ML13046A107

W19-1579-004, Rev 0, Final Status Survey Plan for the Leslie C. Wilbur Nuclear Reactor Facility at the Worcester Polytechnic Institute

ML13046A107
Person / Time
Site:	05000134
Issue date:	01/31/2013
From:	Darrell Adams TLG Services
To:	NRC/FSME, Worcester Polytechnic Institute
References
W19-1579-004, Rev 0
Download: ML13046A107 (56)

Text

Document W19-1579-004, Rev. 0 FINAL STATUS SURVEY PLAN for the LESLIE C. WILBUR NUCLEAR REACTOR FACILITY at the WORCESTER POLYTECHNIC INSTITUTE Operating License No. R-61 KDocket No. 50-134 I

I! ~

~ ~JI k

1/2-~

~-~JN X

A z<

L Prepared for:

WORCESTER POLYTECHNIC INSTITUTE Prepared by:

TLG Services, Inc.

Bridgewater, Connecticut January 2013

Worcester Polytechnic Institute Reactor Decommissioning - Final Status Survey Plan Document W19-1579-004, Rev. 0 Page ii of v APPROVALS S~\\

Radiation Safety Officer Reactor Facility Director Date David Adams Radiation, Health

& Safety Committee David Messier Date

Worcester Polytechnic Institute Document W19-1579-004, Rev. 0 Reactor Decommissioning - Final Status Survey Plan Page iii of v TABLE OF CONTENTS SECTION-PAGE 1.0 IN TR O D U CTIO N..........................................................................................

1-1 2.0 FACILITY DESCRIPTION RELEVANT TO FSS....................................

2-1 3.0 REMEDIATION WORK AND RADIOLOGICAL CONDITIONS E N CO U N TE R ED...........................................................................................

3-1 4.0 RADIOLOGICAL CONTAMINANTS AND CRITERIA...........................

4-1 4.1 Radioactive Contaminants Identified During Decommissioning........ 4-1 4.2 Characteristics of the Radioactive Contaminates That Potentially Could be Present D uring FSS...............................................................

4-4 4.3 F S S C riteria...........................................................................................

4 -6 5.0 QUALITY ASSURANCE PROGRAM.........................................................

5-1 6.0 FINAL STATUS SURVEY APPROACH....................................................

6-1 6.1 Classification by Contamination Potential..........................................

6-1 6.2 Identification of Survey U nits..............................................................

6-2 6.3 Testing to Demonstrate Compliance....................................................

6-8 6.4 Survey D ata Requirem ents...................................................................

6-8 6.5 Survey L ocations...................................................................................

6-9 6.6 Survey D esign Packages.......................................................................

6-9 6.7 Survey Instrum entation........................................................................

6-9 6.8 Background and Reference Area Measurements...............................

6-11 6.9 Survey Reference System s..................................................................

6-11 6.10 Survey T echniques..............................................................................

6-12 6.10.1 B eta Surface Scans...............................................................

6-12 6.10.2 G am m a Surface Scans..........................................................

6-12 6.10.3 Surface Activity Measurements...........................................

6-13 6.10.4 Removable Activity Measurements.....................................

6-13 6.10.5 Soil Sam plin g........................................................................

6-13 6.10.6 Structural M edia Sam pling.................................................

6-14 6.10.7 E m bedded Pipes...................................................................

6-14

Worcester Polytechnic Institute Document W19-1579-004, Rev. 0 Reactor Decommissioning - Final Status Survey Plan Page iv of v TABLE OF CONTENTS (Continued)

SECTION-PAGE 7.0 DATA EVALUATION AND INTERPRETATION.....................................

7-1 7.1 S am ple A n alysis....................................................................................

7-1 7.2 D ata C onversion....................................................................................

7-1 7.3 D ata A ssessm ent...................................................................................

7-1 7.4 Determining Compliance with Guidelines...........................................

7-1 8.0 FINAL STATUS SURVEY REPORT..........................................................

8-1 9.0 BIBLIO G RA PH Y...........................................................................................

9-1 TABLES 2.1 Exterior View of Washburn Shops and Stoddard Laboratory Building........................................................................

2-2 3.1 Radiological Conditions Encountered During Dismantling...........................

3-5 4.1 Summary of Waste Stream and Pre-FSS Sample C haracterization R esults.................................................................................

4-2 4.2 Summary of Analytical Results for Pre-FSS Samples from the Biological Shield..........

............................................... 4-2 4.3 Summary of the Radionuclides Mixtures Representative of R esidual R adioactivity.................................................................................

4-5 4.4 License Termination Screening Values for Building Surface Contam ination...............................................................

4-6 4.5 License Termination Screening Values for S u rface S oil.................................................................................................

4 -7 6.1 Survey U nit A reas...........................................................................................

6-2 6.2 FSS Survey U nits and Classifications............................................................

6-3 6.3 Instrumentation For WPI Final Status Survey............................................

6-10

Worcester Polytechnic Institute Document W19-1579-004, Rev. 0 Reactor Decommissioning - Final Status Survey Plan Page v of v TABLE OF CONTENTS (Continued)

SECTION-PAGE FIGURES 2.1 Exterior View of Washburn Shops and Stoddard Laboratory B uilding........................................................................

2-2 2.2 Ground Floor WPI Reactor Facility Layout -

M ajor Features Relevant to FSS.....................................................................

2-3 2.3 Reactor Facility First Floor Layout -

M ajor Features Relevant to FSS.....................................................................

2-4 2.4 Artist Rendering of Biological Shield / Reactor Pool Structure (Pre-D & D ).................................................................................

2-5 3.1 Locations of Rem oved Radioactive Item s........................................................

3-3 3.2 Post Removal View Of Former Reactor Pool W ater Treatm ent System Area......................................................................

3-4 3.3 Mid-Remediation View of Reactor Pool Floor (After Liner Removal - Before Concrete Removal)........................................

3-6 3.4 Post Remediation View of Remaining Portion of Thermal Column Liner.... 3-7 3.5 Post Remediation View of Beam Port Tube / Shutter Housing Area............ 3-8 3.6 Post Remediation View of Reactor Pool Floor Area........................................

3-9 3.7 Post Remediation View of Remaining Beam Port Vent Stub, Water Treatment Return Line and Scupper Drain Outlet Pipe.................. 3-10 3.8 Post Remediation View of Remaining Thermal Column Vent an d D rain S tu b s.............................................................................................

3-11 3.9a Post Remediation View of Remaining Ground Floor E xhaust D uct and Plenum............................................................................

3-12 3.9b View of First Floor Exhaust Duct Plenum....................................................

3-13 3.10 Post Remediation Configuration of Biological Shield..................................

3-14 4.1 Location of Waste Stream Characterization and Pre-FSS Samples.............. 4-3 APPENDICES A

GEL Laboratory Analysis Report For W aste Stream and Pre-FSS Samples...........................................................

A-1 B

Gross Beta DCGL for Radionuclide Mixture at WPI................................. B-1

Worcester Polytechnic Institute Document W19-.1579-004, Rev. 0 Reactor Decommissioning - Final Status Survey Plan Section 1, Page I of 1

1.0 INTRODUCTION

The Leslie C. Wilbur Nuclear Reactor Facility (the "Reactor") at the Worcester Polytechnic Institute (WPI) was a light-water cooled and moderated open-pool design reactor. The Reactor was licensed (License R-61, Docket 50-134) by the U.S. Nuclear Regulatory Commission (NRC) to operate at a power level of 10 Kilowatt (KW) thermal.. The Reactor began operation in 1959 and 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. WPI discontinued routine operation of the Reactor on June 2007, and submitted an application to the NRC for a Possession-only license, which was granted on August 26, 2008. A Decommissioning Plan (DP) was prepared and submitted to the NRC for approval:

approval to decommission was granted in December 2010. Fuel was removed from the facility in July 2011, which allowed initiation of decommissioning activities. During the Fall of 2011, the Reactor pool was drained, the Reactor systems de-energized and the facility cleared of furniture and classroom equipment. Preparations were then made for decommissioning the Reactor facility, and dismantling commenced in July of 2012.

Dismantling work was completed in October of 2012: off-site shipment of the wastes is scheduled to occur in 2013.

The planned decommissioning activities are described in Decommissioning Plan for the Leslie C. Wilbur Nuclear Reactor Facility at the Worcester Polytechnic Institute, Document W19-1579-003, Revision 1, September 2009. The objective of the decommissioning is to remove radiological materials and equipment associated with WPI's licensed operations, such that radiological conditions satisfy NRC criteria for unrestricted use of the facility and thus permit termination of the NRC license. A Final Status Survey (FSS) will be performed to demonstrate that these NRC criteria have been satisfied. This document describes the FSS Plan.

This FSS plan fulfills WPI's commitment to develop and submit a FSS Plan, as described in "Proposed Final Status Survey Plan" (Chapter 4.0) of the September 2009 Decommissioning Plan. This FSS Plan was prepared in accordance with the guidelines and recommendation presented in NUREG-1757, Consolidated NMSS Decommissioning Guidance, and NUREG-1575, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM). The process emphasizes the use of Data Quality Objectives (DQOs) and Data Quality Assessment (DQA), along with a quality assurance/quality control program. The graded approach concept will be followed to assure that survey efforts are maximized in those areas having the greatest potential for residual contamination or the highest potential for adverse impacts of residual contamination.

This Plan incorporates project-specific information relative to post-remediation facility conditions and known characteristics of potential residual radiological contaminants, guidelines for residual building surface and soil contamination levels, sampling and measurement methods, survey unit identification and classification, and data evaluation techniques. A Quality Assurance Project Plan (QAPP), applicable to the FSS Activities, will be utilized.

Worcester Polytechnic Institute Document W19-1579-004, Rev. 0 Reactor Decommissioning - Final Status Survey Plan Section 2, Page 1 of 5 2.0 FACILITY DESCRIPTION RELEVANT TO`FSS The Reactor is located is housed within a portion of the Washburn Shops and Stoddard Laboratories Building, located on the WPI campus, between West and Boynton Streets.

Figure 2.1 provides an exterior photograph of the Washburn Shops and Stoddard Laboratories Building. It is a four-story brick building, constructed in the 1860's. The reactor room is located within a portion of the lower two levels of the building. The reactor room's perimeter walls are generally constructed of concrete block, with interior partition walls constructed of wood framing and sheet rock. The footprint of the reactor room is approximately 90 feet (27.4m) long and 21 feet (6.4m) wide; the ground floor has a ceiling height of approximately 10 feet (3m), and the first floor (located above the ground floor) has a ceiling height of approximately 12 feet (12.7m). The first floor of the facility includes a 9 foot by 21 foot office room, and a small (12 foot by 8 foot) reactor tool closet. The remainder of the room is an open area, housing class instruction areas and the upper portion of the reactor's biological shield (situated in the center of the room). The ground floor facility includes two partitioned areas - a 6 foot by 8 foot dark room and a 6 foot by 16 foot radioactive material storage area - and the lower portion of the reactor's biological shield. The ground level floor is constructed of poured concrete, covered with vinyl tile, and the first floor level is supported by large timbers, with wooden planks covered with vinyl tile (except the reactor office which is carpeted). Figures 2.2 and 2.3 provide the layouts of the ground and first floors, respectively, indicating major features relevant to the FSS.

The reactor's biological shield is an 18 foot by 16 foot by 14 foot high reinforced concrete structure, which contained the reactor pool and experimental facilities (a beam port and a thermal column). The reactor core was located in pool an 8 foot by 8 foot by 15 foot deep open topped water filled pool within the biological shield, lined with 1/4 inch Aluminum.

The beam port consisted of an 8 inch diameter steel and Aluminum tube embedded in the east wall of the biological shield. The Thermal Column consisted of a 40 inch square cross-section cavity that penetrated the West wall of the biological shield, which was lined with

/ inch thick Aluminum. The Thermal Column was filed with graphite blocks prior to the conclusion of remedial activities. The biological shield also contains a number of embedded, small diameter (1-2 inch ID) Aluminum pipes that were used for ventilating the BP and TC ventilation, supplying and returning pool water for filtration and demineralization, and draining pool scuppers. An artist's cut-away rendering of the constructed biological shield / pool structure is provided in Figure 2.4.

Reactor room air is exhausted via a stack on the building's roof. Exhaust air intake is via two ducts, located on the North wall on each of the reactor room floors. Two drains to the sanitary sewer system are located on the ground floor.

Worcester Polytechnic Institute Document W19-1579-004, Rev. 0 Reactor Decommissioning - Final Status Survey Plan Section 2, Page 2 of 5 FIGURE 2.1 EXTERIOR VIEW OF WASHBURN SHOPS AND STODDARD LABORATORY BUILDING

Worcester Polytechnic Institute Document W19-1579-004, Rev. 0 Reactor Decommissioning - Final Status Survey Plan Section 2, Page 3 of 5 FIGURE 2.2 GROUND FLOOR WPI REACTOR FACILITY LAYOUT -

MAJOR FEATURES RELEVANT TO FSS North Remaining Portion of Thermal Column Uner Reactor Bio-Shield Centerline Area of Removed Neutron Activated Reactor Core Box and Thermal Column EAST - WEST--->

Liner

Worcester Polytechnic Institute Document W19-1579-004, Rev. 0 Reactor Decommissioning - Final Status Survey Plan Section 2, Page 4 of 5 FIGURE 2.3 REACTOR FACILITY FIRST FLOOR LAYOUT -

MAJOR FEATURES RELEVANT TO FSS North

Worcester Polytechnic Institute Reactor Decommissioning - Final Status Survey Plan Document W19-1579-004, Rev. 0 Section 2, Page 5 of 5 FIGURE 2.4 ARTIST RENDERING OF BIOLOGICAL SHIELD / REACTOR POOL STRUCTURE (PRE-D&D)

I

Worcester Polytechnic Institute Document W19-1579-004, Rev. 0 Reactor Decommissioning - Final Status Survey Plan Section 3, Page I of 14 3.0 REMEDIATION WORK AND RADIOLOGICAL CONDITIONS ENCOUNTERED The remediation work described in Section 2.3.1.2 of the DP has either been completed or was not required. Based upon in-process radiological surveys conducted during and following the remedial work activities, it is believed that radioactivity has been sufficiently removed from the facility to meet license termination criteria.

Approximately 300 cubic feet of equipment and materials was removed during the remediation process. Figure 3.1 shows the locations of the items that were found to contain a majority of the detectable radioactivity. Specifically, the following items were dismantled or demolished and removed as waste, and are no longer present at the facility:

1. Reactor Fuel and PuBe Startup Source

2. Reactor Core Structure

a. Reactor Core Box and Grid

b. Control Blade Drive Shafts and Drive Mechanisms (4)

c. Regulating Blade

d. Safety Blades (3)

e. Ion Chambers and associated hardware (3)

f. Startup Counter and Guide Tube Assembly

g. Sample Irradiation Track

h. Reactor Suspension Posts (4)

3. Thermal Column Contents - Graphite Blocks (164)

4. Reactor Pool Equipment

a. Beam Port Tube Extension

b. Beam Port Shutter and Shutter Housing (embedded in the Biological Shield Concrete)

c. Reactor Core Box Locating Rails

d. Fuel Racks

5. Reactor Pool Water Treatment System (see Figure 3.2 for a photographic view of the area from which this system was removed)

a. Main Ion-Exchange Column

b. Auxiliary Ion-Exchange Column

c. Circulating Pump

d. Filter Vessel

e. Hold Up Tank

f. Associated metal and plastic piping and associated gauges, valves and instruments

Worcester Polytechnic Institute

-'Document W19-1579-004, Rev. 0 Reactor Decommissioning - Final Status Survey Plan Section 3, Page 2 of 14

6. Portions of the Pool and Thermal Column Aluminum Liners (that were neutron activated)

a. An approximate 6 foot by 5 foot section of the pool floor liner under the Reactor core area (see Figure 3.3 for a photographic view of this area after the liner was removed, but before removal of the underlying activated concrete)

b. Approximately 3 feet of the 5 foot long Thermal Column Liner (closest to the reactor) (see Figure 3.4 for a post-remediation photographic view of this area)

c. An approximate 2 foot by 3 foot section of the pool wall liner surrounding the Beam Port Tube and Shutter Housing (see Figure 3.5 for a post-remediation photographic view of this area)

7.

Portions of the Concrete Biological Shield (that were neutron activated)

a. An approximate 5 foot by 4 foot by 1 foot deep section of the pool floor concrete under the reactor core area (which exposed the underlying soil) (see Figure 3.6 for a photographic view of this area)

b. An approximate 1.5 foot by 2 foot by 0.75 foot deep section of the pool wall concrete surrounding the Beam Port Shutter Housing area (see Figure 3.5 for a photographic view of this area)

8. Exhaust vent

a. BP vent pipes, booster fans and duct work (see Figure 3.7 for a photographic view of the remaining vent stubs embedded in the biological shield)

b. TC vent pipes, booster fans and duct work (see Figure 3.8 for a photographic view of the remaining vent stubs embedded in the biological shield)

c. Main exhaust duct header (see Figures 3.9a and 3.9b for photographic views of the remaining portions of this duct header on the ground and first floors, respectively)

9. Other - Sink and drain piping During the decommissioning process, radioactive surface contamination was not found on any structural surface at the Reactor facility. Surface contamination within equipment and systems, which had a theoretical potential for being in contact with reactor-produced radioactivity, was found to be extremely limited, with radioactive contamination only being only found in the reactor pool water treatment system's resin / filter media. The majority of the radioactivity that was encountered during the decommissioning process was found in the various materials that were in close proximity to the reactor core and were subjected to neutron irradiation. These materials were located in the bottom of the reactor pool (i.e., the reactor core box and internal, the pool's aluminum floor liner directly under the reactor core box and the underlying concrete bio-shield floor) and the adjacent experimental features (i.e., the Beam Port shutter housing, Thermal Column graphite and aluminum thermal column liner). These items essentially comprised all of the radioactive materials that were found during conduct of the dismantling work. Figure 3.10, a cross-sectional view of the biological shield, shows the post-remediation configuration of the reactor / biological shield, as it will be encountered during the FFS. This figure specifically

Worcester Polytechnic Institute Reactor Decommissioning - Final Status Survey Plan Document W19-1579-004, Rev. 0 Section 3, Page 3 of 14 indicates the location of remaining materials and where activated aluminum liners and concrete have been removed. Table 3.1 summarizes the radiological conditions that were encountered during the removal of the radioactive equipment and structural materials.

FIGURE 3.1 LOCATIONS OF REMOVED RADIOACTIVE ITEMS Bea Port Shutter Housing / Shutter and 1 -

Surrounding Liner and Blo-Shield: Neutron Activated Aluminum and Concrete Thermal Column Liner Extension; Neutron Activated Aluminum 1

r:raphite Blocks: Neutron Activated Reactor Locating Rails and F..

loor Liner: Neutron Activ ited Aluminum

Worcester Polytechnic Institute Document W19-1579-004, Rev. 0 Reactor Decommissioning - Final Status Survey Plan Section 3, Page 4 of 14 FIGURE 3.2 POST REMOVAL VIEW OF FORMER REACTOR POOL WATER TREATMENT SYSTEM AREA

Worcester Polytechnic Institute Reactor Decommissioning - Final Status Survey Plan Document W19-1579-004, Rev. 0 Section 3, Page 5 of 14 TABLE 3.1 RADIOLOGICAL CONDITIONS ENCOUNTERED DURING DISMANTLING

~~AktA, RADIOILOGCALCODIION Activation on the pool floor liner was encountered directly under the reactor core box area. There, the maximum contact exposure rate was approximately 0.015 mr/hr, with detectable gamma activity observed over an approximately lm 2 area.

Reactor Pool Aluminum Liner Activation on the pool wall liner was encountered within a limited area around the (floor and walls) beam port, with a maximum contact exposure rate of approximately 0.018 mr/hr being found inside the Aluminum beam port shutter housing.

No other indications of activation or surface contamination were observed on the pool liner.

Thermal Column Aluminum The maximum activation was encountered on the liner closest to the reactor core, with Liner a contact exposure rate of -1.5 mr/hr. Indications of activation on the TC liner decreased to non-detectable at about 2.5 feet out from the edge of the reactor core.

After the activated portion of the Aluminum floor liner was removed, activated concrete was encountered below the liner. There, the maximum contact exposure rate was approximately 0.035 mr/hr, with detectable gamma activity observed over an, approximately 1m2 area.

Concrete on the east wall surrounding the embedded beam port shutter housing was found to exhibit maximum contact exposure rates of 0.018 mr/hr (after removal of the Biological Shield Concrete Walls beam port shutter housing). There were no other indications of activation or surface contamination elsewhere on the portions of the interior biological shield walls (where the aluminum liner had been removed).

Activation of the beam port structure was encountered within the Aluminum beam port BeaPorg Tshutter housing to which the beam port tube was connected; The shutter housing had a maximum contact exposure rate of approximately 0.015 mr/hr A maximum contact exposure rate of 0.030 mr/hr was observed on the demineralizer while filled with resin that had been used to purify reactor pool water over the Rytear Woperating life of the reactor. With the exception of used filter cartridges, all internal System surfaces of the system exposed by the dismantling process were found to be free of detectable surface contamination.

Two small radioactive items were (possibly pieces of an irradiated flux wire) were found on the floor near the stairs area. The pool water treatment system was also located in Ground Floor, East of Biological this area, which contained reactor pool water and known radioactive resin. No other Shield Centerline radioactive contamination or unexplained elevated radioactivity was encountered in this area.

Radioactive Material Storage Bags of contaminated operational era trash had been stored in this room. Historically Room, Ground Floor radioactive sealed sources were also stored in this area.

Historically a classroom area. Sealed sample vials containing irradiated materials were First Floor Reactor Room handled on a laboratory bench in this area. No other radioactive contamination or unexplained elevated radioactivity was encountered in this area.

Reactor Tool Closet, First Floor This room was used to store tools and supplies for reactor maintenance, and contained a radioactive material disposal sink. However, no radioactive contamination or unexplained elevated radioactivity was encountered in this area.

Worcester Polytechnic Institute Document W19-1579-004, Rev. 0 Reactor Decommissioning - Final Status Survey Plan Section 3, Page 6 of 14 FIGURE 3.3 MID-REMEDIATION VIEW OF REACTOR POOL FLOOR (AFTER LINER REMOVAL - BEFORE CONCRETE REMOVAL)

Worcester Polytechnic Institute Document W19-1579-004, Rev. 0 Reactor Decommissioning - Final Status Survey Plan Section 3, Page 7 of 14 FIGURE 3.4 POST REMEDIATION VIEW OF REMAINING PORTION OF THERMAL COLUMN LINER

Worcester Polytechnic Institute Document W19-1579-004, Rev. 0 Reactor Decommissioning - Final Status Survey Plan Section 3, Page 8 of 14 FIGURE 3.5 POST REMEDIATION VIEW OF BEAM PORT TUBE / SHUTTER HOUSING AREA

Worcester Polytechnic Institute Document W19-1579-004, Rev. 0 Reactor Decommissioning - Final Status Survey Plan Section 3, Page 9 of 14 FIGURE 3.6 POST REMEDIATION VIEW OF REACTOR POOL FLOOR AREA

Worcester Polytechnic Institute Reactor Decommissioning - Final Status Survey Plan Document W19-1579-004, Rev. 0 Section 3, Page 10 of 14 FIGURE 3.7 POST REMEDIATION VIEW OF REMAINING BEAM PORT VENT STUB, WATER TREATMENT RETURN LINE AND SCUPPER DRAIN OUTLET PIPE

Worcester Polytechnic Institute Reactor Decommissioning - Final Status Survey Plan Document W19-1579-004, Rev. 0 Section 3, Page 11 of 14 FIGURE 3.8 POST REMEDIATION VIEW OF REMAINING THERMAL COLUMN VENT AND DRAIN STUBS

Worcester Polytechnic Institute Document W19-1579-004, Rev. 0 Reactor Decommissioning - Final Status Survey Plan Section 3, Page 12 of 14 FIGURE 3.9a POST REMEDIATION VIEW OF REMAINING GROUND FLOOR EXHAUST DUCT AND PLENUM

Worcester Polytechnic Institute Document W19-1579-004, Rev. 0 Reactor Decommissioning - Final Status Survey Plan Section 3, Page 13 of 14 FIGURE 3.9b VIEW OF FIRST FLOOR EXHAUST DUCT PLENUM

Worcester Polytechnic Institute Document W19-1579-004, Rev. 0 Reactor Decommissioning - Final Status Survey Plan Section 3, Page 14 of 14 FIGURE 3.10 POST REMEDIATION CONFIGURATION OF BIOLOGICAL SHIELD

Worcester Polytechnic Institute Document W19-1579-004, Rev. 0 Reactor Decommissioning - Final Status Survey Plan Section 4, Page 1 of 9 4.0 RADIOLOGICAL CONTAMINANTS AND CRITERIA 4.1 RADIOACTIVE CONTAMINANTS IDENTIFIED DURING DECOMMISSIONING Samples of waste materials were obtained during the remediation process for 10CFR61 characterization purposes. As a minimum, each unique waste stream that had the potential for containing different mixtures of radionuclides was sampled. This data is useful for identifying the potential radiological contaminants that might remain in various structural materials that will be present at the time of the FSS.

Five samples were analyzed by an outside laboratory: the samples consisted of the following materials:

• Stainless Steel Regulating Blade: Neutron Activated

" Aluminum from the Thermal Column liner: Neutron Activated

" Concrete from the Biological Shield: Neutron Activated

" Graphite from the Thermal Column: Neutron Activated

" Resin from the Reactor Pool Water Treatment System Table 4.1 summarizes the analytical results for the radionuclides that were positively identified and attributable to Reactor operation in these waste stream samples.

In addition, two pre-FSS concrete samples were taken from the remaining biological shield, at locations likely to represent peak radionuclide concentrations currently present anywhere in the biological shield concrete. These samples were obtained at the end of remediation work for the purpose of providing a preliminary indication that sufficient materials had been removed. The samples were obtained from the remaining edge of the pool floor demolition area and from the pool's interior east wall adjacent to the Beam Port tube area. Table 4.2 provides a summary of the analytical results for the radionuclides that were positively identified in these samples and which are attributable to Reactor operation.

A complete presentation of the laboratory results is presented in Appendix A. The locations from which the seven samples (five waste stream and two pre-FSS) were obtained are shown in Figure 4.1.

Worcester Polytechnic Institute Document W19-1579-004, Rev. 0 Reactor Decommissioning - Final Status Survey Plan Section 4, Page 2 of 9 TABLE 4.1

SUMMARY

OF WASTE STREAM AND PRE-FSS SAMPLE CHARACTERIZATION RESULTS

• ri Smple ID No.

WPTI #1 IWPI #2 WPI #3 '

'WPI # 4 WPI #5 WPI #8 WPI #9 Stainless Aluminum Concrete Graphite Ion-Steel Material (neutron (neutron (neutron (neutron exchange Concrete Concrete activated) activated) activated) activated)

Resin Contents Remaining Remaining Biological g

biological Regulating Thermal Bioloc Contents of reactor biological shield Item Sampled control blade column floor and of Thermal pool water shield floor-near liner Column treatment wall-near wall reactor system beam port core area Purpose of Waste Waste Waste Waste Waste stream stream stream stream Pre-FSS Pre-FSS char.

char.

char.

char.

Radionuclide aiCi/g AiCi

/g i/g tCi/g ACi/g itCi/g ACi/g Fe-55 4.38E-01 8.69E-03 Ni-63 3.47E-01 Co-58 2.27E-06 Co-60 2.24E-01 6.70E-04 3.39E-06 3.38E-06 1.55E-07 7.79E-07 5.85E-07 Zn-65 3.36E-05 Cs-134 3.92E-07 1.85E-07 Cs-137 3.01E-04 Eu-152 1.52E-05 4.02E-04 2.65E-06 2.26E-06 Eu-154 8.06E-07 2.57E-05

• Includes detected radionuclides only and excludes naturally occurring radionuclides -

see Appendix A for complete reporting of analytical results TABLE 4.2

SUMMARY

OF ANALYTICAL RESULTS FOR PRE-FSS SAMPLES FROM THE BIOLOGICAL SHIELD SAMPLE ID, SAMPLE LOCATION (Location Shown Co-60' Cs-1g)

,Eu-152

........1,...............

(Pc i/g (P ig

>*c/g),

Interior East Bio-Shield Wall, WPI # 8 0.78 0.19 2.65 Adjacent to Beam Port Interior Bio-Shield Floor, Edge WPI # 9 0.59 ND 2.26 of Demolition Area I

Worcester Polytechnic Institute Document W19-1579-004, Rev. 0 Reactor Decommissioning - Final Status Survey Plan Section 4, Page 3 of 9 FIGURE 4.1 LOCATION OF WASTE STREAM CHARACTERIZATION AND PRE-FSS SAMPLES

Worcester Polytechnic Institute Document W19-1579-004, Rev.O Reactor Decommissioning - Final Status Survey Plan Section 4, Page 4 of 9 4.2 CHARACTERISTICS OF THE RADIOACTIVE CONTAMINATES THAT POTENTIALLY COULD BE PRESENT DURING FSS The radionuclide mixtures represented by four of the five waste streams and the two pre-FSS samples are likely to be representative of any residual radioactivity that might be present at the time the FSS is performed. All potentially activated stainless steel was dismantled intact (i.e., without segmentation), and removed from the facility. Accordingly, no residual radioactivity due to this material can be expected to be present. The activated Aluminum radionuclide mixture has the potential for being found in the remaining portions of floor and lower walls of reactor pool liner and thermal column liner that were in close proximity to the reactor core. Likewise, the activated concrete radionuclide mixture (three samples) has the potential for being found in the remaining portions of the biological shield walls and floor that were in close proximity to the reactor core. All irradiated graphite was removed from the Reactor facility, and will not be present during the FSS. The contaminated resin was removed along with the reactor pool water treatment system's equipment and piping and is no longer present, but would likely be representative of any contamination found on surfaces that had been wetted by pool water. It is also slightly possible that if surface contamination were found on structural surfaces that it could be due to dust or debris created during the process of removing the activated graphite, activated aluminum liners or activated bio-shield concrete. With the exception of the Fe-55 contained in activated aluminum, all potential sources of radioactivity at WPI (i.e., activated concrete and surfaces wetted with pool water) are comprised entirely of one or more gamma emitters (i.e., Co-60, Cs-134, Cs-137, Eu-152 and Eu-154) that can be readily identified using portable detection equipment and the mixture readily quantified by gamma spectroscopy methods. The one exception (i.e., the radioactivity that may be found in Aluminum liners) does contain a hard-to-detect radionuclide (Fe-55); however, Co-60 which represents 7.13% of the total radioactivity in activated aluminum, can be readily detected and used to scale the hard-to-detect radionuclide Fe-55.

Table 4.3 provides a summary of the radionuclides mixtures that would be representative of residual radioactivity that could be encountered at the WPI facility during the FSS

Worcester Polytechnic Institute Reactor Decommissioning - Final Status Survey Plan Document W19-.1579-004, Rev.0 Section 4, Page 5 of 9 TABLE 4.3

SUMMARY

OF THE RADIONUCLIDES MIXTURES REPRESENTATIVE OF RESIDUAL RADIOACTIVITY Original Radionuclide C o sitf~

i i u en t '

Con.aminantPotential Locations at WPI Constitent Contaminant

' Ii::]*:

Source(fractional abundance)

Aluminum liners within the reactor pool in close proximity to the reactor core, thermal column liner and beam port tube.

Note: a remote possibility exists for loose surface Aluminum contamination being present at locations where (or Fe-55 (0.9251),

(neutron activated) near where) liners where segmented (such as the pool Zn.-65 (0.0036) floor and underlying soil surface beneath the reactor core area that became exposed when the liner and concrete above it was removed - it is also remotely possible that this soil could have become directly neutron activated).

Concrete Biological Shield, in close proximity to the reactor core and beam port shutter and tube.

Note: a remote possibility exists for loose surface Co-60 (0.1717),

Concrete contamination being present at locations where (or Cs-6134 (0.0202),

(neutron activated) near where) the activated concrete that was Eu-152 (0.7677),

demolished and removed (such as the pool floor and Eu-152 (0.7677)

Eu-154 (0.0404) underlying soil surface beneath the reactor core area that became exposed when the concrete above it was removed - it is also remotely possible that this soil could have become directly neutron activated).

None expected - all graphite within the thermal column was removed during decommissioning.

Graphite Note: a remote possibility exists for loose surface Co-60 (0.0078),

(neutron activated) contamination being present at locations where (or Eu-152 (0.9325),

near where) the activated graphite was handled and Eu-154 (0.0596) removed, such as the remaining thermal column liner and adjacent floor surfaces.

Surface that were wetted by reactor pool water (such as interior pool liner surfaces, drain and water Radioactivity in treatment system piping embedded within the Co-60 (0.0005),

Reactor Pool biological shield, scupper drains and operating floor Cs-137 (0.9995)

Water surfaces and, floor wall surfaces adjacent to the removed pool water treatment system).

Worcester Polytechnic Institute Reactor Decommissioning - Final Status Survey Plan Document W19.-1579-004, Rev.0 Section 4, Page 6 of 9 4.3 FSS CRITERIA Future uses of the former Reactor facility at WPI have not yet been completely defined.

However, it is likely that the facility will continue being used in some capacity for academic programs: as such, license termination will be based on the un-restricted release scenario.

In its DP, WPI committed to the use of NRC screening values in lieu of development of site-specific DCGLs. These screening values were presented Tables 2.2 and 2.3 of the DP for building surface and surface soil, respectively, and are presented herein as Tables 4.4 and 4.5. (Note: soil screening values for Nb-94 and Eu-152 were revised for DP Table 2.3 on September 25, 2012 (DP revision 2) to meet the radionuclide concentration limits in the MOU between NRC and EPA regarding consultation level).

TABLE 4.4 LICENSE TERMINATION SCREENING VALUES FOR BUILDING SURFACE CONTAMINATION Hydrogen-3 (Tritium) 1.2E+08 Carbon-14 14C 3.7E+06 Sodium-22 22Na 9.5E+03 Sulfur-35 35S 1.3E+07 Chlorine-36 36C1 5.03E+05 Manganese-54 54Mn 3.2E+04 Iron-55 55Fe 4.5E+06 Cobalt-60 6 0 Co 7.1E+03 Nickel-63 63Ni 1.8E+06 Strontium-90 9°Sr 8.7E+03 Technetium-99 99Tc 1.3E+06 Iodine-129 1291 3.5E+04 Cesium-137 137Cs 2.8E+04 Iridium-192 S

9 2Ir 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-to-100 percent range),

may calculate site-specific screening levels using RESRAD-BUILD Version 3.0 (Ref. 34).

• • Units are disintegrations per minute (dpm) per 100 square centimeters (dpm/100 cm 2).

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 mSv/y~r (25 mrem / 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 - Final Status Survey Plan Document W19-.1579-004, Rev.0 Section 4, Page 7 of 9 TABLE 4.5 LICENSE TERMINATION SCREENING VALUES FOR SURFACE SOIL K':

SURFACE SOIL SCREENING RADIONCLIDE*VALU.ES.**FOR RADIOUCLIE*

SMBOL UNRESTRICTED RELEASE Hydrogen-3 (Tritium) 3H 1.1E+02 Carbon-14 14C 1.2E+01 Sodium-22 22Na 4.3E+00 Sulfur-35 35S 2.7E+02 Chlorine-36 3C1 3.6E-01 Calcium-45 45Ca 5.7E+01 Scandium-46 46Sc 1.5E+01 Manganese-54 54Mn 1.5E+01 Iron-55

• 5 Fe 1.OE+04 Cobalt-57 57Co 1.5E+02 Cobalt-60 60Co 3.8E+00 Nickel-59 59Ni 5.5E+03 Nickel-63 63Ni 2.1E+03 Strontium-90 90Sr 1.7E+00 Niobium-94 94Nb 3.OE+00 ****

Technetium-99 99Tc 1.9E+01 Iodine-129 1291 5.OE-01 Cesium-134 14Cs 5.7E+00 Cesium-137 13Cs 1.1E+01 Europium-152 15 2Eu 7.OE+00 ****

Europium-154 15 4Eu 8.OE+00 Iridium-192 1921r 4.1E+01 Lead-210 2°Pb 9.OE-01 Radium-226 226Ra 7.OE-01 Radium-226+C 226Ra+C 6.OE-01 Actinium-227 22'Ac 5.OE-01 Actinium-227+C 227Ac+C 5.OE-01 Thorium-228 22Th 4.7E+00 Thorium-228+C 228Th+C 4.7E+00 Thorium-230 230Th 1.8E+00 Thorium-230+C 230Th+C 6.OE-01 (Note: Table 4.5 is continued on next page)

Worcester Polytechnic Institute Reactor Decommissioning - Final Status Survey Plan Document W19.-1579-004, Rev.0 Section 4, Page 8 of 9 TABLE 4.5 (continued)

LICENSE TERMINATION SCREENING VALUES FOR SURFACE SOIL SURFACE SOIL SCREENING RALIOINUCbllE*

VALUES** FQR UNRESTRICTED RELEASE Thorium-232 232TH 1.(1E+00 Thorium-232+C 232TH+C

1. 1E+00 Protactinium-231 23 1Pa 3.OE-01 Protactinium-231+C 23 1Pa+C 3.OE-O01 Uranium-234 2 34 U 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+00 Plutonium-241 241Pu 7.2E+O1 Americium 24 1Am 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 (NUREG/CR-5512 Volumes 1, 2, and 3).

• • These values represent surface soil concentrations of individual radionuclides that would be deemed in compliance with the 25 mren/year (0.25 mSv/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).

• • • Screening values are in units of (pCi/g) equivalent to 25 mrem/year (0.25 mSv/year).

To convert from pCi/g to units of Becquerel per kilogram (Bq/kg) divide each value by 0.027. These values were derived based on selection of the 90th 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.

• • • • From: MOU Table 1: Consultation Triggers for Residential and Commercial /

Industrial Soil Contamination, "Memorandum of Understanding Between The Environmental Protection Agency and The Nuclear regulatory Commission2 "

Worcester Polytechnic Institute Document W19-1579-004, Rev.O Reactor Decommissioning - Final Status Survey Plan Section 4, Page 9 of 9 In order to satisfy the surface contamination criteria, measurement of gross beta surface activity will be based upon application of the Sum-of-Fractions rule for the radionuclides potentially present. The radionuclide mixture found in the reactor pool water treatment system's resin should be representative of any radioactivity that might have become dispersed and deposited on the Reactor facility structural or equipment surfaces. As such, that radionuclide mixture will be the basis for application of a gross beta DCGL equivalent to the DP Table 2.2 surface contamination criteria. Based upon that mixture (99.95% Cs-137 and 0.05% Co-60), the gross beta DCGL would be 23,700 dpm/1OOcm 2, based upon the SOF rule. The derivation of this value is provided in Appendix B.

However, while there is no indication of this occurring, it is remotely possible that a surface at WPI Reactor facility could have become contaminated with a different mixture of radionuclides during the dismantling work process, where some of the activated materials (aluminum, concrete and graphite) could have taken on a dispersible form. If any readily detectable surface contamination is detected during the FSS, at locations not likely associated with surfaces wetted by pool water, the gross beta DCGL given above will be evaluated for appropriateness based on the location of the survey unit and knowledge of the various radionuclide mixtures that could affect that survey unit.

Satisfying the soil contamination criteria (either for soil or. activated materials) will be demonstrated by the Sum-of-Fractions approach. The Sum-of-Fractions of contaminant concentrations divided by their respective Default Screening Values must therefore be

<Unity (i.e., <1.0).

Worcester Polytechnic Institute Document W19-.1579-004, Rev.O Reactor Decommissioning - Final Status Survey Plan Section 5, Page I of 1 5.0 QUALITY ASSURANCE PROGRAM A FSS Quality Assurance Project Plan (QAPP), appropriate for implementing the final status survey and developing associated documentation, will be employed. The QAPP shall incorporate the appropriate regulatory requirements applicable to the planning and conduct of radiological surveys necessary for the termination of the Reactor license and the release of the site for unrestricted use.

Worcester Polytechnic Institute Document W19-1579-004, Rev.O Reactor Decommissioning - Final Status Survey Plan Section 6, Page I of 15 6.0 FINAL STATUS SURVEY APPROACH The objective of the FSS is to demonstrate that remedial actions have been effective in removal/reduction of radiological materials and contamination, and that the post-remediation radiological conditions satisfy the NRC-approved criteria for termination of the WPI Reactor License and enable future use of the WPI facility without radiological restrictions. The FSS will be performed using the guidelines and recommendations presented in NUREG-1757 and MARSSIM. FSS activities will be performed by trained and qualified personnel, using properly calibrated equipment, sensitive to the potential contaminants, following documented operating procedures. The QAPP will contain a list of the procedures applicable to this FSS.

6.1 CLASSIFICATION BY CONTAMINATION POTENTIAL For the purposes of guiding the degree and nature of FSS coverage, areas are first classified as impacted or non-impacted (i.e., areas that are considered likely or unlikely to have residual radioactivity from licensed activities). Non-impacted areas do not require further evaluation. For impacted areas MVARSSIM identifies three classifications of areas, according to contamination potential, as follows:

" Class 1 - Areas that have, or had prior to remediation, a potential for radioactive contamination (based on site operating history) or known contamination (based on previous radiation surveys) above the DCGL.

Examples include: site areas previously subjected to remedial actions; locations where leaks or spills are known to have occurred, and; former waste storage areas.

• Class 2 - Areas that have, or had prior to remediation, a potential for radioactive contamination or known contamination, but are not expected to exceed the DCGL.

Examples include: locations where radioactive materials were present in an unsealed form; potentially contaminated transport routes; areas handling low concentrations of radioactive materials, and; areas on the perimeter of former contamination control areas.

" Class 3 -

Any impacted areas that are not expected to contain any residual radioactivity, or are expected to contain levels of residual radioactivity at a small fraction of the DCGL, based on site operating history and previous radiation surveys. Examples include: buffer zones around Class 1 and Class 2 areas, and; areas with a very low potential for residual contamination, but having insufficient information to justify a non-impacted classification.

The bases for classification are the facility history (including the initial Historical Site Assessment and radiological monitoring conducted during characterization), and remedial activities.

Once approval of the FSS Plan is obtained (through a subsequent license amendment request to the NRC), changes to the classification of an area may be made as long as the classification is changed to one of higher contamination potential. A license amendment

Worcester Polytechnic Institute Reactor Decommissioning - Final Status Survey Plan Document W19-.1579-004, Rev.O Section 6, Page 2 of 15 pursuant to 10CFR50.90 shall be obtained if the change would decrease an area classification (i.e., impacted to non-impacted, Class 1 to Class 2, Class 2 to Class 3, or Class 1 to Class 3).

6.2 IDENTIFICATION OF SURVEY UNITS Impacted areas are divided into survey units for implementing the FSS. A survey unit is a portion of a facility with common contaminants and contamination potential and contiguous surfaces or areas. Table 6.1 lists the MARSSIM-suggested survey unit areas applicable to the WPI facility.

The area of individual survey unit will follow these suggested maximum sizes.

TABLE 6.1 SURVEY UNIT AREAS Cls Recom en*ded Surv'ey.Unit Areas' C lass StSructures Laind 1

Up to 100 m 2 Not applicable to WPI 2

100 to 1,000 m 2 Not applicable to WPI No limit Not applicable to WPI Based on historical assessments made during preparation of the DP, and radiological monitoring conducted during remedial activities, the survey units for the WPI Reactor facility areas that are currently expected to be included in the FSS, is presented in Table 6.2, along with the estimated areas, anticipated contamination potential classifications, and a description of the as-found radiological conditions and the remedial activities (if any) conducted within each area. Classifications and survey unit boundaries may change based on results as the FSS progresses. If classifications or boundaries change, the survey of the survey unit will be redesigned and the survey and data evaluation repeated.

Worcester Polytechnic Institute Document W19.-1579-004, Rev.O Reactor Decommissioning - Final Status Survey Plan Section 6, Page 3 of 15 TABLE 6.2 FSS SURVEY UNITS AND CLASSIFICATIONS Building SreUnt(U)MARSSI1M Apr.

gArea Survey Units (SU)

Class' Pre-Remediation Radiological Status Survey Unit Area "Class Size 1.0 Biological Shield /

Reactor Pool 1.1 Reactor Pool Aluminum Liner (floor and walls) 1 Activation on the pool floor liner was encountered directly under the reactor core box area. There, the maximum contact exposure rate was approximately 0.015 mr/hr Activation on the pool wall liner was encountered within a limited area around the beam port, with the maximum contact exposure rate of approximately 0.018 mr/hr, being found inside the aluminum beam port shutter housing.

No other indications of activation or surface contamination were observed elsewhere on the pool liner.

47 M 2

1.2 Thermal Column Aluminum Liner The maximum activation was encountered on 2 m 2 the liner closest to the reactor core, with a contact exposure rate of -1.5 mr/hr. Indications of activation on the TC liner decreased to non-detectable at about 2.5 feet out from the reactor core.

1.3 Interior Concrete Floor 1

After the activated portion of the Aluminum 5 M2 floor liner was removed, activated concrete was encountered below the liner. There, the maximum contact exposure rate was approximately 0.035 mr/hr.

1.4 Interior Concrete Walls 1

Concrete on the east wall surrounding the 43 m 2 embedded beam port shutter housing was found to exhibit maximum contact exposure rates of 0.018 mr/hr (after removal of the beam port shutter housing). There were no other indications of activation or surface contamination elsewhere on the portions of the interior biological shield walls (where the aluminum liner had been removed).

1.5 Beam Port Tube (- 6" dia. Metal tube embedded in the concrete biological shield) 1 Activation of the beam port structure was encountered within the aluminum beam port shutter housing to which the beam port tube was connected; the interior of the shutter housing had a maximum contact exposure rate of approximately 0.15 mr/hr.

1.2 m long

Worcester Polytechnic Institute Reactor Decommissioning - Final Status Survey Plan Document W19-1579-004, Rev.O Section. 6, Page 4 of 15 TABLE 6.2 (continued)

FSS SURVEY UNITS AND CLASSIFICATIONS Approx.

-Buldng MA R

S I Survey Units (SU)

MARSSM Pre-Remediation Radiological Status Survey Unit Area

'Class A re a.*

'*i i

• *i:*

"" i :...

....:-. S ize 1.0 Biological Shield /

Reactor Pool (cont.)

1.6 Soil Under Removed Concrete Floor Area 1

Activated concrete above the soil was rubblized and removed.

-1 m 2 1.7 Top of Biological Shield (Operating 1

No radioactive contamination or unexplained 39 m 2 floor and stub walls (sides and top) elevated radioactivity was encountered in this surrounding the reactor pool, lower area.

South room wall above Bio-shield and Reactor bridge spanning the pool) 1.8 Embedded Piping (in biological shield) 1.8.1 Drain line from pool scupper 1

No radioactive contamination was encountered

~10 m long drains (I"ID) at the two scupper drain basins or the cut pipe stub emerging from the east bio-shield wall.

1.8.2 Return line from pool water 1

No radioactive contamination was encountered

'3.7 m long treatment system (1"ID) at the cut pipe stub emerging from the east bio-shield wall.

1.8.3 Vent line from beam port 1

No radioactive contamination or unexplained

-1.8 m long tube (I"ID) elevated radioactivity was encountered in this item.

1.8.4 Drain line from Beam port 1

The BP end of this line was neutron activated "1.5 m long Shutter Housing (1"ID) where it joined the BP shutter housing. No radioactive contamination was encountered at the cut pipe stub emerging from the east bio-shield wall.

1.8.5 Feed line from pool to pool 1

No radioactive contamination was encountered

-3 m long water treatment system (2"ID) at the cut pipe stub emerging from the bio-shield east wall.

1.8.6 Vent line from thermal 1

No radioactive contamination or unexplained

-1.2 m long column (1"ID) elevated radioactivity was encountered at the TC end or at the cut pipe stub emerging from the west bio-shield wall.

1.8.7 Vent line from thermal 1

No radioactive contamination or unexplained

-1.2 m long column (1"ID) elevated radioactivity was encountered at the TC end or at the cut pipe stub emerging from the west bio-shield wall.

1.8.8 Vent line from thermal 1

No radioactive contamination or unexplained

-1.2 m long column (1"ID) elevated radioactivity was encountered at the TC end or at the cut pipe stub emerging from the west bio-shield wall.

1.8.9 Vent line from thermal 1

No radioactive contamination or unexplained

-1.2 m long column (1"ID) elevated radioactivity was encountered at the TC end or at the cut pipe stub emerging from the west bio-shield wall.

1.8.10 Vent line from thermal column (2"ID) 1 No radioactive contamination or unexplained elevated radioactivity was encountered at the TC end or at the cut pipe stub emerging from the west bio-shield wall.

-1.8 m long

Worcester Polytechnic Institute Reactor Decommissioning - Final Status Survey Plan Document W19-1579-004, Rev.O Section, 6, Page 5 of 15 TABLE 6.2 (continued)

FSS SURVEY UNITS AND CLASSIFICATIONS Aprox.

Building Survey Units (SU)

MARSSIM Pre-Remediation.Radiological Status*

Survey Unit Area Class Size 1.0 1.9 Exterior Walls of Biological Shield Biological 1.9.1 North Wall 2

No radioactive contamination or unexplained 22 M 2 Shield /

elevated radioactivity was encountered in this Reactor area.

Pool 1.9.2 West Wall 1

No radioactive contamination or unexplained 18 M 2 (cont.)

elevated radioactivity was encountered in this area, however this wall has the access opening into the Thermal Column, from which activated graphite was removed, and the TC was used for pool access during D&D activities.

1.9.3 Thermal Column Shield Door 1

No radioactive contamination or unexplained 5 M 2 elevated radioactivity was encountered on this door; however, the inside of this door was in close proximity to the activated graphite 1.9.4 East Wall 1

No radioactive contamination or unexplained 19 m 2 elevated radioactivity was encountered in this area; however portions of this wall are in close proximity to the removed reactor water treatment system.

2.0 2.1 South Lower Wall and Floor, West 1

No radioactive contamination or unexplained 65 m 2 Ground of Biological Shield Centerline elevated radioactivity was encountered in this Floor area, however portions of this wall are in close Reactor proximity to the Thermal Column access Facility opening, from which activated graphite was removed, and the area was used for radioactive waste handling during D&D activities.

2.2 North and West Lower Walls, 2

No radioactive contamination or unexplained 16 m 2 West of Biological Shield Centerline elevated radioactivity was encountered in this area.

2.3 South Lower Wall and Floor,East 1

Two small radioactive items were (possibly 104 m 2 of Biological Shield Centerline pieces of an irradiated flux wire) were found on the floor near the stairs area. The pool water treatment system was also located in this area, which contained reactor pool water and known radioactive resin. No other radioactive contamination or unexplained elevated radioactivity was encountered in this area.

2.4 North and East Lower Walls, East 2

No radioactive contamination or unexplained 21 M 2 of Biological Shield Centerline elevated radioactivity was encountered in this area.

2.5 Upper Walls, Ceiling, and Exterior 2

No radioactive contamination or unexplained 100 M 2 of Suspended Equipment Surfaces, elevated radioactivity was encountered in this West of the Biological Shield area.

centerline (excluding the areas separately covered by SU 1.9.1 and SU 2.6).

.1 Worcester Polytechnic Institute Reactor Decommissioning - Final Status Survey Plan Document W19-1579-004, Rev.O Section 6, Page 6 of 15 TABLE 6.2 (continued)

FSS SURVEY UNITS AND CLASSIFICATIONS Building I

Approx nArea Survey Units (SU)

MARsSil Pre-Remrediation Radiological Status SL.Urvey Unit, Size 2.0 2.6 Upper Walls, Ceiling, and Exterior of 2

No radioactive contamination or unexplained 166 m 2 Ground Suspended Equipment Surfaces, East of elevated radioactivity was encountered in this Floor the Biological Shield centerline area.

Reactor (excluding areas separately covered by Facility SU 1.93).

(cont.)

2.7 WPI RAM Storage Room, Lower 1

Two bag of contaminated operational trash had 29 m 2 Walls and Floor been stored in this room. Historically radioactive sealed sources and samples were also stored in this area.

2.8 WPI RAM Storage Room, Upper 1

Two bag of contaminated operational trash had 44 m 2 Walls, Ceiling, and exterior of been stored in this room. Historically suspended equipment surfaces radioactive sealed sources and samples were also stored in this area.

3.0 3.1 Lower Walls and Floor of Reactor 3

This area was physically separated from the 36 m 2 First Floor office reactor room, used for administrative purposes, Reactor with no known radiological work activities Facility conducted in this area. No radioactive contamination or unexplained elevated radioactivity was encountered in this area.

3.2 Upper walls and ceiling of Reactor 3

This area was physically separated from the 66 m 2 Office reactor room, used for administrative purposes, with no known radiological work activities conducted in this area. No radioactive contamination or unexplained elevated radioactivity was encountered in this area.

3.3 Lower Walls and Floor, West of 2

Historically a class room area, where check 67 m 2 Biological Shield Centerline (excludes sources and sealed samples vials containing lower portion of South wall above Bio-irradiated materials were stored in a cabinet.

shield No other radioactive contamination or unexplained elevated radioactivity was encountered in this area.

3.4 Upper walls and ceiling, West of 2

No radioactive contamination or unexplained 1'35 m 2 Biological shield centerline elevated radioactivity was encountered within this area.

3.5 Lower Walls and Floor, East of 1

Historically a class room area, sealed sample 105 m 2 Biological Shield Centerline (excludes vials containing irradiated materials were lower portion of South wall above Bio-handled on a lab' bench in this area. No other shield radioactive contamination or unexplained elevated radioactivity was encountered in this area.

3.6 Upper walls and ceiling, East of 2

No radioactive contamination or unexplained 188 m 2 Biological shield centerline elevated radioactivity was encountered within this area.

Worcester Polytechnic Institute Reactor Decommissioning - Final Status Survey Plan Document W19-.1579-004, Rev.O Section 6, Page 7 of 15 TABLE 6.2 (continued)

FSS SURVEY UNITS AND CLASSIFICATIONS Building SMARSSI

.Approx.

Ara Survey Units (SU)

.Pre-Remediation Radiological Statu sIus, Survey Unit Area Cass 3.0 3.7 Reactor Tool closet lower walls and 1

This room was used to store tools and supplies 27 m 2 First Floor floors for reactor maintenance, and contained a Reactor radioactive material disposal sink. However, no Facility radioactive contamination or unexplained (cont.)

elevated radioactivity was encountered in this area.

3.8 Reactor Tool Closet, upper walls and 1

No radioactive contamination or unexplained 18 M 2 ceiling elevated radioactivity was encountered within this area.

4.0 4.1 Exhaust ventilation duct interior 1

No radioactive contamination or unexplained Not Other surfaces (exterior to the reactor room) elevated radioactivity was encountered within /

Applicable Areas near this system, before or during

/items decommissioning.

4.2 Floor drain and sanitary sewer line 1

No radioactive contamination or unexplained Not elevated radioactivity was encountered within Applicable this system. Prior to decommissioning the reactor pool water had been batch discharged to the sanitary sewer system, with water sampling and analysis indicating no detectable radioactivity.

4.3 HVAC units - 3 units (room re-circ.

1 No radioactive contamination or unexplained Not Only) elevated radioactivity was encountered within /

Applicable near this system, before or during decommissioning.

Worcester Polytechnic Institute Document W19-.1579-004, Rev.O Reactor Decommissioning - Final Status Survey Plan Section 6, Page 8 of 15 6.3 TESTING TO DEMONSTRATE COMPLIANCE The Null Hypothesis for the statistical test to demonstrate compliance with project criteria is "Residual radiological contamination levels exceed project criteria." The objective of the FSS is to reject this Null Hypotheses, by demonstrating that, at a Type I (a) decision error level of 0.05 (i.e., 95% confidence level), the contamination does not exceed criteria. The Type II (p3) decision error level is also 0.05.

Because there are multiple potential contaminants, compliance with concentration criteria for soil or activated materials will be evaluated using the Sum-of-Fractions approach. For both the soil and building surface surveys, the sign test is the appropriate statistical test of compliance. However, if all data within a survey unit are less than the DCGL levels, the Null Hypothesis will be rejected, and no statistical test will need to be performed.

6.4 SURVEY DATA REQUIREMENTS The parameter known as the "relative shift" is used to establish the number of data points needed to demonstrate that residual contamination criteria have been satisfied. The relative shift, which effectively describes the distribution of final sample data, is calculated as follows:

(1) A/

(DCGL-LBGR) /

Where:

A / a

= relative shift DCGL = cleanup level (Section 4.3).

LBGR = lower bound of the gray region and is defined in the DQOs as 50 percent of the DCGL.

Where final sample data are not 'yet available, MARSSIM guidance (Section 5.5.2.2) assigns a value of one-half of the DCGL for the LBGR.

= standard deviation of the sample concentrations in the survey unit. Where final sample data are not yet available, MARSSIM guidance (Section 5.5.2.2) is to use a value of 30 percent of the DCGL.

Using the equation for relative shift and MARSSIM guidance for situations where final sample data are not yet available, the relative shift for design purposes is (1-0.5) / 0.3 for a value of 1.67. Based on the relative shift of 1.67 and Type I and Type II decision errors of 0.05, the number of required data points from each survey unit to perform the sign test, as obtained from MARSSIM guidance (Table 5.5) is 17.

Once actual sample data are collected the MARSSIM DQO process requires a retrospective assessment of the selected LBGR and a values to confirm that an adequate number of data points were obtained for final evaluation.

Worcester Polytechnic Institute Document W19-1579-004, Rev.O Reactor Decommissioning - Final Status Survey Plan Section 6, Page 9 of 15 6.5 SURVEY LOCATIONS MARSSIM recommends a random-start systematic triangular measurement or sampling pattern for FSS of Class 1 and Class 2 survey units. This type of triangular pattern will be used for this FSS, except where dimensions and/ or other factors related to a specific survey unit require use of an alternate pattern. The spacing (L) between data points on a triangular pattern is determined by:

L = [(Survey Unit Area) / (0.866 x number of data points)]

To simplify the designation of data points while assuring a sufficient number of data points are obtained for statistical purposes, the value of L is rounded to the nearest whole meter. If the systematic pattern does not provide sufficient data points to satisfy the number determined in Section 6.4, additional data points will be identified, using a random-number technique.

However, where there are Class 1 and 2 survey units with structural surfaces of < 10 M 2,

these will not be evaluated on a statistical basis as previously described: instead, a minimum of four measurements (or samples) will be obtained from such areas at locations having the highest probability of exhibiting residual contamination, based on judgment using prior knowledge of contaminant distribution, with a bias towards finding peak concentrations. The resulting datum will be compared individually with the applicable DCGL(s).

For FSS of Class 3 survey units, measurement or sampling data points will be judgmental, based on professional opinion. These data points will be biased to locations considered to have the highest probability of residual contamination, with additional locations chosen to provide distributed survey unit coverage.

6.6 SURVEY DESIGN PACKAGES FSS survey unit designs will be prepared and documented in a survey design package.

Multiple survey units, having similar history, classification, and conditions, may be covered by one design package. These design packages will include survey unit maps and drawings, classification, scan frequency, data point locations, a description of unusual/unique conditions that may require deviation from standard survey techniques, and alternative techniques to be used in such cases.

6.7 SURVEY INSTRUMENTATION Table 6.3 provides a list of radiological survey instrumentation that will be used to implement the WPI Reactor FSS.

The instruments will be maintained and calibrated in accordance with WPI procedure RPP-13 "Calibration and Quality Control of Portable Radiological Survey Instruments."

Worcester Polytechnic Institute Document W19-1579-004, Rev.0 Reactor Decommissioning - Final Status Survey Plan Section 6, Page 10 of 15 TABLE 6.3 INSTRUMENTATION FOR WPI FINAL STATUS SURVEY Detector

• Sensitivity (dpmnu/10 Mt,cm 2 except as noted)'

Apliaro Static.

Make /'

Make IModel Aplcto Type.~

Moe Scanning~ Count Surface Beta contamination Bicron: DP-6A Bicron: Labtech 350 Scintillation scan and 1500 measurement Surface contamination Geiger-scan and Mueller Ludlum: 44-9 Ludlum: 2200 measurement 3500 1050 (15.5 cm2)

(for small or constricted areas)

Geiger-Ludluri: 44-40 Ludlum: 2200 Smear sample na 120 Mueller counting Area gamma Nal (2" x 2")

Ludlum: 44-10 Ludlum: 12 na na scanning Gamma NaI scanning-(05 Ludlum: 44-62 Ludlum: 12 small na na (0.5" x I")

small diameter pipe-In-situ and Alpha Ludlum 732-1 PC-sample Nal (2" x 2")

Based Gamma Ray screening na na Spectroscopy System radionuclide screening

• Alternate instruments of equivalent capability may be substituted as necessary For simplicity in application to FSS, instrument response (efficiency) will be based on NIST-traceable sources, with CI-36 (beta Eave. = 252 keV and Emax = 714 keV) being the source of choice to best represent the various radionuclide mixtures. The beta energies of CI-36 is representative of the dominant potential contaminants at WPI (i.e., Cs-137 @ beta

Eave, 195 keV and Emax. = 1167 keV and Eu-152 @ beta Eave.

288 keV and Emax. = 1840 keV) and thus will provide conservative overestimates of the contaminant mixture gross beta activities. The typical 4 pi efficiency for the 100 cm 2 DPS detectors to be used for direct beta surface contamination measurements is - 25% and - 23% for the 15.5 cm 2 GM

Worcester Polytechnic Institute Document W19-1579-004, Rev.0 Reactor Decommissioning - Final Status Survey Plan Section 6, Page 11 of 15 pancake detectors that will be used count smear samples. Actual instrument efficiencies determined at the time of calibration will be used for data conversions and documented in the FSS packages.

Detection sensitivities are estimated using the guidance in MARSSIM and NUREG-1507, Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Fields Conditions (NRC, 1997b). Instrumentation and survey techniques are chosen with the objective of achieving detection sensitivities of less than or equal to 50% of the criteria for structural surfaces, for both scanning and direct measurement. This assures identification of areas potentially exceeding the established release criteria.

Measuring instruments are calibrated on an annual basis and whenever the accuracy of the equipment is suspect. Calibration is performed using standards traceable to NIST or an equivalent standard organization. Instruments will be suitably marked or otherwise identified to indicate calibration status.

Operational and background checks will be performed at the beginning and end of each day of FSS activity and whenever there is reason to question instrument performance, in accordance with RPP-13.

6.8 BACKGROUND

AND REFERENCE AREA MEASUREMENTS Based upon the radionuclides likely to be present and the materials to be evaluated at the WPI Reactor facility during the FSS, it will not be necessary to evaluate radionuclide concentration data relative to a reference background area.

Background response (where non-radionuclide specific measurements are made) will be determined for each instrument used during the FSS.

A set of reference background measurements will be obtained for each instrument being used. Reference area and background requirements will be identified at the time of individual survey unit FSS design, as background levels have been found to be variable in some parts of the facility, due inherent shielding provided by its some of the structures and/or elevation within the building. Representative background responses will be obtained at various locations in the Washburn Shops-Stoddard laboratory building (which houses the reactor room), that have similar structural configurations as the various survey units and are far enough away from the Reactor facility so as to be unaffected.

6.9 SURVEY REFERENCE SYSTEMS A grid system will be established on surfaces to provide a means for referencing measurement and sampling locations. On Class 1 and 2 structure surfaces, a 1-m interval grid will be established; a 5-rn interval grid will be established on Class 3 structure surfaces.

Grids are assigned alphanumeric indicators to enable survey location identification. Structure grids are referenced to building features. Maps of survey areas

Worcester Polytechnic Institute Document W19-1579-004, Rev.O Reactor Decommissioning - Final Status Survey Plan Section 6, Page 12 of 15 will include the grid system identifications. Systems and surfaces of less than 20 m 2 will not be gridded, but survey locations will be referenced to prominent facility features.

6.10 SURVEY TECHNIQUES Data collected for FSS of structural surfaces will consist of scans to identify locations of residual contamination, direct measurements of beta surface activity, and measurements of removable beta surface activity. FSS of survey units with potential neutron activated materials areas will consist of gamma scans to identify locations of highest potential residual contamination, followed by obtaining samples of the structural materials, and then analysis for potential contaminant concentrations.

Additional measurements and samples will be obtained, as necessary, to supplement the information from these typical survey activities. Survey techniques are described in more detail in the following sub-sections.

6.10.1 Beta Surface Scans Beta scanning of structure surfaces will be performed to identify locations of residual surface activity. Scintillation detectors will be used for beta scans. Hand-held 100 cm 2 scintillation detectors will be used for most surfaces. Large floor areas and other large accessible horizontal surfaces are not present at the WPI Reactor facility, and floor monitors with large area gas proportional detectors may not provide an advantage over hand scanning techniques. However, they may be used optionally if they are later deemed advantageous. Scanning will be performed with the detector within 0.5 cm of the surface (if surface conditions prevent this distance, the detection sensitivity for an alternate distance will be determined and the scanning technique adjusted accordingly). Scanning speed will be no greater than 1 detector width per second.

Audible signals will be monitored and locations of elevated direct levels identified for further investigation.

Minimum scan coverage will be 100 percent for Class 1 surfaces, 25 percent for Class 2 surfaces, and 10 percent for Class 3 surfaces. Coverage for Class 2 and Class 3 surfaces will be biased towards areas considered by professional judgment to have highest potential for contamination.

6.10.2 Gamma Surface Scans The primary purpose for gamma scanning of areas and surfaces will be to identify locations of highest potential residual surface activity due to neutron activation. NaI gamma scintillation detectors (2 inch x 2 inch) will be used for these scans. This will be performed inside the reactor pool - biological shield structure (including the small exposed soil surfaces under the reactor core box area). Initial gross gamma count rate scanning will be performed by moving the detector in a serpentine pattern over the structural surfaces in question, while advancing at a rateof approximately 0.5 m per second. The distance between the detector and the surface will be maintained within 5 cm of the surface. Audible signals will be monitored for indications of elevated radioactivity.

Worcester Polytechnic Institute Document W19-1579-004, Rev.0 Reactor Decommissioning - Final Status Survey Plan Section 6, Page 13 of 15 Additionally, general area gamma scanning will be performed throughout the remainder of the facility to provide additional assurance that unforeseen sources of radioactivity are not present.

Where indications of elevated count rate are present, additional evaluations may be made using a portable NaI gamma spectroscopy system, with the detector collimated, to better define the boundaries of the affected area and the cause of the elevated radioactivity.

Locations where elevated radioactivity due to reactor originated radionuclides has been identified, further investigation will be performed (see 6.10.5 and 6.10.6).

Minimum scan coverage will be 100 percent for Class 1 surfaces, 25 percent for Class 2 surfaces, and 10 percent for Class 3 surfaces. Coverage for Class 2 and Class 3 surfaces will be biased towards areas considered by professional judgment to have highest potential for contamination.

6.10.3 Surface Activity Measurements Direct measurement of beta surface activity will be performed at designated locations using a 100 cm 2 dual phosphor scintillation detector. Measurements will be conducted by integrating the count over a 1-minute period. Where surface conditions may limit the use of the 100 cm 2 detector (such as inside floor drain basin or pipe), smaller area GM detectors may be used, with counting times adjusted to attain desired sensitivities.

6.10.4 Removable Activity Measurements A smear for removable activity will be performed at each direct surface activity measurement location. A 100 cm 2 surface area will be wiped with a - 2 inch diameter paper filter or cloth, using moderate pressure. Smears will be analyzed onsite for gross beta activity using a shielded 15.5 cm 2 GM pancake detector, sample holder and rate meter / scaler.

6.10.5 Soil Sampling Soil sampling will be limited to the floor area of the biological shield where the concrete floor was removed, which exposed the underlying soil surface. Sampling locations will be selected based on gamma scan results (to find the location of highest potential activity) or evenly distributed over the soil surface if no apparent elevated gamma scan activity is found. Samples of surface (upper 15 cm) soil will be obtained from selected locations using a hand trowel or bucket auger. Approximately 500-to-1000 g of soil will be collected at each sampling location. If reactor-originated radioactivity is detected, additional samples at a greater depth (e.g., 15-to-30 cm) will be obtained to verify that radionuclide concentrations are not increasing / or are limited to the top surface layer. Soil samples will be analyzed by gamma spectroscopy and, depending upon the resulting radionuclide mixture identified, hard-to-detect radionuclide concentrations will be scaled-in using the ratios presented in Section 4.

• *I*':5 *,*: * *,,

Worcester Polytechnic Institute Document W19-1579-004, Rev.0 Reactor Decommissioning - Final Status Survey Plan Section 6, Page 14 of 15 6.10.6 Structural Media Sampling Structural media sampling will be performed in and around the reactor pool /,biological shield to determine radionuclide concentrations in media potentially made radioactive by neutron irradiation. In general, this will be limited to the aluminum floor / wall liner within the reactor pool, the aluminum liner within the thermal column and beam port tube, and underlying concrete of the biological shield, thermal column and beam port areas. Sampling locations will be selected based on gamma scan results (to find the location of highest potential activity) or at the theoretical points of maximum activation based on media with the closest proximity to the reactor core if no apparent elevated gamma scan activity is found. Samples will be obtained from selected locations using core drills.

Approximately 2-to-3 inch diameter cores will be collected at each sampling location (complete thickness in the case of aluminum liner material, and - 15 cm long increments for concrete. In the case of the concrete biological shield, if reactor-originated radioactivity is detected, additional samples at a greater depth (e.g., 15-to-30 cm or beyond) will be obtained to verify that radionuclide concentrations are not increasing / or are limited to the top surface layer. Samples will be analyzed by gamma spectroscopy and if the material in question has been determined to contain hard-to-detect radionuclides (see Section 4), the hard-to-detect radionuclide concentrations will be scaled-in using the radionuclide ratios presented in Section 4.

6.10.7 Embedded Pipes Embedded pipes (and vents and similar items) within and around the biological shield will be evaluated for the presence of residual radioactivity. Based upon knowledge obtained while removing the non-embedded portions of these pipes and similar items during the remediation work, these items are not likely to contain any detectable radioactivity.

During the remediation work, detectable radioactivity was not found to be present in any of the items wetted by pool water (except where it became concentrated on filters and the de-mineralizer resin). As previously discussed, any potential radioactivity within these pipes would likely result from being wetted with reactor pool water, having the readily detectable 99.95% Cs-137 and 0.05% Co-60 radionuclide mixture. Similarly, removed non-embedded portions of thermal column and beam port drains, vents, booster fans and duct work were all found to be free of detectable radioactivity. Also, as previously discussed, any potential radioactivity within the thermal column related items would likely result from the presence of activated graphite dust, having the readily detectable 0.78% Co-60, 93.25% Eu-152 and 5.96% Eu-154 radionuclide mixture.

One or more techniques will be used to locate confirm that residual radioactivity within these pipes is not present. Wherever possible, direct beta radioactivity will be measured where smaller thin-window GM detectors will fit in. When possible, cloth swabs will be pulled through the length of each embedded pipe, and the swabs checked for the presence of removable radioactivity (by direct beta frisk and / or radionuclide specific NaI gamma spectroscopy counting; and, wherever accessible, the interior surfaces will also be checked for the presence of non-removable contamination by inserting a small diameter NaI detector (- 0.5 inch diameter) into the item and making a series timed gamma counts along the accessible length of the pipes.

Worcester Polytechnic Institute Document W19-1579-004, Rev.0 Reactor Decommissioning - Final Status Survey Plan Section 7, Page I of 2 7.0 DATA EVALUATION AND INTERPRETATION 7.1 SAMPLE ANALYSIS Smears for removable activity will be analyzed onsite laboratory gross beta activity. Soil, concrete, and aluminum samples will be screened onsite for gamma emitters using a portable, shielded two-by-two (inch) Nal gamma spectroscopy system. Based upon that screening, samples with the highest apparent reactor-originated radioactivity (within an applicable survey unit), for each type of media, will be selected for off-site laboratory gamma spectral analysis to determine the concentration of gamma emitting radionuclides, and depending upon the resulting radionuclide mixture identified, hard-to-detect radionuclide concentrations will be scaled-in using the ratios presented in Section 4.

Radioactivity in any screened samples found without lesser amounts of apparent radioactivity, that are not sent for outside laboratory analysis, will have their radioactive concentrations estimated by normalizing their screening results using screening to laboratory-result ratios for the samples that were sent to the outside laboratory.

The Sum-of-Fractions for the contaminants of concern will be determined based on the criteria presented in Table 4.5, with less than unity being the criterion for final evaluation that decommissioning criteria have been satisfied.

7.2 DATA CONVERSION Measurement data will be converted to units of dpm/100 cm 2 or pCi/g for comparison with guidelines and/or for statistical testing.

7.3 DATA ASSESSMENT Data will be reviewed to assure that the type, quantity, and quality are consistent with the survey plan and design assumptions. Data standard deviations will be compared with the assumptions made in establishing the number of data points. Individual and average data values will be compared with guideline values and proper survey area classifications will be confirmed. Individual measurement data in excess of 50 percent of the guideline level in Class 2 areas and in excess of 25 percent of the guideline in Class 3 areas will prompt investigation. Patterns, anomalies, and deviations from design assumption and plan requirements will be identified. As required, survey units will be reclassified, remediated, and/or resurveyed.

7.4 DETERMINING COMPLIANCE WITH GUIDELINES If all individual surface contamination values for a survey unit are less than the guideline level, that survey unit will satisfy the criterion and no further evaluation is necessary, the null hypothesis is rejected, and the survey unit meets the established criteria. If any individual value for a survey unit is greater than the guideline value, that survey unit will be further evaluated using the Sign test. If the Sign test indicates that a survey unit does not meet the established criteria, then the null hypothesis is accepted.

Worcester Polytechnic Institute Document W19-1579-004, Rev.0 Reactor Decommissioning - Final Status Survey Plan Section 7, Page 2 of 2 If all Sum-of-Fractions values for a survey unit with radionuclide concentration data are less than 1, that survey unit will satisfy the criterion and no further evaluation is necessary; the null hypothesis is rejected, and the survey unit meets the established criteria.

If the Sum-of-Fractions value for any sample with a survey unit is greater than 1, the survey unit will be evaluated using the Unity Rule Sign test. If this test indicates that the survey unit does not satisfy the criterion: then the null hypothesis is accepted.

Worcester Polytechnic Institute Reactor Decommissioning - Final Status Survey Plan Document W19-1579-004, Rev.0 Section 8, Page 1 of 2 8.0 FINAL STATUS SURVEY REPORT A FSS Report describing the survey procedures and findings will be prepared for submission to the NRC in support of license termination. The survey report will provide a complete record of the facility's radiological status and a comparison to the site release criteria.

The survey report will include survey data and overall conclusions, which demonstrate that the WPI Reactor Facility meets the radiological criteria for unrestricted use.

Information such as the number and type of measurements, basic statistical quantities, and statistical test results will be included in the report. The survey report will contain additional detail to enable an independent or third party re-creation and evaluation of the survey results and a determination as to whether the site release criteria have been met.

The following outline illustrates a general format that may be used for the FSS Report and may be adjusted to provide a clearer presentation of the information. The level of detail will be sufficient to clearly describe the final status survey program and certify the results.

Information to be submitted:

• A summary of the results of the final status survey,

" A discussion of any changes that were made in the FSS from what was proposed in the FSS Plan or other prior submittals,

" A description of the method by which the number of samples were determined for each survey unit,

" A summary of the values used to determine the numbers of samples and a justification for these values (refer to Section 6.4),

• The results for each survey unit including:

1. Number of samples taken for the survey unit.

2. A map or drawing of the survey unit showing the reference system and random start systematic sample locations for Class 1 and 2 survey units, and random locations shown for Class 3 survey units and reference areas.

3. Measured sample concentrations.

4. Statistical evaluation of the measured concentrations.

5. Judgmental and miscellaneous sample data sets reported separately from those samples collected for performing the statistical evaluation..

6. Discussion of anomalous data including any areas of elevated direct radiation detected during scanning that exceeded the investigation level or measurement locations in excess of the DCGL.

7. A description of follow-up actions and results.

8. A statement that a given survey unit satisfied the DCGL.

Worcester Polytechnic Institute Reactor Decommissioning - Final Status Survey Plan Document W19-1579-004, Rev.O Section 8, Page 2 of 2

" A description of any deviations from initial survey design and survey techniques,

" A description of the investigation and follow-up actions when the FSS fails to demonstrate that the criteria have been satisfied.

Worcester Polytechnic Institute Document W19-1579-004, Rev.O Reactor Decommissioning - Final Status Survey Plan Section 9, Page 1 of 1 9.0 BIBLIOGRAPHY

1. Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM), NUREG-1575 (Rev. 1), U.S.

Nuclear Regulatory Commission, 2000.

2.

Decommissioning Plan for the Leslie C. Wilbur Nuclear Reactor Facility (Rev 2.), Worcester Polytechnic Institute, September 25, 2012.

3.

Consolidated NMSS Decommissioning Guidance. NUREG-1757, U.S. Nuclear Regulatory Commission, 2000.

4.

Manual for Conducting Radiological Surveys in Support of License Termination, NUREG/CR-5849 (draft), U.S. Nuclear Regulatory Commission, 1992.

5.

Minimum Detectable Concentrations with Typical Radiation Survey Instruments for Various Contaminants and Field Conditions, NUREG/CR-1507, U.S. Nuclear Regulatory Commission, 1997.

Worcester Polytechnic Institute Document W19-1579-004, Rev.0 Reactor Decommissioning - Final Status Survey Plan Appendix A, Page I of 1 APPENDIX A GEL Laboratory Analysis Report For Waste Stream and Pre-FSS Samples WPI 10CFR61 and Pre FSS analysis resu (NOTE: Click on the Adobe link above for a full copy of the GEL Laboratory Analysis Report)

Worcester Polytechnic Institute Reactor Decommissioning - Final Status Survey Plan Document W19-1579-004, Rev.0 Appendix B, Page 1 of 1 APPENDIX B Gross Beta DCGL for Radionuclide Mixture at WPI Section 4.0 indicated that the reactor pool water radionuclide mixture that would be representative of surface contamination at the WPI facility. The fractional contributions of radionuclides in that mixture are, 0.9995 for Cs-137 and 0.0005 for Co-60.

Each of the radionuclides in this mixture decays to some extent by emitting beta particles.

The abundance (A) of beta emissions per decay is 1.0 for C0-60 and 0.85 for Cs-137.

To develop a gross-beta DCGL for the structural surfaces, the fractional contribution (f) of each of the radionuclide contaminants to the total mixu is divided by the Default Screening DCGL for that radionuclide, 7,100 dpm/OOcm 2 for CO-60 and 28,000 dpm/100cm 2 for Cs-137. The gross DCGL was then calculated by:

Gross Beta DCGL = fraction of beta emitters (i.e., fx A)

F (f/DCGL)

Gross Beta DCGL =

0.850075 (0.0005/7,100) + (0.9995/28,000)

The resulting gross beta DCGL value is 23,765 dpm/100 cm 2 (23,700 rounded).