ML20236K777
| ML20236K777 | |
| Person / Time | |
|---|---|
| Site: | 07000824, 07000778 |
| Issue date: | 07/30/1987 |
| From: | NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | |
| Shared Package | |
| ML20236K757 | List: |
| References | |
| NUDOCS 8708070246 | |
| Download: ML20236K777 (26) | |
Text
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JUL 3 01987 l
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I SAFETY EVALUATION REPORT BY TiiE DIVISION OF FUEL CYCLE, MEDICAL, ACADEMIC, AND COMMERCIAL USE SAFETY RELATED TO THE MATERIALS LICENSE RENEWAL FOR THE BABC0CK & WILC0X COMPANY NAVAL NUCLEAR FUEL DIVISION NNFD RESEARCH LABORATORY LYNCHBURG, VIRGINIA DOCKET NO.70-824 LICENSE N0. SNM-778 JULY 1987 j?a"1888R8?8?gy
JUL 3 01987 4
TABLE OF CONTENTS 1
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Page i
I.
Introduction....................................................
1 A.
General....................................................
1 B.
Loca ti o n De sc ri pti o n.......................................
1 C.
License History............................................
1 II.
Possession Limits...............................................
4 III. Authorized Activities..........................................
5 I
IV. ' Facilities......................................................
5-i V.
License Application.............................................
7 A.
Scope of Review...........................................
7 B.
Compliance History........................................
7 i
C.
Current Application........................................
8
)
i i
VI.
Organization and Administrative Procedures......................
8 A.
Organization and Responsibilities..........................
8 B.
Minimum Technical Qualifications...........................
10 C.
Administrative Procedures..................................
11 D.
Audits and Inspections.....................................
12 j
E.
Personnel Training.........................................
12 j
F.
Records....................................................
12 1
VII. Nuclear Criticality Safety......................................
12 A.
Technical Criteria.........................................
13 B.
Organization and Administrative Requirements...............
14 C.
Conclusions................................................
14 VIII. Radiation Safety..............................................
14 A.
Radiation Safety Administration............................
14 B.
Systems of Exposure Controls and Exposure Levels...........
15 C.
Bioassay Programs..........................................
17 0.
Control of Surface Contamination..........................
18 E.
Use of Respiratory Protective Equipment................
19 F.
E f f l ue n t C o n t ro l.........................................
19 G.
Conclusion.................................................
20 IX.
Environmental Monitoring........................................
20 X.
Radiological Contingency Plan..................................
20 XI.
Fire Safety.....................................................
21 ii
JUL 3 01987 TABLE OF CONTENTS (Continued)
Page XII. Plant Decommissioning...........................................
21
. XIII.
Conclusions...................................................
22 iii
JUL 3 01987 I.-
INTRODUCTION A.
General On November 26, 1985, the Babcock & Wilcox Company (B&W), Research and Development Division, Lynchburg Research Center (LRC), filed a renewal application for Materials Licence No. SNM-778.
Since that time, the i
license has remained in effect in accordance with the timely renewal provisions of Subsection 70.33(b) of 10 CFR Part 70.
The LRC is engaged in research and development activities of uranium fuels, overall fuel cycles, and in the examination and testing of irradiated fuels.
By supplement dated April 27, 1987, the NRC was informed that the LRC had been merged into the Naval Nuclear Fuel Division and will be known as the NNFD Research Laboratory (RL).
B.
Location Description l
The NNFD Research Laboratory is located in Campbell County, Virginia, near the James River, approximately 4 miles east of the city of Lynchburg.
Figure 1 shows the location of the Research Laboratory with respect to the Commonwealth of Virginia.
The site also contains the Commercial Nuclear Fuel Plant (CNFP) and the Naval Nuclear Fuel Division (NNFD), which are separately licensed, but also owned and operated by Babcock & Wilcox (see Figure 2).
C.
License History The first of the principal fuel handling buildings, Building A, was constructed in 1956 and licensed in August 1956 under License No. SNM-32.
Subsequently, the Company expanded its nuclear research activities, constructed the Building B complex in 1963, and later constructed Building C.
During this period, separate source material, byproduct material, and special nuclear material licenses were issued.
On July 16, 1966, the activities authorized by these licenses were combined into Materials License No. SNM-778.
Since that time, the license has been renewed twice.
On July 25, 1985, the expiration date was extended to January 1,1986, to allow the licensee time to rewrite its application in accordance with Regulatory Guide 3.52, " Standard Format and Content for the i
Health and Safety Sections of License Renewal Applications for Uranium Proces-sing and Fuel Fabrication." Ten amendments were issued during the period of this license.
By letter dated November 26, 1985, a renewal application was filed, and the license has remained in effect in accordance with the timely renewal provisions of Subsection 70.33(b) of 10 CFR Part 70.
Supplements to the renewal application were submitted on May 27, 1986, and January 20, April 27, and May 26, 1987.
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JUL 3 01987 II.
POSSESSION LIMITS (Proposed)
Material Form Quantity 1.
Uranium enriched to Encapsulated or 3.5 kg con-
>20% in the U-235 isotope irradiated tained U-235 2.
Uranium enriched to Unencapsulated 0.53 kg con-
>20% in the U-235 isotope and unirradiated tained U-235 3.
Uranium enriched >
l 5% to <20% in the~
Encapsulated or 1.2 kg con-irradiated tained U-235 i
U-235 isotope 4.
Uranium enriched >~
and unirradiated tained U-235 Unencapsulated 0.5 kg con-5% to <20% in the U-235 isotope 5.
Uranium enriched >
Encapsulated or 55 kg con-
.711% to <5% in the irradiated tained U-235 U-235 isotope 6.
Uranium enriched >
Unencapsulated 11 kg con-
.711% to <5% in the and unirradiated tained U-235 U-235 isotope 7.
U-233 Any 1 gram 8.
Plutonium Unencapsulated 50 g and unirradiated 9.
Source material Any 6,000 kg
- 10. Fission products Irradiated fuel Quantity
& transuranium contained in elements 4 irradiated fuel assemblies
- 11. Fission products Irradiated fuel 5,000,000 Ci and transuranium j
alements
- 12. Any byproduct Irradiated 50,000 Ci material structural materials &
components
- 13. Byproduct Any 3,000 Ci each material with total not to i
atomic nos. 3 exceed thru 83 1,000,000 Ci 4
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JUL 3 0 BB7 II.
POSSESSION LIMITS (Proposed) (Continued)
Material Form Quantity
- 14. Transuranium Any 20 millicuries elen ants each total not to exceed 100 millicuries
- 15. Cf-252 Sealed sources 4 mg
- 16. Am-241 Sealed sources 30 Ci
- 17. H-3 Sealed sources 100 Ci
- 18. H-3 0xide 3 Ci
- 19. H-3 Ni plated Sc 3 Ci tritide foil III. AUTHORIZED ACTIVITIES The licensee has requested authorization to use byproduct, source, and special nuclear materials in performing research and development activities as defined in Sections 30.4(q) and 70.4(j) of Title 10, Code of Federal Regulations, Parts 30 and 70.
The types of research and development activities performed are identified in both the license conditions and demonstrations sections of the license.
They are associated with the hot cell examination of irradiated and radioactive com-ponents including fuel; activities for other companies or B&W divisions includ-ing laboratory analysis, preparation, and testing of materials and equ'
_nt; preparation and modification of radiation sources; and preparation and uecon-tamination of reactor-related hardware for inspecting, evaluating, and measuring reactor components.
Accordingly, the following license condition is recommended.
Authorized Use:
For use in accordance with the statements, representa-tions, and conditicns of Chapters 1 thru 8 of the license application dated November 26, 1985, and supplements dated May 27, 1986, and January 20, April 27, and May 26, 1987.
IV.
FACILITIES The locations of primary buildings for the Research Laboratory are shown in Figure 3 and are as follows:
Building A housed the Lynchburg Pool Reactor (LPR) and the Critical Experiment Facility (CX-10) both licensed under 10 CFR Part 50.
The building was also authorized as a place of use for SNM materials under License No. SNM-778.
The building is being decontaminated and will eventually be removed from the license as an authorized place of use, l
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l Building B contains a four-cell hot cell facility with its associated operations area, cask handling area, transfer canal and storage pool, and variuus labora-tories associated with the examination of radioactive materials.
It also houses a demineralized for the cleanup of pool water.
Building C was used as a plutonium fuels development laboratory and for research and development of processes for other nuclear fuels.
It is being decontaminated and will eventually be removed from the license as an authorized place of use.
Building J and Annex are used for solid waste storage.
High, intermediate, and low-level wastes may be stored in Building J.
Irradiated fuel wastes are being stored until they are accepted by the Department of Energy in accordance with provisions in the Nuclear Waste Policy Act of 1982.
The Outside Waste Storage Area, adjacent to Building J, is limited to the storage of Type A quantities (10 CFR 71.4) and/or 0.5 g of fissile material in metal containers.
High-level solid waste is also stored underground in storage tubes imbedded in concrete.
The Liquid Waste Disposal Facility is used to treat liquid wastes which are then transferred to the NNFD Waste Treatment Facility for further treatment and disposal to the James River.
Liquids with high concentrations of radioactive materials are solidified and disposed of as solid wastes.
To authorize use of these facilities, the staff recommends the following condition:
Authorized Place of Use:
The licensee's existing facilities at the NNFD Research Laboratory.
V.
LICENSE APPLICATION A.
Scope of Review The safety review of the licensee's application included the application dated November 26, 1985, and supplements dated May 27, 1986; and January 20, April 27, and May 26, 1987.
The safety review also included the compliance history, organization, administration, radiation protection, and nuclear criticality safety programs.
On March 12-14, 1986, a site visit was made by N. Ketzlach, A. L. Soong, and M. A. Young to familiarize the reviewers with the facility's current activities, decontamination activities, and discuss the NRC comments on the November 26, 1985, application.
B.
Compliance History Inspections made by the NRC Region II Staff covering a 4 year period (1981-1985) were reviewed.
During this period, 14 safety inspections were made.
Five items l
of non-compliance were fouad during two of the inspections, resulting in Severity Levels III and IV violations.
None of these violations reflect basic program weaknesses or measurable adverse effects to the employees' health or to the health and safety of the public.
The licensee has been responsive and has taken appropriate action to resolve the problems identified during the inspections and to limit the potential for overexposure of personnel.
7
JUL 3 01937
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C.
Current Application The NRC staff has reviewed the applicant's commitments concerning organization and administration of the nuclear safety program for radiation protection and nuclear criticality safety.
These proposed commitments and special authoriza-tions are in Chapters 1-8 of the renewal application.
The NRC staff also reviewed the safety demonstration portion in Chapters 9-16 of the renewal application.
The following sections contain a description of the principal aspects of the l
organization and safety program, as proposed by the licensee, and the additional license conditions developed by the staff.
VI.
ORGANIZATION AND ADMINISTRATIVE PROCEDURES A.
Organization and Responsibilities The NNFD Research Laboratory line organization is shown in Figure 4.
The Vice President (VP) of NNFO has the ultimate responsibility for ensuring that all operations at the RL are conducted safely and in full compliance with NRC requirements.
The VP has delegated this responsibility to the Manager, Employee, Community, and Regulatory Relations (Manager, EC&RR), who has overall safety responsibility for operations under his direction.
The Manager, Safety and Licensing, the Facility Supervisor, and the Safety Review Committee (SRC) report to the Manager, EC&RR.
Area supervisors report to the Facility Super-visor and are responsible for the safety of all operations in their assigned areas.
The Facility Supervisor is responsible for the safe conduct of all operations at the RL and for ensuring that all applicable operations are conducted in compliance with the license and applicable regulations.
To fulfill these responsibilities, the Facility Supervisor has the authority to stop any opera-tion that is unsafe.
The Facility Supervisor reviews and approves all new Area Operating Procedures and Radiation Work Permits (RWP), and changes thereto, for license and regulatory compliance and facility safety.
The Supervisor, Health and Safety, is responsible for providing adequate facil-ities, procedures, and properly trained personnel to implement the health physics and industrial safety programs.
He is responsible for health physics and industrial-safety activities and reports to the Manager, Safety and Licens-ing.
The Supervisor, Health and Safety, has the authority to stop any opera-tion that he believes is contrary to accepted safety practices or license requirements.
He reviews and approves all new Area Operating Procedures, RWPs, and changes thereto, for radiation safety.
He conducts training programs for new employees and authorized users of radioactive material.
He is responsible for the shipment of licensed material.
The Supervisor, Health and Safety, is a member of the SRC but not a member of the Safety Audit Subcommittee.
A Senior Health Physics Engineer administers activities of the health physics staff and reports to the Supervisor, Health and Safety.
8
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JUL 3 01387
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FIGURE 4 SITE ORGANIZATION NNFD RESEARCH LABORATORY ORAGANIZTION NAVAL NUCLEAR FUEL DIVISION VICE PRESIDENT EMPLOYEE, COMMUNITY, &
REGULATORY RELATIONS MANAGER FACILITY SAFETY SUPERVISOR REVIEW COMMITTEE SAFETY & LICENSING MANAGER l
l l
NUCLEAR LICENSE HEALTH & SAFETY CRITICALITY ACCOUNTABILITY ADMINISTRATOR SAFETY SPECIALIST SUPERVISOR OFFICER SENIOR INDUSTRIAL HEALTH PHYSICS SAFETY ENGINEER OFFICER HEALTH PHYSICS GROUP I
I AREA AREA SUPERVISOR SUPERVISOR 9
JUL 3 01987 The Nuclear Criticality Safety Officer is responsible for ensuring that no operation at the site results in the inadvertent assembly of a critical mass.
He reviews and approves all new Area Operating Procedures, and changes thereto, I
for nuclear criticality safety.
He conducts training programs in criticality safety and performs criticality safety analyses.
He reports to the Manager, Safety and Licensing.
The SRC consists of experts in chemistry, nuclear physics, health physics, and the safe handling of radioactive materials.
The SRC membership is appointed by the Manager, EC&RR, and meets at least four times a year.
{
The SRC's principal functions are reviewing and/or approving all Area Operating Procedures (A0P), new projects, and major changes to existing projects utiliz-ing licensed materials; reviewing the annual report; audit findings; and all i
overexposure and unusual occurrences.
The SRC meeting reports are forwarded to the Manager, EC&RR, and the Vice President, NNFD.
The Licence Administrator reports to the Manager, Safety and Licensing, and is j
responsible for administering the license.
He is the primary liaison with the
{
NRC and other federal, state, and local agencies for matters pertaining to nuclear activities.
He is the coordinator of the SRC and the Safety Audit Subcommittee (SAS) and represents management on both.
He maintains the permanent records of the SRC and is responsible for assuring that appropriate action is taken to correct SAS audit findings.
j B.
Minimum Technical Qualifications 4
l Manager, EC&RR The Manager shall have a bachelor's degree or equivalent, 3 year's experience in the nuclear field, and demonstrate sufficient judgment and experience to manage the function under his direction.
Facility Supervisor The Supervisor shall have a bachelor's degree in science or engineering and 3 years' experience in the use and handling of licensed material or an asso-ciate degree in science or nuclear technology with 5 years' experience in the use and handling of licensed material.
Manager, Safety and Licensing The Manager shall have a bachelor's degree in a technical field and 5 years' experience in the nuclear field.
J Supervisor, Health and Safety The Supervisor shall have a bachelor's degree in a technical field and profes-sional experience at the supervisory level in assignments involving radiation protection.
He must have 4 years' experience, demonstrate proficiency in the application of radiation safety principles, and be knowledgeable in fields related to radiation protection.
10 l
R 30 W Senior Health Physics Engineer The Engineer shall have a bachelor's degree, which includes at least 20 quarter hours of health physics related course work and 2 years of radiation control-related experience or a master's degree plus 1 year of radiation control-related expe rience.
Nuclear Criticality Safety Officer The Nuclear Criticality Safety Officer shall have a bachelor's degree in science or engineering and either 2 years' experience with nuclear criticality safety calculations similar to those associated with RL activities or 1 years' experience with nuclear criticality safety calculations similar to those associated with RL activities, if he has at least an additional 2 years' experience in nuclear reactor physics calculations.
License Administrator The Administrator shall have a bachelor's degree in science or engineering and 3 years' experience in nuclear technology or an associate degree in science or nuclear technology with 5 years' experience in nuclear technology.
7 i
C.
Administrative Procedures The applicant has committed that A0Ps, which incorporate radiation safety and nuclear criticality controls and limits, will be prepared and followed for routine operations.
A0Ps will be approved by the Nuclear Criticality Safety Officer; the Supervisor, Health and Safety; the Facility Supervisor; and the Safety Review Committee.
An RWP is required when an activity is not covered by an A0P and personnel may be exposed to levels of radiation or concentrations of radioactive material in excess of those specified in 10 CFR 20.
RWPs are approved by the Area Supervisor, the Health Physics Supervisor, and the Facility Supervisor.
Technical procedures developed for health and safety or nuclear criticality safety are reviewed and approved by the Senior Health Physics Engineer or the Nuclear Criticality Safety Officer, respectively, or their designated alter-nate who meets the minimum qualifications of the same.
For initiation and review of a proposed addition or change to an A0P, project personnel submit the proposed procedure to the Facility Supervisor.
The Facility Supervisor forwards the submittal for review and approval to the Supervisor, Health and Safety, and to the Nuclear Criticality Safety Officer.
The Super-1 visor, Health and Safety, performs reviews for the radiological and industrial J
safety, and the Nuclear Criticality Safety Officer performs reviews for the nuclear criticality safety.
The nuclear criticality safety evaluations and calculations undergo an independent review by a qualified second party.
Approvals are granted by the Facility Supervisor and the SRC.
Procedure reviews and approvals are documented and reviewed annually by the Facility Supervisor to assure that the manuals contain the most current procedures.
11
JUL 3 01987 D.
Audits and Inspections The licensee has committed to a system of audits and inspections to ensure that plant operations are conducted in accordance with regulations, policies, and procedures.
The nuclear criticality safety audits are conducted once each calendar quarter by the Nuclear Criticality Safety Officer.
The Facility Supervisor performs a weekly nuclear criticality safety inspection of the opera-tions and reports the findings to the Nuclear Criticality Safety Officer.
The Supervisor, Health and Safety, conducts a monthly radiation safety audit.
The Safety Audit Subcommittee (SAS) conducts audits of the operations three times a year and an annual audit of the Health and Safety Group.
The annual audit is performed by a qualified individual who is independent of the Health and Safety Grouo.
The use of personnel separate from the organization being audited helps ensure that the audits are not biased.
Audits are conducted in accordance with written guidance and are documented.
The License Administrator is responsible for ;suring that the appropriate corrective actions are taken to address the audit findings.
E.
Personnel Training New employees receive instruction by qualified experts on basic characteristics of radiation, radiation protection procedures, nuclear criticality safety, and emergency procedures.
All employees receive on-the-job training.
The degree of training given depends on the extent of the employee's contact with hazardous meterials.
The radiation safety training program is administered by the Supervisor, Health and Safety, and the nuclear criticality safety program is administered by the Nuclear Criticality Safety Officer.
Records of training are maintained by those responsible for the identified program.
Retraining in radiation and in nuclear criticality safety is performed annually.
The effectiveness of the radiation and nuclear criticality safety training programs is evaluated and documented by written examinations.
F.
Records In the conditions section of the' license application, the licensee has committed to the maintenance of records for various required actions such as criticality analyses, internal audits, ALARA findings, employee training, routine surveys, and instrument calibration.
Nuclear criticality safety evaluation records are kept for at least 6 months after the termination of the approved operation.
All other records are maintained for at least 2 years or longer if required by regulation.
VII.
NUCLEAR CRITICALITY SAFETY The nuclear criticality safety control system at the RL is based on:
1.
Technical criteria using established policies, analytical methods, data, and safety margins.
2.
Qualified nuclear criticality safety staff with specified responsibility and authority.
12
JUL 3 01987 3.
Administrative requirements for written operating procedures, reviews of nuclear safety analyses, audits of operations, posting of limits, and training.
A.
Technical Criteria The technical criteria that the RL uses to establish the nuclear criticality safety of a proposed, revised, or new operation were provided in detail.
The important criteria are as follows:
1.
The basic policy is the double contingency policy which states that " Proc-ess designs should incorporate sufficient factors of safety to require at least two unlikely, independent, and concurrent changes in process condi-tions before a criticality accident is possible." This policy is endorsed by Regulatory Guide 3.4, " Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors."
2.
Where geometry of the container is not controlling, double batching is always considered possible.
The mass limit is no more than 0.45 of the minimum critical mass.
Mass limits are tabulated in the license condition section of the renewal application as a function of U-235 enrichment for uranium-enriched fuels, for mixtures of uranium-235 and plutonium, and for plutonium in the absence of uranium.
Mass limits have been based on data and calculations reported in the " Nuclear Safety Guide," TID-7016 (Rev 1);
DP-1014; as well as on validated calculations made by B&W using computer codes such as the Monte Carlo Keno Code.
3.
Geometry limits for a fuel assembly may be applied to LWR fuel rods when the effective multiplication constant for the water moderated and reflected assembly is less than 0.95.
The maximum safe slab thickness will be no more than 88 percent of the minimum critical slab.
These margins and those given in Paragraph 2 above are comparable to those used in the Nuclear Safety Guide, TID-7016 (Rev 1).
4.
The optimum (limiting case) conditions of water moderation credible for the system are assumed in setting limits.
5.
Unit and geometry limits are all based on full-water reflection.
6.
The licensee spaces the process equipment and stored units to meet the following general criteria:
a.
The closest approach of one individually subcritical unit to another is limited by mechanical means or clearly delineated criticality zones.
b.
Each individually subcritical unit stored under water is isolated from the adjacent units by at least 12 inches of water.
7.
The licensee spaces the process equipment and stored units to meet the following conditions:
a.
The minimum edge-to-edge spacing between mass limiting units is 8 inches and the minimum center-to-center spacing is 24 inches.
13
JUL 3 01987 b.
The maximum system k is 0.95 at the 95 percent confidence level.
eff B.
Organization and Administrative Requirements In addition to the requirements for qualified staff and established technical criteria, the RL's criticality safety involves several administrative practices:
1.
The Nuclear Criticality Safety Officer reviews and approves all Area Operating Procedures.
2.
Periodic audits are performed to assure compliance with safety require-ments, including quarterly by the Nuclear Criticality Safety Officer, three times a year by the Safety Audit Subcommittee, and weekly by the Facility Supervisor.
3.
Posting of nuclear criticality safety limits.
4.
Training and retraining of operations personnel.
5.
Authorized fuel handling, storage limits, and minimum spacing between units are license conditions.
C.
Conclusions The nuclear criticality safety review and our conclusion that the controls are acceptable are based on the following:
1.
The license conditions section improves clarity, incorporates conditions imposed from the previous license renewal, and ensures continued compli-ance with accepted practice.
The basic policy underlying these conditions is in accordance with Regulatory Guide 3.4, " Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors."
2.
The required recordkeeping system provides the necessary documentation for use by NRC Regional Staff for corroborating the licensee's compliance.
3.
Conformance of the technical criteria for nuclear criticality safety with established U.S. practices.
4.
Validity of the nuclear criticality safety analysis, including the demon-stration section.
5.
The history of safe plant operations with respect to nuclear criticality safety.
VIII.
RADIATION SAFETY A.
Radiation Safety Administration The Manager, Employee, Community, and Regulatory Relations, is responsible for the safety of operations involving licensed material that are carried out at the RL.
It is the responsibility of the Supervisor, Health and Safety (SHS),
to establish and maintain a health physics and industrial safety program, provide technical support services to the laboratory requiring radiation monitoring services, control radiation exposures to the workers, and control 14
JUL 3 01987 effluents released to unrestricted areas.
The SHS reports to the Safety and Licensing Manager and has the authority to cease those operations which are contrary to accepted safety practices or license requirements.
Overall objectives of the program are to ensure adequate containment of radioactive materials and reduce levels of radiation to the workers and the public to meet the As Low As Is Reasonably Achievable (ALARA) goal.
ALARA Commitment i
The licensee has committed to maintain radiation exposures to employees and the gencral public ALARA.
In order to achieve this goal, RL's Safety Review Committee reviews all audit findings anri the annual radiological report, which summarizes exposures, environmental releases, and the ALARA program accomplish-ment.
Committee recommendations are presented to the Manager, Employee, Com-munity, and Regulatory Relations.
It is the responsibility of the SRC Coordina-tor to ensure that the recommendations are being resolved.
Radiation Work Permit An RWP will be prepared for an activity involving licensed materials which could cause personnel exposures exceeding the regulatory limits or which is not covered by an A0P.
An RWP contains estimated radiation levels, radiation pro-tection requirements, and approval signatures by the Area Supervisor, Health Physics Supervisor, and the Facility Supervisor.
B.
Systems of Exposure Controls and Exposure Levels External Exposure External exposures of personnel are controlled and evaluated using personnel dosimeters as required by 10 CFR 20.202(a).
The dosimeters will be read and j
evaluated on a monthly basis.
An action level for an occupational exposure of 300 mrem week with exposare controlled within the 1.25 rem / quarter is estab-lished by the licensee.
i The external exposure data reported by the licensee for 1984 and 1985, as indi-cated by the following table, shows that the annual exposures are typically less than 20 percent of tne 5 rems limit as specified in 10 CFR 20, and no exposure exceeds the permissible limit.
Annual External Radia'. ion Exposure Summary Expressed as Percent Of Individuals In Each Range Annual Exposure Ranges, REM Max. Percent of 1984 1985 Allowable Limit No Measurable Exposure 0
40 58 Measurable <0.1 2
35 26 0.1-0.25 5
15 7
0.25-1.0 20 8
8 1,0-3.0 66 2
1
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JUL 3 01987 Internal Exposure At RL, protection of the operating personnel from excessive internal exposure is provided by:
1.
Ventilation systems designed to limit the concentration of radioactive material in the working area.
2.
An air sampling and analysis program for evaluating the airborne concentra-tions of radioactivity in working areas.
3.
A bioassay program to evaluate any significant deposition of radioactive material in the body.
4.
Protective clothing to minimize direct contact with the radioactive material.
5.
Respiratory protective equipment to limit the inhalation of airborne radioactive material.
l 6.
Surveys to detect the presence and extent of radioactive surface contamination.
7.
Procedures including action levels for investigation, control, and 4
decontamination of contaminated surfaces.
1 Description of Equipment Ventilation Systems i
The ventilation system in Building B, where essentially all the licensed mate-rial will be handled, provides an air pressure gradient so that airflow is directed from the work area into the exhaust system.
The exhaust air from hoods, hot cells, and glove boxes is filtered through HEPA filters and contin-uously monitored for radioactivity before being discharged through the stack.
The ventilation system maintains a face velocity across the hoods, where the licensed materials are used, of at least 100 1pm and maintains the differential pressure across the hot cell face at 0.25-inch of water.
The applicant has committed to maintain the effectiveness of the ventilation system as follows:
1.
The pressure drop across the final HEPA filter will be checked on a weekly basis.
HEPA filters shall be changed when the pressure drop exceeds 4 inches of water, except for the hot cell filters which shall be limited i
to 5 inches of water.
2.
Using the D0P test, the final HEPA filter will be checked on an annual basis or after a final HEPA filter is changed, whichever is earlier.
l 3.
The airflow through the hoods will be checked monthly and the differential l
pressure across the hot cell face will be checked weekly.
4.
On a monthly basis, the airflow direction in the work area will be checked to ensure the air is flowing in a designated direction.
l 16 (L
JUL 3 01987 Monitoring of Air Concentration for Radioactivity The ventilation system at the RL was designed and is operated to move air from areas of low potential concentration to high potential concentration.
In areas where the operation could cause personnel to be exposed to airborne radioactive material, the airborne concentration of radioactivity in the worker's breathing zone air shall be monitored and analyzed.
If an area in which the concentration of airborne radioactivity could exceed 10 percent of the limits specified in 10 CFR 20.103, the area's airborne concentration shall be monitored continuously and analyzed daily for radioactivity.
An area where the potential for airborne concentration of radioactivity is less than 10 percent of the regulatory limits, the atalyzing frequency will be weekly.
When the airborne radioactivity level in a restricted area exceeds 25 percent of the limits specified in 10 CFR 20, the licensee shall investigate the cause.
Furthermore, at least once every 12 months, the location of the fixed air sampler used to determine the breathing zone air concentration level shall be reexamined for its representativeness of workers' exposures.
l In Plant Airborne Activity Levels RL data on the concentration of airborne radioactivity in various work areas l
from 1983 through 1984 show that the average airborne concentration of radio-activity in the Cask Handling Area, All Laboratories, Hot Cell, Waste Storage Area, and Radiochemistry Labs was less than 1 percent of the MPC"# level specified in 10 CFR 20 for restricted areas.
C.
Bioassay Programs Internal exposure is evaluated and controlled by a bioassay program for uranium, plutonium, and fission products.
The program consists of urinary analysis and/
or in-vivo lung counts to determine exposure to radioactive material and lung burden of the material.
The bioassay program for uranium is conducted in accor-l dance with provisions similar to those in Regulatory Guide 8.11, " Application
~
l of Bioassay for Uranium," and provides for routine collection and analysis of workers' urine samples and in-vivo lung counts.
The bioassay sample frequency I
will comply with Regulatory Guide 8.11.
Action shall be taken when the urinary l
results exceed 9 pg/l of uranium or when the in-vivo lung count results exceed j
30 pg U-235.
l The bioassay program for plutonium provides for collection and analysis of urine samples at least every 6 months and in-vivo lung counting at least annually.
A measured result of 0.2 dpm/l in the urine sample or 16 nic Pu by in-vivo lung count would require an investigation of the cause.
At the present time, no formal guides have been developed by the NRC for plutonium bioassay.
RL's pro-gram for plutonium is equivalent to programs of other NRC licensees.
The RL's bioassay program for fission products is consistent with procedures specified in Regulatory Guide 8.26, " Application of Bioassay for Fission and Activation Products", and provides for routine in-vivo lung counting once every year.
Corrective action will be taken if workers' in-vivo lung count results are greater than 10 percent of maximum permissible organ burdens.
RL results of urine analysis for the years 1983 through 1984 indicated that all samples for uranium, except for one, were below the action point. The one resample showed the level had returned below the limit of detection.
All 17
JUL 3 0 B37 plutonium analysis are below the minimum sensitivity of the detecting instru-ments.
In-vivo lung count data for the same period indicated that no indi-vidual had been detected of having a lung burden of radioactive material that exceeded.the action levels for uranium, plutonium, or fission products.
D.
Control of Surface Contamination The facility is zoned to define areas as contaminated and clean.
Each defined area is routinely surveyed for possible contamination using an instrument which is calibrated every 6 months.
The survey frequency and action level for decon-tamination are based on the potential hazards presented by the presence of surface contamination.
Specifications for control of surface contamination are within the range of levels used at other similar types of nuclear facilities.
They are summarized as follows:
Guide to surface contamination control levels Action Level Survey frequency and (Smear) action when the level dpm/100cm2 is exceeded Location Alpha Beta gamma Alpha Beta gamma Unirradiated, Unencapsulated 5000 Weekly 24 hrs Fuel Handling Areas clean up Building B Counting Room 200 2000 Monthly Monthly Hot Cell Operations Area 200 2000 Monthly Bimonthly Electronic Microscopy Lab 200 2000 Monthly Monthly Exit Portal from Controlled Arcas 200 2000 Biweekly Biweekly Cask Handling 22000 Biweekly Radiochemistry Lab 22000 Bimonthly Cafeteria, Snack Bar and Vending Machine Areas Daily Daily At once
- Personnel Body Before leaving contaminated area
- If contamination is found, immediate corrective action shall be taken.
- If contamination is detected, prior approval of the Health Physics Staff is l
required for exiting the area.
1 1
18 j
JUL 3 01987 RL has requested authorization to use encapsulated sealed sources.
To ensure that these sources remain leak tight and are routinely tested, the staff recom-mends the following license condition:
Sealed sources shall be tested in accordance with the enclosed " License Condition for Leak Testing Sealed Sources Which Contain Alpha and/or Beta gamma Emitters."
The staff also recommends the following condition regarding release of equipment to offsite or onsite clean areas.
Release of equipment from the plant site to unrestricted areas or to onsite clean areas shall be in accordance with the enclosed " Guidelines for Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use or Termination of License for Byproduct, Source, or Special Nuclear Material" dated Nay 1987.
E.
Use of Respiratory Protective Equipment RL has a respiratory protection program to limit the workers' intake of radioactive material which is in compliance with 10 CFR 20.103.
F.
Effluent Control Air Effluent All potentially contaminated air from chemical hoods, hot cells, and glove boxes l
in Building B is discharged through at least one set of HEPA filters and released through the 50 meter high stack.
Discharges through the stack are continuously monitored for particulate and gaseous activities and sampled isokenetically to l
ensure a representative sample.
Data reported by the licensee shows that air-borne effluents released from 1982 through 1984 were less than 10 percent of the limits for unrestricted areas specified in 10 CFR 20.
Liquid Effluent All potentially contaminated liquid wastes are first stored in the tanks located in the Liquid Waste Disposal Facility.
The radioactive liqH d is then discharged to the liquid waste treatment plant at the NNFD facility.
The liquid waste is ultimately released to the James River via the NNFD waste treatment plant in l
accordance with 10 CFR 20 limits.
Prior to discharging liquid waste to the I
liquid waste treatment plant, the solution is sampled.
The liquid waste is not released to the treatment plant if the radioactivity in the liquid waste exceeds 25 percent of the MPC values of Table 1, Column 2, Appendix B, of 10 CFR 20.
To ensure that there is no unsafe accumulation of special nuclear material in the storage tanks in the Liquid Waste Disposal Facility, the licensee shall inspect the tank upon each tank voiding or annually, whichever is sooner.
Solid Waste Radioactive contaminated solid wastes will be packaged and shipped to a disposal facility in accordance with regulations (Title 10, Title 49, and requirements of burial site).
Prior to shipment, solid waste packages will be stored at a 19
JUL 3 019B7
)
- pecific-approved location within the facility.
If the package is stored out-j side, it will be stored in a closed metal container and limited to not more than a Type A quantity (10 CFR 71.4) and/or 0.5 grams of fissile materials.
The outside storage area is fenced and locked.
Containment of stored waste will be assured by a quarterly visual inspection by the Supervisor, Health and Safety.
Effluent releases from the RL have been within the regulatory requirements for discharge of radioactivity to unrestricted areas.
Detailed descriptions of the effluent releases and impacts resulting from the overall plant operations were published in the Environmental Assessment (NUREG-1227, December 1986) related to the license renewal.
G.
Conclusion Upon completion of the radiation safety review of the licensee's renewal appli-cation, the staff has concluded that RL has the necessary technical staff to administer an effective radiological safety program.
Conformance by RL to their proposed conditions as well as to those developed by the staff should ensure a safe operation and the detection of unfavorable trends.
IX.
ENVIRONMENTAL MONITORING The staff has analyzed the environmental impact of the continued operation of the Rt.
Action levels and release limits for airborne efn uents are established to maintain dose commitment to the nearest resident well below those allowable by EPA standards.
All potentially contaminated liquid wastes, after sampling, are released to the NNFD Waste Treatment Facility.
Liquic wastes released to the NNFD do not exceed 25 percent of the MPC values specified in 10 CFR Part 20, Appendix B, Table I, Column 2, for restricted are w.
The NNFD has established an environmental monitoring program which, in general, covers the entire 525 acre site occupied by the three B&W licensed facilities in Lynchburg.
The cumulative impacts of the overall BW site from the three facilities were assessed, and the individual dose commitment to the nearest resident is well below the current EPA standard for fuel cycle facilities as specified in 40 CFR Part 190.
The adequacy of RL's environment 1 program was evaluated in the " Environmental Assessment for Renewal of Materials License No. SNM-778", and a Finding of No Significant Impact was published in December 1986.
Based on the recommendations in the Assessment, the following condition is added to the license:
Within 90 days of the date of the license renewal, the licensee shall submit, for NRC's review and approval, a groundwater monitoring pro-gram for the detection of potential leakage of the below grade tanks in the Liquid Waste System (Building B).
The program shall include the location of the wells, monitoring frequency, analysis action levels, and the reporting requirements.
X.
RADIOLOGICAL CONTINGENCY PLAN The applicant has incorporated, by reference, the NRC-approved Radiological Contingency Plan submitted by letter dated November 15, 1983.
In the existing 20
f)L 3 01987 s
license, other requirements for radiological contingency planning were established.
The conditions have been incorporated into the application.
XI.
FIRE SAFETY The development and building construction of the Lynchburg Research Laboratory complex has taken place from 1955 to present.
For the tnree main buildings under consideration in this renewal request, the design and construction efforts took place from 1955 to 1969.
During the period of design and construction for Buildings A, B, and C, there was very little in the way of code requirements or guidance for construction of such a facility.
Virginia did not adopt a state-wide building code until September 1, 1973.
Ca nbell County, in which the RL complex is located, had no building code during this time period.
The only state-wide code directly applicable to building construction prior to 1973 was the Virginia Fire Safety Regulations, enacted in 1949.
The licensee has stated that good accepted engineering and construction practices, coupled with stringent requirements from insuring companies such as Factory Mutual, are used as the main criteria for such facilities; the result-ing structures usually far exceed the minimum requirements of various building codes.
Such is typically the case for all three buildings when examined from the load capacity standpoint as specified in the present Virginia Building Code, BOCA (Building Officials and Code Administrators) 1978, Seventh Edition.
The facility is insured by American Nuclear Insurers and inspected biannually by the Arkwright-Boston Insurance Company on behalf of the Mutual Atomic Energy Reinsurance Pool (MAERP).
A copy of the RL Certificate of Nuclear Energy Liability Insurance from American Nuclear Insurance and a copy of a recent inspection report by the Arkwright-Boston Insurance Company confirming the fire safety of the facility were included in the application.
XII.
PLANT DECOMMISSIONING The RL has updated the Decommissioning Plan approved October 9, 1978, in the license conditions section of the renewal application.
At the time of the renewal application, two programs were underway to decommission both Buildings A and C.
At the end of plant life, the facilities and equipment will be decon-taminated to meet the levels of contamination specified in the NRC " Guidelines for Decontamination of Facilities and Equipment Prior to Release for Unrestric-ted Use or Termination of Licenses for Byproduct, Source, or Special Nuclear Material, May 1987.
In addition, a reasonable effort will be made to further reduce contamination levels to those which are as low as reasonably achievable.
In the Decommissioning Plan, the Babcock & Wilcox Company has provided assurance of ability to cover the decommissioning costs on a continuing basis and that they do not exceed the Company's working capital.
This commitment was reaffirmed in a letter dated July 17, 1986, from L. V. Jordan, Assistant Controller.
21
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JUL 3 01987 The staf f recommends the following license condition to identify the decommis-sioning requirements at the end of plant life and to cover the decontamination of the grounds which were not specifically identified in the Plan:
At the end of plant life, the licensee shall decontaminate the facility and site in accordance with the general Decommissioning Plan submitted with the renewal application dated May 26, 1987, so that these facilities and grounds can be released for unrestricted use.
The financial commitment associated with the Plan, assuring funds for decommissioning, is hereby incorporated as a condition of the license.
XIII.
CONCLUSIONS Upon completion of the staff review of the licensee's application and compliance history, the staff has concluded that the activities authorized by issuance of a renewed license to the NNFD Research Laboratory, subject to the additional condi-tions developed by the FCSS staff, will not constitute an undue risk to the health and safety of the public.
Further, the staff has determined that the application fulfills the requirements of 10 CFR 70.23(a).
The staff has discussed the renewal with Mr. G. Troup, the Region II Project Inspector on June 9, 1987.
He feels the proposed licer.se addresses Region II's concerns and has no objection to the issuance of the renewal.
The staff therefore recommends the license be renewed for a 5 year term in accordance with the statements, representations, and conditions in the RL application dated November 26, 1985; supplements dated May 27, 1986, January 20, April 27, and May 26, 1987; and license conditions recommended by the staff.
P GW An-Lian oong Uranium Fuel Section Fuel Cycle Safety Branch Division of Fuel Cycle, Medical, Academic, and Commercial Use Safety Original Signed Ey Approved by:
Leland C. Rouse, Chief Fuel Cycle Safety Branch A/
OFC: FCUFh1A,
- FCUF
- FCSB[.
NAME: ALSoong/ht/as:VLTharpe: LC' ouse:
DATE:7/d/87
- 7M/87 :7/10/87:
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