ML20076F929
| ML20076F929 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 09/30/1994 |
| From: | Vitkus T OAK RIDGE ASSOCIATED UNIVERSITIES |
| To: | NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| Shared Package | |
| ML20076F927 | List: |
| References | |
| CON-FIN-A-9076 ORISE-94-I-80, NUDOCS 9410180258 | |
| Download: ML20076F929 (78) | |
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x lCONFIRMATORYLSURVEY (OFTHE TURBINE BUILDING, B SITE GROUNDS /AND SITE-EXTERIORS
.SHOREHAM' NUCLEAR POWElf STATION N fBRO'OK-HAVEN,;NEW YORKE
[D6CKET No. 50-322]
T.;]. VITKUS .
1 Prepared for the Division _ of Waste Management y Headquarters Office . .
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The Oak Ridge Institute for Science and Education (ORISE) was established by the U.S. Department of Energy to undertake national and international programs in science and engineering education, training and management systems, energy and environment systems, and medical sciences. ORISE and its programs are operated by Oak Ridge Associated l; Universities (ORAU) through a management and operating contract with the U.S. Department of Energy. Established E in 1946, ORAU is a consortium of 82 colleges and universities.
< NOTICES The opinions expressed herein do not necessarily reflect the opinions of the sponsoring institutions of Oak Ridge ,
Associated Universities.
This report was prepared as an account of work sponsored by the United States Government. Neither the United States l
Government nor the U.S. Department of Energy, nor any of their employees, makes any warranty, expressed or implied, N or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe on privately owned rights. g i
Reference herein to any specific commercial product, process, or service by trade name, mark, manufacturer, or g otherwise, does not necessarily constitute or imply its endorsement or recommendation, or favor by the U.S. Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the U.S. Government or any agency thereof.
E
ORISE 94/l-80 CONFIRMATORY SURVEY OF THE l I TURBINE BUILDING, SITE GROUNDS, AND SITE EXTERIORS SHOREHAM NUCLEAR POWER STATION l
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BROOKHAVEN, NEW YORK I Prepared by T. J. Vitkus I Environmental Survey and Site Assessment Program Energy / Environment Systems Division I, Oak Ridge Institute for Science and Education Oak Ridge, Tennessee 37831-0117 Prepared for the U.S. Nuclear Regulatory Commission Headquarters Office Sponsored by the Division of Waste Management E
FINAL REPORT I
SEPTE51BER 1994 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.
m Oak Ridge Institute for Science and Education performs complementary work under contract g number DE-AC05-760R00033 with the U.S. Department of Energy. ,
I Shoreham Nc' ear Power StsGon . September 9,1994 k
CONFIRMATORY SURVEY OF THE
[.. TURBINE BUILDING, SITE GROUNDS, AND SITE EXTERIORS SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK
( Prepared by Date: k!7!Y '
j T. J. Vitkuf, Project Leader j Environmental Survey and Site Assessment Program l l
I Reviewed by: deuK kd% Date: 4 h 4 'f +
M. J. Laudeman, Radiochemistry Laboratory Supervisor Environmental Survey and Site Assessment Program l Reviewed by: % -
N Date: /4!9_$'
j A. T. Payne, Health & Safety toordinator Environmental Survey and Site Assessment Program Reviewed by: / Date: 9//MM W. L. Beck, Acting Program Director ' 7 '
Environmental Survey and Site Assessment Program 9
Shoreham Nalear Power Suta - Semmber 9.1994
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ACKNOWLEDGEMENTS
}
r- The author would like to acknowledge the significant contributions of the following staff L-' members:
)
( FIELD STAFF !
S. F. Barnett M. A. Henke E. H. Montalvo J. R. Morton D. M. Nugent J. L. Payne LABORATORY STAFF R. D. Condra J. S. Cox M. J. Laudeman CLERICAL STAFF
(: T. T. Claiborne D. A. Cox R. D. Ellis K. E. Waters ILLUSTRATORS M. A. Henke T. D. Herrera Shorehmen Nuclear twer sentina September 9,1994
E -
R TABLE OF CONTENTS PAGE List o f Fi gures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii I List of Tables ............................................ . iv Abbreviations and Acronyms . . . . . . . . . . . . . . . . . . . . . . . . . ............v Introd uction and Site History . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 Site Description .. ................................... ..... .2 Obj ec ti ve s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ......... 3 1
Im I
Document Review and LIPA Procedure Surveillance . . . . . . . . . . . . . . . . . . . . . . . 4 Procedures ................................................ 4 Findings and Results .........................................10 Comparison of Results with Guidelines . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 Summary . ..............................................15 References .. ............................................53 Appendices:
Appendix A: Major Instrumentation Appendix B: Survey and Analytical Procedures E Appendix C: Regulatory Guide 1.86 Termination of Operating Licenses for Nuclear Reactors I
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hreharn Nwicar Power Statmu - September 9.1994 i
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LIST OF FIGURES PAGE FIGURE 1: Location of the Shoreham Nuclear Power Station . . . . . . . . . . . . . . . 16 I FIGURE 2: Plc t Plan of the Shoreham Nuclear Power Station . . . . . . . . . . . . . . . 17 FIGURE 3: Turbine Building, Elevation 15'-Floor Plan and Areas Surveyed . . . . . 18 FIGURE 4: Turbine Building, Elevation 37'6"-Floor Plan and Areas Surveyed . . . 19 FIGURE 5: Turbine Building, Elevation 63'-Floor Plan and Areas S;uveyed . . . . . 20 FIGURF 5: Shoreham Nuclear Power Station, Restricted Area-Exterior Areas Surveyed ..................... .............21 FIGURE 7: Turbine Building, North Condenser Hallway (TB016)-Measurement and Sampling Locations ........... ..... ............22 FIGURE 8: Turbine Building, West Condenser Bay (TB017)-Measurement and Sampling I ocations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 FIGURE 9: Turbine Bc ' ding, Steam Seal Evaporator Room (TB031)-Measurement and Samphng Locations ........................... . . 24 FIGURE 10: Turbine building, Truck Bay (TB035)-Measurement and Sampling Iccatiens ..............................25 FIGURE 11: Turbine Baildicg, Chemistry Laboratory (TB060)-Measurement and Sampling Locations .............................. 26 I FIGURE 12: Turbine Euilding, Re-Heater Area, East (TB081)-Measurement and Sampling Locations .................... . . . . . . . . 27 lE FIGURE 13. Turuine Building Re-Heater Area, West (TB082)-Measurement and i Samp!!ng Locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 l FIGURE 14: Turbine Building, Black Battery Charger Room (TB089)-Measurement g and Sampling Locations ............ ... .............29 5
FIGURE 15: Colt Emergency Diese! Generator Building-Background Measurement Locations . . . . . . . . . . . ... .... ........ . 30 FIGURE 16: Turbine Building, Feedwater Control (SUOO5)-Measurement and Sampling Locations ............. ......... . . . . . 31 Shoreham % dear Power Stauon September 9,194 55
E -
1 LIST OF FIGURES (Continued) l PAGE I FIGURE 17: Turbine Building, Radwaste Interior Drain Pipe System Tank 12 (SUO14 x03)-Measurement and Sampling Locations .......... . 32 I FIGURE 18: Turbine Building, Main Steam Components (SUO24)-Measurement and Sampling I.ocations ..............................33 I FIGURE 19: Turbine Building,15' Elevation, Condensate and Feedwater, Main l
I Condenser, (SUO25 x02)-Measurement and Sampling Locations . . . . . 34 FIGURE 20: Turbine Building, Lube Oil Sump Tank 91 (SUO32)-Measurement and Sampling Locations ..............................35 FIGURE 21: Turbine Building, Extraction Steam Valve 035C (SUO34)-
Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . 36 FIGURE 22: Condensate Transfer and Storage, Condensate Storage Tank (SUO46x02)-Measurement and Sampling Locations ............37 g FIGURE 23: Turbine Building, Influent Drain System, Low Conductivity and 3 Salt Water Drains, Tank 186B (SUO54 x03)-Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38
)
FIGURE 24: Secondary Access Facility, Ventilation System (SUO71)-Measurement and Sampling Locations ..............................39 l
FIGURE 25: Turbine Building,15' and 37' 6" Elevations, Drains (SUO14) and Vents-Measurement and Sampling Locations ..............40 FIGURE 26: Secured Area, Colt Emergency Diesel Generator Building Exterior (SE002)-Measurement and Sampling Locations . . . . . . . . . . 41 FIGURE 27: Secured Area, Chlorine Pumphouse Exterior (SE002)-Measurement and Sampling Locations ..............................42 FIGURE 28: Secured Area West (SG003)-Measurement and Sampling Locations . . . 43 FIGURE 29: Background Soil Sampling and Exposure Rate Measurement Locations . 44 Shoreham Nuclear Power Stauim September 9,1994 iik
LIST OF TABLES PAGE TABLE 1: Summary of Surface Activity Levels . . . . . . . . . . . . . . . . . . . . . . . 45 TABLE 2: Interior Exposure Rates . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 47 TABLE 3: Exterior Exposure Rates and Co-60 Concentrations in Surface Soil Samples .....................................48 TABLE 4: Confirmatory Radiological Status Summary-Structures, Site Grounds, and Site Exteriors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 49 TABLE 5: Contirmatory Radiological Status Summary-Systems . . . . . . . . . . . . 51 Shoreham Nancar Power Stauon . September 9.1994 iV l
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ABBREVIATIONS AND ACRONYMS ac acre ASME American Society of Mechanical Engineers em 2 square centimeter l cpm counts per minute I
dpm/100 cm2 disintegrations per minute per 100 square centimeters EML Environmental Measurements Laboratory EPA Environmental Protection Agency l
ESSAP Environmental Survey and Site Assessment Program I ft2 ha square feet hectare GM Geiger-Mueller ,
km kilometer j LILCO Long Island Lighting Company j LIPA Long Island Power Authority !
m2 square meter MDA minimum detectable activity j mi mile i Nal sodium iodide NIST National Institute of Standards and Technology NRC Nuclear Regulatory Commission ORISE Oak Ridge Institute for Science and Education QA Quality Assurance ;
SNPS Shoreham Nuclear Power Station SE# site exterior survey unit designation SG# site grounds survey unit designation SU# system survey unit designation TB# Turbine. Building structural survey unit designation brd.am Nuclear Power Stanoo . September 9.1994 V
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CONFIRMATORY SURVEY l OF TIIE TURBINE BUILDING, SITE GROUNDS AND SITE EXTERIORS
- SIIOREIIAM NUCLEAR POWER STATION l- BROOKIIAVEN, NEW YORK I
1 INTRODUCTION AND SITE IIISTORY I
The Long Island Lighting Company (LILCO) constructed a boiling water reactor, known as the I Shoreham Nuclear Power Station (SNPS), which was designed to provide a r,ross electrical .
output of 849 Megawatts. Reactor criticality was achieved in February 1985. I.ow power testing, in accordance with U.S. Nuclear Regulatory Commission (NRC) License No. NPF-82 (NRC Docket File No. 50-322), which permitted reactor operations at levels not to exceed 5%
of full power, commenced in July 1985. Reactor operations continued intermittently until !
January 1989, at which time power generating operations were terminated. The total operating history was equivalent to 2.03 effective full power days of fuel exposure. Irradiated fuel, which was a standard low enrichment (2 to 3 % uranium-235) uranium fuel, was subsequently removed from the reactor vessel and placed into the spent fuel pool in August 1989.
Various reactor components, piping systems, and other equipment became radiologically contaminated as a result of reactor operation. The primary contaminants which have been identified during characterization studies include iron-55, cobalt-60, nickel-63, and smaller quantities of tritium, carbon-14, nickel-59, manganese-54, zinc-65, and europium-152.
The Long Island Power Authority (LIPA) was established to decommission the facility and release the site for unrestricted use. LIPA's decommissioning plan was approved for implementation by the NRC in June 1992 and will include decontamination or removal of contaminated portions of the reactor and other plant systems and equipment. A major consideration of the decommissioning plan is to maintain the integrity when possible, of plant structures and systems. Activities involved with the decommissioning and termination surveys will be conducted over an approximate eighteen month period with the fimal phase being removal of the spent fuel from the site. The initial phase involved the termination survey of the internal components of the main turbine; which has since been followed by termination surveys of the Shoreham Nuclear Power Stauon September 9 lW4
E -
1 i
remainder of the structures and systems located within the Turbine Building as well as the site 1
grounds and building exteriors.
l It is the policy of the NRC to perform confirmatory surveys of facilities that have undergone decommissioning and have requested NRC license termination. The NRC Headquarters' Division of Waste Management, formerly the Division of Low-Level Waste Management and Decommissioning, has requested that the Environmental Survey and Site Assessment Program (ESSAP) of the Oak Ridge Institute for Science and Education (ORISE) conduct confirmatory j radiological surveys and related activities for the SNPS decommissioning project as the various decommissioning milestones are completed. The results of the confirmatory survey of the j turbine internal components are the subject of a separate report.2 This report describes the results of the confirmatory process which has been completed for the Turbine Building, site l grounds, and building exteriors.
I
} SITE DESCRIPTION i
l SNPS is located in the Town of Brookhaven, New York on the north shore of Long Island, approximately 80 km (50 mi) east of La Guardia Airport and the confluence of the East River l and Long Island Sound (Figure 1). Reactor and supporting operations were conducted within l a 32.4 ha (80 ac) portion of a larger 202 ha LILCO owned parcel of land that is bounded on the north by Long 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
{
l known as the power block, where radiological controls were necessary (Figure 2). Each of the buildings that are to be addressed during the confirmatory surveys are located here and are shown on Figure 2 as the Turbine Building, the Reactor Building, and the Rad Waste Building.
Turbine Building construction is predominately of concrete and structural steel with a total floor space of 13,500 m2 (145,000 ft2) that is divided between three levels at elevations 15',37' 6",
and 63' (Figures 3 through 5). Surfaces and components within the building remain essentially intact following decommissioning activities. The systems and equipment housed include the turbine generator, main condenser, condensate system, feed water system, extraction heaters, Shoreham Eclear Power Station Sepunber 9.1994 2
part of the off-gas rad waste system, and drain sumps. Site grounds and site exteriors encompass those land areas and building exteriors and roof tops contained within the secured area.
Termination surveys have been performed in accordance with Draft NUREG/CR-5849.3 LIPA has classified plant systems, building surfaces, and outside areas into two categories for survey, which are based on the potential for residual contamination. The two categories,' referred to as affected or unaffected, are defined as follows: "affected areas are those areas which are potentially contaminated or have known contamination, or a system which circulated, stored 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 residual radioactivity." Area classification was determined by radiological use history, environmental monitoring activities, and the results of the previous characterization survey.
Affected and unaffected areas are further subdivided into survey units. Survey units are categorized as structures (floors, walls, ceilings, and exterior surfaces of piping and equipment),
plant systems (equipment and piping internals), and exterior areas (grounds and building exteriors). In addition, affected survey units also have sub-classifications as suspect or non-suspect, and may also be classified as alpha affected if involved with fuel handling or storage.
For the Turbine Building, site grounds, and site exteriors, there were a total of 191 survey units addressed, of which,135 were structures (including building exteriors),48 were systems, and 8 were site grounds. Twenty-four of these survey units were classified by the licensee as )
1 affected. i OBJECTIVES The objectives of the confirmatory activities were to provide independent document reviews, review and perform FM observations of the LIPA procedures for embedded piping surveys, and develop radiological data for use by the NRC in evaluating the adequacy and accuracy of the licensee's procedures and termination survey results.
Shoreham Nudear Power Stauon - Seriember 9, != 3
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I DOCUMENT REVIEW AND LIPA PROCEDURE SURVEILLANCE I ESSAP reviewed LIPA's termination survey procedures and the termination survey release records for those survey units selected for confirmatory survey.'d Documents were reviewed for adequacy, accuracy, completeness, and consistency. In addition, ESSAP reviewed and initiated observational surveillance of the embedded piping procedures for appropriateness and consistency in field application.
I PROCEDURES E During the period November 8 through 12, 1993, an ESSAP team visited the SNPS and performed independent visual inspections, measurements, and sampling of the Turbine Building, site grounds, and site exteriors. Surveys were performed in accordance with a survey plan submitted to and approved by the NRC.* ESSAP randomly selected 8 of the Turbine Building structural survey units,3 of the system survey units and 1 each of the site ground and building exterior survey units for confirmatory surveys. In addition, the NRC site representatives, selected portions of 6 additional system survey units for confirmatory surveys. Survey unit designators are alpha-numeric with the first Egures designating the type of unit, structural (building speciGc), system, grounds, or exteriors, followed by a three digit numeric reference.
Subunits are given an additional two digit designation preceded by X. The survey units selected I and the respective classiGcation for each were:
8 Survey Unit Survey Unit Name Affected(A)/ Structure / System /
5 Unaffected (U) Buildmg Grounds TB016 North Condenser Hallway U structure TB017 West Condenser Bay U structure TB031 Steam Seal Evaporator Room A structure I TB035 Truck Bay A structure TB060 Chemistry Laboratory A structure TB081 Re-Heater Area-East A structure TB082 Re-Heater Area-West A structure sww smie., ro. , si.uon september 9,1994 4
I Affected(A)/ Structure / System /
Survey Unit Survey Unit Name Unaffected (U) Building Grounds TB089 Black Battery Charger Room U structure SU005 Feed Water Control A system SU014 Radwaste-Turbine Building A system l Drain Piping System l SU014X03 Radwaste-Turbine Building A system Drain Piping System, Tank 12 SUO24 Main Steam A system l SUO25X2 Condensate and Feedwater, A system 15' Elevation Main Condenser SUO32 Lube Oil Sump Tank 91 U system SUO34 Extraction Steam Valve 035C A system SUO46 Condensate Transfer and A system Storage SUO54X03 Low Conductivity and Salt A system Water Drains Influent Drain System Tank 186 SUO71 Secondary Access Facility U system Ventilation System SE002 Secured Area-North U structure Buildings SG003 Secured Area West U bldg. grounds Figures 3 through 6 indicate the structural and exterior survey units surveyed.
Shoreham Nuclear Power Statma September 9,1944 5 ,
SURVEY PROCEDURES: INTERIOR The following procaiures apply to interior structural and system survey units.
Reference System LIPA established the grid system used by ESSAP for referencing measurement ar.d sampling locations. The grid size or reference interval established by LIPA for a given survey unit was dependent upon the survey unit classification (affected vs. unaffected) and surface (floor, lower wall, upper wall, ceiling, or equipment). Typically, floor and lower wall grid blocks were 1 m J x 1 m. Upper surfaces, ceiling and equipment were either referenced to these grids or other prominent building features. Systems were referenced by a distance from a specific point, by drawings, or prominent components.
Surface Scans-Structural Units Surface scans for alpha, beta, and gamma activity, were performs ' over 100% of floor, and t
} lower wall surfaces, and up to 50% of equipment surfaces, within each structural survey unit.
Additional scans were performed over portions of upper wall, ceiling, and/or system surfaces as well as locations, such a5 drains, where material may have settled or accumulated. Locations of elevated direct radiation detected by scans were marked for further investigation. Scans were l performed using gas proportional, GM, and/or NaI detectors coupled to ratemeters or ratemeter-scalers with audible indicators.
Surface Activity Measurements j For each structural survey unit, ESSAP performed a minimum of thirty direct measurements for
{ total beta surface activity at randomly selected locations. ESSAP also performed additional direct measurements at locations of elevated direct radiation detected by surface scans. At measurement locations where the average NRC surface contamination guideline was exceeded, the size of the contaminated area and the average activity in the contiguous 1 m2 area was i
Shonham Nwlear Power Simuon - September 9,19W 6 L _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
determined. Alpha surface activity measurements were not required as the siected survey units were not classified as alpha affected and there was no alpha contamination identified by surface scans. Figures 7 through 14 show structural survey unit measurement locations. Measurements were performed using GM and/or gas proportional detectors coupled to ratemeter-scalers. A 1
smear sample for determining removable activity level was collected from each. direct I measurement location.
1 Exoosure Rate Measurements 1
Background exposure rate measurements were made at 10 locations within the Colt Building, which is of similar construction to the Turbine Building but did not have a history of radiological usage (Figure 15). Exposure rate measurements were performed at several direct measurement ]
locations within structural survey units (Figures 7 through 14). All exposure rates were measured at 1 m above surfaces using a pressurized ionization chamber (PIC).
Systems i,
For multicomponent systems, ESSAP requested up to 5 randomly selected access points be opened to each of the system survey units selected for confirmatory survey. Beta and gamma surface scans were then performed within the accessible portions of the system followed by direct measurements and smear samples. Scans and direct measurements were performed using gas proportional, GM, and/or NaI detectors coupled to ratemeters or ratemeter-scalers. The total number of direct measurements performed was dependent upon component size and accessibility and ranged from 3 to 30 measurements (Figures 16 through 24).
Embedded Pinine Confirmation of the radiological status of embedded piping was accomplished primarily through surveillance of the methodologies and review of the procedures utilized. In addition, independent measurements were made within drain openings throughout the Turbine Building (Figure 25).
1 Shoreham NWa Puw e Stauon - September 9,1994 7 j l
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l Comparative Measurements I LIPA was requested to perform direct measurements at 30 ESSAP direct measurement and exposure rate measurement locations. ESSAP instrumentation included GM and gas proportional
~
detectors coupled to ratemeter-scalers for direct measurements and a PIC for exposure rate measurements. The LIPA instrumentation used, which ESSAP selected randomly, included HP-260 and 126 cm2GM detectors coupled to ratemeter-scalers for direct measurements and a micro-rem meter for exposure rate measurements. The LIPA micro-rem readings were converted to R/h using the LIPA developed correlation factor of 3.06 + 1.07 ( rem).
- I SURVEY PROCEDURES: EXTERIOR I
The following procedures apply to exterior site grounds and building exteriors.
I Reference System
'I The grid systems established by LIPA on the site grounds and exterior building surfaces were used by ESSAP for reference.
Surface Scans Site grounds, paved areas, and building exteriors were scanned for gamma activity, while paved areas and exterior building surfaces were also scanned for alpha and beta activity. Scans were performed using NaI, gas proportional, and/or GM detectors coupled to ratemeters or ratemeter-scalers with audible indicators.
Surface Activity P.feasuremer11S ESSAP performed 30 direct measurements for total beta activity on both the paved portions of
,I SG003 and the surfaces of SE002 (Figures 26 through 28). Alpha direct measurements were not required. Direct measurements were made using gas proportional and/or GM detectors g
Shoreham Nuclear Power Station - September 9,1994 8 I
coupled to ratemeter-scalers. A smear sample for determining removable activity was collected from each direct measurement location.
Exposure Rate Measurements i
Background exposure rate measurements were performed at 6 locations within 0.5 to 10 km of the site (Figure 29). Exposure rate measurements were also performed at each soil sampling location within SG003 (Figure 28). Exposure rates were measured at 1 m using a PIC.
Soil Sampline Background soil samples were collected from 6 locations within 0.5 to 10 km of the SNPS (Figure 29). There were five soil samples collected from randomly selected locations within SG003 (Figure 28).
Confirmatory Analysis Four soil samples and 1 septic tank sludge sample, collected by LIPA, were obtained for I
confirmatory analysis.
SAMPLE ANALYSIS AND DATA INTERPRETATION Samples and data were returned to ESSAP's Oak Ridge laboratory for analysis and interpretation. Smears were analyzed for gross alpha and gross beta activity using a low background proportional counter. Soil and sludge samples were analyzed by solid state gamma i spectrometry. The spectra were reviewed for Co-60 as well as any other identifiable photopeaks. Soil and sludge sample results were reported in units of pCi/g. Smear and direct measurement data were converted to units of dpm/100 cm 2. Direct measurements which exceeded background levels were corrected for Fe-55 contribution, which can not be adequately detected with field instrumentation. A correction factor of 1.2 was therefore applied to those surface activity measurements that exceeded background distribution levels. LIPA developed, 1
Shoreham Nucleer Power Stanon - September 9,19W 9 l
. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ - _ _ _ - _ _ _ _ _ _ _ _ - --__L
1 and the NRC approved, the use of this correction factor based on the observed Co-60 to Fe-55 activity ratio identified in characterization samples.*J Exposure rates were reported in R/h.
The 95 % confidence level was calculated for surface activity and exposure rates for each survey unit selected for confirmation. A direct comparison of the ESSAP and LIPA survey unit results and individual soil sample results was performed. Additional information concerning major instrumentation, sampling equipment, and survey and analytical procedures is provided in Appendices A and B.
FINDINGS AND RESULTS DOCUMENT REVIEW ESSAP's review of the termination survey plan indicated that the document provided an adequate description of survey methodologies and general approaches. Comments were provided to the NRC in a January 12,1993 correspondence.' ESSAP's review of the termination survey final
{
report, and release records for those survey units selected for confirmatory survey, indicated that the survey plan had been appropriately followed with no significant deviations. Data were
{
appropriately converted, tested, and presented. Comments which were identified may be summarized as follows:
- Several direct measurements exceeded the average activity guideline. There was no explanation of additional data provided in the report that would demonstrate compliance at these locations with the 1 m2 average activity guideline. However, in the subsequent June 1994 issued Termination Survey Final Report, LIPA provided the 1 m 2average activity data for each of these locations.' A review of this data indicated that the average activity levels were less than the guideline.
- The report would benefit from the incorporation of maps which provide an overall (assembled) system view indicating components, rader than only maps of individual components.
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INTERIOR SURVEY UNITS The results of the confirmatory survey of the interior survey units are discussed below.
I I
Surface Scans !
I Alpha, beta, and gamma surface scans identified one small area of elevated direct beta radiation, measuring less than 15 cm2 in area, on the floor of the Main Condensate Storage Tank (SUO46) {
(Figure 22). A second, small area of elevated direct beta radiation was detected on the eleventh panel of the west wall, west outboard, B side of the Main Condenser (SUO25) (Figure 19). All other surface scans were comparable to background levels.
(
{ Surface Activity Levels
[ The results of total and removable surface activity levels are summarized in Table 1. Total beta activity levels for the structural survey units ranged from -800 to 2,100 dpm/100 cm2 ,
Removable activity levels ranged from -1 to 6 dpm/100 cm2 for alpha and -7 to 16 dpm/100 cm 2 2
for beta. Structural survey unit means ranged from -290 to 370 dpm/100 cm and -1 to )
2 l 1 dpm/100 cm for total and removable beta activity respectively.
Total beta activity levels in the surveyed systems ranged from -910 to 5,800 dpm/100 cm2 . The 2 2 removable activity levels were -1 to 4 dpm/100 cm for alpha and -7 to 16 dpm/100 cm for 2
beta. The mean beta activity levels for systems ranged from -340 to 300 dpm/100 cm for total activity and -1.0 to 1.8 dpm/100 cm2 for removable activity.
Exposure Rates Interior background exposure rates ranged from 4 to 5 R/h and averaged 5 pR/h. Individual gross exposure rates within the Turbine Building ranged from 3 to 7 R/h. The average gross exposure rates for all survey units ranged from 4 to 7 pR/h . Table 2 provides a summary of the interior exposure rates.
Shoreharn Nalear Power Statma - September 9.1994 !
I EXTERIOR SURVEY RESULTS I The following are the ~ sults of the confirmatory surveys of the exterior grounds and building exteriors. Surface Scans Surface scans of the exterior grounds, paved areas, and building exteriors did not identify any locations of elevated direct radiation. E Surface Activity Levels I Surface activity levels for the paved portions of SG003 and the building exteriors of SE002 are l summarized in Table 1. Total beta activity levels ranged from -760 to 1300 dpm/100 cm2 and removable activity ranged from -1 to 10 dpm/100 cm 2 for alpha and -7 to 7 dpm/100 cm 2 for l beta. The mean beta activities were 4 dpm/100 cm2 total and -1 dpm/100 cm2 removable for SG003 and -270 dpm/100 cm2 total and 0.8 dpm/100 cm2 removable for SE002. I Excosure Rates I Background exposure rates ranged from 6 to 9 pR/h and averaged 8 R/h. The gross exposure rates within SG003 ranged from 8 to 9 R/h and averaged 8 R/h. Table 3 provides a summary of exposure rate measurements. Radionuclide Concentrations in Soils Background soil sample concentration levels were less than 0.1 pCi/g of Co-60. The Co-60 concentration levels in the five samples collected from SG003 were less than or equal to 0.1 pCi/g. There were no other radionuclides identified, other than those occurring in nature. I I
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[ ESSAP AND LIPA DATA COMPARISON Comparative field measurements for surface activity and exposure rates were made in one survey unit using the various ESSAP and LIPA instrumentation. Most of the measurements collected from this unit were comparable to background; therefore, a meaningful statistical evaluation could not be performed. ESSAP intended to collect an additional 30 measurements from areas where positive residual activity, that is survey locations where the reported surface activity levels exceed both the background distribution and the minimum detectable activity of the instrumentation, was identified by LIPA. However, a sufficient quantity of these areas could not be located to provide a statistically significant number of measurement locations. The Co-60 concentrations determined by ESSAP in each of the samples obtained for confirmatory analysis were less than 0.1 pCi/g as were the results of the LIPA analysis. [ ( COMPARISON OF RESULTS WITII GUIDELINES ( The confirmatory survey results were compared with both the data provided by LIPA and the generic and site-specific 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. The applicable guidelines are those for beta-gamma emitters of which Co-60
)
and Fe-55 are the primary contaminants at SNPS. The residual surface activity guidelines are: Total Activity 5,000 dpm #-y/100 cm 2, averaged over 1 m2 15,000 dpm #-y/100 cm2 , maximum in 100 cm 2 i Removable Activity 1,000 dpm #-7/100 cm2 Shoseham Nuclear Power Stanon. September 9. Im 13
1 I . I As previously discussed, the detection sensitivities of the field instruments are such that the residual Fe-55 activity can not be detected. Therefore, surface activity measurements were corrected for Fe-55 when appropriate. The mean surface activity level for each survey unit was calculated and the survey unit data tested at the 95% confidence level, relative to the guidelines, in accordance with Draft NUREG/CR-5849. These results are provided in Table 4 and 5. I A comparison of the ESSAP mean activity levels to the LIPA mean activity level showed that in each instance the ESSAP mean was statistically less than or equal to the respective mean determined by LIPA for 16 out of the 19 confirmatory survey units. The ESSAP mean was greater than the LIPA mean for survey units TB017, TB089, and SUO54X03. For TB017 and SUO54X3, the activity levels were indistinguishable from background and the differences are therefore not considered significant. The differences in means observed for TB089 were the result of a higher background level associated with this room. LIPA established a separate l background,189 cpm versus the 129 cpm background used in data conversions for other survey units, for this room while ESSAP did not. The ESSAP decision was based on there being no indications of contamination (surface scans did not identify any locations of direct radiation and all measurements were less than MDA) and the effect of using a different background was inconsequential, relative to the overall status of the survey unit. Surface activity levels within each survey unit satisfied the guidelines at the 95% confidence level. There were no measurements which exceeded the maximum activity guideline. Two direct measurements 2 exceeded the 5,000 dpm/100 cm average guideline, both of which were identified within affected systems. An activity level of 5,100 dpm/100 cm2was detected a; one location within the Main Condenser (SUO25). The elevated activity was confined to an area of approximately 100 cm2 . The average activity within the contiguous 1 m2 area was 2,800 dpm/100 cm2 . The 2 second location measured 5,800 dpm/100 cm and was confined to an area ofless than 15 cm2 on the floor of the Main Condensate Transfer and Storage Tank (SUO46). The average activity in the surrounding 1 m2 area was 1,300 dpm/100 cm2 . All remaining total and removable activity levels were below guideline values. E I Shorcham Nwicar Power Stauon Sepmber 9,1994 14
l 1 Exposure rates were compared with those obtained by LIPA, and tested at the 95% confidence level, relative to the 5 R/h above background guideline currently being used by the NRC (Table 4)', The imerior and exterior exposure rates were both comparable to the respective background exposure rate level and confirmed the findings presented by LIPA. Soil concentrations in the samples collected by ESSAP and the LIPA samples, obtained for confirmatory analysis, were compared with the 8 pCi/g site-specific limit for SNPS produced radionuclides.'8 The Co-60 level in all samples was less than or equal to 0.1 pCi/g. SL MMARY ESSAP performed confirmatory activities of the Turbine Building, site grounds, and site exteriors at the Shoreham Nuclear Power Station in Brookhaven, New York. Confirmatory activities included document reviews, and during the period November 8 through 12, 1993, independent surface scans, surface activity measurements, exposure rate measurements, and soil sampling, were performed. The survey results confirm the results of the LIPA termination surveys. These findings indicate that surface activity levels, exposure rates, and soil concentration levels were below the NRC and site-specific guidelines for release to unrestricted use. Statistical tests of data sets further support the conclusion that each survey unit satisfies the guidelines at the 95 % confidence level. sama smi-, ro.o smu. . s.,iems ,9. io94 15 l,
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UPPER SURFACES AND EQUIPMENT f METERS g FIGbRE Bi Turbine Building, Vest Condenser Bay (TB017) - Measurement and Sonpling Locations Shoreham Nuclear Power $tauon - September 9.1994 23
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UPPER SURFACES AND EQUIPMENT (i METERS > FIGURE 11iMeasurement Turbine Building,d an Sonpling LocotionsChemistry Laborotory (TB060) - brehm Nuclear Power Scanon . September 9. IW 26 f
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1I- UPPER SURFACES AND EQUIPMENT METERS FIGURE 12: Turbine Building, Re-heater Areo, East (TB08D - Measurenent and Sonpling Locations l l Shonham Nudear Power Statmn . September 9.1994 27 l 5 I
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Measurement and Sanpling Locations Shoreham Nucker Power Sistmn . September 9.19M 28
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Background Measurenent Locations Shortham Nuclear Power Station September 9.1994 30
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Measurement and Sanpting Locations Shoreham Nuclear Power Stanon September 9, im 31
SHR74 (1) NORTH VALL 9 DN LEDGE A VEST VALLg g EAST VALL g FLOOR i O l 1 l i SOUTH VALL l l l MEASUREMENT / SAMPLING LOCATIONS S SURFACE ACTIVITY - NOT TO SCALE FLOOR AND VALLS A SURFACE ACTIVITY - TOP GRATE FIGURE 17: Turbine Building, Radweste Interior Drain Pipe Systen, Tank 12 (SU014X03) - Measurenent and Sonpting Locotions Shoreham Nuclear Power Stat on - September 9, im 32
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/ Measurenent and Sanpling Locottons shems.m wa,- ro.- smuon - seru mba 9. im 35
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Shoreham Nudear Power Satmn - Seriember 9,1994 38
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FIGURE 24: Secondary Access Facility, Ventitation Systen (SUO7D - Measurenent and Sampling Locations Shoreham Nuclear Power Stm6on September 9.1994 ' 39
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] FIGURE 27: Secured Area, Chlorine Pumphouse Exterior (SE002) -
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m m-p TABLE 1 h i SUh1 MARY OF SURFACE ACTIVITY LEVEIE [ TURBINE BUILDING, SITE GROUNDS, AND SITE EXTERIORS 7 SIIOREIIAM NUCLEAR POWER STATION
$ BROOKIIAVEN, NEW YORK 5 Number of Total Activity Range Removable Activity Range l I Location
- Measurement (dpm/100 cm 2) (dpm/100 cm2)
Locations Betab Alpha
- Beta d TB016 North Condenser Hallway 30 -570 to 440 -1 to 4 -5 to 4 TB017 West Condenser Bay 30 -680 to 290 -1 to 6 -6 to 5 TB031 Steam Seal Evaporator Room 30 -290 to 860 -1 to 4 -4 to 7 N TB035 Turbine Bldg. Truck Bay 30 -690 to 440 -1 to 4 -5 to 7 TB060 Chemistry Laboratory 30 -430 to 770 -1 to 4 -5 to 6 TB081 Re-heater Area-East 30 -800 to 2,100 -1 to 4 -7 to 16 TB082 Re-heater Area-West 30 -540 to 470 -1 to 4 -4 to 7 TB089 Black Battery Charger Room 30 -140 to 1400 -1 to 4 -6 to 7 SUOO5 Feedwater Control 5 -680 to 80 -1 to 1 -5 to 3 SUO14 Innuent Drain Piping 16 -730 to 040 -1 to 4 -5 to 5 SU014X03 Tank 12 6 -650 to -340 -1 to 4 -7 to 7 SUO24 Main Steam Components 25 -720 to 320 -1 to 4 -7 to 7 SUO25X02 Main Condenser TB-15 30 -910 to 5,100 . -1 to 4 -5 to 8 SUO32 Lube Oil Sump Tank 91 3 -570 to 230 -1 to 1 -4 to 4 SUO34 Valve 035C 3 -490 to -460 -1 to 1 -1 to 0 l
TABLE 1 (Continued) x
SUMMARY
OF SURFACE ACTIVITY LEVELS [ TURBINE BUILDING, SITE GROUNDS, AND SITE EXTERIORS y SIIOREI!AM NUCLEAR POWER STATION 8 BROOKIIAVEN, NEW YORK h Number of Total Activity Range Removable Activity Range
; Location
- Measurement (dpm/100 cm2) (dPm/100 cm2) g Locations Betab Alpha
- Beta d SUO46 Condensate Trans and Storage 30 -390 to 5,800 -1 to 4 -7 to 16 SUO54X03 Tank 186 10 -610 to 470 -1 to 4 -5 to 4 SUO71X70 Secondary Access Vents 30 -600 to 270 -1 to 4 -4 to 13 N SE002 General Bldg. "A" Exterior 30 -760 to 270 -1 to 10 -4 to 7 SG003 Secured Area West, Paved Area 30 76 to 1,300 -1 to 6 -7 to 7 Miscellaneous Turbine Building Vents 4 -400 to 240 -1 to 1 -4 to 3
" Refer to Figures 7 through 14 and 16 through 28. I b
MDAs = 990 to 1300 dpm/100 cm2 l
*MDA = 12 dpm/100 cm 2
MDA = 16 dpm/100 cm2 l
l l l l l~ i
TABLE 2 L INTERIOR EXPOSURE RATES ~ TURBINE BUILDING SIIOREIIAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK Exposure Rat ange p Location" h asar ment + Locations - Background L Colt Building 10 4 to 5 Turbine Building TB016 North Condenser Hallway 9 3 to 4 { TB017 West Condenser Bay 6 4 to 5 TB031 Steam Seal Evaporator Room 15 4 to 5 TB035 Truck Bay 24 4 to 6 TB060 Chemistry Laboratory 13 5 to 6 TB081 Re-heater Area-East 12 5 to 6 TB082 Re-heater Area-West 10 4 to 5 TB089 Black Battery Charger Room 12 6 to 7 [ " Refer to Figures 7 through 15. g Shoreham Nudcar Power Statma - September 9,1W 47
[ TABLE 3 L EXTERIOR EXPOSURE RATES [ AND L Co-60 CONCENTRATIONS IN SURFACE SOIL SAMPLES SIIOREIIAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK [ r Location
- Exposure Rate at 1 m ( R/h) Co-60 Concentration (pC1/g)
L
Background
Location #1 6 < 0.1 [ Location #2 7 < 0.1 _ ( Location #3 9 < 0.1 Location #4 7 < 0.I r - L Location #5 8 < 0.1 Imcation #6 9 < 0.1 Secured Area West (SG003) Location #1 9 < 0.1 L Location #2 8 < 0.1 Location #3 8 0.1 i 0.l b [ Location #4 9 < 0.1 Location #5 ( _ 8 < 0.1
" Refer to Figure 28 and 29.
( b Uncertainties represent the 95% confidence level, based only on counting statistics. ( I l Shoreham Nudear Power Sinu <m - September 9,1994 48
TABLE 4 CONFIRMATORY RADIOLOGICAL STATUS
SUMMARY
-STRUCTURES, l SITE GROUNDS, AND SITE EXTERIORS , SIIOREIIAM NUCLEAR POWER STATION ! BROOKIIAVEN, NY I Radiological TB016 TB017 Survey Unit' TB031 TB035 TB060 Total Beta Activity (dpm/100 cm 2)
# of Direct Measurements 30 30 30 30 30 Mean(X) -62 -130 7 -110 200 LIPA X -63 -480 -16 24 190 F-g 30 -52 77 -27 300 Conditions and 5,000/15,0000 dpm/100 cm2 Guidelines Satisfied Yes Yes Yes Yes Yes 2
Removable Beta Activity (dpm/100 cm)b
# of Smears 30 30 30 30 30 Mean(X) -1.2 0.1 0.1 -0.4 0.1 LIPA X 1.9 1.6 2.9 3.7 1.7 Fa -0.5 1.0 1.1 0.6 1.0 Conditions and 1,000 dpm/100 cm2 Guideline Satisfied Yes Yes Yes Yes Yes l
Exposure Rates at 1 m (pR/h)
# of Exposure Rate Measurements 9 6 15 23 13 Net Mean (X) -0.7 -0.5 -0.1 0.4 0.9 LIPA X -0.1 0.2 0.2 0.1 0.7 l
Fa -0.5 -0.2 0.2 0.6 1.1 I Conditions and 5 pR/h Above Background Guideline Satisfied Yes Yes Yes Yes Yes I LI I s%,. s-w< r-u se. . seriember 9. im 49
TABLE 4 (Continued) CONFIRMATORY RADIOLOGICAL STATUS SUMMA.RY-STEUCTURES, SITE GROUNDS, AND SITE EXTERIORS SIIOREIIAM NUCLEAR POWER STATION BROOKIIAVEN, NY l Radiological Survey Unit *
""**U TB081 TB082 TB089 SE002 SG003 l Total Beta Activity (dpm/100 cnf)
# of Direct Measurements 30 30 30 30 30 l Mcan(X) -290 -84 370 -270 4 LIPA X 450 65 -280 430 440 I Fa -110 -2 480 -180 87 Conditions and 5,000/15,0000 dpm/100 cnf Guidelines Satisfied Yes Yes Yes Yes Yes l
Removable Beta Activity (dpm/100 cuf)b
# of Smears 30 30 30 30 30 Mean(X) 0.6 -0.2 -1.4 0.8 -1.0 l LIPA X 3.5 4.1 0.7 2.1 2.6 Fa 1.9 0.6 - 0.4 1.8 -0.1 Conditions and 1,000 dpm/100 cm2 Guideline Satisfied Yes Yes Yes Yes Yes ,
Exposure Rates at 1 m (gR/h)
# of Exposure Rate Measurements 12 10 12 Net Mean (Y) 0.5 -0.3 2.3 -- -
LIPA X -0.1 -0.5 1.0 - -- Fa 0.7 -0.1 2.4 -- --- Conditions and 5 R/h Above Background Guideline Satisfied Yes Yes Yes -- ---
" Refer to Figures 7 through 14 and 24 through 27.
b All alpha removable activity was less than 12 dpm/100 cm2,
--- = Measurements not performed.
E Shortham Nucicar Power Stahon - September 9,1994 50 J
i-I TABLE 5 CONFIRh1ATORY RADIOLOGICAL STATUS SUALTIARY-SYSTEMS I- TURBINE BUILDING - SIIOREIIAM NUCLEAR POWER STATION BROOKIIAVEN, NY I Survey Unit' Radiological Summary SU005 SUO14-Drains SU014X03 SUO24 SU025X2 Total Beta Activity (dpm/IO0 can
# of Direct Measurements 5 16 6 25 30 MeanCD -340 -340 -490 -240 52 LIPA I , , , ,
380 -NA* -210 350 -120
#= -57 -230 -370 -170 480 Conditions and 5,000/15,000 dprn/100 cm' Guidelines Satisfied Yes Yes Yes Yes Yes Removable Beta Activity (dpm/100 cndh
# of Smears 5 16 6 25 30 l MeanCD LIPA X
- 0.1 2.1
-1.1 NA
-1.0 6.2
-0.8 6.2 0.3 7.2 Ea 3.t 0.3 0.1 0.2 1.4 Conditions and 1,000 dpm/100 cm2 Guideline Satisfied Yes Yes Yes Yes Yes I
I I I I 1 I g , _ _ _ . , _ , , _ _ . , , _ . . _ 51
[ TABLE 5 (Continued) CONFIRMATORY RADIOLOGICAL STATUS
SUMMARY
-SYSTEMS TURBINE BUILDING SIIOREIIAM NUCLEAR POWER STATION BROOKHAVEN, NY (! Radiological
""**U Survey Unit' SUO32 SUO34 SUO46 SUO54X3 SUO71X70 Total Beta Activity (dpm/100 cn/)
# of Direct Measurements 3 3 34 10 30
-410 270 300 -220 -90 Mean(X)
LIPA X 290 300 560 -380 15 F= l
-120 -390 630 7 -29 J
5,000/15,000 dpm/100 cnf Guidelines Satisfied Yes Yes Yes Yes Yes
' Removable Beta Activity (dpm/100 cnf)
# of Smears 3 3 30 10 30 Mean(X) -0.7 -0.5 1.2 -0.4 0.3 LIPA X 2.9 1.9 13.5 4.9 1.0 F= 1.5 0.7 2.7 1.3 2.2 Conditions /1,000 dpm/100 cnf Guideline Satisfied Yes Yes Yes Yes Yes
- Refer to Figure 16 thmugh 25.
b All alpha removable activity was less than 12 dpm/100 cm 2,
*NA = not applicable F
1 sm %, r . s.uo. . s,iemi.,,. im 52 s
I-I - cEs
- 1. Long Island Lighting Company, "Shoreham Nuclear Power Station Site Characterization Program Final Report," May 1990.
I 2. T. J. Vitkus, ORISE, " Confirmatory Survey of the Turbine Internal Components, Shoreham Nuclear Power Station, Brookhaven, New York," July 1993.
- 3. J. D. Berger, Oak Ridge Associated Universities, Draft " Manual for Conducting Radiological Surveys in Support of License Termmation," NUREG/CR-5849, June 1992.
- 4. Long Island Power Authority, "Shoreham Decommissioning Project, Termination Survey Plan, Revision 1,' Apdl,1993.
- 5. Long Island Power Authodty, "Shoreham Decommissioning Project Termination Survey l Final Report, Volumes 1 through 5," September,1993.
- 6. Ietter from T. J. Vitkus, ORISE to D. Fauver, U.S. Nuclear Regulatory Commission, " Final Confirmatory Survey Plan for the Shoreham Nuclear Power Station, Brookhaven, New Yark
- Docket File No. 50-322," November 4,1993.
- 7. Letter from D. N. Fauver, U.S. Nuclear Regulatory Commission, to T. Vitkus, ORISE, July 1,1993.
- 8. Ietter from M. R. Landis, ORISE to D. Fauver, U.S. Nuclear Regulatory Commission, I "Shoreham Decommissioning Project, Termination Survey Plan, Revision O, Shoreham Nuclear Power Station, October 1992," January 12, 1993.
I 9. Long Island Power Authority, "Shoreham Decommissioning Project Termination Survey Final Report," Volume 1, June 1994. I 10. U.S. Nuclear Regulatory Commission, " Guidance and Discussion of Requirements for an Application to Terminate a Non-Power Reactor Facility Operating License," Revision 1, September 1984. I I I
,I I , , _ , _ . . _ . . _ 33
L [ [ r- ' APPENDLX A MAJOR INSTRUMENTATION l i [ [ [ . E [ [ [ l I Stwnham Nuclear Power Staten - September 9,19W i
I-I APPENDIX A MAJOR INSTRUMENTATION l 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. I DIRECT RADIATION MEASUREN ENT Instniments I Eberline Pulse Ratemeter Model PRM-6 (Eberline, Santa Fe, NM) Eberline " Rascal" Ratemeter-Scaler Model PRS-1 (Eberline, Santa Fe, NM) Ludlum Ratemeter-Scaler I Model 2221 (Ludlum Measurements, Inc., Sweetwater, TX) Detectors I 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 j (Ludlum Measurements, Inc., i l Sweetwater, TX) I i Shoreham Nuclear Power Statum September 9, im A-1
E' L I
- l. Ludlum Gas Proportional Detector Model 43-68 Effective Area,100 cm2 l
(Ludlum Measurements, Inc., Sweetwater, TX) I l Reuter-Stokes Pressurized Ion Chamber Model RSS-111 (Reuter-Stokes, Cleveland, OH) i Victoreen NaI Scintillation Detector Model 489-55 3.2 cm x 3.8 cm Crystal l (Victoreen, Cleveland, OH) l LABORATORY ANALYTICALINSTRUMENTATION i h High Purity Extended Range Intrinsic Detectors W Model No: ERVDS30-25195 l (Tennelec, Oak Ridge, TN) Used in conjunction with: i Lead Shield Model G-11 l (Nuclear Lead, Oak Ridge, TN) and I I Multichannel Analyzer 3100 Vax Workstation (Canberra, Meriden, CT) I High-Purity Germanium Detector Model GMX-23195-S,23% Eff. (EG&G ORTEC, Oak Ridge, TN) Used in conjunction with: t lead Shield Model G-16 (Gamma Products, Palos Hills, IL) and Multichannel Analyzer 3100 Vax Workstation (Canberra, Meriden, CT) I Low Background Gas Proportional Counter I Model LB-5100-W (Oxford, Oak Ridge, TN) l L $horchwn Nuckar Power Stat.m - September 9. Im A-2 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ___ _ _ _ . . - - - --- -- J
[ l I' I I I I I APPENDIX B SURVEY AND ANALYTICAL PROCEDURES I I I I I I 'I , i e 4 4 I g
y APPFEDIX B SURVEY AND ANALYTICAL PROCEDURES SURVEY PROCEDURES I Surface Scans Surface scans were performed by passing the probes slowly over the surface; the distance between the probe and the surface was maintained at a minimum - nominally about I cm. A large surface area, gas proportional Door monitor was used to scan the floors of the smveyed areas. Other 2 surfaces were scanned using small area (15.5 cm , 59 cm2 or 100 cm )2 hand-held detectors. Identification of elevated levels was based on increases in the audible signal from the recording and/or indicating instrument. Combinations of detectors and instruments used for the scans were: Alpha - gas proportional detector with ratemeter-scaler ZnS scintillation detector with ratemeter-scaler l Beta - gas proportional detector with ratemeter-scaler
'- pancake GM detector with ratemeter-scaler Gamma -
NaI scintillation detector with ratemeter Surface Activity Measurements Measurements of total beta activity levels were primarily performed using GM detectors with portable ratemeter-scalers. Count rates (cpm), which were integrated over 1 minute in a static position, were converted to activity levels (dpm/100 cm2 ) by dividing the net rate by the 4 x efficiency and correcting for the Shoreham Nuctear Power Stauon Septemtwr 9,19M B-1 l
r I active area of the detector. The beta activity background count rates for the GM detectors ranged from 22 to 42 cpm. Beta efficiency factors ranged from 0.16 to 0.18 for the GM detectors. The effective window for the GM detectors was 15.5 cm2, Surface activity measurements which exceeded the normal background distribution were corrected for the Fe-55 contribution by multiplying the dpm/100 cnf field activity level by a factor of 1.2. The instrument response level at which the detector output could be considered above background was defined as the critical level A). This level was defined for each detector / instrument combination as follows:
**E I' '*""' '" Background count rate 1.96 h ample count time Background count time L* = (Detector Efficiency) (Detector Geometry)
Removable Activity Measurements Removable activity levels were determined using numbered filter paper disks,47 mm in diameter. 2 Moderate pressure was applied to the smear and approximately 100 cm of the surface was wiped. Smears were placed in labeled envelopes with the location and other pertinent information recorded. Exoosure Rate Measurements Measurements of gamma exposure rates were performed using a pressurized ionization chamber (PIC). Soil Samoline Approximately 1 kg of soil was collected at each sample location. Collected samples were placed in a plastic bag, sealed, and labeled in accordance with ESSAP survey procedures. I Shorcham Nwlear Power Stauou - September 9,1994 B-2
ANALYTICAL PROCEDURES Removable Activity { Smears were counted on a low background gas proportional system for gross alpha, and gross beta activity. Gamma Spectrometry Soil and sludge samples were dri:d, mixed, crushed, and/or homogenized as necessary, and a portion scaled in 0.5-liter Marinelli beaker or other appropriate container. The quantity placed in the beaker [ was chosen to reproduce the calibrated counting geometry and ranged from 556 to 1238 g of ( material. Net material weights were determined and the samples counted using intrinsic germanium cetectors 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 detennining the activity of ( the radionuclide of concern was: f Co-60 1.173 MeV Spectra were also reviewed for other identifiable photopeaks. UNCERTAINTIES AND DETECTION . LIMITS The uncenainties associated with the analytical data presented in the tables of this report represent the 95 % confidence level for that data. These uncertainties were calculated based on both the gross 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 report. Shoreham Nwlear Power Sinton Seriember 9,19w B-3
r-l Detection limits, referred to as minimum detectable activity (MDA), were based on 2.71 plus 4.66 times the standard deviation of the background count [2.71 + (4.66/BKG)]. When the activity was I 1 determined to be less than the MDA of the measurement procedure, the result was reported as less than MDA. Because of variations in background levels, measurement efficiencies, and contributions from other radionuclides in samples, the detection limits differ from sample to sample and instrument ; to instrurnent. CALIBRATION AND QUALITY ASSURANCE Analytical and field survey activities were conducted in accordance with procedures from the l following documents of the Environmental Survey and Site Assessment Program: l
- Survey Procedures Manual, Revision 7 (May 1992)
Laboratory Procedures Manual, Revision 8 (August 1993)
- Quality Assurance Manual, Revision 6 (July 1993)
The procedures contained in these manuals were developed to meet the requirements of DOE Order 5700.6C and ASME NQA-1 for Quality Assurance and contain measures to assess processes during their performance. Calibration of all field and laboratory instrumentation was based on standards, traceable to 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 I by the manufacturer. Quality control procedures include: l
- Daily instrument background and check-source measurements to confirm that equipment operation is within acceptable statistical fluctuations.
I
- Participation in EPA and DOE /EML Quality Assurance Progams.
- Training and certification of all individuals performing procedures.
- Periodic internal and extemal audits.
B-4 I Shoreham No. ness Power Staten .. September 9,1994 1
r 1 I I ! l I I I APPENDIX C I REGULATORY GUIDE 1.86, TERMINATION OF OPERATING I LICENSES FOR NUCLEAR REACTORS I I I , I , sh.nham Nmkor Po.er sms,. . seriember 9,1994 f
e I- U.S ATOMIC ENERGY COMMISSION Juna 1974 I REGULATORY GUIDE DIRECTORATE OF REGULATORY STANDARDS I REGULATORY GUIDE 1.86 TERMINATION OF OPERATING LICENSES FOR NUCLEAR REACTORS A. INTRODUCTION important to the safety of reactor opemtion is no longer required. Once this possession +nly license is issued, I Section 50.51, " Duration of license, renewal," of 10 CFR Part 50, " Licensing of Production and Utilization Facilities," requires that each license to operate a reactor operation is not permitted. Other activities from the reactor and placing it m storage (either onsite or offsite) may be continued. production and utilization facility be issued for a specified duration. Upon expiration of the specified period, the A licensee having a possession-only license must retain, license may be either renewed or terminated by the with the Part 50 license, authorization 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 material (10 CFR Part 30, " Rules of General satisfied to termmate an operating license, including the Applicability to Licensing of Byproduct Material"), and requirement that the dismantlement of the facility and source matenal (10 CFR Part 40, " Licensing 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 the health and safety sources are removed from the facility. Appropriate of the public. His guide describes methods and administrative controls and facility requirements are procedures considered acceptable by the Regulatory staff imposed by the Part 50 license % the twhnical for the termination of operating licenses for nuclear specifications to assure that prop iuveillance is I reactors. He advisory Committee on Reactor Safeguards has been consulted conceming this guide and has performed and that the reactor facility .4 maintained in a 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, i entombment, or dismantling. He requirements imposed , 1 When a licensee decides to terminate his nuclear depend on the option selected. 1 reactor operating license, he may, as a first step in the l process, request that his operating license be amended to Section 50.82 provides that the licensee may dismantle , restrict him to possess but t,ot operate the facility. He and dispose of the component parts of a nuclear reactor in l 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 I requirements in that periodic surveillance of equipment disassembly of a reactor and its shipment organization for
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1- further use. The site from which a reactor has been c. Any proposed changes to the technical specifications I removed must be deconamimt~!, as necessary, and inspectal by the Commission to determine whether unrestricted access can be approved. In the case of that reflect the possession-only facility status and the necessary disassembly / retirement activities to be performed.
)
I nuclear power reactors, dismantling has usually been accomplished by shipping fuel offsite, making the reactor moperable, and disposing of some of the radioactive
- d. A safety analysis of both the activities to be accomplished and the proposed changes to the technical l
components. specifications. Radioactive components may be either shipped off-site e. An inventory of activated materials and their for burial at an authonzed burial ground or secured on the location in the facility. site. "Itose radioactive materials remaining on the site must be isolated from the public by physical barriers or 2. ALTERNATIVES FOR REACTORRETIREMENT 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 are considered acceptable by the Regulatory required depends upon (1) the potential hazard to the staff. These are: health and safety of the public from radioactive material remaining 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 radioactivity 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 retumed facility is surveillance, and appropriate security procedures evaluated by considering the amount and type of renwmng should be established under a posseuion only license contamination, the degree of confinement of the remammg to ensure that the health and safety of the public is not I radioactive matenals, the physical security provided by the endangered. confmement, the susceptibility to release of radiation as a result of natural phenomena, and the duration of required 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 internals) within a structure integral with I 1. APPLICATION FOR A LICENSE TO POSSESS BUT NOT OPERATE (POSSESSION-ONLY the biological shield after having all fuel assemblies, radioactive fluids and wastes, and certain selected components shipped offsite. 'Ite structure should I LICENSE) A request to amend an operating license to a provide integrity over the period of time in which significant quantities (greater than Table I levels) of radioactivity remain with the material in the possession-only license should be made to the Director of entombment. An appropriate and continuing Licensing, U.S. Atomic Energy Commission, Washington, surveillance program should be established under a D.C. 20545. 'Ile regt'est should include the following possession only license. information:
- c. Removal of Radioactive. Components and
- a. A description 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 I prevent criticality or reactivity changes and to minimize removed from the site. The facility owner may then releases of radioactivity from the facility. have unrestricted use of the site with no requirement j for a license. If the facility owner so desires, the Note: Section electronicatty reproduced from photocopy. C-2 i
r^ remamder of the reactor facility may be dismantled and barriers in the facility. Sampling should be done along the I all vestiges mmoved and disposed of.
- d. Conversion to a New Nuclear System or a Fossil most probable path by which radioactive matenal such as that stored in the inner contamment regions could be transported to the' outer regions of the facility and I Fuel System. This attemative, which applies only to nuclear power plants, utdizes the existing turbine system with a new steam supply system. The original ultimately to the environs.
- d. An envimamental radiation survey should be I nuclear steam supply system should be separated from the electric generatmg system and disposed of in accordance with one of the previous three retirement performed at least semiannually to verify that no significant amounts of radiation have been released to the environment from the facility. Samples such as soil, altematives. vegetation, and water should be taken at locations for which statistical data has been established during reactor
- 3. SURVEILLANCE AND SECURITY FOR TIIE operations.
RETIREMENT ALTERNATIVES WilOSE FINAL STATUS REQUIRES A POSSESSION-ONLY e. A site representative should be designated to be LICENSE responsible for controlling authorizal access into and A facility which has been licensed under a possession-only license may contain a significant amount f. Admmistrative procedures should be established for of radioactivity in the form of activated and contWne~1 the notification and reporting of abnormal occurrences hardware and structural materials. Surveillance and such as (1) the entrance of an unauthorized person or commensurate secunty should be provided 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 inadvertent exposure of offsite environment.
personnel should be provided by multiple locked barriers. The presence of these barriers should make it extremely g. ' Die following reports should be made: difficult for an unauthorized person to gain access to areas where radiation or contammation levels exceed those (1) An annual report to the Director of Licensing, I specified in Regulatory Position C.4. To prevent inadvertent exposure, radiation areas above 5 mR/hr, such U.S. Atomic Energy Commission, Washington, D.C. 20545, describing the results of the environmental and as near the activated primary system of a power plant, facility radiation smveys, the status of the facility, and an I should be appropriately marked and should not be accessible except by cutting of welded closures or the evaluation of the performance of security and surveillance measures. disassembly and removal of substantial structures and/or shielding material. Means such as a remote-readout (2) An abnormal occurrence report to the Regulatory intrusion alarm system should be provided to indicate to Operations Regiuzml Office by telephone within 24 hours I designated personnel when a physical barrier is penetrated. Security personnel that provide access control to the facility may be umd instead of the physical barriers and of discovery of an abnormal occurrence. The abnormal occurrence will also be reported in the annual report described in the precedmg item. the intrusion alarm systems.
- h. Records or logs relative to the following items i
- b. The physical barriers to unauthorized entrance into should be kept and retained until the license is terminated, !
the facility, e.g., fences, buildings, welded doors, and after which they must be stored with other plant records: access openings, should be inspected at least quarterly to l I assure that these barriers have not deteriorated and that (1) Environmental surveys, locks and locking apparatus are intact. (2) Facility radiation surveys, l Q) Inspections of the physical barriers, and
- c. A facility radiation survey should be performed at )
least quarterly to verify that no radioactive material is (4) Abnormal occurrences. escaping or being tansported through the contWament Note: Section electronically reproduced from photocopy. C-3
. i
- 4. DECONTAMINATION FOR RFIFASE FOR (2) A detailed health and safety analysis indicating that UNRESTRICTED USE the residual amounts of materials on surface areas, together with other considerations such as the prospective {
{ If it is desirol to terminate a license and to eliminate any further su:veillance requirements, the facility should use of the premises, equipment, or scrap, are unlikely to result in an unreasonable risk to the health and safety of be sufficiently dentamim'~i to prevent risk to the public the public. ( health and safety. After the d-tamietion 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 termmated and the facility abandoned or released for establishing that contammation is within the limits specified unrestricted use. He licensee should perform the in Table 1. A survey report should be filed with the decontamination using the following guidelines: Director of Licensing, U.S. Atomic Energy Commission, Washington, D.C. 20545, with a copy to the Director of
- a. The licensee 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. He smvey
- b. No covering should be applied to adioactive 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 (determined by a survey and documented) are below the limits spaified 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 contammation to as low as practicable minimize contamination prior to any such covering. levels;
- c. He radioactivity of the interior surfaces of pipes, (3) Describe the scope of the survey and the general drain lines, or ductwork should be determined by making procedures followed; and measurements at all traps and other appropriate access
{ points, provided contammation at these locations is likely (4) State the finding of the survey in units specifial in
- to be representative of 2 contamination on the interior of Table 1.
the pipes, drain lines, or ductwork. Surfaces of premises, aluipment, or scrap which are likely to be contnmim'~I After review of the report, the Commission may j but are of such size, construction, or location as to make inspect the facilities to confirm the survey prior to granting
)
the surface maccessible 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 alternatives are acceptable for reactor facility retirement. excess of the limitr. specified. This may include, but is not If minor disassembly or "mothballing" is planned, this linuted to, special circumstances such as the transfer of could be done by the existing operating and maintenance premises to another licensed organization that will continue procedures 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 9 50.59. premises, equipment, scrap, and radioactive contammants and the nature, extent, and degree of residual surface If major structural changes to radioactive components 1 contamination. of the facility are planned, such as removal of the pressure vessel or major components of the primary system, a Note: Section electronica9y reproduced from photocopy. C-4 4 1 __ _______________________________.___._______._.________.___.____._____________.___._____________.___________________U
C l dismantlement plan including the informatico requuwl by j I i 50.82 should be submitted to the Commmion. A dismantlement plan should be submitted for all the altematives of Regulatory Position C.2 except mothballing. However, minor disassembly activities may still be ) performed in the absence of such a plan, provided they are permitted by existing ogwating and maintenance I 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 precautions to be taken.
- c. A safety analysis of the dismantling activities including any effluents which may be released.
- d. A safety analysis of the facility in its ultimate status.
Upon satisfactory review and approval of the dismantling plan, a dismantling order is issued by the Commission in accordance with i 50.82. When dismantling is completed and the Comnussion has been notified by letter, the appropriate Regulatory Operations Regional Office inspects the facility and verifies completion in accordance with the dismantlement plan. If residual radiation levels do not exceed the values in Table 1, the Commission may terminate the license. If possession <mly 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. I I I I , Note: Section electronically reprodoced from photocopy. C-5
g. I TABLE 1 ACCEI'rABLE SURFACE CONTAhUNATION LEVEIS Nuclide' Average
- Maximum" Removable
- U-nat, U-235 U-238, and associated decay products 5,000 dpm a/100 enf 15,000 dpm a/100 cd 1,000 dpm a/100 cd Transuranics, Ra-226, Ra-228,
'11-230, Th-228, Pa-231, Ac-227,1-125, I-129 100 dpm/100 cd 300 dpm/100 cd 20 dpm/100 cnf Th-nat, Th-232, Sr-90, Ra-223, Ra-224, U-232,1-126, I-131,1-133 1,000 dpm/100 cd 3,000 dpm/100 cul 200 dpm/100 cnf '
Beta-gamma emitters (nuclides with decay modes other than alpha emissien or spontaneous fission) except Sr-90 and others I noted above. 5,000 dpm Sy/100 cnf 15,000 dpm Sy/100 cnf 1,000 dpm Sy/100 enf I
'Where surface contamination by both alpha- and beta-gamma-emitting nuclides exists, the limits established for alpha- and beta-gamma-emitting nuclides should apply independently. "As used in this table, dpm (disintegrations per minute) mams the rate of emission by radioactive material as determined by I correctmg the counts per minute observed by an appropriate detector for background, efficiency, and geometric factors associated with the instrumentation. ' Measurements of average contammant should not be averaged over more than 1 square meter. For objects ofless surface area.
I the average should be derived fo each such object.
'The maximum contammation level applies to an area of not more than 100 cnf. 'Ihe amount of renovable radioactive material per 100 cm 2of 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 instrument of known efficiency. When rerrvable contammation on objects of less surface area is determined, the pertinent levels should be reduced proportionally and the entire surface should be wiped.
I I i l I Note: Section electronically reproduced from photocopy. C-6 I
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