ML20195C422
| ML20195C422 | |
| Person / Time | |
|---|---|
| Site: | 07001113 |
| Issue date: | 06/03/1999 |
| From: | Vaughan C GENERAL ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| NUDOCS 9906080125 | |
| Download: ML20195C422 (23) | |
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GENuclear Energy Generat DecV:c Company PO Brn 190. Wdmmg vn, NCdw?
9106b 5000 June 3,1999 Director Office of Nuclear Material Safety & Safeguards U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001
Subject:
License Revision to Chapter 5.0 - Respiratory Protection
Reference:
NRC License SNM-1097, Docket 70-1113
Dear Sir:
GE's Nuclear Energy Production (NEP) facility in Wilmington, N.C., hereby submits an amendment to the respiratory protection program requirements of our NRC Materials License SNM-1097. The changes are intended to maintain compatibility with Occupational Safety and lieslth Administration (OSIIA) requirements, and to allow us the flexibility to utilize the most appropriate, Federally approved respiratory protection equipment available.
In January 1998, OSliA rewrote the Federal regulations for respiratory protection [29 CFR 1910]. At our facility we utilize respiratory protection primarily for radiation protection purposes, but we also use it for protection against other airborne contaminants. The NRC and OSIIA regulatory requirements are for the most part compatible; however, the rewrite of 29 CFR 1910 has created some inconsistencies between our NRC license and the OSIIA regulation. The proposed changes will allow us to continue to have one respiratory protection program that meets the requirements of both the NRC and OSIIA, rather than administering dual programs.
In June 1995, the National Institute for Occupational Safety and Health (NIOSII) updated and modemized the Federal regulation for certifying air-purifying particulate respirators [42 CFR 84]. Under the new regulation, new classifications for air-purifying particulate respirators were created, stringent performance standards were established and rigorous quality assurance requirements were codified. The proposed license changes would allow us to select respiratory protection equipment based on the NIOSli certification, to the extent authorized under 10 CFR 20 Appendix A; and to utilize and maintain the equipment in accordance with the NIOSIt device specific conditions, rather than utilizing a generic maintenance and testing program for all equipment.
Attachment i provides a description of the changes by section and page. provides the Revisions By Chapter, the Table of Contents and Chapter 5.0 in its entirety.
All page changes involved in this revision have been dated 6/3/99. Vertical lines ( l ) in the right hand column indicate where changes have been made,
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. Director, NMSS June 3,1999 Page 2 of 2
' Six copies of this submittal are being provided for your use.
Please contact me on (910) 675-5656 or Rick Foleck on (910) 675-6299, if you have any questions or would like to discuss this matter further.
Sincerely, GE NUCLEAR ENERGY Charles M. Vaughan Manager Facility Licensing
/zb Enclosure cc:
CMV-99-025 S. Soong, NRC-HQ D. A. Ayres, NRC-Atlanta l
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.o Director, NMSS June 3,1999 Page1of2 ATTACIIMENT 1 Description of Revisions Page(s)
Section(s)
Description 1
Revisions by Chapter Changed the date on the Table of Contents, Revisions By Chapter and Chapter 5.0 to 6/3/99.
2 Table of Contents Changed the page number for Section 5.7 from 5.7 to 5.8.
5.10 5.10.1 In the second paragraph, added the option of using a quantitative fit test method, and clarified that fit tests are only performed for tight fitting face piece respirators.
The reason for this change is that the quantitative method is a more reliable means of determining an adequate fit, and it is compatible with the requirements of 29 CFR 1910. Fit tests are not performed for loose fitting equipment such as hoods or helmets.
5.10 5.10.2 Replaced the description of equipment types with a commitment to use NIOSH certified equipment in accordance with the protection factors in 10CFR 20 Appendix A.
This change will give us the flexibility to employ the 3
most appropriate equipment available that is allowed j
under Federal regulations. In June 1995, the National Institute for Occupational Safety and Health (NIOSli) updated and modemized the Federal regulation for certifying air-purifying particulate respirators [42 CFR 84]. The respirators certified under this new regulation are tested under much more demanding conditions than under the old regulation [30 CFR 11],
and they provide increased worker protection. While the current equipment in use at our facility meets the new standard, we would like to be in a position to take i
full advantage of new technologies as they are offered by the manufacturers, to the extent that they are authorized under 10 CFR 20 Appendix A.
I
.I Director, NMSS June 3,1999
' Page 2 of 2 ATTACIIMENT I Description of Revisions Page(s)
Section(s)
Description 5.10 5.10.3 Replaced the general maintenance and testing requirements with a commitment to meet the manufacturer's specifications for maintenance of the equipment.
This change is needed because each respiratory protection device has a unique NIOSH certification and instructions that stipulate the proper care and use of the device. Following these instructions insures that the equipment performs as designed.
The new NIOSH regulation also requires manufacturers to exercise rigorous quality control, so there is no longer a need for end users to test new equipment.
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Director, NMSS June 3,1999.
Page1of1 ATTACHMENT 2 -
Revisions By Chapter Table of Contents and I
Chapter 5.0 in its entirety Changes in this chapter are indicated with a vertical line ( l ) in the right hand column.
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Director,'NMSS June 3,1999'
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.' ATTACHMENT 2 Revisions By Chapter Table of Contents and Chapter 5.0 in its entirety Changes in this chapter are indicated with a vertical line ( l ) in the right hand column.
1 A
o REVISIONS BY CHAPTER Application Application Page Date Page Date l
TABLE OF CONTENTS l
l CHAPTER 6.0 l
I through 3 06/03/99 l
1 through 36 09/19/97 l REVISIONS BY CHAPTER l l
CHAPTER 7.0 I
1 06/03/99 l
1 through 3 06/05/97 l
CHAPTER 1.0 l
l CHAPTER 8.0 l
1 through 22 05/13/98 1 through 5 06/05/97 l
CHAPTER 2.0 l
l CHAPTER 9.0 l
1 through 11 03/10/98 1
06/05/97 n
l CHAPTER 3.0 l
CHAPTER 10.0 l
1 through 12 06/05/97 1 through 16 06/05/97 l
CHAPTER 4.0 l
CHAPTER 11.0 l
i 1 through 8 06/05/97 1
06/05/97 l
CHAPTER 5.0 l
l APPENDIX l
1 through 13 06/03/99 l
1 06/05/97 LICENSE SNM-1097 DATE 06/03/99 Page DOCKET 70-1113 REVISION 7
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TABLE OF CONTENTS
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Section Title Page REVISIONS BY CHAPTER I
CHAPTER 1.0 GENERAL INFORMATION 1.1 Facility and Process Description 1.1 1.2 Institutional Information 1.7 1.3 Special Authorizations and Exemptions 1.10 CIIAPTER 2.0 ORGANIZATION AND ADMINISTRATION 2.1 Policy 2.1 2.2 Organizational Responsibilities and Authority 2.1 2.3 Safety Committees 2.10 CHAPTER 3.0 CONDUCT OF OPERATIONS 3.1 Configuration Management (CM) 3.1 3.2 Maintenance 3.2 3.3 Quality Assurance (QA) 3.4 3.4 Training and Qualification 3.6 3.5 Iluman Factors 3.7 3.6 Audits and Assessments 3.7 3.7 Incident Investigations 3.9 3.8 Records Management 3.10 3.9 Procedures 3.11 Cil APTER 4.0 INTEGRATED SAFETY ANALYSIS 4.1 Integrated Safety Analysis 4.1 4.2 Site Description 4.1 4.3 Facility Description 4.1 LICENSE SNM-1997 DATE 06/03/99 Page DOCKET 70-1113 REVISION 1
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Section Title Page 4.4 Process Description 4.2 4.5 Process Safety Information 4.2 4.6 Training and Qualifications of the ISA Team 4.2 4.7 ISA Methods 4.2 4.8 Results of the ISA 4.3 4.9 Controls for Prevention and Mitigation of Accidents 4.4 4.10 Administrative Control of the ISA 4.7 CIIAPTER 5.0 RADIATION SAFETY 5.1 ALARA (As Low As is Reasonably Achievable) Policy 5.1 5.2 Radiation Safety Procedures and Radiation Work Permits (RWPS) 5.1 5.3 Ventilation Requirements 5.2 5.4 Air Sampling Program 5.3 5.5 Contamination Control 5.5 5.6 External Exposure 5.7 i
5.7 Internal Exposure 5.8 l
5.8 Summing Internal and External Exposure 5.9 5.9 Action Levels for Radiation Exposures 5.9 5.10 Respiratory Protection Program 5.9 5.11 Instrumentation 5.10 CII APTER 6.0 NUCLEAR CRITICALITY SAFETY 6.1 Program Administration 6.1 6.2 Technical Practices 6.5 6.3 Control Documents 6.29 6.4 Criticality Accident Alarm System 6.36 CilAPTER 7.0 CIIEMICAL SAFETY 7.1 Chemical Safety Program 7.1 LICENSE SNM-1097 DATE 06/03/99 Page DOCKET 70-1113 REVISION 1
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TABLE OF CONTENTS
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Section Title Page 7.2 Contents of Chemical Safety Program 7.1 CIIAPTER 8.0 FIRE SAFETY 8.1 Fire Protection Program Responsibility 8.1 8.2 Fire Protection Program S.1 8.3 Administrative Controls 8.2 8.4 Building Construction 8.2 8.5 Ventilation Systems 8.3 8.6 Process Fire Safety 8.3 8.7 Fire Detection and Alarm Systems 8.3 8.8 Fire Suppression Equipment 8.4 8.9 Fire Protection Water System 8.4 8.10 Radiological Contingency and Emergency Plan (RC&EP) 8.5 8.11 Emergency Response Team 8.5 CIIAPTER 9.0 RADIOLOGICAL CONTINGENCY AND EMERGENCY PLAN 9.1 CIIAPTER 10.0 ENVIRONMENTAL PROTECTION 10.1 Air Effluent Controls and Monitoring 10.1 10.2 Liquid Treatment Facilities 10.1 10.3 Solid Waste Management Facilities 10.2 10.4 Program Documentation 10.2 10.5 Evaluations 10.3 10.6 Off-site Dose 10.3 10.7 ALARA 10.4 CIIAPTER 11.0 DECOMMISSIONING 11.1 APPENDIX A.1 LlCENSE SNM-1997 DATE 06/03/99 Page DOCKET 70-1113 REVISION 1
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o CH APTER 5.0 RADIATION SAFETY 5.1 ALARA (AS LOW AS IS REASONABLY ACIIIEVAHLE) POI. ICY The GE-Wilmington standard of care for occupationally exposed individuals is to maintain exposures below the limits established by the U.S. Nuclear Regulatory Commission. Beyond the standard of care, the GE-Wilmington professional staff has a commitment to an ALARA program which is delineated by documented plant practices. Area Managers are responsible for implementing the ALARA program via engineered controls and supervision of operations personnel. The radiation safety function ensures that occupational radiation exposures are maintained ALARA via timely exposure monitoring and interaction with production personnel.
An annual ALARA review is conducted by the Wilmington Safety Review Committee as described in Chapter 2.0. The Radiation Safety Committee, also described in Chapter 2.0, meets monthly to maintain a continual awareness of the status of projects, performance measurements and trends, and the current radiation safety conditions of shop activities.
5.2 RADIATION SAFETY PROCEDURES AND RADIATION WORK PERMITS (RWPS)
Routine work perfonned in radiation controlled areas is administered by the use of standard procedures described in Chapter 3.0. Non-routine activities, particularly those performed by non-GE employees, which generally are not covered by documented procedures, are administered by the RWP system. The RWP system is described in documented plant practices.
Radiation Work Permits are issued by a radiation safety technician or supervisor for non-routine operations not addressed by an operating procedure when special radiation control requirements are necessary. The RWP specifies the necessary radiation safety controls, as appropriate, including personnel monitoring devices, protective clothing, respiratory protective equipment, special air sampling, and j
additional precautionary measures to be taken. RWPs are reviewed by radiation safety supervision.
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5.1 J
o The RWP requirements are reviewed by each affected individual and a copy is made av,ailable to the radiation safety function throughout the duration of the activity.
Work is monitored by the radiation safety function as required. RWPs have expiration dates and the status ofissued RWPs is reviewed on a weekly basis by a radiation safety technician or supervisor.
5.3 VENTILATION REQUIREMENTS 5.3.1 INTER-AREA AIR FLOW DESIGN Ventilation equipment is designed to provide air flow from areas oflesser potential contamination to areas of higher potential contamination. Direction of air flow between areas is checked monthly or after significant changes to the ventilation system. Ifinsufficient air flow results in airbome concentrations greater than 10 DAC, then the affected processes are shut down. Specific facilities and capabilities of ventilation systems are detailed in Table 5.1.
5.3.2 ENCLOSURES AND LOCALIZED VENTILATION lloods and other localized ventilation designs are utilized to minimize personnel exposure to airborne uranium. Activities and process equipment that generate airbome uranium are designed with filtered enclosures, hoods, dust capturing exhaust ports and other devices which maintain air concentrations of radioactivity in work areas such that personnel exposures are below 10 CFR 20 limits under normal operating conditions.
Air flows through hood openings and localized vents are maintained in accordance with Table 5.1. Additionally, differential pressure indicators are installed across exhaust system filters to monitor system performance. The flows and differential pressures are checked monthly or after significant changes to the ventilation system.
Ifinsufficient air flow results in airborne concentrations greater than 10 DAC, then the affected processes are shut down in accordance with plant procedures.
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5.2
5.3.3 EXIIAUST SYSTEM Po'tentially contaminated air is exhausted through high efficiency filter media which are at least 99.97% efficient for removal of 0.3 micron particles. HEPA filters in the exhaust system are equipped with a device for measuring differential pressure.
Differential pressures greater than four inches of water are investigated. In no case will filters be operated at a differential pressure which exceeds the manufacturer's ratings for the filter.
i Water scrubbers or other appropriate devices are provided where necessary to treat effluents before filtration. Such scrubbers are installed so that effectiveness of filters is maintained.
5.3.4 AIR RECIRCULATION Room air may be recirculated within the uranium processing areas after being filtered. Room air recirculated within areas where airborne concentrations are likely to exceed 0.1 DAC is filtered by HEPA filters and/or water scrubbers.
5.4 AIR SAMPLING PROGRAM 5.4.1 AIR SAMPLING Air samples are continuously taken from each main process area where airborne concentrations are likely to exceed 0.1 DAC when averaged over 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> to assess the concentrations of uranium in air. The air samples are collected in such a way that the concentrations of uranium measured are representative of the air which workers breathe. Air sampling results and individual personnel exposure assignments are monitored by the radiation safety function to evaluate the effectiveness of personnel exposure controls.
Fixed filter sampling points utilized for personnel exposure assignments are evaluated for representativeness annually and as part of each radiation safety function review for licensed process or equipment changes that may alTect airborne concentrations. Evaluations of air sampling representativeness are performed in accordance with the methods and acceptance criteria in Table 2 of Regulatory Guide 8.25, " Air Sampling in the Workplace".
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Filters from air samplers are changed each shift during normal operating periods or at mpre frequent intervals following the detection of an event that may have released airbome uranium, based upon knowledge of the particular circumstances. Filters are not changed as frequently during periods when no work is in progress. The filters are processed to determine the uranium concentration in air for each area.
Each air sampler is equipped with a rotameter to indicate flow rate of air sampled.
These rotameters are calibrated or replaced at least every 18 months.
Air sampling results in excess of 2.5 DAC (8 hr sample) and not resulting from a j
specific known cause are investigated to determine the probable cause. Operations or equipment will be shut down, and immediate corrective action will be taken, at locations where an air sample exceeds 10 DAC without a specific known cause.
Corrective actions are implemented and documented based on the frequency and magnitude of events causing releases of airborne uramum.
J Routine air sampling is supplemented by portable air sample surveys as required to evaluate non-routine activities or breaches in containment. Based on these surveys, additional radiation protection requirements for the particular operation may be established.
5.4.2 AIR SAMPLING ADJUSTMENTS l
Adjustments to Derived Air Concentration (DAC) and Annual Limit ofIntake (ALI) values in process areas to reflect actual physical characteristics of the airborne i
uranium are made in accordance with written operating procedures. GE-Wilmington j
site specific information on characteristics of airborne uranium is documented in internal records. For those areas in which adjusted ALI/DACs are applied, controls are established to limit soluble uranium intakes using air sampling and urinalysis.
Assigned air adjustments are not made to All/DACs for operations, locations or incidents where the airborne uranium physical characteristics are not documented.
Established airborne uranium limits in each area that adjusted ALI/DACs are used pursuant to the above authorization are reassessed by the Radiation Safety function at one quarter of the work locations at 6-month intervals, selecting different locations each time.
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5.4 J
If the reassessed limit (ALI) has changed by more than 30% from the previously established limit for an area, the limit for that area is either re-established or replaced with a default value based upon 1 micron AMAD.
In addition, a reassessment is performed following process or equipment changes likely to affect the airborne particle size distribution.
5.5 CONTAMINATION CONTROL 5.5.1 SURVEYS Routine contamination survey monitoring is performed for uranium process and manufacturing areas including non-controlled areas such as hallways and lunch rooms immediately adjacent to controlled areas. Removable contamination measurements are made based on the potential for contamination in these areas and operational experience. Survey frequencies are determined by the radiation safety function. Survey results are compared to action guide values as specified in plant procedures and appropriate responses are taken.
The minimum survey frequencies and maximum removable contamination action levels are as follows:
Action Limit 2
Area Freauency (dom a/100 cm )
Controlled Areas (Floors & Other Weekly 25,000 Readily Accessible Surfaces)
Eating Areas used primarily by Weekly 2220 Contro!!ed Area Personnel Non-controlled Areas Monthly 2220 When contamination levels in excess of action limits are found, mitigating actions are taken within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Personnel contamination surveys for external contamination on clothing and the body are required by personnel when exiting the change rooms. If contamination is found in excess of background levels, the individual attempts self-decontamination at LlCENSE SNM-1997 DATE 06/03/99 Page DOCKET 70-1113 REVISION 2
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i the facilities provided in the change rooms. If decontamination attempts are not j
supcessful, decontamination assistance will be provided by the radiation safety function. If skin or personal clothing is still found contaminated above background levels, the individual may not leave the area without prior approval of the radiation protection function.
5.5.2 ACCESS CONTROL Routine access points to controlled areas are established through change rooms.
Each change room includes a step-off area provided between the hot and cold sides.
Instructions controlling entry and exit from controlled area are posted at the entry points. Personnel survey meters are provided in the step-off area of each change room for use by personnel leaving the controlled areas. Posted instructions address the use of the survey meters and appropriate decontamination methods.
Alternate access points to controlled areas are established for specific activities that are not accommodated by the change rooms. Such access is governed by approved procedures, or Radiation Work Permits, which establish controls to prevent the spread of contamination to non-controlled areas.
5.5.3 PROTECTIVE CLOTHING Protective clothing is provided to persons who are required to enter the controlled areas where personnel contamination potential exists as determined by the radiation safety function. The amount and type of protective clothing required for a specific area or operation is determined by operational experience and the contamination potential. Available clothing includes caps, hoods, laboratory coats, coveralls, safety glasses, boots overshoes, shoe covers, rubber and cloth gloves and safety shoes.
The minimum clothing requirement for airborne controlled area entry is as follows:
Inspectors and Visitors Only Area Workers Observing Operations Shoe covers or work area shoes Shoe covers Coveralls Laboratory coats Head covers Ilead covers Rubber gloves Rubber gloves (as needed)
Safety glasses Safety glasses LICENSE SNM-1997 DATE 06/03/99 Page DOCKET 70-1113 REVISION 2
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l Trie protective clothing is removed upon exit in the controlled area change rooms.
In laboratory areas where uranium is handled the minimum protective clothing requirement for entry is a laboratory coat and safety glasses.
5.5.4 LEAK TESTING OF PLUTONIUM ALPIIA SOURCES The sources when not in use shall be stored in a closed container adequately designed and constructed to contain plutonium which might otherwise be released during storage.
The sources shall be tested for loss of plutonium at intervals not to exceed 110 days, using radiation detection instrumentation capable of detecting 0.005 pCi of alpha contamination.
If any survey or measurement performed as required by the preceding paragraph discloses the loss of more than 0.005 pCi of plutonium from the source, or if a source has been damaged or broken, the source shall be deemed to be losing plutonium. The licensee shall immediately withdraw it from use, and cause the source to be decontaminated and repaired, or disposed ofin accordance with the Commission regulations.
Records of test results shall be kept in units of microcuries and maintained for inspection by the Commission.
Notwithstanding the periodic test required above, any plutonium alpha source containing not more than 0.1 pCi of plutonium is exempted from the above requirements.
5.6 EXTERNAL EXPOSURE Deep-dose equivalent and shallow-dose equivalent from external sources of radiation are determined by individually assigned dosimeters. Personnel dosimeters are exchanged quarterly and processed by a National Voluntary Laboratory Accreditation Program (NVLAP) accredited vendor. The capability exists to process dosimeters expeditiously if there is an indication of an exposure in excess of established action guides. Action guides for external exposures are established in LICENSE SNM-1097 DATE 06/03/99 Page DOCKET 70-1113 REVISION 2
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i plant procedures. Maximum radiation exposure action levels are specified in Section 5.9.
External exposures may be calculated by the radiation safety function on the basis of data obtained by investigation when the results ofindividual monitoring are unavailable or are invalidated by unusual exposure conditions.
5.7 INTERNAL EXPOSURE Intakes are assigned to individuals based upon one or more types of measurements as follows: air sampling (described in Section 5.4), urinalysis and in vivo lung counting.
Intakes are converted to committed dose equivalent (CDE) and committed effective dose equivalent (CEDE) for the purposes oflimiting and recording occupational doses. Action levels are established in plant procedures to prevent an individual from exceeding the occupational exposure limits specified in 10 CFR 20. Maximum radiation exposure action levels are specified in Section 5.9. Control actions include temporarily restricting the individual from working in an area containing airborne radioactivity, and actions are taken as necessary to assure against recurrence.
5.7.1 URINALYSIS PROGRAM l
The urinalysis program is conducted primarily to evaluate the intake of soluble l
uranium to assure that the 10 CFR 20 intake limit of 10 mg is not exceeded.
l Individuals assigned to work in areas where soluble airborne uranium compounds are present in concentrations that are likely to result in intakes in excess of 10 percent of the applicable limits in 10 CFR 20 are monitored by urinalysis. The minimum sampling frequency for these individuals is biweekly. Urinalysis may also be used to monitor individuals involved in non-routine operations, perturbations or incidents.
Urine sampling frequencies and action levels are established in plant procedures based on the appropriate biokinetic models for the uranium compounds present.
Results above the applicable action level are investigated. Urinalysis action levels are based on maximum mdiation exposure action levels specified in Section 5.9.
Results that exceed action levels result in a temporary work restriction for the individual to prevent additional exposure and allow a more accurate assessment of the intake.
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5.7.2 IN VIVO LUNG COUNTING Routine in vivo lung counting frequencies are established for individuals who normally work in areas where non-transportable uranium compounds are processed.
Baseline and termination counts are performed when feasible. Lung counting frequencies are based upon individual airborne exposure assignments and previous counting results. The minimum count frequency is annual for individuals with an assigned intake greater than 10 percent of the Annual Limit on Intake (ALI).
Appropriate actions are taken based upon m vivo lung counting results to ensure the ALI will not be exceeded. If an individual's lung burden indicates an intake greater than the applicable action level, the individual is temporarily restricted from working in areas containing airborne uranium. In vivo lung counting action levels are based on the maximum radiation exposure action levels specified in Section 5.9.
5.8 SUMMING INTERNAL AND EXTERNAL EXPOSURE Internal and external exposures determined as described in the preceding sections of this application are summed in accordance with the requirements of 10 CFR 20 for the purposes oflimiting occupational doses and recording individual monitoring results.
5.9 ACTION LEVELS FOR RADIATION EXPOSURES Work activity restrictions will be imposed when an individual's exposure exceeds 80% of the applicable 10 CFR 20 limit; i.e.,0.8 ALI,1600 DAC-Ilours,4.0 rem CEDE,4.0 rem TEDE,4.0 rem DDE,40 rem CDE and 40 rem SDE.
5.10 RESPIRATORY PROTECTION PROGRAM The respiratory protection program shall be conducted in accordance with the applicable portions of 10 CFR 20. Respiratory protection equipment specifically approved by the National Institute for Occupational Safety and 11ealth (NIOSli) is utilized.
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l 5.10.1 QUALIFICATIONS OF RESPIRATOR USERS i
In'dividuals designated to use respiratory protection equipment are evaluated by the I
medical function and periodically thereafter at a frequency specified by the medical function to determine if the individual is medically fit to use respiratory protection devices. If the individual has no restrictions, he is provided respiratory training and j
fitting by a qualified instructor. Additional training on the use and limitations of j
self-contained breathing devices is provided to designated individuals.
j An adequate fit is determined for tight fitting face piece respirators (not loose fitting equipment such as hoods or helmets) using either a quantitative fit test method or a j
qualitative (irritant smoke) method. Mask fits are re-evaluated annually.
5.10.2 RESPIRATORY PROTECTION EQUIPMENT Only NIOSli approved respiratory protection equipment is utilized. Protection factors specified in 10 CFR 20 Appendix A are used for selecting the proper equipment and estimating personnel exposures.
5.10.3 EQUIPMENT MAINTENANCE Respiratory protection equipment is cleaned, serviced, tested and inspected in accordance with the instructions specified by the manufacturer per the NIOS11 certification for each respiratory protection device.
5.11 INSTRUMENTATION Appropriate radiation detection instruments are available in sufficient number to ensure adequate radiation surveillance can be accomplished. Selection criteria of portable and laboratory counting equipment is based on the types of radiation detected, maintenance requirements, ruggedness, interchangeability and upper and lower limits of detection capabilities. The radiation safety function annually reviews the appropriateness of the types ofinstruments being used for each monitoring function. Table 5.2 lists examples of the types and uses of available instrumentation.
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5.11.1 CAllBRATION In'strumentation is calibrated before initial use, after major maintenance, and on a routine basis at least six months following the last calibration. Calibration consists of a performance check on each range scale of the instrument with a radioactive source of known activity traceable to a recognized standard such as the National Institute of Standards and Technology (NIST).
Prior to each use, operability checks are performed on monitoring and laboratory counting instruments. The background and efficiency oflaboratory counting instruments are determined on a daily basis when in use.
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7 TABLES.I SPECIFIC FACILITIES & CAPABILITIES OF VENTILATION SYSTEMS Facility Alarms. Interlocks & Safety Features Pumose Hoods Air flow during operation 2 80 linear feet Prevents spread of radioactive per minute materials Effluent air filtered with HEPA filters Prevents release of radioactive materials to environs High Velocity Local Air flow designated to maintain an Prevents spread of radioactive Ventilation average of 200 linear feet per minute materials from work area to immediate room area UF, Vaporization Vented enclosure Provides containment in event Chambers of cylinder rupture or abnormalleakage Recirculating Air Air filtered in potentially contaminated Removes essentially all Systems & Exhaust zones with HEPA filters or water contaminants from room and Air Systems scrubbers exhaust to environs Pressure drop indicator set to alarm at Maintains adequate circulation 24" H 0AP across final filter for removal of dust and 2
contaminants from the room air Effluent air filtered with HEPA filters Prevents release of radioactive materials in environs LICENSE SNM-1097 DATE 06/03/99 Page DOCKET 70-1113 REVISION 2
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TABLE 5.2 TYPES & USES OF AVAILABLE INSTRUMENTATION (TYPICAL)
Tyne Tvnical Rance Routine Use DOSE ' RATE METERS GM Low Range 0.01 mR - 2000 mR Area Dose Rate Survey, Shipment Survey GM Ifigh Range 0.1 mR - 1000 R Emergency Monitoring Ion Chamber - Low Range 0.1 mR - 10 R Area Doe Rate Survey, Shipment Survey lon Chamber -liigh Range 1 mR - 1000 R Emergency Monitoring ALP 11A SURVEY METERS 50 cpm - 2 x 10' cpm Direct Personnel &
Equipment Surveys NEUTRON METERS 0.5 mR - 5 R Special Dose Rate Surveys LABORATORY INSTRUMENTATION Automatic air sample counter N/A Lab Analysis Fixed geometry Geiger-Mueller N/A Lab Analysis counter Scintillation Counter N/A Lab Analysis in Vivo Lung Counter N/A Lung Deposition Measurements LICENSE SNM-1997 DATE 06/03/99 Page DOCKET 70-1113 REVISION 2
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