ML20081A693
| ML20081A693 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 02/17/1995 |
| From: | Laudeman M, Payne A, Vitkus T OAK RIDGE ASSOCIATED UNIVERSITIES |
| To: | NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| Shared Package | |
| ML20081A687 | List: |
| References | |
| CON-FIN-A-9076 NUDOCS 9503150252 | |
| Download: ML20081A693 (67) | |
Text
CONFIRMATORY SURVEY' OF THE-RADWASTE BUILDING,:
SUPPRESSION PO'OL3 PHASE 2fANDl PHASE 3' SYSTEMS; SHOREHAM NUCLEAR POWERLSTATIONL
-BROOKHAVEN, NEW YORK:
[ DOCKET No. 50-322]-
J T.J. VITKUS Prepared for the Divison of Waste Management Headquarters Office U.S. Nuclear Regulatory Commision
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I The Oak Ridge Institute for Science and Education (ORISE) was established by the U.S. Department of Energy to undertake national and intemational 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 Universities (ORAU) through a management and operating contract with the U.S. Department of Energy. Established in 1946, ORAU is a consortium of 88 colleges and universities. ,
f NOTICES t 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 Government nor the U.S. Department of Energy, nor any of their employees, makes any warranty, expressed or implied, 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. Reference herein to any specific commercial product, process, or service by trade name, mark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement or recommendation, or favor .
by the U.S. Govemme;tt 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.
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i CONFIRMATORY SURVEY OF THE RADWASTE BUILDING, SUPPRESSION POOL,
'; PHASE 2, AND PHASE 3 SYSTEMS !
SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK i
Prepared by.
T. J. Vitkus ,
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Environmental Survey and Site Assessment Program '
Energy / Environment Systems Division
, Oak Ridge Institute for Science and Education !
Oak Ridge, Tennessee 37831-0117 Prepared for the U.S. Nuclear Regulatory Commission Headquarters Office 't Sponsored by the -
Division of Waste Management ;
FINAL REPORT ,
FEBRUARY 1995 i i
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.
t Shortham-BroaLhaven, NY - February 16,1995
I i
I CONFIRMATORY SURVEY OF THE RADWASTE BUILDING, SUPPRESSION POOL, AND PHASE 2 SYSTEMS SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK v t
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e Prepared by:
T. J. Vitkus,(frpject Leader Date: d!/S Y Environmental Survey and Site Assessment Program l Reviewed by: .u"kk@E
- Date: KlMf W !
M. L Laudeman, Radiochemistry Laboratory Supervisor j '
Environmental Survey and Site Assessment Program f)
Reviewed by: [h h we.
t A. T. Payne, Quality Assuranice Officer Date: _/b/97 Q
Environmental Survey and Site Assessment Program Reviewed by: _ O Date: 4 //7/95 W. L Beck, Acting Program Director Environmental Survey and Site Assessment Program ShBrathaven. h f . February 16,1995
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ACKNOWLEDGEMENTS {
The author would like to acknowledge the significant contributions of the following staff members:
FIELD STAFF E. H. Bjelland .
E. H. Montalvo J. R. Morton L. Payne -l J.
LABORATORY STAFF :
R. D. Condra J. S. Cox i M. J. Laudeman ,
CLERICAL STAFF D. A. Adams R. D. Ellis ,
K. E. Waters ILLUSTRATOR ~ ;
T. D. Herrera f
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Shortham-Bmthaven. NY . February 16,1995 1
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- o. . , a TABLE OF CONTENTS j l
PAGE 1 m
6 i i: List of Figures . . . . . . . . . . . . . .. .................................ii' '
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=t List of Tables .... ... . ...... . . . ....... . ..... . .. ........ . .... . ivi !
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. Abbreviations and Acronyms . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v - i e b '1ntroduction and Site History . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I-
_ . Site Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 i i
' - - Obj ectives . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 '
t Document Review and LIPA Procedure Surveillance . . . . . . . . . . . . . . . . . . . . . . . 4 1 i
Procedures ................................. ............... 4 .f i
Findings'and Results ........................................... 10 _!.:
j Comparison of Results with Guidelines . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12' ]
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. S u m mary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 l r
References ................................................49 1
Appendices: !
1 Appendix A: Major Instrumentation i
'i Appendix B: Survey and Analytical Procedures 1
Appendix C: Regulatory Guide 1.86 Termination of Operating Licenses l for Nuclear Reactors !
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LIST OF FIGURES PAGE FIGURE 1: Location of the Shoreham Nuclear Power Station . . . . . . . . . . . . . . . 17 FIGURE 2: Plot Plan of the Shoreham Nuclear Power Station . . . . . . . . . . . . . . . 18 FIGURE 3: Radwaste Building, Elevations 15'6"/19'6" Floor Plan-Surveyed Areas ...............................19 FIGURE 4: Radwaste Building, Elevation 37'6" Floor Plan-Surveyed Areas ...................................20 ,
FIGURE 5: Radwaste Building, Elevation 50'6"/52'6" Floor Plan- -
Surveyed Areas ...................................21 FIGURE 6: Reactor Building, Elevation 8'0" Ficor Plan, Suppression Pool-Surveyed Areas ...............................22 FIGURE 7: Radwaste Building, Radwaste 15' North Hallway (RW013)-
Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . 23 [
t FIGURE 8: Radwaste Building, Cation / Anion Regen. and Resin Storage Tanks Area (RW017)-Measurement and Sampling Locations . . . . . . . 24 FIGURE 9: Radwaste Building, Liner Fill Stations /BW Storage Rooms (Cubicle A) (RWO23)-Measurement and Sampling Locations . . . . . . . 25 FIGURE 10: Radwaste Building, Waste Evap./ Regen.Evap. Distil. Room (RWO40)-Measurement and Sampling Locations . . . . . . . . . . . . . . . 25 FIGURE 11: Radwaste Building, Radwaste 37' North Hallway (RWO42)-
Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . 27 FIGURE 12: Radwaste Building, "B" RW O/G HEPA After Filter Area (RWO69A)-Measurement and Sampling Locations . . . . . . . . . . . . . . 28 FIGURE 13: Radwaste Building, "C&D" Dryer Skid Area (RWO69B)-
Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . 29 FIGURE 14: Radwaste Building, Off Gas Desiccant Dryers Area (RWO72)-
Measurement and Sampling Locations . . . . . . . . . . . . . . . . . . . . . . 30 FIGURE 15: Reactor Building, Suppression Pool, NW Quadrant (SP004)
Lower Walls and Floor-Measurement and Sampling Locations . . . . . . 31 Sheham Broalhaven. NY . February 16,1995 ii
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LIST OF FIGURES (Continued)
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[ FIGURE 16: Reactor Building, Suppression Pool, EW Cuadrant (SP004) Upper Walls-Measurement and Sampling Lx:atians . . . . . . . . . . . . . . . . . 32 FIGURE 17: Reactor Building, Suppression Pool, Area Inside Vessel Pedestal (SP005)-Measurement and Sampling Locations . . . . . . . . . . . . . . . . 33 FIGURE 18: Reactor Building, HPCI Valves E41-01V 3049 and 3050 (SUO12)-Measurement and Sampling Locations ...............34 FIGURE 19: Radwaste Building, Radwaste Influent Drain Sy.; tem, Drain Sump Tank-054 (SU14 x09)-Measurement and Sampling Locations ..............................35 FIGURE 20: Radwaste Building, Equipment Drain System, Drain Sump Tank-071 (SUO14 x 10)-Measurement and Sampling locations ..............................36 FIGURE 21: Radwaste Building, Radwaste Equipment / Components, Flat Bed Floor Drain Filter 1G11-FL-012 (SUO14x12)-Measurement and Sampling Imcations ..............................37 FIGURE 22: Radwaste Building, Radwaste Equipment / Components, Waste Collector Tank 1G11-TK-10A (SUO14 x12)-
Measurement and Sampling IAcations . . . . . . . . . . . . . . . . . . . . . . 38 FIGURE 23: Reactor Building, Reactor Water Clean-up System Components (SUO15)-Measurement and Sampling Locations ...............39 FIGURE 24: Radwaste Building, Condensate Demineralizers, Tank 1N52-DE-002E (SUO43)-Measurement and Sampling locations ...............40 menon.m. rn . mory i6, ms iii
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. LIST OF TABLES PAGE TABLE 1: Summary of Surface Activity Levels . . . . . . . . . . . . . . . . . . . . . . . 41 TABLE 2: Interior Exposure Rates . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 43
- i Embedded Piping Program Summary . . . . . . . . . . . . . . . . . . . . . .
TABLE 3: 44 i
TABLE 4: Confirmatory Radiological Status Summary-Structures . . . . . . . . . . . . 46 ;
TABLE 5: Confirmatory Radiological Status Summary-Systems . . . . . . . . . . . . . 48 i
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a .j Shareham-Brookhaven, NY . Febmary 16,1995 iV
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a ABBREVIATIONS AND ACRONYMS o
,, TASME- American Society of Mechanical Engineers em2 - square centimeter
' cpm - counts per minute DOE' Department of Energy -
dpm/IO0 cm2 disintegrations per minute per 100 square centimeters
- EML - Environmental Measurements Laboratory EPA Environmental Protection Agency
ESSAP Environmental Survey and Site Assessment Program ft2 square feet s ha hectare
'i F' GM ' Geiger-Mueller -
km kilometer L, critical level l
L LILCO Long Island Lighting Company L LIPA Long Island Power Authority . ~ a l'
m meter -
l , m2 square meter MDA minimum detectable activity NaI sodium iodide a
- NIST National Institute of Standards and Technology -
NRC- Nuclear Regulatory Commission ORISE Oak Ridge Institute for Science an'd Education '-
QA -Quality Assurance QC . Quality Control RW# Radwaste Building structural survey unit designation SNPS Shoreham Nuclear Power Station SP#- Suppression Pool structural survey unit designation SU# system survey unit designation R/h microroentgens per hour .
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hBrookhavve, NY February 16,1995 V
r L2 CONFIRMATORY SURVEY OFTHE RADWASTE BUILDING, SUPPRESSION POOL, PHASE 2, AND PHASE 3 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), which was designed to provide a gross electrical l output of 849 megawatts. Reactor criticality was achieved in February 1985. Low 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 l
-I 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.
i Various reactor components, piping systems, and other equipment became radiologically f contaminated as a result of reactor opw tion. The primary contaminants that were identified i during characterization studies included iron-55 (Fe-55), cobalt-60 (Co-60), nickel-63, and smaller quantities of tritium, carbon-14, nickel-59, manganese-54, zinc-65, and europium-157.8
-i 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 includes decontamination (grindi2g, hign pressure washing, etc.) or removal of contaminated portions of the reactor and other plant systems and equipment. A major con:;ideration of the decommissioning plan is to maintain the integrity, when possible, of plant structures and systems. Activities involved with the decommissioning .
i and termination surveys will be conducted in 4 phases. The initial phase involved the termination survey of the internal components of the main turbine, which has since been !
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ShBrookhsven. NY - February 16. 1995
p followed by termination surveys of the remainder of the structures and systems located within the Turbine Building as well as the site grounds and building exteriors. Phase 2 included the termination survey of the reactor suppression pool and several systems. Phase 3 involved the termination surveys for the Radwaste Building. Phase 4 will address the Reactor Building.
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 has requested that the Environmental Survey and Site ,
Assessment Program (ESSAP) of the Oak Ridge Institute for Science and Education (ORISE) conduct confirmatory radiological surveys and related activities for the SNPS decommissioning project as the various decommissioning milestones are completed. The results of the confirmatory survey of Phase 1, the turbine internal components and the Turbine Building, Site Grounds, and Building Exteriors, are the subject of separate reports.2.3 This report describes the results of the confirmatory process that has been completed for the Radwaste Building (Phase 3), Suppression Pool (Phase 2), and Phase 2 systems.
SITE DESCRIPTION SNPS is located in the Town of Brookhaven, New York on the north shore of Long Island, approximately 80 km (50 mile) east of La Guardia Airport and the confluence of the East River and Long Island Sound (Figure 1). Reactor and supporting operations were conducted within a 32.4 ha (80 acre) 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 known as the power block, where radiological controls had been necessary (Figure 2). Each of the buildings that have been or will be addressed during the confirmatory surveys are located here and are shown on Figure 2 as the Turbine Building, the Reactor Building, and the Radwaste Building. Radwaste Building construction is predominately of !
2 concrete and structural steel with a total floor space of approximately 4,700 m (51,000 ft2) l which is divided between three levels at elevations 15'6"/19'6", 37'6", and 50'6"/52'6" w n,. u. ny ra, ,y a. m s 2 )
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I (Figures 3 through 5). The systems and equipment housed within the Radwaste Building include the condensate demineralizers, the liquid radwaste system, the solid radwaste storage area, the ;
crane and truck bay, the makeup water treatment plant, chemical support system, and a portion of the off-gas radwaste system. The Suppression Pool is located on the 8' level of the Reactor Building and is constructed of steel plate (Figure 6). Surfaces and components within the ,
buildings remain essentially intact following decommissioning activities. I Termination surveys have been performed in accordance with Draft NUREG/CR-5849.* 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."5 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 ifinvolved with fuel handling or storage.
For the Radwaste Building, Suppression Pool, and Phase 2 Systems, a total of 77 survey units were addressed, of which 60 were structures and 17 were systems. Sixty-five of these survey units were classified by the licensee as affected.
OBJECTIVES ,
The objectives of the confirmatory activities were to provide independent document reviews, review and perform field observations of the LIPA procedures for embedded piping surveys, and Shortham-BrooLhaven, NY - Febnnry 16, 1995 3
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- develop radiological data for use byl the NRC in evaluating the adequacy and accurat une . I f
Llicensee's procedures and termination survey results.
q DOCUMENT REVIEW AND LIPA PROCEDURE SURVEILLANCE $ l q
l l .ESSAP reviewed'~LIPA's. termination survey procedures and the termination survey. release: ;
i records for those survey units s&cted for confirmatory survey.54 Documents were reviewed
- for adequacy, accuracy, completeness, and consistency. Data were reviewed fEr appropriateness !
of calculations and interpretations relative to the guidelines. In addition, ESSAP reviewed the < l
, applicable procedures and records for both the calibration ofinstruments used in, and the survey _.
data generated for, three representative sections' of embedded piping. Together with ,
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s observational surveillance of the resurvey of one.section of selected embedded piping? the
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documentation was evaluated for appropriateness and consistency in field application. j l
PROCEDURES .!
q During the period August 22 through 25,1994, an ESSAP team visited the SNPS and performed independent visual inspections, measurements, and sampling of' the Radwaste1 Building, . j Suppression Pool, Phase 2, and Phase 3 Systems. Surveys were performed in accordance with '
a survey plan submitted to and approved by the NRC.7 Nine structural' survey units and either q the complete or components of four system survey units were selected for confirmatory surveys.- 1 Survey units were selected either randomly by ESSAP or based on recommendations of the NRC :
site representative. Survey unit designators are alpha-numeric with the'first figures designating the type of unit, structural (building specific, RW=Radwaste, SP = Suppression Pool), or system, :
followed by a three-digit numeric reference. Subunits are given an additional'two-digit j
- designation preceded by X. The survey units selected and the respective classification for each ]!.!
'were: ;
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L Affected (A)/. Structure / System /
Survey Unit - Survey Unit Name Unaffected (U) Building Grounds RW013 Radwaste IS' North Hallway A structure RW017 ' Cation / Anion Regen and Resin 'A structure
- j. Storage Tanks Area RW023 Liner Fill Stations /BW Storage A structure Rooms (Cubicle A)
RWO40 Waste Evap./ Regen. Evap. Distil. A structure Room RWO42 Radwaste 37' North Hallway A structure RWO69 "B" RW O/G HEPA After Filter A structure Area RWO72 Off Gas Desiccant Dryers Area A structure SP004 Suppression Pool- NW Quadrant A structure SP005 Suppression Pool- Area Inside A structure Vessel Pedestal SU012 High Pressure Coolant Injection A system Valves E41-01V-3049 and 50 SU014X09 Radwaste Influent Drain System, A system Drain Sump Tank-654 SU014X10 Radwaste Building Equipment A system Drain System, Drain Sump Tank-071 SUO14X12 Radwaste Equipment / Components, A system Flat Bed Floor Drain Filter 1G11-FL-012 and Waste Collector Tank 1G11-TK-10A SUO15 Reactor Water Cleanup System A system SUO43 Condensate Demineralizers, Tank A system IN52-DE-002E Figures 3 through 6 indicate the structural survey units surveyed. Confirmatory surveys for SU012, SU014X09, SUO14X10, SUO14X12, and SUO43 involved individual component (s) rather than the entire survey unit.
Shareham-Brookhaven. NY - rebruary 16, IMS 5
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p Field survey activities were conducted in accordance with the applicable sections of the ESSAP Survey Procedures and Quality Assurance Manuals. The following procedures apply to survey
_ units selected for independent confirmatory surveys.
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 survey unit classification (affected vs. unaffected) and surface (floor, lower
, wall, upper wall, ceiling, or equipment). Typically, floor and lower wall grid blocks were 1mx1m. Upper surfaces and equipment were referenced to either these grids or other prominent building features. Systems were referenced by either a distance from a specific point,
. 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.
Additional sms were ; trformed over portions of upper wall, ceiling, and/or system surfaces where material may have settled or accumulated. Locations of elevated direct radiation detected by scans were marked for further investigation. Scans were performed using gas proportional, GM, and NaI detectors coupled to ratemeters or ratemeter-scalers with audible indicators.
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 measurements at locations of elevated direct radiation detected by surface scans. Alpha surface activity measurements were not required because the selected survey units were not classified as alpha affected, and no alpha SWBetdhma. NY . February 16,1995 6
F contamination vas identified by surface scans. Measurements were performed using gas propoKonal or GM detectors .oupled to ratemeter-scalers. A smear sample for determining removable activity levels wts collected from each direct measurement location. Figures 7 through 17 show measuremcat and sampling locations.
Exoosure Rate Measurements Exposure rate measurements in structural survey units were performed at each accessible floor grid block where ESSAP had performed direct surface activity measurements. All exposure rates were measured at 1 m above surfaces using a pressutized ionization chamber (PIC).
Figures 7 through 14 show measurement locations. Background exposure rates were previously determined during the confirmatory survey of the Turbine Building.'
Systems LIPA provided access poi..m into each system or system component listed on page 5 of this report. Beta and gamma surface scans were performed within the accessible portions of each system or component, followed by 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 4 to 30 measurements per system. Scans and direct measurements were performed using gas proportional, GM, and/or NaI detectors coupled to ratemeters or ratemeter-scalers. Figures 18 through 24 show measurement and sampling locations. l Embedded Pioing i
l Confirmation of the LIPA embedded piping termination survey program was accomplished i through on-site review and observation of the LIPA procedures for instrument calibration and embedded piping survey design and performauce, review of the data and records for the Shoreham-BroaLhavca. NY . February 16,1995 7
termination surveys of selected embedded piping subunits, and independent measurements and sampling of selected sections of piping. Specific procedures for each activity are described below.
The calibration records for the detector / instrument combinations, referred to as " pipe crawlers,"
used during termination survey data acquisitions for three selected embedded piping subunits were compared with and evaluated against the required calibration and operational check-out procedures and standards. These LIPA instrumentation procedures included the following:
Control of Health Physics Instrumentation (SP Number 61XG80.01), Chi-Square Test and Control Chart Review (SP Number 61X081.01), Detector Calibration (SP Number 66X020.11),
and Instrument Calibration (SP Number 66X022.02). In addition, the recalibration of one detector / instrument combination was observed. .ESSAP then performed independent measurements and calculations to confirm the level of, and distribution of, Co-60 activity on two of the custom sources that LIPA used for the calibration check on the pipe crawler detector ,
assemblies. This information was used to evaluate the appropriateness of the LIPA stated detector efficiencies. Once these parameters were established, ESSAP confirmed the LIPA computer-generated surface activity measurement data conversions to disintegrations per minute 2
per 100 square centimeters (dpm/100 cm), critical level calculations, and the action level calculations used by LIPA for identifying " hot spots" while surveying embedded piping.
Three sections of embedded piping were selected for confirmatory evaluation and independent surveys. Embedded piping subunit selection was based on recommendations of the NRC site representative. ESSAP performed surface scans and direct measurements for beta activity over a length of each selected pipe run, nominally 4 m in either direction of the access point. Access was gained through sumps, drains, traps, or openings where pipe sections had been removed. ;
A total of 17 direct measurements were made. The surface activity data collected were then compared to the LIPA-generated data from the corresponding section of each pipe run and the NRC surface activity guidelines. In addition, for each of these subunits, the Termination Survey j Design (LIPA Procedure SP Number 67X001.10) was compared to the field records and evaluated for appropriateness and consistent implementation. LIPA was also requested to Shoreham-Broalhaven, NY . February 16. 1995 8
resurvey portions of the selected embedded piping subunits and the results compared with the original termination survey data.
SAMPLE ANALYSIS AND DATA INTERPRETATION Samples and data were returned to ESSAP's Oak Ridge, Tennessee laboratory for analysis and interpretation. Smears were analyzed for gross alpha and gross beta activity using a low background proportional counter. Smear and direct measurement data were converted to units of dpm/100 cm 2. Because Fe-55 cannot be adequately detected with field instrumentation, a correction factor of 1.2 was applied to those surface activity measurements that exceeded background distribution levels, referred to as the critical level (Id. LIPA developed, 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.53 The 95% confidence level (p.), in accordance with NUREG/CR-5849, 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 was also performed. Exposure rates were reported in microroerigens per hour ( R/h).
Additional information concerning major instrumentation and survey and analytical procedures is provided in Appendices A and B.
DATA EVALUATIONS AND COMPARISONS The results of each survey unit sampled were statistically tested. The goal of the test was to determine, with a given confidence level, whether the LIPA survey data was not biased low compared to ESSAP. The null hypothesis was that in a survey unit, surface activities as calculated by LIPA were greater than or equal to those determined by ESSAP, i.e., H : LIPA ,
2: pESSAP. This hypothesis as tested at the 95% confidence level (0.05 level of significance). If the hypothesis was rejected at that confidence level, the alternative hypothesis was accepted i.e., 3H : LIPA < ESSAP. The test statistic, t, was calculated using the following equation:
Shareharn-Brouthavta, NY - February 16,1995 9
m X~X z- L I (n,-1)S, + (nt -1)Sj fn, + nt '
% "z * "1 ' 2 r "s"t ,
where:
5 is the LIPA surface activity mean for a survey unit 5 is the ESSAP surface activity mean for the same survey unit no is the number of LIPA direct measurement data points nn is the number of ESSAP direct measurement data points St, Sa are the standard deviations.
The calculated t was then compared to the critical value of Student's t-distribution (one-tailed) 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.
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 thoce survey units selected for confirmatory survey, indicated that the survey plan had been approprittely followed with no significant deviations. Data were appropriately converted, tested, and presented.
SURFACE SCANS Alpha, beta, and gamma surface scans identified one small area of elevated direct beta radiation on a support bracket in SP004. The surface activity level at this location was 2
3,800 dpm/100 cm . All other surface scans were comparable to background levels.
%nson, e. ny - ram-y 16. ms 10
SURFACE ACTIVITY LEVELS The results of total and removable surface activity levels are summarized in Table 1. The data reported below and in Table 1 is the difference between the gross field sample counts and area !
background. The difference is then corrected for detector efficiency and geometry, sample count time, and contributions from Fe-55 (when the net count rate exceeded the background distribution [I,]). Actual values are reported including negative surface activity levels, which occurred when the field count rate was less than the background. Of 414 measurements, 160 exceeded the I,.
Total beta activity levels for the structural survey units ranged from -310 to 3,800 dpm/100 cm2 .
The highest direct measurement was on the SP004 support bracket discussed previously. ;
Removable activity levels ranged from -1 to 6 dpm/100 cm2 for alpha and -7 to 81 dpm/100 cm2 for beta. The mean residual activity in structural survey units ranged from 14 to 510 dpm/100 cm 2 and -1.3 to 3.4 dpm/100 cm2 for total and removable beta activity, respectively.
2 Total beta activity levels in the surveyed systems ranged from -760 to 1,600 dpm/100 cm . The 2 2 removable activity levels were -1 to 8 dpm/100 cm for alpha and -6 to 14 dpm/100 cm for 2
beta. The mean beta activity levels for systems ranged from -310 to 98 dpm/100 cm for total activity and -1.3 to 1.7 dpm/100 cm2 for removable activity.
EXPOSURE RATES As determined during the Turbine Building survey, the interior background exposure rates ranged from 4 to 5 pR/h and averaged 5 R/h at I m. Individual gross exposure rates within i the Radwaste Building ranged from 4 to 6 pR/h at 1 m. The average gross exposure rates for all survey units ranged from 4 to 5 R/h at 1 m. The net survey unit exposure stes ranged j from -1 to 1 pRh at 1 m. Table 2 provides a summary of the exposure rates.
l l
&whe Bem, NY - Ferwy M, M5 11 1
l EMBEDDED PIPING Review of the embedded piping instrumentation calibration and calibration check; operational source cross-checks; survey procedure review and surveillance; and review of dat& conversions, critical levels, and action levels showed that all facets of the embedded piping survey program were performed in accordance with procedures and that the radiological status of embedded piping had been accurately presented. Specifically, ESSAP performed cross-checks of calibration source activity as well as LIPA reported detector efficiencies using representative pipe calibration geometries. The action level, in counts per minute, that LIPA developed and used for determining when small areas of residual activity in excess of the guidelines were present in a section of pipe was also confirmed. A confirmation of the embedded piping survey design was also conducted. For this confirmation, ESSAP evaluated the LIPA requirements for total number of measurements in a given pipe section and compared the required number to the actual number of measurements that LIPA performed. Because of the unique design of the pipe-crawler detector assemblies, ESSAP also evaluated the appropriateness of surface activity data conversions taken from gross field counts through the final reported surface activity levels. The final evaluations of the LIPA embedded piping program included field observation of embedded piping survey procedures, a comparison of original data and data generated when LIPA resurveyed three sections of pipe, and independent measurements by ESSAP for comparison with LIPA data. Table 3 provides a summary of the information developed during these evaluations.
Surface scans performed in each piping section did not identify any locations of elevated direct radiation. The ESSAP total activity results for each subunit are as follows: SUG14X02 (#366),
the range was -320 to 120 dpm/100 cm2 and the mean was -20 dpm/100 cm2 ; SU014X09 (#850),
the range was -360 to 40 dpm/100 cm2 and the mean was -140 dpm/100 cm 2; and SU016X01
(#20), the range was 400 to 690 dpm/100 cm2 and the mean was 540 dpm/100 cm2 , l I
COMPARISON OF RESULTS WITII GUIDELINES l I
The confirmatory survey results were compared with both the data provided by LIPA and the !
1 NRC guidelines for release to unrestricted use. The NRC's Regulatory Guide 1.86 provides the Samham-BemAhaven, NY - February 16, 1995 12 )
l
y . ,. ,
p >
.F ~ 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.-
i 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 4 5,000 dpm #-y/100 cm 2, averaged over 1 m 2 15,000 dpm S-y/100 cm2 , maximum in 100 cm2 L
Removable Activity 1,000 dpm #-y/100 cm2 As previously discussed, the detection sensitivities of the field instruments are such that the residual Fe-55 activity cannot be detected. Therefore, total and removable 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 ( ,
or upper confidence level [UCL]), relative to the guidelines, in accordance with Draft NUREG/CR-5849. These results are provided in Tables 4 and 5.
No direct measurements exceeded the average or' maximum total activity guideline. Overall,.
surface activity levels within each survey unit also satisfied the guidelines at the 95% confidence
~
level. All removable activity was below guidelines at the 95% confidence level.1The maximum removable activity identified was 81 dpm/100 cm2 .
Ascomparison of the ESSAP mean surface activity levels to the LIPA mean. activity livels -
- showed that the ESSAP mean was statistically less than or equal to the respective mean determined by LIPA for 6 of the 15 confirmatory survey units. The conditions established have therefore been satisfied for these survey units. However, the ESSAP mean was greater than the :
' LIPA mean for survey units RW013, RW017, RWO23, RWO42, RWO72, SP004, SUO43, SU014X09, and SUO14X10; therefore, additional evaluation was necessary.
Shoreham-Breathoven NY February 16,1995 13
I L
Of the nine units where the ESSAP total surface activity mean was higher than the LIPA mean, LIPA qualified the release record data for survey units RW013, RW017, RWO23, SP004, and SUO43 as containing excessive negative measurements. According to LIPA, this was due to y observed background levels being lower than the generic site backgrounds that were used for surface activity measurement conversions. 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 3,800 dpm/100 cm2 in SP004, and the maximum survey unit mean and UCL, found in RWO23, was 300 dpm/100 cm2 and 400 dpm/100 cm2 , respectively. Although the condition of the ESSAP mean being less than the LIPA mean was not satisfied for these survey units, both termination and confirmatory survey surface activity levels are below the 5,000 dpm/100 cm2average guideline.
The ESSAP surface activity means for survey units RWO42, RWO72, SU014X09, and SU014X10 were statistically higher than those for LIPA, and LIPA had not qualified the data for these units as biased low. Therefore, the LIPA surface activity levels, mean activity levels, and UCLs for all remaining survey units were evaluated to determine the potentialimpact on the LIPA reported status of the Phase 2 and Phase 3 survey units, relative to the guidelines. The maximum LIPA survey unit mean of 2,317 dpm/100 cm2 (SUO67) and maximum UCL of 812 dpm/100 cm2(RWO61) were within acceptable criteria. The maximum observed difference 2
of ESSAP and LIPA means was 330 dpm/100 cm . If this difference in activity levels were applied to the above, overall surface activity levels would not be significantly altered and the conclusions reached, that the radiological status of the Phase 2 and 3 survey units satisfies the guidelines, would remain valid.
For embedded piping, ESSAP's surface scans and direct measurement results did not indicate the presence of residual surface activity within the embedded piping, with most surface activity levels comparable to background and all levels below the minimum detectable activity (MDA) of the instrumentation. The comparison of the LIPA and ESSAP embedded piping direct measurement data indicates two discrepancies. First, the LIPA mean activity levels (Table 3) for two of the three sections of pipes investigated do not agree for the original survey and the resurvey of SU014X02 (#366) and SU014XO9 (#850). The probable cause of the difference in Simham-BronLhaven, NY Fetnvary 16,1995 14 l
a s
the means of the LIPA surveys may be attributed to variations in background levels. The difference in the average activity levels between the initial survey and resurvey of SU014X02
(#366) was 22 counts per minute (cpm) (3.7 cpm per detector), and for SU014X09 (#850) the difference was 9 cpm (1.5 cpm per detector). The initial survey results and resurvey results for SU016X01 (#20) were comparable.
3 The second discrepancy was identified when the comparison of the means indicated that the ESSAP mean was statistically greahr than the LIPA mean for each of the three data sets. The observed difference in the surface activity level means between LIPA and ESSAP may be the result of the differences in detector geometry. Each LIPA direct measurement location consists of a 93 square centimeter (cm2 ) area. LIPA's reported total activity is the additive activity contribution from six locations distributed around the interior circumference of each 0.3 m section of embedded piping where a measurement was performed, whereas ESSAP performed 2
one direct measurement, representing a 15.5 cm area, every 0.3 to 0.6 m. Based on these factors, LIPA's conclusions as to the overall radiological status of the SNPS embedded piping appear to be appropriate.
Exposure rates were compared with those obtained by LIPA and tested at the 95% confidence level, relative to the 5 pR/h above background guideline currently being used by the NRC (Table 4).' The Radwaste Building exposure rates were comparable to background exposure rate levels and confirmed the findings presented by LIPA.
i
SUMMARY
ESSAP performed confirmatory activities for the Radwaste Building, Suppression Pool; and l Phase 2 systems at the Shoreham Nuclear Power Station in Brookhaven, New York. '
Confirmatory activities included dccument reviews, and during the period August 22 through )
25, 1994, independent surfr.ce scans, surface activity measurements, exposure rate j measurements, and operaticaal surveillance were performed.
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- that. total and removable surface ~ activity. levels and' exposure rates were below4 the NRC -
' f ; 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.
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$ EXPOSURE RATE i METERS FIGURE 14: Radwaste Building, Off Gas Desiccant Dryers Area (RWO72) -
Measurement and Sampling Locations Shoretam.Brookhaven. NY - February 16,1995 30
'258-015 (2)
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g SINGLE-POINT N UPPER WALLS AND EQUIPMENT o 6 m
METERS FIGURE 15: Reactor Building, Suppressior Pool, NW Quadrant (SP004)
Lower Walls and Floor - Measurement and Sampling Locations samim..ane. ny - ra,n=y 26, ms 31
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LOCATIONS g SINGLE-POINT p UPPER WALLS O FELT 6 m
0 METERS I
-l FIGURE 16: Reactor Building, Suppression Pool, NW Quadrant (SP004) '
Upper Walls - Measurement and Sampling Locations Stacham Brathaven. NY . February 16,1995 32 l l
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Measurement and Sampling Locations w a - u - e., w y . rebn - y a , a95 34 1
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1 METERS hGURE 19: Radwoste Building, Rodwaste influent Drain System, Drain Sump Tank 054 (SU14XO9) - Measurement and Sampling Locations waev., n . v.dro.,y 26,1995 35
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METERS FIGURE 20: Radwaste Building Equipment Drain System, Drain Sump Tank-071 (SUO14X10) - Measurement and Sampling Locations w-war.am., tu . raory 26,1995 36
7-
-258-018 (3)
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N MEASUREMENT / SAMPLING h I l
LOCATIONS 4 DIRECT MEASUREMENT h
H FIGURE 21: Radwaste Building, Radwaste Equipment Drain i dter IG11-FL-012 (SUO14X12) / Components,
- Measurement and Flat Bed Floor Sampling Locations Shortharn-Brix4 haven, NY - Feb uary 16,1995 37
. . I 258-021 (4) i 1
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LOCATIONS i
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FIGURE 22: Rodwoste BuilJing, Rodweste Equipment / Components, Waste Collector Tank 1G11-TK -10A (SUO14X12) - Measurement and Sampling Locations ShlhtmLhaven. NY . Febnwy 16,1995 38
258-016 (2)
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n JL MEASUREMENT /SAMPUNG T
LOCATIONS
$ DIRECT MEASUREMENT b 0 6 m
i METERS FIGURE 23: Reactor Building, Reactor Water Clean-up System Components (SUO15) - Measurement cnd Sampling Locations ss=s .n,.as.m. rn . varu.ry a, im 39
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i
[ FIGURE 24: Rodwaste Building, Condensate Demineralizers, Tank- 1N52-DE-002E
- (SUO43) - Measurement and Sampling Locations j
l mm m . r*=,y is, im 40 ;
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I TABLE 1 1
SUMMARY
OF SURFACE ACTIVITY LEVELS '
RADWASTE BUILDING, SUPPRESSION POOL, AND PHASE 2 SYSTEMS .
f SIIOREIIAM NUCLEAR POWER STATION' BROOKHAVEN, NEW YORK Removaole Number of Total Activity Range Activity Range Location" Measurement (dpm/100 cm2) (dpm/100 cm 2)
Locations Betab.c Alpha d Beta' RW013,15' No 1h' Hallway- -
Floor 11 10 to 540 -1 to 1 -5 to 1 Lower Walls 9 -260 to 260 -1 to l' -4 to 4 '
Upper Walls 5 -210 to 100
-1 to 3 -5 to 3 Equipment 5 -260 to 450 -1 to 1 -3 to 8 RW017f 15' CAT /AN/REGN Tank Floor 10 -200 to 360 -1 to 3 -3 to 9 ' ,
Lower Walls 10 -100 to 190 -1 to 1 -4 to 3 .
Upper Walls 7 -100 to 38 -1 to 1 -5 to 0 :
Equipment 3 -200 to -14 -1 to 3 -4 to 3 RWO23,-19' Cubicle A, Liner Stations Floor 9 370 to 1100 -1 to 1 -4 to 3 3 Walls 11 -250 to 490 -1 to 3 -5 to 5 !
Upper Walls 3 -170 to 230 1 0 to 8 '
Equipment 7 170 to 200 -1 to 1 -5 to 3 RWO40,15' Evaporation Distil Room ,
Floor :
13 220 to 440 -1 to 3 -5 to 9 Walls 7 87 to 330 -1to.1 -4 to 3 Upper Walls 4 -26 to 180 -1 to 3 -1 to 5 Equipment 6 -43 to 660 -1 to 3 -5 to 1 RWO42, 37' North Hallway -
Floor 14 450 to 920 -1to'1 -4 to 4
) IAwer Walls 6 230 to 460 -1 to 5 -4 to 3 Upper Walls 3 190 to 280 -1 to 5 0 to 3 Equipment 7 i
-150 to 520 -1 to 1 -5 to 3 I ;
Shoreham Dnxtbaven, NY - February 16, 1995 41
TABLE 1 (Continued)
SUMMARY
OF SURFACE ACTIVITY LEVELS
- RADWASTE BUILDING, SUPPRESSION POOL, AND PHASE 2 SYSTEMS SHOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK Removable Number of Total Activity Range Activity Range Location" Measurement (dpm/100 cm2) (dpm/100 cm2)
Locations Betab,c Alpha d Beta'
~
RWO69, HEPA After FilterfArea 4 ^
Floor 11 110 to 490 -1 to 4 -3 to 6 Lower Walls 9 -62 to 130 -1 to 3 -4 to 10 Upper Walls 5 -48 to 95 -1 to 1 -4 to 3 Equipment 5 -210 to 140 - -1 to 1 -1 to 3 RWO72, Gas Desiccant Dryer Area' Floor 11 410 to 920 -1 to 3 -4 to 8 Lower Walls 9 250 to 550 -1 to 3 -5 to 5 ,
Upper Walls 3 270 to 380 -1t3 -5 to 1 Equipment 7 81 to 580 -1 to 3 -5 to 3 SP004 Suppression Pool'NW: Quad.~ 31 -310 to 3800 -1 to 6 -5 to 81 i SP005, Su'ppression Pool In. Pedestal 35 -210 to 350 -1 to 3 -7 to 8 SUO12,' Hi Pressure Coolant Irdection - 4 -440 to -20_0 -1 to 1 -5 to 8 SUO14x09fTink-54[ Influent Drain ' 31 -320 to 870 -1 to 1 -3 to 6 SUO14x10, Radwaste Equipment Dsains.. 11 -760 to 1600 -1 to 1 -4 to 6 l SUO14x12,-Radwaste EquipmehdComp.- 50 -720 to 440 -1 to 5 -6 to 14 I
~
SUO15 Reactor: Water Cleanup System 0 30 -520 to 1100' -1 to 5 -4 to 5 SUO43, Condensate Demineralizer: 12 -270 to 300 -1 to 8 -5 to 3
" Refer to Figures 7 through 24.
b Beta activity levels Corrected for Fe-55 contribution as appropriate.
'MDAs = 250 to 1100 dpm/100 cm 2, d
MDA = 12 dpm/100 cm2,
'MDA = 16 dpm/100 cm 2.
I shhamumven, NY - funury 16,1W5 42 i
i i
TABLE 2 -
INTERIOR EXPOSURE RATES .
RADWASTE BUILDING SIIOREIIAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK Number of Measurement Net Exposure Rates i tion.
Locations at 1 m (pR/h)b RW013 11 -1 to 0 RW017 6 -1 to 0 '
RWO23 8 -1 to O RWO40 6 -1 to 0 RWO42 13 0 to 1 RWO69 6 0 to 1 RWO72 11 0
- Refer to Figures 7-14.
b S te background exposure rate was 5 pR/h. ,
f I
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1 w a-e.=,xv.ran ,ya. m3 43 l
TABLE 3 EMBEDDED PIPING PROGRAM
SUMMARY
SHOREIIAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK
~
Calil$ ration L Pipe Efficiency l 4" Pipe l LIPA 6 detector Pipe Crawler Assembly 0.134' ESSAP HP-260 0.146b '
8" Pipe l LIPA 6 detector Pipe Crawler Assembly 0.152' ESSAP HP-260 0.163 b Description LIPA Expected / Actual
- ESSAP Confinnation 4" Pipe Crawler Calibration Operational Check 44/48 (dpm/cm 2) 44 (dpm/cm2)
Source Activity (dpm/100 cm2 )
Action Level (for a 15,000 dpm/15 cm2 hot spotd) 2408 cpm /830 cpm' 2410 cpm J
- Survey Design-Required Number of Measurements Confinned Number of Suncy nit (Per SP No. 66x020.11) Measurements SU016x01 (#20) 14 15 )
SUO14 x02 (#366) 28 30 SUO14 x09 (#850) 26 28
. Data Conversionf "
LIPA ESSAP Confirmation Location Surface Activity Critical 12 vel Surface Activity Critical Level (dpm/100 cm2) (dpm/100 cm 2) (dpm/100 cm') (dpm/100 cm )
20-1-3 -290 270 -290 270 ;
20-5-2 -298 270 -298 270 366-2-1 -382 175 -382 175 366-5-2 -271 181 -271 181 366-9-3 -338 178 -338 178 ShorehanBrathaven, NY - February 16,1995 44
. TABLE 3 (Continued)
EMBEDDED PIPING PROGRAM
SUMMARY
SIIOREIIAM NUCLEAR POWER STATION l BROOKIIAVEN, NEW YORK Data Conversion'(Continued)'
LIPA ESSAP Confirmation Location Surface Activity Critical Ievel Surface Activity Critical Level (dpm/100 cm2) (dpm/100 cm2) (dpm/100 cm 2) (dpm/100 cm 2) 850-3-2 -508 257 -508 257 850-5-1 -556 254 -556 254 850-6-3 -411 263 -411 263 850-8-1 -556 254 -556 254 Survey Rasults -
Survey Unit Data Origination SU0l6x01 (#20) SU014x02 (#366) SUO14x09 (850)
LIPA Termination Total Activity Range -468 to 169 -401 to -111 -379 to 16 Survey (dpm/100 cm ) l
-245 -240 -196 Mean (dpm/100 Activity) cm LIPA Resurvey Total Activity Range -395 to -121 -469 to -271 -581 to -315 ;
(dpm/100 cm2)
-245 -337 -488 Mean (dpm/100 Activity) cm ESSAP Survey Total Activity Range 400 to 690 -320 to 120 -360 to 40 (dpm/100 cm2) 540 -20 -140 Mean (dpm/100 Activity) cm l
- Per Table 4.3 Shoreham Decommissioning Project Termination Survey Plan, December 1993.
b Determined using an LIPA NIST traceable flexible mylar Co-60 source placed inside of a 4" and 8" piping ]
l spool piece.
- Determined from 8/19/94 LIPA Multiple G-M Detector Assembly Calibration Check Data Sheet (SPF66x020.11) for detector assembly IZ12-100PC4-0001, i d
Action level developed to identify small areas of contamination measuring less than 100 cm2 and matching a pipe-crawlers individual " pancake" detector active area.
'830 cpm is the LIPA alarm setting for the 6 detector assembly /ratemeter-scaler combinhtion, where additional investigation of possible hot spots is required.
' Data points analyzed were representative of infctmation collected during the resurvey of embedded piping subunits performed by LIPA. Data was tracked for accuracy from the field records to the final computer generated report.
SthBrookhaven, N)r . Tchruary 16,1995 45
TABLE 4 CONFIRMATORY RADIOLOGICAL STATUS
SUMMARY
-STRUCTURES, RADWASTE BUILDING AND SUPPRESSION POOL SHOREHAM NUCLEAR POWER STATION l
BROOKHAVEN, NEW YORK 1.
Radio:ogical Survey Unit"
"***U RWO13 RWO17 RWO23 RWO40 RW642 Total Beta Activity (dpm/100 cal) :
- of Direct Measurements 30 30 30 30 30 Mean(I) 82 '14 300 240 430 l l
LIPA X -170 -130. 13 110 280 !
Fa 140 60 400 290 510 5,000/15,0000 dpm/100 cnf Yes Yes Yes Yes Yes Guidelines Satisfied l
R5 movable' Beta Activity'(dpm/100 cnf)E >.
I !
- of Smears 30 30 30 30 30 ;
Mean(X) -0.5 -0.1 -0.4 0.2 -1.3 j LIPA I 7.5 2.6 3.9 3.2 0.3 i Fa 0.4 0.9 0.6 1.2 -0.5 1,000 dpm/100 cm2 Yes Yes Yes Yes Yes Guideline Satisfied Exposure Rates at 1 m'( R/h)' >
- of Exposure Rate Measurements 11 6 8 6 13 Net Mean (X) -0.7 -0.9 0.4 -0.5 0.2 )
Net LIPA X -0.2 0.5 0.4 0.0 0.2 !
Fa .L -0.7 -0.1 -0.2 0.3 5 R/h Above Backgrouno Yes Yes Yes Yes Yes Guideline Satisfied Shorcham-Brodhaven. NY . February 16. IM5 46
~:
l
[-
^
TABLE 4 (Continued) l CONFIRMATORY RADIOLOGICAL STATUS
SUMMARY
-STRUCTURES, ;
RADWA.STE BUILDING AND SUPPRESSION POOL SIIOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK Radiological Survey Unit' Sununan RWO69 RWO72 SP004 SP005 Total' Beta Xctitltp (dpm/100 cuf)1 ' '
^
- of Direct Measurements 30 30 31 35
- Mean(I) 93 510 110 100 LIPA I 360 280 -220 76 ;
- = 140 580 350 150 5,000/15,0000 dpm/100 cm' Yes Yes Yes Yes Guidelines Satisfied .
.t Rem'ovable' bee %5tivliy (dpm/100 cm)b.l 2 ,
- 3
- of Smears -30 30 31 35 :
Mean(X) 0.3 0.1 3.4 -0.2-LIPA X 3.4 2.9 9.0 5.0 F= 1.2 1.2 8.0 0.6 !
1,000 dpm/100 cm2 Yes Yes Yes Yes j Guideline Satisfied Exposme Rates at 1 m (pR/h)L : '
- of Exposure Rate Measurements 6 11 l
Net Mean (I) 0.3 0.0 - -
)
Net LIPA X 0.1 0.1 - -
F= 0.4 0.2 - -
5 pR/h Above Background Yes Yes - -
Guideline Satisfied
' Refer to Figures 7 through 17.
b All alpha removable activity was less than 12 dpm/100 cm2 .
- = Measurements not performed.
l l
sene=, in . Febru.ry 16. 1995 47 i
TABLE 5.
g CONFIRMATORY RADIOLOGICAL STATUS
SUMMARY
-SYSTEMS l SIIOREHAM NUCLEAR POWER STATION BROOKHAVEN, NEW YORK Radiological Survey Unit'.
Summary SUO12 SU014x09 SUO14x12 SUO14x10 SUO15 SUD43 Total Beta Astivity (dpm/100 Enf)I
- of Direct Measurements 4 31 11 50 30 12 Mean(X) -310 -48 26 80 -32 ' 66 LIPA X 150 -170 -170 11 87 4.6 F- -190 42 170 180 59 160 5,000/15,0000 dpm/100 cnf Yes Yes Yes Yes Yes Yes Guidelines Satisfied
~
Removahid Beta Activity (dpin/100 cuf)b,
- of Smears 4 11 7 50 25 12 Mean(X) 0 0.9 -1.3 1.7 0.2 -0.3 LIPA X 7.2 13.9 6.6 13.1 5.8 , 6.7 1
F= 6.9 2.8 1.3 2.6 1.0 _1.0 2
1,000 dpm/100 cm Yes Yes Yes Yes Yes Yes Guideline Satisfied l
' Refer to Figures 18 through 24. ,
b All alpha removable activity was less than 12 dpm/100 cm2, i
I sheham-Dmthaven, NY February 16,1995 48
g
- *' I
- .g ,
v: ..
- . . .I g; : REFERENCES j
- a. ,
- 1. Long Island Lighting Company, "Shoreham Nuclear Power Station Site Characterization.. !
Q r. . Program Final Report,":May 1990. ';
c .
- 2. T.t J. Vitkus, .ORISE, ;" Confirmatory Survey of the Turbine Internal . Components,-
~
l Shoreham Nuclear Power Station, Brookhaven, New York," July 1993. ;
a i
- 3. T. J. Vitkus, ORISE, " Confirmatory Survey of the Turbine Building, Site Grounds, 'and ~
s Site Exteriors, Shoreham Nuclear Power Station, Brookhaven,- New York," September 1994. ,
, a
' 4. J. . D.1Berger, Oak _ Ridge Associated Universities, Draft " Manual for Conducting Radiological Surveys in Support of License Termination," NUREG/CR-5849, June'1992 . ,
- 5. Long Island Power Authority, "Shoreham Decommissioning Project, Termination Survey .
Plan, Revision 1," April,1993. .
- 6. Long Island Power Authority, "Shoreham Decommissioning Project Termination Survey.
j Final Report, Volumes I through 5," September,1993.
- 7. Letter from T. J. Vitkus, ORISE to D. Fauver, U.S. Nuclear Regulatory Commission, 4 ;
Final Confirmatory Survey Plan for the Shoreham Nuclear Power Station, Brookhaven,- 1 New York - Docket File No. 50-322," November 4,1993.' .
i
- 8. Letter from D. N. Fauver, U.S. Nuclear Regulatory Commission, to T.'Vitkus, ORISE,'
~
j July 1,'1993. !
I
- 9. Letter from M. R. Landis, ORISE to D. Fauver, U.S.' Nuclear Regulatory Commission, i "Shoreham Decommissioning Project, Termination Survey Plan, Revision _O, Shoreham .l Nuclear Power Station, October 1992," January 12, 1993. 1
- 10. U.S. Nuclear Regulatory Commission, " Guidance and Discussion of Requirements for an j Application to Terminate a Non-Power Reactor Facility Operating License," Revision 1, . ;
September 1984. .!
l
.i i
l i
1 31 wm m . ran-y is, ms 49 ,
1 t
APPENDIX A MAJOR INSTRUMENTATION i
l l
1 i
l I
l l
SWBmolhaven, NY - 1%ruary 16,1995
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.
DIRECT RADIATION MEASUREMENT l l
Instruments Eberline Pulse Ratemeter 3 Model PRM-6 (Eberline, Santa Fe, NM) ,
Eberline " Rascal" Ratemeter-Scaler -
l Model PRS-1 (Eberline, Santa Fe, NM) ,
Ludlum Ratemeter-Scaler Model 2221 I
(Ludlum Measurements, Inc.,
Sweetwater, TX)
Detectors l Eberline GM Detector Model HP-260 ;
Effective Area,15.5 cm2 l (Eberline, Santa Fe, NM) l l
Ludlum Gas Proportional Detector Model 43-37 Effective Area,550 cm2 l (Ludlum Measurements, Inc.,
Sweetwater, m i
Ludlum Gas Proportional Detector -
Model 43-68 Effective Area,100 cm2 (Ludlum Measurements, Inc., .
Sweetwater, TX) sh=hma=uma, NY February 16,1995 A-1 l
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Reuter-Stokes Pressurized Ion Chamber
. Model RSS-111
, (Reuter-Stokes, Cleveland, OH) . ;
Victorcen Nal Scintillation Detector
. Model 489-55 3.2 cm x 3.8 cm Crystal .
(Victorcen', Cleveland, OH)
LABORATORY ANALYTICAL INSTRUMENTATION I
- Low Backgroun'd Gas Proportional Counter
- Model LB-5100-W-r- . (Oxford, Oak Ridge, TN) ;
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APPENDIX B SURVEY AND ANALYTICAL PROCEDURES j
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APPENDIX B SURVEY AND ANALYTICAL PROCEDURES L
-SICVEY PROCEDURES Surface Scans IJ Surface scans were performed by passing'the probes slowly lover the surface; the. distance a between the probe and the surface was' maintained at a minimum - nominally about I cm.' A-large surface area,' gas proportional floor monitor was used to scan the floors of the surveyed' 2
areas. Other 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.
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Beta -
- gas proportional detector with ratemeter-scaler
- pancake GM detector with ratemeter-scaler N
u' Gamma -
Nal scintillation detector with ratemeter Surface Activity Measurements
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Measurements of total beta activity levels were performed using GM and gas proportional detectors with portable ratemeter-scalers.
1 V .l L Count rates (cpm), which were integrated over 1 minute in a static position, were converted to I activity levels (dpm/100 cm2 ) by dividing the mt rate by the 4 x efficiency and correcting for L the active area of the detector. The beta activity background count rates for the GM and gas wm=. m - rehru.<y is.1995 B-1
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proportional detectors ranged from 22 to 34 cpm and from 129 to 174 cpm, respectively. Beta l cfficiency factors ranged from 0.16 to 0.17 for the GM detectors and from 0.21 to 0.23 for the l 1
gas proportional detectors. The effective window for the GM and the gas proporticaal detectors 1
were 15.5 cm2 and 100 cm 2, respectively. I Surface activity measurements which exceeded the normal background distribution were 2
corrected for the Fe-55 contribution by multiplying the dpm/100 cm 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 (Is). This level was defined for each detector / instrument combination as follows:
""P I' ' ""# '" + ackground count rate 136 h ample count time Background count time L* = (Detector E.[ficiency) (Detector Geometry)
Removable Activity Measurements Removable activity levels were determined using numbered filter paper disks, 47 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 pertinent information recorded.
Exposure Rate Measurements Measurements of gamma exposure rates were performed using a pressurized ionization chamber (PIC).
ANALYTICAL PROCEDURES Bemovable Activity .
Smears were counted on a low background gas proportional system for gross alpha, and gross -
beta activity.
Shortham-Bmoithaven, NY - Ftbruary 16,1995 B-2
UNCERTAINTIES AND DETECTION LIMITS ;
The uncertainties 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 ,
I uncertainties, associated with sampling and measurement procedures, have not been propagated into the data presented in this report.
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)]. Although data is reported as actual values, including negative values, in the document text and tables, the MDAs for 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 radionuclide in samples, the detection limits differ from the sample to sample and instrument to instrument.
CALIBRATION AND QUALITY ASSURANCE Calibration of all field and laboratory instnimentation 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 by the manufacturer.
i Analytical and field survey activities were conducted in accordance with procedures from the following documents of the Environmental Survey and Site Assessment Program:
- Survey Procedures Manual, Revision 8 (December 1993)
- 12boratory Procedures Manual, Revi:: ion 8 (August 1993)
- Quality Assurance Manual, Revision 6.1 (November 1993) ms u. xy . rearu.rr16.1995 B-3
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The proccJures 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.
Quality control procedures include:
- Daily instrument background and check-source measurements to confirm that equipment operation is within acceptable statistical fluctuations.
- Training and certification of all individuals pet.htming procedures.
- Periodic internal and external audits.
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APPENDIX C REGULATORY GUIDE 1.86, TERMINATION OF OPERATING LICENSES FOR NUCLEAR REACTORS i
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U.S. ATOMIC ENERSY COMMISSION June 1974 REGULATORY GUIDE DIRECTORATE OF REGULATORY STANDARDS REGULATORY GUIDE 1.86 t
TERMINATION OF OPERATING LICENSES FOR NUCLEAR REACTORS !
t 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 CFR Part 50, " Licensing of Production and Utilization from the reactor and placing it in storage (either onsite Facilities," requires that each license to operate a or offsite) may be continued.
production and utilization facility be issued for a specined duration. Upon expiration of the specified A licensee having a possession-only license must 1 period, the license may be either renewed or terminated retain, with the Part 50 license, authorization for by the Commission. Section 50.82,
- Applications for special nuclear material (10 CFR Part, 70, "Special termination oflicenses," specifies the requirements that Nuclear Material"), byproduct material (10 CFR Part must be satisfied to terminate an operating license, 30, " Rules of General Applicability to Licensing of including the requirement that the dismantlement of the Byproduct Material"), and source material (10 CFR facility and disposal of the component parts not be Part 40, " Licensing of Source Material"), until the inimical to the common defense and security or to the fuel, radioactive components, and sources are removed health and safety of the public. This guide describes from the facility. Appropnate administrative controls methods and procedures considered acceptable by the and facility requirements are iraposed by the Part 50 Regulatory staff for the termination of operating license and the technical specifications to assure that licenses for nuclear reactors. The advisory Committee proper surveillance is performed and that the reactor on Reactor Safeguards has been consulted concerning facility is maintained in a safe condition and not this guide and has concurred in the regulatory position. operated.
B. DISCUSSION A possession-only license permits various options rad procedures for decommissioning, such as Wnen a licensee decides to terminate his nuclear mothballing, entombment, or dismantling. The reactor operating license, he may, as a first step in the requirews imposed depend on the option selected.
process, request that his operating license be amended to restrict him to possess but not operate the facility. Section 50.82 provides that the licensee may The aDantage to the licensee of convertmg to such a dismantle and dispose of the component parts of a possession-only license is reduced surveillance nuclear reactor in accordance with existing regulat' ions.
requirements in that periodic surveillance of equipment For research rertors and critical facilities, this has USAEC REGULATORY GU1 DES c,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,
Beguletary Gu6 des are issued to desenbe and make available to the public desared to the U.S. Atomic Energy Commissiert Washington, D.C. 20545. ;
methods a septsbie to the AEC tegulatory staff of imp 6ementmg specihc parts Attentson: Dirutor of Reguistory Standards. Comments and suggestions for of the commiss on's agulations. to ortoinste techneques used by the staff in arnprwwnents in these gu6 des are sncouraged and should t>e sent to the ovatuetmo specshe probeems or postutsted accidents. or to prwide guidance to ' ^ 8Y nu hing%
appheants. Regulatory Gu6 des are not substitutes for regutstions and S w w me"*'n M ut out in gwon wN tab er pro e basis or { Tw gwdes are issued in tw vanowing ten beoed divwons.
mouwte to the issuance or continuance of a perma or scense by the
- 1. Power Reactors s. Products
- 2. Rourch and test Reactors 7. Transportation
- r" a' r: o=/~e:"n,='n e,=r" '" -""-'- U ,E6*f',; (E% '"
Note: Section electronically reproduced from photocopy. C-l
. usually meant the disassembly of a reactor and its b. A description of measurcs that will be taken to shipment organization for further use. The site from prev:nt criticality or reactivity changes and to which a reactor has been removed must be minimize releases of radioactivity from the facility.
decontaminated, as necessary, and inspected by the Commission to determine whether unrestricted access c. Any proposed changes to the technical can be approved. In the case of nuclear power specifications that reflect the possession-only facility reactors, dismantling has usually been accomplished by status and the necessary disassembly / retirement shipping fuel offsite, making the reactor inoperable, activities to be performed.
and disposing of some of the radioactive components.
- d. A safety analysis of both the activities to be accomplished and the proposed changes to the Radioactive components may be either shipped technical specifications.
off-site for burial at an authorized burial ground or secured. on the site. Those radioactive materials e. An inventory of activated materials and their remaining on the site must be isolated from the public location in the facility, by physical barriers or other means to prevent public access to hazardous levels of radiation. Surveillance is 2. ALTERNATIVES FOR REACTOR necessary to assure the long term integrity of the RETIREMENT barriers. The amount of surveillance required depends upon (1) the potential hazard to the health and safety of Four alternatives for retirement of nuclear reactor the public from radioactive material remaining on the facilities are considered acceptable by the site and (2) the integrity of the physical barriers. Regulatory staff. These are:
Before areas may be released for unrestricted use, they must have been decontaminated or the radioactivity a. Mothballing. Mothballing of a nuclear reactor must have decayed to less than prescribed limits facility consists of putting the facility in a state of (Table 1). protective storage. In general, the facility may be left intact except that all fuel assemblies and the The hazard associated with the returned facility is radioactive fluids and waste should be removed evaluated by considering the amount and type of from the site. Adequate radiation monitoring, remaining contamination, the degree of confinement of environmental surveillance, and appropriate security the remaining tadioactive materials, the physical procedures should be established under a security provided by the confinement, the susceptibility possession-only license to ensure that the health and to release of radiation as a result of natural phenomena, safe *y of the public is not endangered.
I and the duration of required surveillance.
- b. In-Place Entombment. In-place entombment i C. REGULATORY POSITION consists of sealing all the remaining highly radioactive or contammated components (e.g., the
- 1. APPLICATION FOR A LICENSE TO POSSESS pressure vessel and reactor internals) within a ;
BUT NOT OPERATE (POSSESSION-ONLY structure integral with the biological shield after LICENSE) having all fuel assemblics, radioactive fluids and wastes, and cenain selected components shipped A request .o amend an operating license to a offsite. The stnicture should piovide integrity over
[. possession-only license should be made to the Director the period of time in which significant quantities of Licensing. U.S. Atomic Energy Commission, (greater than Table 1 levels) of radioactivity remain Washington, D.C. 20545. The request should include with the material in the entombment. An the following information: appropriate and continuing surveillance program should be established under a possession-only
- a. A description of the current status of the facility, license.
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- c. Removal of Radioactivo. Components and - b. The physical barrices to unauthorized entrance Dismantling. All fuel assemblies, radioactive fluids into the facility, e.g., fences, buildings,' welded doors, and waste, 'and other materials having activities and access openings, should be' inspected at least above accepted unrestricted activity levels (Table 1) quarterly to assure that these barriers have not-should be removed from the site. The facility deteriorated and that locks and locking apparatus are owner may then have unrestricted use of the site intact.
with nu requirement for a license. If the facility owner so desires, the remainder of the reactor c. A facility radiation survey should be performed facility may be dismantled and all vestiges removed at least quanerly to veri'y that no radioactive material and disposed of. is escaping or being transported - through the contamment barriers in the facility. Sampling should
- d. Conversion to a New Nuclear System or a be done along the most probable path by which Fossil Fuel System. This alternative, which applies radioactive material such as that stored in the inner only to nuc! car power plants, utilizes the existing containment regions could be transported to the outer turbine system with a new steam supply system. regions of the facility and ultimately to the environs.
The original m. clear steam supply system should be separated from the electric generating system and d. An environmental radiation survey should be disposed of in accordance with one of the previous performed at least semiaanually to verify that no three retirement alternatives. significant amounts of radiation have been released to the environment from the facility.' Samples such as
- 3. SURVEILLANCE AND SECURITY FOR THE - soil, vegetation, and water should be taken at locations RETIREMENT ALTERNATIVES WHOSE for which statistical data has been established during _
FIN A L STATUS REQUIRES A reactor operations.
POSSESSION-ONLY LICENSE
- e. A site representative should be designated to be A facility which has been licensed under a responsible for controlling authorized access into and possession-only license may contain a significant movement within the facility.
amount of radioactivity in the form of activated and contaminated hardware and structural materials. f. Administrative procedures should be established Surveillance and commensurate security should be for the notification and reporting of ' abnormal provided to assure that the public health and safety are occurrences such as (1)the entrance of an unauthorized not endangered. person or persons into the facility and (2) a significant
- a. Physical security to prment inadvenent exposure change in the radiation or contammation levels in the of personnel should be provided by multiple locked facility or the offsite environment, barriers. The presence of these barriers should make it extremely difficult for an unauthorized person to gain g. The following reports should be made:
access to areas where radiation or contaminationlevels exceed those specified in Regulatory Position C.4. To (1) An annual report to the Director of prevent inadvertent exposure, radiation areas above Licensing, U.S. Atomic Energy Commission, 5 mR/hr, such as near the activated primary system of Washington, D.C. 20545, describing the results of the a power plant, should be appropriately marked and environmental and fachity radiation surveys, the status should not be accessible except by cutting of welded of the facility, and an evaluation of the performance of closures or the disassembly and removal of substantial security and surveillance measures.
structures and/or shielding material. Means such as a remote-readout intrusion alarm system should be (2) An abnormal occurrence report to the provided to indicate to designated personnel when a Regulatory Operations Regional Office by telephone physical barrier is penetrated. Security personnel that within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of discovery of an abnormal provide access control to the facility may be used occurrence. The abnormal occurrence will also be instead of the physical barriers and the intrusion alarm reported in the annual report described in the preceding systems. iter;.
Note: Section electronically reproduced from photocopy. C-3
- h. Records or logs rel:tive to the following items d. Upon request, the Commission may authorize a should be kept and retained until the license is licensee to relinquish possession or control of premises, terminated, after which they must be stored with other equipment, or scrap having surfaces contaminated in plant records: excess of the limits specified. This may include, but is not limited to, special circumstances such as the (1) Environmental surveys, transfer of premises to another licensed organization that will continue to work with radioactive materials.
(2) Facility radiation surveys, Requests for such authorization should provide:
(3) Inspections of the physical barriers, and (1) Detailed, specific information describing the (4) Abnormal occurrences. Premises, equipment, scrap, and radioactive contaminants and the nature, extent, and degree of
- 4. DECONTAMINATION FOR RELEASE FOR residual surface contamination.
I- UNRESTRICTED USE (2) A detailed health and safety analysis indicating If it is desired to terminate a license and to that the residual amounts of materials on surface areas, eliminate any further surveillance requirements, the together with other considerations such as the facility should be sufficiently decontaminated to prevent prospective use of the premises, equipment, or scrap, risk to the public health and safety. After the are unlikely to result in an unreasonable risk to the decontamination is satisfactorily accomplished and the health and safety of the public, site inspected by the Commission, the Commission may authorize the license to be terminated and the facility e. Prior to release of the premises for unrestricted abandoned or released for unrestricted use. The use, the licensee should make a comprehensive licensee should perform the decontamination using the radiation survey establishing that contamination is following guidelines: within the limits specified in Table 1. A survey report should be filed with the Director of Licensing, U.S.
- a. The licensee should make a reasonable effort to Atomic Energy Commission, Washington, D.C. 20545, eliminate residual contamination, with a copy to the Director of the Regulatory Operations regional Office having jurisdiction. The
- b. No covering should be applied to radioactive report should be filed at least 30 days prior to the surfaces of equipment of structures by paint, plating, or Pl anned date of abandonment. The survey report other covering material until it is known that should:
contamination levels (determined by a survey and documented) are below the limits specified in Table 1. (1) Identify the premises; )
In addition, a reasonable effort should be made (and l documented) to further minimize contamination prior to (2) Show that reasonable effort has been made to any such covering. reduce residual contamination to as low as practicable levels;
- c. The radioactivity of the interior surfaces of pipes, drain lines, or ductwork should be determined (3) Describe the scope of the survey and the general i by making measurements at all traps and other procedures followed; and 1 appropriate access points, provided contamination at these locations is likely to be representative of (4) State the finding of the survey in units specified j contamination on the interior of the pipes, drain lines, in Table 1.
, or ductwork. Surfaces of premises, equipment, or l scrap which are likely to be contaminated but are of After review of the report, the Commission may such size, construction, or location as to make the inspect the facilities to confirm the survey prior to surface inaccessible for purposes of measurement granting approval for abandonment. j should be assumed to be contaminated in excess of the I permissible radiation limits. l Note: Section electronicapy reproduced from photocopy. C-4
- 5. REACTOR RETIREMENT PROCEDURES As indicated in Regulatory Position C.2, several alternatives are acceptable for reactor facility retirement. If minor disassembly or "mothballing" is planned, this could be done by the existing operating and maintenance procedures under the license in effect.
Any planned actions involving an unreviewed safety question or a change in the technical specifications should be reviewed and approved in accordance with the requirements of 10 CFR I 50.59.
If major structural changes to radioactive components of the facility are planned, such as removal of the pressure vessel or major. components of the primary system, a dismantlement plan including the information required by 9 50.82 should be submitted to the Commission. A dismantlement plan should be submitted for all the alternatives 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 operating and maintenance 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 9 50.82. When dismantling is completed and the Commission has been notified by letter, the appropriate Regulatory Operations Regional Office inspects the tacility 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-only license under which the dismantling activities have been conducied or, as an alternative, may make application to the State (if an Agreement State) for a byproduct materials l license.
Note: Sectbn electronicauy reproduced from photocopy. C-5 j b
TABLE 1 ACCEPTABLE SURFACE CONTAMINATION LEVELS Nuclide' Average *# Maximum'd Removable *#
U-nat, U-235, U-238, and associated decay products 5,000 dpm a/100 cm 2 15,000 dpm a/100 cm2 1,000 dpm a/100 cm2 hi Transuranics, Ra-226, Ra-228, Th-230, Th-228, Pa-231, Ac-227, I-125, I-129 100 dpm/100 cm2 300 dpm/100 cm2 20 dpm/100 cm2 Th-nat, Th-232, St-90, Ra-223, Ra-224, U-232, el26,1-131,1-133 1,000 dpm/100 cm2 3,000 dpm/100 cm2 200 dpm/100 cm2 Beta-gamma emitters (nuclides with decay modes other than alpha emission or spontaneous fission) except Sr-90 and others noted above. 5,000 dpm #y/100 cm2 15,000 dpm #y/100 cm2 1,000 dpm Sy/100 cm2
'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) means the rate of emission by radioactive material as determined by correcting the counts per mi'2ute observed by an appropriate detector for background, efficiency, and geometric factors associated with the instrumentation.
' Measurements of average contaminant should not he averaged over more than 1 square meter. For objects ofless surface area, the average should be derived for each such object.
"The maximum contamination level applies to an area of not more than 100 cm2 . j
'The amount of removable radioactive material per 100 cm2 of surface area should be determined by wiping that area with I 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 removable contamination on objects ofless surface area is determined, the pertinent levels should be reduced proportionally and the entire surface should be wiped. .
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