding this submission, please feel free to contact me in my office at (508) 831-6919, or e-mail micurlev@wpi.edu.

Thank you.

Director, Nuclear Reactor Facility

Document W19-1579-003, Rev. 0 DECOMMISSIONING PLAN for the LESLIE C. WILBUR NUCLEAR REACTOR FACILITY at the WORCESTER POLYTECHNIC INSTITUTE Operating License No. R-61 Docket No. 50-134 J

Prepared for:

WORCESTER POLYTECHNIC INSTITUTE Prepared by:

TLG Services, Inc.

Bridgewater, Connecticut March 2009

Worcester Polytechnic Institute Reactor Decommissioning Plan Document W19-1579-003, Rev. 0 Page ii of viii APPROVALS Radiation Safety Officer David Adams Date Date Reactor Facility Director Radiation, Health

& Safety Committee David Messier Date

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Page iii of viii TABLE OF CONTENTS SECTION-PAGE 1.0

SUMMARY

OF PLAN..................................................................................

1-1 1.1 Introdu ction...............................................

............................................ 1-1 1.1.1 O verview.....................................................................................

1-1 1.1.2 Decommissioning Plan Provisions...........................................

1-10 1.2 B ackground.........................................................................................

1-11 1.2.1 Site and Facility H istory..........................................................

1-11 1.2.2 R adiological Status..................................................................

1-12 1.2.3 Reactor Facility Status................................

1-13 1.2.4 Reactor Decommissioning Overview..................

1-14 1.2.5 Estimated Cost..................................

1-15 1.2.6 Availability of Funds.......,...............................

1-15 1.2.7 Program Quality Assurance 1-15 2.0 DECOMMSSIONING ACTIVITIES..........................................................

2-1 2.1 Decom m issioning Alternative..............................................................

2-1 2.2 Facility Radiological Status 2-1 2.2.1 Facility Operating History....................................................

2-1 2.2.2 Current Radiological Status of the WPI Facility......................

2-2 2.3 Decom m issioning Tasks......................................................................

2-8 2.3.1 Activities and.Tasks.

I 2-8 2.3.2 Sch edule....................................................................................

2-18 2.4 Decommissioning Organization and Responsibilities.......................

2-19 2.4.1 WPI Reactor Director.............................

2-20 2.4.2 WPI Radiation Safety Officer 2-22 2.4.3 Decommissioning Consultant..........

............ 2-22 2.4.4 Contractor Assistance.............................

2-23 2.5 Training Program......

2-24 2.5.1 General Site Training 2-25 2.5.2 Radiation Worker Training 2-25 2.5.3 Respiratory Protection Training............................................ 2-25 2.6 Decontamination and Decommissioning Documents and Guides.... 2-26 2.7 Facility Release Criteria................................................................

2-26 3.0 OCCUPATIONAL AND PUBLIC HEALTH AND SAFETY............. 3-1 3.1 Radiation Protection....................................

3-1 3.1.1 Ensuring As Low As Reasonably Achievable (ALARA)

Radiation Exposures 3-1 3.1.2 Health Physics Program.... I.......................................................

3-3 3.1.3 Radioactive Materials Controls......................

3-12

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Page iv of viii TABLE OF CONTENTS (Continued)

SECTION-PAGE 3.1.4 D ose E stim ates.........................................................................

3-13 3.2 Radioactive Waste Management........................................................

3-13 3.2.1 Radioactive W aste Processing.................................................

3-13 3.2.2 Radioactive W aste Disposal.....................................................

3-14 3.3 General Industrial Safety Program..................................................

3-16 3.4 Radiological Accident Analyses..........................................................

3-17 4.0 PROPOSED FINAL RADIATION SURVEY PLAN.............................

4-1 4.1 Description of Final Status Survey Plan.............................................

4-1 4.1.1 Review and Approval of Final Status Survey Plan.................. 4-3 4.1.2 Means for Ensuring that all Equipment, Systems, Structures and Site are Included in the Survey Plan.............. 4-5 4.1.3 Means for Ensuring that Sufficient Data is Included to Achieve Statistical Goals...........................................................

4-5 4.2 Background Survey Results.................................................................

4-5 4.3 Final Release Criteria - Residual Radiation and Contamination L ev els.....................................................................................................

4 -5 4.4 Measurements for Demonstrating Compliance with Release Criteria.............................................

4-6 4.4.1 Instrumentation - Type, Specifications and Operating Conditions......................................

4-6 4.4.2 Measurement Methodology for Conduct of Surveys................ 4-6 4.4.3 Scan Surv eys.............................................................................

4-9 4.4.4 Soil Sam pling.............................................................................

4-9 4.4.5 Sam ple A nalysis.......................................................................

4-10 4.4.6 Investigation Levels....................................

.......................... 4-10 4.5 Methods to be Employed for Reviewing, Analyzing, and A uditing D ata......................................................................................

4-10 4.5.1 Laboratory/Radiological Measurements Quality A ssurance................................................................................. 4-10 4.5.2 Supervisory and Management Review of Results.................. 4-11 5.0 TECHNICAL SPECIFICATIONS..............................................................

5-1 6.0 PHYSICAL SECURITY PLAN...................................

6-1 7.0 EMERGENCY PLAN.......................................

7-1 8.0 ENVIRONMENTAL REPORT...................................................

............... 8-1

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Page v of viii TABLE OF CONTENTS (Continued)

SECTION-PAGE 9.0 CHANGES TO THE DECOMMISSIONING PLAN..........................

...... 9-1

10.0 REFERENCES

10-1 TABLES 1.1 Profile of the W PI LCW NRF Reactor.............................................................

1-9 1.2 Radiological Decommissioning Cost Summary...........................................

1-15 2.1 List of Expected and Potential Radionuclides................................................

2-8 2.2 License Termination Screening Values for Building Surface C ontam ination.....................................................................................

2-28 2.3 License Termination Screening Values for Surface Soil..............

2-29 3.1 Typical Health Physics Equipment and Instrumentation.............

3-6 3.2 Estimated Concentrations of Radioactivity in the Regulating Control Blade (as of December 2008)............................................................

3-14 8.1 Summary of Environmental Impact Assessment..

.................. 8-3 FIGURES 1.1 State and Local Maps of the Area Surrounding WPI....................................

1-2 1.2 WPI Campus Map..........................................

1-3 1.3 Photographic View of the Reactor Facility Building Exterior.......................

1-4 1.4 Photographic Views of the Reactor Facility Lower Level..............................

1-5 1.5 Photographic Views of the Reactor Facility Operating Level........................

1-7 1.6 Rendering of the Reactor Pool, Biological Shield and Experimental Facilities...........................................................

1-8 2.1 Reactor Core Drawing and Photographic View.............................................. 2-5 2.2 Reactor Pool and Biological Shield Cross-Section Drawing..........................

2-7 2.3 LCWNRF Decommissioning Schedule...................

2-19 2.4 WPI-RDP Decommissioning Organization........................

2-21 APPENDICES A

Letter of Financial Commitment.......................................

A-1 B

Revised Technical Specifications for Decommissioning........................... B-1

Worcester Polytechnic Institute Reactor Decommissioning Plan Document W19-1579-003, Rev. 0 Page vi of viii REVISION LOG Rev. No.

CRA No.

Date Item Revised Reason for Revision 0

N/A 03/27/2009 N/A Original Issue

Worcester Polytechnic Institute Reactor Decommissioning Plan Document W19-1579-003, Rev. 0 Page vii of viii ACRONYMS and ABBREVIATIONS ALARA ANSI ASME CDE CFR Cm COMPASS cpm D&D DAC DCGLw DCGLemc DECON Director dpm DQO EDE ENTOMB FSS g

GM GST HEPA HP LCWNRF MARSSIM MA mR mrem MSHA mSV MW NIOSH NQA NRC OSHA pCi POL ppm QA QAPP QC As low As Is Reasonably Achievable American National Standards Institute American Society of Mechanical Engineers Committed Dose Equivalent Code of Federal Regulations centimeter COMPASS Computer Code Version 1.0.0 (development sponsored by the NRC) counts per minute Decontamination and Decommissioning Derived Air Concentration (see 10 CFR 20)

Derived Concentration Guideline Levels Elevated Measurement Comparison DCGL Decontamination Decommissioning Option WPI Reactor Director disintegrations per minute Data Quality Objective Eye Dose Equivalent (see 10 CFR 20)

Entombment Decommissioning Option Final Status Survey gram, a unit of mass Geiger-Mueller General Site Training High Efficiency Particulate Air (Filter)

Health Physics Leslie C. Wilbur Nuclear Reactor Facility Multi-Agency Radiation Survey and Site Investigation Manual, NUREG-1575 Massachusetts milli-Roentgen, 10-3 Roentgen millirem, 10-3 rem U.S. Mine Safety and Health Administration milli-Sievert (unit of dose equivalence, see 10 CFR 20), 10-3 Sievert Megawatt National Institute for Occupational Safety and Health Nuclear Quality Assurance U.S. Nuclear Regulatory Commission Federal Occupational Safety and Health Acts pico-curie, a unit of radioactivity (2.22 disintegrations per min.) 10-12 curie Possession-Only License parts per million Quality Assurance Quality Assurance Project Plan Quality Control

Worcester Polytechnic Institute Reactor Decommissioning Plan Document W19-1579-003, Rev. 0 Page viii of viii ACRONYMS and ABBREVIATIONS (Continued) rem RHSC RSO RWP SDE Sv TEDE WPI WPI-RDP Roentgen equivalent man (rem) is a unit of absorbed radiation dose: 100 rem = 1 Sv Radiation Health and Safety Committee Radiation Safety Officer Radiation Work Permit Shallow Dose Equivalent (see 10 CFR 20)

Sievert (unit of dose equivalence, see 10 CFR 20)

Total Effective Dose Equivalent (see 10 CFR 20)

Worcester Polytechnic Institute WPI LCWNRF Reactor Decommissioning Project

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 1, Page 1 of 21 1.0

SUMMARY

OF PLAN

1.1 INTRODUCTION

1.1.1 Overview This Decommissioning Plan (DP) is for the Leslie C. Wilbur Nuclear Reactor Facility (herein referred to as "LCWNRF", or "Reactor Facility"). The LCWNRF is situated on the Worcester Polytechnic Institute (WPI) campus in Worcester, Massachusetts, approximately 40 miles west of Boston. The LCWNRF is housed in a portion of the Washburn Shops and Stoddard Laboratories Building, located between West and Boynton Streets. The geographical location of WPI campus is shown on Figure 1.1. Figure 1.2 provides an overall map of the WPI Campus. An exterior photograph of the Washburn Shops-Stoddard Laboratories Building, which houses the reactor facility, is presented in Figure 1.3.

Figures 1.4 and 1.5 present photographic views from the two levels of the LCWNRF. Figure 1.6 presents a rendering of the reactor pool, biological shield and experimental facilities.

The LCWNRF provided graduate and undergraduate students with reactor operating experience and experimental practice in the fields of nuclear engineering, metallurgy, chemistry and physics, as well as irradiation services for other teaching, medical and industrial institutions. WPI discontinued routine operation of the reactor on June 30, 2007 (Ref. 1) and submitted an application to the Nuclear Regulatory Commission (NRC) for a possession-only license (Refs. 2 and 3). On August 26, 2008 NRC subsequently granted Amendment No. 11 to Operating License No. R-61, placing the LCWNRF reactor in a Possession-Only License status (Ref. 4).

WPI desires to proceed with decommissioning the LCWNRF and the termination of the associated reactor license.

Accordingly, WPI has filed the appropriate decommissioning amendment requests with NRC along with this DP.

The LCWNRF is contaminated with limited amounts of radioactive material, principally contained in activated metallic core components.

Scoping surveys indicate that practices employed by WPI to minimize the spread of contamination were effective and contamination was not found outside of the reactor pool.

The goal of the proposed D&D activities is termination of WPI's NRC License R-61, Document No. 50-134 and release of the LCWNRF for "unrestricted use." The term "unrestricted use" means that there will be no future restrictions on the use of the Washburn Shops and Stoddard Laboratories building other than those imposed by the City of Worcester zoning ordinances.

WPI holds a separate Massachusetts Agreement State byproduct material license for possession and use of radioactive materials at other campus locations: that license will not be affected by this decommissioning.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 1, Page 2 of 21 This DP has been prepared using the guidance and format recommended in NUREG-1537, Rev. 0, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors (Ref. 5) and incorporates additional guidance from NUREG-1757, NMSS Decommissioning Standard Review Plan (Ref. 6). A summary profile for the LCWNRF reactor is provided in Tables 1.1 and 1.2.

FIGURE 1.1 STATE AND LOCAL MAPS OF THE AREA SURROUNDING WPI (a) State Map

-I-(b) Local Map

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 1, Page 3 of 21 FIGURE 1.2 WPI CAMPUS MAP Ar Ferce and Aerospace Studies 6 Boynton Hall I

IFulle Laborato rIe 2 Aden Memorial 7

CampusCenter 12 WPI tif Sciences and 3 A umnn Gym 8 Campus Police Bioeinineering Center 4 Atwater Kent Laboratories 9 Collegiate Religious Ctr 13 Goddard Hall 5 Bartgett Center 1 0 20 Trowbridge Road 14 Gordon Library Gate ay16 Health Center PHulgis House 18 Hfgins Laboratories 19 Hughes House 20 Internaltonal House

-21 teppren House (One Drury Lane) 22 Kaven Hall h23 OASIS A

24 Glin Htall 25 Plant Services 26 Power House 27 Pronct Center 28 Salisbury Laboratories 29 Skull To b 30 Stratton Hall N ::21 Student Development and Counseling Center 32 Washburn Shops and Stoddard Laboratories are 33 85 Prescott Street STUDENT RESIDENCES A

DanIels Hall B 16 Elbtlrdge gare.., raw C t1SwOrSl Apaltments S

6d le D

ounders Hal 4

E Fuller Apartments PUnity House GInsillute Hall t

Morgan Hall

,j Samilmd *i~ Halt K 22 Schusslir Stoddard Complex 25 Trowbridge N ast Hall "5 0GREEK HOUSES G1 Alpha Chi One 5fl frG2 Alpna Gamma PeGt G3 Alpha Tea Omega 3

-G4 Alpha XI'Delta 2G Lambea C1I Alpha G6 Phi Gamma Delta G7 PhiKappaTheta G8 Phi Sigma Kappa HZ Edi-~p FO G9 Phi Sigma Sigma IN H-l, i1U GIO Sigma Alpha Epsilon G I Sigma Phi Epsilon GI 2 Sigma Pi JHl 5G13 Tau Kappa Epsilon G14 Theta Chi G 15 Zeta Psi Leslie C. Wilbur Nuclear Reactor Facility (Located in the Washburn Shops and Stoddard Laboratories Building)

Worcester Polytechnic Institute Reactor Decommissioning Plan Document W19-1579-003, Rev. 0 Section 1, Page 4 of 21 FIGURE 1.3 PHOTOGRAPHIC VIEW OF THE REACTOR FACILITY BUILDING EXTERIOR

Worcester Polytechnic Institute Reactor Decommissioning Plan Document W19-1579-003, Rev. 0 Section 1, Page 5 of 21 FIGURE 1.4 PHOTOGRAPHIC VIEWS OF THE REACTOR FACILITY LOWER LEVEL (a) Thermal Column Shield Door (b) Pool Water Treatment System & Effluent Hold-up Tank

Worcester Polytechnic Institute Reactor Decommissioning Plan Document W19-1579-003, Rev. 0 Section 1, Page 6 of 21 FIGURE 1.4 (continued)

PHOTOGRAPHIC VIEWS OF THE REACTOR FACILITY LOWER LEVEL (c) Beam Port Access Plug

Worcester Polytechnic Institute Reactor Decommissioning Plan Document W19-1579-003, Rev. 0 Section 1, Page 7 of 21 FIGURE 1.5 PHOTOGRAPHIC VIEWS OF THE REACTOR FACILITY OPERATING LEVEL (a) Operating Floor - East View (b) Operating Floor - West View

Worcester Polytechnic Institute Reactor Decommissioning Plan Document W19-1579-003, Rev. 0 Section 1, Page 8 of 21 FIGURE 1.6 RENDERING OF THE REACTOR POOL, BIOLOGICAL SHIELD AND EXPERIMENTAL FACILITIES

Worcester Polytechnic Institute Reactor Decommissioning Plan Document W19-1579-003, Rev. 0 Section 1, Page 9 of 21 TABLE 1.1 PROFILE OF THE WPI LCWNRF REACTOR

.:REAQCTOR MATER SILS Fuel Uranium aluminum alloy, 19.75% enriched Moderator High purity light water.

Reflector High purity light water and graphite Coolant High purity light water Control Boral and stainless steel Structural material Aluminum Shield Water filled aluminum lined concrete pool STRUCTUTHAL DIMENSIONS Pool 8 feet by 8 feet by.15 feet deep Core (active portion) 15 inches by 15 by 24 inches high Grid box 9x6 array of 3-inch modules Beam port one, 6 inches diameter Thermal column One, 40 inches by 40 inches in cross-section STRATEGIC MATERIALS Burn-up Approximately 1% U-235 THERMAL CHAR* CTERISTICS ý(Calculated)

Operating power 1 kW (maximum 1959-1967) and 10 kW (maximum 1968-2007)

Temperature, water 1300F (maximum)

Maximum power peaking factor 2.17 Maximum hot channel factor 1.51 Maximum heat flux 400 Btu/hr-ft2 Specific power (clean, cold) 2.5 Watt/g U-235 Maximum gamma heat in core 11 Watt/liter UN-,CLEAR CHtACTERISTICS'(Calated),

Average thermal flux 8.3x1010 n/cm 2 sec Average fast flux 11x10 10 n/cm 2 sec Maximum operating excess reactivity 0.5% delta k/keff Critical, mass 4.0 kg U-235 Temperature coefficient 1.63x10-4 delta k/keff per degree C Void coefficient

-2.4x10-3 delta.k/kff per 1% void.

Prompt neutron lifetime 61.2 microseconds Delayed neutron fraction 1.0.0077

~jJ

~

CONROL SAFETYrELE NT Number 3 vertical blades Dimensions 10.5 inches wide by 40.5 inches long by 0.375 inches thick Material Boral Reactivity Control 3.5% delta k/kff each, minimum Total worth; 3 blades 12% delta k/k"r Maximum withdrawal rate 7.5 inches/minute (Note: Table 1.1 is continued on next page)

Worcester Polytechnic Institute Reactor Decommissioning Plan Document W19-1579-003, Rev. 0 Section 1, Page 10 of 21 TABLE 1.1 (continued)

PROFILE OF THE WPI LCWNRF REACTOR CONTROL REGULATINGELEME NTS" Number 1 vertical blade Dimensions 10.65 inches wide by 40.5 inches long by 0.125 inches thick Material Stainless steel Reactivity Control 0.7% delta k/kfir Maximum withdrawal rate 3.8 inches/minute STANDARD FUEL Number of elements 21 + 1/3 for minimum critical loading Fuel alloy Uranium-Aluminum Clad material Aluminum Fuel enrichment U-235 19.75% enriched Number of plates per element 18 Plate thickness 0.060 inches (1.52mm)

' COOLING SYSTEM.

Coolant Pool water Type cooling Natural convection Temperature 130OF (maximum)

Purity required 1 ppm - 0.5x10 6 ohm-cm 1.1.2 Decommissioning Plan Provisions This DP provides the following information associated with the WPI LCWNRF Reactor Decommissioning Project (WPI-RDP):

" A description of the present radiological condition of the LCWNRF and site environs

" A description of the planned approach to be employed

• Descriptions of the methods that will be utilized to ensure protection of the health and safety of the workers and to protect the environment and the public from radiological hazards

" A description of the physical security and material accountability controls that will be in place during the various project phases

" A description of radioactive waste management and disposal

" A cost estimate for decommissioning the LCWNRF and a discussion of the source of funding for these activities

" A decommissioning project schedule

" A description of the applicable quality assurance program

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 1, Page 11 of 21

" A description of the training program to be established for personnel performing work in support of the project

" An Environmental Report concerning the expected impact of performing the decommissioning activities

1.2 BACKGROUND

1.2.1 Site and Facility History The LCWNRF was made possible by a grant of $150,000 from the U.S. Atomic Energy Commission in June 1958. The LCWNRF reactor was constructed by the General Electric Company as a standard 1-kW (thermal) open-pool training reactor, and. first achieved criticality on December 18, 1959.

The reactor license was upgraded to 10-kW (thermal) in 1967, and the LCWNRF reactor first achieved 10-kW operation on January 31, 1968.

The LCWNRF reactor license was renewed in 1982, which extended the operating license for 20 years. In 1989, the reactor was converted from HEU fuel (93% U-235) to LEU fuel (19.75% U-235). No modifications were made to the reactor other than the fuel itself. In February 1992, a new solid-state-based control panel replaced the original vacuum-tube-based control panel.

The control rod drive motors and position sensors were also upgraded at that time.

The primary use of the LCWNRF reactor was in support of WPI Nuclear Engineering-based laboratory activities and student projects. The reactor provided undergraduate and graduate students, under close supervision of qualified personnel, with reactor operating experience and experimental practice in the fields of nuclear engineering, metallurgy, chemistry, and physics.

The LCWNRF is located in the Washburn Shops. and Stoddard Laboratories Building, an academic facility on the WPI campus about one mile from the center of Worcester, Massachusetts.

That* building is located at the top-of a small hill, surrounded by other academic buildings. The nearest dormitories are located more than 500 feet away in various compass directions.

The reactor is a light-water cooled, and moderated, heterogeneous reactor, fueled with approximately 4 kg of 19.75% enriched U-235. The core is located in the center of a pool of de-mineralized water eight feet square by 15 feet deep.

There is a minimum of ten feet of water above the top of the core.

A typical core configuration is based on 21 full elements, plus one-third of an element, in a rectangular array.

The one-third element consists of a removable-plate element with six plates loaded (of the normal 18 plates). A one-curie Pu-Be start-up source occupies one module adjacent to the active core. Three Boral safety blades and a single, manually actuated stainless-steel regulating blade control the

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 1, Page 12 of 21 reactivity. The blades move vertically in two shrouds, extending the length of the core. Figure 2.1 depicts the configuration of the reactor core components.

The core is moderated, reflected, and cooled, by light water that is circulated by natural convection.

The thermal column side of the core is also reflected by graphite.

Core elements are contained in a grid box enclosed on four sides to confine the flow of cooling water to the channels between and surrounding the elements. The grid box and contents, as well as the blade drive mechanisms, are supported by a suspension frame from a reactor bridge. The cold, clean, core with control blades has less than 0.5% excess reactivity. The safety blades, because of their location and large surface area, have a total shutdown reactivity worth of approximately 12% delta kfkeff.

There is an in-core sample irradiation device, which consists of a sample holder on a track arranged so that the contents of the sample holder may be remotely placed adjacent to the core, from the bridge above. An additional area radiation detector is located on the reactor bridge approximately one meter above the pool surface.

There are two experimental facilities, external to the reactor core: the beam port, and thermal column, depicted in Figure 2.2. The beam port is provided as a means of access to fast neutrons.

The beam port is an air-filled, six-inch diameter aluminum tube, which extends from the reactor core face and terminating in a flange at the biological shield face.

In its now permanent inactive state, it is shielded by a shutter inside the pool and a shield plug in the tube at the biological shield. An area radiation detector is located near the exterior of the beam port.

Provisions are made for venting gasses generated in the port to the building exhaust system, and for collecting and draining any seepage that may accumulate between the port tube and surrounding biological shield concrete. Such collecting and draining occurs via tubes embedded in the biological shield.

The thermal column is a graphite-filled, horizontal penetration through the biological shield that provided a means of access to thermal neutrons. The column consists of an aluminum case, 40 inches square in cross section, and about six feet long. Access to the interior of the thermal Column is by a dense, 5.5-foot thick, 2.5-foot square concrete shield door at the biological shield face. The shield door is mounted on wheels that run on rails set in the concrete floor, and can be moved perpendicular to the shield face. Provisions are made for venting gasses generated in the thermal column to the building exhaust system, via tubes embedded in the biological shield. An area radiation detector is located on the external pool wall adjacent to the thermal column door.

1.2.2 Radiological Status An historical site assessment and radiological surveys were conducted at LCWNRF during the summer of 2008. Results of those efforts indicated that radioactivity is

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 1, Page 13 of 21 essentially limited to the interior of the reactor pool and the associated experimental facilities. The maximum exposure rate encountered was at the geometric center of the reactor core, predominantly from the activated stainless-steel regulating blade, with an exposure rate of 125 milliRem per hour at 13 cm through water (with a calculated in-air exposure rate of about 17 milliRem per hour at one meter).

Aside from the regulating blade, the radiation exposure rates encountered were low, and originated from other activated core components, a portion of the beam port liner near the reactor core and small metallic hardware items within the thermal column, limited to no more than a few milliRem per hour. Actual exposure rates from these other items may be lower, as radiation from the regulating blade and the spent fuel still stored in the pool may have interfered with the radiation measurements.

Based upon the direct radiation exposure rate measurements, a radiological inventory was calculated for the activated materials. The majority of the residual radioactivity at the LCWNRF is contained in the activated regulating blade, with an estimated 56 mCi of radioactivity; a mixture of Mn-54 (<1%), Fe-55 (68%), Ni-59

(<1%), Co-60 (27%) and Ni-63 (4.2%). Based on direct radiation measurements, the radioactivity in the remaining structures, systems and components is not expected to exceed a small fraction of that total.

Further radiological characterization of these radioactive items cannot be completed until the start of decommissioning, when destructive sampling techniques can be employed.

Radioactive surface contamination was not found on building or structural surfaces exterior to the reactor pool, the reactor pool's gutters and drains, floor drains, nor was it detected on a limited number of accessible interior surfaces of the ventilation and reactor water cleanup systems. Additionally, no elevated radioactivity levels were observed from NaI detector gamma scans made on these items or areas.

1.2.3 Reactor Facility Status The Reactor is currently permanently shut down, with the fuel removed from the core and stored in racks in the reactor pool. By letter dated June 27, 2007 (Ref. 1),

WPI certified to NRC that it was progressing toward decommissioning and that it would defuel and not operate the reactor after June 30, 2007.

By letter dated August 13, 2007 (Ref. 2), and subsequent correspondence (Ref. 3), WPI submitted a request to NRC for an amendment to the Facility Operating License No. R-61 (including Appendix A, thereto, Technical Specifications for the Worcester Polytechnic Institute Reactor), to place the LCWNRF reactor in Possession-Only-License status. On August 26, 2008 (Ref. 4), NRC granted WPI's requested License Amendment No. 11, for Possession-Only License status.

WPI has notified the

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 1, Page 14 of 21 Department of Energy (DOE) of the Institute's desire to return the used reactor fuel to DOE (Ref. 7).

1.2.4 Reactor Decommissioning Overview Prior to implementing the decommissioning actions described herein, WPI will have cleared LCWNRF of extraneous furnishings and materials not directly associated with the reactor. In addition, the used nuclear fuel will have been returned to the DOE.- The majority of the remediation will focus on components within the reactor pool, the biological shield surrounding the reactor, and connected support systems (e.g., exhaust ventilation system, pool water treatment system and floor drains adjacent to the reactor pool). No structural contamination in other areas of the facility is known or suspected, such that only minor remediation efforts, if any, are anticipated for the building structures. Additionally, no contamination of soil or ground water is known or suspected; therefore, decommissioning of the LCWNRF can be accomplished without dismantlement or breaching of the building. After termination of the NRC license (thus allowing unrestricted use), the affected portions of the building will be restored for reuse as an educational facility. The general activities planned to complete the DP objectives are:

" Remove the activated regulating blade from the reactor core

" Drain the reactor pool

" Remove the remaining reactor core components from within the reactor pool

" Remove the beam port extension, liner and shutter

" Remove the graphite blocks and activated portions of the aluminum liner from the thermal column

" Remove the activated / contaminated portion of the aluminum pool liner, and survey and remove any affected underlying biological shield concrete

• Remove the reactor pool water treatment system

• Remove the exhaust ventilation ducts associated with the beam port and thermal column

• Survey and remediate the floor drains, if required

• Perform additional decontamination and dismantlement of the structure and equipment in accordance with this DP

• Prepare the decommissioning-generated material for free-release or disposal, as appropriate (i.e., either decontaminate and release the material as non-radioactive waste, or package the material for transport as radioactive waste)

• Ship all radioactive waste off-site to a licensed waste processor or disposal facility Perform and document a Final Status Survey and submit a request to NRC for termination of the reactor license

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 1, Page 15 of 21 1.2.5 Estimated Cost This estimated to cost for the radiological decommissioning of the WPI LCWNRF, including a 25%. contingency, is $2,594,000. A cost breakdown is given in Table 1.2.

TABLE 1.2 RADIOLOGICAL DECOMMISSIONING COST

SUMMARY

(2009 dollars*)

LABOR EQUIPMENT, RADWASTE.

OPERATION PERSON-PLUS.

C

-HIPPINGL HOURS

.TRAVEL.&.-

.N.R.CTS SHIPPING &

TCOST

!PE"" DIEM "

& SUPPLIES" DISPOSAL**

COS

."*, PER.D E 7

1 1.7 Mobilization 400

$64,000

$64,000 Decontamination and 9,060

$1,317,000

$316,000

$131,000

$1,764,000 Dismantling Demobilization License Termination 1,440

$247,000

$247,000 Activities Totals 10,900

$1,628,000

$316,000

$131,000

$2,075,000 25% Contingency

$519,000 GRAND TOTAL

$2,594,000 Costs have been rounded to the nearest thousand; person-hours have been rounded to the nearest ten.

5* The estimate for low-level radioactive waste disposal is based upon the assumption that the radioactive waste will be buried at the EnergySolutions, LLC disposal site (formerly the Envirocare of Utah site).

• • • Included in Decontamination and Dismantling operations.

1.2.6 Availability of Funds WPI is committed to providing the funding for decommissioning of the WPI LCWNRF as evidenced by the letter of financial commitment shown in Appendix A.

1.2.7 Program Quality Assurance A quality assurance program will be implemented throughout the decommissioning effort to assure that work is performed as intended in a controlled manner and does not endanger public safety, and to assure the safety of the decommissioning staff.

A Quality Assurance Project Plan (QAPP) will be developed to incorporate the applicable portions of Code of Federal regulations (CFR) 10 CFR 50, Appendix B (Ref. 8) and 10 CFR 71, Subpart H (Ref. 9). In addition, the QAPP will utilize a graded approach that bases the level of controls on the intended use of the results and the degree of confidence needed in their quality. ANSI/ASQC E4-1994 (ASQC 1995, Ref. 10) and NUREG-1575, Appendix K (MARSSIM, Ref. 11) will be used to provide guidance in quality-related activities, including the development of a QAPP and the collection and evaluation of environmental data.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 1, Page 16 of 21 Quality Assurance (QA) efforts during the decommissioning period will include the following:

" Performing QA functions for procurement

" Qualifying suppliers

" Auditing project activities

" Monitoring worker performance for compliance with work procedures

" Verifying compliance of radioactive shipments with appropriate procedures and regulations

" Performing dimensional, visual, nondestructive examinations or other required inspection services to assure compliance with work plans

" Maintaining auditable files The QAPP will be prepared and implemented by a Decommissioning Consultant and/or contractors, and will be approved and overseen by WPI's management. The program will be implemented by written procedures throughout the decommissioning project.

The management staff of those organizations participating in the QAPP shall regularly review the status and adequacy of that part of the QAPP that they are implementing. All changes to the QAPP shall be governed by measures commensurate with those applied to the original issue. The QAPP will incorporate the items discussed in the following subsections.

1.2.7.1 Quality Assurance Responsibilities WPI, its Decommissioning Consultant and contractors shall have the responsibility, authority and organizational freedom to:

" Identify quality problems

" Take action to stop unsatisfactory or unsafe work and control further processing, delivery, installation or use of nonconforming items

" Initiate, recommend or provide solutions to correct quality issues, unsafe practices or non-conformances

" Verify implementation of solutions 1.2.7.2 Quality Requirements for Instrumentation Calibration Field instruments and associated detectors shall be calibrated on a semi-annual basis using National Institute of Standards and Technology (NIST) traceable sources. Calibration equipment and laboratory instruments shall be calibrated on an annual basis. Equipment calibrations will be performed by a qualified vendor.

Calibration labels showing instrument identification number, calibration date and calibration due date shall be attached to all field and laboratory instrumentation.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 1, Page 17 of 21 Response Testing All instrumentation will be inspected and source-checked daily prior to use to verify calibration status and proper operation. Control checks and/or source-check criteria will be established prior to the initial use of the instruments.

Maintenance Limited maintenance, such as changing Mylar windows, high-voltage cables, etc.,

may be performed on-site by qualified personnel.

Following any limited maintenance, calibration checks may be performed on site by qualified personnel.

Major repairs and ensuing re-calibrations will be performed by a qualified vendor.

Record Keeping Calibration and maintenance records, or copies of these records, shall be maintained on site where they will be available for review. The results of the daily instrument functional checks will be recorded on separate log sheets for each instrument and maintained on site.

1.2.7.3 Sampling and Analysis Quality Control Sample Collection Direct surface beta measurements, removable contamination measurements, gamma exposure rates, soil sampling and any specialized measurements will be performed to provide data required to meet the regulatory guidance provided in:

" Title 10, Code of Federal Regulations, Part 20, Section 1402 (10 CFR 20.1402), Radiological Criteria for Unrestricted Use (Ref. 12)

" NUREG-1575, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM) (Ref. 11)

" NUREG-1727, NMSS Decommissioning Standard Review Plan (Ref. 6)

Sample QC Quality Control (CQ) samples will be obtained for minimum of 10% of all samples collected for radionuclide-specific analysis. The QC samples will be a combination of split, duplicate, blank, and/or spiked samples.

Independent replicate QC measurements for direct measurements will be provided for a minimum of 5% of the measurements associated with the FSS.

Sample Identification Direct surface beta measurements, removable contamination samples, exposure rates, and any specialized measurements will be identified as to location, type of measurement, specific instrument and probe used, sample time and date (as appropriate) and name of the person collecting the data.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 1, Page 18 of 21 Volumetric samples (e.g., concrete, soil, etc.) will be identified with a unique sample number, sample location, depth of sample, sample time and date as appropriate, and the name of the person collecting the sample.

Sample Chain-of-Custody Sample chain-of-custody shall be initiated for those samples being sent off site for analysis or transferred to another organization for analysis. A sample Chain-of-Custody Record will be generated which will document the sample identification and sample transfer and will accompany the sample during shipping to the new custodian of the sample.

Sample Analysis Vendor laboratories shall be on a QA Approved Suppliers List for WPI or its contractors for the type of analytical services being provided. WPI has the ultimate responsibility for ensuring that decommissioning sample analysis specifications and laboratory capabilities meet data quality requirements.

Sample Documentation Sample identification information, sample Chain-of-Custody Records, sample analysis results, vendor laboratory qualification records, or copies of these records, shall be maintained on site where they will be available for review.

1.2.7.4 Record Keeping Measures shall be established to control the issuance of documents and changes to documents that prescribe activities affecting quality, such as procedures, drawings and specifications. These measures shall ensure that documents and changes to documents are reviewed for adequacy, approved for release by authorized personnel and distributed to and implemented at the location where the prescribed activity is performed.

Procedure Control Procedures shall be controlled to ensure that current copies are provided to personnel performing the prescribed activities. Procedures shall be independently reviewed by a qualified person and shall be approved by a management member of the organization responsible for the prescribed activity.

Significant changes to procedures shall be reviewed and approved in the same manner as the original.

Radioactive Shipment Package Documents All documents related to a specific shipping package for radioactive material shall be controlled by appropriate procedures. All significant changes to such documents shall be similarly controlled.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 1, Page 19 of 21 Final Survey Documents All documents related to the FSS shall be controlled by appropriate procedures. All significant changes to such documents shall be similarly controlled.

This documentation would normally include a Survey Plan, Survey Packages, Survey Results and Survey Report.

1.2.7.5 Handling, Storage and Shipping Approved procedures shall be utilized to control the handling, storage and shipping of radioactive materials.

Radioactive Material Storage Areas shall be provided in the Reactor facility for storage of radioactive material to ensure physical protection and control of the stored material. The handling, storage and shipment of radioactive material shall be controlled through the following requirements:

" Procedures shall be provided for handling, storage and shipping operations.

" Established safety requirements concerning the handling, storage and shipping of packages for radioactive material shall be followed.

" Shipments shall not be made unless all tests, certifications, acceptances and final inspections have been completed.

Shipping and Packagini Shipping and packaging documents for radioactive material shall be consistent with pertinent regulatory requirements.

1.2.7.6 Quality Assurance Records Sufficient records shall be maintained to furnish evidence of activities important to safe decommissioning as required by code, standard, and specification or project procedures. Records shall be identifiable, available and retrievable. The records shall be reviewed to ensure their completeness and ability to serve their intended function.

Requirements shall be established concerning record collection, safekeeping, retention, maintenance, updating, location, storage, preservation, administration and assigned responsibility. Requirements shall be consistent with applicable regulations and the potential for impact on quality and radiation exposure to the workers and the public. Typical records would include:

" Decommissioning Plan

" Procedures M Reports

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 1, Page 20 of 21

" Personnel qualification records

" Radiological and environmental site characterization records, material free-release records and FSS records

" Dismantlement records

" Inspection, surveillance, audit and assessment records Health and Safety Related Activities Records that have a potential for impact on quality and radiation exposure to the workers and the public include the following:

• Work Permits

• Work Procedures

• Contamination Survey Reports

• Radiation Survey Reports

• Airborne Survey Reports

• Counting data or air samples and gamma spectrum analysis

• Instrument calibrations

• Source inventory and storage

• Radioactive material inventory and storage

• Shipment records

• Incidents and accidents

" Confined space entry permits

" Monitoring records for oxygen deficient and explosive atmospheres Personal Records Typical. records containing personal information that may impact quality and radiation exposure to the workers and the-public are as follows:

" Dosimetry Records

" Bioassay analysis

" Respiratory protection qualifications (medical/clearance and fit tests, if required)

" Training records

" Visitor logs and exposure information

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 1, Page 21 of 21 1.2.7.7 Audits Audits shall be implemented to verify compliance with appropriate requirements of the Quality Assurance Project Plan and to determine the effectiveness of the plan.

The audits shall be performed in accordance with written procedures or checklists by trained and qualified personnel not having direct responsibility in the areas being audited.

Audit Reports Reports of the results of each audit shall be prepared. These reports shall include a description of the area audited, identification of the individual responsible for implementation of the audited provisions and for performance of the audit, and identification of discrepant areas.

The audit report shall be distributed to the appropriate level of management and to those individuals responsible for implementation of audited provisions.

Audit Corrective Action Measures shall be established to ensure that discrepancies identified by audits are resolved. These measures shall include notification of the manager responsible for the discrepancy and verification of satisfactory resolution. Discrepancies shall be resolved by the manager responsible for the discrepancy.

Higher levels of management shall resolve disputed discrepancies. Follow-up action, including re-audit of deficient areas, shall be taken as indicated.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 2, Page 1 of 30 2.0 DECOMMISSIONING ACTIVITIES The decommissioning alternative selected for the WPI Reactor is the removal of the facility from service and reduction of the residual radioactivity to a level that will permit unrestricted termination of the reactor license and beneficial reuse of the affected rooms within the building. The facility will be surveyed and left in place.

2.1 DECOMMISSIONING ALTERNATIVE The objective of the decommissioning is the regulatory release of the Reactor facility located within the Washburn Shops and Stoddard Laboratories building, allowing for their unrestricted use.

On this basis the safe storage (SAFSTOR) and entombment (ENTOMB) decommissioning options were considered inappropriate to WPI's future plans, and DECON is the decommissioning alternative selected by WPI.

SAFSTOR poses essentially the same potential risks and environmental impacts as the proposed project, but for a much greater period of time. This alternative would necessitate continued surveillance and maintenance of the Reactor facility over a substantial time period during which the radiological risks would continue to exist.

ENTOMB would necessitate continued surveillance and maintenance of the Reactor facility over a substantial time period. During this period, the radiological risks would continue to exist.

DECON is the decommissioning option chosen by WPI. To the extent possible, decontamination of facility equipment and structural components will be conducted so as to minimize radioactive waste. Structural portions of the building found to be radiologically contaminated and/or activated shall be dismantled and/or decontaminated, as necessary.

This would be followed by an extensive and comprehensive final radiation and contamination survey demonstrating-that the LCWNRF meets NRC criteria for release to unrestricted use. The results of this final survey will be documented in a report to be submitted to NRC in support of a request that the site be released for unrestricted use and the reactor license terminated.

2.2 FACILITY RADIOLOGICAL STATUS 2.2.1 Facility Operating History

" Initial reactor criticality: December 12, 1959

" Reactor operation (at 1KW): 1959 to 1967

" Conversion from 1 KW to 10KW: 1968 0 Reactor operation (at 10 KW): 1968 to 2007

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 2, Page 2 of 30

" Typical reactor operation: - 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> per year

" Conversion from HEU to LEU fuel: 1989

" Certification of permanent reactor shut-down: 2007

" Possession-only License Amendment: August 26, 2008

" Request return of reactor fuel to DOE: September 2008

" Return of reactor fuel to DOE: To Be Determined The LCWNRF was constructed at WPI to provide graduate and undergraduate students with reactor operating experience and experimental practice in the fields of nuclear engineering, metallurgy, chemistry and physics, as well as irradiation services for other teaching, medical and industrial institutions. The integrated power generated during operation of the WPI Reactor is estimated at approximately 41 MW-hours.

2.2.2 Current Radiological Status of the WPI Facility 2.2.2.1 General Routine radiological surveys show that the radiation levels and contamination levels measured at the LCWNRF have consistently been low and limited in extent.

Radiological measurements performed in the summer of 2008, confirmed that only minor quantities of residual radioactivity or radioactive contamination are present.

The information indicates that the radioactive portions of the facility are primarily confined to the reactor internals and reactor pool. The majority of the residual radioactivity at the facility contained in structures, systems and components, is found in a single item, an activated stainless-steel regulating blade (as discussed in section 1.2.2. Other than the exposure rate emanating from the regulating blade (125 mR/hr at 13 cm, through water) radiation exposure rates at the reactor facility were found to be very low and limited to portions of the reactor core box and grid, the reactor core end of the beam port and the reactor side of the graphite thermal column interior, that were slightly neutron activated. Maximum radiation levels on these non-regulating blade components were found to be in the range of one to three mrem per hour on contact.

Radiation and radioactive contamination was not detected outside of the reactor pool or experimental facilities in the biological shield.

Estimates of the radioactivity inventory can be determined by considering the constituent elements of the material in question and calculating the duration of exposure to the neutron flux and the energies of the incident neutrons. However, in the case of the WPI Reactor, with its low and intermittent usage, direct measurements are generally more reliable and will be used to guide the actual removal and/or dismantlement of components.

Characterization of radionuclide mixtures and concentrations will be performed at the time of actual removal and dismantlement, as this requires destructive sampling of structures, systems and components that are currently needed to support safe storage of the used fuel, and

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 2, Page 3 of 30 would not be permitted by the possession only license. This information will be obtained during the process of decommissioning and will further define the basis for specifying the necessary safety measures and procedures for the various dismantling, removal, decontamination, and waste packaging and storage operations so that exposure to personnel is maintained ALARA.

2.2.2.2 Principal Radioactive Components This listing is based upon process knowledge and direct radiation measurements.

The most radioactive components to be handled and processed during the WPI LCWNRF Reactor Decommissioning Project (WPI-RDP) are those that have become neutron activated.

The reactor core structure was principally constructed of Aluminum, with a small amount of stainless-steel (i.e., the single regulating blade and small fasteners, such as bolts and pins). Based upon the elemental constituents of these materials, only the stainless steel items should show a significant amount of neutron activation. In the case of the biological shield and experimental facilities, there are elemental constituents within the concrete, steel rebar and graphite that can become activated; however due to their distance from the reactor core, and the low power level and low usage of the reactor, there should not be a significant amount of activation. This pattern of neutron activation was confirmed by the gamma exposure rates measured in and around the reactor core and pool structures.

A summary of the most significant radiological characteristics found at the LCWNRF are as follows (see Figures 2.1 and 2.2 for location and configuration of these items):

The gamma exposure rate at the geometric center of the reactor core, between the three Boral Control Blades and one stainless-steel Regulating Blade was measured at 125 mR/hr (at 13 cm), through water). The Regulating Blade is believed to be the principal the source, and based on its composition and geometry, an in air exposure rate of 17 mR/hr at 1 meter is estimated. That 6.9 Kg (884 cubic centimeter) blade is estimated to contain 56 mCi of radioactivity; Mn-54 (<1%), Fe-55 (68%), Ni-59 (<1%), Co-60 (27%)

and Ni-63 (4.2%).

Ion Chambers and lower Suspension Posts, located above the top corners of the active core region were found to have contact gamma exposure rates of about 2 mR/hr (under water).

Gamma exposure rates from the Reactor Core Box, at mid-core plane elevation, on the exposed West, South and North sides, were found to be 4, 7 and 15 mR/hr, respectively, measured through 30 cm of water. Based on the core configuration and the radiation level pattern these radiation levels are likely influenced by the activated stainless steel regulating control blade and the spent fuel that is currently stored in racks on the North and South pool walls, and not due to significant activation of the Aluminum Core Box.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 2, Page 4 of 30

" Gamma radiation levels were measured, in air, inside the six-inch diameter Beam Tube Extension, with 0.4mR/hr where the tube penetrates the pool wall and 4 mR/hr where the extension tube terminates at the reactor core face.

" Activated stainless steel Heli-Coil hardware embedded in thermal column graphite blocks were found with gamma exposure rates of 200 microR/hr on contact.

Gamma exposure rate measured on contact with select Thermal Column graphite blocks, ranged from background-to-10 microR/hr above background.

As expected, the graphite blocks closest to the reactor core were those found with the detectable radiation levels.

2.2.2.3 Radionuclides The radionuclides expected to be present, or that may possibly be present, in the WPI Reactor at the time of decommissioning, are listed in Table 2.1. This list of radionuclides is principally based on the assumption that reactor operation resulted in neutron activation of reactor core components and other integral hardware or structural members situated adjacent to, or in close proximity to, the reactor core.

Specific materials that are considered to have been exposed to neutron activation include principally aluminum, and small amounts of stainless steel, graphite and concrete. Co-60 and Fe-55 are the principal radionuclides expected to be present in detectable quantities at the WPI LCWNRF.

Table 2.1 also contains other radionuclides that generically can be present at a research reactor, but are not expected in detectable quantities at the LCWNRF, such as fission products from the fuel and Europium radioisotopes from activated graphite and concrete.

The determination of actual radionuclides and concentrations present as residual activity in structures surrounding the reactor, and there content in waste materials will be based upon direct measurements and sampling performed during the decommissioning process.

Worcester Polytechnic Institute Reactor Decommissioning Plan Document W19-1579-003, Rev. 0 Section 2, Page 5 of 30 FIGURE 2.1 REACTOR CORE DRAWING AND PHOTOGRAPHIC VIEW CONTROL ROD

-ION CHAMBER (a) Reactor Core - Artist Rendering

Worcester Polytechnic Institute Reactor Decommissioning Plan Document W19-1579-003, Rev. 0 Section 2, Page 6 of 30 FIGURE 2.1 (continued)

REACTOR CORE DRAWING AND PHOTOGRAPHIC VIEW (a) Reactor Core Photo - Overall View (b) Reactor Core Photo - Close-up View

Worcester Polytechnic Institute Reactor Decommissioning Plan Document W19-1579-003, Rev. 0 Section 2, Page 7 of 30 FIGURE 2.2 REACTOR POOL AND BIOLOGICAL SHIELD CROSS-SECTION DRAWING i~ tz<4y

Worcester Polytechnic Institute Reactor Decommissioning Plan Document W19-1579-003, Rev. 0 Section 2, Page 8 of 30 2.3 DECOMMISSIONING TASKS 2.3.1 Activities and Tasks WPI Reactor facility decommissioning activities will be conducted in four phases, as follows:

u Phase One: Pre-Decommissioning Activities

" Phase Two: Decommissioning Mobilization Activities

" Phase Three: Decontamination and Dismantling Activities 0 Phase Four: Site Restoration Activities TABLE 2.1 LIST OF EXPECTED AND POTENTIAL RADIONUCLIDES NUCLID HALF-LIFE*

MEANS OF

(years)

PRODUCTION 3H*

12.28.

Activation 14C 5,730 Activation 54Mn 0.856 Activation 55Fe 2.73 Activation 60Co 5.27 Activation 59Ni 76,000 Activation 63Ni 100 Activation 63Zn 0.67 Activation 90Sr 29.1 Fission 94Nb 20,000 Activation 99Tc 213,000 Fission 125Sb 2.76 Fission 1291 15,700,000 Fission 134Cs 2.07 Activation 137Cs 30.17 Fission 152EH 13.48 Activation 154Eu 8.8 Activation 155Eu 4.96 Activation

• Radionuclide Half-Life values are from "The Health Physics and Radiological Health Handbook" (Ref 13).

Phase One and Phase Four activities are considered outside of the scope of this DP:

however, they are described briefly below to provide an overall context to this DP.

Phase One Phase One activities are those tasks performed to prepare the facility for future decommissioning activities. This work will be performed under the authority of the existing possession-only license and may be performed before NRC approval to decommission is granted. This work will include general cleanup of the facility,

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 2, Page 9 of 30 such as removal of non-contaminated equipment and materials (non-reactor related) situated throughout the WPI LCWNRF: such materials will be collected,

surveyed, and appropriately dispositioned in accordance with established procedures. This would generally include removal of no-longer needed furniture, laboratory and office supplies. Non-radiologically contaminated ceiling and floor tiles and insulation, which may contain asbestos, may be removed during this phase to facilitate future decommissioning activities. At no time during Phase One will any of the reactor's structures, systems or components be dismantled or decontaminated. It is also expected that the used nuclear fuel would be returned to DOE during this phase.

Phase Four.

Phase Four activities are those tasks performed after the facility's NRC license has been terminated, allowing for unrestricted use. During this phase, the facility will be restored and remodeled for educational use, such as classroom or office space.

While detailed plans have not yet been devised, it is expected that the remaining portions of the biological shield, the decontaminated reactor bridge and control console will be demolished, and disturbed structural surfaces repaired.

Phases Two and Three activities encompass the radiological decommissioning tasks that will be performed in accordance with the provisions of this plan, as described herein.

WPI anticipates that this work will be performed by experienced and qualified contractors, managed by a Decommissioning Consultant, and overseen by the RHSC (see Sections 2.4.3 Decommissioning Consultant and 2.4.4 Contractor Assistance).

2.3.1.1 Phase Two: Decommissioning Mobilization Activities Phase Two will commence after a Decommissioning Order has been granted by NRC and the used nuclear fuel has been removed from the reactor pool. The tasks in this phase will include preparatory work needed to fulfill the administrative and physical requirements of this DP, prior to the conduct of Phase Three decontamination and decommissioning (D&D) activities.

WPI anticipates that Phase Two will be implemented in two sub-phases; off-site and on-site mobilization.

Off-site mobilization activities will include detailed project planning and engineering, preparation of detailed operating procedures, and procurement of specialty equipment, supplies and subcontractors.

On-site mobilization will include training and qualification of workers and physical preparation of the facility for phase three decommissioning activities. Training and qualification of workers will be performed as described in Section 2.5 Training Program.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 2, Page 10 of 30 Systems and components to be dismantled will be de-energized and isolated at this time in order to assure the safety of the decommissioning workers. Temporary electric power, lighting and other utilities will be installed as needed. Material-handling equipment, such as ramps, gantry cranes, etc., will be installed as required to support the decommissioning work.

General radiological controls will be installed to support the decommissioning work.

Contamination control boundaries and protective barriers, personnel contamination frisking stations and step-off pads will be setup to isolate the active work areas from the un-involved portions of the facility. Facilities for workers to change into protective clothing will be set up. Portable air samplers, continuous air monitors, area radiation monitors, containment tents and High-Efficiency Particulate Air (HEPA) filtered ventilation units will be setup as required to support the decommissioning work. A Health Physics (HP) counting room will be set up for sample analysis to support radiation protection, waste management and final site survey (FSS) work. A secure radioactive material storage area will be setup, where waste material can be accumulated, packaged and prepared for shipment to the disposal facility. Additionally, a low-background area will be set up for conducting free-release surveys and secure storage of materials prior to release.

During this phase, characterization of structures, systems and components will be performed on those areas that could not be performed prior to receipt of the decommissioning order (e.g., destructive sampling of activated reactor core and biological shield materials).

2.3.1.2 Phase Three: Decontamination and Dismantling Activities Phase Three activities will be comprised of ten discrete tasks involving the physical dismantling of contaminated and activated equipment and removal or decontamination of structures, plus three additional tasks associated with:

0 The preparation, packaging and transportation of radioactive waste for disposal (see Section 3.2 Waste Management) 0 Demobilization and preparation of the building for final surveys

• Performance of a FSS to verify compliance with facility release criteria (see Sections 2.7 Release Criteria and 4.0 Proposed Final Radiation Survey Plan)

The general philosophy concerning the dismantling and removal of contaminated or activated equipment Will be to use non-destructive disassembly techniques, such as un-bolting, wherever possible. Such techniques eliminate creation of contaminated particles and debris that can create airborne and contamination migrations risks. If un-bolting can not be performed (e.g., if frozen bolts are encountered) only mechanical segmentation techniques will be employed in order to minimize the creation of airborne dusts, mists or fumes: examples of such mechanical segmentation techniques include sawing and shearing.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 2, Page 11 of 30 In addition, whenever possible, equipment will be removed as whole items to minimize exposure times and creation of loose contamination. The thirteen main D&D tasks are discussed in the following pages.

Task 1: Remove Reactor Core Structure The objective of this first -task will be the removal of the reactor core support structure from the pool. The major items include the following:

" Reactor Core Box and Grid

" Control Blade Drive Shafts and Drive Mechanisms

" Regulating Blade

" Safety Blades

" Ion Chambers and associated instrumentation hardware

" Start-up Counter Assembly and Guide Tube Assembly

" Sample Irradiation

" Reactor Suspension Posts The reactor core structure is suspended by a four-post aluminum suspension frame, attached to the bridge spanning the top of the pool. Activation levels and the resulting radiation exposure rates emanating from the core structure are low - in the range of a few mrem/hr. Because of the low radiation levels, remote underwater disassembly of the structure will not be required. Additionally, because the core structure is small, it will not need to be segmented to fit into waste containers.

The first items to be removed from the core will be the regulating blade, which contain the bulk of the reactor's radioactivity and are the largest contributor to radiation exposure rates. The four safety / control blade drive mechanisms and drive shafts will be decoupled at the reactor bridge. The control drive shafts and the attached blades will be lifted out of the core box by workers latching onto the blades with long-handled tools. The blades and shafts Will be raised to the surface of the pool and decoupled. The control blades, one of which is the regulating blade which has the highest estimated radiation levels of 17 mR/hr at one meter, will be manually placed into steel waste container boxes. The control blade drive shafts will be manually cut by mechanical means (saws or shears), and sized to fit into the waste boxes.

Exposure rates from the remaining pool items will be sufficiently low as to safely allow draining of the pool prior to further disassembly of the reactor core. The pool will be drained in a controlled manner using the existing liquid effluent discharge protocols. Pool water will be drained to temporary batch tanks, sampled, analyzed and treated (if necessary) and then released to the sanitary sewer system. Radiation

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 2, Page 12 of 30 levels above the pool water will be monitored as the water level is lowered to assure that no un-expectedly higher sources of radiation are exposed.

Loose surface contamination levels on the exposed surfaces will be checked as the pool water is drained, to assure that no sources of potential airborne contamination are exposed.

If significant loose contamination is indicated, the exposed surfaces will be washed down with water sprays and the loose contamination will be collected along with the drained pool water.

After the pool has been drained, the reactor core box will be removed. The core box consists of the core box sides, grid plate, control blade shrouds, and other small, miscellaneous hardware items. The entire reactor core box-structure is small in size (about 36 inches high by 21 inches wide. by 28 inches long) and will be removed whole, without the need for segmentation.

After removal of the core box structure, radiation levels should be low enough to allow manned entry into the pool. The reactor core structure will be unbolted from the suspension posts and floor locating lugs using hand tools, and slid laterally to clear the suspension frame assembly. A gantry crane will be set up to span the top of the reactor pool and will be used to hoist the core box out of the pool.

Once hoisted out of the pool, the core box will be placed directly into a waste container staged nearby.

After removal of the core box, workers will then remove the suspension frame, ion chambers and start-up counter / guide assembly. These items form one integral structure, with the ion chambers being located inside of three of the four hollow corner posts that comprise the lower portion of the suspension frame assembly.

While the suspension assembly is hanging from the reactor bridge above the pool, workers will manually segment the assembly into manageable-sized pieces using saws or shears.

Segments will be hoisted out of the pool using the previously installed gantry crane, and placed into waste containers staged nearby.

Task 2: Remove Contents of Thermal Column This task removes the contents of the thermal column. The thermal column, a horizontal, aluminum-lined penetration through the biological shield, is filled with graphite blocks. The individual graphite blocks have an approximately four-inch square cross-section, and vary in length up to about three feet long. The graphite blocks are stacked in alternating perpendicular rows; some of the blocks contain embedded stainless Helicoils to facilitate their removal using a threaded T-bar.

During scoping surveys, radiation exposure rates emanating from the graphite and Helicoils were found to be very low - 0.01 and 0.2 mR/hr on contact, respectively.

Smear samples taken on the graphite did not reveal loose graphite dust or radioactive surface contamination.

The graphite blocks will be manually un-stacked and placed into waste containers staged nearby.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 2, Page 13 of 30 Task 3: Remove Reactor Pool Equipment The objective of this task will be to remove the remaining activated equipment from the reactor pool. The equipment is expected to exhibit very low radiation levels, generally less than one mR/hr. Much of pool equipment is either attached to the aluminum pool liner or embedded into the underlying concrete structure. Reactor pool equipment includes the following items:

" Beam port extension

" Beam port shutter and shutter housing

" Core box locating rails

" Spent fuel racks The beam port extension is a six-inch diameter aluminum tube that extends between the interior face of the biological shield and the reactor core box. It will be unbolted from the beam port shutter assembly for removal from the pool. The beam port shutter is a movable shield inside the beam port shutter housing: the beam port shutter and its attached lifting rod will be removed from the housing. The beam port extension and the beam port shutter (and lifting rod) will then be hoisted out of the pool using a gantry crane and placed into a waste container staged nearby.

The remaining items will require destructive removal techniques that have a potential for creating loose debris. Accordingly, engineering controls will be used as required to contain and collect any potentially contaminated debris. Such controls may include enclosing the top of the pool with tarpaulins and using HEPA-filtered ventilation to maintain inward air flow and the use of HEPA vacuums for collection of debris. The sequence of equipment removal will be determined by an ALARA evaluation, with the items with the highest exposure rates being removed first.

The beam port shutter housing is an aluminum box welded to the aluminum pool liner and partially embedded in the biological shield concrete. The aluminum liner surrounding the housing will be cut and removed by sawing, drilling or chisel cutting.

Jack hammering and concrete drilling will be used to break away surrounding concrete to free the housing. The housing will then be saw-cut from the remaining beam port tube still embedded in the biological shield.

The core box locating rails are two aluminum T-channels attached to the pool liner floor that provided support for the reactor core box. The rails will be detached from the pool liner by sawing, drilling or chisel cutting. The two spent fuel racks are also attached to the liner on the pool floor, and will also be detached from the pool liner by sawing, drilling or chisel cutting.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 2, Page 14 of 30 Task 4: Remove Reactor Pool Water Treatment System The objective of this task will be the disassembly and removal of piping and equipment comprising the pool water treatment system. This system is external to the reactor pool, and is comprised of the following items:

" Ion exchange column

" Filter housing

" Pump

" Hold-up tank

" Piping (metal and plastic, all less than two-inch diameter)

Radiological surveys conducted in 2008 did not reveal any radiation exposure rates above background from this equipment, and internal contamination levels are expected to be low, if at all detectable. These are small items, the largest of which is the ion-exchange column, approximately 7.5 feet long by 1.5 feet in diameter.

The equipment will be drained (and will have been previously de-energized during initial decommissioning set-up activities). Generally, the equipment will not need to be segmented to fit into waste containers. Connecting piping and anchor bolts will be saw-cut, sheared or un-threaded.

Catch basins, drop cloths and HEPA vacuums will be used to control and capture any potentially contaminated debris.

Task 5: Remove Liners and Embedments from Biological Shield The objective of this task will be the removal of contaminated or activated portions of the liner from within the pool, thermal column, beam port and gutter drains; and removal of contaminated piping embedded within the biological shield. The exact amount and location of liner that will need to be removed cannot be determined at this time. Characterization will take place after the pool has been drained and activated equipment removed, at which time direct radiation and contamination measurements can be made to direct the work activities.

At the time of construction, the liners were pre-fabricated from 1/4-inch thick sheets of aluminum, and used as the interior form for casting-in-place the concrete biological shield. Removal will entail pre-cutting the liner in place and prying the pieces free from the underlying concrete. The aluminum liner will be mechanically cut (e.g., sawing, drilling or chisel cutting). The manageable-sized pieces will be mechanically pried free using pry bars, jack hammers or hydraulic spreaders.

The beam port liner embedded in the biological shield (and potentially activated concrete surrounding it) will be removed by over coring around the beam port liner, at a diameter of approximately 16-to-18 inches.

A contractor specializing in concrete cutting and coring will be utilized for this operation. Once over cored, the core containing the liner will be withdrawn from the biological shield using chain

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 2, Page 15 of 30 falls and a portable gantry crane. The resulting core, approximately 5.5 feet in length, may be cross-cut into smaller, more manageable segments. The segments will then be placed into waste containers.

Concrete coring and -cross cutting will necessitate the use of water for cooling and lubrication, and to prevent the generation of any potential airborne dust.

The resulting slurry will be controlled, collected and captured as a waste material.

Berms, catch basins and plastic drapes, as well as wet / dry HEPA vacuums will be used to capture and collect the slurry. The resulting slurry will be allowed to congeal and further dried by mixing with cement, if required, and placed into waste containers.

Potentially contaminated embedded pipes within the biological shield include the thermal column and beam port vents and drains, the inlet and outlet pipes for the pool water treatment system, gutter drain lines and the beam port shutter lifting rod sleeve. There may be other, non-system-related embedments present (such as electrical conduits) that will be evaluated for removal on a case-by-case basis. Such pipes (or conduits) may or may not be contaminated. The interior of the embedded pipes will be checked for contamination at accessible openings, using swabs and small-diameter detectors.

If contamination is detected, or they can not be adequately accessed or contamination levels quantified, the embedded pipes will be assumed to be contaminated above release criteria. If these items require removal, they will either be over cored or the surrounding concrete jack hammered to expose the items for removal.

Task 6: Decontaminate Concrete Biological Shield The objective of this task will be to determine if activated or contaminated biological shield concrete is present, and if found removed to meet release criteria. By the start of Task 6, all contaminated and activated items will have been removed from the biological shield. The-remaining concrete will be surveyed and sampled for activation and surface contamination. Added attention for characterizing activation will be given near the beam port, thermal column and floor beneath the core box.

Due to the reactor's low power level and limited use, extensive and deep neutron activation is unlikely, if at all present. Added attention for characterizing surface contamination will be given if signs of water leakage are observed, and near locations having higher probabilities of leakage, such as liner seams and penetrations.

While activated or contaminated concrete is not expected, if it is found above release criteria, it will be removed by mechanical means, such as surface chipping, jack hammering or hydraulic splitting. This process would be iterative, where material would -be removed, resulting surfaces characterized, and the procedure repeated until achievement of release criteria.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 2, Page 16 of 30 These removal techniques have a potential for creating loose or airborne dust.

Accordingly, engineering controls will be used as required to control, contain and collect any potentially contaminated dust and debris. Such controls may include enclosing the top of the pool and other openings, use of HEPA-filtered ventilation for inward air flow, HEPA vacuum collection of debris, and water misting.

Task 7: Remove Liquid Effluent Discharge Equipment The objective of this task will be to gain access to the interior of the liquid effluent discharge equipment for the purpose of radiological characterization, to determine if remediation is required, and if necessary removing them as radioactive waste.

The liquid effluent discharge equipment consists of a few feet of small-bore piping, which connects the hold-up tank (removed in Task 4) and reactor pool water discharge pipe to a locked valve, which in-turn connects to a pipe that discharges to the sanitary sewer system. The piping is partially embedded in the concrete floor of the building's lower level, with the valve located in a floor pit.

The radiological condition of the interior of this piping and valve is not known at this time.

Based on historical site assessment information and scoping survey results, these items may not be radiologically contaminated.

Pipes and the valve will be mechanically segmented, at non-embedded locations, to provide access to interior pipe surfaces. The interior of the pipes will be checked for contamination at accessible openings, using swabs and small-diameter detectors. If contamination is detected, or they can not be adequately accessed or contamination levels quantified, it will be assumed that the piping is contaminated above release criteria.

Otherwise, they will be abandoned in-place.

If these items require removal, they will either over cored or the surrounding concrete jack hammered to expose the items for removal. Additionally, if radioactive contamination is found, the first sewer manhole receiving discharges from this piping will also be checked for radioactive contamination and decontaminated or removed as required.

Task 8: Remove Exhaust Ventilation Equipment The objective of this task is the removal of ventilation ducts that could possibly containing internal contamination, and the evaluation of down-stream building ventilation equipment to determine if additional remediation is warranted.

The exhaust ventilation equipment consists of about 50 linear feet of small-diameter (< 4-inch), light-gauge metal duct work, with two inline electric booster fans, that are connected to the beam port and thermal column vent pipes embedded in the biological shield. During reactor operation the ducts drew a vacuum on the beam port and thermal column such that any gasses generated (e.g., Ar-41, half-life 1.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />) would be routed to the building's main exhaust ventilation system.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 2, Page 17 of 30 The ventilation duct will be segmented and removed using mechanical methods, such as saws or shears, and placed into waste containers. The duct will be removed back to the point where it enters the building's main exhaust ventilation system.

The interior of the building's main ventilation system will be checked for contamination at accessible openings, using swabs and small-diameter detectors.

While not expected, if contamination is detected, or the system's interior can not be adequately accessed or contamination levels properly quantified, the ventilation system will be assumed to be contaminated above release criteria and removed.

Otherwise, it will be left in-place. If radioactive contamination is found in the ventilation ducts, the discharge stack on the roof (and surrounding roofs) surfaces will also be checked for radioactive contamination and decontaminated or removed as required.

Task 9: Remove Remaining Equipment The objective of this task is to survey, evaluate and/or remove any fixtures or equipment remaining within the WPI LCWNRF, where surface contamination could have been missed or hidden in, under or behind equipment or fixtures.

Such remaining items may include sinks, old tools, work benches, etc.

Task 10: Decontaminate Building Interior Surfaces The objective of this task is to provide added assurance that the facility is ready for conducting the FSS.

This task will principally involve taking radiological measurements to identify any contamination within the facility that may have been missed during prior tasks. If any remaining surface contamination is identified it will be remediated at this time.

Based upon the results of the historical site assessment and scoping surveys, any additional structural decontamination is expected to be minimal.

Task 11: Prepare Waste Packages for Transportation and Disposal The objective of this task will be the final disposition of the radioactive wastes generated during the decommissioning.

Due to the relatively small amount of waste expected to be generated (less than 600 cubic feet), only one truck shipment should be required. As such, containers of waste materials will be staged and prepared for shipment at the conclusion of the D&D activities. Section 3.2 describes the waste management activities to be performed during this task.

Task 12: Demobilize and Prepare Reactor Rooms for Final Status Survey During this task, the WPI facility will be prepared for the FSS. Contractors' tools and equipment used during the decommissioning will be surveyed, decontaminated (if necessary), and removed from the facility. Final housekeeping will be performed to remove debris, dust or protruding objects from surfaces, which would otherwise

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 2, Page 18 of 30 interfere with the final survey. Ladders, scaffolding and equipment to be used for the final survey will be staged.

Task 13: Perform License Termination Activities During this task, the FSS Plan will be prepared, implemented and the final report prepared and submitted to NRC with a request for termination of the R-61 license.

Section 4.0, Proposed Final Radiation Survey Plan, describes these activities.

2.3.2 Schedule The decommissioning project schedule is presented as Figure 2.3.

The January 2010 start date shown on the schedule was chosen for illustration purposes only, with the actual start date based on the timing of NRC approval of this DP and the DOE removal of the fuel. Due to the facility's small size, the tasks listed above will essentially be performed in series, by a single work crew. Some task work may be performed in parallel to level work loads, where radiation and industrial safety interferences do not exist. The duration of actual D&D work is expected to be less than 12 months. However, due to WPI being an active educational institution, some gaps in the work may need to be scheduled so as to avoid disruption of WPI's normal activities. Additionally, changes to the schedule may be made at WPI's discretion as a result of resource allocation, availability of a radioactive waste burial site, ALARA optimization and force majeure considerations. Notwithstanding the foregoing, WPI does intend to complete the decommissioning project as soon as possible.

Based on the information in this Decommissioning Plan (DP), WPI anticipates that a formal request for termination of Facility License No. R-61 will be submitted to NRC within 24 months after the approval of the DP is received from NRC.

Worcester Polytechnic Institute Reactor Decommissioning Plan Document W19-1579-003, Rev. 0 Section 2, Page 19 of 30 FIGURE 2.3 LCWNRF DECOMMISSIONING SCHEDULE 2

3 5

6 7

16 F12 17 18 r--

15 16 13 16 19

1. Phae : Pe.Oecw missogActW~ies (Pesfoonedpuioi toNRC P

Appiwi

2. PhasecDe ussiog Moi*znActWMte 21 Off-seNMolzIA 22 On-skeMobitMo

3. NwHae:D minafad D~smalfn9Actele 3.1 TaAI 1:ReMwReatoi Noe Stritia 312 Task t2Raw Coltea d ThesnnwlCoinimn 3.3 TaskIPomRaol IEJeNt 3.4 Task t kmw Pool 4 W Nes "i~

tSsem 15 Tat 5RemMwRe~tePool ineisand 86MddeDiaiiand Pipe 3.6 Task t Deceannmne Conc*

Bkloogcal Shield 3.1 TakRamblafuilkohchaipeEquipment 3.8 Ta&Rmkit~w dVell Monyem 3.9 TMOkROM ninNRm~

Eq ieoit 3.10 Tatk It.Deceia~nhoi BidI lntefio SinIaces 3.11 Taskt i: WastetManeMe 3.12 Task 12 uob~e a Ptepe Me

~ifofFSS 3.13 Task 13Peifofnn kewTeimkialinAct~s 4 Phaswt:Ste Restotatn1Post Lk.TeuiR) tr 1,2010 r22O1O 1r3.2MO 4 r42MO Ql0,31 12, Dec A Feb Mar AVr May Am A Aug Sep O~ NO ec J IFeb M& AVr 2.4 DECOMMISSIONING ORGANIZATION AND RESPONSIBILITIES WPI is committed to, and retains ultimate responsibility for, full compliance with the Institute's existing NRC reactor licenses and applicable regulatory requirements during decommissioning. WPI's policies and goals will be followed to ensure high standards of performance in accomplishing the decommissioning tasks.

WPI intends to utilize experienced and qualified consultants and contractors to perform the physical decommissioning work. WPI will select the contractors after NRC's approval of the DP.

The Reactor Director (from the Office of the Vice President for Finance and Operations) and the WPI Radiation Safety Officer (RSO),

with assistance of the Radiation Health and Safety Committee (RHSC) will monitor performance of the consultants and contractors to ensure that the decommissioning work is being performed safely and according to federal, state and local regulatory requirements (NRC, EPA, DOT, etc.).

Consistent with WPI policy, the RHSC has certain responsibilities to review and approve policies, procedures, programs and facilities pursuant to the safe use of

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 2, Page 20 of 30 radiological materials and radiation-producing equipment. The RHSC's jurisdiction will extend to all decommissioning activities dealing with radioactive material and radiological controls.

The RSO and the RHSC will approve consultant and contractor plans, policies and procedures used during the decommissioning as described in this DP.

The planned organization for the WPI LCWNRF Reactor Decommissioning Project (WPI-RDP) is shown in Figure 2.4. Individuals performing the functions may vary over the project duration; however, the specified functions will be maintained.

Consultants, contractors and subcontractors performing work under this DP will be required to comply with applicable WPI policies, procedures and regulatory commitments.

2.4.1 WPI Reactor Director The WPI Reactor Director functions include:

" Controlling and maintaining safety during decommissioning activities and protecting of the environment

" Reporting of performance

" Approving minor changes to the DP and procedures (which do not change the original intent and do not involve an un-reviewed safety question)

" Oversight and coordination of WPI functional groups, and contractors and consultants

" Ensuring that the conduct of decommissioning activities complies with applicable regulations and is in accordance with WPI licenses The minimum qualifications for the WPI Reactor Director are:

" Currently holds (or previously held) position of Reactor Director

" Ten years of management experience in the field of regulatory compliance

• Familiarity with the WPI LCWNRF Appropriate training in radiation.protection, nuclear safety, hazard communication, environmental protection and industrial safety

Worcester Polytechnic Institute Reactor Decommissioning Plan Document W19-1579-003, Rev. 0 Section 2, Page 21 of 30 FIGURE 2.4 WPI-RDP DECOMMISSIONING ORGANIZATION WORCESTER POLYTECHNIC INSTITUTE Board of Trustees WPI Radiological Protectbio Technicians Laborers Laboratory Services LEGEND:

Line of Responsibility Line of Communication -

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 2, Page 22 of 30 2.4.2 WPI Radiation Safety Officer The WPI Radiation Safety Officer (RSO) shall be responsible for providing radiological support in the decommissioning of the WPI LCWNRF. This function ensures that the activities involving potential radiological exposure are conducted in compliance with the applicable licenses, federal and state regulations, and WPI standard operating procedures. The position includes responsibility for maintaining the surveillance and monitoring program and for HP radiological protection procedures.

The RSO will have oversight of all D&D operations. The scope of his oversight will include all D&D operations that involve work with'systems or materials that have a radiological component. The minimum qualifications for this position are:

• An advanced degree in health physics or a related field

• Ten years operational experience related to radiation safety

• Currently holds (or previously held) position of RSO The RSO is responsible for ensuring that:

" Radiological controls are in place prior to and during any work involving radiation

" Applicable license conditions are satisfied

" Applicable state and federal regulations are met The RSO has the authority to:

• Implement any actions necessary to ensure that radiological controls are implemented and followed

• Immediately stop or modify radiological work determined to be unsafe 2.4.3 Decommissioning Consultant WPI may elect to use a decommissioning consultant to assist WPI with the management of the decommissioning.

The consultant would provide project management and day-to-day oversight of the contractors, as well as providing WIPI with technical expertise needed for the project.

While more than one individual may provide assistance to WPI, the minimum qualifications for the lead consultant are:

• An advanced degree in health physics, radiological engineering or a related field

• 10 years supervisory experience related to radiation safety

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 2, Page 23 of 30 10 years operational experience related to radiological decommissioning, including completion of a minimum of one NRC-licensed research reactor decommissioning project 2.4.4 Contractor Assistance WPI management, with assistance from its Decommissioning Consultant, will select qualified contractors to perform the physical decommissioning of the LCWNRF. In addition to planning and implementing the physical decommissioning activities, the contractors (which may include the Decommissioning Consultant's organization) will provide required HP support, radiation surveys, and waste packaging, processing, and shipping.

Contractors will be selected after WPI's receipt from NRC of the license amendment authorizing decommissioning.

The selection process will be based on a defined scope of work and selection criteria for contractor qualifications, experience, and reputation. The required contractor qualifications and experience shall include the following:

a.

Demonstration of experience in the performance of the following tasks, as applicable to their roles and responsibilities:

" Integration of decommissioning, dismantlement and demolition plans

" Waste management and other methods used to minimize final waste disposal costs

" Radiological decontamination and remediation of facilities and equipment

" Use of survey equipment and protocols suitable for compliance with current NRC or MARSSIM survey criteria

" Development and execution of radiation protection, environmental protection and industrial safety programs that will be used during the D&D

" Selection, design and/or procurement of appropriate containers and packaging for radioactive and hazardous waste, and transportation to approved treatment and disposal facilities

" Performing license termination surveys on a project of similar size and scope

" Waste management (packaging, manifesting and transportation) of radioactive waste

b.

The contractor selected for waste management must be able to provide a QA program that meets the requirements under Code of federal Regulations (CFR) 10 CFR 71, Packaging and Transportation of Radioactive Material, Subpart H, Quality Assurance.

Worcester Polytechnic Institute Document W1941579-003, Rev. 0 Reactor Decommissioning Plan Section 2, Page 24 of 30

c.

The contractors will be required to provide qualified personnel and programs, including but not limited to the following areas of expertise, as applicable to their roles and responsibilities:

" Project management

" Health Physics

" Radiological engineering with MARSSIM survey experience

" Radioactive waste management

" Industrial and occupational safety

" Civil and mechanical engineering 0 Quality assurance

" Construction supervision

" Project controls

" Decontamination and waste handling

" Radiological safety engineering

" Worker training and qualification 2.5 TRAINING PROGRAM Individuals (employees, contractors and visitors) who require access to the work areas or a radiologically restricted area will receive training commensurate with the potential hazards to which they may be exposed.

Radiation protection training will be provided to personnel who will be performing remediation work in radiological areas or handling radioactive materials.

The training will ensure that decommissioning project personnel have sufficient knowledge to perform work activities in accordance with the requirements of the radiation protection program and accomplish ALARA goals and objectives.

The principle objective of the training program is to ensure personnel understand the responsibilities and the required techniques for safe -handling of radioactive materials and for minimizing exposure to radiation.

Training records will be maintained which will include trainee's names, dates of training, type of training, test results, authorization for protective equipment use, and instructor's name.

Radiation protection training provides the necessary information for workers to implement sound radiation protection practices.

The following are examples of the training programs applicable to remediation activities.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 2, Page 25 of 30 2.5.1 General Site Training WPI will be implement a general training program designed to provide orientation to project personnel and meet the requirements of 10 CFR Part 19, Notices, Instructions and Reports to Workers: Inspection and Investigations (Ref. 14).

General Site Training (GST) will be required for all personnel assigned on a regular basis to the decommissioning project. GST will include:

" Project orientation/access control

" Introduction to radiation protection

" Quality assurance

" Industrial safety

" Emergency procedures 2.5.2 Radiation Worker Training All individuals directly associated with the WPI-RDP will be required to undergo Radiation Worker Training (RWT), which will include the following topics:

" Fundamentals of radiation

" Biological effects of radiation

" External radiation exposure limits and controls

" Internal radiation limits and controls

" ALARA Program (program, objectives, investigational limits, keeping doses ALARA)

• Contamination limits and controls

• Management and control of radioactive waste

• Use of personal protective equipment Personnel who have documented equivalent RWT from another site may be waived from taking training except for training on site specific requirements such as, WPI's administrative limits and emergency response actions. Personnel will be required to pass the written examination and practical exercises.

2.5.3 Respiratory Protection Training Generation of airborne radioactivity and extensive loose surface contamination during the decommissioning activities is not expected, and use of respiratory protective devices is unlikely to be needed.

However, if personnel whose work assignments do require the use of respiratory protection devices, they will receive respiratory protection training in the devices and techniques that they will be required to use. The training program will follow the requirements of:

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 2, Page 26 of 30 0 10 CFR 20 Subpart H, Respiratory Protection and Controls to Restrict Internal Exposure in Restricted Areas (Ref. 15) 0 Regulatory Guide 8.15, Acceptable Programs for Respiratory Protection (Ref.

16) 0 NUREG 0041, Manual of Respiratory Protection Against Airborne Radioactive Materials (Ref. 17) 0 29 CFR 1910.134, Respiratory Protection (Ref. 18)

Training would be provided by either the Decommissioning Consultant or a specialty contractor and would consist of a lecture session and a simulated work session. Personnel who have documented equivalent respiratory protection training may be waived from this training. Use of respiratory protective devices will require specific medical evaluation and respiratory device fit testing.

2.6 DECONTAMINATION AND DECOMMISSIONING DOCUMENTS AND GUIDES Health physics, industrial safety criteria and other standards that guide the activities described in this DP are discussed in the following sections of this DP:

• Section 3.1 Radiation Protection

• Section 3.2 Radioactive Waste Management

• Section 3.3 General Industrial Safety Program

• Section 3.4 Radiological Accident Analyses Relevant documents and guides used are noted therein.

2.7 FACILITY RELEASE CRITERIA The proposed decommissioning alternative that has been presented in this DP does not necessitate the dismantlement of the building that houses the WPI reactor. The results of the historical site assessment and radiological surveys indicate that the building structures (exclusive of the reactor pool./ biological shield structure) may be directly releasable, without the need for decontamination.

This section provides the specific criteria for unrestricted release of the WPI LCWNRF. The final release survey will use the Derived Concentration Guideline Levels (DCGL's) developed from the current NRC guidance for license termination in 10 CFR Part. 20, Subpart E, Radiological Criteria for License Termination, Standards of Protection Against Radiation (Ref. 19).

1-0 CFR 20.1402, Subpart E, Radiological Criteria for Unrestricted Use (Ref. 12),

allows termination of a license and release of a site for unrestricted use if the residual radioactivity that is distinguishable from background radiation results in a

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 2, Page 27 of 30 total effective dose equivalent to an average member of a critical group that does not exceed 25 millirem (0.25 mSv) per year and the residual radioactivity has been reduced to levels that are as low as is reasonably achievable (ALARA).

The current NRC guidance for acceptable license termination screening values (meeting the 10 CFR 20.1402 criteria) for common radionuclides for building surface contamination and surface soil contamination are presented in NUREG-1757, Volume 1, Appendix B, Consolidated NMSS Decommissioning Guidance, Decommissioning Process for Materials Licensees, (Ref. 20). As such, an ALARA analysis is not needed.

As stated in NUREG-1727, Appendix D, NMSS Decommissioning Standard Review Plan (Ref. 6), "in light of the conservatism in the building surface and surface soil generic screening levels developed by the NRC staff, the staff presumes, absent information to the contrary, that licensees or responsible parties that remediate building surfaces or soil to the generic screening levels do not need to demonstrate that these levels are ALARA."

Upon completion of the decontamination and remediation activities (e.g., see Section 2.3 Decommissioning Tasks), WPI will complete a FSS of the WPI reactor facility using the method described in NUREG-1575, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM) (Ref. 11). To support the license termination request, the results of the survey(s) will be summarized in a report which will be submitted to NRC, as required by NUREG 1537 (Ref. 5).

If it is impractical or not possible to satisfy release criteria (or conclusively demonstrate that they have been met), the location/item will be treated as radioactively contaminated and dispositioned as low-level radioactive waste.

Based upon radiological surveys and historical site assessment, there is no known or suspected potential for surface soil contamination and release criteria for soil will likely not be needed. However, if small amounts of concrete with volumetric residual radioactivity need to be released, the volumetric soil release criteria may be used for that purpose.

The release criteria will be based upon use of the sum-of-fractions unity rule, based on relative concentrations of radionuclides on or in the material and their respective release criteria if more than one radionuclide is present. If additional screening values are required for nuclides not included in Table 2.2 or Table 2.3, they will be calculated using the NRC's COMPASS computer code (Ref. 21) for planning and assessing MARSSIM site surveys, with default values.

Worcester Polytechnic Institute Reactor Decommissioning Plan Document W19-1579-003, Rev. 0 Section 2, Page 28 of 30 TABLE 2.2 LICENSE TERMINATION SCREENING VALUES FOR BUILDING SURFACE CONTAMINATION ACCEPTABL ESCREENING iLEVELS*

RADiONUCLIDE, SY. MEL*

FORUNIRESTRICTED RELEASEr, Hydrogen-3 (Tritium) 3H 1.2e+08 Carbon-14 14C 3.7E+6 Sodium-22 22Na 9.5E+03 Sulfur-35 35S 1.3E+07 Chlorine-36 36C1 5.03+05 Manganese-54 54Mn 3.2E+04 Iron-55 55Fe 4.5E+06 Cobalt-60 60Co 7.1E+03 Nickel-63 63Ni 1.8E+06

'Strontium-90 90Sr 8.7E+03 Technetium-99 99Tc 1.3E+06 Iodine-129 1291 3.5E+04 Cesium-137 137Cs 2.8E+04 Iridium-192 1921r 7.4E+04

• Screening levels are based on the assumption that the fraction of removable surface contamination is equal to 0.1. For cases when the fraction of removable contamination is undetermined or higher than 0.1, users may assume, for screening purposes, that 100 percent of surface contamination is removable, and therefore the screening levels should be decreased by a factor of 10. Alternatively, users having site-specific data on the fraction of removable contamination, based on site-specific resuspension factors, (e.g., within 10 percent to 100 percent range) may calculate site-specific screening levels using COMPASS Version 2 (Ref 21).

• • Units are disintegrations per minute (dpm) per 100 square centimeters (dpm/lO0 cm2). One dpm is equivalent to 0.0167 Becquerel (Bq). Therefore, to convert to units of Bq/m2 multiply each value by 1.67. The screening values represent surface concentrations of individual radionuclides that would be deemed in compliance with the 0.25 mSvlyr (25 mrem I year) unrestricted release dose limit in 10 CFR 20.1402. For radionuclides in a mixture, the "sum of fractions" rule applies (see Part 20, Appendix B, Note 4).

Worcester Polytechnic Institute Reactor Decommissioning Plan Document W19-1579-003, Rev. 0 Section 2, Page 29 of 30 TABLE 2.3 LICENSE TERMINATION SCREENING VALUES FOR SURFACE SOIL SURFACE SOIL'SCREENING RADIONUCLIDE*-

SMoL vALS,*

FOR UNRESTRICTED RELEASE (.PCi/g)"*~

Hydrogen-3 (Tritium) 3H 1.1E+02 Carbon-14 14C 1.2E+1 Sodium-22 22Na 4.3E+00 Sulfur-35 35S 2.7E+02 Chlorine-36 36C1 3.6E-01 Calcium-45 45Ca 5.7E+01 Scandium-46 46SC 1.5E+01 Manganese-54 54Mn 1.5E+01 Iron-55 55Fe 1.OE+4 Cobalt-57 57Co 1.5E+02 Cobalt-60 6oCo 3.8E+00 Nickel-59 59Ni 5.5E+03 Nickel-63 63Ni 2.1E+03 Strontium-90 9 0Sr 1.7E+00 Niobium-94 94Nb 5.8E00 Technetium-99 99Tc 1.9E+01 Iodine-129 1291 5.OE-01 Cesium-134 134Cs 5.7E+00 Cesium-137 137Cs 2.8E+4 Europium-152 152Eu 8.7E+00 Europium-154 154Eu 8.OE+00 Iridium-192 192Ir 4.1E+01 Lead-210 210pb 9.OE-01 Radium-226 226Ra 7.OE-01 Radium-226+C 226Ra+C 6.OE-01 Actinium-227 227Ac 5.OE-01 Actinium-227+C 227Ac+C 5.OE-01 Thorium-228 228Th 4.7E+00 (Note: Table 2.3 is continued on next page)

Worcester Polytechnic Institute Reactor Decommissioning Plan Document W19-1579-003, Rev. 0 Section 2, Page 30 of 30 TABLE 2.3 (continued)

LICENSE TERMINATION SCREENING VALUES FOR SURFACE SOIL SURFA:CESOI.SCREENING RADIONUCLIDE*.

SYMBOL VALUE-S!ý

,OR UNREST.RICTED Thorium-228+C 228Th+C 4.7E+00 Thorium-230 230Th 1.8E+00 Thorium-230+C 230Th+C 6.OE-01 Thorium-232 232TH 1.1E+00 Thorium-232+C 23 2TH+C 1.1E+00 Protactinium-231 2 3 1Pa 3.OE-01 Protactinium-231+C 23 1Pa+C 3.OE-01 Uranium-234 234U 1.3E+01 Uranium-235 235U 8.OE+00 Uranium-235+C 235U+C 2.9E-01 Uranium-238 238U 1.4E+01 Uranium-238+C 238U+C 5.OE-01 Plutonium-238 238Pu 2.5E+00 Plutonium-239 239Pu 2.3E+O00 Plutonium-241 241pu 7.2E+01 Americium 241AM 2.1E+00 Curium-242 241Cm 1.6E+02 Curium-243 243Cm 3.2E+00

• Plus Chain (+C) indicates a value for a radionuclide with its decay progeny present in equilibrium. The values care concentrations of the parent radionuclide, but account for contributions from the complete chain of progeny in equilibrium with the parent radionuclide (NUREGICR-5512 Volumes 1, 2, and 3).

• • These values represent surface soil concentrations of individual radionuclides that would be deemed in compliance with the 25 mremlyear (0.25 mSvlyear) unrestricted release dose limit in 10 CFR 20.1402. For radionuclides in a mixture, the "sum of fractions" rule applies (see Part 20, Appendix B, Note 4).

• • • Screening values are in units of (pCilg) equivalent to 25 mremlyear (0.25 mSvlyear). To convert from pCilg to units of Becquerel per kilogram (Bq/kg) divide each value by 0.027. These values were derived based on selection of the 9 0 t' percentile of the output dose distribution for each specific radionuclide (or radionuclide with the specific decay chain). Behavioral parameters were set at the mean of the distribution of the assumed critical group. The metabolic parameters were set at "Standard Man" or at the mean of the distribution for an average man.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 3, Page 1 of 17

.3.0 OCCUPATIONAL AND PUBLIC HEALTH AND SAFETY 3.1 RADIATION PROTECTION 3.1.1 Ensuring As Low As Reasonably Achievable (ALARA) Radiation Exposures Decommissioning activities at the WPI LCWNRF involving the use and handling of radioactive materials will be conducted in a manner such that radiation exposure will be maintained As Low As Reasonably Achievable (ALARA), taking into account the current state of technology and economics of improvements in relation to the benefits.

ALARA Prog ram A documented ALARA evaluation will be required for specific work evolutions if it is likely that 5% of the applicable dose limits (collective dose) for any of the following may be exceeded:

" Total Effective Dose Equivalent (TEDE) (5 rem)

" The sum of the Deep Dose Equivalent (DDE) and the Committed Dose Equivalent (CDE) to any individual organ or tissue other than the lens of the eye (50 rem)

" Eye Dose Equivalent (EDE) (15 rem)

" Shallow Dose Equivalent (SDE) to the skin or any extremity (50 rem)

The ultimate responsibility for assuring radiation protection and maintaining exposures ALARA during decommissioning remain with the WPI Reactor Director and Radiation Safety Officer (RSO).

Methods for Occupational Exposure Reduction Various methods will be utilized during-the Decommissioning Project to ensure that occupational exposure to radioactive materials is kept ALARA.

The methods include the use of Radiological Work Permits (RWIPs), use of protective equipment, radioactive material handling techniques, and work practices as described in the following subsections. Work will be performed in accordance with NRC regulations and this Decommissioning Plan (DP), and implemented in accordance with approved written procedures.

Radiological Work Permits (RWPs)

A Radiation Work Permit (RWP) will be used for the administrative control of personnel entering or working in areas that have radiological hazards.

Work techniques will be specified in such a manner that the exposure for all personnel, individually and collectively, are maintained ALARA. RWPs will not replace work procedures, but will supplement procedures.

Radiation work practices will be

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 3, Page 2 of 17 considered when procedures are developed for work that will take place in a radiologically controlled area.

Project RWPs will describe the job to be performed, define protective clothing and equipment to be used, and personnel monitoring requirements.

RWPs will also specify any special instructions or precautions pertinent to radiation hazards in the area. Such information will include listing the radiological hazards present, area dose rates and the presence and intensity of hot spots, loose and fixed surface radioactivity, and other hazards as appropriate.

The Health Physics (HP) organization will ensure that radiation, surface radioactivity and airborne surveys are performed as required to define and document the radiological conditions for each work activity.

RWPs for work activities with low dose commitments may be approved by the HP technician or HP supervisory personnel. RWPs for work activities with potentially high dose commitment or significant radiological hazards will be approved by the RSO or Decommissioning Consultant. Use of RWPs will be governed by written and approved procedure. Examples of topics covered by RWP implementing procedures include:

" Requirements, classifications and scope for RWPs

" Initiating, preparing and using RWPs

" Extending expiration dates of an RWP

" Terminating RWPs Respiratory Protection and TEDE ALARA Evaluations WPI anticipates that dismantling activities for the more radioactive components can be performed by unbolting or unlatching, without the need for segmentation.

However, if it is required, segmentation activities will be minimized and limited to mechanical cutting (e.g., sawing or shearing) which does not generate airborne dust, fumes or smoke. If there ever is a potential for generating airborne radioactivity, the use of engineering controls will be the first choice with respect to controlling the hazard.

Engineering controls will be used to mitigate the airborne radiological hazard at the source. Such controls may include contaminant capture with HEPA-filtered vacuums.

While extremely unlikely, if circumstances were to exist where engineering controls are not practical or may not be sufficient to prevent airborne concentrations that constitute an airborne radioactivity area, and worker access is required to such areas, respiratory protective equipment will be utilized to limit internal exposures.

Any situation wherein workers are allowed access to an airborne radioactivity area, or allowed to perform work that has a high degree of likelihood to generate airborne radioactivity exceeding a value of 0.1 Derived Air Concentration (DAC), the decision

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 3, Page 3 of 17 to allow access will be accompanied by the performance of representative measurements of airborne radioactivity to assess potential worker intake.

The results of DAC-hour tracking and air sample results for intake will be documented in accordance with appropriate regulations and implementing procedures.

Control and Storage of Radioactive Materials The WIPI-RDP's HP Program will establish radioactive material controls that ensure:

" Deterrence of inadvertent release of licensed radioactive materials to unrestricted areas

" Confidence that personnel are not inadvertently exposed to licensed radioactive materials

" Minimization of the volume of radioactive wastes generated during the decommissioning All material leaving the Restricted Area will be surveyed to ensure that radioactive material is not inadvertently released from the WPI LCWNRF. Refer to Section 3.1.3 Radioactive Material Controls, for a description of the specific survey methods that will be used.

3.1.2 Health Physics Program General WPI will use written, approved procedures to control the implementation of work during the WPI-RDP. All procedures will be developed and approved in accordance with written policy and procedure.

The WPI RSO, with the assistance of a Decommissioning Consultant, will have full authority to act in all aspects to ensure protection of workers and the public from the effects of radiation. Procedures for conduct of the decommissioning HP program will be will reviewed and approved according to WPI policy.

Audits, Inspections, and Management Review

.Physical decommissioning -work will be subject to continuous oversight by WPI's RSO or Decommissioning Consultant. Additionally, aspects of the project may be evaluated and reported by an applicable contractor's Quality Assurance program through audits, assessments and inspections of various aspects of decommissioning performance, including HP, as described in Section 1.2.5 Program Quality Assurance.

Audits of the decommissioning HP program will be conducted in accordance with the requirements of Code of Federal Regulations (CFR) 10 CFR 20 (Ref. 15).

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 3, Page 4 of 17 Additional independent assessments or management reviews may be performed when deemed appropriate by the WPI Reactor Director.

Health Physics Equipment and Instrumentation Project HP equipment and instrumentation will be suitable to permit ready detection and quantification of radiological hazards to workers and the public. They will be chosen so as to ensure the validity of measurements taken during remediation and final release surveys.

The selection of equipment and instrumentation to be utilized will be based upon detailed knowledge of the radiological contaminants, and concentrations that are expected to exist based on scoping survey measurements or as known from process knowledge of the working history of the WPI LCWNRF. Equipment and instrumentation selection also takes into account the working conditions, contamination levels and source terms that are reasonably expected to be encountered during the performance of decommissioning work, as presented in this DP.

The following sections present examples of the typical equipment and instrumentation that may be used during the decommissioning. It is anticipated that the Decommissioning Consultant and contractors will be supplying equipment for their own use, and while capabilities will remain equivalent, specific manufacturers and model designations may vary from those examples listed herein.

Criteria for Selecting Survey and Monitoring Equipment A sufficient inventory and variety of instrumentation will be maintained on site to facilitate effective measurement of radiological conditions, control worker exposure consistent with ALARA, and to evaluate the suitability of materials for release for unrestricted use. Instrumentation and equipment will be capable of measuring the range of dose rates and radioactivity concentrations expected to be encountered during the LCWNRF decontamination and decommissioning activities, including implementation of a FSS.

The project HP staff will select instrumentation that is sensitive to the minimum detection limits for the particular task being performed, but also with sufficient range to ensure that the full spectrum of anticipated conditions for a task or survey can be met by the instrumentation in use. Consumable sampling supplies (e.g.,

smears, air filters, etc.) will conform to manufacturer and/or regulatory recommendations to ensure that measurements meet desired sensitivity and are valid for the intended purpose.

Storage, Calibration, Testing and Maintenance of Health Physics Equipment and Instrumentation Survey instruments will be stored in a common location under the control of WPI-RDP HP personnel. A formal program will be adhered to throughout the project to

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 3, Page 5 of 17 identify and remove from service inoperable or out-of-calibration instruments or equipment as described in HP procedures.

Survey instruments, counting equipment, air samplers, air monitors and personnel contamination monitors will be calibrated at Technical Specification-required intervals, manufacturer-prescribed intervals (if shorter frequency) or prior to first use.

Such equipment will be calibrated against standards that are traceable to National Institute of Standards and Technology (NIST), in accordance with approved calibration laboratory procedures, HP procedures, or vendor technical manuals. Survey instruments will be operationally checked daily when in use. Counting equipment operability will be verified daily when in use. The personnel contamination monitors or friskers will be operationally tested on a daily basis when work is being performed.

Specific Health Physics Equipment and Instrumentation Use and Capabilities Table 3.1 provides details of typical HP equipment and instrumentation that is planned for use in the WPI-RDP. This list is not intended to be all-inclusive or exclusive, and equivalent makes and models may be substituted based on availability.

Policy, Method, Frequency and Procedures The existing framework of WPI's current HP Program will be used as the basis for constructing the WPI-RDP's HP Program.

WPI's existing program will be augmented as necessary using plans and procedures provided by the Decommissioning Consultant and contractors.

Airborne Effluent Monitoring -

During decommissioning activities where a temporary enclosure is in use with a HEPA exhaust system, the ventilation system exhaust points from the temporary barrier will be sampled continuously downstream of the HEPA filtration system. Airborne concentrations within the uncontrolled areas of the building will be maintained at less than environmental airborne concentration effluent limits.

Radiation Surveys - Radiation, airborne radioactivity and contamination surveys during decommissioning will be conducted in accordance with approved HP procedures. The purposes of these surveys will be to (1) protect the health and safety of workers, (2) protect the health and safety of the general public, and (3) demonstrate compliance with applicable license, federal and state requirements, as well as DP commitments.

HP personnel will verify the validity of posted radiological warning signs during the conduct of these surveys. Surveys will be conducted in accordance with procedures utilizing survey instrumentation and equipment suitable for the nature and range of hazards anticipated.

Equipment and instrumentation will be properly calibrated and, where applicable, operationally tested prior to use in accordance with procedural requirements.

Routine surveys will be conducted at a specified frequency to ensure that

Worcester Polytechnic Institute Reactor Decommissioning Plan Document W19-1579-003, Rev. 0 Section 3, Page 6 of 17 contamination and radiation levels in unrestricted areas do not exceed license, federal, state or site limits.

The HP staff will also perform surveys during decommissioning whenever work activities create a potential to impact radiological conditions.

TABLE 3.1 TYPICAL HEALTH PHYSICS EQUIPMENT AND INSTRUMENTATION INSTRUMENT MODEL DETECTOR INSTRUMENT APPLICATION,-

TYPE' RANGE Bicron Micro-Rem meter 1"xl" Tissue equivalent 0-5000 Micro rem/hr Low gamma exposure rates scintillation Ludlum Model 12 count rate meter, 2"x2" NaI wit Ldlm 4-0 sinilaton0-500K cpm Low-level gamma scanning with Ludlum 44-10 scintillation Ludlum Model 12 count rate meter, r"xl" Nal Low-level gamma scanning interior of with Ludlum 44-2 scintillation 0-500K cpm equipment and pipes Ludlum Model 17 Ionization 0-50 R/hr Gamma and Beta exposure rates Ludlu Modl 17chamber Alpha and Bicron labtech dual channel scaler /

beta ratemeter scintillation, 0-1,000K cpm Measurement of surface contamination 100cm 2 Ludlum Model 300 Portable Area GM tube 0.1-1000 mr/hr Portable area radiation alarm Radiation Alarm Ludlum Model 177 Alarming GM tube, Ratemeter pancake, 15 0- 500K cpm Personnel contamination frisking cm 2 Beta Ludlum 2200 ratemeter-scaler, with scintillation, 0-500K cpm, 6 sec.

Beta-counting Ludlum 44-1 12 to 990 min.

-12 cm 2 Ludlum 2200 ratemeter-scaler, with Alpha 050 p,6sc Ludlum 44-2 scintillation, to-500K cmi6se.

Alpha-counting smear and air samples

-12 cm2 Thin end Beta-gamma contamination checks, interior Ludlum Model 12 count rate meter, window GM 0-500K cpm with Ludlum 44-7 tube equipment and pipes Hi-Q Environmental Products TFIA N/A 0-70 CFM High-volume grab air samples Apex Pro 200 N/A 0.8-4 LPM Personnel breathing zone air sampler F&J Model LV-14M Gooseneck "Lo-N/A Low volume air sampling for long-term Vol de Lo o

0.35-3.5 cfm continuous air sampling o-Local continuous particulate airborne Ludlum Model 333-2 air monitor GM 10-103 cpm monitor with alarm capability Personnel Monitoring - Internal and External - External monitoring will be conducted in accordance with approved procedures. Prospective external exposure evaluations will be performed prior to initiating decommissioning activities and whenever changes in conditions warrant. Visitors to the WPI LCWNRF will be monitored in accordance with requirements specified in WPI's HP procedures and according to the radiological hazards of areas to be entered.

Internal monitoring will be conducted in accordance with approved procedures.

This prospective internal exposure evaluation will be performed on an annual basis, at a minimum, or whenever significant changes in planned work evolutions warrant

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 3, Page 7 of 17 additional evaluation. A comprehensive air-sampling program will be conducted during the WPI-RDP to evaluate worker exposures regardless of whether or not internal monitoring is specified.

The results of this air-sampling program will be utilized to ensure validity of specified internal monitoring requirements for decommissioning personnel. If hazards are encountered, at any time during the decommissioning, that may not be readily detected by the preceding measures, appropriate special measures or bioassay will be instituted to ensure the adequate surveillance of worker internal exposure.

Monitoring will be required if the prospective dose evaluation shows that an individual's dose is likely to exceed 10% of the applicable limits, and for individuals entering a high or very high radiation area.

Respiratory Protection -

If required, the WPI-RDP respiratory protection program will include direction for use of National Institute for Occupational Safety and Health/Mine Safety and Health Administration (NIOSH/MSHA) certified equipment. This program will be reviewed and approved by the HP staff to ensure adherence to the requirements :of 10 CFR 20 (Ref. 15).

NIOSH/IMSHA-approved air purifying respirators include full-face-piece assemblies with air purifying elements to provide respiratory protection against hazardous vapors, gases, and/or particulate matter to individuals in airborne radioactive materials areas. The RSO or Decommissioning Consultant will also ensure that the respiratory protection program meets the requirements of 10 CFR Part 20, subpart H (Ref. 15).

Maintenance - When respiratory protection equipment requires cleaning, the filter cartridges will be removed. The respirator will be cleaned and sanitized after use with a cleaner/sanitizer and then rinsed thoroughly in plain warm water in accordance with HP procedures.

Storage - Respiratory protective equipment will be kept in proper working order.

When any respirator shows evidence of excessive wear, distortion or has otherwise failed inspection, it will be repaired or replaced. Respiratory protective equipment that is not in use will be stored in a clean dry location.

Contamination Control - Contamination control measures that will be employed include, as appropriate, the following:

• Worker training will incorporate methods and techniques for the control of radioactive materials, and proper use and donning/doffing of protective clothing

• D&D procedures will incorporate control techniques to minimize spread of contamination during work

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 3, Page 8 of 17

" Radiological surveillance surveys conducted by HP

" Containment devices (e.g., designed barriers, containers and plastic bags, etc.) will be used to prevent the spread of radioactive material

" Physical decontamination of areas or items to mitigate potential contamination migration

" Physical barriers such as Herculite sheeting, strippable paint, and tacky mat step-off pads to limit contamination spread

" Posting, physical area boundaries and barricades

" Step-off pads at the exit point from contaminated areas Personnel entries into radiological contaminated areas will require the use of protective clothing.

Depending upon the conditions outlined in the RWP, such clothing may consist of suitable combinations of the following items:

" Cloth or Tyvek lab coat

" Cloth or Tyvek coveralls

" Cloth or Tyvek hoods or skull caps

" Plastic, cloth or Tyvek calf-high booties

• Rubber, plastic or cloth shoe covers

" Plastic or rubber gloves, which may require cloth liners

• Additional layers of disposable Tyvek coveralls or plastic rain suits P

Face shields, safety glasses, hard hats, safety shoes, ear muffs or other industrial protective devices Access Control - A Restricted Area will be established and properly posted and monitored to prevent unauthorized access.

Engineered Controls - Personnel exposure to airborne radioactive materials will be minimized by utilizing engineering controls such as the following:

0 Ventilation devices - in-place or portable HEPA-filtered ventilation units or vacuum cleaners 0 Containment devices - designed containment barriers, containers, plastic bags, tents, and glove-bags a Airborne source term reduction - application of fixatives prior to handling, misting of surfaces to minimize dust and resuspension Airborne Radioactivity Monitoring - Air sampling will be performed in areas where airborne radioactivity is present or likely.

10 CFR 20.1502(b) requires

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 3, Page 9 of 17 monitoring for the intake of radioactive material if the intake during the year is likely to exceed:

" Ten percent of the annual limit on intake (ALI) for an adult worker, or

" The committed effective dose equivalent 0.1 rem (1.0 mSv) for an occupationally exposed minor or declared pregnant woman WPI will establish monitoring requirements for the project based on prospective estimates of worker intakes and air concentrations. Such estimates will consider the following:

• The quantity of material(s) based upon its physical form and use

• The annual limit on intake for the nuclides of interest

• The release fraction for the radioactive material(s) being handled

• Other factors that may be applicable HP personnel will use technical judgment in determining the situations that necessitate air sampling regardless of results of generalized, prospective evaluations performed for the WPI-RDP.

The purpose of the radiological air sample will be identified prior to specifying a location for an air sampler. The following are a few examples of the variety of reasons that exist for collecting air samples:

" Estimation of worker intakes

" Verification of confinement of radioactive materials

" Early warning of abnormal airborne concentrations of radioactive materials

" Determining the existence of criteria for posting an Airborne Radioactivity Area in accordance with 10 CFR 20 Smoke tubes and buoyant markers (ribbons) may be used to determine airflow patterns in the area. Airflow patterns may be reevaluated if there are changes at the WPI LCWNRF that may impact the validity of the sampling locations. Such factors might include the following:

" Changes in the work process

" Changes in the ventilation system

" Use of portable ventilation that might alter earlier assessments After identifying the purpose for the air sample and establishing flow patterns, air sample locations will be chosen as follows:

• For verification of confinement of radioactive materials:

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 3, Page 10 of 17 o

Locate samplers in the airflow near the actual or potential release point o

More than one sampling point may be appropriate when there are more than one potential or actual release points For estimation of a worker's intake:

o Sampler intakes will be located as close as practical to the worker's breathing zone, without interfering with the work or worker General workplace air sampler intakes will not be placed in or near ventilation exhaust ducts unless their purpose is to detect system leakage during, normal operation, and if quantitative measurements of workplace concentrations are not required. Locations or number of air samplers will be changed when dictated by modifications to facility structure, changes in work processes, or elimination of potential sources.

WPI will maintain a sufficient inventory and variety of operable and calibrated portable and semi-portable air sampling equipment to allow for effective collection, evaluation, and control of airborne radioactive material and to provide backup capability for inoperable equipment. Air sampling equipment will be calibrated at prescribed intervals or prior to use: calibration will be against certified equipment having known valid relationships to nationally recognized standards. Table 3.1 provides a listing of anticipated air-sampling equipment.

When the work being performed is a continuous process, a continuous sample with up to a weekly exchange frequency is appropriate.

Longer sample exchange frequencies may be approved by HP management for situations where airborne radioactive material and nuisance dust are expected to be relatively low. Grab sampling for continuous processes may also be approved by HP management based upon consideration of variability of the expected source term for the facility and process. Grab sampling is the appropriate means of airborne sampling for processes conducted intermittently, and for short duration radiological work that involves a potential for airborne release.

Potential Sources of Radiation or Contamination Exposure Potential sources of radiation or contamination exposure to workers and public, as a result of decommissioning activities, may be assessed by several methods, including:

0 Process and historical knowledge 0 Surveys performed during characterization 0 Pre-work RWP surveys N Previous and current job coverage surveys

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 3, Page 11 of 17 Daily, weekly and monthly routine surveillance surveys Classification of potential sources may also be identified by radionuclide, physical properties, volatility and radioactivity.

Worker exposure to significant external deep-dose radiation fields is considered unlikely during this project, due to the nature and extent of the contaminants present and/or the work precautions and techniques employed. Worker exposure to airborne radioactivity is also unlikely to occur during most dismantling operations/work evolutions unless they involve use of abrasive or thermal cutting methods on activated materials: such cutting methods can volatilize or create loose dispersible contamination. WPI does not plan to use those types of cutting methods on activated materials.

Exposure of the public to external or internal radiation from the WPI-RDP is not considered credible because of the limited levels and extent of radioactivity present and the confinement provided by the facility.

Control of potential sources of radiation exposure to workers and public, as a result of decommissioning activities, will be achieved through, but not limited to, the use of administrative, engineering and physical controls.

Administrative controls consist of, but are not limited to:

" Administrative dose limits that are lower than regulatory limits

" Training

" Radiological surveys In general, physical barriers (such as lockable doors/gates, and, or other applicable barriers) may be used in combination with radiological postings to preclude inadvertent exposure of the general public to radiation or radioactive materials.

Environmental Monitoring During the conduct of decommissioning activities, measurements will be made for radiation and radioactive contamination at locations where direct radiation exposure or contaminant release pathways could exist that could credibly impact the environment or the general public. While releases of radioactive material or increased radiation levels in uncontrolled areas are not expected during decommissioning activities, monitoring will be performed to verify that expectation.

Generally, environmental monitoring will entail placement of dosimeters at selected locations -to monitor for direct radiation exposure.

Such locations will include nearby classrooms, offices or dormitories adjacent -to the LCWINRF frequently occupied by WPI students, staff or visitors. Periodic measurements will be made for radioactive surface contamination at locations Where radioactive contamination

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 3, Page 12 of 17 could migrate out of the facility (e.g., at the LCWNRF's doorway exits and outward air flow paths).

Continuous particulate air sampling will be made at selected locations representative of nearby adjacent uncontrolled areas.

Health Physics Policies for Non-WPI Personnel WPI intends to use non-WPI personnel (contractors) to assist in the decommissioning.

Contractors who will work with licensed radioactive materials will be required to:

" Attend and satisfactorily complete an appropriate radiation safety course

" Provide required exposure history information

" Read and sign an applicable RWP and comply with instructions

" Follow all special instructions given by HP 3.1.3 Radioactive Materials Controls The WPI-RDP radiation protection program will establish radioactive material controls that ensure the following:

" Prevention of inadvertent decommissioning radioactive waste (licensed) material release to uncontrolled areas

" Assurance that personnel are not inadvertently exposed to radiation from licensed radioactive decommissioning waste materials

" Minimization of the amount of radioactive waste material generated during decommissioning.

Pool water releases will be processed using existing pool water treatment equipment and analysis protocols to ensure that discharges to sanitary sewerage will meet the requirements of 10 CFR 20.2003 (Ref. 22).

The potentially contaminated equipment, materials, instrumentation, and tools that are used or encountered during the decommissioning will be handled as described below:

" Items may be surveyed and released on site as clean if the residual radioactivity is less than the values specified in Table 1 of NRC Regulatory Guide 1.86, Termination of Operating Licenses for Nuclear Reactors, June 1974 (Ref. 23)

" Items may be shipped directly for disposal as radioactive waste

" Items may be shipped to a licensed radioactive material processing facility for survey and release, decontamination followed by survey and release, or shipment for disposal as radioactive waste

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 3, Page 13 of 17 Items may be shipped to a licensed facility for holding until they are utilized on another project involving radioactive materials

" Items may be transferred to WPI's Agreement-State radioactive materials license 3.1.4 Dose Estimates The total projected occupational exposure to complete the decommissioning of the Reactor facility is conservatively estimated to be less than 0.5 person-rem.

This estimate is provided for planning purposes only. Detailed exposure estimates and exposure controls shall be developed during detailed planning of the decommissioning activities. Area dose rates used for this estimate are based on 2008 scoping survey data, process knowledge and current survey maps (where available).

The dose estimate to members of the public as a result of decommissioning activities is estimated to be negligible: site perimeter controls will restrict members of the public from entering the area where decommissioning activities are taking place.

This is consistent with the estimate given for the "reference research reactor" in NUREG-0586, Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities (Ref. 24).

In NUREG-0586, the dose to the public during decommissioning (DECON) and truck transport transportation of radioactive waste from the reference research reactor is estimated to be "negligible (i.e., less than 0.1 person-rem)."

3.2 RADIOACTIVE WASTE MANAGEMENT 3.2.1 Radioactive Waste Processing The processes of decontamination and dismantlement of the WPI LCWNRF will result in solid low-level radioactive waste, consisting principally of activated metals, and a small amount of contaminated metals and concrete debris.

No mixed.or hazardous waste is expected to be generated. Lead paint coatings and asbestos-containing materials are likely to be present in the facility, but are not anticipated to be significantly present on materials likely to become radioactive waste. No soil remediation is anticipated which would result in solid radioactive waste. Solid, low-level radioactive waste will be handled (processed and packaged), stored and disposed of in accordance with applicable sections of the Code of Federal Regulations (CFR), disposal site Waste Acceptance Criteria, Massachusetts Department of Environmental Quality requirements, WPI Licenses and Permits, and the applicable implementing plans and procedures.

Radioactive waste processing includes waste minimization or volume reduction, radioactive and hazardous waste segregation, waste characterization, neutralization, stabilization, solidification and packaging.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 3, Page 14 of 17 3.2.2 Radioactive Waste Disposal Low-level radioactive waste will be processed and packaged for disposal at a licensed low-level waste site such as the EnergySolutions, LLC (formerly Envirocare of Utah) site. The volume of low-level radioactive waste is estimated at 600 cubic feet (packaged volume).

All wastes resulting from and dismantling and decontamination are estimated to have concentrations of radioactive materials well below 10 CFR 61 Class A waste limits. The most highly radioactive item of waste arising from the decontamination and dismantling of structures, systems and components will be the activated regulating control blade. The blade is estimated to contain a total of 56 mCi, representing the majority of the waste radioactivity to be generated during the decommissioning program. Table 3.2 provides the estimated concentrations of radioactivity in the regulating control blade.

TABLE 3.2 ESTIMATED CONCENTRATIONS OF RADIOACTIVITY IN THE REGULATING CONTROL BLADE (as of December 2008)

RAD:*IONCLIDE,

CONCENTRATION, Ci/Cubic Meter Mn-54 0.61 Fe-55 43.22 Ni-59 0.05 Co-60 16.91 Ni-63 2.69 All other wastes materials are expected to have radioactive concentrations a small fraction of those listed for the regulating blade. The one Curie Pu-Be source, which would exceed 10 CFR 61 Class A limits, will either be returned to DOE or a source manufacturer, or transferred to WPI's agreement state license until a suitable disposal option becomes available.

10 CFR 61, Licensing Requirements for Land Disposal of Radioactive Waste, Subpart D -

Technical Requirements for Land Disposal Facilities (Ref. 25),

establishes minimum radioactive waste classification, characterization and labeling requirements. WPI's adherence to those requirements will be ensured through the implementation of project packaging and characterization procedures, Disposal Site Waste Acceptance Criteria for the selected disposal site(s) and the Project-Specific Quality Assurance Plan. Training/Qualifications will be provided for project waste management personnel to assure conformance to applicable 10 CFR 61 requirements (Ref. 25) as stated in the specific implementing procedures and plans.

Audits and surveillances will be conducted per the Project-Specific Quality Assurance Plan based on American Society of Mechanical Engineers (ASME) NQA-1 (Ref. 26) and the requirements of 10 CFR 71 (Ref. 9).

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 3, Page 15 of 17 10 CFR 71, Packaging and Transportation of Radioactive Material, establishes requirements for packaging, shipment preparation and transportation of licensed material.

WVPI is licensed by the NRC/ State of Massachusetts to receive, possess, use and transfer licensed byproduct and source materials. 10 CFR 71 requirements will be met through the implementation of the WPI-RDP approved packaging and shipping procedures. Training will be provided to assure conformance to applicable 10 CFR 71 requirements.

Quality Assurance for waste packaging and transportation will conform to 10 CFR 71 Subpart H (Quality Assurance) requirements through the implementation of a WPI-RDP project-specific quality assurance plan.

10 CFR 20.2006, Transfer for Disposal and Manifests, establishes requirements for controlling transfers of low-level radioactive waste intended for disposal at a land disposal facility; establishes a manifest tracking system; supplements requirements concerning transfers and record keeping; and requires generator certification that transported materials are properly classified, described, packaged, marked and labeled, and are in proper condition for transport. These requirements will be met through the implementation of WPI-RDP packaging and shipping procedures with the oversight of WPI's RHSC and Decommissioning Consultant.

Radiological wastes will be disposed of at disposal sites per the applicable disposal site's Waste Acceptance Criteria. Associated implementing plans and procedures will reflect the characterization, processing, removal of prohibited items, packaging and transportation requirements. Appropriate documentation will be submitted to designated disposal sites including, as required, certification plans, qualification statements, assessments, waste stream analysis, evaluations and profiles, transportation plans, and waste stream volume forecasts. Waste characterization, waste designation, waste traceability, waste segregation, waste packaging, waste minimization, and quality assurance and training requirements of the designated disposal sites will be incorporated in implementing procedures to assure conformance to disposal site requirements.

Generator State (Massachusetts) and Treatment/Storage/Disposal Facility State (Utah) requirements for radioactive waste management will be incorporated into plans and procedures to assure conformance with applicable state regulations, licenses and permits.

Applicable state regulations include Massachusetts Department of Environmental Quality requirements and Utah Department of Environmental Quality rules (R313) for the control of ionizing radiation reflected in Envirocare's Utah Radioactive Material License, UT 2300249.

Radioactive waste will be staged in designated controlled areas at the LCWNRF in accordance with the requirements of 10 CFR Part 19 (Ref. 14) and 10 CFR Part 20 (Ref. 19). Measures will be implemented through plans and procedures to prevent the spread of contamination, limit radiation levels, prevent unauthorized access,

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 3, Page 16 of 17 prevent unauthorized material removal, prevent tampering, and prevent weather damage.

The designated areas will be controlled by Radiological Work Permits (RWP).

Radioactive waste material will be packaged for shipment per 10 CFR 61 (Ref. 25) and 49 CFR (Ref. 27), and the designated disposal site's Waste Acceptance Criteria and held in interim storage (staged) until shipped.

Due to the relatively small amount of waste to be generated during the WPI-RDP, it is expected that only one truck shipment will be needed. As such, the waste will be staged on-site until all the waste products have been generated. Radioactive material storage areas will be contained inside posted restricted areas according to WPI-RDP procedures and consistent with 10 CFR 20.

3.3 GENERAL INDUSTRIAL SAFETY PROGRAM WPI, along with its contractors, shall be responsible for ensuring that the WPI-RDP complies with all applicable occupational health and safety requirements.

The primary functional responsibility is to ensure compliance with the Federal Occupational Safety and Health Acts (OSHA) of 1973 (Ref. 28).

Specific responsibilities include:

" Conducting an industrial training program to instruct employees in general safe work practices

" Reviewing decommissioning project procedures to verify adequate coverage of industrial safety and industrial hygiene concerns and requirements

" Performing periodic inspections of work areas and activities to identify and correct any unsafe conditions and work practices

" Providing industrial hygiene services as required

" Advising Project Management personnel on industrial safety matters and on the results of periodic safety inspections All personnel working on the WPI-RDP will receive Health and Safety training in order to recognize and understand the potential risks involving personnel health and safety associated with the work at the WPI LCWNRF. The Health and Safety training implemented on the WPI-RDP is to ensure compliance with the requirements of NRC (10 CFR), the EPA (40 CFR), and OSHA (29 CFR). Workers and regular visitors will be familiarized with plans, procedures and operation of equipment to conduct themselves safely. In addition, each worker must be familiar with procedures that provide for good quality control.

Section 2.5 Training Program, provides additional information.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 3, Page 17 of 17 3.4 RADIOLOGICAL ACCIDENT ANALYSES The inventory of radioactive material expected to be present at the start of decommissioning is significantly smaller than that postulated for the accident scenarios applicable for when the reactor was operational. The reactor fuel will have been removed prior to the start of decommissioning, and the majority of the residual radioactivity (less than 60 mCi) is in the form of activated metals which are not readily dispersible. As such, the existing operational accident analysis continues to bound and by far exceeds the consequences of any accident that could credibly occur during the decommissioning period. As such, a new accident analysis is not required.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 4, Page 1 of 11 4.0 PROPOSED FINAL RADIATION SURVEY PLAN Based upon the results of an historical site assessment and scoping surveys of the Leslie C. Wilbur Nuclear Reactor Facility (LCWNRF), Worcester Polytechnic Institute (WPI) intends, to the extent practicable, to completely remove equipment and items that had become neutron -activated or that contain radioactive contamination.

Such removed materials will be disposed as radioactive waste.

WPI anticipates that most - if not all - of the building structures will be free of residual radioactivity, and will not require decontamination.

The Final Status Survey (FSS) Plan (and subsequent FSS Report) discussed in this section deals with release of the WPI reactor building structure and surrounding environs for unrestricted use.

4.1 DESCRIPTION

OF FINAL STATUS SURVEY PLAN The purpose of the FSS is to demonstrate that the radiological condition of the LCWNRF structures is at or below established release criteria (see Section 2.7). It is anticipated that NRC will then terminate the R-61 reactor license, thereby allowing unrestricted use of all areas of the former LCWNRF.

WPI will develop its FSS Plan using the guidance provided in NUREG-1757, Volume 2, Consolidated Decommissioning Guidance, Characterization, Survey, and Determination of Radiological Criteria, (Ref. 29) and NUREG-1575, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM, Ref. 11).

A FSS Plan will be developed after decommissioning contracts are awarded. The basis for the design of the FSS is MARSSIM. The overall goal of the FSS design is to ensure surveys are planned and conducted in such a manner that ensures the proper decision is made as to whether or not to accept or reject the Null Hypothesis (see further discussion on Null Hypothesis later in this section). The COMPASS computer code (Ref. 21) will also be used as part of the survey planning and survey assessment process.

The major inputs into the FSS planning process are:

" The development of Data Quality Objectives (DQOs) (see Section 4.1)

" The designation of survey units within each area classification (see Section 4.4.2)

M The review of contaminants and establishment of Derived Concentration Guideline Levels (DCGL) (see Sections 2.7 and 4.3)

The selection of appropriate survey instrumentation (see Section 3.1)

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 4, Page 2 of 11 The survey plan will serve as the guidance document for development of the survey package instructions used during implementation of the FSS. The facility will be sectioned into survey units according to guidance provided in MARSSIM. Plots, diagrams, and facility layout drawings will be developed to illustrate the classification of the survey units. In addition to a FSS Plan, survey work packages will be developed for each survey unit. Each survey package will include survey-unit-specific instructions, describe the survey unit size, grid spacing, and scan area prescribed, and specify a prescribed number of static measurements (including the location and spacing).

As the survey progresses, reevaluation of the survey plan may be necessary based on newly acquired survey data. If a condition is discovered that is not encompassed by the survey plan, the survey plan may undergo revision to address the condition.

The condition with the revised survey plan will be fully disclosed and provided to the Decommissioning Consultant and RSO for review and concurrence prior to further performance of the FSS as it applies to the revised information.

To prevent recontamination of the clean areas, WPI will institute administrative and physical access controls of surveyed areas deemed radiologically clean.

Administrative control of surveyed areas will be accomplished by written instruction contained in the final survey plan and by training of project personnel.

Physical control of surveyed areas may be accomplished by placing rope barriers, locking doors where possible, etc. and placing signs to notify personnel regarding the condition of an area.

The FSS Plan will contain the criteria used to assess all final survey data including the statistical tests performed. The Plan will also state conclusions to be drawn based upon statistical test results.

The FSS Plan will be developed according to the guidance provided in MARSSIM and based upon the following assumptions:

• The interior rooms of the LCWNRF are impacted (determined from the results of operational history, scoping survey data and professional judgment).

• The screening activity levels for building surfaces are based on the assumption that the fraction of removable surface contamination is equal to 0.1.

The fraction of activity that is removable will be verified.

If the removable fraction exceeds 0.1 then more decontamination will be performed or site-specific DCGL values will need to be applied.

• Decision Errors - there are two types of decision errors applied to analytical results: Type I (a) and Type II (6) errors. A Type I error, or false positive, is the probability that Null Hypothesis is rejected when it should be accepted.

A Type II error, or false negative, is the probability of determining that Null

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 4, Page 3 of 11 Hypothesis is accepted when it should be rejected. The probability of making decision errors can be controlled by adopting an approach called hypothesis testing.

The Null Hypothesis is treated like a baseline condition and is defined by MARSSIM as:

Null Hypothesis (H.) = residual radioactivity in the survey unit exceeds the release criterion.

This means that the site or survey area is assumed contaminated until proven otherwise. For the purpose of this final survey, Type I or a error will be set at 0.05 (or 5 percent) and Type II or B error will be set at 0.05 (or 5 percent).

For the LCWNRF it is likely that WPI will not need to evaluate survey data relative to a reference background area. However, if there were a need to evaluate survey data relative to a reference background area, WPI would utilize either the Sign test or the Wilcoxon Rank Sum (WRS) test to determine if any residual contamination in a survey unit exceeds the DCGL.

The selection process for determining which test method to use will be documented.

" Any background count rate information required for input into the survey planning process will be obtained prior to performing the FSS. The standard deviation in the background count rate will be calculated based on the data obtained. With the COMPASS code used to automatically calculates the standard deviation.

The background information is stored as part of the general site information and not on a survey unit specific basic.

" Once the final survey has been performed, survey data will be converted to DCGL units and compared to the DCGLs.

Individual measurements and sample concentrations will be compared to DCGL levels for evidence of small areas of elevated activity. Data will then be evaluated using the appropriate statistical test method (i.e., Sign test or WRS test) to determine if they exceed the, release criterion.

" If the release criterion has been exceeded (i.e., Null Hypothesis is proven true) WPI management, with assistance from the Decommissioning Consultant, will determine appropriate further actions. If all data points are less than the DCGL levels, no statistical test need be performed.

4.1.1 Review and Approval of Final Status Survey Plan WPI is utilizing Method 1 from NUREG-1757, Volume 2,

Consolidated Decommissioning Guidance, Characterization, Survey, and Determination of Radiological Criteria (Ref. 29), to submit information to NRC on facility radiation surveys.

WPI has submitted information to NRC on release criteria, characterization surveys, and operational surveys as part of this DP. In addition, WPI has committed to using the MARSSIM approach in developing the final

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 4, Page 4 of 11 radiological survey. Once the design of the final radiological survey for the site has been completed, WPI will submit information to NRC on FSS Design.

The FSS Plan will include sufficient information to allow NRC to determine that the FSS Design is adequate to demonstrate compliance with the radiological criteria for license termination. The information will include:

" A brief overview describing the FSS Design.

" A description and map or drawing of impacted areas of the site, area, or building classified by residual radioactivity levels (Class 1, Class 2, or Class

3) and divided into survey units, with an explanation of the basis for division into survey units. Maps will indicate compass orientation.

" A description of the background reference areas and materials, if they will be used, and a justification for their selection.

" A summary of the statistical tests that will be used to evaluate the survey results, including the elevated measurement comparison, if Class 1 survey units are present, a justification for any test methods not included in MARSSIM, and the values for the decision errors (and) with a justification for values greater than 0.05.

A description of scanning instruments, methods, calibration, operational checks, coverage, and sensitivity for each media and radionuclide.

For in-situ sample measurements made by field instruments, a description of the instruments, calibration, operational checks, sensitivity, and sampling methods, with a demonstration that the instruments, and methods, have adequate sensitivity.

• A description of the analytical instruments for measuring samples in the laboratory, including the calibration, sensitivity, and methodology for evaluation, with a demonstration that the instruments and methods have adequate sensitivity.

• A description of how the samples to be analyzed in the laboratory will be collected, controlled, and handled.

• A description of the FSS investigation levels and how they were determined.

" A summary of any significant additional residual radioactivity that was not accounted for duringý site characterization.

" A summary of direct measurement results and/or concentration levels in units that are comparable to the DCGL and, if data is used to estimate or update the survey unit.

A summary of the direct measurements or sample data used to both evaluate the success of remediation and to estimate the survey unit variance.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 4, Page 5 of 11 4.1.2 Means for Ensuring that all Equipment, Systems, Structures and Site are Included in the Survey Plan Any items left in the LCWNRF after remediation will be radiologically surveyed prior to removal to ensure that radioactive (i.e., licensed) materials are not inadvertently removed from the facility (see Section 3.1.3). When it is impractical or not possible to decontaminate an item such that it exhibits no discernable facility-related activity when surveyed following methods presented in Section 3.1.3, the item will be treated as radioactive waste. Items that exhibit no discernible facility-related activity will be disposed of as uncontaminated waste.

The systematic approach to the WPI LCWNRF Reactor Decommissioning Project (WPI-RDP) will ensure that every item or structural component in the LCWNRF is specifically evaluated for release before beginning the FSS. The FSS will group the LCWNRF into three classes (as suggested in MARSSIM) to ensure adequate survey coverage in support of a license termination request and subsequent release of the property for unrestricted use.

4.1.3 Means for Ensuring that Sufficient Data is Included to Achieve Statistical Goals WPI will develop the WPI-RDP FSS Plan using the guidance presented in NUREG-1757, Volume 2, Rev. 1, Consolidated Decommissioning Guidance, Characterization, Survey, and Determination of Radiological Criteria (Ref. 29) and NUREG-1575, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM, Ref. 11).

Using such regulatory guidance will ensure that NRC-recommended statistical goals are satisfied.

4.2 BACKGROUND

SURVEY RESULTS The FSS Guideline values for residual activity are taken to be levels above the naturally occurring background radiation.

However, if the FSS results are significantly below the release guideline levels, the regulations allow the licensee to opt not to use background subtraction. WPI has opted to not subtract background for direct surface contamination measurements due to naturally occurring radioactive materials that may be present in construction materials. However, WPI will account for background when gamma exposure rate measurements are made.

4.3 FINAL RELEASE CRITERIA -

RESIDUAL RADIATION AND CONTAMINATION LEVELS The criteria for release of the LCWNRF for unrestricted use, after completion of the decommissioning activities described in this DP, are presented in Section 2.7. In summary, they are:

a.

The generic screening values provided in Tables H.1 and H.2 of Volume 2 in NUREG 1757 (Ref. 29), and

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 4, Page 6 of 11

b.

A limit specified in 10 CFR 20.1402, Radiological Criteria for Unrestricted Use (Ref. 12) 4.4 MEASUREMENTS FOR DEMONSTRATING COMPLIANCE WITH RELEASE CRITERIA 4.4.1 Instrumentation - Type, Specifications and Operating Conditions Instrumentation utilized during the Final Release Survey (and equipment and materials survey) will be selected based upon the need to ensure that site residual radiation will not exceed the release criteria. In order to achieve this goal, WPI will select instrumentation that is sensitive to the isotopes of concern and that is capable *of measuring levels below the guideline values for those isotopes.

Instrumentation available for the FSS and its respective detection range capability is presented in Table 3.1 of this DP.

Instrumentation sensitivities will be determined following the guidance of NUREG-1507, Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions (Ref. 30) using nominal literature values for background, response and site conditions.

Refinements to these detection sensitivity estimates will be made, as necessary, on the basis of actual instrument response and background data gathered during site survey activities.

Instrumentation used in the surveys will be calibrated against sources and standards that are traceable to National Institute of Standards and Technology (NIST), and representative of the isotopes encountered at the LCWNRF. When used, instruments will be operationally tested daily, or prior to each use, whichever is less frequent.

Instruments. will not be used in conditions that are not in conformance with manufacturer recommendations.

4.4.2 Measurement Methodology for Conduct of Surveys This DP has been developed under the presumption that the LCWNRF will have been decontaminated to the extent practicable prior to the FSS. The WPI-RDP FSS Plan will include several steps to calculate the number of measurements and samples required, according to MARSSIM guidance, to release the site without restrictions. These steps include:

E Classify survey units 0 Specify the decision error 0 Determine the DCGLs 0 Calculate the relative shift (COMPASS, Ref. 21, may be utilized) 0 Obtain the number of samples per survey unit (COMPASS, Ref. 21, may be utilized) 0 Estimate the sample grid spacing 0 Perform evaluation for small areas with elevated radioactivity

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 4, Page 7 of 11 0 Determine if the number of samples is reasonable Classify Survey Units The impacted areas of the LCWNRF will be divided into three classes of survey units. Class 1 is an area with the highest potential for contamination.

Class 2 is an area that has a low potential for delivering a dose above the release criteria and has little or no potential for containing small areas of elevated activity. Class 3 is an area with the lowest potential for contamination.

The FSS Plan and survey work packages for each survey unit will include a discussion regarding relevant historical, characterization and in-process survey information and evaluations, to support each survey units' classification.

Specify the Decision Error There are two types of decision error (applied here to analytical results): Type I (alpha) and Type II (beta).* A Type I error is described as the probability of determining that a result is above a criterion when it actually is not (false positive).

A Type II error is described as the probability of determining that a result is below a criterion when it actually is above it (false negative). Both types of error will be set at 0.05 (5%).

Determine the DCGL MARSSIM defines the derived concentration guideline level (DCGL) as the radionuclide-specific concentration within a survey unit corresponding to the release criterion. The radionuclides known or potentially existing at the LCWNRF include 60Co and 55Fe, the presumed predominant radionuclides, and possibly traces of 3H, 14C, 54Mn, 63Ni, 152Eu, 154Eu, and 155Eu. The DCGL values are discussed in Section 2.7.

As stated in NUREG-1757 (Ref. 29), "in light of the conservatism in the building surface and surface soil generic screening levels developed by the NRC staff, the staff presumes, absent information to the contrary, that licensees or responsible parties that remediate building surfaces or soil to the generic screening levels do not need to demonstrate that these levels are ALARA."

Gross activity DCGLs and DCGLEMC (elevated measurement comparison DCGLs) will be developed during the planning stage for the-FSS. Such planning will include consideration of "hard-to-detect" radionuclides for surfaces.

The site-specific DCGLs will be calculated based on the relative fraction of each radionuclide in the expected radionuclide mix. Once developed, the values will be included in the FSS plan.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 4, Page 8 of 11 Calculate the Relative Shift The relative shift is defined as A/o where A is the DCGL - LBGR (Lower Bound of the Gray Region) and a is the standard deviation of the contaminant distribution.

In order to calculate the relative shift, the DCGL must be determined and two assumptions must be made to estimate the LBGR and the standard deviation of the measurement distribution. MARSSIM suggests that the LBGR be set at 50% of the DCGL but can be adjusted later to provide a value for the relative shift between the range of 1 to 3. The standard deviation may be calculated from preliminary survey data, prior surveys of similar areas and materials or the standard deviation of a reference background area.

It should be noted that ai represents the standard deviation prior to release after all area decontamination is thought to be complete. If no reference data is available to make a reasonable estimate of the background standard deviation, MARSSIM suggests using 30% of the mean survey unit background. For the LCWNRF, data from the facility characterization or from post-remediation surveys will be used to calculate the standard deviation value for each survey unit. The value for LBGR is input into COMPASS, and the relative shift is automatically calculated and reported.

Obtain the Number of Samples per Survey Unit Once the relative shift is determined the calculated value, A/o, will be used to obtain the minimum number of measurements or samples necessary to reject the Null Hypothesis based upon the initial assumptions and justify that the survey unit meets the requirements for release for unrestricted use.

MARSSIM Table 5-3 contains the number of samples or measurements necessary for the given decision errors (a and B), and the calculated relative shift (A/u), when dealing with non-radionuclide-specific measurements or when the radionuclide is present in the background. The value N/2 from MARSSIM Table 5-3 represents the number of samples or measurements to be collected in each survey unit and the reference background area. MARSSIM Table 5-5 provides the number of measurements of samples for the case in which the radionuclide is not in the background.

Estimate the Sample Grid Spacing The grid spacing for the measurement and samples is estimated in two ways, depending upon the shape of the grid (i.e., either triangular or rectangular grid). If a triangular grid is used, the grid spacing is estimated as follows:

L = (AI0.866N) 1 2 Where:

L = Distance between measurement locations A = Survey unit Area N = Number of measurements

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 4, Page 9. of 11 If a square grid is used, the spacing is estimated as follows:

L = (A/N) 1 2 Perform Evaluation for Small Areas with Elevated Radioactivity After the grid spacing has been calculated, the area between samples can also be calculated. For example, if for a square grid the grid spacing is 10 m, then there could be an undetected pocket (area) of elevated radionuclide concentrations 100 M 2.

Adjustments to the grid spacing (i.e., additional sampling) may be necessary depending on the following three factors:

• The class of the survey unit

• The ability to scan for the radionuclide

" The minimum potential size of the elevated activity that could produce an exposure above the dose or risk criterion Determine if the Number of Samples is Reasonable After the number of samples per unit has been calculated, it will be evaluated to determine if the number is reasonable. It is possible, even if MARSSIM guidance is strictly followed, that there are too few samples to produce the desired level of comfort. The site managers and health physicists will be responsible for evaluating whether the number of samples is reasonable. If it is determined that the number of samples is inadequate or excessive, the data quality objectives will be reevaluated.

4.4.3 Scan Surveys Following remediation and prior to conducting sampling, beta scans for surfaces and structures and gammascans for environs will be performed over 100% of the Class 1 surfaces, 50% of the Class 2 surfaces and 25% of the Class 3 surfaces. A scanning response exceeding an action level (based on Section 6.8.2 of NUREG-1507) will be investigated/sampled/re-surveyed and, if necessary, remediated.

If remediation is performed, scanning shall be repeated to demonstrate effectiveness of the remediation.

4.4.4 Soil Sampling While soil sampling is not anticipated, if soil samples are later determined to be necessary, they will be obtained to a depth of 15 cm; samples will be packaged and uniquely identified in accordance with chain-of-custody and site-specific procedures.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 4, Page 10 of 11 4.4.5 Sample Analysis Samples will be transferred to a radio-analytical laboratory for analyses in accordance with documented laboratory-specific standard methods. In accordance with MARSSIM, analytical techniques will provide a minimum detection level of 50% of the individual radionuclide DCGLw (or DCGLEMC) values for all primary contaminants.

If analyses indicate residual activity exceeding guideline levels, further remediation will be performed; as required, and scans and sampling of the remediated area will be repeated.

4.4.6 Investigation Levels Radiation levels identified by scans that indicate potential residual radioactive contamination above background will be investigated to identify the source, level and extent of such residual activity.

Areas that contain residual radioactivity concentrations of individual radionuclides, or sum-of-ratio concentrations above respective guideline values, will be remediated, reclassified (as necessary) and re-surveyed.

4.5 METHODS TO BE EMPLOYED FOR REVIEWING, ANALYZING, AND AUDITING DATA 4.5.1 Laboratory/Radiological Measurements Quality Assurance During decommissioning survey activities, many direct and indirect measurements and sample media samples will be collected, measured and analyzed for radiological contaminants.

The results of these surveys will be utilized to evaluate the suitability of the material or item for release to unrestricted use, or whether decontamination of structures, components, and the surrounding site have achieved the desired result. Sample collection, analysis, and the associated documentation will adhere to written procedures and meet NRC guidance, as well as comply with recognized industry recommendations and good practices. Laboratories selected to analyze decommissioning samples will be approved by WPI and listed on the contractor's QA Approved Suppliers List.

The selected contractor will need to implement a QA program that meets the requirements under 10 CFR 71, Packaging and Transportation of Radioactive Material, Subpart H, Quality Assurance (Ref. 9). In addition, the contractor's QA program must meet the applicable criteria from 10 CFR 50, Appendix B (Ref. 8) and ASME NQA-1 (Ref. 26). One of the applicable criteria that must be included is a QA Approved Suppliers List; the contractor will be required to maintain the QA Approved Suppliers List.. WPI will assess the effectiveness of the QA program either through direct audits performed by WPI's Reactor Director, RSO, or Decommissioning Consultant, or by the acceptance of audits performed by other organizations.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 4, Page 11 of 11 Organizations that perform radiological monitoring measurements recognize the need to establish quality assurance programs to assure that radiological monitoring measurements are valid. Such programs are established to:

(1) Readily identify deficiencies in the sampling and measurement processes to those individuals responsible for these activities so that prompt corrective action can be taken, and (2) Routinely monitor the survey and laboratory measurement results in order to assure that results and conclusions are valid 4.5.2 Supervisory and Management Review of Results Health Physics technicians who are trained and qualified will conduct radiological surveys. In addition, senior-level Health Physicists other than the individual that performed the survey will review radiological surveys and sample results. FSS data will also be independently reviewed by the Decommissioning Consultant.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 5, Page 1 of 1 5.0 TECHNICAL SPECIFICATIONS In August 2008, NRC granted WPI a license amendment to remove its authorization to operate the reactor (Ref. 4). That amendment authorizes WPI's possession only of the reactor and used fuel, and amended the Technical Specifications to remove operational requirements not needed for a Possession-Only License (POL).

After the nuclear fuel is removed from the reactor and shipped off-site and NRC issues the Decommissioning

Order, WPI can commence decommissioning operations. At that time, many of the technical specifications for the POL will no.

longer apply.

Proposed revisions to the Technical Specifications for the decommissioning period are given in Appendix B.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 6, Page 1 of 1 6.0 PHYSICAL SECURITY PLAN The existing physical security plan that is approved by NRC will continue to be implemented during the decommissioning period.

That existing plan meets the requirements in NUREG-1537, Chapter 17 (Ref. 5). All radiologically controlled areas are secured from unauthorized entry.

The LCWNRF is normally locked during non-working hours, if unattended. WPI maintains routine, periodic police surveillance of the reactor site.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 7, Page 1 of 1 7.0 EMERGENCY PLAN As required by NRC, WPI has a Reactor Facility Emergency Plan for responding to emergencies at the reactor facility.

The purpose of the Emergency Plan is to minimize the effect of any emergency on the public, personnel, reactor facility and the environment surrounding the facility. Removal of spent fuel from the site will significantly reduce the potential for significant release of radioactive material off site. Any airborne or liquid releases due to decommissioning activities would have negligible impact off-site.

The most likely accident scenario is a contaminated and/or injured individual. This scenario is adequately addressed by the existing emergency plan.

Training will be provided to key personnel to ensure their familiarity with the Emergency Plan and their expected responses.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 8, Page 1 of 5 8.0 ENVIRONMENTAL REPORT This section presents a summary of information relevant to the Environmental Review process, as described in NUREG-1748 (Ref. 32). Based upon the work scope and approach described in this decommissioning plan, the potential for observing negative impacts to the environment during the decommissioning of the WPI research reactor are small or not applicable. As such, the information presented herein has been intentionally limited, in scope, to that necessary for demonstrating the lack of significant environmental impact. In that context, impacts are defined as being effects upon the following environmental resources: land use, aquatic ecology, terrestrial ecology, water use and quality, air quality, waste, human health, socioeconomic, transportation, aesthetics, historical and archeological resources, and environmental justice.

Due to several factors, many of the potential environmental impacts that would normally be associated with a decommissioning project are not be applicable to the WPI decommissioning program. Such factors include: the small size of the facility, the limited scope of the planned D&D work, the short duration of the proposed work, and the small radiological inventory within this facility.

Decommissioning work at the WPI reactor facility will be limited to the reactor's structures; systems and components associated with the reactor's biological shield /

pool, that are housed within a classroom area with a 1,200 ft 2 footprint. No changes will be made to the building's structure and no soil or ground water contamination is present. Decommissioning will principally entail disassembly of small metallic components, with a very limited amount of destructive removal work. As such, there will be no disruption or contact with the outside environment. This decommissioning work is not expected to cause any perceivable negative environmental impacts. The rational for this conclusion is based on the fact that the decommissioning work scope is small and of limited extent, has very low

..inherent radiological hazards due to a small inventory of radioactive materials, and will produce a small amount of radioactive waste materials. These statements are supported by the following determinations obtained during preparation of the Decommissioning Plan:

Nature and Extent of the Decommissioning Work Scope

" The scope of the decontamination and dismantling work is limited to an 8-foot by 8-foot by 15-foot deep reactor pool, within a five-foot thick concrete biological shield.

" Radioactive contamination within the facility is low and limited to the interior. of the reactor pool and a small demineralizer system.

" No decontamination is expected to be required for the building structures.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 8, Page 2 of 5

" There will be no radiological work external to the room which houses the reactor and there will be no outdoor radiological work.

" Soil and ground water contamination is not present at the facility.

" The majority of the D&D work will entail disassembly of small, slightly activated or contaminated components by non-destructive method (e.g.,

unbolting or unthreading) and simple saw cutting of slightly contaminated metals. No thermal cutting or otherwise airborne-producing cutting techniques will be used or permitted.

" No structural changes will be made to the building that houses the reactor room.

" During the course of the D&D work, no perceivable impacts (e.g., noise or staging of materials and equipment) external to the building are expected.

" Radiological D&D work is estimated to require less than 160 work-days of effort.

" Decommissioning work can be accomplished by a very small, hands-on work crew, generally 3 or 4 individuals, with an estimated 6,800 person-hours of work; plus 4,100 person-hours of associated technical and management support.

Facility Radiological Conditions

• The reactor was of very low power, with limited use (approximately 40 MW-hours over 40 years) which resulted in the production of very little residual radioactivity by neutron activation.

• The extent of radioactivity within the facility is limited, with the majority of the radioactivity contained within a single, activated stainless steel regulating control blade having a mass of about 7,000 grams, with an estimated 56 mCi of radioactivity (15 mCi of Co-60). Radiation levels from the blade are estimated to be 17 mrem/hour at one meter (air), based on a measured 125 mrem/hr at 13 cm (through water).

" All other radioactive reactor core components and experimental facilities are estimated to contain significantly lower levels of radioactivity.

" No significant activation of the concrete biological shield is expected.

" Worker radiation exposure is estimated to total less than 0.5 person-rem.

" Generation of airborne radioactivity is not expected nor will it be permitted, and no air effluents containing radioactivity are anticipated.

Worcester Polytechnic Institute Reactor Decommissioning Plan Document W19-1579-003, Rev. 0 Section 8, Page 3 of 5 Radioactive Waste Generation and Disposal

" Less than 600 cubic feet of solid radioactive waste are estimated to be produced, containing less than 60 mCi of radioactivity. The majority of this radioactivity (56 mCi) is contained in a single 7 Kg stainless steel control blade, comprised of activated metals (Mn-54, <1%; Co-60, 27%; Fe-55, 68%,

Ni-59, <1%; and Ni-63, 4.2%).

" All radioactive wastes are expected to be 10 CFR 61 Class A wastes.

Based upon the preceding information, WPI has assessed the potential for environmental impacts on environmental resources. The results of that assessment are presented in the Table 8-1. The results of this assessment use following definitions as a measure of environmental impact:

NONE-Not applicable, environmental effects on the resource are not possible or credible given the site specific conditions.

SMALL-Environmental effects are not detectable or are so minor that they will neither destabilize important attributes of the resource.

MODERATE-Environmental effects are sufficient to alter noticeably, but not destabilize important attributes of the resource.

LARGE-Environmental effects are clearly noticeable and are sufficient to destabilize important attributes of the resource.

TABLE 8.1

SUMMARY

OF ENVIRONMENTAL IMPACT ASSESSMENT E

IONMENTA

' IMUPACT JUSTIFICATION Land Use None No outdoor work will be performed. Current use of the property - a university campus - will not change as a result of the decommissioning.-

Ecology: Aquatic None No outdoor work will be performed. No contact will be made with the aquatic environment.

Ecology: Terrestrial None No outdoor work will be performed. No contact will be made With the terrestrial environment.

(Note: Table 8.1 is continued on next page).

Worcester Polytechnic Institute Reactor Decommissioning Plan Document W19-1579-003, Rev. 0 Section 8, Page 4 of 5 TABLE 8.1 (continued)

SUMMARY

OF ENVIRONMENTAL IMPACT ASSESSMENT ENVIRONMENTAL IMPACT JU*S FICATION

-*... RESO.URCE :'

i Water Use and Quality Small The decommissioning will require disposal of approximately 7,000 gallons of pool water via the public sewer system. Historically the pool water has not contained any detectable radioactivity. As a precaution, pool water will be filtered, sampled and analyzed for radioactive content, and batch released to assure that 10 CFR 20 effluent limits are not exceeded.

Air Quality None Radioactive contamination levels are very low within the facility, with the majority contained in solid, activated metallic components that can be dismantled without the need for segmentation. Dismantling techniques that could produce dust, mists or fumes will be prohibited. Air quality within the reactor classroom will be monitored and sampled for airborne radioactivity to assure and verify that no emissions of radioactivity to the environment are possible.

Use of internal combustion powered equipment, such as compressors or generators, that would generate emissions are not planned.

Waste Small Less than 600 cubic feet of radioactive waste is estimated to be generated. No hazardous or mixed wastes are expected to be generated. All waste materials are estimated to be a small fraction of 10 CFR 60 Class A upper limits. Wastes are planned to be disposed of at the EnergySolutions, LLC facility in Clive, Utah.

Human Health Small Occupational radiation exposure to five decommissioning workers is estimated at less than 0.5 person-rem: to be incurred over 6,800 person-hours.

No radiation exposure to the general public is expected via direct radiation or air and liquid effluents.

There is a small risk of worker injury during performance of the decommissioning work, similar to that of construction work. The probability of an injury leading to a fatality is very small. Based upon the frequency of fatalities for construction workers, 10.3 per 100,000 workers per year, as reported by the Bureau of Labor Statistics for 2007 (Ref. 33), the probability for a fatality for this decommissioning project, with 6,800 person-hours of construction type work, is estimated to be 0.00035.

(Note: Table 8.1 is continued on next page)

Worcester Polytechnic Institute Reactor Decommissioning Plan Document W19-1579-003, Rev. 0 Section 8, Page 5 of 5 TABLE 8.1 (continued)

SUMMDARY OF ENVIRONMENTAL IMPACT ASSESSMENT ENVIRONMENTAL"I'.

JSIICTO RESOURCE AU S*IFI...I.

Socioeconomics Small No workers will be displaced by shutdown and decommissioning of the reactor. The reactor facility has a minimal part time staff (an RSO andan SRO),

who have other duties at WPI. Decommissioning of the reactor will likely have a positive socioeconomic impact, in that these staff will be available for increased educational activities.

Additionally, a small and temporary positive economic impact may be provided to the local area through employment of decommissioning workers, purchase of materials, supplies and equipment, and spending on living expenses at hotels and restaurants.

Transportation Small Only one truck shipment is anticipated: to transport the radioactive waste to the disposal site.

Aesthetics None All decommissioning work will be limited to the interior of the reactor classroom. There will be no outdoor work performed. No outdoor storage of material or equipment is anticipated. No excessively noise producing dismantling techniques is anticipated.

Historical and None No structural or appearance changes will be made to Archeological the building. Decommissioning work is limited to one classroom within the building.

Environmental Justice None The decommissioning work will not involve contact with, or disruption of, any surrounding neighborhoods.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 9, Page 1 of1 9.0 CHANGES TO TIE DECOMMISSIONING PLAN WPI requests that changes to the Decommissioning Plan (DP) be allowed with local approval by the Reactor Director and RHSC, and without prior NRC approval, unless an un-reviewed safety question is involved. An un-reviewed safety question involves:

1.

The increase in probability of occurrence or the increase of consequences of an accident or malfunction of equipment important to safety compared to that situation previously evaluated in the original Safety Analysis Report (SAR) for the Leslie C. Wilbur Nuclear Reactor Facility, or

2.

The possibility for a type of accident or malfunction differing from those previously analyzed in the SAR, or

3.

The reduction in margin of safety as defined in the SAR.

Reports and records of changes to the DP, and retention of documents, will be in accordance with the applicable portions of 10 CFR 50.59 (Ref. 31).

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 10, Page I of 3

10.0 REFERENCES

1.

WPI discontinued routine operation of the reactor on June 30, 2007.

Certification letter sent to the NRC on 6-27-07 (copied to Alexander Adams) signed by David Adams, Mike Curley, and Jeffrey Solomon.

2.

WPI submitted an application to NRC for a possession-only license.

Certification letter sent to the NRC on 8-13-07 (copied to Alexander Adams) signed by David Adams, Mike Curley, and Jeffrey Solomon.

3.

WPI submitted additional correspondence to NRC in support of an application for a possession-only license. Certification letters sent to the NRC on 11-19-07, 4-10-08, and 5-5-08 (copied to Alexander Adams) signed by David Adams, Mike Curley, and Jeffrey Solomon.

4.

NRC Granted WPI a Possession-Only License Amendment No. 11 to R-61.

Received at WPI 9-2-08.

5.

Nuclear Regulatory Commission (U.S.), "Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors, NUREG-1537 Rev. 0, February 1996.

6.

Nuclear Regulatory Commission (U.S.), "NMSS Decommissioning Standard Review Plan," NUREG-1727, September 2000.

7.

WPI requests DOE take possession of the fuel Email from David Adams to Douglas K. Morrell on 6-25-07; verification email from Douglas K. Morrell to David Adams 6-25-07.

8.

10 CFR 50, Appendix B, "Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," Code of Federal Regulations.

9.

10 CFR 71, Subpart H, "Packaging and Transportation of Radioactive Material - Quality Assurance," Code of Federal Regulations.

10.

ANSI/ASQC E4-1994, "Specifications and Guidelines for Quality Systems for Environmental Data Collection and Environmental Technology Programs,"

American Society for Quality Control, 1995.

11.

Nuclear Regulatory Commission (U.S.), "Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM)," NUREG-1575, Rev. 1, August 2000.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 10, Page 2 of 3

12.

10 CFR 20.1402, "Radiological Criteria for Unrestricted Use," Code of Federal Regulations.

13.

B. Shleien (editor), "The Health Physics and Radiological Health Handbook, Revised Edition, Scinta, Inc, Silver Spring, MD, 1992.

14.

10 CFR Part 19, "Notices, Instructions and Reports to Workers: Inspection and Investigations," Code of Federal Regulations.

15.

10 CFR 20 Subpart H, "Standards for Protection Against Radiation -

Respiratory Protection and Controls to Restrict Internal Exposure in Restricted Areas," Code of Federal Regulations.

16.

Nuclear Regulatory Commission (U.S.), "Acceptable Programs for Respiratory Protection," Regulatory Guide 8.15, Rev. 1, October 1999.

17.

Nuclear Regulatory Commission (U.S.), "Manual of Respiratory Protection Against Airborne Radioactive Materials," NUREG-0041, Rev. 1, January 2001.

18.

29 CFR 1910.134, Subpart I, "Occupational Safety and Health Standards -

Personal Protection Equipment - Respiratory Protection," Code of Federal Regulations.

19.

10 CFR 20 Subpart E, "Standards for Protection Against Radiation Radiological Criteria for License Termination," Code of Federal Regulations.

20.

Nuclear Regulatory Commission (U.S.),

"Consolidated NMSS Decommissioning

Guidance, Decommissioning Process for Materials Licensees," NUREG-1757, Volume 1, Rev. 2, September 2002.

21.

COMPASS Code Version 1.0.0, Developed by Pacific Northwest National Laboratory, Richland, Washington, under the sponsorship of the U.S. Nuclear Regulatory Commission for implementation of MARSSIM in support of the decommissioning license termination -rule (10 CFR Part 20, Subpart E).

BNWL-1264, March 1970.

22.

10 CFR 20.2003, Subpart K, "Standards for Protection Against Radiation -

Disposal by Release into Sanitary Sewerage," Code of Federal Regulations.

23.

Regulatory Guide 1.86, "Termination of Operating Licenses for Nuclear Reactors," Directorate of Regulatory Standards, U.S. Atomic Energy Commission, Washington, D.C., June 1974

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Section 10, Page 3 of 3

24.

Nuclear Regulatory Commission (U.S.), "Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities," NUREG-0586, Office of Nuclear Regulatory Research, Washington, D.C., 1988

25.

10 CFR Part 61, Subpart D, "Licensing Requirements for Land Disposal of Radioactive Waste - Technical Requirements for Land Disposal Facilities,"

Code of Federal Regulations.

26.

ASME NQA-1-2000, "Quality Assurance Requirements for Nuclear Facility Applications," American Society of Mechanical Engineers, New York, NY, 2001.

27.

49 CFR Parts 100-177, "Research and Special Programs Administration, Department of Transportation," Code of Federal Regulations.

28.

Federal Occupational Safety and Health Acts (OSHA) of 1973, Title 29 United States Code, Chapter 15, "Occupational Safety and Health."

29.

Nuclear Regulatory Commission (U.S.),

"Consolidated NMSS Decommissioning Guidance - Characterization, Survey, and Determination of Radiological Criteria," NUREG-1757, Volume 2, Rev. 1, 2003.

30.

Nuclear Regulatory Commission (U.S.),

"Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions,"

NUREG/CR-1507, Final, 1997, Washington, D.C.

31.

Regulatory Guide 1.187, "Guidance for Implementation of 10 CFR 50.59, Domestic Licensing of Production and Utilization Facilities - Changes, Tests and Experiments," U.S. Nuclear Regulatory Commission, Washington, D.C.,

November 2000.

32.

NUREG-1748, "Environmental Review Guidance for Licensing Actions Associated with NMSS Programs," U.S. Nuclear Regulatory Commission, August 2003.

33.

"National Census of Fatal Occupational Injuries in 2007," Bureau of Labor Statistics, United States Department of Labor.

Worcester Polytechnic Institute Reactor Decommissioning Plan Document W19-1579-003, Rev. 0 Appendix A, Page 1 of 2 APPENDIX A LETTER OF FINANCIAL COMMITMENT

Worcester Polytechnic Institute Reactor Decommissioning Plan Office of dhe Eectie Vice President Document W19-1579-003, Rev. 0 Appendix A, Page 2 of 2 100 Institute Road Worcester, MA 01609-2280, USA 508-831-5288, Fax 508-831-5064 www.wpi.edu March 25, 2009 U. S. Nuclear Regulatory Commission, 1, Jeffrey S. Solomon, Executive Vice President and Treasurer of Worcester Polytechnic Institute (WPI),

which is located at 100 Institute Road, Worcester, MA 01609-2280. This letter is being provided to assure WPI's financial commitment to fund the radiological decommissioning of the LeslieC. Wilbur Nuclear Reactor Facility that is described in the Decommissioning Plan to which this letter is attached.

As documented in WPI's Decommissioning Plan, the radiological decommissioning cost, including a 25%

contingency, is estimated to be $2,600,000. These funds are available and will be obtained from WPI's unrestricted capital project funds. As Executive VicePresident and Treasurer of WPI, I commit to spending these funds to accomplish the decommissioning.

As Executive Vice President and Treasurer of WPI, I have the authority to commit and spend these funds for the decommissioning of the Leslie C. Wilbur Nuclear Reactor Facility. I certify the statements made in this letter are true and made freely under penalty of perjury.

Sinprely Jeffrey Solomon Executive Vice President and Treasurer Worcester Polytechnic institute

Worcester Polytechnic Institute Reactor Decommissioning Plan Document W19-1579-003, Rev. 0 Appendix B, Page 1 of 14 APPENDIX B REVISED TECHNICAL SPECIFICATIONS FOR DECOMMISSIONING (Edited from the 9-2-08 Possession Only License Amendment)

Worcester Polytechnic Institute Reactor Decommissioning Plan Document W19-1579-003, Rev. 0 Appendix B, Page 2 of 14 APPENDIX A TO LICENSE NO. R-61 TECHNICAL SPECIFICATIONS FOR THE WORCESTER POLYTECHNIC INSTITUTE REACTOR DOCKET NO. 50-134

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Appendix B, Page 3 of 14 TABLE OF CONTENTS Page 1.0 D EFIN ITION S.......................................................................... I 2.0 SAFETY LIMITS AND OPERATING RESTRICTIONS................... 3 2.1 Safety Lim its........................................................................... 3 3.0 SURVEILLANCE REQUIREMENTS............................................. 4 3.1 Frequency of Surveillance..................................

4 3.2 Action to be Taken...................................................................

4 3.3 R adiation D etection...................................................................... 4 4.0 SITE AND DESIGN FEATURES...............................................

5 4.1 Site...................................................

5 4.2 Restricted Area and Exclusion Area................................................. 5 4.3 Reactor Building and Ventilation System....................................

5 4.4 Reactor Core............................................

5

5. ADMINISTRATIVE AND PROCEDURAL REQUIREMENTS.............. 6 5.1 Facility Administrator....................................

6 5.2 Radiation, Health, and Safeguards Committee..................

6 5.3 Radiological Safety Offi cer............................................................. 6 5.4 Fire Protection..........................................

6 5.5 Procedures.........

........ 6 5.6 Operating Records........

............................... 7 5.7 Reports...............................................

7 5.8Annual Operating Reports............................................... 8 5.9 F uel Storage.................................

...............................................8

Worcester Polytechnic Institute Reactor Decommissioning Plan Document W19-1579-003, Rev. 0 Appendix B, Page 4 of 14 The actual values of dimensions, measurements, and other numerical values may differ from values given in these specifications to the extent of normal construction and manufacturing tolerances, or normal accuracy of instrumentation.

1.0 DEFINITIONS Cold. Clean, Critical Condition: Since xenon and samarium effects arc negligible thisA re.actor-in its normnal operations, th-e termfi comld, clcan, cr-itical- -sh-all referf to the conditionm of t~he recor cor when-R it iss at the norm-al amnb-i-ent water-temperature of 7-0 to !79'F and free of any exper-iments th-at coeuldd affect r-eactivity.

Cr-itical Reactor Oper-ati-on: Cr4itical r-eactor oper-atiomn-shaIAll -e-fer to any situation wheLPn mPore than 12 fu-el-elele-men-ts-arie lo-aded in the core and any contr-ol blade is withdrawn more than 6 in.

Exncerimcnt: An exncr-iment shall mean any ann~aratus. device. or material installed in the cor or ineternal exper-imental facilities thiat is not a normal part of those faefihtiPS~

Movable Exp~er-iment:. A movsxable exEper-iment is one that may be in-aserfted, r-emoved, or manipuated while the reactor is critical.

Operable: An instrument or channel is operable when the instrument or channel will be operational once it is energized.

Operational: An instrument or channel is operational when that instrument or channel is installed, energized, and in all other respects performing the monitoring and safety functions for which it was intended.

Reacetor Operation.:.Reactor-operation shall be any codiio wherein either-the Reactor Safety System: The reactor safety system is that combination of con~trol c-1h-aDnnels, &and aRssoci4ated-circu04-itry' that forms the automatic protective system for-the iniitiated.

Reactor Scram: Reactor scr-amf shall -be the rapid i-nse-rtion-of the three control blades into the coAre by either-of the following methos

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Appendix B, Page 5 of 14 (1)

Relay (

,,ow),-ram:

Rcactor relay eram (slow

,ram, shall be instigated by the relay. eca ircuits which control cur-rent inputs for-the trip amplifierF.

inter-ruption of this cur-rent shall de energize the ecr-am-magnet&.

(2)

Electr-onic (fiast) secram: Reactor-electr-onic scr-am (fast scr-am) shall be caused by the application of Suifficei-en-t negative bias in the trip amplifierf to terFminate current to the ecr-am magnets-.

Readily Available on Call: Readily available on call shall mean an individual on duty within a reasonable driving time (1/2 hr) from the reactor building, that can be contacted to fulfill duties as specified in the implementing plans and procedures.

Reportable Occurrence: A reportable occurrence is any of the following conditions:

(1) a safety system setting less conservative than the limiting setting established in the Technical Specifications; (2) operation in violation of a limiting condition for operation established in the Technical Specifications; (3) a safety system component malfunction or other component or system malfunction that during operation could, or threatens to, make the safety system incapable of performing its intended safety functions; (4) release of fission produc-tc firomf _a failedl f&uel element; (5) an uncontrolled or unplanned release of radioactive material from the restricted area of the facility; (6) an uncontrolled or unplanned release of radioactive material that results in concentrations of radioactive materials within the restricted area in excess of the limits specified in Appendix B, Table 1 of 10 CFR Part 20; (7) an uncontr-olled or-unanticipated change in reactivity in excccc of 0.5% Ak/k; (8) conditions arising from natural or man-made events that affect or threaten to affect the safe operation of the facility; and (9) an observed inadequacy in the implementation of administrative or procedural controls such that the inadequacy causes or threatens to cause the existence or development of an unsafe condition in connection with the operation of the facility.

Worcester Polytechnic Institute Reactor Decommissioning Plan Document W19-1579-003, Rev. 0 Appendix B, Page 6 of 14

.7~II Ii iZJV~.'hi ~..iiimi: I Thpe-r-eatno nhal1-,be-deeme t

i-4An Qhp-shudWn condition-ifno control regulating *blade is withdrawn from its fully inserted position-Or i-f the4-re arc less than 12 fue1 l-elemfflenits loa -ded-on the gr-id plate.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Appendix B, Page 7 of 14 2.0 SAFETY LIMITS AND OPERATING RESTRICTIONS 2.1 Safety Limits Radiation Alarms: Upon indication of radiation levels in excess of 50 mrems/hr (-20 mr.ems.hr for fuel storag)* area monitors shall actuate audible evacuation alarms in the reactor room and in the second and third floor areas above the reactor pool.

Radiation Levels: The maximum radiation levels 1 m above the pool surface and at the surface of the concrete shield, when the beam port and thermal column are closed, shall be less than 50 mrems/hr.

Water Level: The minimum depth of water -above thne to.p of the e-nd bx* of the fuel elmns in the r-eactor pool sh-all be 10 ft. When-- fuel-i-s not in the pool, 11e-w-e lee t

shP1 1 all be mai:nta-i:n-ed to keep radiation-m leveels a8s low as reasonably achievable (AL=ARA). The pool water-level d-etect-or and alar-m will be operational, with the al-ar-m set-point being a water-drop of less than or equal to 1 foot. The pool water l evel4 detector-and alarm may be by passed provided- -A lic-ensed senior-reactor-operator-is present in the facility-.

the pool water are not met.

(1) pH less than 8.0 and greater than 6.0 (2) resistivity greater-than 5- -x

-.011M I em (3) pool water activity less than 10-5*Ci/n4 Water Temp.eratur-e: The ximum bulk water temperature of the reactor pool s-.h -All1 be 110.F and the minimumfl 1

Shall e 40

.F, when.fuel is in the pool.

Worcester Polytechnic Institute Reactor Decommissioning Plan Document W19-1579-003, Rev. 0 Appendix B, Page 8 of 14 3.0 SURVEILLANCE REQUIREMENTS 3.1 Frequency of Surveillance Quarterly: The area radiation monitoring systems and the pool water-level swith*

shall be checked and ensured to be operational quarterly.

Semiannuially:; A t least semiannually, dur-ing pos-se-ssionIn of the facility, a reacto-r inspection shall be performed consisting of (1) Pool water-p4 sh-all be mie-asurfled-n-d-conBductivAity and pH devicese shall be ealibr-ated-.

3.2 Action to be Taken If maintenance or recalibration is required for any of the items, it shall be performed and the instrument shall be rechecked before being placed back in service.

3.3 Radiation Detection Area Monitors: Area radiation sensors capable of detecting gamma radiation in the range of 0.1 to 100 mrems/hr shall be installed near the beam port, demineralizer, thermal column door, Aw-l storage area, and less than 1 m above the core pool surface, until these items have been removed or decontaminated. Upon indication of radiation levels in excess of 50 mrems/hr (20 mremsnhr for-fuel storage) these monitors shall actuate audible alarms in the reactor room and in the second and third floor areas above the reactor pool. Area monitors shall be calibrated on a semi-annual basis.

Portable area monitors capable of detecting gamma radiation in the range of 0.10 to 50 mrems/hr may temporarily replace fixed area monitors described above provided that the required alarms are operational.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Appendix B, Page 9 of 14 Portable Monitors: During fuel handling or other operations involving or potentially involving sources of radiation, operable portable survey instruments shall be readily available to the reactor operator for measuring beta-gamma exposure rates in the range 1.0 mremsfhr to 50 remslhr, and fast plus thcrm.a! neutron dose rate from.

0.04 to 1,000 mr-,mJ One or more portable survey instruments for measuring beta-gamma exposure rates in the range 10 mrems/hr to 50 rems/hr will be kept available to the reactor staff in an external location (normally the security office) to facilitate obtaining radiation readings if a reactor radiation alarm should be activated.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Appendix B, Page 10 of 14 4.0 SITE AND DESIGN FEATURES 4.1 Site The reactor and associated equipment is housed in the Washburn Laboratories located between West Street and Boynton Street on the campus of Worcester Polytechnic Institute in Worcester, Massachusetts.

4.2 Restricted Area and Exclusion Area The reactor room shall constitute a restricted area as defined in 10 CFR Part 20 and shall be controlled by partitions and normally locked doors. In addition, two small areas, one each on the second and third floors of Washburn Laboratories, directly above the reactor control drives, shall become restricted areas whenever the radiation levels in any of the rooms exceed those specified in 10 CFR 20.1301. The exclusion areas, as defined in 10 CFR Part 100, shall consist of the reactor room and the areas above the reactor.

4.3 Reactor Building and Ventilation System The reactor shall be housed in a closed room that is designed to restrict leakage.

The ventilation system shall provide at least two changes of air per hour in the reactor room whenever the reactor is operating.

4.4 Reactor Core Fuel Elements: Standard fuel elements shall be flat plate type consisiting4o urnimaluminm alloy clad with alu minum. The width and depth of e.AchL fu el ePlem*e PmIt hAAI*

,. be23 i n. x 3 in. EAch e

-em en1t shall have an active length of 24n.

There-tshall be -A mlaximum of 10 g of U 235 in each fuel plate and not more than 170 g of U 235 in an-y fuel elePm-en--t.P-H T-he fuel. sh-al b

.e e

nriched to r

th&an, 0*%* U 2*35.

SSt-Aan da-r-d fuel elements hav&e 18 fuel plates,, each pl-ate 1.52-mmHO thipck-with a clad thiclmess of 0.38 1 m oneah side. No fuael elements may be installed in the core.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Appendix B, Page 11 of 14 5.0 ADMINISTRATIVE AND PROCEDURAL REQUIREMENTS 5.1 Facility Administrator The Director of the Nuclear Reactor Facilities shall have full responsibility for maintaining the facility in a safe configuration. The Director shall report to the Provost and shall be responsible to the Radiation, Health, and Safeguards Committee for conformance to the facility license provisions and all local and NRC safety regulations. The Director also shall be responsible for proper maintenance of such records and operating practices as the Committee may deem necessary for the safe storage of the facility.

5.2 Radiation, Health, and Safeguards Committee A Radiation, Health, and Safeguards Committee (RHSC) shall review, approve and document all proposed modifications affecting reactor safety and procedures, pursuant to 10 CFR 50.59. This committee also shall conduct, at least quarterly, reviews of operations, equipment performance, records, and procedures. The Committee shall establish written procedures regarding review methods, quorums, and subcommittees, and it shall maintain written records of its activities. The members of the Committee shall be appointed by the Provost of Worcester Polytechnic Institute (WPI) and a majority shall be WPI faculty members.

5.3 Radiological Safety Officer A Radiological Safety Officer shall be appointed to serve on the Radiation, Health, and Safeguards Committee and to review and approve all proposed procedures am4

n. ie-""s concerning radiological safety. The Radiological Safety Officer shall advise the Director of the Nuclear Reactor Facilities of rules, regulations, and procedures relating to radiological safety and shall routinely conduct radiation surveys.

5.4 Fire Protection The licensee shall provide heat or ionization-type smoke detectors, which will alarm when there is a fire in the reactor room. At least two such detectors shall be operable at all times.

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Appendix B, Page 12 of 14 5.5 Procedures Detailed written procedures shall be provided for all normal decommissioning operations eofthe -ee...

, operation of supporting facilities, maintenance operations, radiation protection, emper-imenits and emergency plans and operations. These procedures shall be approved by the Radiation, Health, and Safeguards Committee before they are implemented.

Temporary procedures that do not change the intent of the initial approval procedures may be authorized by two members of the facility staff at least one of whom shall be a licensed senior operator. Such procedures shall be subsequently reviewed by the Radiation, Health, and Safeguards Committee, 5.6 Operating Records In addition to records required elsewhere in the license application, the following records shall be generated and kept of:

(1) maximum radioactivity released or discharged into the air or water beyond the effective control of the licensee as measured at or before the point of such release or discharge; (2) maintenancc operations involving sjubstitution or rceplaccment of rcactor-cqimnt or-corponents; and (3) tests and measurements performed pursuant to the Technical Specifications.

Old records pertaining to operation of the reactor, including power levels, emergency shutdowns, inadvertent scrams, experiments, and in core irradiations, shall be kept for purposes of decommissioning.

5.7 Reports In addition to.reports otherwise required under this license and applicable regulations-(1)

The licensee shall inform the Commission of any incident or condition relating to the safe storage of the facility that prevented or could have prevented a safety system from performing its safety function as described in the Technical Specifications. For each such occurrence, WPI shall promptly notify, by telephone or telegraph, the Administrator of the appropriate NRC Regional Office Listed in Appendix U of 10 CFR Part 20 and shall submit within 10 days a report in writing

Worcester Polytechnic Institute Document W19-1579-003, Rev. 0 Reactor Decommissioning Plan Appendix B, Page 13 of 14 to the Director, Division of Waste Management and Environmental Protection (DWME.P), with a copy to the Regional Office.

(2)

The licensee shall report to the Director, DWMEP, in writing within 30 days, any observed occurrence of substantial variance of conditions from performance specifications contained in the Safety Analysis Report or the Technical Specifications.

(3)

The licensee shall report to the Director, DWMEP, in writing within 30 days, any occurrence of significant changes in transient or accident analysis as described in the SAR.

5.8 Annual Operating Reports A report covering the previous year shall be submitted to the Administrator of the appropriate Regional Office not later than March31 of each year. It shall include:

(1)

Operations Summary: a summary of issues having safety significance occurring during the reporting period, including:

(a) changes in facility design (b) performance characteristics (e.g., equipment and fuel performance)

(c) changes in operating procedures that relate to the safety of facility operations (d) any abnormal results of surveillance tests and inspections required by these Technical Specifications (e) a brief summary of those changes, tests, and experiments that did not require authorization from the Commission pursuant to 10 CFR 50.59 (d)(2)

(f) changes in the plant staff serving in the positions of Reactor Facility Director, SRO, RSO, ARSO, or Radiation, Health, and Safety Committee members; (2) Maintenance: a discussion of corrective maintenance (excluding preventative maintenance) performed during the reporting period on safety related systems and components; (3)

Changes, Tests, and Experiments: a brief description and a summary of the safety evaluation for those changes, tests, and experiments that were carried out without prior Commission approval, pursuant to the requirements of 10 CFR 50.59 (d)(2); and

Worcester Polytechnic Institute Reactor Decommissioning Plan Document W19-1579-003, Rev. 0 Appendix B, Page 14 of 14 (4)

Radioactive Effluents Releases: a statement of the quantities of radioactive effluents released from the plant.

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rj-s-hall be designed to contain not more th-an 1 fue eleme-&n-ts. IV/hen-the reacetor-contains a cr-itical mass, all addiIonl fuel elements not in the cor-e shall be locke in place except as aultho&rized-by the lice-nsed-senioroerato in charge.

A fuel element shall not be stor-ed outside of the r-eactor-pool unless it produces radiation dose levels of less than 100 M--remsiAhrF -At. the storage container-surface..

Storage container-s of fuiel elements shall be locked closed when unattended All fuel element transfers sh-all1 be coqnducted by a staff of not less than the persnswhich shall include a licen.-4sed senior operator-in char-ge and a RSO. Staf me.mbersA willcontinuRously monitor-the operations using appropr-iate radiation monitoring and cor-e nuclear instruamentation.