ML19257D351
| ML19257D351 | |
| Person / Time | |
|---|---|
| Site: | 07000824 |
| Issue date: | 12/04/1979 |
| From: | Olsen A BABCOCK & WILCOX CO. |
| To: | Rouse L NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| 14901, NUDOCS 8002040182 | |
| Download: ML19257D351 (58) | |
Text
l' PGA 70- 924 Babcock &Wilcox aesza,cn.no o..iopment oi isio.
P.O. Box 1260, Lynchburg, Va. 24505 Telephone: (804) 384-5111 December 4, 1979 Mr. L. C. Rouse, Chief Fut1 Processing and Fabrication Branch Division of Fuel Cycle and Material Safety U. S. Nuclear Regulatory Commission Washington, D. C.
20555
Reference:
BAW-381, " Demonstration and Conditions for License SNM-778",
December, 1978.
License: SNM-778, docket 70-824 Jear Mr. Rouse:
The application for renewal of the referenced license was submitted to you in December, 1978.
Since that time revisions one and two, dated September and October, 1979 respectively, have been submitted in response to comments by members of your staff.
Attached is revision three to the reference, responding to additional comments from your staff. An instruccion sheet for inserting these replacement pages in your copies of the reference is also attached.
I was requested to furnish evidence of insureability, the latest report of an inspection performed by or for the insurance company and evidence that the building covered in the renewal application were constructed in accordance with state and local codes and standard. Responses to the items above are attached.
The replacement pages for the referenced document are Xerox copies.
I will forward a printed replacement in the near future.
Yours very truly, Babcock & '411cox Company Lynchburg Research Center Arne F. Olsen License Administrator AF0:ccf 1868 115 800204o l 3 2_
Attachment 14301 The Babcock & Wilcox Company / Established 18
Allendale Mutual Insurance Company CERTIFICATE OF INSURANCE We hereby certify that insurance coverage is now in force with our Company as outlined below. This Certincare does not amend, extend or otherwise alter the terres and conditions of insurance coverage contained in this Policy.
Title Of Insured:
BABC0CK & WILCOX COMPANY AND ITS SUBS 101 ARIES Policy:
Effective Empires Binder No.
MAERP PD October 1,1979 October 1.1980 RG-1166 Description & Location Of Property Covered: (fxcept as otherwise specincolly excluded)
SEE ATTACHED SHEET NO. 2 0F CERTIFICATE.
(includes Insured's interest in improvements & Betterments)
Perils insured: (Subject to the conditions of our Standard Form)
Fire, fightnmg, wind, hail, sprinkler leakage, liquid damage, emplosion, riot, civil commotion, vandalism, malicious mischiet, acts of civil authority, vehecies, aircra't, sonic boom, smoke, escaped molten materials, radioactive contamination, collapse and volcanic ensption.
To Whom it May Concern:
Th is is to certify that this Company instires the subject location against 1oss to Property Damage subject to the terms and conditions of the above Policy.
1868 116 Account No. 1-30368-ALLENDALE MUTUAL INSUR ANCE COMPAN 830 Morris Turnpike Short Hills, New Jersey 07078 BY
~
Date:
November 2, 1979 A h.:edSs. nature 187.A (11/7G N. Y. Region R. J. Gaffney
BA'BCOCK & WILCOX COMPANY Page 2 of 2 Binder No. RG-1166 Account No. 1-30368 Description of Location Location No. 1 Index No. 45336.3 Lynchburg, Virginia Nuclea,- Reactor Fuel Elements, Critically Test & Assembly Facilities, Nuclear Development Center, Fuel Lab including Fuel & Commercial Fac. Plants; on premises bounded by U.S. Route 460, a line 200 f t. N/W of the Critical Experiment Lab.,
Bldg. No. 11 and a line 1,000. ft N/E of the Nuclear Development Center and a line 500 ft S/E of the Fuel Element Bldg., Bldg. No. 1, including the Pump House located 3,500 ft, north of the Filter Plant, Bldg. No. 7 on the East side of the James River and the Tank and Appurtenances located approximately 1,500 ft South of the Gare House, Bldg. No. 9.
1868 117
-%~
l Authorized Underwriter (}'
R. J. Gaffney
AIIendate Insurance Allendate Park, Johnston, Rhode Island 02919
@$0@@@@@70 0 @ @@@@@T MAERP MUTUAL ATOMIC ENERGY REINSURANCE POOL
- Dsz 45336.30 THE BABCOCK & WILCOX COMPANY "RESEARCH CENTER" Fw Rte 360
^cc U"Y 01-30368 Lynchburg, Virginia 24505 March 29, 1979 R. C. Cribble o.
sy c=rwacevim Mr. C. Bell, Mgr. Fac.
No gN a 4 Yes s Aaan a issata = tory AREAS srRmxLEREn 55 a..,,...,.
IO M
sourEurErneo W ATER sVFFLT ig5attafectory Yes MADtTENANCE & PEPAIR 195a11stactory ALL YALVEs FOUND OPEN is satisfactory sVPERT1310N HRE EQUIPMENT IS atisfactory CRTnCALITT CONTROL 5
is satisfactory FLANT EMERGENCT ORGANIEADON & W ATCRMEN isssetafW NUCLE AR REACTOR OPERATION RADict30 TOPE MAND 1JNG
SUMMARY
This is a well maintained nuclear materials development center.
GENERAL REMARKS Plant mechanics have installed automatic sprinklers in the storage room above the hot cells using pendant heads on ordinary hazard pipe schedule and supplied by connection Since the heads are too far below to the 2 in. piping supplying small hose stations.
ii the roof deck, standard upright heads are to be installed in the upright pos t on.
(77-9-3)
Plant mechanics have installed automatic sprinklers in the Furnace Annet using pendant Since heads on ordinary piping supplied by the sprinkler system in the adjacent room.d (77-9-4) the pendant heads are teo far from the roof, upright heads are to be installe.
The previously reported small hydraulic press has been relocated to a second-story room The pipes are enclosed and the reservoir and pump unit has been relocated to the roof.
in heavy hose. This arrangement is acceptable, Due to cost, an approved storage cabinet for the 6 cartons of four,1-gal jars of j
Due to value acetone (F.P. O'F) has not been provided in the central stores area.
of storage, at least an ordinary storage cabinet should be provided.
3 MVFP' s
i 1868 118 uw b
"Research Center" Index No. 45336.30 The mock-up generator has been enclosed with metal walls, but they do not reach the ceiling.
An automatic sprinkler supplied by the system at the ceiling will be provided beneath it.
Several areas contain cartons of equipment, records, and empty, with the most found in the electronics shop in the southeast end of Building D.
This is a continuing problem and attention will be given to it.
With the new construction completed, most areas are not as congested as previously noted.
S iBa68 II9
Fcrm LRC-138 Research and Development Division LYNCHBURG RESEARCH CENTER babcock & W.lCOX LYNCHBURG, VIRGINlA i
To A. F. OLSEN, LICENSE ADMINISTRATOR N '"
C. E. BELL, MANAGER, FACILITIES File No.
Cust.
or Ref.
Date Subj.
BUILDING CONSTRUCTION DECEf1BER 3,1979 m, i.,+, +..-
a +...a
..bi,
ir.
In response to your request, I have reviewed the criteria used in the construction of the buildings at the Lynchburg Research Center in which NRC licensed material is handled, used or stored.
I have determined that Buildings A, B, C, J, and the Liquid Waste Disposal Facility were con-structed in accordance with the codes and standards of the Commonwealth of Virginia and the County of Campbell, in effect at the time of con-struction.
f br A
C. E. Bell CEB/jhc 1868 120
INSTRUCTION SHEET FOR BAW-381 DEMONSTRATION AND CONDITIONS FOR LICENSE SNM-778 Amendment 0 Revision 3 November,1979 REMOVE OLD PAGE INSERT NEW PAGE Me Rev.
Date Page Rev.
Date 4-5 2
10/79 4-5 2
10/79 4-6 2
10/79 4-6 3
11/79 4-17 1
9/79 4-17 1
9/79 4-18 2
10/79 4-18 3
11/79 4-23 1
9/79 4-23 1
9/79 4-24 2
10/79 4-24 3
11/79 4-41 2
10/79 4-41 2
10/79 4-42 1
9/79 4-42 1
9/79 4-43 1
9/79 4-43 3
11/79 4-44 1
9/79 4-44 1
9/79 4-49 2
10/79 4-49 2
10/79 9/79 4-50 3
11/79 4-50 9/79 4-50a A-7 2
10/79 A-7 3
11/79 A-8 2
10/79 A-8 2
10/79 A-9 2
10/79 A-9 3
11/79 A-10 2
10/79 A-10 3
11/79 A-11 2
10/79 A-11 3
11/79 A-12 2
10/79 A-12 2
10/79 A-13 2
10/79 A-13 2
10/79 A-13a 3
11/79 A-13a 1
9/79 A-13b 3
11/79 A-14 2
10/79 A-14 2
10/79 A-17 2
10/79 A-17 2
10/79 A-18 2
10/79 A-18 3
11/79 A-19 2
10/79 A-19 3
11/79 A-20 2
10/79 A-20 2
10/79 A-21 2
10/79 A-21 3
11/79 A-22 2
10/79 A-21a 3
11/79 A-22 3
11/79 A-23 2
10/79 A-23 3
11/79 A-24 2
10/79 A-24 3
11/79 1868 121
INSTRUCTION SHEET FOR BAW-381 - DEMONSTRATION AND CONDITIONS FOR LICENSE SNM-778 PAGE 2 - November, 1979 REMOVE OLD PAGE INSERT NEW PAGE Page Rev.
Date h
Rev.
Date A-25 2
10/79 A-25 3
11/79 A-26 1
9/79 A-26 3
11/79 A-26a 1
9/79 A-27 2
10/79 A-27 2
10/79 A-28 1
9/79 A-28 3
11/79 A-29 1
9/79 A-29 1
9/79 A-30 1
9/79 A-30 3
11/79 A-31 1
9/79 A-31 3
11/79 A-32 2
10/79 A-32 2
10/79 A-33 2
10/79 A-33 3
11/79 A-34 1
9/79 A-34 1
9/79 A-39 1
9/79 A-39 1
9/79 A-40 1
9/79 A-40 3
11/79 C-33 2
10/79 C-33 3
11/79 C-34 2
10/79 C-34 2
10/79 1868 122
4.3 SAFETY REVIEW COMMITTEE The responsibilities and functions of the Safety Review Committee are given in Appendix D of this document.
4.4 (A.9.5) NUCLEAR SAFETk 4.4.1 (A.9.5.lhdministrative Requirements The ultimate responsibility for nuclear safety rests with the Director. However, first-line responsibility is with the facility supervisor supported by the nuclear safety officer.
It is the facility supervisor's responsibility to see that personnel follow all nuclear safety procedures and to question any doubtful prccedure.
The nuclear safety officer is generally responsible for establishing nuclear safety limits and nuclear safety considerations in operating procedure, processes, and the like. His duties are shown more specifically in the following statement.
The position of uclear safety officer has been established at the Lynchburg Research Center. It will be this officer's responsibility to ensure, as far as possible, that no operations in the Lynchburg Research Center can lead to the inadvertent assembly of a critical mass. To this end he will review all new procedures which involve the handling of special nuclear materials as well as changes in old procedures, observe operations, inaugurate educational programs if and when he deems them necessary, and carry out confirming criticality calculations.
This appointment does not in any way relieve facility supervisors of their responsibilities for ensuring safe operations in their areas, nor will it eliminate the necessity for the reviews by the Safety Review Committee required by the_li, cense.
Once a month the nuclear safety officer or qualified person designated by him will inspect all LRC operations where special nuclear materials are being processed.
Other areas shall be inspected less frequently; however, all areas shall be inspActed'at least twice a year.
He shall consider area License No.
SNM-778 Docket No.70-824 Date October 1979 Page 4-5 Amendment No.
Revision No.
2 1868 123
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Babcock & Wilcox
operations when scheduling these inspections and shall, if necessary, schedule his inspection at more frequent intervals. His consideration should include inspection of new facilities, inspection of hazardous non-routine operations, an audit of nuclear safety records, a check for area posting and a review of current practices.
A written report is to be filed with the Director quarterly with a copy to the LRC license administrator.
Prior to submission of the report, he shall discuss any findings with the appropriate facility supervisor. The report shall be brief, concerning itself with inspections made during the quarter and with the nuclear safety activity of the quarter.
The following information is to be included.
e Areas visited.
Operations observed, e
Unsafe practices or situations noted, e
Nuclear safety activity of the quarter (brief summary).
e e Recommendations.
Resolution of previous recommendations.
e
- 4. 4. 2(A.9. 5.5) Nuclear Isola tion Special nuclear material at the LRC is isolated.from all other special nuclear material for nuclear criticality safety urposes if any of the three conditions (or equivalent) listed in A.9.5.5 are met.These three isolation criteria are accepted industrial practice for maintaining nuclear criticality safety.
It is recognized that 12 inches of high density concrete may not be adequate as isolation between two large parallel slabs of SNM; this does not describe any SNM configuration at the LRC and will not be permitted without additional evaluation and NRC approval.
4.4.3 (A.9.5.6) Building A Note: The Lynchburg Pool Reactor and Critical Experiment Facility are nuclearly isolated from the units specified in this license. A minimum of 5 feet of high density concrete and/or water separates them from these units.
License No.
SNM-778 Docket No.70-824 Date November 1979 4-6 page Amendment No.
Revision No.
3
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Babcock & Wilcox
Cases 2 through 5 represent removal of rods " uniformly" through the assembly while Case 6 represents the removal of eight rods clustered about the center. Rods removed are shown schematically in Figures 4-1 and 4-2 for Mark B and C assemblies, respectively.
The calculations reported above demonstrate the safety of an assembly under varying stages of dismantlement; grouping of 75 fuel rods under water and confined with a 8.6-inch square merely describes a dismantled assembly and is also safe.
The safety of withdrawing an assembly and its associated rod storage position partially into the cell is demonstrated safe by comparison to a series of KENO runs made for pool storage at a reactor site. The KENO physics code (a Monte Carlo code) is described in SNM License No. 1168, Appendix A to Section III, pages 11 and 12.
To demonstrate the safety of flooding a reactor site storage pool filled with fresh Mark B fuel assemblies, an array of fuel assemblies 14 units wide, infinitely long, and reflected on the sides and bottom by concrete was calculated by KENO.
Each assembly was spaced 1 foot from the other on the concrete reflector, as appropriate.
Four cases at different degrees of pool flooding were evaluated and are described in Table 4-6.
Table 4-6.
Reactivity for an Infinite by 14-Unit Array of Fuel Assemblies Water Height eff Fully flooded 0.951 0.006 3/4 0.946 0.007 1/2 0.928 0.007 0 (dry) 0.506 0.004 License No.
SNM-778 Docket No.70-824 Date September 1979 Page 4-17 Amendment No.
Revision No.
1 3,
euuom#
1868 12-Babcock & Wilcox
The confidence levels quoted above are one standard deviation, K
f r the fully flooded condition is higher than that calculated by eff PDQ because of simplifications made in running the cases. The series of runs were to demonstrate safety of a partially flooded pool, a much more restrictive condition than partial withdrawal into one cell. The similarity in the Mark B and '.fark C nuclear characteristics and the simplifying assumptions assr;e these calculations are also vaid for the Mark C assembly type.
4.4.4. i.4 (A.9.5.7.5.4) Assembly and Machine Shop and Development Test Areas Assemblies of either Mark B or C disassembled in air are far less reactive than the cases listed in Table 4-5.
Either assembly type may be disassembled in air. A safe reactivity level (<<0.95) is assured provided I
the handling in Section A.9.5.7._5 4 is followed.The cond.tions stated in Section A.9.5.7.5.'4 are based on KEN 0 calculations that show:
1.
Two assemblies in air 21 inches or more apart are nuclearly safe.
2.
Fuel pins at a maximum enrichment when optimumly moderated are fully reflected in an infinite slab have a Keff = 0.95 if the slab is no more than 4 inches thick.
3.
Fuel rods in any configuration or number, up to the number in the assembly, when limited to the confines of the assembly size are no more reactive than the intact assembly (Ref. Table 4-5).
4.4.4.5.5 (A.9.5.7.5.5) Hot Cell Operations -- Work within the hot cell will, by and large, follow existing controls. One fuel rod will have a maximum of 2,230 grams of uranium or about 90 grams of U-235 at 4.0 wt%
nominal enrichment. Five rods represent, therefore, 450 grams of U-235 -
considerably less than the permitted unit of 850 grams U-235 at 4.0 wt%
enrichment or less. Two work stations are now authorized in Cell No. 1.
1868 126 4-18 License No.
SNM-778 Docket No.70-824 Date November 1979 Page Amendment No.
Revision No.
3 Babcock & Wilcox
4 It is located where, in the judgment of the health physicist, it will provide the most adequate service.
The alarm setting for a plutonium area is about the equivalent of one 40-hour work week at 1 MPC and cannot be set lower dua to randon levels. Alarm levels in other areas are set in accordance with the particular operation, the material being handled, and its potential hazard. The actual levels are then set as law as possible commensurate with local radiation levels.
2.
Portable Continuously Indicating This type of air monitor is placed as close as practicable to a particular operation and usually has the same alarm-level setting as the fixed continuously indicating monitor.
4.5.2.2.2 Air Sampling Program -- Air sampling can be separated into two categories: fixed and portable. Selection of the category is at the discretion of healtn physics personnel.
1.
Fixed A central vacuum system with up to 100 sampling points has been installed in Building C.
The sampling points are located as close as practicable to operator stations to permit continuous sampling of breathing zones. Although these samples 'are usally checked weekly, the frequency may vary as the situation dictates. Samples are usually counted after waiting for randon daughters to decay, but if a particular operation is in question, the samples may be counted after a shorter period, and an appropriate radon decay factor applied.
2.
Portable Where the use of a fixed air sampling is not practical, a portable sampler is used.
This sampler is also used throughout the LRC to monitor special situations.
The samples are checked and evaluated like fixed air samples.
Air samples are counted on a low background proportional counting system. Appropriate factors for background activity and detector efficiency are applied to give results in disintegrations per minute. These results are License No.
SNM-778 Docket No.70-824 Date September 1979 Page 4-23 Amendment No.
Revision No.
1 1868 127
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Babcock & Wilcox
then divided by the toal air flow volume of the system to obtain activity per unit volume (uCi/ml) which can be readily compared to MPC values for particular isotopes.
The true flow rate is determined by the following method:
Initial ~ flow rate + Final flow rate = True Flow Rate 2
4.5.2.3 Liquid Sampling - Liquid sampling consists of liquid waste tank samples, Lynchburg Pool Reactor (LPR) samples, hot cell storage pool, the experimental pool and pool gate samples. These liquid samples are taken at intervals determined by the health physicist.
Liquid waste tanks are sampled quarterly, prior to release to the Naval Nuclear Fuel Division's waste treatment system or whenever an unknown quantity of fissile material has been released to the tanks.
Prior to sampling, the tank contents are thoroughly mixed and sampled.
Two of the waste tanks are treated with strontium carbonate to precipitate radio-strontium and they are mixed, but time is allowed for the SrCo3 to settle before the sample is taken.
The LPR primary water is sampled each week during which the reactor has operated. The primary and secondary coolant is sampled monthly regard-less of operations. These samples are taken by the operations staff.
Sampling intervals of the other pools are dependent on use and are estab-lished by health physics.
Known amounts of sample liquid are extracted, evaporated to dryness and counted. Results obtained are recorded in activity per unit volume.
These records are retained by the health physics group.
4.5.2.4 Environmental Sampling - Environmental sampling of areas sur-rounding the LRC is performed to evaluate changes in levels of radioactivity of air, water and vegetation. As a minimum the program includes:
e One continuous on-site background air sample.
e Monthly water samples of the James River up and down stream of the liquid discharge point.
e Continuous sampling of rain water on site.
e Quarterly samples of river silt and near-river vegetation.
License No.
SNM-778 Docket No.70-824 Date November 1979 Page 4-24 Amendment No.
Revision No.
3 1868 128 Babcock & Wilcox
T 4.5.11 Respiratory Protection Program The health physics group is responsible for the implementation of the respiratory protection progros.
The primary objective of a respiratory protection program is to limit the inhalation of airborne radioactive or other hazardous materials. This objective is normally accomplished by
'.c application of engineering controls, including the use of process, containment. and ventilation equipment. When such controls are nct feasible or cannot be applied, respiratory protective devices must be used.
The program will include the following:
1.
Procedures governing the selection, fitting, and use of respirators.
2.
Procedure for training of users of respiratory protection.
3.
Procedure for respirator decontamination, maintenance, and storage.
4.
Medical surveillance for users of respiratory protection.
5.
Regular inspection and evaluation of the program to determine its continued effectiveness.
4.5.12 Audit The health and safety supervisor or his designee perform audits monthly.
A written report is to be filed with the Director, LRC quarterly with a copy to the License Administrator. The audits ate conducted in accordance with a written plan. An example of the contents of an audit plan is:
1.
Each month the following itens are audited.
a.
Records of shipments and receipts of radioactive material are reviewed for completeness.
b.
LRC-151 " Work Order" forms are reviewed to ensure that appropriate signatures have been entered.
c.
Selected work areas are inspected to ensure proper posting, labeling and storage of radioactive material, and safety practices and procedures are being followed.
License No.
SNM-778 Docket No.70-824 Date October 1979 Page 4-41 Amendment No.
Revision No.
2 1868 129
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Babcock & Wilcox
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2.
During the year the following items are covered:
a.
An evaluation of the respiratory protection program.
b.
An evaluation of employee whole body exposure.
c.
An evaluation of environmental releases.
d.
An evaluation of bioassay results.
e.
An evaluation of air borne radioactivity.
f.
An evaluation of environmental monitoring.
g.
An audit of health physics records.
4.5.13 ALARA The management at the Lynchburg Research Center is committed to the philosophy of maintaining radiation exposure to levels that are as low as reasonably achievable. This commitment is made known to all employees within the first month af ter they report to work.
Employees who's work assignment involve exposure to licensed material are provided with the initial training and annual retraining that reinforces this commitment and provides the employee with the fundamental knowledge necessary to assist in implementing the ALARA principle.
The implementation of the radiation protection program is the responsibilty of the health and safety supervisor. The qualificaltions of this position are presented in Section 2.
These qualifications ensure that the individual holding the position is imminently capable of dealing with potential problems encountered at the LRC.
The supervisor has the authority to suspend operations until corrective action is taken if he observes practices which could result in a definite hazard.
The health and safety supervisor is responsible for reviewing personnel exposures to assure that ALARA principles are being applied.
Table 4-8 is a summary of whole body radiation exposure for the calendar year 1977 and 1978.
It can be seen by this data that there is an increase in exposure dose in 1978 as compared with 1977. This is an example of a trend in exposures that resulted from an increase in the level of work in licensed activities rather than a reduction in the application of I
License No.
SNM-778 Docket No.70-824 Date September 1979 Page 4-42 Amendment No.
Revision No.
1 186,8 130 Babcock & Wilcox
9 1
ALARA principles. This type of trend is not unusual in research and development where the icvel and types of work will vary from year to year.
In this case the increased exposure is attributable to an increase in hot cell work and support activities from 5.75 manyears of effort in 1977 to 9.75 manyears in 1978 and an increase in critical experiment operations over the same period.
During this period no abnormal occurrences resulted in internal or external exposures.
Air sample data presented in Table 4-9 for plutonium and uranium areas l
historically have averaged less than 1% of MPC with an occasional sample indicating 2-3% of MPC. Bioassays including invivo for alpha emitters historically indicate zero exposure. Occasionally a sample will indicate detectable activity but analyses of subsequent samples have not confirmed an internal deposition.
Air sample data for hot cell work has shown that some air samples including the respiratory protection factor utilize a significant portion of 40 MPC hours per work week.
However, invivo counting indicates the presence of only trace quantities of gamma emitters.
In most cases < 1% of a lung burden is indicated and up to 3% of a lung burden in a few cases.
An example of the application of ALARA principles is the improved method of fuel transfers into and out of the hot cell.
Previously these transfers were made in air through a cell door resultin'g in an exposure of approximately 100 mR per transfer. By performing these transfers under
, water the exposures have been reduced to less than 10 mR per transfer.
Respirator protection equipment for making entries into the hot cell has been filter masks. The installation of a supplied air respiratory system is planned.
Installation will get under way in the near future at a cost of approximately $70,000. This new system will result in an increase in the protection factor to werkers of from 50 for filter masks to 2000 for the supplied air system.
License No.
SNM-778 Docket No.70-824 Date November 1979 Page 4-43 Amendment No.
Revision No.
3 1868 151 Babcock & Wilcox
4.6 FIRE SAFETY 4.6.1 Fire Prevention 4.6.1.1 Housekeeping -- The facility supervisors are responsible for the day-to-day safety of operations at the LRC. This responsibility includes assuring that fire loading in their facility is maintained at the practicable minimum and work and storage areas are kept clean.
4.6.1.2 Area Operating Procedures -- In exercising their review and approval responsibility for area operating procedures, the facility supervisors shall assure that new project and operations involve the least practicable amount of combustible and flassable materials.
In exercising its review and approval responsibility, the Safety Review Committee also reviews new projects and area operating procedures for fire safety.
4.6.1.3 Flammable Liquids 4.6.1.3.1 General -- Flammable liquids for use in the laboratories or work areas are stored in flammabic liquid storage cabinets. Volatile liquids may be stored in glass or plastic containers outside a finnmnble liquid storage cabinet of not more than 1 liter capacity for laboratory Quantities exceeding 1 liter may be located outside a flammable liquid use.
storage cabinet but only in safety cans, 4.6.1.3.2 Hot Cells and Glove Boxes -- Hot cells and glove boxes are limited to quantities of volatile material which, when vaporized and mixed throughout the volume of the hot cell or glove box, would not result in the accumulation of an explosive mixture. The determination of the maximum permissible concentration is made by calculation.
An example of the calculational method is presented below for Methyl Alcohol in Hot Cell No. 1 License No.
SNM-778 Docket No.70-824 Date September 1979 Page 4-44 Amendment'No.
Revision No.
1 il2
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SNM-778 Docket No.70-824 Date October 1979 Page 4-49 Amendment No.
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Table 4-9 Annual Average Airborne Activity Concentrations (pci/ml)
Area Average Maximum Average Maximum 1977 1977 1978 1978 Building B (MPC = 1 x 10 ')
-14
-1
-10
-14 Radiochemistry Lab 3 x 10 8,6 x 10 1 x 10 7 x 10
-14
-1'
-14 Hot Cell Control Room 1 x 10 7 x 10-1 x 10 3 x 10
-1
-13
-12 Cask Handling Area 6 x 10 3.6 x 10-5 x 10 7 x 10 Building C (MPC = 1 x 10-10)
-15
-13
-15
-14 Lab 15 1 x 10 1 x 10 1 x 10 3 x 10
-15
-13
-15
-13 16 1 x 10 2 x 10 1.x 10 1.8 x 10
-15
-13
-15
-14 17 1 x 10 2.7 x 10 1 x 10 3 x 10
-15
-15
-15
-15 18 1 x 10 9 x 10 1 x 10 4 x 10
-15
-14
-15
-14 19 1 x 10 4 x 10 1 x 10 2 x 10
-15
-13
-15
-14 20 1 x 10 1.2 x 10 1 x 10 2 x 10
-15
-15
-15
-14 22 1 x 10 7 x 10 1 x 10 1 x 10
-15
-14
-15
-13 27 1 x 10 1 x 10 1 x 10 2.3 x 10
-15
-15
-15
-14 44 1 x 10 8 x 10 1 x 10 1.4 x 10
-15
-14
-15
-14 50 1 x 10 1 x 10 1 x 10 4 x 10 License No.
SNM-778 Docket No.70-824 Date November 1979 Page 4-50 Amendment.No.
Revision No.
3 1868 134 Babcock & Wilcox
Figure 4-1.
Fuel Rod Removal Schematic - Mark B Case 2 Case 3 4 rods removed 12 rods removed Q
- K
. (
g
_ X
.__g X
X O
O X
O O
I O
X O
O o
l I
Case 4 Case 5 24 rods removed 36 rods removed Q
X_
.._X_.
g
- -<X-- - -f X
X X
i X
O O
X O
O i
X
' X O
X
! X O
X O
O X
X X
X X
a
\\1.ocation of removed fuel rods Case 6 8 rods removed Instrument Tube (1 per assembly)
M
- 4.
XX O
O coacrat aad cuid tub-(16 per assembly)
O O
l Rod 4
t, License No.
SNM-778 Docket No.70-824 Date September 1979 Page 4-50a Amendment No.
Revision No.
1868 135 Babcock & Wilcox
control of activities and for the safe handling of licensed materials within the areas of their supervision. To assist these individuals, the Director will appoint the health and safety supervisor, the nuclear safety officer, and the nuclear materials manager (NMI). All of these, including facility supervisors, are considered key positions.
A. 5. 2 Position Requirements The minimum requirements for these key positions are given in the following paragraphs.
A.5.2.1 Laboratory Manager - He shall be appointed in accordance with Company policy.
A.5.2.2 Section Manager - The section managers shall have a BS degree and three years post graduate work or equivalent training in the pertinent technical field.
They are responsible for the safe performance of projects under their purview. Section Managers shall have demonstrated knowledge in the application of radiation and nuclear criticcliev safety associated with i
their projects.
A.S. 2. 3 Facility Supervisor - He must demonstrate to Company manage-ment proficiency in the application of good principles of radiation protection, industrial safety, and nuclear safety as related to the activities expected in his area of responsibility. He must have a minimum of three years' related experience and either hold a degree in his related work or five years' additional experience in the use and handling of licensed material.
The facility supervisor shall be responsible for reviewing and approving area operating procedures to ensure that amoung other things that the concerns of the industrial safety officer are incorporated, and terminating any operation he deems contrary to license conditions, area operating procedures, or general safety conditions.
License No.
SNM-778 Docket No.70-824 Date November 1979 Page A-7 Amendment No.
Revision No.
3 I
1868 136 mc Babcock & Wilcox
l A.S.2.4 Health and Safety Supervisor - He must have a BS degree in a technical field and professional experience in assignments involving radiation protection at a supervisory level. He must have five years' experience and demonstrate proficiency in the application of radiation safety principles and be knowledgeable in fields related to radiation protection.
A.5.2.5 Nuclear Safety Officer - He must be technically trained and have a BS degree in science or engineering. He must have either two years' experience with nuclear criticality safety calculations similar to those associated with LRC activities, or he must have one year's experience with nuclear criticality safety calculations similar to those associated with LRC activities if he has at least an additional two years' experience in nuclear reactor physics calculations. All nuclear criticality safety cal-culations that require a second party reviewer shall be reviewed by an individual whose qualifications are the same as those required above for the nuclear safety officer.
A. 5. 2. 6 Nuclear Materials Manager - He must demonstrate to Company management his knowledge of the principles necessary for the accountability and safeguarding of special nuclear material.
A.S.2.7 License Administrator - He shall have a BS degree in science or engineering and three years' experience in nuclear technology or an AS degree in. science or nuclear technology and 12 years' experience in nuclear technology.
License No.
SNM-778 Docket No.
7_0-824 Date October 1979 Page A-8 Amendment No.
Revision No.
2 1868 137 Babcock & Wilcox
4 A.S.2.8 Industrial Safety officer - The industrial safety officer shall have at least one year's experience in radiation and industrial safety He shall be familiar with the codes and requirements of the Occupational Health and Safety Act of 1970 and the National Fire Protection Association. The Bab-cock & Wilcox Company's fire protection engineer is available to the industrial safety officer for consultation in fire protection when the need for additional expertise is required.
The industrial safety officer is responsible for admini-stration of the industrial safety program, conducting, at least bimonthly, in-spectionsiof the facilities to insure that industrial safety and fire protection requirements are being met, and revieiving projects and procedures for indus-trial safety and fire protection considerations.
A.S.2.9 Health Physics Engineer - The health physics engineer shall have a BS degree which shall include at least 20 quarter hours of health physics related course work or the equivalent in work experience. The health physics engineer shall be responsible for the administration of the radiological and environmental monitoring programs, the respiratory protection program, the radioactive waste disposal program and routine monitoring for radiological protection.
A.6 ADMINISTRATIVE CONTROL Procedures control and provide the necessary direction to coordinate and ensure the safe operation of activities at the LRC.
In addition to formal procedures, numerous routine activities are conducted in accordance with established practices developed through years of operational experience.
Many of these have never been produced in written form since the basis for ensuring safe operation rests with the experienced scientist, engineer, and technician. Work involving the use of licensed material that is not covered by an approved written procedure shall be performed pursuant to a Radiation Work Permit (RWP).
RWP's shall be reviewed and updated quarterly.
License No.
SNM-778 Docket No.70-824 Date November 1979 Page A-9 Amendment No.
Revision No.
3 1858 138 Babcoch Wilcox
The LRC procedures are defined as folloss.
A. 6.1
,Semonstration and Conditions for License SNM-773 This document presents the procedures necessary to show compliance with the license conditions.
1.
Additions or changes to the procedures in the demonstration sections of
" Demonstration and Conditions for License SNM-778" are submitted to the LRC license administrator, who, with the concurrence of the Director, laboratory managers and facility supervisor, aay implement these changes or additions. Prior to implementation, all changes or additions are referred to the Safety Review Committee.
s.
A. 6. 2 Area Operating Procedures Area operating procedures are' facility procedures for the safe handling of licensed material.
1.
Additions or changes to area operating procedures arc sub-mitted to the representive facility supervisor.
The facility supervisor forwards the submittal to the supervisor, health and safety, and nuclear eafety officer for review and approval.
The health physicist and nuclear safety officer shall be independent of the operation under review. Subsequent to these approvals he shall review and approve the addition or change. The Facility Supervisors shall consider the recommendations of the Industrial Safety Officer.
i 2.
All area operating procedures and revisions thereto shall be approved by the Safety Review Co=mittee except:
Revisions correcting spelling or typographical errors.
a.
b.
Revisions in the procedure form or format.
Procedures and revisons that govern the use of equipment c.
that does not have an effect on the use or handling of licensed material.
3.
Area operating procedures shall be followed and shall be available in each operations area.
License No.
SNM-778 Docket No.70-824 Date November 1979 Page A-10 Amendment No.
Revision No.
3 186o8 139 Babcock & Wilcox
4.
Area operating procedure manuals shall be issued to specified individuals. New and revised procedures are distributed in accordance with a document control system which assures that the manuals contain only the most currently. approved procedures.
Facility Suoervisors shall review the crocedures for their facility annually to insure that procedures are up to date and applicable.
i 5.
General health physics and nuclear criticality safety procedures shall be established, maintained and followed for all operations involving the processing, handling and storage of licensed material.
A.6.3 License SNM-778 Requests for changes to License SNM-778 are made to the license administrator, who, with the concurrence of the Director, the laboratory managers, and the facility supervisors, will forward the request to the U.S. Nuclear Regulatory Commission. Requests for license amendments will also be referred to the Safety Review Committee.
A.6.4 Safety Audits A.6.4.1 Nuclear criticality safety and radiation safety audits shall be per-formed monthly by the nuclear safety efficer and the supervisor, health and safety respectively, or similarly qualified persons designated by them.
"The audits shall be conducted in accordance with a written plan and documented.
t A written report of the audits shall be filedquarterly with the Director, LRC and with a copy of the License Administrator including the areas in-spected, findings, and status of action taken to correct previous findings.
A.6.4.2 The nuclear safety officer and the supervisor, health and safety shall have the authority to suspend operations until action is taken to correct observed practices which could result in a definite hazard.
License No.
SNM-778 Docket No.70-824 Date November 1979 Page A-ll Ame.dment No.
Revision No.
3 1660 140 Babcock & Wilcox
A.6.5 Radiation Work Permit Radiation Work Permits (RWP) shall be prepared for operations or maintenance work not covered by an area operating procedure and which in-volves entry into a system containing SNM or where 3 potential for release of contamination exists such that the airborne radioactivity concentration to which employees are exposed from the proposed operation or work is likely to exceed tha concentration in Appendix B, Table 1, of 10 CFR 20 or the
~
potential external radiation exposure to which employees are exposed from the proposed operation or work is likely to exceed 100 mR in one work week.
The RWPs shall specify the necessary radiation safety controls including but not limited to respiratory protection, special air sampling, and special local ventilation.
All RWPs shall be signed by representatives of the appropriate line supervision and the responsible representative of health physics prior to the start of operations except that during off-shift hours approval of the supervisor of health and safety or uls designee may be obtained via telephone.
A.7 SAFETY REVIEW COMMITTEE A.7.1 Membership The Safety Review Committee (SRC) shall be comprised of at least five technical members having experti'2 in chemistry, radiological catety, industrial safety, and nuclear engineering. The members shall be appointed by the Director, who shall also appoint the chairman, alternate chairman, and the members who serve on the Audit Subcommittee. The Director shall appoint a management representative to be a committee member who shall serve as the committee coordinator.
License No.
SKH-778 Docket No.70-824 Date October 1979 Page A-12 Amendment No.
Revision No.
2 1868 141 Babcock & Wilcox
A.7.2 Function The SRC reports to the Director.
It reviews and approves new Projects and major changes to existing projects that utilize licensed materials. It reviews and approves all area operating procedures pursuant to Section A. 6.2.2.
The Safety Audit Subcommittee (SAS) shall perform audits of LRC operations to cssure compliance with safety requirements.
~'
It shall be the responsibility of the license administrator to act on the recommendations o? the safety audit subcommittee and confirm the acticns taken.
A.7.3 Frequency of Meetings The SRC shall meet at least four times annually. The SAS shall conduct audits of the LRC at least three times annually.
A.7.4 Documentation Minutes of the SRC actions shall be submitted to the Director, LRC, and the permanent records of the SRC shall be kept by the committee coordinator. A report of the SAS's audits showing the items audited, findings and disposition of action taken on previous findings shall be sent to the SRC chairman who shall forward the report to the Director, LRC, with comments as appropriate.
A.7.5 Annual Report An annual report by the supervisor, health and safety shall be made to and reviewed by the SRC reviewing employee exposures and effluent release data to determine:
- 1. If there are any upward trends developing in personnel exposures for identifiable categories of workers or types of operations or effluent releases.
- 2. If exposures and releases might be lowered in accordance with the concept of as low as reasonable achievabic.
- 3. If equipment for effluent and exposure control is being properly used, maintained, and inspected.
License No.
SNM-778 Docket No.70-824 Date October 1979 Page A-13 Amendment No.
Revision No.
2 A
i868i I42 Babcock & Wilcox
The report shall include a review of audit and inspections performed by Supervisor, Health and Safety during the previous 12 months and review data from the following areas:
- 1. Employee exposures
- 2. Bioassay results
- 3. In-plant airborne radioactivity
- 4. Environmental monitoring
.. The Satecy Review Committee recoccendations on the annual report shall be cade to the Director, LRC subsequent to completing its review of the report.
10rn iUVO l4' November 1979 License No.
SNM-778 Docket No.70-824 Date page A-13a Amendment No.
Revision No.
3 J
Babcock & Wilcox
A.8 FACILITIES Operations within the cross-hatched areas in Figures A-1 and A-2 are licensed under the provisions of 10 CFR 50 for the operation of nuclear reactors and critical experiment facilities are not covered by License SNM-778 except for operations with sources, byproduct, and special nuclear material non-reactor related.
Records shall be maintained to control the quantity and identity of licensed material possessed within the cross-hatched areas pursuant to License SNM-778.
All facilities used in the handling of source, byproduct, and special nuclear materials will be maintained in a safe condition and will meet the minimum requiremests stated in the conditions of this license. The facilities to be maintained at this site include but will not be limited to the following.
- 1. Hot Cells.
- 2. Storage facilities for radioactive materials.
- 3. Underwater transfer facilities.
- 4. Fume hoods.
- 5. Glove boxes, including open-port or open-face designs.
- 6. Waste handling facilities.
- 7. Laboratories and machine shops.
The design features and the specifications of these facilities are covered in sections A.8.1 and A.8.2 of this appendix.
Emergency power sources with both automatic and manual starters are provided. As a minimum, these power sources provide pour to emergency lighting, the stack exhaust system fans, hot ec11 exhaust fans, the stack radiation monitoring system, criticality monitoring system, and evacuation alarm. Emergency power sources shall be tested weekly in accordance with written procedures which shall specify the acceptance criteria. The test shall ensure that if normal site power is interrupted the emergency system senses the interrupticn and supplies power to the vital loads.
The minimum frequency for testing the evacuation alarm shall be biweekly.
License No.
SNM-778 Docket No.70-824 Date Novetrber 1979 Page A-13b Amendment No.
Revision No.
3 4'
1868 144 BabcockiWilcox
(
I The minimum frequency for checking the pressure drop.across filters, direction of air flow in working areas, or at entr,ances to' hoods and hood face velocitic's shall be in accordance with.the following schedule.-
- 1. Pressure drop across filters - weekly.
- 2. Air flow directions - monthly.
- 3. Hoo4 face velocities -- monthly.
- 4. REPA filters shall be replaced prior to,the differential pressure across the filter exceeding the limit of 4 inches of water.
Final HEPA filters shall be. tested in ac'cordaInce with Regulatory Guide 3.2 " Efficiency Testing of Air Cleaning Systems Containing Devicas
.for Removal of Partic,les," dated August 1973.
A.8.1 Hot Cells Ventilation in the hot cell operations area provides air flow toward the increasingly hazardous areas. The operations area (workJng area outside of hot cells 1.and 2) is maintained ut the highest relative pressure of any operations area..
The area of lowest pressure is within the cells. A pres-sure differential (delta-P) of at least 0.25. inch of dater ac.ross the cell face shall be mainta'ined.
- A velogity of 100 ifm shall be maintained thr' ugh any opened door by o
.' dither of two fans, which automatically or manually energize ~when the delta-P falls below Or25 inch of water. A sufficient volume of air is circulated through this facility to maintain an ambient temperature below 120 F.
Two stages of.HEPA filters in series and single stage fire-resistant prefilters are used to filter hot cell exhaust air. The prefilters are located within. each hot cell and the 'HEPA filters are* located above the hot cell on the roof slab of the isolation area. All connecting ducting
~
and dampers are constru'cted of steel. The system incorporates a standard automatic fire damper that will interrupt thd air fl'ow immediately upstream of the HEPA filter in the event that the temperature in the exhaust ducting License.No. SNM-778 Cocket No.70-824 Date October 1979 Page A-14 Amendment'No.
Revision No.
'2 4
1868 i4.5
~
Babcock & Wilcox
4.
Pumping.c pouring through a service line.
5.
Pouring into a container inside the pass-in box.
6.
Removing the glove-box face inside a " tent".
7.
Sending in from the pass-through tunnel between glove boxes.
Current methods for removing materials from the glove boxes include the following:
1.
Bagging procedures.
2.
Pumping or pouring through service line.
3.
Transferring to a clean container inside th*e pass-in box.
4.
Removing the glove box face inside a " tent".
5.
Sending out into the pass-through tunnel between glove boxes.
The following types of containers will be used for transferring plutonium between glove boxes: Tygon tubing, metal tubing, metal liner, plastic bottles, metal cans, plastic bags, and others approved in accordance with the preceding paragraph.
Plutonium-bearing waste material shall be stored in plastic bhgs in a DOT container approved for off-site shipment of radioactive material.
Plutonium not in solution or in glove boxes shall be stored in prescribed storage areas in double-bagged plastic or metal containers.
Solutions shall be stored in the vault in vented plastic bottles.
Small quantities of plutonium-bearing solutions shall be stored in uavented plastic bottles according to special procedures approved by the SRC.
A. 8. 2. 5 Open-Port Glove Boxes and Fume Hoods Used for Plutonium or Transuranium Elements - The open-port glove box and fume hood operations are limited as follows:
1.
Plutonium powders are limited to less than 50 micrograms of plutonium.
2.
Plutonium metal and solutions are limited to 200 milligrams of plutonium.
3.
The maxinum permissible contamination level for the interior 2
of open-port glove boxes is 10,000 dpm/100 cm,
License No.
SNM-778 Docket No.70-824 Date October 1979 Page A-17 Amendment No.
Revision No.
2 1868 146 Babcock & Wilcox
Detailed opetating procedures for these boxes and hoods will be developed
- ~
by operations personnel and approved by the SRC. All boxes sha21 have an air velocity of at least 100 ifm through the open face.
m-A.9 SAFETY A.9.1 Emergency Plan Appendix C is the emergency plan for the LRC.
A.9.2 Radiation Safety Fat ility supervisors will approve written area operating procedures encompassing radiation safety developed by the line organization. A health physicist, appointed by the Director, shall assist the facility supervisors in this responsibility.
A.9.2.1 Training - The radiation safety program is administered by the supervisor of health and safety. The contents of the courses are described in Section 4.5.8.
Program I shall be presented to each new employee within 30 days of reporting to work. Personnel shall not be permitted to work with licensed material unsupervised until they are authorized and trained in radiation and nuclear criticality safety. Programs II and III are presented i
to employees who are selected by their section manager to be designated as an authorized user of radioactive material (i.e., employees who may handle licensed material unsupervised, health physics technicians and emergency team monitors) on an as-needed basis.
Retraining of authorized users of radioactive material is performed annually. Attendance and the identify of the inetructor shall be docu-mented. The effectiveness of the training program shall be evaluated on the basis of a written examination and documented.
A.9.2.2
_ Personnel Protection - Personnel monitors (film badges, dosimeters, or other suitable devices) are provided to measure the radiation exposure of visitors and employees. Personnel dosimeters issued pursuant to 10 CFR 20.202 shall be read on a monthly basis.
The use of respiratory protective equipment shall be in accordance with 10 CFR 20.103.
License No.
Smi-778 Docket No.70-824 Date November 1979 Page A-18 Amendment No.
Revision No.
3 Babcock R6hiled]
The employee's line supervisor shall be responsible for keeping exposures below 300 millirems per week and 1250 millirema per quarter.
The supervisor of health and safety may approve weekly exposures above 300 millirems, but the quarterly limit of 1250 millirems shall not be exceeded without the approval of the Director.
If an employee has received the quarterly limit and the Director has not authorized exceeding the limit, the employee's work shall be restricted to prevent further exposure for the remainder of the quarter.
A.9.2.3 Bio-Assay Program A.9.2.3.1 Uranium Bio-Assay Program - Health Physics shall administer the bio-assay program for uranium. The sampling frequency shall comply with Tables 2 and 3 of Regulatory Guide 8.11. " Application of Bioassay for Uranium",
dated June, 1974. The program shall follow Regulatory Guide 8.11 except as follows:
1 When an employee is not at the LRC during a period when the bioassay counting service is on site, a special counting (make-up) shall not be performed for routine exposure control monitoring.
2.
Bioassay samples (urinanalyses and in-vivo lung counting) shall be analyzed for plutonium and uranium if the sample involves an employee working in an area where both plutonium mad uranium may be present in air. However, the health and safety supervisor may decide to analyze for Pu only if Pu and U are mixed and it can be shown that analyzing for Pu would be more sensitive.
License No.
. november 1979 Page A-19 Amendment No.
Revision No.
3 J
1868 148~
Babcock & Wilcox
A.9.2.3.2 Plutonium Bio-Assay Program - Health Physics shall administer the bio-assey program for plutonium. All personnel who routinely work in plu-tonium handling areas shall be subject to the plutonium bio-assay program. The minimum frequency for urine sampling shall be six months. The action levels and actions to be taken are presented in Table A-1.
The minimum frequency for invivo counting (lungs) shall be annual. The action levels and actions to be taken are presented in Table A-1.
More frequent analyses may be performed if other data (air samples, floor and clothing contamination) indicates that workers are being exposed. The Supervisor, Health and Safety shall determine this increase in frequency.
A.9.2.4 Air Sampling -- Air samples shall be taken in all areas of the LRC in which operations are conducted that might cause personnel to be exposed to airborne radioactive materials.
Any of these areas in which the concentration of airborne radioactive material exceeds 25% of the applicable MFC shall be continuously monitored for as long as the process that causes the airborne activity is in progress.
Permanently mounted air sampling equipment normally used to determina con-centrations in a worker's breathing zone shall be evaluated for representa-tiveness at least once every 12 months and whenever any licensed process or equipment changes are made. In addition, the location of air samplers shall be checked out at the beginning of operations in en area that has been shut down for more than 6 months to verify the representativeness of the air sampling.
License No.
SNM-778 Docket No.70-824 Date October 1979 Page A-20 Amendment No.
Revision No.
2
- 1868, 149 Babcock & Wilcox
Table A-1 Plutonium Bio-Assay Action Criteria (W and Y compound exposure)
Analysis Action Action to be Taken Level (a) Urinalysis
< 0.2 dpm/L None Urinalysis
> 0.2 dpm/L 1.
Resample the individual within 5 working days.
2.
Verify if area surveys support the analysis results.
3.
If #2 verifies the analysis, investi-gate the cause and correct.
4.
If exposure is confirmed by #1. in-vestigate to determine how the exposure was incurred, and correct.
If exposure exceeds 50% of the maximum permissible annual dose, the worker shall bd restricted from further exposure until the Supervisor, Health and Safety authorizes the lifting of this restriction.
Jrinalysis
> 4 dpm/L 1.
Restrict the individual from further Pu work.
2.
Resample the individual within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 3.
Initiate investigation.
4.
The Supervisor, Health and Safety only, may lift the work restriction.
(b)
Invivo
< l.6 X 10- Ci None
> 1.6 X 10- Ci 1.
Restrict the worker from further ex-posure.
2.
Resample the individual within 10 workin days.
3.
Determine if area surveys support the analysis results.
4.
If #3 vetifies the analysis, investigatt the cause and correct.
License No.
SNM-778 Docket No.70-824 Date November 1979 Page
.i-21 Amendment No.
Revision No.
3 1868 150 Babcock & Wilcox
5.
If exposure is confirmed by #2, the Supervisor, Health and Safety shall determine the organ dose.
If the confirmed exposure exceeds 50% of the maximum permissible annual dose, the worker shall be restricted from further exposure until the Supervisor, Haalth and Safety authorizes the lifting of this restriction.
6.
The restruction in #1 may be lifted by the Supcrvisor, Health and Safety if the results of analysis performed under
- 2 fails to confirm the analysis.
License No.
SNM-778 Docket No.70-824 Date November 1979 Page A-21a Amendment No.
Revision No.
3 1868 151 Babcock & Wilcox
Permanently mounted air samples used to determine concentration in a worker's breathing zone shall be changed and the filters counted according to the following schedule.
- 1. Process areas during normal operations - once/ shift.
- 2. All areas during periods when normal operations are shut down (maximum interval) -- 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
3.
Samples that are less than 10% of the MPC weekly.
If any one sample exceeds 25% of the MFC one sample change per shift shall be required until the situation is corrected and at least one month's samples are less than 10% of the MPC.
The minimum sample flow rate for permanently mounted air samplers shall be 8 Lpm, except in the rare case when a_ personnel air sampler is used as a temporary replacement for a permanently mounted air sampler, in which case 1.8 Lpm minimum flow rate is permitted.
A.9.2.5 Detection and Survey Equipment -- The performance specifica-tions and sensitivities of the radiation detection and survey equipment used at the LRC are listed in Table A-2.
The calibration interval for portable survey instruments and criticality monitors shall be at least semiannual.
In sampling emissions at low emission rates, the probability of missing particles is great (inversely the probability of collecting particles is low). With isokinetic sampling there is a direct flow ratio of 50,000 cfm, the stack.. flow rate, to a minimum of 2 cfm, the sampler flow rate. Therefore, there is a one in 25,000 chance of collecting particles.
The stack monitor shall be a minimum flow rate of 2 cfm.
} Obb3 I)
License No.
SNM-778 Docket No.70-824 Date November 1979 Page A-22 Amendment No. _
Revision No.
3 Babcock & Wilcox
Table A-2.
LRC Detection Equipment Instruments
_ Sensitivity Range IJindow Thickness 2
Low-range GM (4)
- Beta, gamma Background, mR/h 30 mg/cm 2
Intermediate-range Beta, gamma mR/h, R/h Beta - 1 mg/cm 2
ion chambers (4)
Gamma - 1 g/cm 2
High-range ion (2)
Gamma Up to 500 R/h
>100 mg/cm chambers 2
Proportional (3)
Alpha, beta Backgtound, 1 mg/cm counter 500,000 counts /
min Proportional (1)
Neutron, fast Background, N/A counter and thermal 5000 mR/h 2
Counting room Alpha, beta Background,
<1 mg/cm proportional (1) 100,000 counts /
counter min Counting room gamma Gamma Background up N/A spectrometry, sodium (1) iodide and Ge(Li) 2 Air particulate Alpha, beta Background up
<1 mg/cm monitors (4)
Stack monitoring
- 1) Alpha & beta Background to 106 <1 mg/cm2 equipment (1) particulate counts / min 5
2
- 2) Gaseous beta, Background to 10 30 mg/cm gamma counts / min Personal monitors **
Gamma, beta, 30 mR to 500 R Variable neutron (special cases only)
Pocket dosimeters *** Gamma 0 to 200 mR N/A Direct-reading pocket Gamma 0 to.200 mR N/A dosimeters (12)
O to 500 mR N/A Plutonium wound Gamma 0.004 pCi Thin monitor (1)
Portable air samples Air particulate N/A N/A (6) collection only
- Number of instruments available for use.
- One for each radiation worker.
- Two for each radiation worker. May be substituted with a TLD if it is read at least month.
License No.
SNM-778 Docket No.70-824 Date November 1979 Page A-23 Amendment No.
Revision No.
3 1868 153 Babcock & Wilcox
If the entire particle is radioactive, the activity of the particles varies with the cube of diameter of the particle. However, usually a radio-active particle attaches itself to a nearby non-radioactive dust particle (approximately 1 micron in diameter) which is much larger physically than the radioactive particle. As shown above, if one of these particles gets by the absolute filter, its chances of being detected are only one in 25,000.
In the case of our plutonium handling area, each glove box has an absolute filter at each box exhaust. These filters are so effective that af ter fears of glove bos operations there is no detectable activity in the main off-gas line before coming to the final absolute filters. This is a very strong indication that the plutonium particles are not gecting by the absolute filters.
A. 9. 2. 6 Area Surveys - Health physics shall administer an area survey program that includes smear, air, and environmental sampling.
A.9.2.6.1 Smear Surveys - Health Physics shall perform smear surveys at the below indicated frequencies in the areas listed.
Health Physics shall determine and direct the action to be taken to protect personnel and reduce the levels of contamination below those specified. Decontamination to reduce levels of contamination shall be performed as soon as practicable after discovery.
License No.
SNM-778 Docket No.70-824 Date November 1979 Page A-24 Amendment No.
Revision No.
3 1868'154 Babcock & Wilcox
Area Frequency Action Level
-(DPM/100cm2)
Alpha Unirradiated, Unencapsulated fuel handling areas Weekly 5000 Building B counting Laboratory Monthly 200 Building A Laboratories Monthly 200 Hot Cell Operations Monthly 200 Area Scanning electron Microscopy Lab.
Monthly 200 Exit portals from controlled areas Twice Weekly 200 Beta -
Building A Laboratories Monthly 2000 Building B Counting Monthly 2000 Laboratory Scanning Electron Monthly 2000 Microscopy Lab Hot Cell Operations Area Twice Monthly 2000 Cask Handling Area Twice Monthly 22000 Radiochemistry Lab Twice Monthly 22000 Exit portals from Controlled Twice Weekly 2000 Areas License No.
SNM-778 Docket No.70-824 Date November 1979 Page A-25 Amendment No.
Revision No.
3 1868 155 Babcock & Wilcox
Large area smears are used to survey many square meters of surface area.
To deterttine if these smears indicates that an action level has been exceeded, the assumed area covered shall not exceed im2, l
Fixed contamination that, in the opinion of the Supervisor, Health and Safety, does not substantially contribute to a worker's exposure, shall be posted and its location and radiation level recorded in the Health Physics log book and its removal shall be scheduled as soon as practicable.
Fixed contamination that, in the opinion of the Supervisor, Health and Safety, may substantially contribute to a workers exposure shall be posted and removed as soon as practicable.
A.9.2.6.2 Environmental Surveys - Environmental surveys shall con-sist of river water samples taken monthly below the B&W outfall and mud samples taken semiannually below the B&W outfall.
A.9.2.6.3 Air Surveys - Reference A.9.2.4 License No.
SNM-778 Docket No.70-824 Date November 1979 Page A-26 Amendment No.
Revision No.
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A.9.3 Effluent Control A.9.3.1 Air Monitoring 1.
Potentially contaminated exhaust air from hoods, hot cells, and glove boxes shall be exhausted through at least one set of absolute filters and shall be dir.;harged through a single stack which is continuously monito.ed for particulates and gaseous activity.
The activity level of these effluents is continuously displayed and recorded in Building B health physics area. The monitors are capable of detecting radioactivity at or above the allowed release rate and are equipped with audible and visible alarms which are actuated in the Building B health physics area.
Exhaust systems, that cannot be practically discharged to the atmosphere through the 30 meter stack, and where there exists a reasonable probability that the discharges to the atmosphere could exceed 10% of the applicable MPC for an unrestricted area, will be monitored by taking a continuous particulate sample.
- 2. ~ The stack monitor minimum flow rate shall be 2 cfm.
l 3.
Exhaust air from areas in which there is no airborne radioactive material may be exhausted directly to the roof either with or without continuous sampling if approved by the SRC.
4.
Room air in areas equipped with room air monitors may be exhausted to the roof through absolute filters if the con-centration of airborne radioactive material is below the appropriate MPC for an unrestricted area if approved by the SRC.
5.
Exhaust systems that cannot be practically discharged to the atmosphere through the 50 meter stack, and where there exists a reasonable probability that the discharges to the atmosphere could exceed 10% of the applicable MPC for an unrestricted area, will be monitored by taking a continuous particulate sample.
6.
A dilution factor for discharge of radioactive material through l
a 50 meter stack as calculated by Gifford's solution to the diffusion equation is used. Pasquill's turbulance Type D will be used to calculate an average dilution factor for the year and Pasquill's turbulance Type C will be used to calculate the l
average dilution factor during a planned released during a favorable meteorological period.
7.
The stack blower shall be provided with an emergency back-up l
Lystem for use in the event of an a-c power failure.
I License No.
SNM-778 Docket No.3-824 Date September 1979 Page A-26 a Amendment No.
Revision No.
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8.
The stack samplers shall be operated isokinetically and con-tinuously, except for short periods to permit equipment repair or calibration, to assure a representative sample.
l A.9.3.2 Action Levels - Table A-3 presents the action levels for airborne releases. Health physics shall respond to each alarm, monitoring the operation producing the release and shall, if necessary, shut down the operation to prevent exceeding the release limit. When an action level speci-fled in Table A-3 is exceeded for four consecutive time periods, the operation producing the release shall be shutdown until the cause is corrected.
I i
Table A-3. Stack Release Limits and Action Levels Release Product Release Limit Action Level Beta particulate 3 mci /yr.
200pci/ week Alpha particulate 20pci/yr.
Ipci/2 weeks (long lived) 3 Kr-85 3500 Ci/yr.
70C1/ week H-3 130 Ci/yr.
3Ci/ week I-131 6 mci /yr. or 200pci/ week 300pci/ week License No.
SNM-778 Docket No.70-824 Date October 1979 Page A-27 Amendment No.
Revision No.
2 1868 158 Babcock & Wilcox
A.9.3.3 Liquid Waste 1.
Low-level radioactive liquid wastes are collected in under-ground tanks. Wastes are discharged to the liquid waste treatment plant at B&W's Naval Nuclear Fuels Divisien (NNFD).
Prior to discharging wastes to the NNFD the solution shall be mixed and sampled. The sample shall be representative of the liquid to be discharged. The ratio of the sum of the composite materials shall not exceed 25% of the MPC values of Table I. Col,2, of 10 CFR20, Appendix B, for release to the NNFD waste treatrent system. The limiting values in water shall be determined in accor-dance with the note at the end of Appendix B, 10 CFR 20.
~
Process liquid wastes are collected and stored indoors. These liquid wastes are solidified and handled as dry waste prior to shipment for off-site disposal.
2.
Concentrations of the mixtures of radioisotopes will be below those defined in Appendix B, 10 CFR 20, averaged over one year.
Measurements are taken at or before the point of discharge to the_ James River.
3.
Storage tanks are inspected visually upon each tank voiding to ensure that there are no unsafe accumulations of special nuclear material.
4.
The fissile content of the vaste tanks shall be less than 0.01 gm/ liter. The tanks shall be analyzed (a) quarterly, (b) whenever emptied, (c) whenever an unknown quantity of fissile material has been released to the tanks.
5.
A 10,000 sq ft pond collects storm drainage from Buildings B and C.
This pond shall be sampled quarterly for activity.
A. 9. 3. 4 Dry Wastes - Radioactive dry wastes containing fissile material or high to intermediate-level wastes requiring shielding to meet the require-ments of 10 CFR 20 are stored indoors in DOT-approved containers suitable for off-site shipment and outdoors for short periods of time incidental to transfer or sff-site shipment.
All other dry wastes ('i.e., LSA wastes and fissile exempt waste) may be stored in a locked, interior fenced, paved outside area in DOT-approved shipping containers. These must be in closed containers (usually 55-gallon drums) where weather exposure would not cause any leakage or spread of con-tamination and must be placed on pallets or the equivalent. The storage area is under positive control and surveillance by health physics, and the License No.
SNM-778 Docket No.70-824 Date November 1979 Page A-28 Amendment No.
Revision No.
3 1868 159 Babcock & Wilcox
i area vill receive at least. a quarterly visual and radiological inspect on.
waste is not stored for long periods of time.
This is acequate since A gamma monitor in the waste storage building meets the requirements for a criticality monitor as defined in 10 CFR 70.24.
Each drum with fissile material is gamma scanned on a drum counter to verify the fissile content. The drum monitor has a limit of sensitivity of 2 grams U-235 and 2 grams of Pu.
Each waste container will be labeled so that the contents of each drum can be identified and drums will not be opened outside without specific approval of the LRC health physicist and the facility supervisor.
A.9.3.5 Environmental Sampling -- The environment surrounding the LRC and the Mount Athos plant site is sampled periodically to determine whether the radiation and radioactive material levels in the area surrounding the site have changed or increased as a result of the operations at this location.
The frequency of these surveys will be a function of the work taking place on the site. Environmental samples are collected and analyzed by LRC personnel or qualified commercial concerns. As a minimum, the environmental program will include the following:
1.
One continuous on-site background air sample.
2.
Monthly water sampling of grab samples taken from the James River below the discharge point.
3.
Continous sampling of rain water.
4.
Annual grab sampling of river silt and plant life.
A. 9.4 Special Conditions A. 9. 4.1 Restricted Area - The restricted area at the LRC is defined as any area within a building or enclosed by chain-link fence or a single-strand fence and properly marked with radiation area signs.
License No.
SNM-778 Docket No.70-824 Date september 1979 Page A-29 Amendment No.
Revision No.
1 1868 160 Babcock & Wilcox
A.9.4.2 Records Retention - The following records will be maintained for at least two years.
1.
Records showing the results of surveys made to evaluate the radiation hazard incident to the production, use, release, disposal, or presence of radioactive materials or other sources of radiation, as required by 10 CFR 20.401 (b).
2.
Records used to prepare Form NRC-4 " Occupational External Radiation Exposure History" as required by 10 CFR 20.102.
3.
Records showing receipt, transfer, export, and disposal of by-product material.
4.
Records showing the results of leak tests of sealed sources as required by license.
5.
Records showing the receipt, inventory, and transfer of speical nuclear materials as required by 10 CFR 70.51.
6.
Records showing the receipt, transfer, export, and disposal of source material as required by 10 CFR 40.61 7.
Records of the SRC meetings.
8.
Records of instrument calibration.
9.
Safety audit records.
10.
Records of training and retraining.
11.
Records shall be maintained to reflect the quantities of material in categories A-E and I-Q specified in sec*: ion A.3.
When personnel mon!.toring is required by 10 CFR 20.202, as required by 10 CFR 20.401(a), records showing the radiation exposure of each individual will be retained in accordance with 10 CFR 20.401(c).
A. 9. 4. 3 Area Decontamination - Areas previously used for handling radioactive material (including contaminated glove boxes and equipment) will be removed from service, held in a standby condition, or decontaminated in accordance with current health physics procedures and special procedui as and practices approved by the SRC. These procedures and practices will describe the following activities.
License No.
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Revision No.
3 1868 161 Babcock & Wilcox
1.
The health physics protection and monitoring requirements during removal snd while glove boxes and equipment are in standby condition.
2.
Cleaning procedures.
3.
Package preparation procedures.
4.
Means and methods to prevent the leakage of contamination during package preparation and during storage.
5.
Procedures for removal of off-gas lines and packaging of these and other contaninated equipment.
A.9.4.4 Decommissioning -- At the end of plant life the LRC shall decontaminate the facility and site in accordance with Appendix F, LRC Decom issioning Plan.
A.9.4.5 Water Level Control -- A float-type alarm switch is provided in the storage pool adjacent to the hot cell. If the water level exceeds either the high (3 inches below the overflow) or low limit (29 inches be-low the overflow), an alarm is actuated in the Building B health physics area. A switch shuts off the pool recirculator pump in the event of loss of pool water.
A.9.4.6 Audit of Shipping Procedures -- The health physics group is responsible for auditing all shipping records (monthly) to ensure that pro-cedures are adequate and that the LRC follows these procedures in preparing shipments of radioactive materials. The SAS audits the health physics group periodically.
A.9.4.7 Health Physics Procedures - written and approved health physics shall be established, maintained and followed to include as a minimum procedures for:
- 1. Alarm source check and calibrations
- 2. The conduct of surveys
- 3. Sample counting
- 4. Counting system calibration.
License No.
SKH-778 Docket No.70-824 Date November 1979 Page A-31 Amendment No.
Revision No.
3 s. C 1868 16?~
Babcock & Wilcox
A.9.5 Nuclear Safety A.9.5.1 Administrative Requirements -- Facility supervisors are responsible for the overall administration of the nuclear safety program for their areas. Facility supervisors advise the nuclear, safety officer of their nuclear safety requirements. All postings, training programs, and other written instructions containing nuclear safety rules shall be approved by the facility supervisor. Each facility supervisor maintains a record of the number of units authorized and present in his material balance area.
The nuclear safety officer or person designated by him provides nuclear safety evaluations (including calculational support) as requested or as he deems necessary; these calculations are used to support the license conditions. Nuclear safety evaluations and calculations are reviewed by a qualified individual as defined in A.5.2.4; results of the review are documented and maintained with the nuclear safety evaluations for the same period of time that the evaluations are maintained. He will maintain records of these calculations for at least six months af ter termination of the approved process. The nuclear safety officer performs monthly inspections in areas where SNM are being processed. Other areas are inspected less frequently; however, all areas are inspected at least twice a year. This inspection includes an audit of nuclear safety records, a check for area posting, and a review of current practices. A written quarterly inspection report, noting any items of non-conformance, is given to the Director of the LRC.
The facility supervisors advise the Director of corrective action.
The SRC shall consider the administration of the site nuclear criticality safety program as outlined in Appendix D.
A.9.5.2 Training -- All authorized users shall be trained in nuclear safety by the nuclear safety officer or by a qualified person designated by him. All authorized users shall be retrained annually by the nuclear safety officer or by a qualified person designated by him. The instructor shall be identified and attendance shall be documented for each training session.
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SNM-778 Docket No.70-824 Date October 1979 Page A-32 Amendment No.
Revision No.
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Subjects to be covered in training sessions will normally include:
a description of nuclear criticality safety in general, the effects of a criticality accident, and specific controls used at the LRC to maintain nuclear safety. The effectiveness of the training programs shall be evaluated based on the results of written examinations.
A.9.5.3 Posting - Each unit containing SNM is posted and material in process or storage is governed by the posted limits.
A.9.5.4 General - The Double Contingency Principle as defined in the American National Standard ANSI N16.1-1975 is followed in establishing the basis for nuclear criticality safety of all operations.
Where structural integrity is necessary to provide assurance for nuclear criticality safety in any operation, the design and construction of those structures will be evaluated with due regard to load capacity and foresee-able abnormal loads, accidents and deterioration. This engineering activity is the responsibility of the LRC's Facilities Department.
A.9.5.5 Nuclear Isolation - When nuclear isolation is required, the unit or units isolated shall be separated from all other SNM by one of the following (or equivalent) conditions.
1.
Twelve inches of water.
2.
Twelve inches of concrete with density of at least 140 lb/ft provided that the isolated unit or units cannot be representable as a slab which interacts with the other SNM primarily through its major face.
3.
The edge-to-edge separation of 12 feet, or the greatest distance across an orthographic projection of either accumulation on a plane perpendicular to a line joining their centers, which-ever is larger.
License No.
SNM-778 Docket No.70-824 Date November 1979 Page A-33 Amendment 30.
Revision No.
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d' 1868 164 Babcock & Wilcox
A.9.5.6 Building A N(TIE: Building A houses the Lynchburg Pool Reactor and the Critical Experiment Facility which are licensed under R-47 and CX-10, respectively. The following limits apply to the other units in the building.
A.9.5.6.1 Cencral - Building A is limited to 40 units.
(Units will l
conform to the mass limits in A.9.5.6.2.)
A unit shall be one of the following:
1.
A separate laboratory, room, or work area.
2.
A transfer cart where SNM is separated from adjacent units by at least 8 inches edge-to-edge and 24 inches center-to-center. More than one unit may be on a cart provided the l
Preceding edge-to-edge and center-to-center values are l
maintained.
3.
A processing bench, glove box, furnace, fume hood, or other similar process equipment or container separated from adjacent units by at least 8 inches edge-to-edge and 24 inches center-to-center.
A.9.5.6.2 Mass Limits - Each unit in Building A is limited to one of the following:
1.
Mass limits for Pu and U-235 mixtarts Pu, wt%
Limit, grams Total Fissile l
0 350 1 to 20 313 20 to 40 283 40 to 60 258 60 to 80 237 80 to 100 220 License No.
SNM-778 Docket No.70-824 Date September 1979 Page A-34 Amendment No.
Revision No.
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A. 9. 5. 7. 5. 3 Work Area of Pool Under Hot Cell No. 1-This work area position under Cell No.1 is used to unload irradiated fuel assemblies from shipping casks and to dismantle both irradiated and unirradiated fuel assemblies. The following conditions govern operations in the work area.
1.
Only one assembly at a time is permitted.
2.
An associated storage position will be permitted for fuel rods which have been removed from the assembly.
3.
The assembly and associate rod storage position will be separated from each other and any other fissile material by a minimum surface-to-surface separation of 1 foot.
4.
Fissile material and fuel rods in the associated rod storage position area each restricted to the 8.6 inch square.
5.
Only one fuel rod at a time may be removed from or inserted into the assembly or the rod storage position. A maximum of 75 rods is permitted in the rod storage position.
6.
The fuel assembly may be completely dismantled by withdrawing one fuel rod at a time from the assembly; during all stages of dismantlement, the partially dismantled assembly will be maintained within the confines of the 8.6 inch square.
7.
The assembly and its associated rod storage position may be withdrawn from the pool into the cell. Free water drainage from both the assembly and rod crorage position as well as 1 foot separation from other fissile materials is assured.
A.9.5.7.5.4 Assembly and Machine Shop and Development Test Area -- The work areas on the first floor of Building B are used to disassemble unirradiated fuel assemblies for testing. The following conditions govern operations in the work area.
1.
Only one assembly at a time is permitted to be dismantled.
2.
An associated storage position will be permitted for fuel rods which have been removed from the assembly, and will be spaced and stored as stated in items 3 and 4 below.
License No._
SNM-778L Docket No.70-824 Date September 1979 Page.A-3 9 Amendment No.
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=..
3.
The assembly and associated rod storage position will be separated from each other and any other fissile material by a minimum surface-to-surface separation of 21 inches.
4.
The associated rod storage Josition will be no larger in any dimension than the fuel assembly. There will be one such storage position for each assembly to be dismantled.
Rods may be stored or handled in a slab up to 4 inches thick provided the slab is separated from other fissile material by a minimum of 12 feet.
5.
Only one fuel rod at a time may be removed from or inserted into the assembly or any rod storage position.
Only one rod may be in transit to any one location at a time.
6.
The fuel assembly may be completely disassembled by withdrawing one fuel rod at a time from the assembly; during all stages of disassembly, the partially dis-assembled assembly will be maintained within the confines of the assembly whether damaged or undamaged during assembly.
7.
Fuel rods may be remceed one at a time from this area as required. These rods are subject to all fuel handling requirements pertinent to the area they are in (A.9.5.6, A.9.5.7 and A.9.5.8).
8.
Assemblies may be handled and dismantled under water in these areas (mock-up pool and development test area pool) subject to the same requirements of the hot cell pool (A.9.5.7.5.3).
A.9.5.7.5.5 Hot Cell Operation - Fuel rods removed from irradiated Mark B or Mark C assemblics are examined including destructive examination in the hot cell. Operations in the hot cell are governed according to the following condition.
1.
Five rods, separated from other fissile material in the hot cell by 1 foot may be stored in the hot cell.
2.
In addition to the five stored rods, another unit limited to the values of A.9.5.6.2 may be present in the cell.
In this unit, under mass control, rods may be destructively examined.
License No.
SNM-778 Docket No.70-824 Date November 1979 Page A-40 Amendment No.
Revision No.
3 186JB 167 Babcock & Wilcox
se -
The annual drill shall include as a ninimum a test of the communications links and notification procedures for early warning of the public between the LRC and state and local emergency units. The emergency 'ontrol organization, as directed by the emergency officer, will critique each drill.
A report summarizing each prepla,nned drill will be submitted to the Director. The emergency officer is responsible for preparing this report.
All training sessions, drills, and exercises will be documented.
The documentation shall include evaluations and follow up on corrective actions.
C.8.1.3 Emergency Planning Coordinator -- The LRC coordinates off-site emergency assistance with the two other B&W facilities located at the Mount Athos site. The NNFD has assumed the responsibility for coordinating all off-site emergency assistance with the three Mount Athos B&W facilities.
The LRC's emergency officer or his designated alternate is the Emergency Planning Coordinator and is assigned the responsibility of ensuring that the LRC's off-site emergency assistance requirements are properly coordinated by the NNFD emergency planning coordinator.
C.8.2 Review and Updating of the Emergency Plan and Emergency Procedures The LRC emergency officer shall conduct and document an annual review of the E=ergency Plan and the Emergency Procedures.
In addition, when drills and/or exercises indicate a need to change the plan and/or procedures, the change to the plan and/or prncedures shall be made by the emergency officer in a timely manner.
The emergency planning coordinator will review and update, through the NNFD planning coordinator, all written agreements for local ser-vices each two years.
C.8.3 Emergency Equipment and Supplies Emergency equipment at the LRC can be grouped into the following broad categories:
d) Portable radiological equipment a) Fixed meteorlogical systems e) Portable fire extinguishing equipesnt b) Fixed radiological systems
() Fixed fire detection systems ff,,Emergencylockerfirstaidequipment License No.
SNM-778 Docket No.70-824 Date November 1979 Page C-33 Amendment No.
Revision No. 3 14901 1868 168 Babcock & Wilcox
a m.
The maintenance, surveillance testing, and inventory of emergency equipment and supplies identified above and in detail in Chapter C.7 - Emergency Faci-lities and Equipment - is provided for in approved group operating and emer-gency procedures.
Portable Emergency equipment and supplies listed above and emerge,cv locker first aid equipment is inventoried and inspected quarter-ly.
Fixed 'mergency systems as listed above are inspected and/or tested in accordance with approved procedures or manufacturer's specifications.
The evacuation alarm system is tested by-weekly.
The test results of the alarm system are documented.
Emergency equipment and supplies are stored in emergency equipment lockers located in Buildings A and B.
Emergency equipment for Building C is located in the Building B emergency equipment locker.
In addition, first aid kits are located at various locations throughout the LRC site.
These kits are in the custody of qualified first aiders.
C.9 RECOVERY C.9.1 Post-Emergency Recovery Plan The post-emergency recovery plan is dependent upon the type of emergency which affected the LRC.
Normally, recovery operations following an emergency condition at the LRC will not be considered as part of the Emergency Plan, but rather as part of general operations for restoring the facility to a normal s ta tus.
Recovery operations will be directed by the emergency officer working closely with the Manager of the Facilities Department. Recovery operations will not be performed by a recovery team per se, but rather by groups within the Facilities Department and the technical laboratories which contain the required expertise to bring the operation of the LRC back to a normal status. Assistance as required will be requested from the other B&W facilities located at the Mount Athos site.
The LRC will rem:in in a post-
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O Revision No.
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1868 169 Babcock a Wilcox
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