ML20237B357
| ML20237B357 | |
| Person / Time | |
|---|---|
| Site: | 05000054, 07000687 |
| Issue date: | 12/05/1994 |
| From: | CINTICHEM, INC. |
| To: | |
| Shared Package | |
| ML20237B359 | List: |
| References | |
| PROC-941205, NUDOCS 9808180205 | |
| Download: ML20237B357 (86) | |
Text
I r
l l
FINAL STATUS SURVEY PLAN AND REPORT CINTICHEM, INC. DECOMMISSIONING PROJECT (RESEARCH REACTOR AND RADIOCHEMICAL PROCESSING FACILITIES) e, l
O 9908190205 941205 PDR ADOCK 05000054 y
[
SECTION TITLE PAGE #
1.0 BACKGRCUND INFORMATION 1
1.1 Reason for Decommissioning 1
1.2 Management Approach 1
2.0 SITE DESCRIPTION 3
2.1 Type and Location of Facility 3
'2.2 Ownership 4
2.3 Facility Description 4
2.3.1 Buildings Decommissioned 4
2.3.2 Grounds 7
3.0 OPERATING HISTORY 12 3.1 Preconstruction 12 3.2 License History 12 (D
q j 3.3 Processes 13 3.3.1 Mo99 Production 13 3.3.2 Reactor Operation 16 3.4 Waste Disposal Practices 16 3.4.1 Solids 16 3.4.2 Liquids 16 3.4.3 Gaseous Effluent 17 3.5 Incidents and spills 18 3.5.1' Hot Cell Exhaust Duct 18 3.5.2 Reactor Primary Coolant Leaks 19 4.0 DECOMMISSIONING ACTIVITIES 20 4.1 Decommissioning Objective 20 4.2 Pre-Decommission Radiological Characterization 20 l
4.2.1 Contaminated Structures 20 C
f I
)
SECTION TITLE PAGE #
1.0 BACKGROUND
INFORMATION 1
l l
1.1 Reason for Decommissioning 1
l 1.2 Management Approach 1
2.0 SITE DESCRIPTION 3
l 2.1 -
Type and Location of Facility 3
2.2 Ownership 4
l 2.3 Facility Description 4
2.3.1 Buildings Decommissioned 4
l 2.3.2 Grounds 7
3.0 OPERATING HISTORY 12 3.1 Preconstruction 12 3.2 License History 12
,O().
3.3 Processes 13 3.3.1 Mo99 Production 13 l
3.3.2 Reactor Operation 16 3.4 Waste Disposal Practices 16 l
3.4.1 Solids 16 3.4.2 Liquids 16 3.4.3 Gaseous Effluent 17 i
3.5 Incidents and Spills 18 3.5.1 Hot Cell Exhaust Duct 18 3.5.2 Reactor Primary Coolant Leaks 19 4.0 DECOMMISSIONING ACTIVITIES 20 4.1 Decommissioning Objective 20 4.2 Pre-Decommission Radiological
( )
Characterization 20 4.2.1 Contaminated Structures 20
_4.2.2 Contaminated Systems and Equipment 22 4.2.3 Activated Components and Structures 24 4.3 Decontamination Procedures 24 4.3.1 General 24 4'. 3. 2 Decommissioning Organization 25 4.3.3 Final Survey Organization 25 5.0 INTERIM SURVEY PROCEDURES 26 5.1 Intent of Survey 26 5.1.1 Scope 26 5.1.2 Purpose 26 5.1.3 Intended Use of Interim Surveyed Area 27 5.1.4 Use of Interim Survey Data 28 5.2 Remediation Work Prior to the Survey 28 O'
5.2.1 Hold-up Tank / Primary Coolant Storage Tank 28 5.2.2 SK Storage Tanks 29 5.3 Interim Survey Overview 29 5.3.1 Survey Objectives 29 5.3.2 Identity of Contaminants 30 5.3.3 Organization and Responsibilities 32 5.3.4 Training 33 5.3.5 Laboratory Services 33 5.3.6 General Survey Plan 33 5.3.7 Tentative Schedule 34 5.3.8 Survey Report 34 j
5.4 Survey Plan and Procedures 34
'N 5.4 1 Instrumentation 34 (V
g 5.4.2 Survey Plan 36
5.4.3 Background Level Determinations 40 5.4.4 Sample Analysis 41 5.5 Data Interpretation 41 5.6 Report 42 I
1 i
{
i
i A,s.
. LIST OF FIGU'RES 2.1 Decommissioning Site Location 2.2 Site Plan 2.3 Fractor Building Plan and Elevations 2.4 Hot Lab Plan and Elevations 2.5 10 Mile Radius Site Map
- 2. 6.
Site Plan - Topography 2.7 Bedrock Contours 2,8 Bedrock Fracture Zones 2.9 Overburden Ground Water 2.10 Bedrock Ground Water 2.11 Surface Water Drainage 3.1 Mall Tanks, 001 Outfall tv/
3.2 Reactor Building Exhaust Ventilation 3.3 Hot Lab Exhaust Ventilation 3.4 Hot Cell Exhaust Header Leak 4.1 Plan View of Contaminated Areas in Buildings 4.2 Plan View of Surface Soil Contamination 4.3 D & D Organization Chart 5.1 Organization Chart for Interim Surveys 5.2 HUT Tank and Storage Tank Survey Area l
l 4
L l
i LIST OF TABLES 4.1 a. b. and c.
Summary of Structural contamination l
4.2 Activated Reactor Components and Structure f
5.1 Cintichem Site Soil Release Criteria 5.2 Overview of Major Activities and Tasks 5.3 Instrumentation Used for Survey Activities l
5.4 Listing of Survey Areas / Units and Analysis Frequency l
1 O
. LIST OF ATTAC8KENTS-
- Attachment A' Reg. Guide 1.86 Release Criteria Attachment B N.Y. Code Rule 38 Table V (1994)
' Attachment C-Soil Criteria Attachment'D Letter to NRC:from J. J. McGovern dated August 30, 1994 -
SUBJECT:
Request for Modification of Regulatory Guide 1.86 Surface Contamination Limits-for Fe-55 and H-3 per SECY-94-145 lO
.m L
O l
INTERIM TERMINATION SURVEY PLAN AND REPORT
(
l.0 BACKGROUND INFORMATION 1.1 Reason for Decommissioning The Cintichem, Inc. research reactor and radiochemical processing
{
facilities were constructed during the period from 1957 to 1960.
At the time operations ceased in 1990 the facility had been operated for 30 years.
Two basic defects in the plant were l
discovered in early 1990 that effectively rendered the facility functionally obsolete.
Concurrent with the discovery of these j
l
- defects, the sale of a radiopharmaceutical business that was partially conducted in this f acility was under negotiation.
In view of these circumstances it was decided to terminate the radiochemical processing operations permanently and to proceed with decommissioning according to the requirements of Title 10 CFR Parts 50 and 70 and NYS Code Rule 38.
1.2 Management Approach The regulations pertaining to the decommissioning of the reactor provided three options for satisfying the decommissioning requirements and minimum acceptance criteria for terminating the facility license.
The regulations pertaining to decommissioning special nuclear materials facilities were less specific with regard to perf ormance criteria.
The agreement-state regulations A
pertaining to the decommissioning of the by-product material
\\
)
facilities provide minimum acceptance criteria for releasing the facility for unrestricted industrial use.
Cintichem has chosen the primary decommissioning objective of terminating all licenses and allowing unrestricted future use of the site.
This option had been chosen because it provides the best opportunity for returning the site to productive use.
Under the current local zoning ordinance that would be industrial use.
Cintichem managed the decommissioning project with direct involvement of TLG Services personnel holding some. key management positions within the decommissioning project organization.
TLG
- Services, Inc.
had extensive experience in decommissioning nuclear facilities.
The Cintichem decommissioning proj ect has been funded by the parent company, Hoffmann-La Roche, Inc..
Funding was assured through the parent company guarantee method prescribed by NRC regulations.
1 Quality assurance for the final survey process was prescribed in j
several documents that are listed below and included in this report by reference.
Cintichem Decommissioning Pr oj ect Quality Assurance l
Plan / Manual (POL-009 dated March 1991) d i
)
l 1
l JJM/83.94B-Page 1
1 l
INTERIM TERMINATION SURVEY PLAN AND REPORT Cintichem He al th, Safety and Environmental Affairs Department Measurement QA Manual s.
Detailed Health Physics Department Q.C. Procedures Detailed Health Physics Department Survey Procedures.
All surveys have been performed according to detailed implementing procedures that were issued according to the referenced QA plans.
All personnel working on the decommissioning project have been trained in basic radiation protection and industrial safety.
Specialized training has been provided to all employees as they were assigned specific tasks.
Records of training have been maintained per the requirement of the QA Plans and implementing procedures.
l
.fX i
n.U l
JJM/83.94B Page 2 l
l J
INTERIM TERMINATION SURVEY PLAN AND REPORT i
i (n) 2.0 SITE DESCRIPTION v
2.1 Type and Location of Facility l
The Cintichem site consists of 100 acres of land.
About 30 acres i
of the site is developed while the remainder consists of mountainous terrain.
Figure 2.1 depicts the site in plan view with the location of all of the buildings.
{
The six principle buildings at the plant site are:
Building 1 - Reactor Building 2 - Hot Laboratory (adjoining the reactor building)
Building 3 - Engineering, Maintenance and Warehouse (Currently used as base of operations for the decommissioning project.)
Building 4 - Administration and Pharmaceutical Manufacturing Building 5 - Utilities Services Building 6 - Class A Waste Processing and Storage C'
The site is located in the town of Tuxedo which is in the
()1 southeastern corner of Orange County, New York.
The precise location is on the west side of County Route 84 (Long Meadow Road), one mile south of the intersection of County Route 84 and NY State Route 17A.
Currently, only decommissioning operations are being conducted on site.
Prior' to decommissioning, radio-chemical and radiopharmaceutical production operations were conducted.
The radiochemical production ope ra tion predominantly involved the separation of Mo-99, Xe-133, and I-131 from the fission products of U-235.
Other short-lived radiochemical were produced by neutron activation but in significantly lower quantities compared to the processing of fission products.
The radiopharmaceutical production operation consisted of manufacturing multi-curie Mo-99/Tc99m radioisotope generators and packaging Xe-133 gas for use in diagnostic nuclear medicine.
The radiochemical operations were conducted in the reactor and hot laboratory (buildings 1 and
- 2) and were terminated in February 1990.
The radiopharmaceutical operations were conducted in building 4 and were terminated in June 1993.
i This final survey report pertains only to the radio-chemical production facilities in buildings 1
and 2
and associated structures, the waste processing / storage in building 6 and the affected and unaffected areas of the site (refer to Figure 2.2).
n
}
The decommissioning and survey of the former radiopharmaceutical wj JJM/83.94B Page 3 C
l INTERIM TERMINATION SURVEY PLAN AND REPORT l
iq I
pr'oduction facilities are regulated under the NY State Code Rule 38 and will be conducted under a separate plan and reported separately to the NY State Department of Labor.
2.2 Ownership l
The history of ownership of the site is as follows:
I 1
Owner Parent Company Time Period Cintichem, Inc.
Hoffmann-La Roche, Inc.
May 1990 to present
]
Cintichem, Inc.
Medi-Physics, Inc.
Aug. 1985 to May 1990 Union Carbide April 1981 Sub. B Union Carbide Corp.
to Aug. 1985 l
Union Carbide Corp. None 1960 to April 1981 l
All of the above companies are privately held corporations except l
Union Carbide Corp. which is a public stock company.
.Q 2.3 Facility Description
%J 2.3.1 Buildings Decommissioned Building 1, the reactor building, measures 70' wide x 92' long x 57' high from the lower floor level.
It is constructed of reinforced concrete that is set into an excavation from the side of.the adjacent mountain.
The west side is against the rock face of the excavation whereby the wall was formed against the uneven face of the excavated bedrock that extends to the full height of the building.
The north side abuts building 2 and about 1/2 of this wall is below grade, backed by concrete fill to bedrock.
The south side is similar to the north in that the upper portion extends above grade while the lower portion is also backed by concrete fill to bedrock and some soil.
Much of the fill behind the south wall contained cylindrical storage ports that opened to the lower level of the reactor room.
The east side is a formed wall from top to bottom that is about 1/2 above grade and the lower half is backfilled with soil and loose rock from the l.
original excavation.
There are two openings at the base of this wall that provide access to underground structures extending east of the building.
One of these underground structures is a tunnel that opens to the roadway at the lower level of the site.
The other structure consisted of two reinforced concrete chambers that extended beyond the southeast corner of the reactor room.
The inner chamber housed the reactor primary coolant pumps, heat
,p exchangers and deionizer and also some electrical controls for q,)
l L
JJM/83.94B Page 4 l
INTERIM TERMINATION SURVEY PLAN AND REPORT i
O auxiliary equipment.
The outer chamber served as the primary coolant hold-up tank (refer to Figure 2.3).
The roof is composed of reinforced concrete with built-up water proofing on top.
4 Auxiliary structures associated with the reactor included a
100,000 gallon primary coolant storage tank and a secondary coolant forced convection cooling tower that were located south of the reactor building (refer to Figure 2.1).
The reactor was immersed in a pool of demineralized water that was located in the center of the reactor room.
The reactor pool consisted of two sections, namely the stall and the pool.
The pool was formed of reinforced concrete with an embedded steel j
liner.
The nominal dimensions of the pool structure were 20' wide x 40' long by 34' deep.
The reactor core was located in the stall section during operation.
Stainless steel and aluminum piping (10" diameter) penetrated the pool. and stall floors and lining and ran under the reactor room floor to the pump room and hold-up tank in the southeast corner of the building.
The concrete walls of the stall section of the pool contained a network of small diameter piping and ducting that contained pool water and ventilation air for reactor auxiliary equipment.
Two small hot
- cells, a
radiochemical laboratory, a counting laboratory, offices and the reactor control room were located on p
the upper level of the reactor building.
Health Physics counting laboratories, a lavatory and an electronics instrument shop were located on the mezzanine.
Two counting rooms and apparatus for an isotope production loop were located on the lower level.
Building 2,
the hot laboratory, is constructed partly of reinforced concrete and partly of steel-frame and masonry (refer to Figure 2.4).
The west side is constructed of reinforced concrete that essentially rests on bedrock.
The south end of this wall forms the outer wall of the canal that connected the 1
reactor pool with the hot cells and it is half below grade.
The l
north end of-this wall is also half below grade where it forms the west side of the waste evaporator room and the filter room.
The middle of this wall is on grade.
The south side of this building is constructed essentially of reinforced concrete and it abuts the reactor building.
The north and east sides are of steel frame and masonry construction whereby the steel columns are supported on concrete piers that extend below grade to various depths (some to bedrock, some to virgin soil).
The roof consists of steel decking on steel beams with built-up roofing on l
top.
The original roof was supported by columns that were embedded in the hot cells and rested on the hot cell foundations.
l As part of the decommissioning
- process, a
steel frame was constructed over the building and the roof load (plus the second floor load) was transferred to the overhead girders of this f rame to enable removal of the original roof support columns during the demolition of the hot cells inside of the building.
The load on d
this steel f rame rests on new footings that were placed along the l
east an.d west sides of the building.
l JJM/83.94B Page 5
INTERIM TERMINATION SURVEY PLAN AND REPORT A waste storage cell room extends to the north of building 2.
U The foundation of this room is a concrete monolith containing 100 cylindrical storage cells arrangad in a honeycomb-like array.
The dimensions of this concrete monolith is 40' x 40' x 8' and it is set directly on bedrock.
The storage cells were interconnected by exhaust ventilation ducts.
The upper chamber of this storage cell room is constructed of steel framework and masonry walls with a steel decked and built-up roof.
The previously mentioned canal at the southern end of this building connects the reactor pool with the hot cells in building 2.
It is constructed of reinforced concrete, it is 4' wide and 12' deep and it extends along the west side of the hot laboratory building from hot cell 1 to the reactor pool (approximately 100').
It was filled with reactor primary coolant water and it l
was used to transfer radioactive materials between the reactor and hot cells.
Five major hot cells were located approximately in the center of building 2.
The cells were constructed of high density concrete.
Networks of piping, conduit and ducting were embedded in the I
walls and floors of these cells.
Large vitreous tile exhaust i
ventilation ducts penetrated the lower walls and foundations of the cells and connected to a ventilation header that was located in soil under the cell ope rating area floor.
The header ran
/Q underground to two filter houses at the north end of the i
i V
building.
A contaminated liquid waste collection, storage, evaporation and ion exchange plant was located in two basement rooms and on the main floor level of this building just north of the hot cells.
A clean condensate water storage tank that received water from the liquid waste evaporator system was located on the north end of the second level.
A machinery space for supply ventilation and domestic services was located on the south side of the second level.
Three radiochemical laboratories that were used for l
electroplating U-235 irradiation targets were located on the west i
side of the second level.
l A machine shop and an electrical equipment control room were located on the south end of the first level of this building.
The exhaust ventilation equipment for the reactor and hot laboratory was located in the northwest corner of the first level.
This equipment was connected to a 3' diameter by 400' long steel exhaust stack that ran to the crest of the mountain west of the building.
The stainless steel lined decon room was l
located adjacent to the exhaust ventilation equipment room and it was primarily used for interim storage of contaminated equipment that was removed from the hot cells, n
Building 6 is a building, slab on grade, constructed of masonry
(\\~}
' walls and a wood framed roof with asphalt shingles.
It was used for storage of Class A waste packaged for shipment to a disposal facility.
JJM/83.94B Page 6
INTERIM TERMINATION SURVEY PLAN AND REPORT
(
I 2.3.2 Grounds
\\_)
Plant Area and Topography The Cintichem site is located within the Town of Tuxedo in Orange County, New York.
Tuxedo is in the extreme southeastern corner of Orange County approximately six miles north of the New Jersey state line.
The plant site is located on 100 acres of land in an area known as Sterling Forest and is about 3h miles northwest of the village of Tuxedo Park.
Features within ten miles of the site are shown in Figure 2.5.
The plant is situated on the eastern side of the 100 acre property, on the eastern slope of Hogback Mountain at a nominal elevation of 800 feet above sea level.
The plot plan of the site showing topographical features is presented in Figure 2.6.
Demography The plant is located in a thinly populated area.
The closest occupied off-site area is the Laurel Ridge housing development which contains l'12 houses at a minimum distance of 1,100 feet east of the reatcor building.
A second. development, consisting of 27 houses an/ called Clinton Woods, is located 3,200 feet to the north.
The.e are no other housing developments within 1.5 miles.
7 The following table shows the population distribution, in 22.5*
compass sectors (The north sector is centered on true north but includes 11'15' on either side of true north, a total of 22.5*.
Likewise, all other sectors embrace an arc of 22.5 ),
out to 5 miles.
The table indicates the most heavily populated areas to be to the north, south-southeast and west-northwest of the site within the 5 mile radius.
The population density of these sectors is due to housing developments in the towns of Southfields, Tuxedo and Monroe /Chester respectively.
Population within Sector 5 miles
- N 614 NNE 346 NE 108 ENE 187 E
107 ESE 330 SE 132 SSE 3,878 S
i
,s WSW 125 W
68 L
\\_)
WNW 878 NW 192 NNW 190 JJM/83.94B Page 7 i
b-
INTERIM TERMINATION SURVEY PLAN AND REPORT m
)
- Based on 1980 Census of Population and Housing.
Little w'
development has occurred within five miles since 1980.
Geology Overburden Deposits The site is underlain by overburden deposits comprised of.
unconsolidated glacial
- deposits, man-emplaced backfill and weathered bedrock.
Tne original site topography was extensively altered during construction of the facility.
Large volumes of crystalline bedrock comprising the east slope of Hogback Mountain were blasted and removed to facilitate construction of the foundations of Buildings 1 and 2.
Topographically-lower parts of the site were backfilled with the rock rubble derived from the blasting.
Buildinga 3 and 5,
and the paved parking area in between were constructed within rubble-backfilled areas.
The native unconsolidated material consists predominantly of silty sand and gravel with boulders deposited by past glacial or glaciofluvial processes.
This material thickens toward the east, and is exposed at the surface east and northeast of the paved areas on site.
Based on the monitor-well installation logs, as well as pre-construction
- borings, these deposits attain a
thickness of about 35 to 40 feet beneath the eastern part of the C;
site.
l Bedrock Geology Competent bedrock occurs at depths ranging from grade exposure to more than 40 feet below grade (bg).
Bedrock crops out behind buildings 1 and 2,
and the bedrock surface slopes toward the east.
A map of the bedrock surface was prepared based on data obtained from a monitor-well installation program in 1990 - 1991, as well as pre-construction diagrams and borings (Figure 2.7).
The bedrock geology in the vicinity of the site was studied by Hotz (1952) and further discussed by Offield (1967).
The bedrock is comprised of pre-Cambrian foliated metamorphic
- rocks, including amphibolite and hornblende-feldspar gneias.
The structural relationship of these units is not well understood, although the site appears to occupy the western flank of sn
{
overturned syncline whose axis trends to the north-northeast.
Isachsen and McKendree (1977) indicate that a fault strikes north-northeast roughly parallel to Long Meadow Road immediately
)
along the eastern edge of the Cintichem property.
The direction of fault displacement is not reported.
Stereo aerial photographs of the site vicinity in 1940 and 1951 that were taken prior to the construction of the Cintichem
,. m
- facility, exhibit evidence of fracture traces within the
{
r
)
boundaries of the Cintichem property which may be the fault J
identified by Isachsen and McKendree.
The fracture trace evident l
JJM/83.94B Page 8 i
L
INTERIM TERMINATION SURVEY PLAN AND REPORT OQ in the aerial photos trends northeast, and extends from a swamp near the southern end of the south parking area to the vicinity of the retention pond east of building 3.
The strong " layered" nature (i.e.,
regional foliation pattern) of the bedrock is visible on all of the stereo aerial photographs.
The strong visibility of the foliation pattern is a result of differential weathering of the bedrock surf ace.
Generally, the more resistant bands of bedrock f orm higher elevation features, such as ridges and upland protuberances into the Indian Kill Reservoir, while the less resistant bedrock forms depressions, auch as valleys and large gullies.
The foliation pattern visible in the aerial photographs trends to the north-northeast, and dips to the east.
Site Bydrogeology Ground water occurs in both the overburden deposits, comprised of unconsolidated rock rubble, native soil materials and weathered bedrock surface (overburden aquifer),
and in the crystalline bedrock (bedrock aquifer).
Ground water in the overburden aquifer occurs at depthe ranging f rom 1.5 to 28 feet bg (below grade), and in the bedrock aquifer at depths between 13 and 60 feet bg.
b A comparison of ground-water elevations measured at monitor wells
'V completed in the overburden and the bedrock indicate vertical i
hydraulic gradients between the overburden and bedrock aquifers.
In the vicinity of building 2,
there is a negative (downward) differential vertical hydraulic gradient between the overburden and bedrock aquifers in the area along the upper roadway with a downward vertical flow potential f rom the overburden aquifer into i
the bedrock aquifer.
l In
- contrast, to the east of building 3,
monitor well data indicate a
positive head differential hydraulic gradient indicating upward flow potential from the bedrock aquifer into the overburden aquifer.
Based on the ground-water elevation contour maps constructed for the overburden and bedrock aquifers (Figures 2.9 and 2.10), the interpreted transition zone between downward and upward vertical hydraulic gradients occurs along the topographic slope between buildings 2 and 3.
To the west of this l
line, the overburden aquifer recharges the bedrock aquifer; to the east of the line, the bedrock aquifer discharges into the overburden aquifer.
The actual degree of ground-water exchange between these two aquifers would be governed by the relative i
l hydraulic conductivities (ground-water transmissive capability) 1 l
of each of the respective aquifers.
In addition, the location of the transition zone would be expected to shift upgradient or downgradient in response to changes in the bedrock aquifer p
recharge rate.
N.)
I JJM/83.94B Page 9
INTERIM TERMINATION SURVEY PLAN AND REPORT Overburden' Aquifer Ground water in the overburden aquifer exists under unconfined or
" water-table" conditions (i.e.,
ground-water surface exists at atmospheric pressure).
The overburden aquifer across most of the site ranges be twee n 2 and 10 feet thick.
To the east of the paved areac the overburden aquifer thickness increases f rom 10 feet to more than 30 feet.
Ground-water elevation measurements collected from wells i
completed within the thinly-saturated overburden aquifer (the areas under buildings 1 and 2) indicate highly variable water levels over time.
Ground-water elevations in the more thickly-saturated zone (areas east of the paved areas) are generally less variable.
Based on the ground-water elevation contours, the direction of ground-water movement in the overburden aquifer is toward the east-northeast, in general conformance with the surface topography.
The lateral hydraulic gradient in the surficial aquifer ranges between 0.07 ft/ft (vertical feet per horizontal l
foot) between buildings 3 and 5 to 0.20 ft/ft along the uppe r roadway east of buildings 1 and 2.
Bedrock Aquifer Ground water in the bedrock aquifer occurs primarily within and flows through bedrock fractures.
As a result, the orientation of the bedrock f ractures can locally control the mode and direction I
of ground-wate r movement.
Fracture density in bedrock generally decreases with depth.
At the Cintichem site, ground water in the l
shallowest bedrock f ractures occurs under water-table conditions and appears to be in direct hydraulic communication with the l
overburden aquifer.
Ground water in the deeper bedrock fractures l
appears to exist under water-table to confined (ground-water surface under higher-than-atmospheric pressure) conditions.
Ground-water elevation measurements obtained f rom wells completed in the bedrock suggest that the bedrock aqui fe r consists of a series of interdependent f racture systems associated with each of the larger fracture zones that transect the site (refer to Figure 2.8).
Large differences in ground-water elevation and gradient may occur between different fracture systems.
For example, ground-water elevations and gradients in the fracture system west of FZA appear to be distinct from those encountered in the fracture system east of it.
An interpretive map of ground-water elevation distributions in the bedrock aquifer is presented in Figure 2.10.
The hydraulic gradient within the fracture system east of FZA is p
0.05 ft/ft.
The potentiometric surface (surf ace to which water Q
would rise due to hydrostatic pressure) within this aquifer zone slope s to the north-northeast, which is consistent with the expected regional ground-water flow direction.
The hydraulic JJM/83.948 Page 10
l INTERIM TERMINATION SURVEY PLAN AND REPORT I
l
[N-gradient -in the fracture system west of FZA has not been calculated.
However, based on the predominant bedrock fracture pattern, ground water in the western bedrock zone is likely to l
parallel that in the bedrock east of FZA.
l Outfalls l
Surface water drainage flows generally from southwest to l
northeast both on the surface and through a network of storm-water drainage piping into a small retention pond (refer to l
Figure 2.11).
Water in the pond may be discharged by pumps to outf all 001 or by gravity to outf all 002.
Outfall 001 consists of an 8" diameter industrial liquid waste pipe that carries waste water from liquid waste holding tanks (10K tanks and SK tankers) l and discharges it into the Indian Kill stream downstream of the Indian Kill Reservoir.
Outfall 002 consists of a surface stream that flows from a gate valve at the retention pond outlet and discharges into the Indian Kill Reservoir.
Currently all water from the retention pond is discharged to outfall 001.
When storm water flow exceeds the capacity of the 8"
diameter industrial liquid waste pipe, storm water from the southern part of the site is diverted away from the retention pond at S-7.
i O
- V I
i I
l 4
q l
l JJM/83.94B Page 11
eb
~
asu r
CC T
I 3
}{4
~
r uu 84 t
P*
8 4
p
=
~
h
~
8s,s
/
?
h eg p
k tP 4
?
3 f
?
,., 7, 3y d
m 4N,Q; N
O Q,,,
p I
4P T
U A
E
%5g 4^
C O
f L
se y
=
H D
E T
~
r I
S
.~
9 lh%
Ow, SE
. w R
G l'
p=w ^
/[/
, u C
N a
A
,. v I
a 2
0 N
~e 9
G y
I x
D g
gP R
s M
P I
PA M
M f
D C
E D
O.
/
ao m e
gy y2.
p c
i I
J
(,
\\
}
Q
}
l l
APPRoc 100 Atd b
1
\\
\\
\\
l 1
/
\\
\\
t
\\
i C,.
a
/
N '*
g re a
< o,7 4
9 j
.,,s,x
,n,.-
- 0 D
N,s CN"t DN"
\\
is. e
} ;M&"
n, su e.=e i- -
/%
., i 9;4-e e,
5 y
- 5 f/
f
\\\\\\\\\\e\\\\\\\\\\\\\\,\\
\\
\\
k O
I 4
(_'__-------
e l
M.NN l
l APERTURE CARD l
b VBb & Q Apertune w b
I I
j
,,, ~. - ~. < -
g
(
f FIGURE 2.2 kI 3_
- TA 7+50V I
I LOCATIDh5 VHERE BIASED TINAL r
SURVEYS SH/A t. SE CONDUCTED
- %"M=
/
NORTH g msma+: 4 = w -
~r d*
'e
_l _
ITA 2*DOV v
l 9/2)/93l ADD NAMC INDIAM MIt.t. RCf. L RCH3VC HATCH FROM POND j A APPR.
D..T E l DESCR'PTION l P.EV cAo ru J cw si 2353 REVISIONS 2400 ST.
j co*wCJG Ig/25,2E CINTICidEM, INC.
p om:w
- ,.a mee.
~
uses,ogggg,jn.
r a ru s:s. va. e nn w.sr llll@
c-o FINAL BIASED 53' SURVEY PLAN C 136C2353 v4LEl' = 200Tfl Sacri
.a. e.
p;"
Q 4
n e
,4 m
_r r
4 r
A r
a
/
t R
,4 u
N 3
I m
n O
4 D
o T
,4 T
a C
c A
W A
,E r
V E
n
=
T f
N
,E n
w I
7 b
R
,A L
t e
E D,
T L
, A E
- h wne au
. E
, S i
r,
{ l-
,4 v
a
, a "u-ra>- -
r x~
,4
=
r m
=
g,
a Ew.
n
% w_
n
,5
-.A r
-m-
-, +
_e A 7." a
=
=
n=
'M
.I w g, 7e
_,w
'c
]
m-s Ne-l I
~
n =s _
.i x
1 7-j sa N
( 4 o
n
=
E sr va r
a av c
=
n
=
-s n
m c
a e
s.
=
v l
ue u
s
]
nAI Y iu s
I T
c aL s
t
\\a3 V
l-ss mR a
r r
aR y
n nT u
m!L
^i-
_m tr t
l c
a e
a w
i l
p ie t
eA c
O a
a a
B c
r
=
r
=
a e A.
a
=
sea
=
N L',
,i!
U M
,s
/
NN q
m
- B _8 u
2 3
1o 7
H n
s' 0
T e
i,,i ir i
7 R
U c.
.a pi u
D sa cn N
c g
s g
Wa-
,,.i4,
. ~
o
- .4 gI U
s
[
C
,C C
g R
A
/
I R
U I
D Ye#
I y, K;.
IL*
~
D 7P C
F U h R
- 3
$t gl IE '0 T
W="" i A
N M
E s
9 A rj t
j I1 E
N B
AL ZE E i
~
Z V
r U E
0 R-8 Os M
A L
D t LFr 8
Ut L
a 8
Cv D
- u B.
sa a
D P
U L Tt r
E"1 j,g, e
e J
Mm i
t A
Ex n= 5 Bu
- i r
B
_2 N g 7e B
A D
C' c
i p-
'E
)J i
l
[
l l
I!
I Tr0 r
q j
A
E8
- 848r, rra-i da s
-i t TN A
-I m
2 E
n m
g z.
.f.a l
5 b
'~'
fg
.. 5
~;l
~'
.=5 j
g$ i
--2 5
.$.:=. g
- 'g
~
m.
J'
.i::p s
s g
g 4
o
', '*i C%
C i
~
ga 1
g j
dEdi
^
r y,. u -
i l-o r";7im!
l j 8
a s
I jggj+jgg80%iPp2 11 3
c :
o-cse 1-
- t d
au i
9.-
a p
h=
=
l
\\x -)
l I.
.+
t,gg '%
}
i
.~.'p I
d.Nhs,. I i
e
'7 9
l ~g~~
.g? = = = g" g
l g
-a g
g a
J_],]g=_h__%
a
-=
k a
91 g
a -li
- E-
,i 9
l u
3 n
C g
.M l ~e l,
W!
h T6C i,
a d
lENIC h
_e
_ [-]
g 3
e ig.
=
dl g
- - 6#,
I, d
D
- -) l 3
5 i-a i
i y
c
,i d
,. k !I_
E l
j I
1 I in a
s 7
' ds.1' j.,
L_
8 4
i h
, e 7
t (sy 1
i s
S 91-, l rt_Ai
'i-e g
i
' _. s n
n r
~
1 L
F
~
o l
i h
Vt
'au i
er
' }..
4 lg f
y-
-f, II###
W
)
.y
~
]. '
si e
/s..
- ~='
r j
,, t g
g f
Y Qi
\\
h [/
4
,/?
s,,N.,
'~'
g f
f'-
'N
[/
/
./
N 4
w-
,1 m
s 2 b
l I-I r
A g
7 d
[
N
/
wm 4
4 x..
f
[$
~
j t
t
(
i 1
A x
Features Within 10 Miles of the Site Figure 2.5 0
i w
O e
O m
e
})
g l
.)
.I.
. Y h h h i\\
ILkWAL %
B@%WW:
%vg/<oy; gr ( g7 s
ia!!!
l r
%,*'/
seece L
q
/
'I t,,,
O
~
N y
- id i
s.
ps.~
'e*
x
\\
H:e.'
i
}.
't
-y tl' n
J\\ W@(y (
~~
\\e',;y,,
~
e
}(
'{..
3
![68 e
D:
=
.ir g"
- m) t 1
c z
D
};i:.
s !!t s
O
\\;
u k
z
- i m
. a x O
U o
.-m
@!;lifl...-,.l.\\- a U
i /4... t/iii..
I l
i j}
Ai.
V ai a
g.45 i
g S:
gl a _,.
~~
N
.m y
1 s..sgw x
n cs
\\
- 1. -
)
o
\\.
O.0..
w we"..
bI."u rcui.
~
l ii!$-
i i
Q 2
_n ',-
e s
nT e
I D
w a
(W i 2 mn p
Ni
. Q, :=-
w
- r. :;..a y
n:
1 I
s 1 9 i~
m a
z g
s i
--J 5-
-.- >L T, $1
/
)
n.-
' 41 ill9 9
I 8
1 l/~Ib!gp-Amagi e
1 a
=
/
2 f:f 4_
,4-i z
e:
a t
f.
=-
(
E]*
3 i j N
i MM9f)b C
~
i a =
m J
{
i
=
Aj s t y d e/
' ':Li!
o t
uJ t
w.
/
q.f i i si a N
g l4
/^\\
(
l
.s s
as A
j
%I I
- =
i_-
.N ll*~ I x
l c) 1
R,i LI L
I K
[:l N
A
- .i D
N
- _ i-
/
.!.. :.f I
l
&.l/4.,ClhI-Q
- ~
i,4.*.
[l:
o
. / '-
l
~..
- .a!. lNi ::a'l\\; d
../..
i i
- p,,' j.*
i w
(
l' a.
> j%:W~' *.
.g \\ :
.W:Q. l
- l
![ ::
/,.
E j.g *:
f
.A~5
- g
- a 4 :f:!3.g sX r.u
\\
~
- \\.
- ~
a ;,r". & =;
? ;
!. b '
(,,
T. ;
S
%*r.3 gY^
c.- Q y 4
s' E
i>
b",,
'g.
/
N
- @m a$_ j
/
9*/.
I
,g,,
O
=~
s
~
whia /k%h[j t
Ol Z
- g'-
> a I.f w r'
c.
~~..
W5pk~
Q f1 L9.,'?
E
'(.-
t
?
R
,Ni3.!.
2(
U h
~
T
_55 E
3p c
u: Y__ +wi. d.,5
~
C G
=
yl f.
A g-i s5: % n {!,g h j :' f,
K 1g_~ Q R
4 D
= _ '
L F
r$
I ac = Y i U
="_sgyuWD 3: :G:m.: d!,.
e p-C
,~9~-
.sy&g.pke,~lC f
.'g.
B O
, 4:c R
w f,,
/C.:
D
%.yv
,N
-m ~ '_. v $
E h >:. m y&
s A g-j s.- '
e.
,t t
s"'
'-lln?hs..N U
? 4 g..' T.
h-6 m 1Y
.N 8
i, l.
. :, ?
2 l.
\\Nj g,r.
r
=$n q%t t
E
-*2 s
R
^
7 G
=i$
., s%l.
~I N
y,i.v:Q\\-Q n; G
D aL
- 2ly$ r I
mI F
G U
Qs D
k h
o t
i E
h, %' '., r
'A/
x N
O
!" l(
- .;,4 p/ *i C
IO N R UR Z
)
$,y.n
.T
(~
)
ED D
s G
N EE RE MI y
N
\\
D c-MU
.y!
I T
I
~
ZE CE s',,,& ~*. s% >' !.:
5 ll A
T F
AF A
C E
T N R N u
O I
R O
L M
U I F
I
..r j E
E h~~*' s_p*
Mg.
L T
IL C
N T
T
~
e m
i.'
!. '! llk R
R g "-
N E
E L
IO A
E
~
L A
H H
,m i
E E
E R
R d~.'.x?.;i..b $ d' W
F W A
)e P
L W
R I
T D P
4 l
.f[, W:.
P o
G A
F D
G
.[\\::
R N
O N E A E 1
Y E H H
D O
IL E
R RS R
L
- D IT P
S
/
O P
V A A OA N
M R
F P
ND N O A
U L
P
(
I
(
D E
M
-S S
V A
M G
E 4
6 W
n.
L e
1 S
n
$ A
\\
u"
.1
=
5 N
~
O.
c
- ,e...g fr j
2 t
,y
- =..
i W
s.
g g
,i, 5
<')
1
-.A i
s i
s' A
.g
.,$~
. g..
Z i
a D
sq, \\ \\ g g. \\.,.c.
\\g.
O b$y,tsgs\\N'--=.<I,?f"*i 22+.
5
..g
\\'
=%
2 I
. / '.', T/', > s\\
- -\\
\\
' ' ~
.s,\\i
} \\'U ~~ 2
\\
\\\\
N-
/,,
d I
'Jlt,.% k'(q.',
e i
U
,.8 h-i\\v-11 ja.
a 1
s
,fg ':, q..g4g;s s.
t
\\s., -
(
m l
. i y'.
'l i i
/
/
E j
's Y '. - \\
s-pe N. b. %,%. \\
, \\l.
.\\ \\.s \\ ) 'i h
' /.,.
... ' (, i} y, 4.g4 S 'h si.'
W l
s s'-
.\\\\
n
'\\ N'.:. $ '};
l
\\L
's f
~
O
.', '. Q..{,. ) r N.,S Qf*.?'Q lgl W.Q~rR q, Q
- . :,'h)..
-\\.\\;
Il\\
3 kN;;. -
\\s 5 I:kh i lE.'.I w lf;-
W l
T '. d(i.iN M i; M h t.
' (ij) ' TM i E
f <'.
i j
,,, l /
i.
-s i! 1
! !!;,e
- l I
d i.k.
s m
,v-l
=-
.3
- l
.y
.\\.,
9,,.*j f<
.u. :.
J r,
i ". i s.,. -
- e i
t
\\ Y.!
,.I' h,a,:f;"N o
il,
's
!-- Jq v
N
\\
[][.}A* * :!35[-~
I J d
- h[,T_['}&n- 'bs. '_1yg l
f
_'% -G,'
/
g h
h,::::.I p];e il
~
f
/
't:.~. ' @
~
l', i.- #i_
. ' ' +
'- '{*~' "'y'.-} -)#
i U
/
j,i. 'I ;.., l-E l
l E,
l' ll'
@f g
o,
.)
]f j
i./
Ua G
...! : 1.'
1S r*,[, h i
- l l
Ya
?
h$
i Y~;
- i. Q f.Ni.gf ! hij{ 93_ A
!! j e/.o e,\\
- x.. $:
o m
~
en s
r, m y 2,,
g p'~.
~
y4
.,j
)
,(/03:-
48 i
t,
' 'y
,l 40" j;i 15 a
83
\\3
+!
}
k g /t?l3f"
/
( '.(
\\ -.... -
0E 5
\\
l A
i 4\\J
- w-I l
N r
m f"'g u
/*
)
c
(/
^ O....
.,s d
1
.\\
5
.i
- r c
5 t
b b
...1.
1 ; j..
Hu i 'g ' t \\.
4
.g,
_- o
');
Mh;.is2.'[.k\\.\\\\Y\\h.h'5::.7~$~~31
., -s..
i Q
5
- orYe
'.R\\\\N~~~"~I-j) i t; L \\ \\ '
li, ','. ' -
\\
\\\\
4
. * \\.
.. -- ~~-%~M O
.. s s'
s
/.,
e,.. qs
\\
s
..... /.::= =.
' q/s,7- ~,,
.f s
} z\\;. s.\\ \\ {
g',,__...,'.. 4.*.,
a q,j),
1,,..
.' N )o,t.\\,,..
};., c..,. x
,n.
8
- s. - t, y
. \\\\v
.. 5.N. %. v's \\\\n. v. -, 9. 41.,i C '-
l
\\
\\s
, /,.*-
u d
O
.t).,t.m,h,x~...m'N.sm.
R g.
/
6Ws.9. e.s y..
s
. s y :ug,X,\\
's x...
&a s.
-, @, 4. ? %.l i n @d ' m ~ %aVs u
- 9.. s.
m a
- q., y.
8 5;
w@t
\\q/9 i
.,; x.,. -.
ti
.t..i t.
~
Y,'
y.
.n s,
~... s.t s.y u
s,
.,, T-
~~:,
y<
. c D;\\. c f
'g
\\
. b,.!!;(. 8 v.
Yi h. / :s'. 9,. q
- h
'p.jp; (w~%qM[%@g.$; 3,4(Mb "
l'l 'Wkm6<ty;![lu ' p, N p_?-
h Y I'.
--)
ro c k (i -
Ni a
L.
I
.ti!
. p 12 gs m
u, 3
- y. y i
g.x x
~
aa wi s
- s. y
!bh!'[!bgd i'( k.., l.F-4 -- k.,s 3,-
g 1
N \\ih. f ps.
m_d n
. c n y.
?:.. m #' 4,, 5k. u". -st.-1 w.,, m [sp[*
i)<. J!? 8.,.
\\. w
. /N J ' g A.. a.[;
I
' s,.h'..,
/. l','. ~p.-
'.@~ M c
. ~. ~.
..x.=., /
. /
'l Q dFL/
- T 1 f /. -
i
~
~'
N
.'... p gh" -
,. I i
5.
y 5
)' J
-- N,' q ~
I.
'r 1
i i7 ~
Ei
'I-S
- d. i. (*
b
!)
/'
y
/
j V
/[, " [.' [h]! y/ 1 j
'~~ NQ,:
l' s ' /
O*
O OO 1
I 3!
M r3 f.d / / 'l (f I
\\[. '. / 3l
/
7.i f Il e =
F 1.<_ // - ~ ~
g "$<f 55 I
!f F sGAJ e
- I : *.,l.y. @ i, is"]..
.. u. b.&p ry -dA a, '1 ir
'.s__ T.h ;.
l 8
Sa,
=
~'t,f M.'. bR !4N. 5,i '* -
=5E.E5 8
=
3 s
pt '
_f-
-***'C**~~,/
! 1.\\.7 j
c'
['* *
- 2-~
._. OOC o o
3=
Q_ o f. \\.sT ' '(.
.,;p' a
- s
- \\
yc.p-r
.; r '
.,/
/.
t g
Ly O p Oc g E<ywC T.
55_
ih M. ( _
8 3 d g a,i s E u.
3 oas,g,g z
/' {]. @,.t 3
=
% [
s..-
\\. g r s /:,y; li,. l. k-.]- i /p a 6". h
j M u = ) j
. '. 'S'M
$ 5 55 $3 E
> + ea c o u.
o
/p O
a
'fl'f'gg
!g. -
~-
a'sO aby,_== o 4
u
/k-4
@ i5 5m? i e
/
i'e.
o
,t 'N,,,%g' $ ' /,g. A ' )
E 5
f-s J
1 y, g
(
\\
- 3 f
4 h
=3 O O
\\
l C
O < C --
g Z
{
L W
%.,/
2 y n 0
- C O
't e
/%.
.,a -s 1
e N
h h
l
\\
l
~
u, ll ll k" :~if (I'\\
bl U
i
\\ Q f it
qjf l
i,.t i
=
@[A \\s r '/,.
z i
l,i.,
,s $s i<
=
n i
~
a l l i
ay ms\\
/
=,
i,
- k #
\\\\
l g
jy M
~
E=h'k's 8
N.
O u,
lll' % 55 $!
/
O M i.
)Y p i
+m(/
sR\\
l: a\\ v\\\\
/
/ *'f
\\
\\ '
g j!
a g
\\-
o c$p
\\'
+
4@
\\
v
\\\\
\\
/
2 s /
e,0
$$$ll lO Y
3
/
0 g
y v '\\g
/
i
INTERIM TERMINATION SURVEY PLAN AND REPORT l
lb
\\
3.0 OPERATING HISTORY l':
l 3.1 Preconstruction Prior to the commencement of construction in the late 1950's, the Cintichem site was part of a larger, 27 square-mile, tract of land known as Sterling Forest.
This area was sparsely populated with very little development prior to this time.
The principle industry. in the area consisted of iron mining and smelting that was carried on be twee n the late 1700's and the late 1800's.
Other than this the land was primarily used for recreation.
l 3.2 License History The Cintichem facility is currently licensed by the NRC per 10 CFR Part 50 and Part 70 under license numbers R-81, Docket 50-54 and SNM 639, Docket 70-687.
Under the Agreement States Program the f acility is licensed for possession of by-product material under N.
Y.
Code Rule 38, license #0729-0322.
A brief history of operations is as follows-1 1961 Reactor achieved initial criticality and routine operation at 5 MW commenced on a limited duty cycle.
Experimental work commenced in neutron activation analysis, radioisotope
("'
production and neutron spectroscopy.
1 l \\
1963 Routine production of radioisotopes for medical applications commenced.
Initial radionuclides included I-131, Mo-99, Au-
- 198, I-125, Fe-59, Hg-197, Hg-203 and others that were produced by thermal neutron activation.
Some fast neutron activation products such as P-32 were also produced in the reactor.
All of the radioisotopes were processed in the hot cells in building 2.
i 1971 Routine production of Mo-99 by separation f rom the fission products of uranium-235 commenced.
Xenon-133 and iodine-131 were also eventually separated as byproducts from this process.
The reactor duty cycle increased to 95% at this time.
1990 Operations in the reactor terminated in February.
Most operations in the hot cells also were terminated at this time but limited operations with Mo-99 only continued until November when all opera tions with radioactive materials ceased in buildings 1 and 2.
At this time the predominant operation in the facility was the production of Mo-99, Xe-133 and I-131 by separation f rom the fission products of U-235.
Some other isotopes were being produced by the neutron gamma reaction such as I-125, Fe-59, Cr-51 and Ir-192.
Of all the processes carried out at this facility, the t
fission-product Mo-99 production process had the most d
significant impact on the decommissioning project because of the inclusion of some long-lived radioisotopes (i.e. Cs-137, Sr-90,-Ce-144, Ru-103, Zr-95 and Nb-95).
JJM/83.94B Page 12 E----___-._____
INTERIM TERMINATION SURVEY PLAN AND REPORT l
(r.)
3.3 Processes l
v 3.3.1 Mo99 Production The Mo-99 prcduction process involved manuf acturing irradiation targets containing up to 25 gms U-235 as U02 Irradiating these targets in the reactor core at a nominal integrated fission power of 2.4 MWH pc r target, separating the useful fission product radionuclides of Mo-99, I-131 and Xe-133 from the dissolved target processing the waste fission products for disposal and recovering the unused U-235.
Each stage of this process involved work with radioactive materials in varied physical and chemical forms that resulted in different residual contamination j
conditions that had to be addressed in the decommissioning work.
j The irradiation target for this process consisted of a stainless steel cylindrical capsule that contained a film of UO2 (93%
U-235) on the inner wall surface.
The U02 film was deposited by an electroplating process.
I The essential steps of this process included:
dissolving UO2 or U308 feed material in HNO3 to form uranyl I
nitrate (UO2(HNO3)2 + H2O)
{
t'3 converting uranyl nitrate to uranyl acetate (UO2(C2H202)2 +
Q H201 by adding oxalic acid electroplating 002 f rom the uranyl acetate solutiori onto the inner wall surface of the stainless steel (type 304) target j
tube sealing this target tube by welding plugs in the tube ends, decontaminating and leak-testing the sealed target assaying the target to determine the U-235 total }oading and j
uniformity of loading ($ 25 gm total), < 60 mg/cm 25%)
[
I irradiating the target in the reactor core to produce a l
maximum fission power of 13 KW for no more than 336 hours0.00389 days <br />0.0933 hours <br />5.555556e-4 weeks <br />1.27848e-4 months <br /> dissolving the UO2 in the target with sulfuric acid (H2SO4) to f orm uranyl sulf ate (UO2SO4 + H2O)
I traping the fission product gases and dissolver vapors on adsorbants and cryotraps.
(I-131 adsorbed on Cu, Xe-133 condensed in cold trap with liquid N2) 1 precipitating the Mo-99 f rom the UO2SO4 and fission product solution with alpha-benzoin, oxime (alphaBen) m (V) filtering the Mo-99, alphaBen precipitate I
j dissolving the Mo-99, alphaBen precipitate in sodium i
hydroxide to form sodium molybdate (Na2 moo 4)
JJM/83.94B Page 13 l
t
[
INTERIM TERMINATION SURVEY PLAN AND REPORT V]
purifying the sodium molybdate by passing the solution over
[
ion exchange and chromatographic separation media of silver
,impre,gnated carbon, zirconium oxide and activated carbon i
separating the I-131 f rom the copper adsorber by washing the l
adsorber with water and sodium hydroxide solution l
(
I separating the Xe-133 f rom the cryotrap by adsorbing it on chilled activated carbon and purifying it by absorbing H 0, l
2 CO2 and N2 on suitable media and cryoseparating 02, N2 and other gases l
recovering the U-235 as UO2 by converting the waste 002SO4 to uranyl acetate and barium sulfate by mixing it with j
- acetate, decanting the uranyl acetate from the i
barium sulfate and drying and calcining the uranyl acetate j
to uranium dioxide 1
At the time fission product radiochemical operations ceased early in 1990, the Mo-99 product that was being processed amounted to about 7,000 Ci's per week.
The total amount of mixed fission products that was handled at this level of production was approximately 160,000 Ci's per week at processing time.
The i
longer lived fission product radionuclides such as Cs-137 and Sr-90 that are among the few remaining at the end of the gm decommissioning project were being produced at the rate of 7 Ci j
and 6.5 Ci per week respectively.
They existed as compounds of sulfates, sulfites or sulfides in the Mo-99 separation process.
The amount of U-235 handled in the f acility was about 20 Kg per year.
The amount of unirradiated UO2 for isotope production targets that was kept in inventory was limited to about 3.5 Kg by license constraints.
About 75%
of the target uranium was recycled through the U.
S.
DOE Savannah River reprocessing plant and the balance of it was shipped to licensed waste repositories.
Strict accountability requirements and the high value of this material dictated ve y close control of it.
The effectiveness of this control program was demonstrated by the f act that, when the final accounting records were closed at the shipment of the final U-235-bearing waste,100 g of U-235 was measured in excess of the book inventory.
Essentially all SNM had been accounted for.
Each step of the Mo-99 separation process was carried out in a different location within the plant.
Operations with unsealed uranium oxide target material or uranium solutions were carried out in the hot laboratory.
Electroplating was done in the three laboratories in the upper level of the hot lab.
Plated targets i
were sealed by fusion welding in the welding lab on the first floor of the hot laboratory just south of the gamma facility.
Target uranium remained sealed throughout the remainder of the production process until the post-irradiation processing that was
[]
done in the hot cells in building 2.
The counting lab on the
(/
lower level of the reactor building was used for uranium l
accountability assay analyses by titration and by neutron activation analysis.
JJM/83.94B Page 14
/
L
l i
l INTERIM TERMINATION SURVEY PLAN AND REPORT I
i
[
Containment systems for solids, liquids and gaseous forms of radioactive materials existed throughout the facility where they were handled.
Work with contaminated or radioactive materials
!~"
was confined to hot cells, fume hoods, glove boxes, shielding l
caves or other containment systems that were appropriate to
{
control the work being performed.
]
l l
The reactor primary coolant system (approximately 200,000 l
gallons) was contained in the concrete structure consisting of the pool, stall, canal and hold-up tank.
Primary coolant was segregated from the secondary coolant by a
stainless steel tube / sheet heat-exchanger.
An aluminum 100,000 gallon storage tank, located south of the reactor building, was used for the temporary storage of primary coolant as required by maintenance l
operations on the reactor or the primary water system.
The pool j
and stall sections.of the coolant system contained an embedded j
t steel liner.
All other sections of the system had single containment of either aluminum or stainless steel piping or concrete vessels.
Contaminated liquid wastes were collected from sources in i
laboratories and from floor drains and were d' :ceted to a contaminated liquid waste treatment system located in building 2.
Contaminated waste water was treated by evaporation and/or ion exchange.
Contaminated concentrates were solidified and the clean effluent was sampled and analyzed prior to release.
Air ef fluent f rom potential sources of airborne contamination was filtered at least once with an appropriate filter before the air was released to the e nvironme nt.
The reactor building had sources of airborne particulate and iodine contamination that was filtered by HEPA and/or activated carbon at the fume hood or glove box sources.
All of the reactor room air was exhausted through a main exhaust header and it was monitored before release to the exhaust stack.
If airborne contamination was detected in excess of action levels the normal building ventilation could be shut off and the building could be isolated under negative pressure so that all building air could be exhausted through the emergency ventilation system' that had additional HEPA and activated carbon filtration.
I l
The hot lab building had two main exhaust pathways.
Air from l
f ume hoods and glove boxes was filtered through HEPA filters at each source and a " polishing" HEPA filter in the main exhaust i
filter room adjacent to the exhaust fan room.
Exhaust air from l
the hot cells was filtered once through HEPA and twice through activated carbon filters in the hot cell filter room and it was
(
filtered through a third activated carbon and a final HEPA filter in the main exhaust filter room.
All exhaust air was continuously monitored and discharged to the stack on the crest of the mountain to the west of the building.
U JJM/83.94B Page 15
INTERIM TERMINATION SURVEY PLAN AND REPORT
[G]
3.3.2 Reactor Operation Operation of the reactor had the second-most significant impact (second to Mo-99 production) on the decommissioning project.
The reactor was operated in the stall section of the biological shield at a steady-state powe r level of 5
MW from initial criticality in October 1961 through February 1990.
The duty cycle was
> 95%
since 1971 and the total accumulated power amounted to more than 900,000 MWH.
Activated reactor components and some of the bioshield in the stall had to be dismantled and disposed of.
All of the surfaces of the components that contained or were contacted by primary coolant had to be remediated.
3.4 Waste Disposal Practices 3.4.1 Solids Solid wastes, qualifying as Class A waste, were packaged for disposal at the point of generation in appropriate packaging for shipraent to disposal facilities.
Contaminated compactable trash was collected in receptacles designated for such waste and it was kept segregated from non-radioactive trash.
Such contaminated trash was compacted into 55 gallon drums (DOT Spec.
17H) for shipment to a proper disposal facility.
Class A waste rm that was packaged for disposal was stored in building 6.
Solid
()
wastes, qualifying as Class B wastes, were packaged into high integrity containers that were suitable for disposal at the disposal facility.
Class B waste that was packaged f or disposal was stored in the storage cell facility at the north end of building 2.
3.4.2 Liquids l
Liquid radioactive waste was solidified in concrete and placed into appropriate shipping containers for disposal.
Liquids with a high concentration of radioactivity was usually solidified at the source.
Liquids with low concentrations of radioactivity were collected in a liquid waste storage and processing system.
Building 1 had a liquid waste collection system that directed liquids from the hot sinks and cup drains in the upper and lower counting labs as well as from floor drains and some auxiliary l.
equipment throughout the building into a sump in the pump room.
l This waste water was then pumped to a liquid waste storage tank, designated as the Tl
- tank, in the basement of building 2.
i Similarly, in building 2, contaminated waste water was directed to the T1 tank through a network of drain piping from the j
electroplating labs, the radiochemistry lab and the hot cells.
l The water that collected in the T1 tank was processed by evaporation and ion exchange in the liquid waste treatment system located in building 2.
Contaminated concentrates and ion
(
)
exchange resins from this liquid waste treatment system were
(._/
solidified in concrete and packaged appropriately for disposal.
Clean or decontaminated wate r f rom this system was first tested JJM/83.94B Page 16
INTERIM TERMINATION SURVEY PLAN AND REPORT O'
and released from the clean condensate hold tank and then it was collected along with other clean process waste water in two 5,000 l
gallon underground storage tanks located in the mall area east of the reactor building tunnel.
Water in these tanks was sampled and assayed for radioactive contaminants prior to release to the 001 outfall (ref. Figure 3.1).
Three 10,000 gallon storage tanks, located east of building 5 were used occasionally during operation to hold storm water surges that had to be stored and analyzed prior to release.
These tanks were also used as alternates for the 5,000 gallon hold tanks during the last stages of the decommissioning project.
Effluent from these tanks was also released to the 001 outfall.
3.4.3 Gaseous Effluent Reactor exhaust air was filtered at the sources of potential airborne contamination in laboratories and other containments within the building (ref. Fig. 3.2).
Appropriate filters such as HEPA or activated charcoal were used at these locations as required.
Exhaust air was directed out of the building through a header to the exhaust fan room on the west side of building 2.
The main exhaust fan then pumped the exhaust to the main exhaust stack that ran westerly to the crest of the northern slope of Hogback Mountain immediately behind the building.
Reactor exhaust air was continually monitored.
In the event of the presence of airborne contamination in the reactor main exhaust
- duct, the normal ventilation system could be discontinued, and the building could be isolated by maintaining a negative pressure within the building while filtering exhaust air through the emergency exhaust system that contained HEPA and activated carbon filters.
During decommissioning all filters and ducting at the exhaust ventilation intakes were abandoned and a new HEPA filter bank was installed with sufficient capacity to filter all exhaust air that was discharged from within the building.
Monitoring of all exhaust air continued until the exhaust system was finally dismantled at the end of the decommissioning project.
The hot lab exhaust system consisted of two main pathways (ref.
Fig.
3.3).
Exhaust air from fume hoods in the electroplating laboratories, the radiochemistry laboratory and minor hot cells in the charging area behind the cells was discharged through a main header on the roof into the main filter room west of the building.
Exhaust air f rom the hot cells, the waste storage pits and certain glove boxes in the radiochem lab was directed into the hot cell filter room where it was filtered by one HEPA and two activated carbon filter banks and then it was directed to the main filter where it passed through a third activated carbon and final HEPA filter.
The filtered effluent was continuously monitored and discharged to the exhaust stack.
JJM/83.94B Page 17
INTERIM TERMINATION SURVEY PLAN AND REPORT O
cell' exhaust air was carried to the filter rooms via an d
. Hot underground, vitreous tile duct.
'3.5 Incidents and spills Two incidents leading to termination of radiochemical production operations. were significant factors affecting the decommissioning project.
3.5.1 Hot Cell Exhaust Duct The. hot' cell ' exhaust system consisted' of a main duct that ran underground from the main filter room to the hot cell filter room
' and then. to a header that extended under the hot cell operating area floor (ref. Fig. 3.4).
Lateral. ducts extended f rom each hot
- cell and. the T1 room. and j oined. the header under the operating area floor.
The lateral duct intakes were at the base of the north wall - of each hot cell and the ducting ran 'down under the
-cell foundation to the level of the header under the floor.
This ducting was constructed of 4'
sections of vitreous clay tile resulting in many joints in the duct system.
In late 1989, contamination. (predominantly I-131) was discovered in ' ground water that entered a sump located in the northeast corner of the T1 room.
This sump was used to collect ground 0.
water. that accumulated around the base ~ of the T1/ evaporator rooms during
. periods of heavy precipitation.
Upon further
- investigation some. contamination was also found in the site storm drain system at manhole S-4 between buildings 3 and 5, indicating the existence. of contamination beyond the confines of the hot laboratory building.
Further investigation and remedial actions revealed the source of the contamination to be the hot cell exhaust duct system upstream of the hot cell filter room.
~ The evidence indicated that voids existed in the ground around the. foundation of the T1 room.
Pressure measurements at various
' locations in these voids and in the exhaust duct' system indicated that-air.was leaking f rom the hot _ cell exhaust duct upstream of the. hot cell f11ter room and it was. passing through these void spaces around the T1 room and re-entering the hot cell exhaust
, duct? system at s ome point under the hot cell filter room or downstream of it.
During excavation later-in the decommissioning project it was discovered that during construction this area had
- been-backfilled with large pieces of broke _n rock (shot-r ock) from the original. pre-construction bedrock excavation and it was placed so that the interstitial. spaces between the shot-rock were not completely filled with soil, thus creating the channel or path of least resistance by which the air had bypassed the hot cell f11ter room and ducting.
1 L
-JJM/83.94B Page 18
- - _ _ _ _ =
INTERIM TERMINATION SURVEY PLAN AND REPORT l
! f) 3.5.2 Reactor Primary Coolant Leaks j
Q During the investigation into the possible causes for the ground water contamination found under the Tl
- room, leaks were discovered in two locations in the reactor primary coolant containment system (ref. to Fig. 3.4).
An area of unconsolidated l
concrete was found at the base of the north wall of the gamma f acility and also in the southwest corner of the hold-up-tank.
Later, during remediation work in the decommissioning project, a
third leak was discovered in the south end of the canal where it joined the pool.
I 1
I l
i i
I JJM/83.94B Page 19
i l'
sJ'
@D 0
lNot A n rett p@
1
_R E s t a vo u n p
8i j,
O'01
/
i
)
D '4 '
I I
l'
\\
/L T
)
S 4
Sa:npling Poln g f
f cmpff1 Pit
.I[
Afor 001 L
1 00
-Sampiing Polnt I
for 002 h.
I Ml
\\
__')lHold I
lTants
- Samples For 001 Are j'
50 106 ti irid d
- f /66 ' A ' C6mbo s i t e 4
Thesc samples are taken J
from the hold tanks.and samp!Ing pt c sd,lch a re
/
y flow proportions l.
O l
,y
?
)
's _. '
FIGURE 3.1 MALL TASKS, 001 OUTFALL J
t.
N s
L.
_.J r
llCLLL i g
7 y
e L A33tATURi-1
-E-l 1
D i
I C
r 11 ca*'
I L
i L
M a:n.
l C
~
!"IN
,/
omcc 1
m w
i 0
r-
)
C J
E-
\\
_y I/
N:
am:c uh) i I
\\
~
/
R l C
g amec, a au E--
stAM rust rLoca lgl g
/
It M ll \\ \\
OPEN TD 781.12
[
rLa0R EL. 781' L5'
,,a ecm t
ngcp mo' "$'-;'
O
, ~'NN:7M
-- A 4 \\
1-+-
cas x
c-s',
U W,EAL7N U
U U
J V
g U 781' 1.5"
- rr 5
ranucs eS LASGRATDRY UP g P.,u i.m m, 4 VALv ;
gg:r,.
n G D LP TAPE Q 771' 7.S' 5
B OACc 0
5 i
/
NORTH s
i.AvA1cRv v
o
+-!
= ".
a in' r.s.
Tunncu er Vas
]
llllllllllll
' ' =,
,, _i_
l
/
-==
(evertms my,.
is g-to seu.c en 8"
+[-
l
,4 Il E
- " " ! ' "'I'"'! "" !
MEZZANINE FLDOR EL.798' 3' ouT cL.
A33VE & CAM TUDE TLDUR i
LOVER LEVELS, REACTOR EL. 781' 1.5' & 771' 7.5' n(L)
FIGURE 3.2 REACTOR BUILDING EXHAUST VENTILATION Rtr. 6847-1-H2 9\\20\\94 DVGS\\2962 t
i aa i iia!! 8 l ~~
it UF
~
i
~
~
-5
~
~
i 4--
7 c
)
.e Z
- u.. w.11 1
3 3
m
..*8 af I
=
1 i /
i r, a3 R I [!".
E' jf
{
j f F
H r
Z
-W y*
=
{
,+..)
W y
h W. $,a
't
.s e ni n
~
=
2 M%
u.
IX l lllllll N
Cn
<r 1
y
{
- j d
%g v;.gs n
-.o ;;'
n i
s
- )'
G) n.-
o i
b
/vs u.
)
,.-~
,f.
i
- w e
3
- b s
a m5 l
5 g
d I
i...
T,
j 3..~;*.....-
g In ts i
l i
+.o\\
lo e"'"
t r-1
?
i i; ;
ja N e
1 L..:
g d
l Et=
hlg
~
k s
x v
g h $
5
]!: f
.+
i g
I a
i 7[e EP..e.,g 9 p
. u,;
^
i
'0D
~T I
~ f{ d{ pb E". 001d
,fD>
i
'i i
9g.
e I
f' ll
-~e E
i n
1 f e, n
1 I
(-
/ f1 1
[
mae I'
4 f
J 1,
<c.!
>I 6g
[-
s i
4 e
'Y
/j Q
li sf
@g.m (i
I.
T i1-(
3-6/pl
,f
' ji; s
.n 4
w v a.f u
'm-1 m-c
- q'~, !
r m\\\\
73
_.-m
' I ) I:
y N'
E N
/-
w a
E II
,I s!
<C' s"r 9
i*
I Ej 1=
E i.
w y [;
C l hje)..
' 1{
}
7
'l
.5 m
c O
35 e
=
j iI
/h'
~
i u
(
d
,\\
e 2 cs E
)
I;1 n :
k YM I 5 E I 2 ~3 i 'a-i.!
j
("
[
HXT EU H
LL&
E M
CA M
O 0
TM O
5 H
M 0A N
O HG E
A O
F R.
L 4
A
~
C
~
3 1
S
~
l T
E R
K d
C
- 4
. *
- i 0
U G
A
'(w ::,-j T
I F
S
+
l O
M
-(-
T T
O
'?4 C
O U
R D
R 2
T O
S T
G U
A T.*
N a
A l
R C.*
n I
rc D
A H
O U.*
I m
L k.
s P
D.*
X s
e t *-
U a
p
, - *E S.*
te ly t
A t
E p
D V
T 3
n A
I j
- U.-
t T
e n
c c
k e
A m
t
+
e s
a H.*
o S
a y
n t
L X. '
- S o
t E
e i n ip o
A r
t r
a M
t ti
=
n o
a T
a u
MI is tn n
n P
w e
a A
n G
i V
r C
c.
d tn I
1 2
3 l
i i
t S
n S
t e
A 1*n P
n Y
B 1
R E M GO N
E NT L
O T
.:*.*j*:.::. E Z
I C
S D
. N A
L Y L
E
. N I S IU R U
O T
SN
- ls.
D
,, M O
, O T
S DI E A A
O S T L
E
. * ;: j H E
AI P
INT R
R
~
U N
A
- J*
M E E
~
2 M
T A V U
A T C
/
NM R
P E
F OO U
K K
O N
A E CR S
F A
E C D L
T)
FL A EG A
F I
T O
TNl T
P U D
R C'T U H A E U E L N
8 E T S
D R C D
PU E
D S S T
n, L
E N U UE O
A s
O P
E SSR P -
H S
U G
S E
L i
INTERIM TERMINATION SURVEY PLAN AND REPORT 4.0 DECOMMISSIONING ACTIVITIES 4.1 Decommissioning Objective The objective of the decommissioning project is to decontaminate and dismantle the reactor and hot laboratory buildings and their accessory structures and remediate land areas as necessary to allow termination of Nuclear Regulatory Commission and New York State Department of Labor licenses and New York State Department of Environmental Conservation pex nits and orders.
The criteria that must be satisfied in order to achieve this goal are as follows:
I.
Buildings, Furnishings and Equipment - must meet the current criteria in Reg. Guide
- 1. 86 and/or N. Y.
Code Rule 38 Table 5, whichever is more restrictive (refer to Attachments A and B).
II.
Grounds - must meet the criteria specified in NRC SNM 639 license condition G dated August 26, 1993 as amended October 17,.1994 (refer to Attachment C).
4.2 Pre-Decommission Radiological Characterization 1
4.2.1 Contaminated Structures
! /~'T U
This section summarizes the radiological contamination status of general structural surfaces and area exposure rates within the reactor and hot lab buildings that were originally identified as contaminated.
I 1
Structural Surfaces The original characterization of the reactor and hot lab buildings was conducted by dividing them into manageable survey areas based upon the likelihood of similar contamination levels, deposition patterns and radionuclides mix.
The location of these l
survey areas where structural contamination was found is shown in Figure 4.1.
Table 4.1 summarizes the total surface direct beta l
contamination
- levels, removable beta and alpha contamination l
levels and general area exposure rates at one meter above the l
floor for the areas shown to be contaminated in the figure.
Each j
set of determinations (mean and range values) was based upon 30 evenly spaced sample points within each sample population.
In general, the highest total surface beta contamination levels were found on the chaf 533,000 dpm/10 0cm,)ging area floor behind the hot cells (up to in the hot cell conveyor station area (up to 420,000 dpm/100cm )g in the transfer canal (up to 1,634,000 dpm/100cm ),
on the reactor pool surfaces (up to 621,000 dpm/100cm')
and in the lower pump room (up to 242,000 dpm/100cm ).
The removable fraction of this surface contamination was generally less than five percent.
l JJM/83.94B Page 20 1
INTERIM TERMINATION SURVEY PLAN AND REPORT General building area (exclusive of the uranium labs), alpha x
contamination (as found on smear samples) was either nonexistent or-very
- low, with the highest single removable alpha contamination level, of 4,310 dpm/100cm',
being found on the transfer canal wall.
As would be expected, highest surface contamination levels within an area were generally found on the floor and on miscellaneous horizontal surfaces.
Generally accessible area gamma dose rates, measured one meter above the floor surface ranged f rom background up to 3,600 uR/hr.
The radionuclides mix was found to vary between the reactor and hot lab buildings.
Table 4.2 presents a
summary of gamma isotopic ratios for composite smear samples that were selected f rom the various contaminated areas, shown in Figure 4.1.
Sr-90 activity was present in proportions to that of Cs-137 ranging from 1:1 to 4:1.
The isotopic mix comprising surface contamination varied considerably by location.
The overall predominant radionuclides, in descendin Cel44, Csl37, K40, Zr95, Co60, Nb95, Rul03, Zn65, g order weret and Ag108 Trace amounts of Sc46, Mn54, Co57, CoS8, Sn113, Sb124, Sb125, Cs134, Eu152, Eu155 and Ir192 were also occasionally present.
In general, activation products were more prevalent in the reactor building and fission products more predominant when approaching the hot cell areas.
O)
(
Bot Cells The hot cells
(#1 to 5). located in the hot lab building had internal exposure rates ranging from 50 to 1,000 R/hr as of Ju}y 1990.
A small piece of a smear sample representing (1 cm )
i taken inside of hot cell #1 (7-16-90) was found to have 1.2 uci i
of Nb9b, 0.59 uCi of Zr95, 1.27 uCi of Mo99, 0.12 uCi of Rul03, 0.0041 uCi of Sb125, 0.0062 uCi of Cs134, 0.067 uCi of Cs137, 1.33 uCi of Cel44, 0.13 uCi of Ir192 and 1.53 uCi of Tc99m.
l Obviously, only the Csl34, Csl37 and Cel44 were significantly present at the start of decommissioning due to radioactive decay of the shorte r half-life radionuclides.
A conservatively high estimate of total surfacedpm/100 cm} nation levels contam in the hot cells amounted to 2.2 x 10' The estimate was based on 6 months of decay with Sr-90 activity assumed equal to Cs-137, and a 10% smear retention factor.
Other Areas Radioactive contamination was present beneath portions of the reactor and hot lab buildings.
This contamination was found in or on soil, rubble backfill and on some of the underlying bedrock.
These contaminated areas were adjacent to the buried holdup tank, under the primary coolant storage tank, under or l
around the canal / gamma pit, and in soil surrounding the buried
(_)d hot cell exhaust system and the subsurf ace T-1/ evaporator rooms.
The backsides and/or undersides of structures in these areas were also contaminated.
JJM/83.94B Page 21 lL__ _
INTERIM TERMINATION SURVEY PLAN AND REPORT I
' h Some surface soils were found to be contaminated in the fenced V
area behind building 6 (Csl37) at the outf all of S-5 (Csl37 and Co 0) and downstream of the outfall at S-3 (Csl37)
(refer to 6
Figure 4.2.)
4.2.2 Contaminated Systems and Equipment This section summarizes the original radiological status of the systems and equipment within the reactor and hot laboratory facilities..
Contact gamma dose rates were taken on each component and at periodic sections of pipe or duct runs.
Components were opened for direct internal. measurement of contamination levels where' non-destructive access could be obtained.
Samples of sediment were obtained when found inside of i
opened components and analyzed by gamma spectroscopy.
j Primary Reactor Cooling System Twenty contact gamma dose readings were taken on the primary reactor cooling systems and equipment.
The external gamma dose rates ranged from 14 uR/hr to 2,900 uR/hr on contact, with an average dose rate of 1,060 uR/h r.
Two components, the primary side of the heat exchanger and the primary pump were opened for direct measurement of internal contamination levels within the primary system, and 337,000 and 809,000 dpm/100 cm' beta was L
found, res pe ctively.
A sample of the internal sediment was taken along with a
direct beta contamination measurement.
This sediment sample was analyzed by gamma spect r ome t ry.
The radionuclides comprising this internal contamination and their relative contributions were found to be; Sc46, 152;, Co60, m; 4%
Zn65 8%; Nb95, 22%; Ag110m, 10%, Cel44, 34%; Eu 2%; Ir192, 3%- and the following radionuclides at less than 1%
each
- Mn54, 59, Rul03, Ag108, Snll3, Sbl24, Sb 25, Csl34, and Csl37.;
This l
-Fe isotopic mix was determined to be present as of 27 July 1990.
The secondary side of the heat exchanger was opened for direct contamination measurement to determine if pr to secondary Less than 1000 dpm/100 cm} mary leaks had occurred.
beta was found on the secondary side of the heat exchanger.
Primary Reactor Cooling Purification System Seven gamma dose readings were taken on the primary reactor l
cooling purification system and equipment.
The external gamma dose rates ranged from 500uR/hr to-4,000uR/hr on contact, with an average gamma dose rate of 1,606uR/hr.
A flange cover plate was removed and a direct beta contamination measurement was taken on
{
the back pide of the plate.
A contamination level of 13,000 dpm/100 cm beta was found.
Reactor Building Air Exhaust System (VO Twenty-one gamma dose readings were taken on the reactor building air exhaust system.
The external gamma dose rates ranged from 21uR/hr to 1,400uR/hr on contact, with an average gamma dose rate JJM/83.94B Page 22
l INTERIM TERMINATION SURVEY PLAN AND REPORT l
l l
I on contact of 340uR/hr.
Two components were opened for direct b
measurement of internal surface contamination.
The total surface direct beta contamination levels were 1600 dpm/100cm' and l
1900 dpm/10Ocm*.
l Hot Laboratory Building Air Exhaust System Sixteen gamma dose readings were taken on the hot laboratory building air exhaust system.
The external gamma dose rates ranged from 34 uR/hr to 2,200 uR/hr on contact, with an average gamma dose rate on contact of 460 uR/h r.
Four exhaust system components were opened for direct measurement of internal surface contamination.
Total surface direct beta contamination levels ranged from 1100 dpm/100 cm' to 20,400 dpm/100 cm'.
Hot Cell Air Exhaust System (up stream of polishing filters)
One gamma dose reading was taken on the hot laboratory building hot cell air exhaust system at a point down stream from the hot cell filter bank, where surrounding soil had been previously excavated.
The external gamma dose rate was 235uR/hr on contact.
This system was in operation and therefore, the internal surfaces of the system were not characterized, however the internal contamination of this system was similar to the inside of the hot cells.
p i
h Exterior Air Discharge Duct and Stack Ten gamma dose readings were taken on the building air exhaust systems located exterior to the reactor building and the hot i
laboratory.
The external gamma dose rates ranged from 10uR/hr to 30 uR/hr on contact, with an average gamma dose rate on contact of 18 uR/hr.
The base of the exhaust stack was opened for direct measurement of internal surface contamination.
A tiotal surface direct beta contamination level of 75,700 dpm/100 cm was found.
]
A small sediment sample was obtained from the base of the stack.
Non-quantitative results (July 27, 1990) indicated the following j
proportions of gamma emitters being present in the stack and duct work:
K-40, 80%; Co-60, 7%; Cs-137, 6%; Ag-108%, 4%; Ag110m, 2%
and; Cs-134, Mn-54, Zn-65 and Nb-95 comprising the remaining 1%.
Sr-90 was presumed to be present in amounts approximately equal to that of Cs-137.
Waste Water Evaporator System l
Systems and equipment within the T-1/ Evaporator room were characterized for typical external contact exposure rate only because the system was in active use.
The emergency surge tank
~
(Tl) had a gamma dose rate of 200 mR/hr to 500 mR/hr.
The
/
equipment within the evaporator room had a gamma dose rate of k
approximately 10 mR/hr on contact.
JJM/83.94B Page 23 1
INTERIM TERMINATION SURVEY PLAN AND REPORT Storage Cells Air Exhaust System Three gamma dose rates were obtained on the air exhaust duct from storage cell header located within the crawl space beneath the radwaste laydown/ shipping area.
This duct discharged to the first filter bank of the hot cell exhaust system.
The external gamma dose rates ranged from 5900 uR/hr to 9500 uR/hr on contact, with an average gamma dose rate on contact of 6800 uR/hr.
A 900 uR/hr dose rate was found on the exhaust duct header sump.
The sump was opened for direct beta contamination measurement.
A contamination level of 132,700 dpm/100 cm' was found.
4.2.3 Activated Components and Structures Isotopic concentrations, curie content and radiation dose rates were calculated f or components and structures that were activated during operation of the Cintichem reactor 1 A neutron activation analysis was performed to provide an estimate of the neutron-j induced radioactivity in core components, structures and the surrounding bioshield walls.
The components that were addressed in this study are detailed in Table 4.2.
i o
Reactor core support tower; o
Reactor grid plate (and locator pins);
o Plenum; o
Core outlet assembly; o
Beam tubes; Thermal column and thermal column lead shield assembly; o
o Pneumatic rabbit assembly; o
Concrete bio-shield.
4.3 Decontamination Procedures l
4.3.1 General The general methods of decontaminating the various areas within the facility and on site are summarized here briefly.
The detailed procedures that were used in specific areas of the plant or on specific tasks in the project are presented in Section 5 of 1 Radionuclides and Dose Rate Analysis for Neutrotrinduced Radioactivity in the Cintichem Reactor", prepared by TLG Engineering, Inc., October 1990.
JJM/83.94B Page 24 i
INTERIM TERMINATION SURVEY PLAN AND REPORT this report in order to relate survey procedures that were used in each area to the specific characteristics of the work done and to the 11nal conditions existing in each survey area.
Equipment and furnishings were either disposed of at a
licensed facility or they were decontaminated either on site or at an off-site decontamination service facility.
The choice depended upon the amount and type of contamination, the size of the contaminated item and the relative costs associated with each option.
Structural surfaces were either scarified, washed or dismantled and discarded depending upon the above mentioned factors for decontaminating equipment.
Underground piping and tanks were removed and either discarded or decontaminated depending upon the same aforementioned factors.
Contaminated soils and rock were excavated and discarded.
Some mixed wastes and some wastes from the reactor structure remained on site for lack of access to a proper disposal facility.
(T
/~
The Chem-Nuclear, Inc.
disposal facility in Barnwell, South
)
Carolina and the Envirocare, Inc.
disposal facility in Clive, Utah were used as the final disposal facilities.
Alaron, Inc. in Wampum, Pennsylvania, received most of the dismantled equipment and furnishings for decontamination and volume reduction by various methods.
Chem-Nuclear, Inc.
in Channahan,
- Illinois, reduced the volume of most of the low concentrated dry waste by super-compaction of drums that were preliminarily compacted at Cintichem.
Hot Cell Se rvice s in
- Seattle, Washington, decontaminated most of the lead by CO2 pellet blasting.
Some mixed waste was incinerated by DSSI in Knoxville, TN.
4.3.2 Decommissioning Organization Figure 4.3 is a diagram of the basic Decommissioning Pr oj ect Organization.
4.3.3 Final Survey Organization Figure 5.1 is a diagram of the basic Final Survey Organization.
1 O
V JJM/83.94B Page 25 l
l
Q*
S A
'l f
- xx' -
m a.i g
ip !-
r 0
,d l
I
\\
fl3 Q
_i IN 1
l
}
d 9 y I!3 1.
iTC-x\\\\\\\\.____l
'x
.I w !
P g!!j!
!r i,
q j
!.I;!!!! ![l ie i
"*s % %x i
,i u
r i
J.
' k'N i
LLL.! L I
e t,re y
g b
I '
n 3l f'
W
- +,
=
E cd w
!u dia V
ji.ll i
- g,,,-
b5 [L 7'
y':. ' N,l
-*fft E
i
~"'4[2-l!l!
- =gI(
1 dI mg v
e 1
- . n 1,Q,}i
.m o
f a
a y
d y_
g ll ;
s f
- s
.i
......r q b i
s si n i i
i L___. Li 9
m.i g
~
l4{r'g+p:
t-4 c
g-z e -
6, gf.
1 i m E
s e
ja L_
,a 8
}
u.[
,s' l
9
~
F i
e c _ 4 h_,t _.a _h
_ d L "'
h ic la U" H sju
- lfi ihi i
i J
IA @_$$[
sm
,1 w_
c
'J U
U U
U U
L'
-U
,d "g
s 6g5 s
t
_ _sg!il -[
-,p& p (f,! ; \\ WO fsg---
a m
1 N
s o
I e
$Nxx hx'.'! l=
l t
~~T" s
l 8
n 4-ja g
i a
s O
$WS# bww 0
i Tygggg
% egg e
E
)
E55 5!
s s
se es
{
$'I 5S
'I 2
U,~3 -
1.:
4 i
e I
s n
n 3
a
=r e
%s 4;i H.1'x1 d
r a
l z
i Oc dl
=
J 4
~
s ct e-d,<f 3
/
g s
LJ a A'
HE i
- e 0
8,6 g:f
- u.
3 l
5 5
L 1
m E
] !
-h n V\\ w, P
\\
J '[
LU s
psio
%P o
d.
$ll I 0
h id\\\\\\\\\\\\xxx
UQ SAEM l I l!
pgg g
a l I l!
mg U
m, P
g SO q
NI m
OG p
ITT Il AR RO EP PP OU 1 I l g S
I l lI BAL I I l I I I l I I M
R I
E f
G l
l g N
A P
W N
RA U
O I I I i A&
1I G
A A
EE O
C M
G R
B C
N L
O AS G
R E
MH i
N 0
I J
I 9
V O
f r
N R
N S
C I I~ lE_1 l
?_
P U
O I I I l I l 1 I I l
I II TN EM ER i I I US Y
AD E
R V
EO R
C U
YE i
S ER V
D RD L
UN E
S_ A I
F I l l
lI N
I II OG INGIT UR DO EE RE O
I T
A A
R A
TD D
A N D_A i I i I i
i e*
Cintiches T A B L(
I Sussary of Structural
, /~%
g' l
SETA CONTAMI N A710N t
AREA ll DO5E Raft ll Direct ll l
No-1:
Naas ll s e s s e s ll s a s s e e ns a s s e n e s s a n s sis, a s e s s s en e.... s sll...../hr
- l nesa ll Ranse ll nee up
......nll seses.. l;............. l;
....e 3 1 ll Cell Plug Area floor ll 30 90 ll 3200 ll 0 24700 ll 3 2 ll C ell Plus Area Lomer Wall N/A ll 460 ll 0 9050 ll 4-1 ll Upper Level Hot Lab Floor ll ND ll 2l00 ll 0-12600 ll 4-2 ll Upper Level Hot Lab Lower Wall ll N/A ll 140 ll 0 1100 ll 4-3 !! Upper Level Hot Lab Upper Wall N/A ll 30 ::
0 520 4 4 ll Upper Level Hot Lab Ceiling ll N/A 40 ::
0 55C ll 6*1 ;; Waste Pit floor ll l'0 3600 ll 30000 ll 0-135003 ll 8,
6-2 !! Waste Pit Lower Wall N/A ll 620 ll 0 1930 ll 6-3 !! Waste Pit Upper Wall ll N/A 8 ll 500 ll 0-4510 6 4 ll Waste Pit Ceiling ll N/A ll 270 ll 0-1840 ll 6 5 ll Waste Pit Horizontal Surface 1:
N/A l l "" 10330 ll 0-69000 ll f
71 ll Shipping Area Hot Lab floor 0-1340 !!
22000 !;
0 323000 ll
{
7-2 ll Shipping Area Hot Lab Lower Wall ll N/A ll 1740 ll 0-20320 !!
j 7-3 ll Shipping Area Hot Lab 1*pper Wall N/A ll 13200 ll 0 362000 :l
)
7 4 !! Shipping Area Hot Lab Celling ll N/A ll 1110 ::
0 29360 ll 7-5 ll Shipping Area Hot Lab Horizontal surface!!
N/A 2340 ll 0-8320
(
9-1 ll Charging Area rioor ll 490 3090 ll 83620 ll 0-533000 ll 19 9-2 ll Charging Area Lower Wall l1 N/A 1320 ll 0-16060 ll B 3 ll Charging Area Upper Wall N/A ll 1030 ll 0 12170 ll 9-4 !! Charging Area Ceiling ll N/A ll 710 ll 0 4400 j
9 5 ll Charging Area Horizontal Surface ll N/A ll 1650 ll 0-19320 l C
10-1 ll Conveyor Station Hot cell rioer ll ND ll 35230 ll 1900-420000 ll C
10-2 ll Conveyor Station Hot Cell Lower Wal!
N/A ll 3900 ::
750-26000 !!
G 10 3,;l Coveyor station Hot Cell L'pper Vall N/A ll 90 l; 0 1160 ll 121 !! Exhaust ran Room Floor 70-150 ll 7370 ll 2230-21810 ll 9
12 2 ll Exhaust Tan Room Lower Wall lt N/A 610 ll 0-5670 3
12 2 ll tahaust ran Roos Upper Wall ll N/A 200 ll 0-1150 ll 15-1 :: Oper Area Manipulator Inc Floor 130 1590 ll 16000 ll 0 137000 ll 3@
15-2 ll Oper Area Manipulator Enc Lower Wall ll N/A ll 3150 l1 0 18570 ll i
15-3 !! Oper Area Manipulator Inc Upper Well ll N/A ll 50 ll 0-160 ll g/
15-4 ll Oper Area Manipulator Enc Ceiling N/A 150 ll 0 570 ll 15-1 !! Canal Floor ll 310-1990 ll 206100 ;l 0-553000 ll 6@
18-2 !! Canal Lower Wall ll N/A ll 296700 ll 5700-1634000!!
11@
21-1 ll Target Welding Shop Floor 10 35 ll' 0 ll 00 21-2 ll Target Welding Shop Lower Wall ll N/A ll TO ll 0 1090 1;
2 21 3 !! Target Welding Shop Upper Wall l:
N/A ll 60 ll 0 1120 ll 2
21-4 ll Target Welding shop Ceiling ll N/A 50 ll 0-1400 1
21-5 !! farget Welding Shop Horizontal Surface !!
F/A ll 0 ll 0-0 ll 21-6 !! farget Welding shop ll N/A ll 65 !;
0-840 ll 23-1 ll Old Reactor Cave Area Floor 230-2990 !!
5200 !!
0-31590 !!
16 26-1 ll Reactor 3rd Floor floor ll 10-990 ll 1320 ll 0 5730 ll 2
26-2 ll Reactor 3rd Floor Lower Wall N/A 1600^ll 0 18000 ll
{
26 3 ll Seactor 3rd floo'r Upper Wall l!
N/A ll 140 !:
0-2140 26-4 !! Reactor 3rd Floor Ceiling ll N/A l!
460 ll 0-4730 ll 26-7 l! Reactor Pool Cutter.
ND ll 251000 ll 63970-550000 ll 8
26-8 !! Reactor Pool Walls ll FD 485000 ll 323000-621000 !!
C 30-1 ll 781.12 ' tiev Reactor Area Floor l*
60-1090 ll 990 ll 0 9580 ll 19 30-2 ll 781.12' Elev Reactor Area Lower Wall N/A ll 1100 ::
0-21160 ll 3
30-3 ll 731.12' Elev Reactor Area Upper yall ll N/A ll 100 ll 0-560 ll 8
32-1 ll Lower Pump Room floor ll 240-790
$2250 l' 0 142000 ll SB 32-2 !! Lower Pump 800e Lower Wal!
N/A ll 450 ;*
0 214C ll 0
32-3 ll Lower Pump Roon Upper Wall N/A ll 730 ll 0-5050 ll 8
32 4 ll Lower Pump Roos Ceiling
- l N/A ll 190
- :
0 1810 ll 32-5 : Lower Pump Room Horizontal surface ll N/A 5190 ll 0 54260 ll 33 1 :: bpper Pump Room floor ll 210-1990 ll 3090 l*
0 17030 ll 33 2 ll Upper Pump Room lower Wall l:
N/A ll 1150 ::
0 8930 ll B
33 3 ll cpper Pump Roon Upper Wall ll W/A 110 ll 0 7210 1 01 33 4 ll Upper Pump Room Ceiling N/A ll 50 ll a.560 ll 1(
N0ft:
Direct Instrument contaelnation levels corrected for na Removal seest data corre cted for instrument background <
Dose rate data corrected for bachground f rom natual roc N/A a Not Applicable
?
FD a Not Determined w.-
4 * --
- ulumume-bM Sg eiu-emm-,w w egg af hy,PJW"% $WIM Eds*N -
e 4
- 4 w 9
- te * * -8
'.-4 4
4
- I
(1.instlen Lesels Nt7 otn/ loo CH*2 t
ALPHA CONTAMINA71oM 4ble Direct Removable pin Mean mense nean sanse ::
- ......se
- 140-1190 FD ko 3 ::
0-20 5 620 FD FD 1 ::
0-10 30 1500 FD ND Is ::
3-50 0-410 FD HD 4 ::
0-50 i
5-Iso FD FD 2 ::
0-7 o-30 FD ko 1 ::
0-10 9
c-10000 ::
0 0-0
<1 ::
0-3
)
i
- M '
li-F:
APERTURE j
8 c-260 FD W
't
<1 ::
0-20 CARD 8
o-540 o
0-0 10 ::
0-70 8
o-3470 o
0-0 3 :: 'o.70 B.
0-4180.::
o 0-0 7 ::
0-150 ::
3
- c 2soo o
0-0 2 ::
0-30 4 ISO Avgjg g 0-210 No FD
<1 ::
0-3 gm gfd 2-5240 o
0-0 27 ::
0-190 ::
2 26850 !:
FD HD 20 ::
o-310 ::
0 180 FD ND 1 ::
0-10
)
c-250 FD j!
ND
<1 ::
0-3 l
c-7 FD ND
<1 ::
0-7 l
70-1710 FD FD 22 ::
0 110 ::
}
40-3060-FD No 14 ::
o-so 2 5870 FD FD 10 ::
0 150 ;;
o-340 1:
FD ND 2 !!
o-13 150-2500 ND WD 11 !!
o.50 0-730 FD FD 4 ::
0-30 20-220 FD WD ll 2 !!
o-lo 340-16570 ::
'o 0-0 2 ::
0-7 0-3140 FD FD 41 ::
0-7 0 780 No No 2 ::
07 0-10 Fo ND 1 ::
o7 20-2706o :
o 0-0 31s ::
3-1150 ::
1920-7s200 ;;
o 0-0 640 :: 150-4310 ::
0-0
- 30s p-5610 ::
o ::
0-0 7 275 48 0-160 ::
4 ::
0 35 7-275 23 0-134 ::
4 ::
0-35 2-40 FD Fo 2 ::
0-10 c-o FD FD 0 ::
0-o.
- ^
00 FD FD 0 ::
0-0 ' !:
0-1780 No ND 2 ::
0-20 2 60 Fo No 1 ::
0-10 l
o-170 FD FD 1 ::
o-7 o-0 FD FD o ::
0-0 0-G No ND 0 ::
o-40 o 12o FD FD
<1 ::
0-7 o-so No HD
<1 ::
0-3 240-2560 FD Fo 14 ::
o-40 l
O-240 FD No 1 ::
o-7 o-tso ND No 1 ::
0-10 0-40 Fo ND 1 ::
0-7 0 ::
FD No o ::
0-0 00
-FD FD o ::
0-0 5-670 Fo ND I !!
o7 2-1090 ND
- j FD 3 ::
03 s
} rad material backstound.
Detien.
l j
I e
e e - e b
.s
Cantiches TABLE 4.lb 1:otopic Coogesition o f 5tructurg y I
!! CAF.nA EMITTi>C isofor:C RATIO AREA ll ll ll Ho.
ll Na.e
- l
>4-22 x-40 ll SC-46 l:
MN-54 ll
......:l
........................................ll.........lll rr l...........
.........ll
.........ll
,,,,i 3-1 ll Cell Plug Area rioor ll
- l 85.42::l ll 3-2 ll Cell Plus Area Lower Wall ll ll ll ll ll 4 1 ll Upper Level Hot Lab rioor ll j;
ll 4-2 ll Upper Level Hot _ Lab Lower Vall ll ll
- l l*
ll 4-3 ll Upper Level Hot Lab Upper Wall ll ll ll ll 44 :: Upper Level Hot Lab Ceiling ll ll ll ll 6-1 ll Weste Pit rioor ll ll ll ll ll 6 2 ll Waste Pit Lower Wall ll ll l
- l ll 6-3 ll Waste Pit Upper Wall ll ll ll ll ll 6-4 ll Waste Pit ceiling ll ll
- l ll 6-5 ll Waste Pit Horizontal Surface ll ll 65.401 ll I
7-1 ll Shipping Area Hot Lab rioor ll ll 7 2 ll snipping Area Hot Lab Lower Wall ll ll 1:
ll ll ll 7 3 j; Shipping Area Hot Lab Upper Wall ll ll ll ll ll 7-4 ll Shipping Area Hot Lab Ceiling ll ll ll ll 7-5 ll Shipping Area Hot Lab Horizontal Surf ace ll
- l ll 9-1 ll Charging Area rioor l'
ll ll ll 9-2 ll Charging Area Lower Wall ll ll ll ll 9-3 ll Charging Area Upper Wall ll
'll ll ll ll 9-4 ll Charging Area ceiling ll l:
ll ll l
9-5 :: Charging Area Horizontal surface ll ll ll ll ll 10-1 ll Conveyor Station Hot cell rioor ll ll ll ll ll 10-2
- Conveyor Station Hot Cell Lower Wall ll ll ll ll ll t
10 3 ll Coveyor Station Hot Cell Upper Wall ll ll ll ll 12-1 ll Exhaust ran Room floor ll ll ll ll 12-2 l* Exhaust ran Room Lower Wall ll ll ll ll ll j
12 '4 :l tahaust ran Room Upper Wall l:
ll l
ll ll 15 1 ll oper Area Manipulator Enc rioor ll ll ll ll
- l
)
15-2 :: over Area Manipulator rne Lower Wall ll ll ll ll ll 15-3 :: over Area Manipulator Enc Upper Wall ll ll ll ll ll 4
15-4 ll Oper Area Manipulator Enc Ceiling l*
ll ll ll ll
(.
1s 1 ll Canal rioer ll
- l ll l!
18 2 :l Canal Lower Wall ll 0.09:ll 15.05::l 0.90 l:
1.09:ll 21 1 ll Target Welding shop rioor ll ll ll ll 21-2 ll Target Welding Shop Lower Wall ll ll 21-3 ll Target Welding Shop Upper Wall ll ll ll 21-4 ll Target Welding shop Ceiling ll
- l ll ll ll 21 5 ll Target Welding shop Horizontal surface ll ll ll ll l!
21-6 ll Target Welding shop ll ll ll ll ll 23-1 !! Old Reactor Cave Area floor ll ll ll l
26-1 !! Reactor 3rd rloor floor ll ll ll ll 26-2 ll Reactor 3rd rioor Lower Wall l!
ll ll ll 26-3 !! Reactor 3rd rioor Upper Wall ll ll ll ll ll 1
26-4 ll Reactor 3rd floor ceiling ll ll ll ll 26-7 !! Reactor Pool Cutter ll ll 14.17:ll 2.67:ll 1.26 ;;
00 26-8 ll Reactor Pool Walla ll 0.22*;
12.16:ll 3.27 ll 1.38 ll 30-1 ll 781.12' Elev Reactor Area ripor ll ll I:
30-2 ll 781.12' 21ev Reactor Area Lower Wall ll ll ll
~ll 30 3 :
781.12' tiev Reactor Area Upper Wall ll ll ll ll 32-1 ll Lower Pump Room rioor ll ll ll ll 32-2 ll Lower Pump Room Lower Wall ll ll ll ll 32-3 ll Lower Pump Room Upper Wall ll ll ll ll ll 32-4 l* Lower Pump Room Colling ll ll ll ll 32-5 ll Lower Pump Room Horizontal Sur' ace ll ll ll ll ll 33-1 :: Upper Pump Room rioor ll ll ll ll 33 2 ll L'pper Pump Room Lower Wall
'l ll lI ll 33-3 ll Upper Pu=p Poom Upper Wall ll ll ll ll 33-4 ll Upper Pump r.oom Ceiling ll ll ll ll x
""**------,.h-
~p:1- --
~-urg.yCa,.__.,f;,-- y **:~ pt..*frAW.-r.;,. cst-&.,0 '. *l7d LW.m-w.cMg., T'Z *o
-- -- e
-.r--
s% _, _r__1 :.. - - - -
.----=-%--
6
,,,4 Wgg- -
C--..
5
Dinstian 1.
- : 'co.57 ::
co 58 ::
co.so 28 65 ::
zu.ss ::
us.95 :: au.103 :-
e........
1.35:::
2.00:::
0.15:::
2.35:::
7.so:::
s.53 ::
1.14:::
33.3os.
c m
8 0.42 ::
0.212::
1 95:::
2.74:::
0.19:::
pg Q 7.01:::
s.os:::
8 8
AISO Availwie org<
8-Anertas. ay 8
4.2c:::
4.27:::
1.51:::
8 8
8 8
8 8-8.-
0.70:::
19.42 ::
18.18:::
g.
- 2. s s.'.'
8 8
8 7.41:::
5.01%::
7.40:::
0.50:::
3 0.02:::
s.65:::
5.30:::
1.00 ::
3.17:::-
0.54:::
3 0.10:::
7.04:::
15.1a:::
4.81:::
s.24:::
0.76:::
l s.7s:::
4.38:::
1.35:::
4.so:::
1.07:::
l 16.98:::
f 4.
e
- l. l l
~.-.-
=
u =-..
- -r.-
. =* ' m==s F-., w _ _
. w..m
.,,........ u. w,.
l Cintiches r
t ~%
t AREA
' cec *.A CMITTINC !$070PIC RATIO
. "..... l :
Name 3 ) !! Cell Plus Area floor.......................................,:;l;,,,,,,.....,,,,,,,5l; Ac los ::
.l
,,l..
,,,,,,,,,..,,,,,,,,,l,l; 3-2 ll Celt rius Area tower Walt
'l I
l l
- l
'l ll 4-1 l! Upper Level Hot Lab Tloor 4 2 !! Upper Level Hot Lab Lower Wall ll ll 4-3 ll Upper Level Hot Lab Upper Wall ff fh
((
ll
{l l
4-4 :: Upper Level Hot Lab Ceiling
{
6 1 ll Waste Pit Floor
- =
.l 6 2 j' Waste Pit Lower Wall ll II.
L l
6 3 !! Waste Pit Upper Wall s
ll
{
6 4 ll Waste Pit Ceiling ll 6 5 !! Waste Pit Horizontal surface 0.13:!!,
ll
'!{
((
7-1 *l Shipping Area Hot Lab Floor
- l
- p 1-2 l
- shipping Area Hot Lab Lower Wall s
ll l;
7 3 ll Shipping Area Hot Lab Upper Wall I
ll 7 4 ll Shipping Area Hot Lab Ceiling I
ll 7 511 shipping Area Hot Lab Horizontal Surface!!
9 1 ll Charging Area floor 2.06:ll ll ll j
9 2 !! Charging Area Lower Wall ll g.
t l
9-3 !! Charging Area Upper Wall ll ll j;
9-4 ll Charging Area Ceiling lj ll j;
9-5 ll Charging Area Horizontal Surface ll ll ll ll ll 10-1 !! Conveyor Station Hot Cell floor ll 10-2 ll Conveyor Station Hot Cell Lower Wall j;
ll ll j;
10-3 ll Coveyor Station Hot Cell Upper Wall l!
ll l;
12-1 ll Exhaust Fan Room Floor ll lj j;
}
12-2 !! Enhaust Tan Roca Lower Wall 12-2 !! Exhaust ran Roon Upper Wall ll ll lj ll ll ll
- l
- l 15-1 !! Oper Area Manipulator Enc floor ll lj ll jj jj 15-2 ;! Oper Area fianipulator Enc Lower Wall ll ll ll ll S.
15-3 ll oper Area Manipulator Enc Upper Wall 1
ll
- l ll 15-4 ll Oper Area Manipulator Enc Celling
?
ll ll l*
IS-1 !! Canal floor b
18-2 !! Canal Lower Wall ll ll 0.13 ll 3.85:ll 4.26:ll 0.32:ll 0.08:ll 21-1 *! Target Welding Shop T!oor ll l*
21-2 ;! farget Welding Shop Lower Wall 11 ll 21-3 ll Target Welding Shop Upper Wall ll ll ll 21-4 ll Target Welding Shop Ceiling ll ll ll l
21-5 ll Target Welding Shop Horizontal Surface 21-6 !! farget Welding Shop ll l
11 ll ll,
)
23-1 j; old Reactor Cave Area floor 26-1 ;; Reactor 3rd floor floor ll ll ll ll 26-2 j Reactor 3rd Floor Lower Wall i
ll ll ll ll l
26-3 ;j Reactor 3rd floor Upper Wal!
ll ll l
25-4 ll Reactor 3rd Flo'dr Ceiling 26-7 !! Reactor Pool Cutter ll ll 1
26-8 l! Reactor Pool Walls ll 1.67:ll 8.61:l!
1.30:!l 0.26:!!
(
4.80 ll 21.41:!!
1.51:;!
0.23 ::
(
30-1 !! 781.12' Elev Reactor Area floor ll ll 30-2 j; 731.12' Elev Reactor Area Lower Wall ll ll ll ll 30-3 ;; 7 81.12 ' Eley Reactor Area Upper Wall ll ll ll ll -
32-1 ll Lower Pump Room Floor ll ll 32-2 !! Lower Pump Room Lower Wall ll ll ll ll ll 32-3 !! Lower Pump Roon Upper Wall ll ll ll ll 32 4 ll tower Pump soo Ceiling 11 ll ll 32-5 ll Lower Pump Roog Horizontal Surface ll ll 33-1 :l Upper Pump Roo rioor ll ll ll l:
33-2 ll Upper Pump Room Lower Wall ll ll ll ll 33-3 :: t'pper Pump Roo= Upper Wall
- t 1:
ll ll l:
33 4 ll Upper Pump Room Ceiling ll ll 9
[
j J
~~--.=E : ~ =3 ' =;_q=nm,
-. a -
. _wy
_myw
. _n, w;_. _
.w=....
, m. -, r.-
re w w..a ;_,,,
-~ ~
O k
- - - - ~ -
TABLg 4.}C
!stf opIC Composition of Structural Contamination
- C5-134 :: C5-137 :: CE-144
- 20 152 :: EU 155 *: la.192 ::
TOTAL :
5.::.........::.........::.........::.........;;.........;;.........::.........::
4.ast :
6.23t::
100.00:::
100.00x::
100.00%::
o co.
o.oog o oo
.65:!!
n.,3!!
!! io:::::!!
APERTURE s.12:::
79.7sx::
100.005 '
0.00 8::
0.12%::
5.471 :
22.971 :
0.10:::
100.00%;f.
o. o o..o.
29.19:::
ss.64x::
200.00s::
Also Avallebts o:1 72.12:::
26.282::
1.60%::
100.002::
49.4s:::
50.52:::
100.00:::
Aperture Card 0.00...
51.71:::
36.242;:
100.00:::
0.00...
0.00..
0.00%
- 0. 0 0.,..
0.00..
0.00~.
0.00%.
0.00 oc 0.00%
o.
o o
- 0. 0 0.~
o 0.00..
9.35:::
52.35:::
100.00:::
54.27:::
42.87:::
100.00:::
o.
o 3
0.00%.
0.00%.
3 0.16:::
5.14:::
72.575 :
1.071::
100.00 ::
0.112::
0.96:::
54.20:::
0.30s::
0.30:'
1.111:
100.002::
0.00 0.00.
o.onz.
0.00%
0.00..
o g
0.00:::
o o
0.00-0.002.0.00..
o o
0.00..
o 3
l 0.00%
0.31x;;
0.s7x::
2s.41:::
1.sn::
3.60%::
100.00%;;
0.21x::
0.2st;;
29.78:::
1.06t::
0.21t::
2.06t::
100.00 ::
4.ss:::
78.331::
200.00:::
0.00%::
0.00%.
0.00:::
0.00x::
0.00 0.00..
0.00:::
0.00:::
0.00x.
0.003.
. nn.
W.-w*
g 6
6 6
=.__nn. ~~-
~'
"eg qm e g e ee M *
- se e'# 4 #*
v} eleMagee h,
- { g*.
O aE *,I
l l
1 r~N I
t g
b TA13 LE 4.2 ESTI.\\1ATED RADIONUCLIDES INVENTORY IN ACTIVATED CO.\\lPONENTS aj l
i l
1 l
n..
c.
c...
i.n
......... c...,,
n, n.
...,..:.i......-i.-.
,,,.s....,i....n.......
{
n.
c............
.i....,.....
..n...
ia. ~i... - n......i.
...~i. -
.u..... no.,
i,...~i.i-..:.....,
...io.....,
n 3
...n.....n....
.....n..........
......n.io.....
f_.%
4
... i n..,
i t
.....,.....1 l
i,
. a,.
...u n
3.n 1
i l
(a) i All curie estimates are based upon decay until January 1,1991, t
(b)
Includes 27.2 curies of 109Cd,27.2 curies of 109 mag.13.6 curies of 10SmAg, and 5.48. curies l
(,
of 123mTe, among other isof. opes.
1 1
i l
l
['A 1
i
\\
t\\
(
r i
INTERIM TERMINATION SURVEY PLAN AND REPORT b
5.0 INTERIM SURVEY PROCEDURES
%/
5.1 Intent-of. Survey 5.1.1 Scope Interim surveys are limited in scope to areas that were remediated during the proj ect and could not be maintained accessable for the final survey.
Specifically, the hold-up-tank (HUT) and the SK tank excavations are such areas.
Peripheral areas which could affect the future integrity of the radiological status of the interim-survey areas subsequent to the time of the interim survey.
Specifically, land and structures within the radiologically controlled area (RCA) south of the HUT, and the
's pile from the SK tank excavation are peripheral areas that 4
w2n
,e used to backfill the interim survey areas.
5.1.2 Purpose The purpose of this interim survey is to produce data that will
' demonstrate that these surveyed areas meet the criteria for site release, such that they can be backfilled prior to conduct of the sitewide final termination survey.
The final termination survey will be conducted at a future time upon completion of D&D activities.
Therefore, while the interim survey areas may be deemed as meeting release criteria at the present time, that j
, ('
status will be verified upon the completion of remaining site D&D activities, if appropriate, via a reduced scope final termination survey.
To insure that the interim-survey areas remain in an "as
)
surveyed" condition during the remaining D&D work on other i
portions of the site, they will be isolated from the active areas l
of the RCA, their use will be controlled subsequent to the interim survey, and their radiological condition will be verified at the time of the sitewide final termination survey.
This will be accomplished as follows:
l Isolation of Interim Surveyed Areas i
Physical' barriers will be installed where possible to prevent i
i uncontrolled movement of personnel, equipment,
- material, or surface water runof f originating f rom the active D&D areas from entering the ' interim surveyed area.
A chainlink fence, with 4
. lockable
- gates, will be installed north of the pump
- room, extending between the east wall of the reactor building and the east side RCA security fence.
Personnel, equipment and material will not be allowed entry into the interim surveyed area without a specific need, management approval and the personnel and/or items designated as non-contaminated by standard HP procedure.
A blacktop berm will be installed on the driveway north of the pump room to divert RCA surface water run-off f rom entering the HUT (V
area.
Openings in the pump room south wall (which are open to the HUT excavation) will be sealed with solid coverings, such as l
JJM/83.94B Page 26
INTERIM TERMINATION SURVEY PLAN AND REPORT O
- plywood, steel plates and/or caulking compound.
This will preclude any potential for migration of radioactive material that could be in the pump room, and block personnel and/or equipment from by-passing approved HUT excavation / area entry protocol.
Similar precautions will be taken to protect the integrity of the SK tank backfilled excavation.
Interim Survey of Peripheral Areas Peripheral areas around the excavations that f orm the watershed for the HUT excavation will be included as part of the interim survey.
This will alleviate any concern about radioactive material entering the HUT excavation via surface water run-off from contaminated surfaces.
The peripheral areas include the i
following areas and/or structures:
1 1
o Reactor building roof I
o East, West and south reactor building exterior walls l
Concrete spillway west of the reactor building o
o Pump room roof o
Pump room south exterior walls Reactor water storage tank foundation and excavated fill pipe o
trench o
Reactor secondary water cooling tower o
Asphalt driveway south of HUT l
o Land areas inside RCA south of HUT l
Verification of Interim Survey Status Upon completion of all D&D activities, a
sitewide final termination survey will be conducted to demonstrate compliance with release criteria.
Data generated during the interim survey will be used to demonstrate compliance for those areas within the scope of that survey.
However, the radiological status of these areas must be verified at the time of the final termination survey, since the possibility exists that these areas could be affected by the on going D&D
- work, however unlikely.
Verification will be accomplished by a limited scope radiological survey (see Section 5.1.4).
5.1.3 Intended Use of Interim Surveyed Area t
After completion of the NRC/ORISE confirmatory survey at the convenience of Cintichem, the excavations may be backfilled with material from the peripheral areas that were included in the l
interim surveys.
Access for personnel, equipment and materials will be controlled as indicated in section 5.1.2.
The only D&D related activities that could be envisioned in this area would be the emplacement of demolition related equipment around the periphery of the reactor building or use of roadways through the areas for access to the east side of buildings 1 and 2 at elevation 804' or to the reactor tunnel at elevation 780'.
JJM/83.94B Page 27
INTERIM TERMINATION SURVEY PLAN AND REPORT 9
5.1.4 (O
Use of Interim Survey Data l
At the time of the sitewide final termination survey, the data from the interim survey will be verified as still being representative of the excavations and peripheral areas.
Once accomplished, the verified interim surv=y data will be used to represent the radiological condition of the HUT excavation, the SK tank excavation, and peripheral areas.
Verification will entail conducting a radiological survey of limited scope to generate the same types of data as the interim survey.
These measurements will be of a spot-check type to compare with the findings of the interim survey.
The verification survey will address from 1
to 10%
of the interim surveys measurement locations, with possible additional locations surveyed if there are areas that have been disturbed or otherwise potentially affected by on going D&D activities.
5.2 Remediation Work Prior to the Survey 5.2.1 Hold-up Tank / Primary Coolant Storage Tank The Hold-up tank (HUT) and the primary coolant storage tank and its associated piping contained reactor primary coolant :.ater.
l Tests of the HUT in 1990 indicated that some leakage into the surrounding soil had occurred.
The storage tank was leak tight f7 at the time decommissioning commenced but the task was planned
('~)
around the potential presence of contamination in the soil under the tank and/or along the piping that connected the tank to the pump room.
The interior of the HUT was initially decontaminated by scabbling the ceiling, walls and floor.
Pipe penetrations and obvious cracks in the wall between the HUT and the pump room were removed by core drilling or by sawing.
Approximately 30' of overburden was removed from above and around the tank so that the east, south and west sides were exposed down to the floor level of the tank.
The roof and south wall of the adj oining pump room were also uncovered in this excavation.
During this excavation continuous surveys were done as described in the decommissioning plan supplement (s).
Contaminated soil was not encountered until the excavation reached the roof of the HUT.
All contaminated soil was
- removed, placed in proper shipping containers and disposed of at a licensed facility.
The roof and wall of the HUT were sawn into manageable sections that could be removed for further decontamination as necessary.
The primary coolant pump suction line and sump were dug out from the floor of the HUT and from under the 3' thick wall to the pump room.
Removal of the HUT floor was attempted but after doing core surveys of the concrete / bedrock interface on the underside I
of the HUT floor, it was deemed to be unnecessary.
Some saw I
m
(
)
curfs and hot spots were selectively removed instead.
v JJM/83.94B Page 28 l
l INTERIM TERMINATION SURVEY PLAN AND REPORT (v;
The aluminum primary coolant storage tank and accessory above-ground piping was dismantled in manageable sections for further decontamination.
The soil under the tank floor was found to be contaminated around the sump area in the bottom of the tank.
The soil was excavated down to several feet below grade until residual soil contamination was well below the soil acceptance criteria specified by license condition G (SNM-639).
Underground piping was uncovered while continuous surveys were performed.
No contamination was found around the piping except in the area where it entered the sump of the storage tank.
The pipe was segmented into manageable sections for further decontamination.
5.2.2 SK Storage Tanks Two SK storage tanks were located underground in the Mall area east of the reactor building tunnel at elevation 780'.
The tanks were mounted on saddles and a
concrete pad at elevation approximately 765'.
These tanks were used as storage of processed waste water from buildings 1,
2 and 4.
Water was stored in these tanks so that it could be sampled and assayed for radioactivity prior to release to the environment.
Although most of the water contained in these tanks met unrestricted release criteria (10 CFR 20, App. B Table II), these tanks were treated as if contaminated internally due to residual (3
fixed contamination buildup that had occurred over the (j
operational life (34 years) of the system.
The tanks were excavated by removing soil in layers at approximately 1 foot increments and continuous surveys were performed as described in the Decommissioning Plan supplements.
Some contaminated soil was removed at the ground surface (7 80' level) around the access l
manhole for the north tank and some contaminated soil was removed f rom between the tanks adjacent to the manhole concrete casements at about elevation 770'.
The remainder of the excavated soil was uncontaminated.
Inlet and discharge piping and valves were removed to the extent of the tank excavation, pipes were plugged l
and the ends of remaining pipes will be lef t uncovered when the 1
excavation is backfilled following the termination survey.
l l
5.3 Interim Survey Overview i
5.3.1 Survey Objectives The purpose of this interim survey is to demonstrate that the
{
radiological conditions of the excavation for the Hold-up Tank I
(HUT) and primary coolant storage tank, the SK tank, and the respective surrounding environs satisfy NRC guidelines for j
unrestricted release and therefore, can be isolated from the existing controlled area and backfilled.
The data produced will also be used in conjunction with future data from the final survey to allow the entire site to be released from regulatory
/s) control.
The specific objectives of the survey are to show that:
\\J f
JJM/83.94B Page 29 I
L______
INTERIM TERMINATION SURVEY PLAN AND REPORT l
'}
A.
Surface Activity of Buildings, Structures and Bedrock v'
l.
Average surface contamination levels for each survey unit are within the authorized values (refer to Section 5.3.2) 2.
Small, areas of residual activity, known as " hot-spots" do not exceed three times the average value.
The hot-spot limit applies to areas of up to 100 cm'.
The average activity level within the 1 m' area containing a hot-spot must be within the guideline for the average level.
3.
Reasonable efforts have been made to clean up removable activity and removable activity does not exceed 1,000 dpm/100 cm'.
4.
Exposure rates'in occupiable locations are no more than an average of 5 uR/h above background.
Exposure levels are measured at 1 m from surfaces and are averaged over the survey unit / area.
Individual hot spots will not exceed 10 uR/hr at one meter.
B.
Volume Activity of Soil
,em 1.
Average radionuclides concentrations are within the
(')
authorized limits (see Table 5.1, Soil Criteria).
Averaging is based on a 100 m8 grid area.
2.
Reasonable efforts have been made to identify and remove hot-spots that may (exceed the average limits by greater than a factor of 100/A)h, where A is the area (in m ) of the hot spot, or three times the average 8
limits, whichever is less.
3.
Exposure rates do not exceed 5 uR/h above background at 1 m above the surface.
Exposure rates may be averaged over a 100 me grid areas.
Maximum exposure rates over any discrete area of within the 100 m' area may not exceed 10 uR/h above background.
The above conditions will be demonstrated at a 95% confidence level for each survey unit as a whole.
5.3.2 Identity of Contaminants Based on the knowledge of site operations and the results of
- sampling, two distinct radionuclides contaminant mixtures exist for surfaces in buildings 1 and 2 (reactor and hot laboratory buildings respectively).
These mixtures, decayed to January 1, 1995, consist of the following:
,s
(
)
L.)
l JJM/83.94B Page 30 1
I CINTICHEM SITE SOIL RELEASE CRITERIA RADIONUCLIDES pCi/am H-3 815.4 Mn-54 3.3**
i Fe-55 521920 7 I
Co-60 0.9 l
i Ni-63 57971.0 i
Zn-65 3.7**
l Sr-90 17.4 l
Zr-95 2.5 Nb-95 2.3 Tc-99 1788.3 Ru-106 13.5 i
Ag-108m 1.1 Ag-110m 0.7 Cd-109 63.1 Sb-125 6.5 l
Cs-134 1.8 l
Cs-137 3.8 Ce-144 63.4 Eu-152 2.0 l
Eu-154 1.8*
g Eu-155 99.5*
U-234 30*
U-235 30*
U-238 30*
Pu-238 30*
Pu-239 30*
Pu-241 30*
Cm-244 17.1 l
- Pending NRC approval Excluded from sum-of-fractions rule l
TABLE 5.1
INTERIM TERMINATION SURVEY PLAN AND REPORT O
(a) Reactor Building
\\m/
Radionuclides
- Fractional Abundance Ag-110m 0.126 Ce-144 0.021 Co-60 0.072 Fe-55 0.637 H-3 0.019 Ni-63 0.046 Ru-106 0.016 J
Sr-90 0.002 Tc-99 0.061 (b) Hot Laboratory Radionuclides
- Fractional Abundance Ce-144 0.044 l
Cs-137 0.209 l
Fe-55 0.003 l
H-3 0.004 Ni-63 0.017 Ru-106 0.005 Sb-125 0.005 f
Sr-90 0.711 I
(_
Tc-99 0.001 Radionuclides reported comprising 0.1%
of total radioactivity, results are rounded to 3 significant figures.
For the purposes of this interim survey, where the HUT excavation and surrounding environs are of
- concern, only the reactor building mixture will be encountered.
On that basis the beta-gamma surface contamination limits for the area of concern are (see Attachment D for explanation of limits):
13,500 dpm/100 cm, averaged over 1 m8 8
40,500 dpm/100 cm8, maximum not to exceed 100 cm8 area i
1 1,000 dpm/100 cm8, removable l
The interim survey of the SK storage tark excavation will include the soil from and in 'the excavation and the concrete foundation in the bottom of the excavation.
The beta-gamma surface contamination limits (based on the hot lab mix) are:
1,3 01 dpm/100 cm, averaged over 1 m8 8
('~J
')
3,903 dpm/100 cm, maximum
(< 100 cm8 area) 8 s.
260 dpm/100 cm*, removable JJM/83.94B Page 31
INTERIM TERMINATION SURVEY PLAN AND REPORT O
Alpha contamination (U-235) has not been encountered to date and V
is not' expected _.
In the event that it were found, criteria from Reg. Guide.1.86 would apply:
5,000 dpm/100 cm*,
averaged over 1 m8 15,000 dpm/100 cm8, maximum not to exceed 100 cm8 area 1,000 dpm/100 cm*, removable Soil contaminant mixture makeup and ratios have been found to vary depending upon the particular leak source (e.g. HUT, canal, hot cell exhaust duct), as well as distance from the release point.
As such, no single concentration guideline for the mixture can be specified.
- Instead, individual site-specific radionuclides guidelines were developed by Cintichem and approved by NRC.
Table 5.1 presents this soil criteria.
The sum-of-fractions method will be used where a soil contaminant mixture meets "must not exceed" unity (exclusive of uranium and plutonium radionuclides).
Example:
if radionuclides A,
B and C
are present in concentrations Ca, Cb and Cc, and if soil guideline concentration are Ga, Gb and Gc, then the soil concentrations will not exceed l
unity for the following relationship:
i (Ca/Ga)~ + (Cb/Gb) + (Cc/Gc) <1 5.3.3 Organization and Responsibilities l
-The interim survey will be performed by qualified personnel from Cintichem's
- Health, Safety, Environmental ' Affairs Department (HSEA).
Logistical support will be provided by Cintichem's D&D Operations Department.
Figure 5.1 is an organizational chart for the survey activities.
l The survey organization : will be directed by the Manager of HSEA.
The: HSEA Manager will have the authority to make appropriate changes to the survey plan (subj ect-to established Cintichem procedural revision protocol) as deemed ' necessary as the survey progresses.
Field measurements, sample collection and sample analysis will be performed by the Health Physics Support and Environmental Monitoring groups from the HSEA Department.
j QA/QC responsibilities will be handled by the Project QA Manager.
i-The.
Proj ect Industrial Safety' Specialist, from the HSEA Department,. will provide industrial safety oversight for the survey. process.
Independent data review will be performed by members of the Healh Physics Staff'from the HSEA Department, that do not have direct responsibility for generating data.
i JJM/83.94B Page 32 m
1 INTERIM TERMINATION SURVEY PLAN AND REPORT
!/O Logistical Support will be provided by the D&D Operations group.
'V They will be responsible for gridding survey locations and providing scaffolding erection.
Qualifications of each key team member have been previously reviewed by NRC during routine periodic inspections.
5.3.4 Training Cintichem provides continuing training for its health physics personnel and other workers who may be exposed to radioactive materials.
Training varies according to potential exposure and the nature of the employee's job duties.
In addition to the regular training, special training will be provided on equipment, special techniques, and practices relative to the survey activities for those employees who will be involved in taking radiological measurements and samples.
All members of the survey team will attend an in-house orientation session reviewing radiation protection, survey procedures, and quality assurance activities.
Documentation will be retained in the Cintichem training files.
l 5.3.5 Laboratory Services
-Analytical services for gross alpha / beta levels on smears, and
- P gamma spectroscopy and Sr-90 analysis of bulk samples will be performed by Cintichem's environmental monitoring group in accordance with existing procedures.
These procedures have been reviewed by NRC during past routine inspections.
A contract laboratory, typically,
- Teledyne, Inc.,
will be used for wet chemistry analysis (other than Sr-90)- for soil and/or special bulk samples, when required.
0A/0C programs for both in-house and contractor laboratory services will be monitored by the QA Manager.
5.3.6 General Survey Plan This survey plan consists of systematic processes and procedures that have been deemed acceptable by industry standards and the NRC.
Activities (organized units of work needed to complete a function) have been defined and tasks (specific work assignments within a
specific activity) have been delegated to. the appropriate team members.
Table 5.2 provides a breakdown of activities and tasks that are currently a part of the termination survey plan.
Tasks will be performed in general accordance with guidelines stated in the Manual for Conductino Radiological Survey in Support of License Terminatio _, NUREG/CR-5849.
n tO V
l l
JJM/83.94B Page 33 t________________
DRAFT 4
I
TABLE 5.2 (m,
OVERVIEW OF MAJOR ACTIVITIES AND TASKS l
ACT IVITIES TASKS i
Evaluate contamination
- 1. Review operating history with potential respect to facility use, spills, releases etc.
l
- 2. Review radiological data from
{
scoping, characterization and D&D progress surveys.
- 3. Identify radionuclides of concern and determine guidelines.
- 4. Classify areas as to "affected"
{
and " unaffected".
Establish grid reference
- 1. Install grids.
q system 2
Prepare facility survey maps.
l Determine background
- 1. Measure indoor exposure rates and levels ambient beta-gamma levels.
- 2. Measure outdoor exposure rates.
- 3. Collect background soil samples.
Perform direct
- 1. Conduct surface scans.
measurements
)
if w)
- 2. Determine frequency and locations
'/
of measurements to meet criteria.
{
i
- 3. Conduct surface activity measurements.
- 4. Measure exposure rates.
Collect samples
- 1. Determine f requency and locations of sampling to meet criteria.
- 2. Collect systematic and special samples.
Analyze samples
- 1. Count smears and swabs.
i
- 2. Analyze soil, paint, residue and l
other solid samples for activation and fission product.
l 1
Interpret data
- 1. Convert data to standard units.
- 2. Calculate average levels.
- 3. Compare data with criteria.
- 4. Compute total residue activity I
inventory.
Prepare report
- 1. Construct data tables.
- 2. Develop graphics.
- 3. Prepare text.
- 4. Submit report to NRC.
~~
N) l
INTERIM TERMINATION SURVEY PLAN AND REPORT f) o Section 4.0 -
Planning and Designing the Final Status
(
Survey o
Section 5.0 -
Radiological Instrumentation o
Section 6.0 -
Survey Techniques o
Section 7.0 -
Samples Analysis o
Section 8.0 -
Interpretation of Survey Results l
5.3.7 Tentative Schedule Field activities for the interim survey of the HUT and surrounding areas began in August 1994 and will be completed by the end of October, 1994.
5.3.8 Survey Report l
A report, describing the survey procedures and findings, will be
_ prepared and submitted to the NRC.
Report format and content will follow the recommendations contained in Manual for Conducting Radiological Surveys in Support of License Termination, NUREG/CR-5849.
5.4 Survey Plan and Procedures i
l 5.4.1 Instrumentation j
(a) Reactor Mixture Surveys Table 5.3 lists the instrumentation to be used for the survey activities, along with typical parameters and detection sensitivities for the instrumentation and survey technique for I
both the reactor and hot laboratory ' mixtures.
The reactor I
l building radionuclides mixture is shown in section 5.3.2.
The l
radionuclides Fe-55, H-3 and Ni-63 comprise about 70% of the radioactivity in the mixture, and are not detectable with the field instrumentation being used.
The other 30%
of the radioactivity in the mixture, Ag-110m, Ce-144, co-60, Ru-106, Sr-90 and Tc-99, are detectable with the field instrumentation being used.
For every radiation event detected (counted),
2.33 radiation events are not counted.
Therefore, one count represents 3.33 counts.
As such, a correction or scaling factor must be applied to a net count rate in order to account for all detectable and non-detectable radionuclides present.
This factor, for the reactor building mixture, is a multiplier of 3.33 l
or equivalently a
divisor of 0.3.
The combination of instrumentation and technique were chosen to provide a detection sensitivity for direct survey beta-gamma measurements of 25% or less of the guideline levels for beta-gamma emitters (3,375 dpm/100 cm8).
The basic equation for determining field instrument direct survey detection limits will be:
i OV 1
1 JJM/83.94B Page 34 L - -- ------------- ----- - - --- --
INTERIM TERMINATION SURVEY PLAN AND REPORT MDA
= ((2.71 + 4.65)/ background)/(count time x efficiency x
,U (probe area /100) x scaling factor) j 1
- (NOTE:
The reactor building radionuclides mixture contains radionuclides that cannot be readily detected by field i
instrumentation.
These radionuclides comprise 70% of the mixture (e. g.
As such, a scaling factor of 0.3 in the denominator is used to correct the in-field count rate.)
Sensitivities for scanning techniques are based on movement of the detector over the surface at 1 detector width per second and i
use of audible indicators and count-rate meter to sense changes in instrument count rate.
Data obtained experimentally with the i
equipment and technicians that will be used on this interim survey indicates that qualified surveyors can detect 82% of the j
guideline value (11,111 dpm/100 cm8) with a 90% confidence level.
All instruments will be - calibrated a minimum of once very 3
- months, using NIST-traceable standards.
Calibrations of beta detection instruments will be made with an even mixture of Tc-99 and Cl-36, which most closely resembles the average beta energy for the radionuclides mixture present in the reactor building.
The reactor building radionuclides mixture has an average beta energy of 154 kev for the " detectable" fraction.
An even mixture of Tc-99 and Cl-36 has an average beta energy of 168 kev.
G Operational and background checks will be performed at least once Q
each 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> on instrument use.
(b) Bot Laboratory Mixture Surveys Table 5.3 lists the instrumentation to be used for the survey activities, along with typical parameters and detection sensitivities for the instrumentation and survey technique for both the reactor and hot laboratory mixtures.
The hot laboratory building radionuclides mixture is shown in section 5.3.2.
The radionuclides Fe-55, H-3, Tc-99 and Ni-63 comprise about 2.6% of the radioactivity in the mixture, and are not detectable with the field instrumentation being used.
The other 97.4%
of the radioactivity in the mixture, Ce-144, Cs-137, Ru-106, Sb-125 and Sr-90, are detectable with the field instrumentation being used.
For every radiation event detected (counted),
0.027 radiation events are not counted.
Therefore, one count represents 1.027 counts.
As such, a correction or scaling factor must be applied to a net count rate in order to account for all detectable and non-detectable radionuclides present.
This f actor, for the hot laboratory building
- mixture, is a
multiplier of 1.027 or l
equivalently a
divisor of 0.974.
The combination of j
instrumentation and technique were chosen to provide a detection sensitivity for direct survey beta-gamma measurements of 25% or less of the guideline levels for beta-gamma emitters (325 dpm/100 cm ).
8 V
JJM/83.94B Page 35
i i
INTERIM TERMINATION SURVEY PLAN AND REPORT 1
i O
The basic equation for determining field instrument direct survey i
detection limits will be the same as used in iten (a) above.
j Sensitivities for scanning techniques are based on movement of the detector over the surface at 1 detector width per second and use of-audible indicators and count-rate meter to sense changes in instrument count rate.
Data obtained experimentally with the equipment and technicians that will be used on this interim survey indicates that qualified surveyors can not detect less than the guideline value with a 90% confidence level.
Scanning will be performed, however, with a minimum sensitivity of 2,764 dpm/100 cm (2.1 times the. average guideline).
As such, the 8
frequency of direct measurements will be increased (5 pe r m 8 ).
1 All instruments will be calibrated a minimum of once very 3
- months, using NIST-traceable standards.
Calibrations of beta detection instruments will be made with an even mixture of Sr-90 which most closely resembles the average beta energy for the radionuclides mixture present in the hot lab building.
Operational and background checks will be performed at least once each 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> on instrument use.
(c) Data Correction In-the-field survey instrument readings (counts per time) will be corrected for instrument background, detector efficiency, non-
- O detectable radionuclides, and probe area
- size, to produce a
result in dpm/100 cm* (as previously.shown in item (a) above).
dpm/100 cm8
((gross counts / count time) - (background counts /
=
count time)) x (1/ efficiency x non-detectable scaling factor x probe area /100)
This dpm/100 cm8 value will then be corrected for the specific surface materials natural background.
This will be accomplished by subtracting an average background value that will be determined for each type of material (rock, concrete, asphalt, wood, etc.) in dpm/100 cm8 (see section 5.4.3).
5.4.2 Survey Plan (a) Area Classification For purposes of establishing the sampling and measurement frequency and
- pattern, survey areas have been divided into affected and unaffected areas as indicated in Table 5.4 and shown l
on Figure 5.2 The bases for these classifications are:
o affected areas: Areas that have potential radioactive contamination (based on plant operating history) or known radioactive contamination (based on radiological characterization and/or measurements made during D&D O_
operations).
This includes areas where radioactive materials were used and stored, where records indicate spills or other unusual occurrences that could have resulted in spread of contamination, where radioactive JJM/83.94B Page 36
I INTERIM TERMINATION SURVEY PLAN AND REPORT I
i rx
(
materials leaked f rom systems or structures and where V'
decontamination work has been performed.
Areas i
immediately surrounding or adjacent to these locations are included in this classification because of the l
potential for inadvertent spread of contamination.
o unaffected areas: All areas not classified as affected.
These areas are not expected to contain residual radioactivity, based on a knowledge of site history and l
previous survey information.
i Table 5.4 lists the various survey units and areas for the HUT area surveys and the classification of each.
i (b) Reference Grids Grids will be established for the purpose of referencing locations of samples and measurements, relative to building and other site features.
The gridding intervals are based on the potential for residual contamination in the various plant areas (i.e.
affected or unaffected area classification)
(See Table 5.2.)
All affected area surfaces will be gridded at 1 m intervals; Building surfaces in unaffected areas that have not been contaminated as a result of prior activities will not be gridded; measurements will be referenced to other grid systems or
\\]j
(
to prominent building features.
Affected outdoor land areas will l
be gridded at 10 m intervals; unaffected areas will not be gridded.
The facility will be divided into " survey units" having common
- history, contamination potential, or that are naturally distinguishable f rom other site areas.
These survey units will be sized to assure a minimum of 30 measurement locations each for floors and lower
- walls, other vertical
- surfaces, and other horizontal surfaces.
Unaf fected areas identified by scans or direct measurements or as exceeding guidelines will be reclassified as affected areas and will be gridded and resurveyed accordingly.
(c) Surface Scans Scanning of surfaces to identify locations of residual surface and near-surface soil activity will be performed according to the following schedule:
o Affected Area Structural Surfaces - 100% of surface, beta and alpha radiations o
Affected Area Land Surfaces 100% of surface, gamma radiations (n!
LJ f
f i
JJM/83.94B Page 37
INTERIM TERMINATION SURVEY PLAN AND REPORT l
o Unaffected Area Structural Surfaces - M ne V
o Unaf fected Area Land Surf aces - 10%, gamma radiations Building surface scans will be conducted for alpha and beta radiations.
Scans of exterior building and paved surfaces will be for beta and alpha radiations.
Soil and bedrock surfaces will t
be scanned for gamma radiations only.
i Instrumentation for scanning is listed in Table 5.3.
Scanning speeds will be no greater than 1 detector width per second for alpha and beta detection instruments and 0.5 m pe r second for i
gamma instruments.
Audible indicators will be used to identify locations, having elevated levels.
All scanning results will be noted on standard field record forms; locations of elevated radiation will be identified for later investigation.
Table 5.4 l
lists the specific areas and frequency of surface scan
)
measurements.
(d) Surface Activity Measurements f
j Direct Measurements f
Direct measurements of alpha and beta surface activity will be
]
performed at selected locations using instrumentation described i
f' in Table 5.3.
Measurements will be conducted by integrating
~
l counts over an 0.2 or 1.0 minute period for the reactor and hot s
laboratory mixture respectively.
Direct surface activity measurements will be systematically placed using grids with 1 m intervals on all structural surfaces of affected areas.
Direct measurements, where the reactor mixture is present, will be at a f requency of one per m8 Direct measurements, where the hot lab mixture is present, will be at a frequency of 5 per m8, due to the inability to scan with a sensitivity less than the average guideline value.
l On surfaces of unaffected
- areas, a
minimum of 30 random
. measurements or an average measurement of 1 per 50 m8 of survey unit area, whichever is greater, will be performed for each survey unit.
These locations will include the roof and the above grade walls of the reactor building which are adjacent to the HUT excavation and the cooling tower, as indicated in Table 5.4.
j Removable Contamination Measurements i
I Smears for removable contamination will not be performed at each measurement
- location, as these are all outdoor weathered surfaces.
OV l
JJM/83.94B Page 38 l
E_____________________._.________._.
INTERIM TERMINATION SURVEY PLAN AND REPORT rO (e) Exposure Rate Measurements I
Gamma exposure rates will be measured at 1 m above ground, floor I
or wall surfaces, using a gamma scintillation instrument shown in Table 5.3, calibrated for Cs-137.
Measurements will be uniformly spaced according to the f ollowing f requency:
Structural Surfaces Affected Areas:
1 measurement per 4 m8 Unaffected Areas:
30 measurements per area Grounds Affected Areas:
5 measurements per 100 m8 grid block Unaffected Areas: 30 measurements per area j
Table 5.4 shows the specific measurement frequency for each survey area.
(f) Soil / Sediment Sampling i
Surface Samples (about 500 grams each) of surface soil (0 - 15 cm) will Oi be systematically collected f rom the center and 4 points midway
)
between the center and the block corners for each 10 mx 10 m grid in affected areas.
Thirty samples will be obtained from l
random locations in unaffected areas.
Additionally, sediment will be collected from inside storm drain basin and f rom other natural surface drainage pathways from the affected areas.
(g) Special Measurements and Samples The following lists additional measurements and sampling that will be performed as part of the interim survey.
4 The roof of the reactor building is constructed of concrete, covered with a rubber membrane which is, in-turn, covered with loose gravel.
Thirty direct beta-alpha measurements will be taken on the gravel and roof membrane as previoasly described.
Sediment samples (rock fines and organic matter) that has settled out under the gravel layer, will be taken from 5 10 of the i
direct measurement locations.
Sample locations will be selected from those with the highest direct survey results or at random if no elevated areas exist.
Additionally, direct beta-alpha l
measurements and sediment samples will be taken from roof drain / overflow spouts.
Portions of the pump room exterior south wall and below grade walls of the reactor building are tar covered.
Tar samples will s
[x be taken f rom 5 - 10 of the direct measurement locations for each survey unit.
Sample locations will be selected from those locations with the highest direct survey results or at random if JJM/83.94B Page 39
INTERIM TERMINATION SURVEY PLAN AND REPORT (m) no elevated areas exist.
Gamma spectral analysis on tar samples i
\\d will be qualitative as this material is difficult to prepare for analysis.
The concrete floor of the hold-up tank sits directly on bedrock.
A minimum of 30 concrete cores will be drilled to the concrete-bedrock interface.
Direct beta-alpha measurements will be taken on the cores' bedrock and concrete surfaces.
If elevated results are found, gamma spectral analysis will be performed to identify radionuclides present.
Samples of the bedrock, that surrounds the HUT excavation on the west, south and east sides, will be taken from locations with elevated gamma exposure rate and/or elevated direct contamination measurements.
These samples will be analyzed by gamma spectral analysis to determine if elevated results are due to high background levels of Th-232+D or man-made contamination.
The cooling tower contained secondary reactor cooling water, and therefore is not expected to be an affected area.
In addition to the 30 direct measurements that will be taken to verify this, samples of its wooden pipes, plastic slates and wooden framework will be taken and analyzed by gamma spectral analysis for the presence of radioactive contamination.
(]
5.4.3 Background Level Determinations Background exposure rates will be determined for the outdoor areas by taking a minimum of 30 gamma scintillation microrem measurements at unaffected locations of similar terrain offsite.
Results of background exposure rate will be evaluated to assure that the averages determined are representative of the true averages, using procedures described in NUREG/CR-5849.
Background
soil samples have been previously collected from offsite locations to determine fallout fission product concentrations.
Soil fallout concentrations have been determined to contain Cs-137 at 1.25 pCi/gm at the 90th percentile.
This value has been accepted by NRC as part of the soil criteria review process.
Direct beta-alpha backgrounds will be taken from representative materials (e.g.
bedrock, concrete) and determined and tested as specified in NUREG/CR-5849.
Additional sampling or measurements will be performed if necessary to satisfy criteria.
l Bedrock and/or soil in the HUT excavation as well as bedrock / soil j
at other locations on site are known to contain deposits of natural thorium ore.
These localized ore deposits may cause direct measurements of beta-gamma and alpha surf ace contamination as well as gamma exposure rates to appear to exceed free-release s
(
)
criteria, even when corrected with average background values.
It V
will therefore be necessary to distinguish "real" contamination l
from natural hot spots.
Locations where direct surface i
f JJM/83.94B Page 40 I
L
INTERIM TERMINATION SURVEY PLAN AND REPORT O
contamination measurements appear to be greater than expected average background levels, or release criteria for the material in question, will be sampled and analyzed by gamma spectroscopy.
If the sample result indicates the presence of thorium ore without non-natural radionuclides, the high readings will be dismissed as background.
If non-natural radionuclides are detected, part or all of the high result will be attributed to "real" contamination, dependent upon the ratio of natural and non-natural radionuclides' and the associated decay schemes.
A similar approach will be taken with higher than expected gamma exposure rate results.
When this occurs, gamma spectrum will be taken at the same location with a portable intrinsic germanium detector and multi-channel analyzer.
If the resulting gamma spectrum indicates the presence of Th-232+D without non-natural photopeaks, the higher readings for that area will be attributed to natural background.
If non-natural photopeaks are identified, part or all of the high result will be attributed to "real" contamination, dependent upon the distribution of natural and non-natural photopeaks and the associated dose conversion factors.
5.4.4 Sample A'alysis Smears and swabs for removable contamination will be analyzed for N
' gross
- alpha, gross beta activity.
- Soil, sediment,
- gravel, s
roofing material and other large volume samples will be analyzed by high resolution gamma spectrometer and Sr-90 analysis.
Non-detectable radionuclides (Fe-55, H-3, Tc-99, Ni-63, U,
Pu, etc.
will be calculated on the basis of previously determined radionuclides ratios for reactor mixture and/or hot lab mixture as appropriate.
If found, samples with sufficient activity will be sent to a
contract laboratory for analysis of these non-detectable radionuclides to verify radionuclides ratios.
Existing NRC
- reviewed, laboratory chain-of-custody procedures will be observed for all sample analyses.
5.5 Data Interpretation Measurement data will be converted to units of dpm/100 cm8 (surf ace activity),
uR/h (e xpos ure rates) and pCi/g (soil concentrations) for comparison with guidelines..
Values will be adjusted f or contributions f rom natural background.
Individual l
measurements and soil levels will be compared with " hot-spot" criteria.
Average values for survey units 'will be determined and i
compared with guideline levels.
Data for each survey unit will
)
l be tested-against the confidence level obj e ctive, using guidance and procedures described in NUREG/CR-5849.
l Additional remediation and/or further sampling and measurements j
i will be performed where guidelines are not met or cannot be demonstrated to the specified level of confidence.
Computations and comparisons will be repeated, as necessary.
JJM/83.94B Page 41 u_ _ __-_ -.
INTERIM TERMINATION SURVEY PLAN AND REPORT O
-The average levels will be used to estimate the total residual V
. non-background radionuclides inventory of the HUT and surrounding environs.
5.6 Report A report, describing the procedures and findings of the interim status survey will be prepared and submitted to the NRC.
Data will be summarized in tables.. Measurement and sampling locations will be shown on scale drawings.
All field and analytical data will be archived by Cintichem until such time as the NRC authorizes disposal.
l O
%)
1 l
l O
L JJM/83.94B Page 42 L
f G
/I LH AP I
l'. I' l I l CA IR RG ELC l I I l I I l
I I S
W B
C IS A
N TC LSI AI l
I l I I I-RT PG, PI l
CIE ES AVN Om RRY TED DI NSE 0
L D
0 ET l
i i l
l l
I l l I I I I I I R
S E
N E
A B
I N
A A
l C
i M
l i l L
I IN r
C S
G 3
Q M
E
/
T I
A I
P Q
V H
i E
I i I I
R F
I l l i I l
~
F A
A T
T 1
l l
l I l A
S 1 I I I I I D
T S
N B
N A
A L
E I I 1 A
A L
I PR S
C
/ E H
l I
N I
TEil EN E
I l l l l I 1 I I l
G.
N R
G N
I DA IM E
SO E
CP T
S N
I E
A S.
R M
YV P -
HN I l 1 I I il 1 I
I' i
I PE H&
l I l l I I lUT S
N1 AP Y
S x'
\\
x x'
c s\\
N' N'
-jb x'
N s.
_ N~ \\
_ N l
l i
!i I
ji ll ll1 4
J_
v V
V v
V u
V V
v v
O 0
0
~
r.
e S
4 3
e 0
0 0
+
+
+
9 0
t 1
1 l
1 0
7 0
+
+
+
6 1
+
0 0
A A
A A
A A
A A
A
+
o o
I T
T T
1 N
S S
S 5
T T
A T
A T
r S
S S
S T
.i S
S f[Js fi !I b2 7
G a
Ys D
,-cW
't, Te L
c4 D
6 NSx/
I 9
g E
R G
//4 X
IG2 s2.
ll
!l
.l[
,l l
D l
g Cv/1 J
D_
r Io9 V
/t rO Y/f.,.
H g
R TU m...,
!*j p
ff_.
O 92-a g
S gf
.m
"{
R
-r -
O4 4
RD
[
4 O
Tn A ' W_. :
g
- V,pp_
ll j
c]L I
RI
~
Ce
[cI +5.
- i 7R z.
An
~
=
EO 1
x I
d R
i Y
y-Rr
.V h
m_
L '
.L G
h^~
L m_
D V
y I
ain P
.I L-ROrH l,
!l rs
=
_lI B-i
- f IUR R
h h
r PI l
i2 tam XDU
~
l C
, 3I
[ "j'-
- % ' l g
i u
/
s u
wi $
9 y
4 O=
s X
4 X
5 R
C 5
v
/
P G
Pa TC A
D Ah I
e &$
I RID RL E
G O
R N
ET 1
l A
' ]
jl k
2 Y
E X
.E F
- 5 V
c A
H C
TL 4
4 R
R R
E U
R UL R
R g 'l $
g/
A OA
.O S
Lv IP U
A D
,- glW t g
A e
ll Vx TP {
C R
I F
R T
C X
A IN 4
K Cr R
Cr D
E C
r C
K 4
Ou H T
T RH TL N S 7/
r NN D
UL ID i
L AO gr Cr OA L
A TI Do SV N
C T
- a
- 2!$
C I
[l OU CA jG GV!
g O
A L
R A I
TX B
D OC
= m D
SC A
4 g
G Y
R D
E 0
L V
id i4 ji?; g lC4 l )[
\\
1 *,
B R
9 U
S h
X
/3'
)'
P C
O g#/p s'
R I$
P
/
N O
+
4 A
O g
- c.'
%sd 4
=
s MI $
/
f
+
=
+
/
g N
/
=
T w
L
)
A
/gc *
/
H
=
o P
~
UI J
. /
S g
A
=
Y
=
Cg
//
4
&/' '
V f
i
/
C
/
=
V I
8/'-
I e
R
~
=
D
.,y hE $
=
y,f
=
Mrv n
nc
/
/r c h
oi p
p n
ytt d
d%
y pn d
n n r ci 0
d di e
pp oes 5
49 m
R dd ttn 7
6 8
2 A
u A
aee 3
7t 0
01
/
42
/
e
)
rDS 4
2a 2
2(
N 2
12 N
o
(
b
^
aL 6
6 t
2 2
a nra s o
)
=
=
H ae s gi
(
0 0
e g
i 9
9 68 i hr Pf A
A A
rt e f
6 r
6 r
/
/
56
/
ain 4 e 1
S 1
S N
N 33 N
vwe 1
)
)
b
)
1 a
(
)
c
(
c
(
2 y
(
"n )t 2
mt n
cn 2
c 0i m
an 0
0a c
0u
'y 0
1t 0
0 o t
1
/r 0
1 c e
i
/
ne 1
/
r rrv n
yc
/
n n h
uoi p d n
yi
/
ttt d
p dm m
nn xci 10 d
e pp i es 5
19 72 n
dd Mt n 7
1 8
6 A
u A
)
ee 3
1t 0
00
/
42
/
e rDS 4
1a 2
4(
N 2
12 N
(
o t
ca
=
=
e R
6 6
3 3
)
l l
C C
i
/
/
(
i 9
9 68 a
Pf 9
9 A
'A A
m f
6 c1 6
c1
/
/
56
/
m 1
e 1
T2 1
T2 N
N 33 N
a 4
g n
n p
p h
t c
c i
n n
ws n
u p
5r un p
e 6
p 6
p c
1h pp c
si 3
c 3
c d
cc eg 0
ir 0
g 0
0 0
e 25 0
re k
4 4
5 0R 5
an B
0 5
O 5
1 1 u 02 1
ve s
s a
n n
n h
a a
a i
p t
c s
c t
a l
e s
e s
y n
t a
a b
t s
e i
t t
t t
a m
l e
yN e
a b
t r
i u
s s
N y
s n
n
~
a a
a e
c c
M s
s\\
f e \\
o caA e.
f p
eE ys T
hj38 Ol t
(N
d di pp oes 5
49 m
e P
dd t t n 7
6 8
2 A
u A
aee 3
7t 0
01 W
)
rDS 4
2a 2
2(
4 2 V
e o
2 1 2
(
I ba L
6 6
2 2
a t
n o
)
=
=
r s re Il
(
0 0
s gi i
9 9
e g
Pf A
A A
rt e 68 ihr f
6 t
6 r
U 14 e 1
S 1
S F
V 56 V
ain I
33 vwe t
)
b I
)
)
a
(
c
)
(
(
c
'n ty
(
"a an c
0i
'a
'n )t cn 0
0 a c
0 u
_ y 0
1t 0
0 o t
1
/r 0
1 c e i /
ne 1
/
r rrv n
pc /
nn h
uoi p d n
pi
/
ttt d
t y dm m
na 1
xci 10 d
e pp i es 5
19 72 R
dd Mt n 7
1 8
6 A
u A
ee 3
1t 0
00 V
42
)
rDS 4
1 a 2
4(
V o
P 2
12 (e
t S
tc Y
a
=
=
E e
V P
6 6
R O
)
3 3
U 1
1 C
C S
(
/
/
i 9
9 L
68 a
A Pf 9
9 A
A A
m f
6 t1 6
c1 1
e 1
72 1
T2 W
V 56 V
m C
4 P
33 I
a I
g G
n u
O p
p h
L c
c t
- 3. O n
i n
ws I
n n
p 5r u n p
e D
6 p
6 p
e 1h pp c si 5
3 c
3 c
/ cc eg RA 0
- m 0
ir E
g 0
0 0
e 25 0
re L
k 4
4 5
0R 5
an B R B
0 5
0 5
1 1 u 02 1
ve A O F
T N
O s
s I
a n
n n
h a
a T
a i
t t
c s
c A
t a
l e
s e
s y
T n h a
a b
t s
a N
e t
t a
a rs a
s E
m l
e y
y n
r ar n
s M
e a
b t
t n
t ea n
ay U
r i i a
e mc a.
t u
s s
v v
gc r
s gc yt R
s n
n i i t
u s
t t e T
a a
a t
t ee s
a ee i n S
e c
c c
v o
h a v vo N
M s
s a
a i,
p pt i,
i e t s x
l e t s t g I
f e
e e e ae e
ab ae c
o c
c c
c t
l t p al a
a a
a ii a
ss i i m
e f
f f
f l p n
ss l p l
r r
r r
r a
rn oo a
i 0 y
u u
u u
uf a
r r uf o5 T
s s
s s
qo g
gg qo s2 O
hc 0
l 0
l e
e 1
e Tl n
r 5
n e
0 n o
B 0 n bn 0 a t
L 0 a an 2h c
2h La 2 cr e
c 2 cr h
e t
e e
ncr mez e
l a ez r
o e
ul y d e
tr l y e
rll i gl n
i gl t
caa d na e
n dna
)
u iuc uin e
b uir d
t Bds Lsa s
T Lsa
(
)
r ao w
tt r
o sl ec o
l eal e su nt f
id
,t oa s
e rl a
r ad r 'n h cl g
p ii pn B t ecl o n
l t
ai 2
.i wa n0( t r c on E0 a
3 es d o r
1 rl
,4 t
ni
,1
)
o r
ol I
ec it I
(
G d
t a
,hi am mi wr bn c
e0 pt Nu n
o e
l 0 sn o
t l
ma np r
e t
c1 oi 2d eg i o 1 c e bBhc xu P r Vr xi lm D
Ats 2L uo ' p l B I
t I
D I_
1 2
3 4
5 6
F e
i ADS
)
r o
. i v
r o
l p r
rbh
- >v b
ao p
a Nr L
p e
- yWhi 3"
t r
t a
a o
)
x 3d tw y
H
~ '
'dwW
(
n f
i g
oa o
. 'f.
Pf A
t s
y ' _m3 f
7
/
r 1
e 2
N ee m
4
,h s
y vf a
if g
t u i
ab n
l l
n
} 3ym2,D f
~
el O
re
(
n c
n e
ta t
nn r
y n
i ec O
t g
5ff ii e i /
ct e
rrv i
il h
uoi C
f u I
tt t p
f m xci e
i es 5
K Mt n A
%6 a
ee
/
01 r
rDS N
o 3
t S
a c
t y
e Y
c l
p E
a ey s
V O
e tl R
R ap a
U mp m
iu m
S
)
xs a
(
o g
L i
rr A
Pf A
pe f
C f
7
/
pw o
4 1
e 2
N ao I
p n
G l
o O
l p i
L am t
O a
a n
- 3. ID ee i
5 pr m.
A np yp a
5.2003042 6647 R
E g
c A
)
t xs b
(
en 1.......
L k
R V
py o
B B
3 I
l di A O e
3p nt F
e p
aa T
s d u r
N ns et a
gn O
s r
ae I
0 c
l de rc T
9
,t e
ww on A
e r
e n
t o T
t S
el n
op sc n
ri,
a d
N e
r ool h
ne e
E hM m o t sa c
ig M
e f
o i wa t a U
r sur a
t r
R u
y it e t
nl e
sy
)
s t
it e
io t
wt s
T a i esa t
hv i
oi e
S e
v v
m t
p l t l
N M
i i n d
h m
l n p
I t
ti g n
eg a
ae m
f c
a n
a pi c
d a
o a
tfi yh hi s
i od a
t e
c 95474245 e
l l
l h
- g l
ic m
02334555 p
l axi p
nn b
hi a
1 11 1l11 l r
o uiu l
i a wd r
db sseuuu T
s qmb a
2d t
i g CSCCCEEE u
a rl O
n
,l p
oc 5
i sc mt u
0 r n an 4
s oi c
l o ty t
ui y
nt c,
C md l
un e)
F ea e
oi t B
/
r t
G ca hC m A
a a4 l &
t (M
Ce m
c I
t0
@m r csus 1
au r
dr I
Nt i
e4 ee n
a x
t 2 z
nl o
m, eir i
r al 8c b% mne M c
p 54 7
52 e
ti e
,0 e
m9t 0 iaf sl p
00320501 t
- d A0t mf 2
ia h
l o C5 r c
r u 3
c (a
t aM M7O tt aB
/ e oang 6
r o n
t u
l 8
at g
doye 3
w
/
o dect n t a i y
t en i n a
f r r
l t,
r ki era ot a
i r eo nmt t
u o sc u pG n
r o vt i r t
e n
o ut ic l en i op i i
P c%t e tiht d
r s t
e0 at s e m gl e
t s
r ht1l e i d i u t
sa e
o ge ed p
i d
,r yy2. h M c
e m u e
ar l
o H
c I
tl 0cx n
Pa a
r n eta il e2 n
g v
r apr eut nN vant9 o s o
r o
l i re i cor Oc.c
'c o a
mm f
l t
rdt b nOi*
ti c
eec aa c3 i rd eee e
c 680 e
bl e t mGix sieet cr t s i 40555001 t
nat r r &f3 npsvr aa ri p
56699l 1 1 e
aee beGf en aiOf Os y
nonrbug3 D
Csd GE ea SDBF r s y T I
CZZt RAA i
t ee Gl v t '2
) ) ) )
1
& a D
)
I_
abcdinK ma n
e 7
8
( ( ( (
Fi t
(
1 TABLE 5.4 LISTING OF SURVEY AREAS / UNITS AND ANALYSIS FREQUENCY
(
SZ S[
AFFECTED/
(f CUkVEY AREA / UNIT UNAFFECTED MEDIA C
1.0 EUT Floor 1.1 Concrete surface A
concrete with limited wk pockets of sediment aq 2.0 South Exterior Wall of'Rx Bldg 2.1 Above grade line*
U
-painted concrete 2.2 Below grade line*
A
-bare concrete 3.0 East Exterior Wall of Rx Bldg 3.1 East wall U
-painted concrete 3.2 West wall U
-painted concrete 0.0 Roof of Rx Bldg 4.1 Gravel and sediment U
concrete covered with I
4.2 Membrane cover U
rubber membrane and 4.3 Old roof surface U
gravel / flashing 5.0 South Exterior Wall of Pump Room
~5.1 Above HUT roof elevation A
-tar coated concrete e.2 Below HUT roof elevation A
-bare concrete 6.0 Bedrock Wall of EUT Excavation granite bedrock with w$
6.1 HUT floor to top of HUT roof A
random pockets of a$
elevation soil / sediment 59 j7.0 Exterior Roof of Pump Room A
bare concrete (e)
Qualitative gamma spectroscopy analysis of sample from locations where 90 results will be used to determine if elevated survey results are due to 2 N/A Not appropriate measurement / analysis for media.
- Will not be performed until HUT excavation has been filled in, access to these)
~
. (%/.
JJATAB54.94D
--~~= Page 1 s
ANALYSIS /OUANTITY - FREQUENCY n.
QUALITATIVE L/
DIRECT
/y DIRECT y SCAN GAMMA BMENT and a
/y REMOVABLE W/ SAMPLE AT EXPOSURE SPECIAL 2PgCT (Thne and a p/y ELEVATED RATE AT ANALYSIS /
Dr )
Integrated)
(Scanned) and a LOCATIONS 1 METER SAMPLES go 1/1m2 2
100%
N/A N/A 1/4m 30 Alable concrete floor cores
/A 30 0
N/A N/A 30 N/A
/A 1/1m2 100%
N/A N/A 1/4m2 N/A
/A 30 0
N/A N/A 30 N/A
/A 30 0
N/A N/A 30 N/A 30 gravel 0
N/A N/A 30 roof
'/A 30 membrane O
N/A N/A N/A drains and
'/A 30 membrane O
N/A N/A N/A vents tar at 2
r/A 1/1m 100%
N/A N/A 1/4m2 locations f/A 1/1m2 100%
N/A N/A 1/4m2 of highest direct result ge 2
lilable 1/1m 100%
N/A N/A 1/4m2 rock /
)
debris (a) 2
//A 1/1m 100%
N/A N/A 1/4m2 N/A Ess p/y-a and/or gamma exposure rate results exceed release criteria, sample igh background (Th-232 and D deposits) or contamination.
areas is currently unsafe.
APERTURE CARD Al80 Available on Aperture Card 6
6 L________________-__-_-_-_________---_
)
m TABLE 5.4 LISTIN3 OF SURVEY AREAS / UNITS AND ANALYSIS FREQUENCY AFFECTED/
SURVEY AREA / UNIT UNAFFECTED MEDIA 8.0 Grounds (South of Rx Bldg, within i
1 RCA) i 8.1 spoils pile A
soil d
8.2 Driveway (asphalt)
A asphalt black top 9
8.3 HUT excavation bedrock (above HUT U
granite bedrock with j
roof elevation) random pockets of sediment ik 8.4 Spill way behind Rx Bldg U
bare concrete 4
8.5 All other areas except above and U
soil / rock g.
storage tank and pipe excavation l'
'8.6 East of driveway U
soil
-s 9.0 Outside of RCA Deleted
- ]
Q 10.0 Storage Tank Excavation sj 10.1 Foundation A
bare concrete j!
10.2 Tank excavation and pipe excavation A
soil V
l-jf 11.0 Cooling Tower / Shack 11.1 U
concrete, wood, metal, plastic, sediment i
=
12.0 5K Tank Excavation 12.1 Excavation A
soil 12.2 Tank foundation A
concrete (a)
Qualitative gamma spectroscopy analysis of sample from locations where g results will be used to determine if elevated survey results are due to N/A Not appropriate measurement / analysis for media.
Will not be performed until HUT excavation has been filled in, access tC i
('v JJATAB54.94D Page 2
AN_ALY_ SIS /_QIJ_ ANTI _TY - F_ REQ _ ENCY U
QUALITATIVE BIL/
DIRECT p/y DIRECT y SCAN GAMMA EDIMENT and a
/y REMOVABLE W/ SAMPLE AT EXPOSURE SPECIAL yCPgCT (Time and a p/y ELEVATED RATE AT ANALYSIS /
3 cr )
Integrated)
(Scanned) and a LOCATIONS 1 METER SAMPLES 2
20 N/A N/A N/A 100%
5/100m N/A 2
2 W/A 1/1m 100%
N/A N/A 1/4m N/A koro 30 0
N/A 10%
30 rock /
Wailable debris (a) 30 0
N/A 10%
30 N/A 30 N/A N/A 10%
30 N/A 30 N/A N/A N/A 10%
30 N/A 2
2 N/A 1/Im 100%
N/A N/A 1/4m N/A 2
2 5/100m..
N/A N/A N/A 100%
5/100m 3f3 shoro 30 0
N/A 100%
N/A
- wood, Svailable pipes /
33 0
- slates, where available 5/1m2 N/A N/A N/A 100%
5/100m2 N/A N/A 5/100m2 100%
N/A N/A 1/4m2 N/A
=
css p/y-cr and/or gamma exposure rate results exceed release criteria, sample igh background (Th-232 and D deposits) or contamination.
these areas is currently unsafe.
APERTURE CARD Aleo Ayajgagig og ADerture Card
/
0 O
?
/
h [/
L__
TABLE 5.4 LISTING OF CURVEY AREAS / UNITS AND ANALYSIS FREQUENCY AFFECTED/
SURVEY AREA / UNIT UNAFFECTED MEDIA
{
8.0 Grounds (South of Rx Bldg, within RCA) 8.1 Spoils pile A
soil 8.2 Driveway (asphalt)
A asphalt black top 8.3 HUT excavation bedrock (above HUT U
granite bedrock with roof elevation) random pockets of sediment 8.4 Spill way behind Rx Bldg U
bare concrete 8.5 All other areas except above and U
soil / rock storage tank and pipe excavation 8.6 East of driveway U
soil 9.0 Outside of RCA Deleted 10.0 Storage Tank Excavation 10.1 Foundation A
bare concrete 10.2 Tank excavation and pipe excavation A
soil 11.0 Cooling Tower / Shack 11.1 U
concrete, wood, metal, plastic, sediment
\\..
12.0 5K Tank Excavation 12.1 Excavation A
soil 12.2 Tank foundation A
concrete (a)
Qualitative gamma spectroscopy analysis of sample from locations where q results will be used to determine if elevated survey results are due to E/A Not appropriate measurement / analysis for media.
Will not be performed until HUT excavation has been filled in, access tc
(
)
v l
JJATAB54.94D l
Page 2
ANALYSIS /_ QUANTITY _ F_REQUENCY QUALITATIVE DOIL/
DIRECT
/y DIRECT y SCAN GAMMA DEDIMENT and a
/y REMOVABLE W/ SAMPLE AT EXPOSURE SPECIAL
]ySPgCT (Tbne and a p/y ELEVATED RATE AT ANALYSIS /
& Br )
Integrated)
(Scanned) and a LOCATIONS 1 METER SAMPLES 2
5/10pm N/A 30 N/A N/A N/A 100%
2 N/A 1/Im 100%
N/A N/A 1/4m N/A 5horo 30 0
N/A 10%
30 rock /
available debris (a) 30 0
N/A 10%
30 N/A 30 N/A N/A 10%
30 N/A 30 N/A N/A N/A 10%
30 N/A N/A 1/1m 100%
N/A N/A 1/4m2 N/A 2
2 2
5/100m.
N/A N/A N/A 100%
5/100m N/A 5horo 30 0
N/A 100%
N/A
- wood, 9vailable pipes /
230
- slates, where available 5/1m2 N/A N/A N/A 100%
5/100m2 N/A N/A 5/100m2 100%
N/A N/A 1/4m2 N/A ess p/y_a and/or gamma exposure rate results exceed release criteria, sample ligh background (Th-232 and D deposits) or contamination.
these areas is currently unsafe.
CARD Alto Avanecie on ADerture enrri
/
0 O
l
/
(6
_... _ _ _ _. _ _ _ _. _ _