ML20058J172
| ML20058J172 | |
| Person / Time | |
|---|---|
| Site: | 07000754 |
| Issue date: | 12/08/1993 |
| From: | Cunningham G GENERAL ELECTRIC CO. |
| To: | Pierson R NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| NUDOCS 9312130438 | |
| Download: ML20058J172 (23) | |
Text
p/.
Q) 70-Jsf i
GE Nuclear Energy Gtneni[tedoc Com;&y
' i Ly.wt;;a huaw Gr.::det to ex :sa wrieate.:s Rmd i
maren. CA 94ses l
December 8,1993 3
Robert C. Pierson, Chief Licensing Branch Division of Fuel Cycle Safety & Safeguards Office of Nuclear Materials Safety & Safeguards i
U.S. Nuclear Regulatory Commission Washington, D.C.
20555
References:
- 1) License SNM-960, Docket 70-754.
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- 2) Application for Renewal; April 21,1989.
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- 3) Revised Application for Renewal; December 9,1992.
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Dear Mr. Pierson:
Enclosed are revised pages to Appendix A (proposed license conditions) and Appendix B (demonstration section) to our current application for renewal of License SNM-960. The changes
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resulted from meeting with Mr. Gaskin of your staff.
As the application reflects the revised 10CFR20, we request that the renewal become effective on or ;
j after January 1,1994.
- .j If you or your staff have any questions, please contact me at (510) 862-4330. Thank you.
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Sincerely, j
1 G. E. Cunningham Senior Licensing Engineer l
Ica Enclosures 1
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000'9 W \\..
a, hI 9312130438 93120s-M
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PDR ADOCK 07000754 d C
PDR i
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P CONTENTS Section Page
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1.0 AUTHORIZED ACTIVITIES A-1-1 1.1 Product Processing Operations..
........................ A-1-1 1.2 Laboratory Operations........................................... A-1-1 j
1.3 General Services Operations...................
A-1-2 j
1.4 Waste Treatment.....................
........................... A-1-3 l
2.0 POSSESSION LIMITS A-2-1 i
3.0 DEFINITIONS A-3-1 i
t A-3-1 3.1 Area Manager...........
3.2 Array..............................................
............ A-3-1 l
3.3 Criticality Area.......
......................................... A-3-1 3.4 Criticality Control............................................. A-3-1 3.5 Criticality Limit Area..................
.......... A-3-1 3.6 Fissile Material Accumulation................................... A-3-1 3.7 Full Reflection................................................. A-3-1 2
3.8 Homogeneous System.......
...................................... A-3-2 j
3.9 Heterogeneous System........................................... A-3-2 3.10 Normally Suberitical Values............................
........ A-3-2 3.11 Nuclear Energy Operations....................................... A-3-2 i
3.12 Nuclear Safety....................................................
A-3-2 3.13 Safe Batch......................
................................ A-3-2 A-3-2 3.14 Site Safety Manager........
l 3.15 SNM Custodian....................
.............................. A-3-2 l
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3.16 Special Nuclear Material.....
A-3-2 l
i 3.17 Suberitical Area..................
................. A-3-3 l
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4.0 GENERAL ADMINISTRATIVE REQUIREMENTS A-4-1 4.1 Area Managers..............
................................... A-4-1 4.2 Criticality Safety Component.........................
.......... A-4-1 l
4.3 Radiation Safety Component.
A-4-2 8
4.4 Vallecitos Technological Safety Council......................... A-4-3 4.5 Compliance Policy........
A-4-4 I
4.6 Change Procedures........................
....................... A-4-5 l
t Table 4.1,
" Responsibility Matrix.................................... A-4-4 l
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5.0 CRITICALITY CONTROL ADMINISTRATIVE REQUIREMENTS A-5-1 5.1 Procedures.
................................................ A-5-1 5.2 Analysis Request...
A-5-1 5.3 Criticality Control Analysis.................................. A-5-1 r
5.4 Neutron Physics Advice...................
..................... A-5-2 5.5 Verification of Criticality Safety Analysis Results............. A-5-2 5.6 Records...
............ A-5-2 5.7 Criticality Control Inspections.
......................... A-5-2 l
5.8 Training Program....
A-5-3 i
l 5.9 Criticality Monitoring..........................................
A-5-4
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i Contents L
License No.
p,- g APPENDIX A Date 12/3/93 Amench Sect.W
P r
Section Page 6.0 CRITICALITY CONTROL CONDITIONS - TECHNICAL AND ANALYTICAL REQUIREMENTS A-6-1 6.1 General Requirements........
... A-6-1 6.2 Calculative Methods..
. A-6-1 6.3 Normally Subcritical Values - Individual Accumulations........ A-6-2 6.4 Normally Subcritical Values - Interacting Accumulations......
A-6-3 6.5 Integrity of Structures................................
A-6-4 6.6 Optimum Conditions of Moderation and Reflection...............
. A-6-4 6.7 Conditions of Moderation Other Than Optimum..............
A-6-5 6.8 Conditions of Reflection Other Than Full Reflection...........
A-6-5 6.9 Moderators....
A-6-6 6.10 Nuclear Isolation...
A-6-6 7.0 RADIATION CONTROL ADMINISTRATIVE REQUIREMENTS A-7-1 7.1 Radiation Standards......
A-7-1 7.2 Radiation Control Inspection...
A-7-1 7.3 Radiation Safety Training...
A-7-1 8.0 RADIATION CONTROL CONDITIONS - TECHNICAL AND OPERATING REQUIREMENTS A-8-1 8.1 Air Contamination Control.
A-8-1 8.2 Instrument Capability.
A-8-2 8.3 Contamination Detection.....
A-8-2 A-8-3 l
8.4 Sample Detection..
8.5 Exposure Detection..
..................... A-8-3 B.6 Instrument Calibration.
............. A-8-3 A-B-3 l
8.7 Instrument Check Sources.
A-B-3 l
8.8 Liquid Waste Disposal.
A-8-4 l
8.9 Airborne Effluent Control......
. A-B-4 l
8.10 Contamination-Free Articles.
A-B-5 l
8.11 Bioassay.
A-8-5 l
8.12 Personnel Contamination....
Table 8.1,
" Equipment and Facility Design Criteria and Guidelines".
A-8-6 l
Table 8.2,
" Acceptable Surface Contamination Levels"...............
A-S-7 l
9.0 ON-SITE TRArSFERS A-9-1 9.1 Type A and Unirradiated SNM...
A-9-1 9.2 Type B.
........... A-9-1 10.0 EFFLUENT AND ENVIRONMENTAL SAMPLING A-10-1 10.1 Effluent Sampling.
A-10-1 10.2 Environmental Sampling....
.. A-10-2 Table 10.1,
" Stack Action Levels"......
.......................... A-10-1 Table 10.2, " Liquid Effluent Action Levels"..................
A-10-1 Table 10.3, " Groundwater Action Levels".
A-10-2 Table 10.4,
" Stream Bottom Action Levels"...
..... A-10-2 Table 10.5,
" Vegetation Action Levels".
A-10-3 License No.
SNM-960 Dorist No.70-754 Sect. No.
Contents pg,, g APPENDIX A Date 12/3/93 Ameads Sect.(s)
l i-3.8 Homogeneous System means a system containing compounds of fissile
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materials in the form of uniformly distributed particles and fissile compounds in solution.
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3.9 Heterogeneous System means a system containing compounds of fissile materials in the form of clumped fissile compounds (e.g., fuel rods in water).
I 3.10 Normally Suberitical Values means those maximum values which provide safety under normal conditions of operation.
Further adjustment of such values may be necessary to incorporate safety margins for the 1
activity as it is conducted in the plant allowing for credible mishaps that could occur in the actual plant situation.
3.11 Nuclear Enerov Operations means those components of the General Electric Company engaged in the various aspects of nuclear energy and does not refer to a specific component by title.
3.1' Nuclear Safety means that field of safety comprised of criticality safety and radiation safoty.
3.13 Safe Batch means an accumulation of fissile material which is 45% of the critical accumulation considering enrichment, full reflection, and optimum water moderation consistent with the form of the material.
3.14 Site Safety Manager means an individual designated by upper management as having the responsibility for nuclear safety at the entire VNC site.
3.15 SNM Custodian means an individual designated by an Area Manager who is responsible for maintaining the inventory of CLA's.
l 3.16 Special Nuclear Material means plutonium, Uranium-233, Uranium-235, or any material artificially enriched by any of the foregoing.
Limnse No. SNM-960 70-7.1 MH No.
Sect. No. 3. 0 Page A-3-2 APPENDlX A Date 12/3/93 Amends Sect (s)
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3.17' Suberitical Area means a physically identified area or location
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involving special nuclear materials in quantities of less than 500 l
grams of U-235,' 300 grams of U-233, or 300 grams of plutonium or a prorated combination of such materials under the direction of a single SNM custodian and which is unrelated to any other area where special nuclear materials are handled (a subcritical area is considered unrelated when it meets the isolation requirements of Section 6.10 of this Appendix and is not located in the same room); or an unrelated'-
building or structure under the direction of a single SNM custodian I
I which meets the criteria of 10CFR70.24 (a) e l
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i License No.
SNM-960 4 tNo.
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Pp Ma 70-754 APPENDIX A Oste 12/3/93 g,9 j
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4.3.2 The minimum qualifications of personnel assigned functional f
responsibilities in the Radiation Safety component shall be as follows:
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Manaaer:
B.S. degree in science or engineering with five years experience in assignments involving radiation protection.
b.
Specialists:
B.S. degree in science or engineering with two years experience in assignments involving radiation protection or eight years of experience in health physics or radiation protection.
c.
Monitors: High school with two years experience in handling i
radioactive materials or two years of college and four months experience; also, successful completion of a General Electric certification program which includes written and oral examinations covering radiation protection procedures and l
criticality procedures.
t 4.4 VALLECITOS TECHNOLOGICAL SAFETY COUNCIL l
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The functions of the Vallecitos Technological Safety Council (VTSC) shall i
i include responsibility for review of reportable incidents and the nuclear safety program, for contributing professional advice and counsel on j
criticality and radiation safety policy, and review, as appropriate, of new facilities or major changes to facilities. The VTSC shall review annually the site safety and compliance program performance to include effluent 1
releases and occupational exposures in terms of ALARA and focus on trends for corrective action as necessary.
Its deliberations in any calendar quarter shall be reported in writing to the f
f Site Safety Aanager (SSM) (see Section 3.14) and retained two years.
Should the VTSC choose not to meet in any calendar quarter, the SSM shall be j
i notified in writing. The Vallecitos Technological Safety Council is an i
independent review body and shall consist of at least five senior members of' i
General Electric's technical and/or management personnel appointed by the SSM and shall include competence in the physics, chemistry, and engineering i
disciplines.
License No. SNM-960 Docket No.70-754 g g 4.0 p,, A,4,3 APPENDIX A
> 12/3/93 m g,y 3
4.5 COMPLIANCE POLICY l
General Electric shall establish and maintain a policy statement in writing expressing requirements for compliance with the terms and conditions of special nuclear materials licenses and applicable NRC regulations. VNC shall establish a comprehensive set of standards for operational health and safety, a system of site emergency procedures, and a system of standard operating 1
procedures. VNC is committed to controlling activities involving special nuclear materials in accordance with these approved written procedures and standards.
These documents shall be prepared, reviewed, revised, approved, and implemented in accordance with the matrix contained in Table 4.1.
Violations of radiation protection procedures or criticality safety j
specifications which are of repetitive or serious nature are subject to disciplinary action.
j TABLE 4.1.
RESPONSIBILITY MATRIX l
A-Prepare l
B - Review Regulatory C - Approve Compliange Area Operating l
Function Managers Management 1
Site Safety Standards A,
B, C
A, B,
C A
Nuclear Safety Procedures A,
B, C
Operating Procedures B
A, C
Change Authorization A,
B A,
C B
Emergency Procedures A,
B, C B,
C l
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J 1Standards include radiation satety, criticality safety, industrial safety, and envir nmental.
2Only those related to nuclear safety.
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l Includes radiation, criticality, environmental, and hazardous materials safety functions.
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Ucense No.
SNM-960 Dodet No. 70-754-Sen No.
4.O pg A,4,4 j
i APPENDIX A gate 12/3/93 m gengg}
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4.6 CHANGE PROCEDURES i
i Activities which do not involve a change in license conditions but which require procedures, facilities, or equipment rubstantially different from i
those previously used shall not be initiated until the Radiation Safety component has completed a review and technical evaluation to assure adequacy f
of health and safety features and compliance with license conditions and-ALARA policy where applicable.
l In a criticality limit area, such activities shall not be initiated until the
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Criticality Safety component has completed a criticality analysis or a review of the existing criticality analysis for the proposed activity. Such changed activities shall be initiated in accordance with written procedures issued by the area manager.
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l Licener No. SNM-960 Dodet No.70-754 g,g Page A 5 4.0 l
APPENDIX A
> 12/3/93 gg
i control purposes, the movement of SNM, and general criticality safety work practices.
l 5.7.2 A program of inspection shall be performed by the criticality Safety component to determine that actual operations conform to the physical i
i situations on which the calculations of criticality limits have been based.
Inspection reports shall be furnished to area managers.
Where situations are identified which require corrective action, such reports shall so indicate. Corrective or follow-up' action shall be i
taken in accordance with VNC Safety Standards and/or site Nuclear Safety Procedures.
5 Operations which handle routinely in process more than one safe batch (as defined in Section 3.13) of fissile material shall be inspected on a quarterly basis. All operations shall be inspected on an annual
- basis, i
1 5.8 TRAINING PROGRAM Area managers shall assure that new employees receive instruction'in
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l criticality safety and plant operating and emergency procedures prior to their working with special nuclear materials in a criticality limit area 1
(CLA). A criticality control training program, approved by the Criticality Safety component, shall be maintained to emphasize the need for following l
l criticality control procedures and to aid personnel in understanding the various parameters which are essential to the maintenance of subcritical conditions. The program may be conducted by the Criticality Safety l
l component, some other portion of the Nuclear Safety component, or combined with training performed by operating components.
It may be combined with radiation safety training. A written test shall be completed by each i
employee taking the course. The test shall be evaluated and the results q
l forwarded to the appropriate area manager, j
Employees requiring criticality safety training shall receive refresher training annually.
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l SNM-960 70-754 5.0 Ucense No.
Docket No.
Sect. No.
p,,,.A-5-3 APPENDIX A h 12/3/93 gg
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l 5.9 CRITICALITY MONITORING 5.9.1 No transfers of fissile materials between criticality limit areas shall be permitted in criticality areas required to have a monitor alarm system unless the system is operable. Transfers may resume following repair and verification of the monitoring system's operability.
5.9.2 Exemption from the monitor alarm requirement of Secti;n 70.24 (a) (2) of 10CFR70 is granted for the Remote Handling Operation (RHO) pool and hot cells. At the high-level solid waste storage facility and the storage pit in the RHO pool area, for the purpose of compliance with 70.24 (a) (2) of 10CFR70, the source of a possible accidental condition of criticality may be considered as the accessible surface of the earth or concrete shielding.
5.9.3 Specific exemption from the prescribed preset alarm points (5 to 20 mrem /h) is granted for those monitoring devices where the routine movement of by-product materials could result in gamma fields at the monitoring devices in excess of those that would result in a dose rate of 20 mR/h. The alarm setting for these monitoring devices under the above conditions shall be such as to respond to the postulated criticality but shall not exceed 500 mrem /h.
5.9.4 Exemption from the requirements of Section 70.24 of 10CFR70 is granted for each area in which there is not more than one shipment of packages containing special nuclear materials licensed pursuant to 10 CFR Part 71 for transport outside the confines of the Vallecitos l
Nuclear Center (packages in any shipment which depend on special arrangement for nuclear safety shall be retained in that same arrangement during such storage); or one safe batch of finished reactor fuel rods or assemblies, providing no activities could cause rearrangement of fuel-bearing portions into more reactive configurations; or which meets the requirements of a suberitical area as defined in Section 3.15 of this Appendix.
License No.
SNM-960 Dodet No.70-754 gg P8F A-5-4 5.O APPENDIX A g 12/3/93 m g(g)
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l 7.0 RADIATION CONTROL ADMINISTRATIVE REQUIREMENTS 7.1 RADIATION STANDARDS j
General Electric shall establish and maintain a comprehensive set of standards for operational health and safety, including ALARA considerations.
Such standards shall be reviewed by the Radiation Safety component and the appropriate area managers prior to issuance. The Radiation Safety component shall review the standards annually thereafter.
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7.2 RADIATION CONTROL INSPECTION j
i Activities involving special nuclear material shall be inspected by the Radiation Safety component on a continuing basis. Conditions of an unusual' l
or uncertain nature that could lead to radiological health and safety problems shall be referred to the area manager immediately for correction.
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7.3 RADIATION SAFETY TRAINING l
A training program in radiation protection shall be in effect and shall include training in the requirements of 10CFR19 and 10CFR20, methods of controlling radiation exposure, license requirements, protective methods, and basics of radiation effects. All personnel working in radioactive materials areas or in radiation areas shall receive an indoctrination lecture j
prior to starting work followed by additional training commensurate with the work environment as determined by the area manager.
Personnel working with SNM or regularly assigned to work in radiation areas shall receive refresher training annually.
l License No.
SNM-960 Dodet No.70-754 Sen No.
7.0 p.,,
3,7,y APPENDIX A Osto 12/3/93 Amens Sen(s) l l
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8.0 RADIATION CONTROL CONDITIONS -
TECHNICAL AND OPERATING REQUIREMENTS 8.1 AIR CONTAMINATION CONTROL 8.1.1 The following design criteria for ventilation systems shall be used l
to provide air contamination control.
8.1.1.1 Airflow shall be from areas of lesser contamination to areas.of l
higher contamination.
Potential accident conditions shall be considered.
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8.1.1.2 Duct flow velocities'and design shall be such to minimize possible l
accumulation of contamination.
l 8.1.1.3 Air shall be sampled continuously in normally occupied areas in l
l which dispersible SNM is handled. Samples'are analyzed for gross alpha and gross beta-gamma.
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e 8.1.1.4 Potential airborne radioactivity-producing operations shall utilize l l
close-capture ventilation devices, e.g.,
hoods and high-velocity 1
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local exhaust when comparable with ALARA.
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8.1.1.5 All ventilation exhaust systems for facilities that routinely handle dispersible SNM in quantities in excess of the limits i
specified in Appendix C of 10CFR20 shall be sampled continuously as j
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l proof of filter performance. All HEPA filters for such facilities l
l will be fire resistant.
8.1.1.6 At least one filter in each such stream ( 8'.1. 5) shall be equipped-l with a device for measuring differential pressure that shall be t
read monthly. Filters shall be changed when readings deviate from specification values (.8 to 4.0 inches WG) or following evidence of - l
'f damage.
i License No.
' k No.
8.0 p, A,g,1 APPENDIX A Dets 12/3/93 mw) l
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8.1.2 Smearable contamination in excess of 10,000 dpm/100 cm alpha in normal working areas shall be cleaned up expeditiously.
e 8.2 INSTRUMENT CAPABILITY Dose rate range capabilities of portable inst? 1 mentation shall include the following:
gamma
- 1 to 50,000 mR/h (10-5,000 pR/hr for environmental monitoring) beta
- 4 to 200,000 mrad /h neutrons - 1 to 5,000 mrem /h 8.3 CONTAMINATION DETECTION 8.3.1 Contamination detection capabilities of portable instrumentation
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shall be:
alpha
- 200 to 300,000 disintegrations per minute beta-gamma - twice background to 400,000 disintegrations per minute l on a Geiger-Mueller counter 8.3.2 Contamination detection capabilities of portable instrumentation when used in the smear survey technique shall be 200 disintegrations per minute for alpha and 500 disintegrations per minute beta-gamma.
8.3.3 Detection equipment shall be maintained on site to meet the requirements of Section 8.11 of this Appendix for contamination-free items.
8.3.4 For unconditional release surveys (Section 8.10), laboratory-type counting equipment may be used to meet the release limits of Table 8.2.
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8.0 License No.
P8F A-8-2 l
APPENDIX A
> 12/3/93
% g,)
8.4 SAMPLE DETECTION l
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Sample detection capabilities for laboratory analysis of effluents shall be l
s.1 times the concentrations specified in 10CFR20, Appendix B, Table II.
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8.5 EXPOSURE DETECTION i
Personnel dosimeters shall be capable of detecting gamma, beta, and x-ray radiation. Additional neutron detection capability shall be available as appropriate. Primary dosimeters (film badges, TLD dosimeters) are worn in a l
manner as to record the maximum whole body exposure.
l 8.6 INSTRUMENT CALIBRATION T
Portable monitoring instruments shall be calibrated upon initial acquisition, I
after major maintenance, and at least aviually.
r Fixed gamma area monitors used as detection or warning devices (i.e., not for personnel monitoring) shall be source checked at least annually.
8.7 INSTRUMENT CHECK SOURCES 1
Field check sources shall be available for use in functional' response checks of portable radiation-measuring instrumentation.
8.8 LIQUID WASTE DISPOSAL l
Potentially contaminated liquid wastes shall be collected, solidified, and disposed of as solid waste.
License No. SNM-960 Dodot No.70-754 gg, Page A-8 0.0 APPENDIX A Oste 12/3/93 M h (s)
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t l-l 8.9 AIRBORNE EFFLUENT CONTROL Potentially contaminated airborne effluents shall be released through HEPA filter systems which shall be at least 99.95% efficient for 0.3-micrometer particles. Such effluents shall be limited at the point of release to the atmosphere so that the annual average concentration at the site boundary does not exceed the concentrations specified in 10CFR20, Appendix B, Table II, Column 1.
The limits at the points of release shall be based on a x/Q calculation, reconcentration factors, and the effects of other site stacks.
If the environmental sampling program indicates a reconcentration of these materials, uhe release levels shall be lowered such that the reconcentration trend is reversed.
These represent maximum release limits. Actual normal releases shall be controlled to meet ALARA objectives (see Section 10.0).
8.10 CONTAMINATION-FREE ARTICLES Articles which have been handled, used or stored in areas with a potential for contamination with radioactive materials may be disposed of or transferred to persons not licensed to possess radioactive materials when all of the following conditions are satisfied:
8.10.1 Either all surfaces are accessible for survey or it is reasonable to l assume from the design and usage that no radioactive materials could have contaminated inaccessible surfaces without having contaminated j
the accessible surfaces as well.
8.10.2 Articles are considered contamination free which meet the requirements of Table 8.2.
License No.
SNM 960 Dodet No.70-754 8.O h No.
P8F A-8-4 APPENDIX A Date 12/3/93
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8.11 BIOASSAY Persons who work roatinely in areas where there is potential for internal deposition of radioactive materials shall be subject to determination of the extent of radioactive materials intake by techniques such as analysis of I
urine or whole body counting.
Bioassay data shall be evaluated by the radiation safety function.
8.11.1 whole Body Counting: VNC employees shall receive a whole body count l at least annually. An investigation shall be initiated if a whole body count result exceeds 10% of an ALI.
l 8.11.2 Urinalysis: Urinalysis may be performed on a job-by-job basis when l
radioactive materials which cannot be directly detected by the whole body counter and which are not tagged with isotopes detectable by the whole body counter are handled.
Sampling will begin with a base sample before the project starts and continues on a regular schedule 4
until the project is completed. All samples above background are investigated.
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8.12 PERSONNEL CONTAMINATION i
Personnel contamination shall be kept below 200 dpm/100 cm alpha.
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l 8.O SNM-960 Dodet No.70-754 Sect. No.
Pape A 5 License No.
12/3/93
% gg APPENDIX A Date
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TABLE 8.1.
EQUIPMENT AND FACILITY DESIGN CRITERIA AND GUIDELINES E
Facility Alarms, Interlocks and Safety Features Basis D
E h
Hot Cells Door interlock (operationally checked Prevents both doors from opening simultaneously twice per year).
during normal operation.
Interlock can be 2
y nullified during cell cleanout.
o Stepped plugs in cell penetrations.
Prevents beam-type exposure to personnel at cell face.
Continuous air monitors at face of Give prompt alarm in case of contamination spread
- cells, outside cell.
Continuous radiation level meters in Give immediate alarm in case of high radiation g
Re vicinity of cells and storage pool.
outside cell.
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Check valves on all inlet lines for Prevents release of contaminants to work area 5
liquids and gases.
through backflow.
Effluent air double filtration -
Provides contingency in case of cell fire and HEPA filters, loss of first filter.
> 0.02 in. of H O 6P (instrument Prevents blow-back to working area; positive flow checked for accuracy annually).
through openings around plugs, etc., > 125 ft/ min.
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b U'nderwater Anti-syphon drain.
Prevents water draining out.
f Storage Pools p
Automatic filling.
Prevents loss of shielding by evaporation, etc.
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Hoods Airflow designed to 125' lineal ft/ min Prevents carryout of radioactive materials.
(flow checked quarterly).
Glove Boxes 2 0.5 inch H O AP.
Prevents spread of radioactive material to g
working areas.
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TABLE 8.2.
ACCEPTABLE SURFACE CONTAMINATION LEVELS h
NUCLIDES AVERAGE MAXIMUM REMOVABLE P
h
2 ht 5,000 dpm a/100 cm 15,000 dpm u/100 cm 1,000 dpm a/100 cm Q
associated decay products E
$o Transuranics, Ra-226, Ra-228, Th-230, Th-228, Pa-231, 100 dpm/100 cm 300 dpm/100 cm 20 dpm/100 cm Ac-227, I-125, I-129 Th-nat, Th-232, Sr-90, Ra-223, Ra-224, U-232, I-126, 1,000 dpm/100 cm 3,000 dpm/100 cm 200 dpm/100 cm I-131, I-133 g
Beta-gamma emitters (nuclides y
Rp with decay modes other than 2
2 2
R alpha emission or spontaneous 5,000 dpm Sy/100 cm 15,000 dpm ST/100 cm 1,000 dpm By/100 cm y
fission) except Sr-90 and o
j others noted above.
.e.
- where surface contamination by both alpha-and beta-gamma-emitting nuclides exists, the limits established for alpha-and beta-gamma-emitting nuclides should apply independently.
bAs used in this table, dpm (disintegrations per minute) means the rate of emission by radioactive material as determined by correcting the counts per minute observed by an appropriate detector for background, efficiency, and geometric factors associated with the instrumentation.
F E
- Measurements of average contaminant _should not be averaged over more than i square meter.
For objects of f
less surface area, the average should be derived for each such object.
d The maximum contamination level applies to an area of not more than 100 cm.
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'The amount of removable radioactive material per 100 cm of surface area should be determined by wiping that area with dry filter or soft absorbent paper, applying moderate pressure, and assessing the amount of radioactive material on the wipe with an appropriate instrument of known efficiency. When removable contamination on objects of less surface area is determined, the pertinent levels should be reduced proportionally and the entire surface should be wiped.
I fThe average and maximum radiation levels associated with surface contamination resulting from beta-gamma Y
emitters'should not exceed 0.2 mrad /hr at J cm and 1.0 mrad /hr.at I cm, respectively, measured through not more than 7 milligrams per square centimeter of total absorber.
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DEMONSTRATION FOR SPECIAL NUCLEAR MATERIAL LICENSE RENEWAL FOR THE VALLECITOS NUCLEAR CENTER 1
i APPENDIX B i
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l LICENSE SNM-960
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DOCKET 70-754 MARCH 27,1989 (REVISED DECEMBER 1,1992)
(REVISED DECEMBER 3,1993) l GENERAL ELECTRIC COMPANY VALLECITOS NUCLEAR CENTER 4
P.O. BOX 460 PLEASANTON, CA 94566
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i Section Pace t
8.0 BUILDING 107 8-1 8.1 Location and General Description............................
B-1 8.2 Use......................
8-1 i
t 9.0 DELETED
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l 10.0 CHEMISTRY, METALLURGY AND CERAMICS LABORATORY -
BUILDING 103 10-1 i
5 10.1 Location and General Description............................
10-1 10.2 General Plans and Uses of Material.........................
10-1 10.3 Laboratory Facilities and Equipment.......................
10-1 10.4 Building 103 Procedures....................................
10-5 f
10.5 Building 103 Criticality Control.........
10-7
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Ficure l
10.1 Building 103 Floor Plan (Ground Level)......................
10-8 l
10.2 Building 103 Floor Plan (Second Floor)......................
10-9 j
10.3 Building 103 Vault Layout...................................
10-10 11.0 DELETED I
12.0 BUILDING 105 12-1 12.1 Location and General Description..
12-1 12.2 NTR Facilities.......
12-1 12.3 Other Laboratory Areas (Advanced Nuclear Applications)....
12-1 Figure 12.1 Building 105 Floor Plan......................................
12-2 13.O WASTE HANDLING 13-1 13.1 Radioactive Liquid Waste Evaporator Plant (Building 349)....
13-1 l
13.2 Solid Waste Handling.........................
13-4 l
Figure 13.1 Waste Treatment Plant...........................
........... 13-6 13.2 Hillside Waste Storage.....................................
13-7 1
i Licenu No.
SNM-960 Decket No.70-754 5.ct. No.
Contents
- p.,
12/3/93 Deve Amends $ect.(s) j I
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The facility has a horizontal tube facility for storage of high-level radioactive material contained in sealed encapsulations called awaste liners".
The horizontal tube facility is made_of two rows of 40-foot-long concrete-lined steel pipes mounted horizontally and covered with earth.
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Eleven of the tubes have a 6-inch inside diameter, and seven'have a 10-inch l
inside diameter. The tubes in either row are spaced on 3-foot centers, and the rows are spaced 3 feet apart with.the tubes in.the bottom row offset j
halfway between the tubes in the upper row.
Shielding is provided on the top f
and sides of the facility by a minimum of 6 feet of compacted earth.
l Shielding at the front and back ends consists of 3 feet of concrete in which-I the pipes have been anchored, plus concrete-filled step plugs with a minimum of 3 feet of concrete shielding in the plug. Additional above-ground space f
i for lower level waste or other materials is available within this fenced and posted facility.
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The facility is covered by a prefabricated metal building. Ventilation (max.
i 18,500 cfm) is operated when material is transferred. Air is exhausted through a bank of absolute filters
(> 99.97% efficient) and the system is sampled when vent system is operating.
i 3.5.2 Licuid Wastes l
i Liquid wastes are routed from laboratory sinks and gravity drains leading from sources known to be or potentially contaminated through regulated pipe lines to retention tanks located in each building where such wastes are generated.
Such wastes are transferred periodically to a waste treatment j
l plant for concentrating and solidifying the liquid wastes which are described i
l in Section 13 of this application.
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-L Other liquid waste (excluding sanitary waste) flows through a separate piping-,
system into any three of four 60,000-gallon retention basins. After sampling
[
i and determining that radioactivity, if any, is within permissible-discharge-
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levels (3.0 x 10 gross alpha and 5.0 x 10 gross beta-gamma), the water in j
the basin is released.
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Sanitary wastes are treated, and the waste waters are sprinklered on site.
1 License No.
SNM-960 Decket No.70-754 30 Sect. No.
Pep j
i' 12/3/93 3-4 Dete Amends Sect.(s)
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4
i 8.0 BUILDING 107 i
8.1 LOCATION AND GENERAL DESCRIPTION l
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Building 107 consists of two prefabricated, compartmentalized storage units with three storage cells per unit.
The units are covered by a prefabricated l
l metal building.
8.2 USE l
The cells are used for storage of hazardous materials (nonradioactive). No SNM is used or stored in the facility. Secondary containment is provided in i
case of spills.
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70-754 80 License No.
Sm-960 Docket No.
bet. No.
Pope Dee Amends het.(s) t
r
+
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f
_g LTST OF EQUIPMENT
- 3. A BOILER
)
DVERHE AD 000R
- 2. CHEMICAL FEEDER
- 3. A BOILER FEED A
B 4.
CONTROL PANEL' OFFICE AREA
- 5. B BOILER l
- 6. B BOILER MOT WELL j
- 7. U PUMP r
P B. ROLLEh CONVEYOR i
1 5
- 9. ORY MIXER
- 18. SLURRY T ANK ( J )
FIL
- 13. CHEMICAL TREAT N..T l
l TANK (CTT3
.i 6
l I
- 14. DEION12ERS gDVERHEA0 000R %
- 15. R PUMP g
- 16. O PUMP i
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- 17. EVAPORATOR SYSTEM l
OVERHEAD 000R l
jB P
g I
- 28. VENTILATION SYSYEM 4
- 21. TRANSPORT TANK { TT I
PORTABLE l
- 22. M PUMP
,i
- 23. HOLOUP TANK tHUT) f g
33 24 N PUMP J
25..D PUMP l
8 10
- 26. L PUMP.
- 27. MONITOR TANKS i
p**.-* * * '. **
l P-28 x SUMP PUMP
?
i g
i
,*.'. * * '.. * ' l 12 IN SHIELOING l
l DVERHEAD 000R s
CTT
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]
13 22 24 l
- /
925 O
9
l MUT i
23 l
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.=
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18 19
- ...:.. m 28 Q25 l
1 I
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Mil MT2 f
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27 I
l 27 i
{
l OVERHEAD 000R
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21 l
l g 28 i
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t FIGURE 13.1.
WASTE TREATMENT PLANT.
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I 1.icense No.
SNM-96o Decliet No.
- 10-754 bet. No. Figure 13.1 p,p Date Amends Sect.(s)
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