ML20081A715
| ML20081A715 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 01/27/1995 |
| From: | Vitkus T OAK RIDGE ASSOCIATED UNIVERSITIES |
| To: | NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| Shared Package | |
| ML20081A714 | List: |
| References | |
| CON-FIN-A-9076 NUDOCS 9503150258 | |
| Download: ML20081A715 (92) | |
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CONFIRMATORY SURVEY OF THE REACTOR BUILDING AND PHASE 4 SYSTEMS l
SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK Prepared by T. J. Vitkus l
Environmental Survey and Site Assessment Program Energy / Environment Systems Division Oak Ridge Institute for Science and Education Oak Ridge, Tennesse5 37831-0117 4
Prepared for the U.S. Nuclear Regulatory Commission Headquarters Office Sponsored by the Division of Waste Management f DRAFT REPORT, REVISION 1 l
JANUAR's 1995 l
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, 9503150258 950130 l PDR ADDCK 05000322 W PDR
- This report is based on work performed under an Interagency Agreement (NRC Fin. No.
A-9076) between the U.S. Nuclear Regulatory Commission and the U.S. Department of Energy.
Oak Ridge Institute for Science and Education performs complementary work under contract number DE-AC05-760R00033 with the U.S. Department of Energy.
Shortham Reactor Buildirig and Phase 4 - January 27.1995 h:\essap\ report \shoreham\brookhaven.002 h'
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r ACKNOWLEDGEMENTS-The author would like to acknowledge the significant contributions of the following staff.
members:
FIELD STAFF T. L. Bright-T. D. Herrera A. L. Mashburn _
E. . H. Montalvo J. L. Payne LABORATORY STAFF -
R. ~ D. Condra J. S. Cox
' M. J. Laudeman s
CLERICAL STAFF D. A. Cox R. D. Ellis K. E. Waters ILLUSTRATOR T. D. Herrera l
$horttam Reactor Buildmg and Phase 4. January 27.1995 h:\casap\reporu\thoreham\brookhaven.002
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I-m TABLE OF CONTENTS PAGE ,
List of Figures ' . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv -
4 .
List of Tables . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . vii .
' Abbreviations and Acronyms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. viii e ' Introduction and Site History . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 Site Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 : -
' Obj ecti ves . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 Document Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 Procedures ................................................4 Findings'and Results ..........................................8'-
Comparison of Results with Guidelines . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 References ............................................... 69-
. Appendices:
Appendix A: Major Instrumentation Appendix B: Survey and Analytical Procedures Appendix C: Regulatory Guide 1.86 Termination of Operating Licenses for Nuclear Reactors Shoreham Reactor Building and Phase 4. January 27.1995 iii h:\essap\ reports \shoreham\brookhaven.002
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LIST OF FIGURES 'I PAGE 7
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FIGURE 1: Location of the Shoreham Nuclear Power Station . . . . . . . . ... . . . . . . 18,,
' FIGURE 2: - Plot Plan of the Sh'oreham Nuclear Power Station '. . . . . . . . . . . c . . . , 19.
I FIGURE 3: Reactor Building Floor Plans of 8'-0" and 40'-0" Elevations-Structural Units Surveyed ~ . . . . . . .-... . . . . . . . . . : . . . . . . . . . . . .- 20f FIGURE 4: Reactor Builuing Floor Plans of 63'-0" and 78'-0" Elevations-Structural Units Surveyed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 l FIGURE 5: Reactor Building Floor Plans of 112'-0" and 126'-0" Elevationh-Structural Units Surveyed . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 :
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FIGURE 6: Reactor Building Floor Plans of 150'-9" and 175'-9" Elevations-Structural Units Surveyed . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 '
FIGURE 7: . Reactor Building, Primary Containment 63' NW G/A (PC004)-
Measurement and Sampling Locations '. . . . . . . . . . . . , . _ . _
24 FIGURE 8: Reactor Building, Primary Containment Sub-Pile Room. (PC005)-
. Measurement and Sampling Imcations . . . . . . . . . . . . . . . . . . . . . . 25 FIGURE 9: Reactor Building, Primary Containment 78' NE G/A (PC007)-
Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . . 26 -
FIGURE 10: Reactor Building, Primary Containment 78' SE G/A (PC008)-
Measurement and Sampling Imcations . . . . . . .. . . . . .. .. . . . . . . . . . 27 -
FIGURE 11: Reactor Building, Primary Containment 103' SE G/A (PC012)-
Measurement and Sampling Locations . . . . . . . . . . .- . . . . . . . . . . 28 ;
FIGURE 12: Reactor Building, 8' SE G/A (RB004)-Measurement and -
Sampling locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ' 29 j
1 FIGURE 13: Reactor Building, 8' SW G/A (RB006)-Measurement and j Sampling Locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30' FIGURE 14: Reactor Building, 40' SW G/A (RB014)-Measurement and Sampling Ixcations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 i FIGURE 15: Reactor Building. 78' SE G/A (RB037)-Measurement and _ l Sampling locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 i Shoreham Reactor Bulkling and Phase 4. January 27.1995 iV h:\cssap\ reports \shoreham\brookhaven.002 f
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f, ' LIST OF FIGURES (Continued)
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- FIGURE 16: Reactor Building, West Accumulator Aisle (RB038)-Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33 !
' FIGURE 17: Reactor Building,' Fuel Pool Pipe Tunnel (RB05,7)-Measurement .
and Sampling Locations .............................. 34'
- FIGURE 18: . Reactor Building, RWCU Regen /Non-Regen HTX's Room (RB061)-
Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . . ' 35 q FIGURE 19: Reactor Building, Spent Fuel Storage Pool, Floor and North-and West Walls (RB068)-Measurement and Sampling locations . . . . . . 36 FIGURE 20: Reactor Building, Spent Fuel Storage Pool,' South and East Walls (RB068)-Measurement and Sampling Locations ............ 37.
- FIGURE 21: Reactor Building, Reactor Cavity Floor (RB071 x01)-Measurement and Sampling Locations ..............................38 FIGURE 22: Reactor Building, Reactor Cavity Walls (RB071 kOI),-Measurement .
and Sampling Locations'. .............................39 FIGURE 23: Reactor Building, Dryer / Separator Storage Pool (RB072 x01),-
Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . 40 FIGURE 24: P.cactor Bu'ilding,175' South G/A (RB103)-Measurement and Sampling locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 41 FIGURE 25: Reactor Building,175' NW G/A (RB106)-Measurement and Sampling locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 42 FIGURE 26: Reactor Building, Polar Crane (RB109)-Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 43
- FIGURE 27: Reactor Building, Polar Crane Rail Area (RB109)-Measurement and Sampling Locations ............................44 FIGURE 28: Radwaste Building,52' SE G/A Shield Blocks (RWO73x02 and x03)-
Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . 45.
FIGURE 29: Radwaste Building, 52' SE G/A Shield Blocks (RWO73 x02 and x03)-
Measurement and Sampling IAcations . . . . . . . . . . . . . . . . . . . . . . 46-Shorehara Reactor Buildmg and Phase 4. Jamsary 27,1995 V h:\cssap\ reports \shoreham\broothaven.002
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N .T.i M _ , . LIST OF MGURES (Continued) i PAGE FIGURE 30:~ Radwaste Building,52' SE G/A Shield Blocks (RWO73x02 and. x03) -
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' Measurement Locations . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . 47 FIGURE 31: Reactor Building, Reactor Assembly lower Bowl (SU001)-
Measurement and Sampling Locations' . . . . . . . . . . . . . . . . . . . . . 48 -
. FIGURE 32: Reactor Building, Nuclear Boiler Main Steam Relief Valve (SU002)-
Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . .. . .. . . . 49 -
FIGURE 33: . Reactor Building, CRD Hydraulic Control Components (SU004)-
Measurement and Sampling Locations . . . . . ._ . . . . . . . . . . . . . . . . . 50 l FIGURE 34: Reactor Building, Porous Concrete' Sump Components (SU014)-
Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . 51
. FIGURE 35: Outdoor Building Storm Drain Manways and Trenches Components . _
- (SUO23)-Measurement and Sampling Locations . . . . . . . . . . . . . . . .- 52 :
FIGURE 36: - Reactor Building, Reactor Primary Containment Components (SUO58)-
Measurement and Sampling locations . . . . . . . . . . . . . . . . . . . . . . - 53
' FIGURE 37: - Reactor Building, Reactor Primary Containment Components (SUO58)-
Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . . -54 FIGURE 38: Reactor Building, Ventilation Components (SUO60)-Measurement and Sampling Locations ..............................55 FIGURE 39: Reactor Building, Ventilation Components (SUO61)-Measurement and Sampling Locations ..............................56 shoreham Reactor Building and Phase 4. January 27.1995 yi h:Wsap\ reports \shoreharn\brookhaven 002 -
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LIST OF TABLES -l PAGE R
TABLE 1: Reactor Building Confirmatory Survey Units . . . . . . . . . . . . . . . . . . 58 l i
TABLE 2: Summary of Surface Activity Levels . . . . . . . . . . . . . . . . .. . . . . . . . ,
-: TABLE 3: - Exposure Rates . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 62 TABLE 4: Confirmatory Radiological Status Summary-Structures . . . . . . . . . . . 63
- TABLE 5: Confirmatory Radiological Status Summary-Systems . . . . . . . . . . . . . 67 3
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cABBREVIATIONS AND ACR'ONYMS e l
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MME >
JAmerican Society of Mechanical Engineers J ' 2
.cm . . square centimeter .
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Co-60. . cobalt-60
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cpm . counts per minute ; 4 j
' DOE Department of Energy .
H dpm/100 cm 2J disintegrations per minute per 100_ square centimeters! ]
EML Environmental Measurements Laboratory - -l EPA- ' Environmental Protection Agency [
.ESSAP Environmental Survey and Site Assessment Program - ' ':
- Fe-55 L iron-55 ft2 - square feet . j H-3 tritium ;
-has hectare j GAG' gross activity guideline GM' Geiger-Mueller ' ;
km kilometer- i I, . critical level LILCO Long Island Lighting Company-LIPA Long Island Power Authority <
m - '
meter m2 . square meter- .}'
MDA minimum detectable activity .
Nal sodium iodide !
- Ni-63 nickel-63
NIST National Institute of Standards and_ Technology _ ;
NRC- Nuclear Regulatory Commission
- ORISE Oak Ridge Institute for Science and Education PC# primary containment structural survey unit designation j
pCi/l picoeuries per liter l pCi/g picocuries_ per gram - -l QA Quality Assurance >
QC Quality Control RB# Reactor Building structural survey unit designation RW# Radwaste Building structural survey unit designation '!
SNPS Shoreham Nuclear Power Station :
SU# system survey unit designation j UCL upper confidence level j pR/h microroentgens per hour ZnS zinc sulfide i i
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Shoreham Reactor Building and Phase 4 January 27,1995 viii h:\essap\ reports \shoreham\brookhaven.002 ,
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CONFIRMATORY SURVEY OF THE REACTOR BUILDING AND PHASE 4 SYSTEMS '
SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK INTRODUCTION AND SITE HISTORY
' The Long Island Lighting Company (LILCO) constructed a boiling water reactor known as the Shoreham Nuclear Power Station (SNPS). The plant was designed to provide a gross electrical output of 849 Megawatts and achieved initial criticality in February 1985. The U.S. Nuclear Regulatory Conunission (NRC) License No. NPF-82 (NRC Docket File No. 50-322) issued for the facility allowed. reactor operations at power levels not to exceed 5% of full power. IAw power testing in accordance with the license then commenced in July,1985 and continued intermittently until January,1989, at which time power generating operations were terminated.
The total reactor operating history was equivalent to 2.03 effective full power days of fuel exposure. The irradiated fuel)which was a standard low enrichment (2 to 3% uranium-235)..
uranium fuel, was removed from the reactor vessel and placed into the spent fuel pool in August 1989. Subsequently, the fuel was shipped off site during 1993 and 1994.
Various reactor components, piping systems, and other equipment became radiologically contaminated as a result of reactor operation. The primary contaminants identified during characterization studies were iron-55 (Fe-55) and cobalt-60 (Co-60). Smaller quantities of nickel-63 (Ni-63), tritium (H-3), carbon-14, nickel-59, manganese-54, zinc-65, and europium-152 were also identified (LILCO 1990).
The Long Island Power Authority (LIPA) was established to decommission the facility and release the site for unrestricted use. LIPA developed a decommissioning plan, approved for implementation by the NRC in June 1992, which included decontamination or removal of contaminated portions of the reactor and other plant systems and equipment. A major consideration of the decommissioning plan was to maintain the integrity, when possible, of plant structures and systems.
Shoreham Reactor Building and Phase 4. January 27.1995 h$cssap\ reports \shorcham\brookhaven.002
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The decommissioning and termination surveys were conducted in four phases. Phase 1 included j the termination survey of the internal components of the main turbine, the Turbine Building and associated systems, site grounds, and exterior site structures. Phases 2 and 3, included the Reactor Building Suppression Pool, Phase 2 systems and the Radwaste Building. Phase 4 addressed the remaining portions of the Reactor Building and associated systems.
The NRC Headquarters' Division of Waste Management requested that the Environmental Survey and Site Assessment Program (ESSAP) of the Oak Ridge Institute for Science and Education (ORISE) conduct confirmatory radiological surveys of the SNPS decommissioning project. Since February 1993, ESSAP has completed the confirmatory survey of the turbine internal components, the Turbine Building, site grounds, exterior site structures, Radwaste Building, Suppression Pool, and Phase 2 systems. The results of these surveys are the subject of separate final, or draft reports (Vitkus 1993a; Vitkus 1994a). The final confirmatory survey for Phase 4, the Reactor Building, has now been completed and is the subject of this report.
SITE DESCRIPTION The SNPS is located in the Town of Brookhaven, New York on the north shore of Long Island, ,
l approximately 80 km (50 miles) east of La Guardia Airport and the confluence of the East River and Long Island " rund (Figure 1). The SNPS is located on a 32.4 hectare (80 acre) portion of a larger 202 hectare (ha) parcel ofland owned by the LILCO. The site is bounded on the north by 1.ong Island Sound, on the east by the Wading River Marshland, on the west by other LILCO property, and on the south by Route 25A. A cyclone fence encloses the 8 ha site secured area.
Within this boundary are the buildings and grounds classified as the Restricted Area, also known ,
as the power block, where radiological controls were necessary (Figure 2). Each of the buildings addressed during the confirmatory process are located here and are shown on Figure 2 as the Turbine Building, the Radwaste Building, and the Reactor Building. Construction of the Reactor Building 's predominately structural steel and concrete. Total floor space of the building is l 2 2 approximately 7,800 m (84,000 ft ) divided among 8 levels at elevations 8'-0",40'-0", 63'-0",
78'-0",112'-0",128'-9",150'-0", and 175'-0". The Reactor Building housed the nuclear steam ,
supply system which included the Reactor Pressure Vessel and its associated auxiliary and safety 6
shoreham Reactor Bunding and Phase 4 - January 27,1995 2 h Aessap\repons\shoreham\brookhaven.002
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- systems. . Major stmetural components included the primary containment system, spent fuel .l storage' pool,. dryer / separator pool, polar crane and building sumps. LThe major auniliary and safety systems included the reactor core isolation cooling, high pressure coolant injection, core j spray, stand by liquid control, reactor. water cleanup, fuel pool cooling and clean'up, and j L primary containment atmospheric control systems.
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Termination . surveys have been performed . in' acc6rdance Ewith.' Draft NUREG/CR-5849 '
(Berger 1992). LIPA classified plant systems and building surfaces into two categories that are j based on the potential ~ for residual contamination. The two area categories, ' referred to;as_
j q
affected 'or unaffected, are defined as follows: "affected areas are those areas of the SNPS that f
-l are potentially contaminated or have known contamination, or a system which circulated, stored -l or processed radioactive materials such that they could become contaminated, or experience, [.,
neutron activation, or where records indicated spills or other occurrences may have resulted in ;
contamination; unaffected areas are those portions of the SNPS that are'not expected to contain
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residual radioactivity" (UPA 1994a). ' Area classification was determined by radiological use s
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history, environmental monitoring activities, and the results of the previous characterization 1 i
survey. Affected and unaffected areas are further subdivided into survey units. Survey units _ j are categorized as structures (floors,- walls, ceilings, and exterior surfaces of piping ~ and j equipment), plant systems (equipment and piping internals), and outdoor areas (grounds and j building exteriors). In addition, affected survey units also had sub-classifications as suspect or j; non-suspect, and may also have been classified as alpha affected if involved with fuel handling j or storage. Phase 4 of the decommissioning addressed a total of 117 survey units, of which 100 l 1!
were structures, and 17 were systems.
OBJECTIVES i i
a The objectives of the confirmatory survey' were to provide independent document reviews and - _
radiological data for use by the N'RC in evaluating the adequacy and accuracy of the licensee's - f procedures and termination survey results. l f
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DOCUMENT R' EVIEW' '
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. ESSAP reviewed LIPA's release records for those survey units selected for confirmatory survey
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(LIPA' 1994b). ' Documents were reviewed for adequacy, accuracy, completeness, and i
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' consistency. Data were reviewed for appropriateness of calculations and interpretations relative - 'g to the guidelines. l i
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PROCEDURES .
A survey team from ESSAP visited the SNPS during the period October 31 through November . l 4,1994 and performed independent visual inspections, measurements, and sampling of survey _ j units associated with the Reactor Building. Table 1 lists,'and Figures 3 through 6 show, the' ..
t structural survey units selected for confirmatory surveys. Survey unit designators were alpha- -
numeric with the first characters designating the type of unit, stmeture (building or area specific,' F RB = Reactor Building, PC = Primary Containment, RW = Radwaste Building [for this report, U e
refers specifically to the Bioshield wall blocks that are presently stored in the Radwaste j i
- Building]), or system (designated as SU), followed by a three-digit numeric reference. Subunits ' .
were given an additional two-digit designation preceded by X. Eleven survey units were selected randomly and twenty were selected by the NRC site representative. Field survey activities were l conducted in accordance with a site-specific survey plan submitted to and approved by the NRC- l as well as applicable sections of the ESSAP Survey Procedures and Quality Assurance Manuals f (Vitkus 1994b). The following procedures apply to survey units selected for independent )
confirmatory surveys. Additional information regarding selection of confirmatory survey units and the implementation of this plan may be found in the general site confirmatory survey plan -
(Vitkus 1993b). ,
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SURVEY PROCEDURES Reference System LIPA established the grid system that ESSAP used for referencing measurement and sampling locations. . The grid size or reference interval established by LIPA~ for a given survey unit was dependent upon the classification of the survey unit (affected vs. unaffected) and surface (floor, . j lower wall, upper wall, ceiling, or equipment). Typically, the established grid consisted of
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1 m X 1 m grid blocks on the floors and lower walls (up to 2 m). ESSAP referenced ;
measurement and sampling locations on ungridded surfaces (upper walls and equipment) to the [
floor or lower wall grid. Systems were referenced by drawings or by prominent components.
Surface Scans Surface scans for alpha, beta, and gamma activity were performed over 100% of floor and lower wall surfaces and up to 50% of equipment surfaces within each structural survey unit selected for confirmation. Additional scans were performed over portions of upper wall, ceiling, and i system surfaces, as well a3 :ocations such as drains, where material may have settled or accumulated. Accessible portions of each confirmatory system survey unit were also scanned.
Locations of elevated direct radiation detected by scans were marked for further investigation.
Scans were performed using gas proportional, Geiger-Mueller (GM), zine sulfide (ZnS), and/or ;
sodium iodide (Nal) detectors coupled to ratemeters or ratemeter-scalers with audible indicators. t Surface Activity Measurements !
For each structural survey unit, ESSAP performed a minimum of 30 direct measurements for ;
total beta surface activity. ESSAP also performed additional direct measuremente at locations of elevated direct radiation detected by surface scans. At measurement locations, where the ,
average NRC smface contamination guideline was exceeded, the size of the contaminated area j and the average activity in the contiguous 1 rrl area was also determined. Measurements were {
performed using gas proportional and/or GM detectors coupled to ratemeter-scalers. A smear Shaham Reactor Buildmg and Phne 4. January 27.199 5 h%upVepomhhmhamermumenRU
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sample for determining removable gross alpha and gross beta activity levels were col'.ected from each direct measurement location. Figures 7 through 29 show measurement and simpling locations.
Exposure Rate Measrrements
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l Exposure rate measurements were performed within each survey unit at each accessible floor direct measurement location, excluding units RB071 x01 and RB109 and system interiors which were inaccessible for this type of measurement (Figures 7 through 19,23 through 25, and 30).
.- All exposure rates were measured at 1 m above surfaces using a pressurized ionization chamber (PIC). Background exposure rates were previously determined during the confinnatory survey of the Turbine Building (Vitkus 1994a).
Systems LIPA provided access points into each system or system component listed in Table 1. Beta and ;
gamma surface scans were performed within the accessible portions of each system or I
component, followed by beta direct measurements and smear samples. The total number of
( direct measurements performed and smears collected was dependent upon component size and accessibility and ranged from 5 to 45 measurements per system. Scans and direct measurements were performed using gas proportional, GM, and/or Na1 detectors coupled to ratemeters or l- ratemeter-scalers. Figures 31 through 39 show measurement and sampling locations.
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l Miscellaneous Material Samoles Per the request of the NRC site representative, ESSAP collected a water sample from the porous concrete sump and a sediment sample from the truck bay trench (Figures 34 and 35). ]
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l H Sample Analysis and Data Interpretation i i
Samples and data were returned to ESSAP's Oak Ridge, Tennessee facility for analysis and l interpretation. Smear samples were analyzed for gross alpha and gross beta activity using a low bsckground proportional counter. Direct-measurement and removable activity results were . j 2
converted to units of disintegrations per minute per 100. square centimeters l(dpm/100 cm). f
? Surface activity levels which exceeded'the backgroimd distribution,' referred to as' the critical ;
' level (IJ, were then corrected for those radionuclides that may not be quantified using field ? ]
I instrumentation,'specifically Fe-55, H-3, and/or to a lesser extent Ni-63. The respective' correction factors used were developed by UPA and were based on the fraction of total activity contributed by each radionuclide in each particular survey area (UPA 1994a). Water and sediment samples were analyzed by solid state gamma spectrometry. Spectra were reviewed for f
Co-60 and any other identifiable photopeaks. Results were reported in units of picocuries per : j liter (pCi/l) for water and picocuries per gram (pCi/g) for the sediment. Exposure rates were !
i reported in units of microroentgens per hour ( R/h). Additional information concerning survey .l and analytical procedures may be found in Appendices A and B. i DATA EVALUATIONS AND COMPARISONS i e
The results of each survey unit sampled were statistically tested. The goal of the test was to l
- determine, with a given confidence level, whether the UPA survey data was not biased low compared.to ESSAP. The null hypothesis was that in a survey unit,' surface activities as calculated by UPA were greater than or equal to those determined by ESSAP, i.e., H.: UPA j 2: ESSAP. This hypothesis was tested at the 95% confidence level (0.05 level of' significance). If the hypothesis was rejected at that confidence level, the alternative hypothesis .!
'I was accepted i.e., Hg UPA < ESSAP. The test statistic, t, was calculated using the- i i
following equation: ;j I
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t- T-K z l
.(ng-1) si + (nz-1) sl f n, + nz ' \
) n, + nz -2 , n,nz ,
'where:
K is the LIPA surface activity mean for a survey unit 4 is the ESSAP surface activity mean for the same survey unit L no is the number of LIPA direct measurement data points ng is the number of ESSAP direct measurement data points I
St, Sg are the standard deviations.
The calculated t was then compared to the critical.value of Student's t-distribution (one-ta' led) for the appropriate degrees of freedom at the 95% confidence level (0.05 level of significance).
If the H : LIPA 2: ESSAP was rejected, then ESSAP evaluated additional options.and alternatives and conferred with.the NRC as to the recommended approach.
I FINDINGS AND RESULTS DOCUMENT REVIEW 1
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ESSAP's review of the termination survey final report and release records for those survey units selected for confirmatory survey indicated that the termination plan had been appropriately !
followed with no significant deviations. Data were appropriately converted, tested, and presented.
l 1 j SURFACE ACTIVITY GUIDELINES 1
l l The surface activity results that follow were compared with the following guidelines: l l
l l
Shoreham Reactor Buikting and Phase 4 - January 27.1973 8 h:\cssap\ reports \shorcham\brookhaven.002 I
L__.__.. . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
.y i
, The beta ' guidelines are: !
Total Activity 5,000 B-p dpm/100 cm2, averaged over 1 m2 15,000 B-1 dpm/100 cm2 ; maximum in a 100 cm2 area ;
F :
L- Removable Activity
~
~
1,000 B-y dpm/100 cm2 ,
n .
I j; The alpha guidelines are:
Total Activity j 2
5,000 a dpm/100 cm 2, averaged over 1 m 15,000 a dpm/100 cm2 , maximum in a 100 cm2 area s ;
. Removable Activity j 1,000 a dpm/100 cm2 In addition, the NRC has approved site-specific surface contamination gaidelines for H-3 and ;
Fe-55, particularly in activated concrete and steel (Pittiglio'1994). These guidelines are:
Total Activity ;
2 2 200,000 dpm/100 cm , averaged over 1 m 600,000 dpm/100 cm2 , maximum in a 100 cm2 area 1 t
Removable Activity ]
2-1,000 dpm/100 cm For those survey units involving mixtures of radionuclides 13PA substituted the gross activity guideline (GAG) value, which is based on the " sum of fractions mie" found in Appendix A of l
-l Shmbam Reactor Building and Phme 4 - January 27, im 9 h w apwep-wh-h nw m o m2
+ .. .. ..
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a :3 e"G w
7 Draft NUREG/CR-5849, rather than the above generici NRC guidelines (LIPA 1994a; Berger. .
i1992). These confirmatory survey units, together with the GAG 'values are listed below-n e,. 1 N #
. SU001. Reactor Assembiv Lower Bowl and RB072x01. Drver/Senarator Pooi 11,100 dpm/100 cm2, averaged over.1 m2 33,000 dpm/100 cm2 , maximumi RWO73 x02. Bioshield Concrete 95,900 dpm/100 cm2, averaged over 1 m2 l'
287,700 dpm/100 cm2, maximum :
. . RWO73 x03. Bioshield Steel 76,900 dpm/100 cm2 , averaged over 1 m2 230,700 dpm/100 cm2 , maximum s
RB068. Spent Fuel Pool 9,400 dpm/100 cm2, averaged over 1 m2 28,300 dpm/100 cm2, maximum .
-SURFACE SCANS Surface scans identified a number of locations with above background residual beta surface' activity in three survey units, RB068 (Spent Fuel Storage Pool), RB072x01 (Dryer / Separator Pool), and SUO01 (Reactor lower Bowl). Additional investigation of each location determined that two of the locations, each measuring less than 100 cm2 in' total area, were either at the-average GAG or exceeded the maximum GAG values. Both of these wem located in servey unit .
RB068. The surface activity levels were 9,400 dpm/100 cm2 (average GAG is 9,400-2 2 dpm/100 cm2 ) and 32,000 dpm/100 cm (maximum GAG is 28,300 dpm/100 cm). - LIPA personnel remediated each location and ESSAP then performed post-remedial action surface scans and direct measurements. These surface scans and measurements indicated that areas were adequately remediated. The surface activity levels at the other locations were less than the -
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(, i fapplicable guidelines but'LIPA elected to remediate these areas as well. There were no other :
+
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areas of elevated direct gamma or beta activity detected in the remaining confirmatory survey- I
- units.
]
SURFACE ACTIVITY LEVELS I Li j
The data reported below and in Table 2 is the difference between the gross field sample counts -
and area background. The difference was then corrected for detector efficiency and geometry, sample count time, and contributions from Fe-55, H-3, and Ni-63 as appropriate.(whe'n'the net J f count rate exceeded the background distribution [I,]).' Actual values are reported, including f i
negative surface activity levels, which occurred when the field count rate was less than the background. Of 7% measurements,180 exceeded the Idnd 104 exceeded'the instrumentation minimum detectable activity level.) .
The results of total and removable surface activity levels for each confirmatory survey unit are 1 summarized in Table 2. Total' beta activity levels after additional remediation for the stmetural survey units, excluding RWO73X02 and -X03 which are discussed below, ranged from -760 to .
5,600 dpm/100 cm .2 The highest final direct measurement was located in the Spent Fuel Storage Pool (RB068) which had an average GAG of 9,400 dpm/100 cm 2
. Prior to the additional j remediation, residual activity levels ranged up to 8,600 dpm/100 cm2 (average GAG of 11,100 2
dpm/100 cm2 ) and 32,000 dpm/100 cm2 (maximum GAG of 28,300 cpm /100 cm ) in survey ; j units RB072X01 and RB068, respectively. Removable activity levels ranged from -1 to.8 j 2
dpm/100 cm2 for gross alpha and from -7 to 23 dpm/100 cm for gross beta. The mean residual l activity in structural survey units, as presented in Table 4, ranged from -340 .to -
2 l
450 dpm/100 cm 2and from -0.7 to 3.5 dpm/100 cm for total and removable gross beta activity,-
2 respectively. The mean activity levels, prior to additional remediation, were 670 dpm/100 cm
]!
2 in RB072X01 and 1,500 dpm/100 cm in RB068. l i
I 2
Total beta activity in surveyed systems ranged from -680 to 2,500 dpm/100 cm . The removable ; j 2
activity levels ranged from -1 to 5 dpm/100 cm 2for gross alpha and from -5 to 35 dpm/100 cm .
)
11 h:kssap\ reports \shoreham\brookhaven.002 Shortham Reeciar BulMing and Phase 4 January 27.1995 -)
i i
_. -,.. - . . . _ ~ _ , , . .. __ _ . , , . . _ , ,
i I
h-2 for gross beta, The mean beta activity levels for syst *ms ranged from -70 to 390 dpm/100 cm l 1
for total activity and from -0.3 to 2.8 dpm/100 cm2 for removable gross beta activity (Table 5).
-l Total beta activity levels for the Bioshield Wall Blocks (RWO73X02 and -X03) ranged from -410 2
to 15,000 dpm/100 cm2. The average GAGS for these survey units were 95,900 dpm/100 cm for RWO"7X02 and 76,900 dpm/100 cm2 for RWO73X03. Removable activity ranged from -1 to 1 dpm/100 cm2 for alpha and from -5 to 17 dpm/100 cm2 for beta. The mean beta activity L levels were 6,400 and 55 dpm/100 cm2 for RWO73X02 and -X03, respectively. The mean gross beta removable activity was 0.4 dpm/100 cm2 for both units.
I EXPOSURE RATES The interior background exposure rates, as determined during the Turbine Building survey, ranged from 4 to 5 rR/h and averaged 5 pR/h at 1 m (Vitkus 1994a). Individual gross exposure rates withm the Reactor Building and adjacent to the Bioshield Blocks ranged from 2 to 6 pR/h at 1 m (Table 3). The net m'ean exposure rates for the stmetural survey units, ranged from
-3 to 0.9 pR/h (Table 4).
RADIONUCLIDE CONCENTRATIONS IN MISCELLANEOUS SAMPLES Cobalt-60 concentrations in the Porous Concrete Sump water sample and the Reactor Building Truck Bay trench residue samples were <4.2 pCi/l and <0.2 pCi/g, respectively. No other significant photopeaks were identified. ,
COMPARISON OF RESULTS WITH GUIDELINES i
The confirmatory survey results were compared with both the data provided by LIPA and the NRC guidelines for release to unrestricted use. The NRC's Regulatory Guide 1.86 provides the guidelines for acceptable surface contamination levels used to determine whether a licensed ;
facility may be released to unrestricted use. These guidelines are summarized in Appendix C. j The applicable guidelines are those for beta-gamma emitters and the alpha contamination l l
3 Shoreham Reactor Buildmg and Phase 4. 'inuary 27.1995 12 b:\essap\repons\shoreham\brookhaven.002 1
1 I
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guidelines are those for uranium and associated decay products? The betalgamma guidelines are: d
' ~
' Total Activity 2
5,000 B-y dpm/100 cm 2, averaged over 1 m 2 ,
15,000 B-y dpm/100 cm2 , maximum in a 100 cm2 area i
+
1
~
Removable Activity 1,000 B-y dpm/100 cm2 .
~
]. I The alpha guidelines are:
j i
Total Activity ;
2 2 5,000 a dpm/100 cm , averaged over 1 m -l 15,000 a dpm/100 cm2 , maximum in a 100 cm2area 7
j
, s .t Removable' Activity a
1,000 a dpm/100 cm2 j In addition, the NRC has approved site-specific surface contamination guidelines for H-3 and j
.I Fe-55, particularly in activated concrete and steel (Pittiglio 1994). lThese guidelines are: .
Total Activity ;
1 200,000 dpm/100 cm2, averaged over 1 m2 ;
i 600,000 dpm/100 cm2 , maximum in a 100 cm2 area {
Removabfe Activity l
2 1 , 1,000 dpm/100 cm ?!
i l
For those survey units involving mixtures of radionuclides LIPA substituted the gross activity .
guideline (GAG) value, which is based on the " sum of fractions rule" found in Appendix A of l Draft NUREG/CR-5849, rather than the above generic NRC guidelines (LIPA 1994a; Berger ]
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4 2, "1992). These confirmatory survey. units,~together with the GAG values that the data were -
1l
( ' compared with and tested against, are listed below:
.p o .l SU001. Reactor Asembly Imwer Bowl and RB072xCi.- Qrver/Senarator Pool 2
11,100 dpm/100 cm2, averaged over 1 m 3 eo L 33,000 dpm/100 cm2 , maximua..
[l!
1
~ RWO73 x02. Bioshield Concrete 2
95,900 dpm/100 cm2 , averaged over 1 m ;
287,700 dpm/100 cm2 , maximum
- RWO73 x03c Bioshield Steel 2
76,900 dpm/100 cm2 , averaged over 1 m .
I 230,700 dpm/100 cm2 , maximum RB068. Synt Fuel Pool 2
9,400 dpm/100 cm2, averaged over 1 m
~ 28,300 dpm/100 cm2 , maximum .
2 Removable activity guidelines for each of the above remained at 1,000 dpm/100 cm ,
The exposure rate guideline currently being used by the NRC is 5 pR/hr above background, measured at 1 m above the surface (U.S. NRC 1984).
As previously discussed, the detection sensitivities of the field instruments are such 'that the residual Fe-55 activity cannot be detected. In addition, UPA determined that residual H-3 and/or Ni-63 activity also must be accounted for within several survey units (UPA 1994a).
Therefore for those surveys that UPA identified as such, total and removable surface activity measurements were corrected for Fe-55, H-3, and/or Ni-63. The mean of the corrected surface activity level for each survey unit was then calculated and the survey unit data was tested at the 95% confidence level (upper confidence level [UCL]), relative to the guidelines, in accordance Shoreham Reactor BuBding and Phase 4 - January 27.1995 14 h:\cssap\reporis\shoreham\brookhaven.002
with Draft NUREG/CR-5849 (Berger 1992); These results are provided in Table 4. The ESSAP and UPA data sets were then compared and evaluated relative' to the established L' conditions.
7
? A comparison of the ESSAP mean surface activity levels to the UPA mean activity levels showed that~ the ESSAP mean was statistically less than or equal to the respective mean
' determined by UPA for 24 of the 30 confirmatory survey units, thereby satisfying the data-l'- evaluation and comparison conditions. However, the ESSAP mean surface activity value was-statistically greater than the UPA mean for survey units PC005, PC007, PC012, RB072X01, R_ RWO73X02 and SUO60; therefore, additional evaluation was necessary.
For the three Primary. Containment survey units (PC005, PC007, and PC012) where the ESSAP '
surface activity mean was higher than the UPA mean, UPA qualified the release record data .
as containing excessive negative measurements. According w UPA, this was due to observed, background levels being lower than the generic site backgrounds that were used for surface activity measurement conversio'ns. As a result, the mean survey' unit total activity levels were biased low. The maximum surface activity level obtained by ESSAP for these survey units was 910 dpm/100 cm2 n PC007, and the maximum survey unit mean and UCL, found in PC005 were 310 dpm/100 cm2 and 350 dpm/100 cm2 , respectively. In' addition, the ESSAP total surface activity mean was less than or only slightly higher than the detection limits of the ~ >
instrumentation, in all three units. Baad on these factors, it is ESSAP's opinion that for the above survey units, the observed difference in means would not affect the conclusions UPA' reached for remaining survey units.
- In survey unit RB072X01, the ESSAP surface activity mean of 450 dpm/100 cni2 exceeds the UPA mean of 140 dpm/100 cm 2, . It should be noted that ESSAP performed 5 of the 30 direct measurements at locations of elevated direct radiation identified by surface scans. The surface - y 2
activity at these. locations ranged from 1,800 to 4,000 dpm/100 cm. Inclusion of these -
f measurements in the determination of the average surface activity level results in a mean'that . y I
1 Shoreham Reactor Building and Phase 4 - Jarmary 27.1995 15 h:\essap\ reports \shoreham\brookhaven.002 l-g rmi ______ _ _ _ _ _ _ _ _ _,.
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. 'is biased high and may not be representative of the average survey unit residual activity. level f which UPA reported. j i
2
- The ESSAP mean for survey unit RWO73X02 was 6,400 dpm/100 cm versus the UPA mean -
- of 2,200 dpm/100 cm2. To the extent practical,' ESSAP selected direct measurement locations ,
o
' on the concrete portion of the shield blocks on the basis of ths maximum direct radiation levels ~
noted.while performing surface scans.' UPA selected measurement locations systematically /
, Therefore, similar to RB072X01, the ESSAP.mean for this survey. unit is high relative to the l LIPA determined average, and the difference in means'in not considered significant when~
2 compared to the applicable guidelines (average GAG of 95,900 dpm/100 cm ) for this survey - l unit. :
t For survey unit SUO60, the ESSAP total surface activity mean was 360 dpm/100 cme with a -
l 2
2 maximum level of 2,500'dpm/100 cm . The corresponding UPA values were 174 dpm/100 cm - j and 3,247 dpm/100 cm2 . UPA had not qualified the data for this unit as containing excessive l negative values. Therefore, thetPA surface activity levels, mean activity level, and UCL were -
evaluated.to determine the potential impact on the UPA reported status of this and the other l 3
i Phase 4 survey units, relative to the guidelines. The difference between the ESSAP and UPA .
means was 213 dpm/100 cm2 . If this difference in activity levels were applied to each Phase
- 4 survey unit, overall surface activity levels would not be significantly altered and the ]
conclusions reached, that the radiological status of this and all other Phase 4 survey unit satisfy
{
the guidelines, would remain valid.
Survey unit RB068, the spent storage pool, was an alpha affected area. Although the total activity levels provided in the Tables of this report indicate beta activity, the instrumentation ]
= ESSAP used in this survey unit detects both alpha and. beta radiation. LESSAP evaluaied the confirmatory data developed for this unit, relative to the potential alpha radiation contribution . j at the location with the highest surface activity level. If all of the net activity'present at this l location was to be attributed to alpha contamination, the maximum alpha surface contamination -
2
- level in the survey unit would be approximately 2,600 dpm/100 cm , which is below the average
-i f
2 4 alpha activity guideline of 5,000 dpm/100 cm .
' Shortham Reactor Building and Nse 4 January 27.1995 16 hh@rgwdhwehmWrmAmenR2 -- l
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. Compdrison of the ESSAP survey data directly with the guidelines showed that overall, total
[- ,
surface activity levels within each survey unit satisfied the guidelines at tie 95 % confidence. level
~ ( ; or UCL). The maximum UCL was 2,900 dpm/100 cm2 n survey unit RWO73X02. This -
comparison also included an evaluation and data testing of the preremediation surface activity
~ levils in survey units RB068 and RB072X01. The evaluation determined that prior to-the '
( 2 e additional remediation the average' GAGS (9,400 dpm/100 cm for RB068 and 11,100 dpm/100 2
L cm for RB072X01) were satisfied at the 95 % confidence level and no additional measurements 2-were required to demonstrate compliance. The preremediation UCLs were 3,100 dpm/100 cm for RBM8 and 1,200 dpm/100 cm 2for RB072x01.' There were no final direct measurements.
that exceeded the applicable NRC generic or site-specific GAG average guideline values. -'All l - removable activity was below the guideline at the 95% confidence level.
Exposure rates were compared with those obtainedb' y LIPA, and tested at the 95% confidence level, relative to the 5 R/h above background guideline currently being used by the NRC. ' The .
Reactor Building and Bioshield Block exposure rates were comparable to background exposure l
rate levels and confirmed the findings presented by LIPA.
1 There was no detectable Co-60 in the sediment sample and therefore the site-specific guideline of 8 pCi/g is satisfied. There is not a site-specific guideline for Co-60 in water for comparison; however, the water sample collected did not contain a detectable level of Co-60.
SUMMARY
l The Environmental Survey and Site Assessment Program penimed-confirmatory survey activities for the Reactor Building at Shoreham Nuclear Power Station in Brookhaven,' New -
. York. Confirmatory activities included document reviews, and during the period of October 31 -
through November 3,1994, ESSAP performed independent surface scans, surface activity 4 measurements, exposure rate measurements, and miscellaneous sampling.
ESSAP's confirmatory survey results support the data presented in the LIPA termination survey report. These results indicate that total and removable surface activity levels and exposure rates -
4 Shorcham Reactor Buil&ng and Phase 4 - January 27.1995 17 hhpWpomWrchmWa*hmn@2 l
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were below the.NRC guidelines for release to unrestricted use. Statistical tests of data se
' '. * .2 furthef support the conclusion that each survey unit' satisfies the guidelines at the 9 y
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FIGURE 2"/: Reactor Building, Polar Crane Rail Area (RB109) - Measurement and Sampling Locations Shoreham Reactor Buildmg and Nse 4 - January 27,1995 45 hnessap\ reports \shortham\brookhaven.002 '
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i h i FIGURE 29: Rodweste Building,. 52' SE G/A Shield Blocks (RWO73X02 and X03) - i Measurement and Sompling Locations ; Shoreham Reactor Buildmg and Phase 4 January 27.1995 47 h:bsap\ reports \shoreham\brookhaven.002 [ t
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$ SINGLE-POINT FIGURE 32: Recctor Biulding, Nuclear Boiler Mcin Stecm Relief Valve (SUC02) -
Measurement and Sampling Locations Shoreham Reactor Buildmg and Phase 4. January 27,1905 50 a:sessapsreportsunorehamshrootma 002
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Measurement and Sampling Locctions Shoreham Reactor BuMing and Phase 4 January 27, tw5 51 h;w ap w po m w meham m m e m o.m2
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M( surement and Sampling Locations Shoreham Reactor Building and 1 use 4 - January 27. IM $2 h:\cssap\ reports \shoreham\brookhaven.002
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- Outdoor Building Storm Drains Manways and Trenches, Components (SUO23) -
Measurement and Sampling Locations Shoreham Reactor BuMing and Phase 4 - January 27.1995 53 ne.sapsrepor=whorehamwrooummm2
273-033 (2) a. J J A _ _ M l' > 9 SUD58-14-1 9 SUO58-15-1 l W V SUO58-11 -1 SUO58-16-1 SUO58-17-1 0 J SUO58-12-1 SUO58-11-1 SUO58-6-1 SUPPRESSION POOL SUPPRESSION POOL SUPPRESSION POOL MEASUREMENT / SAMPLING LOCATIONS NOT TO SCALE g SINGLE-POINT FIGURE 36: Reactor Building, Reactor Primary Containment Components (SUO58) -- Measurenment and Sampling Locations Shoreharn Reactor Building and Phase 4 - January 27,1995 54 h:sessapsreportisihorehamsdroothaven.oo:
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l FIGURE 37: Reactor Building, Reactor Primary Containment Components (SUO58) - l Measurement and Sampling Locations Shoreham Reacwr Building and Phase 4 - January 27.1995 55 a:sessapsregorusinorenomsoroornaven.oo2
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i FIGURE 38: Reactor Building, Ventilation Components (SUO60) - ! Measurement and Sampling Locations , surcham acacior suudme we rnase 4. January :7 995 56 b:VuapVepomuhoreham\bnmbam.002 a
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I FIGURE 39: Reactor Building, Ventilation Components (SUO61) - Measurement and Sampling Locations Shoreham Reactor Buhg and Phase 4 January 27,1995 57 m:sess.pweporusshorehamsbrooth,ve .oo:
TABLE 1 REACTOR BUILDING SURVEY UNITS i SELECTED FOR CONFIRMATORY SURVEY SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK , Survey Mected (A)/ Stmdure/ r Survey Unit Name Unit / Component Unaffected (U) System PC004 Primary Containment - 63' NW A structure , PC005 Sub-Pile Room A structure PC007 Primary Containmem - 78' NE A structure PC008 Primary Containment - 78' SE A structure PC012 Primary Containment - 109' NE A stmeture RB004 Reactor Building - 8' SE A structure RB006 Reactor Building - 8' SW A structure RB314 Reactor Building - 40' SW A_ structure RB037 Reactor Building - 78' SW A stmeture , RB038 West Accumulator Aisle A structure RB057 Fuel Pool Clean-up Pumps Room A structure RB061 - RWCU Regen /Non-Regen HTX's Room A structure , RB068 Spent Fuel Storage Pool A structure RB071x01 Peactor Cavity - 150' A structure RB072x01 Dryer / Separator Storage Pool A structure RB103 Reactor Building - 175' A structure , RB106 Reactor Building - 175' A- structure RB109 Polar Crane A structure i RWO73x02 Bioshield Wall Blocks (Concrete) A structure RWO73 x03 Bioshield Wall Blocks (Steel) A stmeture i SUO01 Reactor Assembly Lower Bowl A system i
- SU002 Nuclear Boiler Main Steam Relief Valve A system !
1B21-RV-095G l
%:thara Reactor Building and Phase 4 - January 27 1m 58 h:\cssap\ reports \shorehannbrookhaven 002 i -_ _ . . _l
TABLE 1 (Continued) REACTOR BUILDING SURVEY UNITS SELECTED FOR CONFIRMATGBY SURVEY SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK Survey Survey Unit Name Anected (A)/ Stmetum/ ;
~
Unit / Component Unaffected (U) System f SU004 CRD Hydraulic Control A system SUO14x05 #953 Misc. Embedded Drain Piping A system 1G11-TK-190 I SU014 x12 RB Porous Concrete Sump #P-224A A system SU023 Misc. Building Storm Drains A system SUO58 Reactor Primary Containment A system j SUO60x09 Fuel Pool Cooling Room A system SUO60x13 RWCU Valve Chambers Room A system r"'%0 x 23 Reactor Building Air from Drywell A system IT46-ADV-039A SUO60x24 Reactor Building Vent Dump A system i i k 1 i Shon:havn Reacsar Buildmg and Phase 4 - January 27, im 59 h:\casap\ reports \shortharn\brookhaven.002 I
TAQLE 2 %
SUMMARY
OF SURFACE ACTIVITY LEVELS REACTOR BUILDING E SIIOREIIAM NUCLEAR POWER STATION I BROOKIIAVEN, NEW YORK I a F Number of Total Activity Range Removable Activity I Location" Measurement (dpm/100 cm2) Range (dpm/100 cm2) Il Locations Betab'* Alphad Beta' i
} Reactor Building
[ PC004 Primary Containment 30 -230 to 410 -1 to 1 -5 to 14 25 PC005 Primary Containment, Sub-Pile Room 30 57 to 580 -1 to 3 -4 to 18 5 PC007 Primary Containment 30 -340 to 910 -1 to 5 -4 to 10 PC008 Primary Containment 30 -420 to 270 -1 to 3 -5 to 6 PC012 Primary Containment 30 -100 to 530 -1 to 6 -4 to 9 g RB004 Reactor Building 30 -760 to -9 -1 to 3 -7 to 3 RB006 Reactor Building 30 -640 to -150 -1 to 3 -4 to 8 RB014 Reactor Building 30 -220 to 370 -1 to 3 -5 to 10 RB037 Reactor Building 30 -470 to -61 -1 to 5 -3 to 12, RB038 West Accumulator Aisle 30 -670 to 580 -1 to 6 -5 to 10 RB057 Fuel Pool Clean-up Pumps 30 -450 to -22 -1 to 8 -5 to 9 e RB061 RWCU IITX's Room 30 -570 to 330 -1 go 1 -5 to 5 5 { RB068 Spent Fuel Storage Pool 34 (-300 to 32,000)'8 -1 to 5 -3 to 21 j -300 to 5600 ( RB071 X01 Reactor Cavity 30 -470 to 680 -1 to 3 -7 to 23 I RB072X01 Dryer / Separator Storage Pool 30 (-390 to 8600)'h -1 to 6 -3 to 21
-390 to 4000 g RB103 Reactor Building 30 -500 to 530 -1 to 3 -5 to 12 5 RB106 Reactor Building 30 -190 to 550 -1 to 3 -4 to 6 RB109 Polar Crane and Crane Rail Area 30 -570 to 770 -1 to 3 -3 to 14
TAQLE 2 (Continued) SUAIA1ARY OF SURFACE ACTIVITY LEVELS REACTOR BUILDING E SIIOREHAAI NUCLEAR POWER STATION [r BROOKHAVEN, NEW YORK I Removable Activity r Number of Total Activity Range I Location" hieasurement Locations (dpm/100 cm2) Range (dpm/100 cm2) E Beta" Alphad Beta' F SU001 Reactor Assembly Lower Bowl 30 -160 to 2200 -1 to 3 -5 to 10
}
E SU002 Nuclear Boiler System 5 -38 to 340 -1 to 1 -3 to 1 l js SU004 Hydraulic Control System 18 -460 to 300 -1 to 5 -5 to 3 3 SU014x12 Sump P-224A 6 -110 to 730 -1 to 1 -1 to 5 SUO14x05 Embedded Drain Piping 10 -380 to 340 -1 to 1 -4 to 5 SUO23 Misc. Building Sto'rm Drains 26 -680 to 2400 -1 to 1 -5 to 13 g SUO58 Primary Containment Valves 17 -190 to 1000 -1 to 5 -4 to 35 SUO60 Reactor Building Ventilation 45 38 to 2500 -1 to 5 0 to 35 SUO61 Stand-By Ventilation 19 -76 to 590 -1 to 5 -3 to 12 Radwaste Building RWO73 x02 Bioshield Wall Blocks (Concrete) 25 -130 to 15,000i -1 to 1 -5 to 17 RWO73 x03 Bioshield Wall Blocks (Steel) 25 -410 to 66001 -1 to 1 -5 to 17 w lii
- Refer to Figures 7 through 39.
b f- h1DAs = 200 to 1200 dpm/100 cm2, 1
- Activity levels corrected for Fe-55, H-3, and Ni-63 as appropriate.
f d MDA = 12 dpm/100 cm2, (5 'MDA = 16 dpm/100 cm2, f Represents surface activity levels prior to additional remediation.- h 8 GAG values are 9,400 dpm/100 cm2 average and 28,300 dpm/100 cm2 maximum. i h GAG values are 11,100 dpm/100 cm2 average and 33,300 dpm/100 cm2 maximum. 8 ' GAG values are 95,900 dpm/100 cm2 average and 287,700 dpm/100 cm2 maximum. IGAG values are 76,900 dpm/100 cm2 average and 230,700 dpm/100 cm2 maximum. __ _ - _ - - __ -______-A
TABLE 3 l
' EXPOSURE RATES REACTOR BUILDING -l SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK ;
Location' um r of easurement Net Exposure Rate hge Locations - at I'm (gR/h? Reactor Building PC004 10 -3 to -2 3 PC005 8 -1 PC007 7 -3 PC008 8 -3 PC012 2 -2 RB004 10 -2 to -1 RB006 4 -2 to -1 RB014 10 0 to 1 RB037 9 -1 RB038 6 -1 ; RB057 ___ 6 -1 to O RB061 10 -1 to 0 _ j RB068 9 -1 to 1 l RB071x01 nab NA RB072x01 9 0 to I j RB103 17 0 to 1 I RB106 9 0 to 1 ) RB109 NA NA Radwaste Building > RWO73x02 5 1 RWO73 x03 5 0
" Refer to Figures 7 through 18 23 24 25 and 30, , , , .
Site background exposure rate is 5 pR/h.
'NA=Not Applicable, measurements not performed.
I i 4 l Shorehara Reactor BuMing and Phase 4 - January 27,1995 62 h:\essap\repons\shoreham\brookhaven.002 l
1 j TABLE 4 I CONFIRMATORY RADIOLOGICAL STATUS
SUMMARY
-STRUCTURES l- REACTOR BUILDING l' SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK . l l Radiological Survey Unit
""""*7 PC004 PC005 PC007 PC008 PC012 Total Beta Activity (dpm/100 cm 2)* # of Direct Measurements 30 30 30 30 30 Mean(X) 36 310 100 -120 23 UPA X -38 -26 -78 70 -144.4 Fa 86 350 189 -73 63 Generic Guidelines or GAG Satisfied Yes Yes Yes Yes Yes 2
Removable Beta Activity (dpm/100 cm)a,b
# of Smears 30 30 30 30- 30-Mean(X) '
O.3 0.1 0.7 -0.4 -0.2 UPA X 7.0 9.9 2.7 4.8 7.4 ' Fa -1.4 1.5 1.2 0.5 0.6 1,000 dpm/100 cm2 Guideline Satisfied Yes Yes Yes Yes Yes Exposure Rates at 1 m (gR/h)
# of Exposure Rate Measurements 10 8 7 8 2 Net Mean (I) -2,6 -1.2 .-3.0 -2.8 -2.4 LIPA X -1.0 -0.3 -0.5 -0.9 -0.2 Fa -2.5 -1.2 -2.9 -2.8 -1.8 5 pR/h Above Background Guideline Yes Yes Yes Yes Yes Satisfied Shorchara Reactor Buildmg and it.ase 4 - Jamtary 27.1995 63 hw srepomwx e ch-enmumv un
TABLE 4 (Continued) j CONFIRMATORY RADIOLOGICAL STATUS
SUMMARY
-STRUCTURES ; REACTOR BUILDING j SHOREHAM NUCLEAR POWER STATION ' BROOKHAVEN, STW YORK Radiological Survey Unit RB004 RB006 RB014 RB037 RB038 Total Beta Activity (dpm/100 cm 2)*
# of Direct Measurements 30 30 30 30 30 Mean (I) -290 -340 150 -210 -190 LIPA X -48 -220 56 47 86 F= -250 -300 190 -180 -110 Generic Guidelines or GAG Satisfied Yes Yes Yes Yes Yes ;
2 Removable Beta Acthity (dpm/100 cm)a,b
# of Smears 30 30 30 30 30 Mean(X) -0.7 1.1 0.7 1.5 0.1 LIPA X 3.1 6.6 5.1 5.5 5.1-F= 0.1 1.9 1.8 2.7 1.0 1,000 dpm/100 cm2 Guideline Satisfied Yes Yes Yes Yes Yes Exposure Rates at I m (gR/h) # of Exposure Rate Measun:ments 10 4 10 9 6 Net Mean (I) -1.4 -1.7 -0.7 -0.9 -1.1 [
LIPA X -0.3 0.1 -0.2 -0.1 -0.6 Fa -1.3 -1.1 -0.7 -0.7 -0.9 5 pR/h Above Background Guideline Yes Yes Yes Yes Yes ! Satisfied Shoreham Reactor Building and Phase 4 - January 27,1995 64 h:\essap\ reports \shoreham\brookhaven.002 l
TABLE 4 (Continued) CONFIRMATORY RADIOLOGICAL STATUS SUABIARY-STRUCTURES
, REACTOR BUILDING SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK i
i ! Radiological Survey Unit
"""" U RB057 RB061 RB068 RB071x01 RB072x01 l
Total Beta Activity (dpm/100 cnf)* i # of Direct Measurements 30 30 34 30 30 j Mean(X) -240 -170 (1500)* -43 (670)* 85 450 LIPA X 75 55 290 36 140 F= -210 -110 60 (3100)* (1200)* , 250 760 i Generic Guidelines or G,.1G Satisfied Yes Yes Yes Yes Yes Removable Beta Activity (dpm/100 cnf)a,b
# of Smears 30 30 34 30 30 Mean(X) 0.2 -0.6 2.7 1.5 3.5 LIPA X 3.6 9.2 17 4.6 27 F= 1.5 0.2 4.3 3.4 5.3 l 1
1,000 dpm/100 cnf Guideline Satisfied Yes Yes Yes Yes Yes 1 Exposure Rates at 1 m (pR/h)
# of Exposure Rate Measurements 9 10 9 NA 0.4 1 l
Net Mean (I) -0.8 -0.7 0.0 NA 0.4 ' LIPA X 0.0 -0.1 0.0 -0.3 0.1 F= -0.6 -0.6 0.2 NA 0.6 5 pR/h Above Backgrotuxi Guideline Yes Yes Yes Yes Yes Satisfied Shoreham Reactor Building and Phase 4 January 27,1w5 65 hhapVeponhhorehamernoummM2
~
TABLE 4 (Continued) CONFIRMATORY RADIOLOGICAL STATUS
SUMMARY
-STRUCTURES REACTOR BUILDING SIIOREIIAM NUCLEAR POWER STATION BROOKIIAVEN, NEW YORK Radiological Survey Unit RB103 RB106- RB109 RWO73x02 RWO73x03 Total Beta Attivity (dpm/100 cm")"
# of Direct Measurements 30 30 30 25 25 Mean(X) 130 79 100 6400 55 LIPA X 64 55 85 2200 2700 Fa 210 130 210 2900 520 Generic Guidelines or GAG Satisfied Yes Yes Yes Yes Yes 2
Removable Beta Activity (dpm/100 cm)a,b
# of Smears 30 30 30 25 25 Mean(X) 0.4 1.2 0.8 0.4 0.4 LIPA X 7.2 4.9 6.8 15 5.1 F= 1.7 2.0 1.9 1.8 1.8 1,000 dpm/100 cm2 Guideline Satisfied Yes Yes Yes Yes Yes Exposure Rates at 1 m (pR/h)
# of Exposure Rate Measurements 17 9 NA 5 5 Net Mean (X) 0.5 0.5 NA 0.9 0.9 LIPA X -0.3 -0.2 -0.8 0.3 0.4 Fa 0.5 0.7 NA 1.2 1.2 5 R/h Above Background Guideline Yes Yes Yes Yes Yes Satisfied . ,
' Activity levels corrected for Fe-55,11-3, and Ni-63 as appropriate.
b 2 All alpha removable activity was less than 12 dpm/100 cm ,
' Represents survey unit radiological status prior to additional remediation.
Shoreham Reactor Building and Phase 4 - J. mary 27.1995 66 h;Wsap\ reports \shoreham\brookhaven 002
TABLE 5 CONFIRMATORY RADIOLOGICAL STATUS
SUMMARY
-SYSTEMS REACTOR BUILDING SIiOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK Radiological Survey Unit SUO01 SU002 SU004 SU014x12 .SUO14x05 Total Beta Activity (dpm/100 cm')*
- of Direct Measurements 30 5 18 6 10 Mean(X) 170 110 -70 130 -65 LIPA X 470 260 210 960 -130 F= 250 260 4 400 110 Generic Guidelines or GAG Satisfied - Yes Yes Yes Yes Yes 2
Removable Beta Activity (dpm/100 cm)*6 !
- of Smears 30 5 9 5 8 Mean(X) ,
0.6 0.6 -0.2 0.2 0.2 LIPA X 14 5.2 15 13 14 Fa 1.7 2.0 1.6 2.7 2.4 Conditions and 1,000 dpm/100 cm2 Yes Yes Yes Yes Yes Guideline Sctisfied l i l Shoretuun Reactor Building and Itase 4 - January 27.1995 67 hhprquWhwehmnWmLhavenAU
TABLE 5 (Continued) CONHRMATORY RADIOLOGICAL STATUS
SUMMARY
-SYSTEMS REACTOR BUILDING SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK J Radiological Survey Unit Summary SUO23 SUO58 SUO60 SUO61 Total Beta Activity (dpm/100 cnf)"
# of Direct Measurements 26 17 45 19 Mean(X) 50 32 390 280 LIPA 240 210 170 590 go 240 150 530 360' Generic Guidelines or GAG Satisfied Yes Yes Yes Yes Removable Beta Activity (dpm/100 cnf)*6 # of Smears 15 17 45 19 Mean(X) -0.3 2.2 2.8 1.1 LIPA X NA 7.4 17 18 A 17 6.1 4.5 2.6-1,000 dpm/100 cnf Yes Yes Yes Yes Guideline Satisfied ' Activity levels corrected for Fe-55, H-3, and Ni43 as appropriate.
b All alpha removable activity was less than 12 dpm/100 cm2 , 1 Shorehara Reactor Sullding and Phase 4 - January 27.1M 68 WsapVepomuhoretam%rocumenW2
e, REFERENCES.
. Berger, J. D. Manual for Conducting Radiological Surveys .in Support of. License Termination,.
Draft, NUREG/CR-5849. Nyorod by Oak _ Ridge Associated Universities, Oak Ridge, Tennessee
< for the Nuclear Regulatory Commission. June 1992.
Ieng Island Lighting Company, Brookhaven, New York. Shoreham Nuclear Power Station Site
; Characterization Program Final Report. 1990.
l-long Island Power Authority, Brookhaven, New York. Shoreham Decommissioning Project, L Termination Survey Plan, Revision 3. July 1994a. Ieng Island Power Authority, Brookhaven, New-York. Shoreham Decomnussioning Project-i Tennination Survey Final Report, Volumes 1 and 2. October 1994b. Pittiglio, C. L. letter to A. J. Bortz, long Island Power Authority. June 7,1994. U.S. Nuclear Commission. Guidance and Discussion of Requirements for an Application to-l Terminate a Non-Power Reactor Facility Operating License, Revision 1. September 1984J Vitkus, T. J. Ietter to D. Fauver, U.S. Nuclear Regulatory Commission. November 4,1993. Vitkus, T. J. Confinnatory Survey of the Turbine Building, Site Grounds, and Site Exteriors, Shoreham Nuclear Power Station, Brookhaven, New York. Oak Ridge Institute for Science and Education, Oak Ridge, Tennessee. September 1994a. Vitkus, T. J. Ietter to D. Fauver, U.S. Nuclear Regulatory Commission. October 25,1994b. Shorchern Reactor Building and Phase 4. Jamary 27.1995 69 h:bupWpomWrehmWroonamE2
w .- P .. I. APPENDIX A MAJOR INSTRUMENTATION s i \ Shoreharn Reactor Building and Phase 4 - January 27,1995 b hsap\reportsworcham\brookhaven.002
i
; 1 APPENDIX A MAJOR INSTRUMENTATION The display of a specific product is not to be construed as an endorsement of the product or its manufacturer by the authors or their employers. ,
1 DIRECT RADIATION MEASUREMENT Instruments Eberline Pulse Ratemeter Model PRM-6 (Eberline, Santa Fe, NM) , Eberline " Rascal" Ratemeter-Scaler Model PRS-1 (Eberline, Santa Fe, NM) Ludlum Ratemeter-Scaler Model 2221 s (Ludlum Measurements, Inc., , Sweetwater, TX) Detectors Eberline GM Detector Model HP-260 Effective Area,15.5 cm2 , (Eberline, Santa Fe, NM) I Eberline ZnS Scintillation Detector Model AC-3-7 Effective Area,59 cm2 (Eberline, Santa Fe, NM) Ludlum Gas Proportional Detector Model 43-37 .; Effective Area,550 cm2 (Ludlum Measurements, Inc., Sweetwater, TX) i Shoreharn Reactor Nilding and Phase 4 January 27, tws A-1 hwapveponmorcharownmu veam2 l
Imdlum Gas Proportional Detector 1Model 43 68 < Effective Area,100 cm2 h (Imdlum Measurements, Inc., - L Sweetwater, TX) 1
- Reuter-Stokes Pressurized Ion Chamber Model RSS-111 L - (Reuter-Stokes, Cleveland, OH) - - Victoreen Nal Scintillation Detector .
Model 489-55 3.2 cm x 3.8 cm Crystal
' (Victoreen, Cleveland, OH)
LABORATORY ANALYTICALINSTRUMENTATION High Purity Extended Range Intrinsic Detectors Model No: ERVDS30-25195 (Tennelec, Oak Ridge, TN) Used in conjunction with: Lead Shield Model G-11 (Nuclear Lead, Oak Ridge, TN) and Multichannel Analyzer 3100 Vax Workstation (Canberra, Meriden, CT) High-Purity Germanium Detector Model GMX-23195-S,23% Eff. (EG&G ORTEC, Oak Ridge, TN) Used in conjunction with: lead Shield Model G-16 (Gamma Products, Palos Hills, IL) and Multichannel Analyzer 3100 Vax Workstation (Canberra, Meriden, CT) Iow Background Gas Proportional Counter Model LB-5100-W (Oxford, Oak Ridge, TN) . Shoreharn Reactor Building and Phase 4 - January 21. Im A-2 hAessap\ reports \shoreham\brookhaven.002 -
-1 I
l i l APPENDIX B SURVEY AND ANALYTICAL PROCEDURES s Shoreham Reactor Buikimg and Phase 4 January 27,1995 h:\cssap\repats\shoreham\broothaven.002
p '.1 s, < APPENDIX B
- t. SURVEY AND ANALYTICAL PROCEDURES
. SURVEY PROCEDURES Surface Scans
~
Surface scans were performed by passing the probes slowly over the surface; the distance between the probe mid the surface was maintained at a minimum - nominally about 1 centimeter (cm). A-large surface area, gas proponional floor monitor was used to scan the floors of the surveyed areas. Other surfaces were scanned using small area (15.5 cm2 , 59 cm2 or 100 square centimeters [cm2 )) hand-held detectors. Identification of elevated levels was based on increases in the audible signal from the recording and/or indicating instmment. Combinations of detectors and instmments used for the scans were: s Alpha - gas proportional detector with ratemeter-scaler ZnS scintillation detector with ratemeter-scaler I Beta - gas proponional detector with ratemeter-scaler GM pancake detector with ratemeter-scaler Gamma - Na1 scintillation detector with ratemeter 1 Surface Activity Measurements i I Measurements of total beta activity levels wem performed using Geiger-Mueller (GM) and gas I proponional detectors with portable ratemeter-scalers. 1
)
Count rates in counts per minute (cpm), which were integrated over 1 minute in a static position, 2 wem converted to activity levels in disintegrations per minute /100 square centimeters (dpm/100 cm ) Simham Reactor Building and l'hase 4 - January 27.1995 B-1 hwpveromw=hamwnxumam2 j
7 a f by dividing the net rate by the 4 x efficiency and correcting for the active area of the detector. The beta activity background count rates for the GM and gas proportional detectors ranged from 15 to 37 cpm and from 100 to 207 cpm, respectively. Beta efficiency factors were 0.17. for the GM detectors and ranged from 0.21 to 0.23 for the gas proponional detectors. The effective window areas for the GM and the gas proponional detectors were 15.5 cm2 and 100 cm2, respectively. , Surface activity measurements in all confhmatory survey units, except for RB068,' RWO73x02, RWO73 x03, and SU001 that exceeded the normal background distribution were corrected for the. iron-55 contribution by multiplying the dpm/100 cnf field activity level by a factor of 1.2. For the remaining survey units, correction factors for Fe-55, H-3, and/or Ni-63 were as follows:
)
RB068, measured dpm/100 cm2 times 2.3 RWO73 x02, measured dpm/100 cm2 times 35.9 RWO73 x03, measured dpm/100 cm2 times 24.2 SU001, measured dpm/100 cm2 times 2.6 LIPA based each of these correction factors on the relative concentrations of contaminants in the constmetion material of the various survey units. The instmment response level at which the detector output could be considered above background was defined as the critical level (IJ. This level was defined for each detector / instrument combination as follows: g,95 Sample count rate , Background count rate
\ ample count time Background count time L* = (Detector Efficiency) (Detector Geometry)
I Removable Activity Measurements i Removable activity levels were determined using numbered filter paper disks, 47 millimeters (mm) - in diameter. Moderate pressure was applied to the smear and approximately 100 cm2 of the surface was wiped. Smears were placed in labeled envelopes with the location and other peninent information recorded. , Shoreham Reactor Building and Phase 4 - hnuary 27.1995 B-2 hdessap\ reports \shoreham\brookhaven.002 ~.
x - Exposure Rate Measurements ' Measurements of gamma exposure rates were performed using a pressurized ionization chamber (PIC) set I meter from the surface. Water Sampling Approximately 3.8 liters of water was collected, the sample was tansferred to a plastic container, sealed, and labeled in accordance with ESSAP survey procedures. , Sediment Sampline Approximately 1 kilogram of sediment was collected from the trench, placed in a plastic container, sealed and labeled in accordance with ESSAP survey procedures. ANALYTICAL 1ROCEDURES s . Removable Activity -
'i~
Smears were counted on a low background gas proportional system for gross alpha, and gross beta - activity. Gamma Spahwiietry Sediment and water samples were dried, mixed, cmshed, and/or homogenized as necessary, and a portion sealed in 0.5-liter Marinelli beaker or other appropriate container. The quantity placed in the , beaker was chosen to reproduce the calibrated counting geometry, Net material weights were determined and the samples counted using intrinsic germanium detectors coupled to'a pulse height , analyzer system. Background and Compton stripping, peak search, peak identification, and concentration calculations were performed using the computer capabilities inherent in the analyzer system. The energy peak used for determining the activity of the radionuclide of concern was: Co-60 1.173 MeV Spects were also reviewed for other identifiable photopeaks. Shoreham Reactor Buikfmg and Phase 4 - January 27,1995 B-3 h:kssap\ reports \shoreham\brookhaven.002
- r
y ; W, : l I UNCERTAINTIES AND DETECTION LIMITS ' The uncertainties associated with the analytical data presented in the tables of this report mpresent i the 95% confidence level for that' data. These uncertainties were calculated based on both the gross j sample count levels and the associated background count levels. ~ Additional uncertainties, associated - with sampling and measurement procedures, have not been propagated into the data presented in this i L s report. '! t Detection limits, referred to as minimum detectable activity (MDA), were based on 2.71. plus 4.66 i times the standard deviation of the. background count [2.71 + (4.66VBKG)]. Although data is - l reported as actual values, including negative values, in the document text and table, the MDAs for l total and removable activity levels are provided in the footnotes of applicable tables. Because of variations in background levels, measurement efficiencies, and contributions from other radionuclides - ! t in samples, the detection limits differ from sample to sample and instmment to instmment. - l l CALIBRATION AND QUALITY ASSURANCE j s -! Calibration of all field and laboratory instrumentation was based on standards, traceable to the ~ j National Institute of Standards and Technology (NIST), when such standard were available. In cases where they were not available, standards of an industry recognized organization were used. Calibration of pressurized ionization chambers was performed by the mai.ufacturer. t i i Analytical and field survey activities were conducted'in accordance with procedures from the ; " following documents of the Environmental Survey and Site Acceament Program: f
- Survey Procedures Manual, Revision 8 (December 1993) -
l
- 12boratory Procedures Manual, Revision 8 (August 1993).
- Quality Assurance Manual, Revision 6 (July 1993l !
i i The procedures contained in these manuals were developed to meet the requirements of DOE Order j 5700.6C and ASME NQA-1 for Quality Assuranco and contain measures to assess processes during l thCir perforinance. l Shoreharn Reactor Building and Phase 4 - January 27.1995 B-4 khpVepombhorchamWndimenR2
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* : Quality control procedures includei >
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-* ? Daily instrument background and . check-source measurements -to confum' that ; ;
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-* Participation in EPA and DOFJEML Quality Assurance Programs. -i P
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'" ~*: Training and certification of all individuals performing piuceann. -l A
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*- Periodic ~ internal and external audits.
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f i P 4 APPENDIX C , REGULATORY GUIDE 1.86, TERMINATION OF OPERATING l LICENSES FOR NUCLEAR REACTORS F s
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- shoream ne.cin amm .w n 4 3a,iuary 27,1995 h:\essap\ reports \shoreham\ brook!aven.002 ]
i
U.S. ATOMIC ENERLY COMMISSION June 1974 - REGULATORY GUIDE y
' DIRECTORATE OF REGULATORY STANDARDS L
I' l REGULATORY GUIDE 1.86 TERMINATION OF OPERATING LICENSES FOR NUCLEAR REACTORS A. INTRODUCTION important to the safety of reactor operation is no longer required. Once this possession-only license is issued, Section 50.51, " Duration of license, renewal," of 10 reactor operation is not permitted. Other activities from CFR Part 50, " Licensing of Production and Utilization the reactor and placing it in storage (either onsite or Facilities," requires that each ' license to operate a offsite) may be continued. production and utilization facility be issued for a specified duration. . Upon expiration of the specified period, the A licensec having a possession-only license must retain, license may be either renewed or terminated by the with the Part 50 license, authodzation for special nuclear Commission. Section 50.82, " Applications for termmation material (10 CFR Part, 70, "Special Nuclear Material"), of licenses," specifies the requirements that must be byproduct matenal (10 CFR Part 30, " Rules of General satisfied to terminate an operating license, including the Applicability to Licensing of Byproduct Material"), and requirement that the dismantlement of the facility and source snaterial (10 CFR Part 40, "Iacensing of Source disposal of the component parts not be inimical to the Material"), until the fuel, radioactive components, and common defense and security or to tho. health and safety sources are removed from the facility. Appropriate of the public. This guide describes methods and admmistrative controls and facility requirements are procedures considered acceptable by the Regulatory staff imposed by the Part 50 license and the technical for the termmation of operating licenses for nuclear s},zifications to assure that proper surveillance is , reactors. The advisory Committee on Reactor Safeguards performed and that the reactor facility is maintained in a ] has been consulted concernmg this guide and has safe condition and not operated. concurred in the regulatory position. A possession-only license permits various options and B. DISCUSSION procedures for decommissioning, such as mothballing, entombment, or dismantling. The requirements imposed When a licensee decides to temunate his nuclear depend on the option selected. reactor operating license, he may, as a first step in the process, request that his operating license be amended to Section 50.82 provides that the licensee may dismantle i restrict him to possess but not operate the facility. The and dispose of the component parts of a nuclear reactor in i I advantage to the licensee of converting to such a accordance with existing regulations. For research possession-only license is reduced surveillance reactors and critical facilities, this has usually meant the - requirements in that periodic surveillance of equipment disassembly of a reactor and its shipment organintion for { l USAEC REGULATORY GUIDES k copies of putin hed guidas may be obtained by request ind'catino the 8viaton Regulatory Guidue are issued to duecrite and inske evaHable to the pubnc desired to the U.s. Atomic Energy c_ ' . Washington. D.e. 20545. methods acceptable to the AEC regulatory staff of bnplementing specific parts Attention: Director of Regulatory Standards. Comments and suggestions for of the Commiasson's regulations, to originate techniques imod by the staff in improvements in those guides are encouraged and should be sent to the evaluating specific problems or postuisted accidents, or to provide guidance to # ##TM mopAcants. Regulatory Guldse are not substitutes for regulations and
- compEence with thern is not required. Methods and solutions efferent from those set out in the guidea we to acceptate e they provide e base for the The guides are issued in the folowing ten broad divisions.
requisne to the issuance or continuance of a pomut or acense by the
- 1. Power Reactors 6. Products
- 2. Research and test Reactors 7. Transportation Paekshed guides we be revised penodicany. es appropr6ste, to accommodate 3. Fuels and Materlais Facaittes 8. Occupatkinal Keith comments and to reflect new informauon er experience. "" " A" 'A" Note: Section electronically reproduced frorn photocopy. C-l
m funher use. The site from which a reactor has been c. Any proposed changes to the technical specifications removed must be dwkminated, as necessary, and that reflect the possession-only facility status and the inspected by the Commission to determine whether necessary disassembly / retirement activities to be ; unrestricted acem can be approved. In the case of performed. l nuclear power reactors, dismantling has usually been accomplished by shipping fuel offsite, making the reactor d. A safety analysis of both the activities to be inoperable, and disposing of some of the radioactive accomplished and the proposed changes to the technical
]
components. specifications. I Radioactive components may be either shipped off-site e. An inventory of activated materials and their for burial at an authorized burial ground or secured on the location in the facility.
' site. Rose radioactive materials remammg on the site must be isolated from the public by physical barriers or 2. ALTERNATIVES FOR REACTOR RETIREMENT other means to prevent public access to hazardous levels of radiation. Surveillance is necessary to assure the long Four altematives for retirement of nuclear reactor term integrity of the barriers. The amount of surveillance facilities am considered acceptable by the Regulatory required depends upon (1) the potential hazard to the staff. Rese are:
health and safety of the public from radioactive material remauung on the site and (2) the integrity of the physical a. Mothballing. Mothballing of a nuclear reactor barriers. Before areas may be released for unrestricted . facility consists of putting the facility in a state of use, they must have been decontammated or the protective storage. In general, the facility may be left radiotavity must have decayed to less than prescribed intact except that all fuel assemblies and the radioactive limits (Table 1). fluids and waste should be removed from the site. Adequate radiation monitoring, environmental The hazard associated with the returned facility is surveillance, and appropriate security procedures evaluated by considering the amount and type of remaming should be established under a possession-only limnse contammation, the degree of confinement of the remammg to ensure that the health and safety of the public is not radioactive materials, the physical security pmvided by the endangered. , confinement, the susceptibility to release of radiation as a result of namral phenomena, and the duration of reqmred b. In-Place Entombment. In-place entombment surveillance. consists of sealing all the remammg highly radioactive or contammated components (e.g., the pressure vessel C. REGULATORY POSITION and reactor intemals) within a structure integral with the biological shield after having all fuel assemblies,
- 1. APPLICATION FOR A LICENSE TO POSSESS radioactive fluids and wastes, and certam selected BUT NOT OPERATE (POSSESSION-ONLY components shipped offsite. He stmettre should LICENSE) pmvide integrity over the period of time in which significant quantities (greater than Table 1 levels) of A request to amend an operating license to a radioactivity remam with the matenal in the possession-only license should be made to the Director of entombment. An appmpriate and continuing Licensing, U.S. Atomic Energy Comnussion, Washington, surveillance program should be established under a D.C. 20545. The request should include the following possession-only license.
informanon:
- c. Removal of Radioactive. Components and t a. A desenption of the current status of the facility. Dismantling. All fuel assemblies, radioactive fluids and waste, and other materials having activities above
- b. A description of measures that will be taken to accepted unrestricted activity levels (Table 1) should be prevent criticality or reactivity changes and to mmunize removed from the site. He facility owner may then teleases of radioactivity from the facility. have unrestricted use of the site with no requirement for a license. If the facility owner so desires, the Note: Section electronically reproduced from photocopy. C-2 5L
6,,- r > remainder of the reactor facility may be dismantled and barriers in the facility. Sampling should be done along the g
, all vestiges removed and disposed of. most probable path by which radioactive material such as that stored in the inner contamment regions could be
- d. Convenion to a New Nuclear System or a Fossil transponed to the outer regions of the facility and i Fuel System. This alternative, which applies only to' ultimately to the environs, nuclear power plants, utilizes the existing turbine
. system with a new steam supply system. He original d. An environmental radiation survey should be nuclear steam supply system should be separated from performed at least semiannually to verify that no the electric generatmg system and disposed of in significant amounts of radiation have been released to the accordance with one of the previous three retirement environment from the facility. Samples such as soil, e altematives. vegetation, and water should be taken at locations for which statistical data has bece established during reactor
- 3. SURVEILLANCE AND SECURITY FOR THE operations.
RETIREMENT ALTERNATIVES WHOSE FINAL L STATUS REQUIRES A POSSESSION-ONLY e. A site representative should be designated to be LICENSE responsible for controlling authorized access into and movement within the facility. A facility which has been licensed ' under a possession-only license may contain a significant amount 'f. Administrative procedures should be established for of radioactivity in the form of activated and mnt=inal the notification and reponing of abnormal occurrences hardware and stmetural materials. Surveillance and such as (1) the catrance of an unauthorized person or commensurate security should be pmvided to assure that persons into the facility and (2) a significant change in the the public health and safety are not endangered. radiation or contammation levels in the facility or the
- a. Physical security to prevent inadvenent exposure of offsite environment.
personnel should be pmvided by multiple locked barriers. De presence of these barriers should make it extremely g. -The following repons should be made: difficult for an unauthorized person to gain acem to areas where radiation or contammation levels exceed those (1) An annual repon to the Director of Licensing, specified in Regulatory Position C.4. To prevent U.S. Atomic Energy Commission, Washington, D.C. inadvertent exposure, radiation arcas above 5 mR/hr, such 20545, describing the results of the environmental and as near the activated primary system of a power plant, facility radiation surveys, the status of the facility, and an - should be appropriately. marked and should not be evaluation of the performance of security and surveillance accessible except by cutting of welded closures or the measures. disassembly and removal of substantial stmetures and/or shielding material. Means such as a rennte-readout (2) An abnormal occurrence report to the Regulatory - intrusion alarm system should be provided to indicate to Operations Regional Office by telephone within 24 hours designated pwouc! when a physical barrier is penetrated. of discovery of an abnormal occurrence. The abnormal
- Security personnel that provide access control to the occurrence will also be reported in the annual report facility may be used instead of the physical barriers and described in the precedmg item.
, the intmsion alarm systems.
- h. Records or logs relative to the following items
- b. The physical barriers to unauthorized entrance into should be kept and retamed until the license is tenmaated, the facility, c'.g., fences, buildings, welded doors, and after which they must be stored with other plant records:
access openings, should be inspected at least quanctly to 4 assur: that these barriers have not deteriorated and that (1) Environmental surveys, 1 eks and lockiog apparatus are intact. (2) Facility rn:liation surveys,
- c. A facility radiation survey should be performed at (3) Inspecuans of the physical barriers, and l= least quanerly to verify that no radioactive material is (4) Abnormal occurrences
, escaping or being transponed through the containment
. Note: Section electronically reproduced from photocopy. C-3
- 4. DECONTAMINATION FOR RELEASE FOR (2) A detailed health and safety analysis indicating that - l UNRESTRICTED USE the residual arnounts of n'aterials on surface areas, together with other considerations such as the prospective if it is desired to terminate a license and to climmate use of the premises, equipment, or scrap, are unlikely to any funher surveillance requirements, the facility should result in an unreasonable risk to the health and safety of be sufficiently decontaminated to prevent risk to the public the public.
health and safety. After the decontamination is satisfactorily accomplished and the site inspected by the e. Prior to release of the premises for unrestricted use, Commission, the Commission may authorize the license to the licensee should make a comprehensive radiation survey be teiminated and the facility abandoned or released for establishing that contammation is within the limits specified unrestneted use. The licensee should perform the in Table 1. A survey report should be filed with the decontamucion using the following guidelines: Dimetor of Licensing, U.S. Atomic Energy Commission, Washington, D.C. 20545, with a copy to the Director of
- a. The licence should make a reasonable effort to the Regulatory Operations regional Office having eliminate residual contammation. jurisdiction. He report should be filed at least 30 days prior to the planned date of abandonment. H e survey
- b. No covering should be applied to radioactive report should:
surfaces of equipment of structures by paint, plating, or other covering material until it is known that contammation (1) Identify the premises; levels (determ.ned by a survey and documented) are below the limits specified in Table 1. In addition, a reasonable (2) Show that reasonable effort has been made to effort should be made (and documented) to further reduce residual contamination to as low as practicable minimize contanunation prior to any such covering. levels;
- c. De radioactivity of the interiorJurfaces of pipes, (3) Describe the scope of the stuvey and the general drain lines, or ductwork should be detennined by making procedures followed; and measurements at all traps and other appropriate access points, provided contamination at these locations is likely (4) State the fmding of the survey in units specified in to be representative of contamination on the interior of the Table 1.
pipes, drain lines, or ductwork. Surfaces of premises, equipment, or scrap which are likely to be contammated After review of the report, the Commission may but are of such size, construction, or location as to make inspect the facilities to confirm the survey prior to granting the surface inaccessible for purposes of measurement approval for abandonment. should be assumed to be contammated in excess of the permissible radiation limits.
- 5. REACTOR RETIREMENT PROCEDURES
- d. Upon request, the Commission may authorize a licensee to relinquish possession or control of premises, As indicated in Regulatory Position C.2, several equipment, or scrap having surfaces contammated in altematives are acceptable for reactor facility retirement.
excess of the limits specified. His may include, but is not if minor disassembly or "mothballing" is planned, this limited to, special circumstances such as the transfer of could be done by the existing operating and mamtenance premises to another licensed organization that will continue pmcedures under the license in effect. Any planned to work with radioactive materials. Requests for such actions involving an unreviewed safety question or a authorization should provide: change in the technical specifications should be reviewed and approved in accordance with the requirements of 10 (1) Detailed, specific information describing the CFR I 50.59. premises, equipment, scrap, and radioactive contaminants and the nature, extent, and degree of residual surface If major stmetural changes to radioactive components contammation. of the facility are planned, such as remont of the pressure vessel or major components of the primary system, a Note: Section electronically reproduced from photocopy. C-4
E l dismantlement plan including the information required by : [ l 50.82 should be submitted to the Commission. A dismantlement plan should be submitted for all the . alternatives of Regulatory Position C.2 except anothballing. However, minor disassembly activities may still be performed in the absence of such a plan, provided they are , permitted by existing operating and mamtenance l procedures. A dismantlement plan should include the ] following:
- a. A description of the ultimate status of the facility
- b. A description of the dismantling activities and the i
precautions to be taken.
- c. A safety analysis of the dismantling activities including any effluents which may be released.
i
- d. A safety analysis of the frility in its ultimate l status. I
)
Upon satisfactory review and approval of the dismantling plan, a dismantling order is issued by the Commission in accordance with 6 50.82. When dismantling is completed and the Commission has been notified by letter, the appropriate Regulatory Operations Regional Office mspects the facility and verifies completion in accordance with the dismantlement plan. If residual radiation levels do not exceed the values in Table 1, the Comnussion may turmnate the license. If possession-only license under which the dismantling activities have been conducted or, as an altemative, may make application to the State (if an Agreement State) for a byproduct materials license. Note: Section electronically reproduced from photocopy. C-5
TABLE 1 ACCEPTABLE SURFACE CONTAMINATION LEVELS Nuclide' Average Maximum 6d Removable6 ' U-nat, U-235, U-238, and associated decay products 5,000 dpm a/100 cm2 15,000 dpm a/100 crd 1,000 dpm a/100 ed Transuranics, Ra-226, Ra-228, R-230, %-228, Pa-231 Ac-227, I-125, I-129 100 dpm/100 cid 300 dpm/100 cnf 20 dpm/100 cnf Th-nat, R-232, Sr-90, Ra-223 Ra-224, U-232, I-126,1-131,1-133 1,000 dpm/100 cd 3,000 dpm/100 cnf 200 dpm/100 ed Beta-gamma emitters (nuclides with decay modes other than alpha emission or spontaneous fission) except Sr-90 and others
~
noted above. 5,000 dpm Sy/100 cd 15,000 dpm 67/100 ed 1,000 dpm Sy/100 cd f
'Where surface contammation by both alpha- and beta-gamma-emitting nuclides exists, the limits c::tablished for alpha- and beta-gamma <mitting nuclides should apply independently. 'As used in this table, dpm (disintegrations per minute) means the rate of emission by radioactive material as determmed by correcting the counts per minute observed b'y an appropriate detector for background, efficiency, and geometric factors associated with the instrumentation. ' Measurements of average contammant should not be averaged over more than I square meter. For objects ofless surface area, the average should be derived for each such cyect. 'The maximum contammation level applies to an area of not more than 100 crd. 'Ibe amount of removable radioactive material per 100 cm2of surface area should be determined by wiping that area with dry filter or soft absorbent paper, applying moderate pressure, and assessing the amount of radioactive material on the wipe with an appropriate instmment of known efficiency. When removable contammation on objects of less surface area is determin;d, the pertinent levels should be reduced proportionally and the entire surface should be wiped.
t Note: Section electronically reproduced from photocopy. C-6}}