Text

Forwards Replacement Pages to Appendices to SNM-960, Submitted in Response to Rai,Dtd 990726.Editing Line in right-hand Margin of Revised Pages Indicates Changes
	ML20211N739
Person / Time
Site:	07000754
Issue date:	09/01/1999
From:	Murray B; GENERAL ELECTRIC CO.
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
References
	TAC-L31006, NUDOCS 9909130054
	Download: ML20211N739 (50)
	v • d • e

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GE NucIcar Energy aw re nc nyyy kUIw 'O.\\ % In st Cente?

bIUD Vah'!ilUT OC3d S)nC[, (b OOO6 September.1,1999

- Director, Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555

References:

1.

License SNM-960, Docket 70-754, TAC No. L31006.

2, Letter from Andrew D. Rayland to Ben Murray," Request for AdditionalInformation Materials License Renewal, SNM-960 (TAC No. L31006)"; July 26,1999.

3.

Request for Renewal of SNM-960, B. M. Murray to Director, Office of NMSS, dated January 21,1999.

4.

Letter from B. M. Murray to Director, Office of Nuclear Material Safety and Safeguards, dated July 14,1999.

Dear Sir:

The enclosed replacement pages to the appendices to SNM-960 are submitted in response to the request i

for additional information (Reference 2). Some of the included pages were previously submitted (Reference 4), but are included here to provide a complete response tu the request for additional information. The included pages (Attachment B) are:

Pages i through iii, Table of Contents for Appendix A Pages A-1-1 and A-1-3 of Section 1, Appendix A Page A-2-1 of Section 2, Appendix A Pages A-3-1 through A-3-3 of Section 3, Appendix A Pages A-4-1 through A-4-4, of Section 4, Appendix A Page A-5-4 of Section 5, Appendix A

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Page A-7-1 of Section 7, Appendix A Pages A-8-2 through A-8-5 and A-8-8 through A-8-10 of Section 8, Appendix A Page lii, Table of Contents for Appendix B Pages 1-3,1-4 and 1-5 of Section 1, Appendix B e

Pages 2-4,2-9,2-10, and 2-12 of Section 2, Appendix B e

Pages 3-4 through 3-6 of Section 3, Appendix B e

y Pages 4-1,4-2, and 4-5 of Section 4, Appendix B

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Page 7-6 of Section 7, Appendix B Page 13-5 of Section 13, Appendix B.

e The editing line in the right-hand margin of the revised paps indicates the changes.

Item number 2 of reference 2 requested a copy of the position paper regarding a negative iodine bias in the stack effluent measurements. A copy of this paper is enclosed as Attachment A.

1 QliT.1 d 9909130054 990901 PDR ADOCK 07 7y4

4

-USNRC

September 1,1999 Item number 25 of reference 2 deals with the differences between the VNC fire program and the criteria specified in NFPA 801," Recommended Fire Protection Practice for Facilities 11andling Radioactive Materials",1991 Edition. The Vallecitos Nuclear Center (VNC) fire protection program meets many, but not all, of the applicable criteria. Inspectors for the American Nuclear Insurers (ANI) routinely audit the VNC for compliance with fire protection practices. These audits are very rigorous and assure a compliant fire protection program at VNC that meets the intent of the applicable requirements of NFPA

- 801. The exceptions to the criteria in NFPA 801 are itemized below:

Chapter 2-2.3 through 2-4:

A specific fire hazard analysis does not exist; the program has evolved over 40 plus years of operation. Inspections are performed annually. A fire response plan is documented in the VNC Site Emergency Procedures.

Chapter 3:

The ducting, fire suppression, dampering and emergency power requirements are met for the Building 102 facility, which has the largest potential for radioactive material release. The other buildings on site, with low potential source terms, are not in strict compliance with NFPA.

801.

Chapter 4-4.2 Each fire hydrant does not have an isolation valve m addition to the shutoffvalve, If you or your staff have any questions concerning this application, please contact me at (925) 862-4455.

Thank you.

j Very truly yours, b

L B. M. Murray Senior Licensing Engineer Attachments 1

cc:

W. L. Britz, Fuel Cycle Inspector

' U. S. NRC, Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, Texas 76011-8064 t-l j

1 e

ATTACHMENTA 4

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r AttecA2mcnt A February 18,1999 Memo to Regulatory Compliance File

Subject:

Analysis of Potential Negative Bias in Iodine-131 Samples from VNC Stack Samples The attached evaluation (Attachment A and supporting I-131 release charts) discusses the probable cause and significance of an observed possible negative bias in the reported quantities ofI-131 released from the VNC effluent stacks. The evaluation concludes that a small, but insignificant, bias is possible. Weekly samples display a mix of positive and negative values on a random basis. Future annual reports will use consistent monthly report periods for all of the airborne effluent releases.

b Ben Murray l

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d February 18,1999 Attachment A

-Refer-nce: NRC InWon Report 70-754/98-02, Dated Jan. 4,1999; Item 3.2,

" Observations and Findings" related to I-131 airborne release samples.

3.2 Observations and Findings "The annual report of1997 efluent monitoring and environmental surveillance programs at VNC was reviewed. The inspectors noted that Figures 3 and 5 were

. negatively biasedfor the iodine-131 airborne release samplesfor the entireyear.

Negative results on samples without a source term are expected on a random basis but are typically statistically ofset byperiodicpositive results. An entire year ofnegative results may be indicative ofa calibrationproblem. The licensee had not addressed and could not explain the consistent negative resultsfor the 1997 data. The laboratory calibrationfor the analysis oflodine-131 airborne release samples will be reviewed during a subsequent inspection as inspectionfollow-up item (IFl 70-754/9802-01).

Response

Figure 3 of the 1997 annual summary of the VNC effluent monitoring and environm surveillance programs contains bar charts of the integrated releases of gross alpha, gross beta-gamma, I-131, and noble gas from the Building 102A stack. The time axes of the bar charts are divided into monthly intervals, except for the I-131 releases. The I-131 chart is consolidated into quarterly periods. Figure 5 of the 1997 annual summary of the VNC effluent monitoring and environmental surveillance programs contains three of the same charts for the "other" stacks, except for Building 102A and the NTR. There is no noble gas releases from these "other" stacks. The "other" stacks' consist of Building 103 (Sample No.12), the Waste Evaporator (Sample No. 30), the Waste Storage Facility (Sample No. 34), and the Bunker Containment Building at the Hillside Storage Facility (Sample No. 37).

i The current routine operations at VNC provide very little potential for release ofI-131 from any of the "other" stacks or from the Building 102A stack. The NTR is an operational test reactor which is a likely source of airborne iodine-131. Airbome I-131 is

)

created from the fissioning of" tramp" uranium which is contained in the aluminum skin of the U-Al fuel disks. Therefore, the amount ofI-131 in the charcoal cartridge samples from the NTR is normally a positive value. - However the I-131 content of samples from Building 102A and "other" stack effluents is typically non-detectable. The reason for the lack ofI-131 in these facilities is that irradiated fuel is only handled in Building 102 and at the Hillside storage facility, and the minimum decay time on this fuel is 90 days.

- Ninety day decayed fuel is essentially free ofI-131.

i i

The established method for reporting the stack releases, as measured from the sample media used in the continuous stack sampling systems, is to report a positive or negative net value, i.e., the value of the gross counts minus a background value (converted to appropriate units). Collected I-131 is measured with a GeLi detector and multichannel analyzer. Count peaks on the analyzer from I-131 are predefined as regions surrounding three characteristic energy peaks of the I-131 decay photons (284,364.5, and 637 kev).

A VNC created algorithm (SOURCE.ICAL), used with the counter output, performs an automatic background subtraction from the gross counts in the identified peak regions and calculates the net concentration ofI-131 in the sampled air. The background quantification method used in the VNC counting system, and other counting systems, is to calculate a background as the integrated counts within a trapezoidal area defined by a straight line function between the start and end points of the region ofinterest on the energy spectrum curve. This background count is then subtracted from the total gross counts within the area defined by the actual counts between the end points of the region ofinterest. When the gross counts from a background sample are less than the calculated background counts, a negative net count value is reported.

For a true zero activity and a very large number of counts of the sample, the negative values should statistically equal the number of positive net counts. However, a statistical analysis of the52 samples taken from the Building 102 stack for the year of 1997 yielded a negatively biased distribution of the positive and negr results on the charcoal samples. The mean value of the calculated I-131 releases was approximately -0.53 microcuries per week. This bias should be put into perspective. The bias of-0.53 microcuries is less than 0.001% of the weekly release action levels (from VSS 7.2) of 73 millicuries per' week. This level of bias is not considered statistically significant with respect to the action levels for VNC stack releases considering the systematic errors associated with the stack sampling system (stack and sample flow rates, sample times, sample efficiency, counting statistics, etc.).

A possible cause of the minor negative bias is that the typical shape of a true background energy spectrum may be such that a negative, albeit low, net count will typically result from the analysis. For example: if the shape of the energy spectrum between the defined points ofinterest is convex, i.e., the curve falls below the calculated straight line function, a small negative net value will result from the background function calculation. Because of the nomially irregular shape of a background energy spectrum curve, a negative net count rate is a reasonable possibility.

In conclusion, the anomaly of reporting a negative value for the sum of the 13 samples in a quarter for all four quarters in a year is unexpected, but not impossible. When the weekly sample results for the same annual period are plotted as a 52 group bar chart, it can be seen that there are many positive sample results in addition to the negative results (see attached graphs of Building 102A I-131 releases for 1997 and 1998). The weekly graphs confimi the random, but slightly biased, nature of the sign of the net count values on Building 102A stack I-131 samples.

2 J

For consistency, the I-131 release values in future annual environmental reports will be displayed as monthly values (12 groups) in the bar charts, as are the other categories of release in the report.

Ben Murray i

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ATTACHMENTB 1

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CONTENTS Section Pace i

1.0 A UTH ORI Z E D ACTIVITI ES..................................................................... A 1 1.1 Laboratory Operations............

.... A 1 1.2 General Services Operations.....

............. A-1 -2 3

1.3 Waste Treatment...................

.....................................A-1-3 2.0 PO S S E SS ION LIM ITS...

............................................................................ A 1 3.0 D E FINI TI O N S............................................................................................. A 1 j

i 3.1 Area Manager.........

..........................................A-3-1 1

3.2 A rray...............................

........... A 1 3.3 Criticality Area........

...... A 1 3.4 Criti cality Control.............................................................. A 1 3.5 Criticality Limit Area..

.. A 1 1

3.6 Fissile Material Accumulation..

.. A 1 3.7 Full Re flection..................

................... A 1 3.8 Homogeneous System.........

................... A-3 -2 3.9 Heterogeneous System...........

............. A-3 -2 3.10 Normally Subcritical Values..............

.................................................A-3-2 3.11 Nuclear Energy Operations.........

.............................................A-3-2 3.12 Nuc Iear Safcty.................

............ A-3 -2 3.13 Safe Accumulation..........

..... A-3-2 3.14 Safe Batch......................

.. A-3-2 3.15 Safe Concentration..........

.............. A-3 -2 3.I6 Safe Geometry............

..............................A-3-2 3.17 Safe Mass........

.................................................A-3-2 3.18 Safe Spacing...........

......................A-3-3 3.I9 Safe Volume......

.................................A-3-3 3.20 Site Safety Manager.......

..................... A-3 -2 3.21 SN M C u stod ian........................................

.... A-3-2 3.22 Special Nuclear Material.................................

................... A-3 2 3.23 Subcritical Area.........

........................A-3-3 4.0 G ENERAL ADMINISTRATIVE REQ UIREMENTS...................................................... A-4-1 4.1 Area Managers.......................

...........................A-4-1 4.2 Criticality Safety Component..............

.........................A-4-1 4.3 Radiation Safety Component...................

................... A 2 4.4 Vallecitos Technological Safety Council...

.... A-4-3 4.5 Compliance Policy...........

......... A-4-4 4.6 Chege Procedures....

....... A 5 Table 4.1," Responsibility Matrix"...

....... A-4-4 l

License No. SNM-960 Docket No.70-751 Sect. No. Contents Page i AppendixA Date 8/23/99 Amends Sect.(s) 1.0 & 3.0 l

Section P

-_ age 5.0 CRITICALITY CONTROL ADMINISTRATIVE REQUIREMENTS..........................A-5-1 5.1 Proced ures..............................

......... A 1 5.2 Analysis Request...........

......... A 1 5.3 Criticality Control Analysis......................

........... A 1 5.4 Neutron Physics Advice...........

.......... A-5-2 5.5 Verification of Criticality Safety Analysis Results......

....... A-5-2 5.6 Reco rd s.......................

................ A 2 5.7 Criticality Control Inspectiora............................

............ A-5 -2 5.8 Training Program...................

............ A-5-3 I

5.9 Criticality Monitoring......................................

......................A-5-4 6.0 CRITICALITY CONTROL CONDITIONS - TECHNICAL AND AN A LYTI CA L RE QU IRE M ENTS................................................................................ A 1 6.1 General Requirements...

...... A 1 6.2 Calculative Methods..........

..... A 1 6.3 Normally Suberitical Values - Individual Accumulations.

... A-6-2 j

6.4 Normally Suberitical Values - Interacting Accumulations.

... A-6-3 6.5 Integrity of Structures.........

............ A-6-4 l

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 N uclear I solation.....................................

....... A-6-6 i

i 7.0 RADIATION CONTROL ADMINISTRATIVE REQUIREMENTS.............................A-7-I 7.1 Radiation Stacdards......

.... A 1 7.2 Radiation Control Inspection...........

..... A 1 7.3 Radiation Safety Training.

..A 7-1 8.0 RADIATION CONTROL CONDITIONS - TECHNICAL AND OPERATING RE Q UIRE M ENTS................................................................................................................ A I 8.1 Air Contamination Control.........

........ A-8 1 8.2 Instrument Capability......................

.......... A-8-2 8.3 Contaminatien Detection.......................

..... A-8-2 8.4 Sample Detection...-.

.......... A. 8-3 8.5 Exposure Detection.....

..... A-8-3 8.6 Instrument Calibration..

... A-8-3 8.7 InstrumeM Check Sources..

.... A-8-3 8.8 Liquid Wa te Disposal..

.. A-8-3 8.9 Airborne E@ent Control.

.. A-8-4 8.10 Contamination-Free Articles.....

.. A-8-4 8.11 Internal Exposure.......

..A-8-5 l

8.12 Personnel Contamination.

... A-8-5 i

License No. SNM-960 Docket No.70-751 Sect. No. Contents Page 11 AppendixA Date 8/23/99 Amends Sect.(s) 8.11 i

Section Eage Table 8.1," Equipment and Facility Design Criteria and Guidelines"........................A-8 6 Table 8.2, " Acceptable Surface Contamination Levels"............

.....................A-8-7 Table 8.3, " Acceptable Frequencies for Surveys"..............

......... A 8 Table 8.4," Recommended Action Levels for Removable Surface Contamination".........A-8-9 Table 8.5," Air Sampling Recommendations Based on Estimated intakes and Airborne Concentrations..........................................

................... A 8-10 9.0 O N-S IT E TRAN S FE RS....................................................................................................... A 1 9.1 Type A and Unirradiated SNM..........

................... A 1 9.2 TypeB........................................................................

..... A I i

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t l

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I l

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License No. SNM-960 Docket No.70-754 Sect. No. Contents Page iii AppendixA Date 8/23/99 Amends Sect.(s)

8. 0

1 i

LICENSE CONDITIONS FOR THE VALLECITOS NUCLEAR CENTER Set forth herein are the technical and administrative requirements proposed for governing the receipt, possession and use of the special nuclear material and associated byproduct material (as defined in Section 2.0 of this Appendix), subject to licensing and regulation by the Nuclear Regulatory Commission, in activities other than reactors at General Electric's Vallecitos Nuclear Center near Pleasanton, California.

1.0 AUTHORIZED ACTIVITIES The following types of activities are authorized. The authorized activities consist of research and development analyses, exploration or experimentation of licensed material. Fuel fabrication operations or activities are explicitly excluded from the authorized activities.

1.1 LABORATORY OPERATIONS 1.1.1 Chemical Analysis of the chemical and isotopic composition, concentration and behavior of special nuclear materials by wet chemistry and physical measurement techniques.

1.1.2 Metallureical Physical analyses and testing of physical and metallurgical properties of special nuclear

. materials.

1.1.3 Physics and Health Physics Measurements of radiation and its effects on instruments and on the structure and composition of materials.

License No. SNM-960 Docket No.70-754 Sect. No.

1. 0 Page A-1-1 AppendivA Date 8/23/99 Amends Sect.(s)

1. 0

1.2.4 Decontamination l

Decontamination of equipment and facilities.

1.3 WASTE TREATMENT l

1.3.1 Liouids l

Concentration of the radioactive constituents of liquid wastes by evaporation, chemical l

. treatment, sedimentation, filtration, and ion exchange; solidification and packaging of concentrates for disposal.

1.3.2 Solids l

Packaging and storage of wastes contaminated with or containing nonreclaimable special nuclear

. materials, excluding direct burial in soil.

1 i

License No. SNM-96),_

Docket No.70-754 Sect. No.._ :.0 Page A-1-3 AppenditA Date 8/23/99 Amends sect.(s) 1.2.4.1.3.

1. 3.1. & l. 3. 2

2.0 POSSESSION LIMITS 1

The following quantities of special nuclear materials are authorized at the Vallecitos Nuclear Center.

2.1 U-235

50 kg enriched to less than or equal to 10% for authorized activities. The material may be in the form of irradiated special nuclear material with its ettendant byproduct material and reactor-produced transuranics. The enriched uranium core of the Nuclear Test Reactor (NTR) may be temporarily stored and examined in Criticality Limit Areas in Building 102. Small quantities, s 100 grams, of U-235 enriched to more than 10% may be used in research and design activities. Use of greater quantities outside of the NTR requires NRC pre-approval.

2.2 U-235

4 kg enriched to more than 10% for authorized activities. The material may be in the fomt of irradiated special nuclear material with its attendant byproduct and reactor-produced transuranics.

2.3 Plutonium

100 grams in contained or sealed form in addition to the irradiated quantities as referenced in Subsections 2.1 and 2.2 above.

2.4

.U-233: 100 grams in any form, i

License No. SNM-960 Docket No.70-754 Sect. No.

2. 0 Page A-2-1 AppendixA Date 8/23/99 Amends Sect.(s) 2.1

l 3.0 DEFINITIONS

~

\\

l As used in the proposed license conditions set forth herein:

3.1 Area Manager means the lowest supervisory position fully responsible for the specific activity or function with which the term is associated. Tht generic term ' Area Manager" does not necessarily refer to the title of any specific position in General Electric's system of organization and position nomenclature.

3.2.Arrav means two or more interacting rissile material accumulations.

1 3.3 Criticality Area means any physically identified area or location within which fissile materials are handled under the direction of a single area manager. A criticality area may include more than one criticality limit area.

3.4 Criticality Control means the administrative and technical requirements established to minimize the possibility of achieving inadvertent criticality in the environment analyzed.

3.5 Criticality Limit Area (CLA) means a designated and physically identifiable locality within l

which a specific set of criticality control limits governs the use of fissile materials.

)

3.6' Fissile Material Accumulation means any single accumulation of_ fissile material; for i

exartiple, one can of UO powder, one can of fuel pellets, or one fuel element.

2 3.7 Full Reflection means that degree of reflection equivalent to a tight-fitting shell of greater than 12 inches of water.

l License No. SNM-960 Do/ st No.70-754 Sect. ho.

3. 0 Page A-3-1 AppendixA Date 8/23/99-Amends Sect.(s) 3.5 l

i

l 3.8 Homoceneous Svstem means a system containing compounds of fissile materials in the l

form of unifud; A stributed particles and fissile compounds in solution.

l 3.9 Heteroceneous System means a system containing compounds of fissile materials in the form of clumped fissile compcunds (e.g., fuel rods in water).

3.10 Normally Subentic s1 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 activity as it is conducted in the plant allowing for j

credible mishaps that could occur in the actual plant situation.

3.11 Nuclear Enerev Oneration means those components of ie '3eneral Electric Company engaged in the various aspects cf nuclear energy and does not refer to a specific component by title.

3.12 Nuclear Safety means that field of safety comprised of criticality safety and radiation safety.

3.13 Safe Accumulation means an accumulation of fissile material which is subcritical as described in Section 6.3 of this Appendix.

3.14 Safe Batch means an accumulation of fissil material which is 45% of the critical l

accumulation considering enrichment, fuli reflection, and optimum wau:r moderation consistent with the form of the material.

3.15 Safe Concentration means the maximum concentration of a fissile solution which is subcritical as described in Section 6.3.5 of this Appendix.

3.16 Safe Geometry means the maximum dimension (s) of a physical shape for fissile material which is suberitical as described in Section 6.3.4 of this Appendix.

3.17 Safe Mass means the maximum snass of fissile material in any form which is suberitical as described in Section 6.3.2 of this Appendix.

l l

License No. SNM-960 Docket No.70-754 Sect. No.

3. 0 Page A-3-2 AppendixA Date 8/23/99 Amends Sect.(s) 3.13-3.17

3.18 Safe Spacine means the minimum spacing of suberitical fissile accumulations which are nuclear isolated as described in Section 6.10 of this Appendix.

3.19 Safe Volume means the maximum volume of fissile material which is subcritical as l

described in Section 6.3.3 of this Appendix.

3.20 Site Safety Manacer means an individual designated by upper management as having the l

responsibility for nuclear safety at the entire VNC site.

3.21 SNM Custodian means an individual designated by an Area Manager who is responsible l

for maintaining the inventory of CLA's.

3.22 Special Nuclear Material means (1) plutonium, uranium 233, uranium enriched in the isotope 233 or in the isotope 235, and any other material which the Commission, pursuant to the provisions of section 51 of the act, determines to be special nuclear material, but does not include source material; or (2) any material artificially enriched by any of the foregoing but does not include source material.

3.23 Suberitical Area means a physically identified area or location involving special nuclear l

materials in quantities ofless than 500 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 suberitical 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 which meets the criteria of 10CFR70.24(a).

License No. SNM-960 Docket No.70-754 Sect. No.

30 Page A-3-3 AppendixA Date 8/23/99 Amends Sect.(s) 3 18-3.23

i 4.0 GENERAL ADMINISTRATIVE REQUIREMENTS 4.1 AREA MANAGERS Operations and activities in a specific criticality area or area where the licensed material will be I

used or stored shall be directed by the designated area manager. The responsibility for safe operation and control of activities in the area and for the safety of the environs as influenced by the activities conducted therein shan be vested in this position. An area manager shall be l

proficient in the application of the VNC radiation protection program as it relates to limitations and radiological controls on work activities in this assigned radiation or radioactive materials arcas. Additionally, each area manager of a criticality area sha! be proficient in the application of criticality control procedures and be knowledgeable in the procedures appli able to the criticality area under his management.

1 4.2 CRITICALITY SAFETY COMPONENT i

The Criticality Safety component is defined as that component with designated responsibility to i

provide authoritative professional advice and counsel to area managers on matters of control against accidental criticality and to measure the effectivuess of the criticality control program.

4.2.1 The functions of the Criticality Safety component shall exclude direct responsibility for operations involving the use of fissile materials, and the Criticality Safety component shall not report to an area manager responsible for an area where fissile material is handled other than a suberitical area.

4.2.2 The Criticality Safety component shall include at least one technically trained person with a bachelor's degree in science or engineering and three years experience in the nuclear field, including one year of directly relevant criticality safety experience. Should the Criticality Safety component lack the one year of e.xperience, all analys s for the interim period until the experience is achieved shall be verified by two of the individuals described in 5.5 or by the physics advice function in 5.4 of this Appendix. Criticality Safety staff with less than one year experience will be supervised by qualified Criticality License No. SNM-9Q Docket No.70-754 Sect. No.

4.0 Pcge A-4-1 AppendixA Date 8/23/99 Amends Sect.(s) 4.1 & 4.2.2 i

lW.

Safety. staff or the Manner, Regulatory Compliance. In the event new or changed

~

criticality analyses beyond the bounds of existing analyses are required, additional expertise would be obtained through other Company sources, or consultants.

~

4.2.3. The manager of the Criticality Safety component shall hold a bachelor's degree in science or engineering and have at least five years experience in a responsible position in a mzlear field such as engineering, physics, or chemistry. In addition, the manager will have at least three years experience in criticality safety functions or attend a formal criticality training class if new or changed criticality analyses beyond the bounds of the existing analyses are required.

4.3 RADIATION SAFETY COMPONENT l

The Radiation Safety component is defined as ther component with designated responsibility to l

l provide authoritative professional advice and counsel to area managers at the Vallecitos Nuclear Center on matters of radiation protection and ALARA and to measure the effectiveness of the radiation protection program.

4.3.1 The functions of the Radiation Safety component shall exclude direct responsibility for operations involving the manufacture of nuclear products or,nrocessing of nuclear materials. The Radiation Safety component shall be responsible t) establish and maintain the radiation safety program to ensure the protection of employees at the Vallecitos Nuclear Center and of the community. The radiation safety program shall include as a l

minimum: the evaluation of release of radioactive eff! cents and metorials from the site, establishment of procedures and training programs to control contamination and exposure to individuals, the review of calibration and maintenance activities for radiation detection instruments, maintenance of appropriate rece ds and reports, review of radioactive material handling practhes, and review of cisnge procedures (see Section 4.6), including ALARA considerations.

License No. SNM-960 Docket No.,70-754 Sect. No.

4. 0 Page A-4-2

-AppendixA Date 8)b.l 99 Amends Sect.(s) 4.2.2 & 4.2.3 j

h1

1 4.3.2 The minimum qualifications of personnel assigned functional responsibilities in the

~

Radiation Safety component shall be as follows:

a.

Manater: 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 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 criticality procedures.

4.4 VALLECITOS TECIINOLOGICAL SAFETY COUNCIL The functions of the Vallecitos Technological Safety Council (VTSC) shall include responsibility for review of reportable incidents and the nuclear safety program, for contributmg professional advice and counsel on 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 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 Site Safety Manager (SSM) (see Section 3.14) and retained two years. Should the VTSC choose not to meet in any calendar quarter, the SSM shall be notified in writing. The Vallecitos Technological Safety Council is an independent review body and shall consist of at least five senior members of General Electric's technical and/or management personnel appointed by the SSM and shall include competence in the physics, chemistry, engineering and radiation safety disciplines.

l I

P.

License No. SNM-960 Docket No.70-754 Sect. No.

4. 0 Page A-4-3 AppenditA Date 8/23/99 Amends Sect.(s)

4. 4

r' 4.5 COMPLIANCE M. fCY 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 procedures. VNC is committed to controlling activities involving licensed material in j

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 specifications which are of repetitive or serious nature are subject to disciplinary action.

Table 4.1.

Resoonsibility Matrix A - Prepare Regulatory B - Review Compliance Area Operating C - Approve 2

Function Managers Management Site Safety Standards' A,B,C A,B,C A

Nuclear Safety Procedures A,B,C 2

Operating Procedures B

A, C Change Authorization A, B A, C B

Emergency Procedures A,B,C B, C

' Standards include radiation safety, criticality safety, industrial safety, and environmental.

2Only those related to nuclear safety.

l

' Includes radiation, criticality, environmental, and hazardous materials safety functions.

License No. SNAf-960 Docket No.70-754 Sect. No.

4. 0 Page A-4-4 AppendixA Date 8/23/99 Amends Sect.(s)

4. 5

f 5.9 CRITICALITY MONITORING 1

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 Section 70.24(a)(1) of 10CFR70 is granted for the Building 102 storage pool and hot cells. At the high-level horizontal waste storage facility (bunker) and the storage pit in the Building 102 storage pool area, for the purpose of compliance with 70.24(a)(1) 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 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 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 of assemblies, providing no activities could cause rearrangement of fuel-bearing portions into more reactive configurations; or which meets the requirements of a soberitical area as defined in Section 3.17 of this Appendix.

License No. SNM-960 Docket No.70-754 Sect. No.

S0 Page A-S-4 AppendixA Date 8/23/99 Amends Sect.(s)

S. 9.2

)

7.0 RADIATION CONTROL ADMINISTRATIVE REQUIREMENTS 7.1 RADIATION STANDARDS 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.

7.2 RADIATION CONTROL INSPECTION Activities involving special nuclear material shall be inspected by the Radiation Safety component on a continuing basis. Conditions of an unusual or uncertain nature that could lead to radiological health and safety problems shall be referred to the area manager immediately for correction.

7.3 RADIATION SAFETY TRAINING 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 prior to starting work followed by additional training commensurate with the work environment as determined by the area manager. All radiation workers shall complete a formal training in radiation safety prior to working with licensed material independently.

Personnel working with licensed material or regularly assigned to work in radiation areas shall l

receive refresher training annually.

License No. SNM-960 Docket No.70-751 Sect. No.

7. 0 Page A-7-1 AppendixA Date 8/23/99 Amenas Sect.(s)

7. 3

F t

l 8.1.2 Smearable contamination in excess of 10,000 dpm/100 cm alpha in normal working 2

areas shall be cleaned up expeditously.

l l

8.2 INSTRUMENT CAPABILITY l

Dose rate range capabilities of portable instrumentation shall include the following:

1 to 50,000 mR/h (10-5,000 R/hr for environmental monitoring) gamma i

beta 4 to 200,000 mrad /h 1 to 5,000 mrem /h neutrons i

8.3 CONTAMINATION DETECTION l

8.3.1 Contamination detection capabilities of portable instrumentation shall be:

I alpha 200 to 300,000 disintegrations per minute l

twice background to 400,000 disintegrations per minute beta gamma on a Geiger-Mueller counter 4

8.3.2 Contamination detection capabilities of portable instrume.itation when used in the smear survey technique shall be 200 disintegrations per minute f ar alpha and 500 disintegrations per minute beta-gamma.

8.3.3 Detection equipment shall be maintained on site to meet the requirements of Section 8.10 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.

8.3.5 A routine radiological survey of work areas shall be conducted for determining alpha and beta-gamma surface contamination levels and external exposures.

License No. SNM-960 Docket No. 70 754 Sect. No.

8. 0 Page A-8-2 AppendivA Date 8/23/99 Amends Sect.(s) 8.3.5

I

\\

8.3.6 The frequency of the surveys will be established based on the values presented in Table 8.3, " Acceptable Frequencies for Surveys". The surface contamination action levels for these surveys will be based on the values presented in Table 8.4, " Recommended Action Levels for Removable Surface Contamination".

8.4 SAMPLE DETECTION Sample detection capabilities for laboratory analysis of effluents shall be no greater than the concentrations specified in 10CFR20, Appendix B, Table 2. The detection limits of the VNC Analytical Counting Lab are listed in Section 3.9 of Appendix B.

8.5 EXPOSURE DETECTION 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 manner as to record the maximum whole body exposure.

Dosimeters will be analyzed routinely on a schedule determined by the potential for exposure to penetrating radiation. Typically, radiation workers are assigned to a monthly routine and others are on a quarterly routine. The TEDE action level is 3.5 Rem. If the action level is exceeded, written approval by VNC management is required before additional radiation work will be permitted.

8.6 INSTRUMENT CALIBRATION Portable monitoring instruments shall be calibrated upon initial acquisition, after major maintenance, and at least annually.

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 Field check sources shall be available for use in functional response checks of portable radiation-measuring instrumentation.

License No. SNM-960 Docket No.70-751 Sect. No.

8. 0 Page A-8-3 AppendixA Date 8/23/99 Amends Sect.(s) 8.3.6 & 8.5

8.8 LIQUID WASTE DISPOSAL Potentially contaminated liquid wastes shall be collected, solidified, and disposed of as solid waste.

8.9 AIRBORNE EFFLUENT CONTROL Potentially contamiaated 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 2, 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, the 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 Release of equipment or materials for unrestricted use shall be in accordance with the attached

" Guidelines for Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use or Termination of Licenses for Byproduct, Source, or Special Nuclear Material"; April, 1993.

1 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 assume from the design and usage that no radioactive materials could have contaminated inaccessible surfaces without itaving contaminated 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 Docket No.70-751 Sect. No.

8. 0 Page A-8-4 l

AppendixA Date 8/23/99 Amends Sect.(s) 8.10 1

8.11 INTERNAL EXPOSURE Persons who. work routinely 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 air sampling, urine or whole body counting.

l 8.11.1 Bioassav When bioassay measurement results are used for determining workers' annual intake of j

radioactive material, the bioassay program shall be conducted in accordance with Regulatory

. Guide 8.9, Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay Program.

Bioassay data shall be evaluated by the radiation safety function.

i 8.11.1.1-Whole Body Counting: VNC employees shall receive a whole body count at least l

i i

annually. An investigation shall be initiated if a whole body count result exceeds 10% of an ALI.

8.11.1.2 Urinalysis:

Urinalysis may be performed on a job-by-job basis when radioactive l-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 until the project is completed.

All samples above background are investigated.

-8.11.2 Air Samoline When air sampling measurement results are used for determining workers' annual intake of radioactive material, the air sampling program shall be conducted in accordance with Table 8.5,

" Air Sampling Recommendations Based on. Estimated Intakes and Airborne Concentrations",

taken from Regulatory Guide 8.25, Air Sampling in the Workplace.

8.12 PERSONNEL CONTAMINATION 2

Personnel contamination shall be kept below 200 dpm/100 cm alpha. An individual whose skin or personal clothing is found contaminated above background levels shall not be allowed to exit a restricted area without prior approval of the staff of the radiation safety component.

License No. ' SNM-960 Docket No.70-754 Sect. No.

8. 0 Page A-8-5 Appendix A '

Date 8/23/99 Amends Sect.(s) 8.11-8.12

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TABLE 8A RECOMMENDED ACTION LEVELS FOR REMOVABLE SURFACE CONTAMINATION Type of Radioactive Materiala j

Alpha Emitters Low-Risk High Lower Beta or X Ray Beta or X Ray Toxicity Toxicity Emitters Emitters Type of Surface

( Ci/cm2)

(pCi/cm2)

( Ci/cm2)

1. Unrestricted areasb g0-7 10-7 to-6 10-6 2.

Restricted arease 10-4 10-3 10-3 10-2 3.

Personal clothing worn 10-7 10-7 10-6 10-6 outside of restricted areas 4.

Protective clothing worn 10-5 10-5 10-4 10-4 only in restricted areas

{

aHigh toxicity alpha emitters include Am-243, Am-241, Np-237, Ac-227. Th-230, Pu-242, Pu-238, Pu-240, Pu-239, Th-228, and Cf-252. Lower toxicity alpha emitters include those having permissible concentrations in air greater than that for Ra-226 (s) in 10 CFR Part 20, Appendix B, Table I, Column 1. Beta or x-ray emitter values are applicable for all beta or x-ray emitters other than those cons;dered low risk. Low-risk nuclides include those whose beta energies are less than 0.2 MeV, whose gamma or x-ray emission is less than 0.1 R/h at I meter per curie, and w hose permissible concentration in air in 10 CFR Part 20, Appendix B, Table I, is greater than 10 pCi/ml.

4 bContamination limits for unrestricted (non-contamination-controlled) areas in this table are considered to be compatible in level of safety with those for release of facilities and equipment for unrestricted use, as given in Regulatory Guide 1.86, " Termination of Operating Licenses for Nuclear Reactors", and in " Guidelines for Decontamination of Facilities and Equipment Prior to Release for Unrestricted Use or Termination of Licenses for Byproduct, Source, or Special Nuclear Material", which is available from the Division of Fuel Cycle and Material Safety, Office of Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, D.C.

20555.

cAveraging is acceptable over inanimate areas of up to 300 cm or, for floors, walls and ceiling,100 cm.

2 8

These limits are allowed only in those restricted areas where appropriate protective clothing is worn.

2 Note on Units: The above units of pCi/cm have been used in this table since they are consistent with units adopted as national standards in several other nations and the IAEA; the units of pCi and cm are already used to express concentration in 10 CFR Part 20, and they are readily convenible to SI units by the well-known relation: 1 pCi =

3.7 x 10' dis /sec = 3.7 x 10 Becquerels (Bq). Bey may also be easily convened to other frequently used units of 4

I radiation protection practice, i.e., disintegrations / minute per 100 cm = 2.22 x 10' x (activity expressed in pCi/cm ),

2 2

Note on Skin Contamination: Skin contamination should always be kept ALARA. Exposed areas of the body of persons working with unsealed radioactive materials should always be monitored and should be washed when any contamination is detected. It is imponant, however, that contaminated skin should not be so treated or scrubbed that the chance ofintake of radioactivity into the body is increased.

l License No. SNAf-960 Docket No.70-754 Sect. No.

8. 0 Page A-8-9 l

AppendixA Date 8/23/99 Amends Sect.(s)

Table 8.4

l TABLE 8.5 AIR SAMPLING RECOMMENDATIONS BASED ON ESTIMATED INTAKES AND AIRBORNE CONCENTRATIONS Worker's estimated Estimated airborne annualintake as concentrations as a fraction of ALI a fraction of DAC Air Sampling Recommendations

< 0.1

< 0.01 Air sampling is generally not necessary.

However, monthly or quarterly grab samples or some other measurement may be appropriate to confirm that airbome levels are indeed low.

> 0.01 Some air sampling is appropriate. Intermittent or grab samples are appropriate near the lower end of the range. Continuous sampling is appropriate if concentrations are likely to exceed 0.1 DAC averaged over 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months or longer.

> 0.1

< 0.3 Monitoring ofintake by air sampling or bioassay is required by 10CFR 20.1502(b).

> 0.3 A demonstration that the air samples are representative of the breathing zone air is appropriate if(l) intakes of record will be based on air sampling and (2) concentrations are likely to exceed 0.3 DAC averaged over 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months (i.e.,

intake more than 12 DAC-hours in a week).

Any annual intake

>1 Air samples should be analyzed before work resumes the next day when potential intakes may exceed 40 DAC-hours in I week. When work is done in shifts, results should be available before the next shift ends. (Credit may be taken for protection factors if a respiratory protection program is in place.)

>5 Continuous air monitoring should be provided if there is a potential for intakes to exceed 40 DAC-hours in 1 day. (Credit may be taken for protection factors if a respiratory protection program is in place.)

l l

License No. SNM-960 Docket No.70-751 Sect. No.

8. 0 Page A-8-10 AppendixA Date 8/23/99 Amends Sect.(s) Table 8.5 l

n Ie r

Section pgg 8.0 B UI L D IN G J 0 7..........................................................................

L 8.1 Location and General Description............................

................................8-1 8.2 Use............................................................................................................8-1 9.0 DELETED 10.0 CHEMISTRY, METALLURGY AND CERAMICS LABORATORY, BUILDING 103............................................................................................101 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 L

10.4 Building 102 Procedures........................

....................10-5 10.5 Building 103 Criticality Control........................................................10-7 Figure 10.1 Building 103 Floor Plan (Ground Level)...................................................10-8 10.2 Building 103 Floor Plan (Second Floor)..............................................10-9 10.3 B uilding 103 Vault Layout........................................................................... 10- 10 11.0 DELETED 12.0 BUILDING 105.........................................................................................12-1 12.I Location and General Description.............................................................12-1 d

12.2-NTR Facilities................................................................................. 12-1 12.3 Other Laboratory Areas (Leased Facilities).................................................12-1 Finure 12.1 B uilding 105 Floor Plan...................................................................... 12-1 13.0 WA STE H AND LIN G.............................................................................. 13 - 1 13.1-Radioactive Liquid Waste Evaporator Plant (Building 3 4 9)....................... 13 - 1

-13.2 S olid Waste H andlin g................................................................................ I 3 -4 Ficure 13.1 Waste Treatment Plan t.............................................................................. 13 -6 13.2 Hillside Waste Storage....................................

...................................13-7 L

\\

l l

- License No. SNM-960 Docket No.70-751 Sect. No. Contents Page lii l

Date 8/23/99 Amends Sect.(s)

I2.3 l

h

1.3 GENERAL PLANS AND USES OF SPECIAL NUCLEAR MATERIALS This application requests authorization under Title 10, Code of Federal Regulations, Part 70, to receive and possess the special nuclear material designated in Section 1.4 herein; to receive and possess the special nuclear material and associated byproduct material produced by the irradiation thereof; and to use said special nuclear materials in research and development activities as defined in Section 70.4, in chemical and physical analysis, and examination and investigation of nuclear fuels, associated materials and devices at the Vallecitos Nuclear Center.

1.4 SPECIAL NUCLEAR MATERIAL POSSESSION LIMITS 1.4.1 Vallecitos Nuclear Center The special nuclear materials used in connection with activities authorized by License SNM-960 at the Vallecitos Nuclear Center will not at any time exceed those limits listed in Section 2.0 of l

Appendix A to License SNM-960.

1.4.2 Form and Enrichment Specifications j

)

The majority of the Vallecitos Nuclear Center activities are conducted in facilities and under procedural controls which accommodate any chemical or physical form and any U-235 isotopic 1

content. If specific limitations are placed on these parameters in connection with an individual activity in order to assure the radiation or nuclear safety of that work, the limit is described in the appropriate section of this application entitled, " General Plans and Uses of Materials".

l License No. SNM-960 Docket No.70-754 Sect. No.

1. 0 Page 1-3 Date 8/23/99 Amends Sect.(s)

1. 4. I

1.5 PRINCIPAL VALLECITOS FACILITIES Descriptions of the principal buildings and laboratories in which special nuclear materials are used at the Vallecitos Nuclear Center site are set forth in this section with the primary objective of general orientation. The locations of these facilities are shown in Figure 1.2. The specific activities conducted in each of these facilities and their safeguards equipment and procedures are l

discussed in later sections.

1.5.1 Radioactive Materials Laboratory l

The Radioactive Materials Laboratory (RML) is located in Building 102. This laboratory is a shielded facility equipped with remote manipulators to conduct experiments and analyses with irradiated reactor fuels and other radioactive materials. The facility also includes the Building 102 storage pool and dry pit storage.

1.5.2 Radiochemistry Laboratory Adjacent to RML, on the main floor of Building 102 and providing analytical support to it, is a radiochemistry laboratory equipped with standard chemical and radiochemical apparatus. This laboratory primarily is used to ansJyze samples of materials prepared in the RML.

1.5.3 Metallurev. Chemistry, and Ceramics Buildine A second major laboratory building in the 100 Area is the Metallurgy, Chemistry, and Ceramics i

Laboratory (Building 103). This two-story building consists oflaboratories, variously equipped with laboratory apparatus designed to handle moderate quantities of radioactive materials, and offices. The functions served by this facility are research, development, and analytical chemistry services.

License No. SNM-960 Docket No.70-754 Sect. No.

1. 0 Page 1-4 Date 8/23/99 Amends Sect.(s)

1. 5.1

1.5.4 Buildine 105 Just north of Building 102 is Building 105. The principal facilities located in this building are an experimental reactor (the Nuclear Test Reactor) and laboratories leased to another company. The laboratories in Building 105 use only minute quantities of special nuclear materials, which are possessed and controlled by GE.

1.5.5 Engineerine Shop Building 106 contains various maintenance shops (e.g., machine, carpentry, electric, and instrument calibration facilities).

1.5.6 Solid Radioactive Waste Storage Facility Solid radioactive wastes generated at the various laboratory and facility locations are stored in the solid waste storage facility located approximately midway between the deactivated Vallecitos l

Boiling Water Reactor (VBWR) and General Electric Test Reactor (GETR) areas. This storage j

area includes shielded horizontal tubes for storing 5-inch and 7-1/2-inch-diameter waste liners.

1.5.7 Waste Evanorator Plant The Waste Evaporator Plant is located adjacent to the deactivated VBWR site. This plant is used to concentrate and solidify liquid radioactive wastes generated at the Vallecitos Nuclear Center or other licensed facilities prior to transfer to authorized waste disposal firms or waste burial sites. Such wastes may contain minute quantities of special nuclear material.

1.5.8 Reactors and Auxiliary Facilities The ESADA-Vallecitos Experimental Superheat Reactor (EVESR), the VBWR, and the GETR are deactivated.

License No. SNM-960 Docket No.70-734 Sect. No.

1. 0 Page 1-5 Date 8/23/99 Amends Sect.(s)

1. S. 6

(

b.

Industrial Safety and Hveiene Function. Develop programs to protect the employees from industrial hazards, including operation of medical and safety education programs (located at San Jose).

c.

Environmental Safety Function. Develop programs to protect the employees and the general public from exposure to hazardous materials and assure proper disposal of hazardous wastes (located at VNC and San Jose).

2.6 TECHNICAL PERSONNEL CAPABILITIES The primary responsibility for operational radiation safety for the operations conducted in the various Vallecitos facilities involving special nuclear material rests with the supervisor or manager of each facility. Equally important are the knowledge and experience of personnel in the Nuclear Safety function. Since the issuance of License SNM-960 in 1966, the Commission's Regional office (Region IV and Region V) has inspected VNC to assure that adequate levels of l

technical expertise are maintained in all positions. Rssumds for key personnel are included as Addendum A to this section.

2.7 IMPLEMENTATION OF CRITICALITY CONTROL PROGRAM The program for protection against accidental conditions of criticality is implemented by means of functional responsibility assignments.

Managers whose operations require the use of quantities of special nuclear materials approaching a theoretical minimum critical mass, or greater, are responsible for integrating and measuring the efforts of line and staff participants in this program. The principal participants and their responsibilities are outlined below.

License No. SNM-960 Docket No.70-754 Sect. No.

2. 0 Page 2-4 Date 6/7/99 Amends Sect.(s)

2. 6

1 i

2.9 TRAINING PROGRAMS l

All radiation workers shall complete a formal training in radiation safety prior to working with licensed materialindependently.

l l

2.9.1 Radiation Safety l

y Every new employee at VNC who routinely handles radioactive materials normally receives 9 radiation safety orientation, "New Employees Radiological Safety Orientation", within thirty days of reporting to the site. If an employee's area manager determines that the employee will be exposed regularly or frequently to radiation and/or radioactive materials, the employee is instructed in radiation protection such that he is able to protect himself and is made awre of the degree of hazard involved. A training program entitled, " Radiological Safety At Vallecitos Nuclear Center", is completed by each such employee normally within one year of his starting date at VNC. This course is scheduled and conducted by the Radiation Safety component.

The training course includes the following elements:

)

a.

basic principles of radiation safety,

. b.

Company policies and operating procedures,

radioactive materials handling methods and shielding requirements, c.

d.

emergency procedures, e.

requirements of NRC regulations, and f.

NRC license requirements.

Follow-up training commensurate with the work environment and the employee's work performance is determined by employee supervision. The employee also receives on-going

- training in the form of on-the-job demonstrations, periodic safety meetings, etc.

Employees whose work assignments may include the need for the use of respiratory protection equipment also receive the " Respiratory Protection Training Course" (RPTC). The course

provides both instruction and hands-on experience in the proper use and fitting of the respiratory protective equipment. Note: Respiratory protection equipment is worn only as a precaution.

VNC does not assign protection factors or take credit for License No. SNM-960 Docket No.70-754 Sect. No.

2. 0 Page 2-9 Date 8/23/99 Amends Sect.(s)

2. 9

I-respiratory protection for routine work. Should respiratory protection be mandatory for a given job, the Change Authorization will assure compliance with 10CFR20.

2.9.2 Criticality Every new employee who will work regularly with licensed material in areas in which quantities l

of special nuclear material capable of forming a critical mass (considering such criteria as existing moderation, reflection, enrichment, etc.) are present is instructed in the principles of criticality safeguards and is made aware of the degree of hazard involved. This instruction is normally completed within one year after the employee's starting date at that facility. Untrained employees handling licensed material will be supervised by an employee trained in criticality safety.

I Area supervision is responsible for informing all personnel at work or otherwise present in their I

area of specific procedures for criticality control and for appropriate administrative action to assure compliance with these procedures.

2.9.3 Other Proerams

' Radiation monitors receive the " Radiation Monitoring Technicians Certification Course". This is an intensive training program designed to qualify participants as VNC-certified Radiation

~

Monitoring Technicians.

Other courses are available from the Radiation Safety component on an as-needed basis.

2.9.4 Records Records ofindividuals receiving training courses are kept in a central computerized record file.

1 1

l l

License No. - SNM-960 Docket No.70-754 Sect. No.

2. 0 Page 2-10 I

Date 8/23/99' Amends Sect.(s) 2.9.2

I

'2.10.2 PurDOSe

'Any addition, alteration, deletion or substitution which adds a new capability, performs a different function, modifies performance characteristics, or introduces a hazard not previously l

analyzed requires an independent review by use of the Change Authorization (CA) procedure. A l

Change Authorization is prepared whenever the work involves changes to:

1 Facilities, equipment, or processes so that safety or regulatory compliance considerations i

a.

l differ from those previously an alyzed.

l l'

b.

Radioactive materiallimits, Hazardous or potentially hazardous industrial materials where such change is significant in c.

4 terms of quantities or use.

The independent review ofitems is conducted so that the hazards (both direct and indirect) of the proposal are recognized and appropriate safeguards are provided to eliminate or reduce the probability and ' severity of potential accidents.

While procurement, fabrication, selective installation-or testing, etc., may proceed prior to the final CA review and approval, actual i

implementation of the proposed change should not proceed until this review and approval are received.

l l

The Change Authorization is processed in accordance with a written procedure and reviewed by the Nuclear Safety component and by the Industrial Safety component as appropriate.

l l

License No. SNM-960 Docket No.70-731 Sect. No.

2.0 Page 2-12 l

Date 8/23/99'

. Amends Sect.(s) 2.10.2' l

The facility has a horizontal waste storage facility for storage of high-level radioactive material contained in sealed encapsulations called " waste liners". The horizontal waste storage facility is made of two rows of 40-foot-long concrete-lined steel pipes mounted horizontally and covered with earth. Eleven of the pipes have a 6-inch inside diameter, and seven have a 10-inch inside diameter. The pipes in either row are spaced on 3-foot centers, and the rows are spaced 3 feet apart with the pipes in the bottom row offset halfway between the pipes in the upper row.

Shielding is provided on the top and sides of the facility by a minimum of 6 feet of compacted earth. Shielding at the front and back ends consists of 3 feet of concrete in which the pipes have been anchored, plus concrete-filled step plugs with a minimum of 3 feet of concrete shielding in the plug. Additional aboveground space for lower level waste or other materials is available within this fenced and posted facility.

The facility is covered by a prefabricated metal building. Ventilation (nominal 27,000 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.

3.5.2 Liauld Wastes 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 plant for concentrating and solidifying the liquid wastes which are described in Section 13 of this application.

Other liquid waste (excluding sanitary waste) flows through a separate piping system into any three of four 60,000-gallon retention basins. After sampling and determining that radioactivity, if any,is within permissible discharge levels (3.0 x 10 Ci/ml gross alpha and 5.0 x 10 pCi/ml 4

gross beta-gamma), the water in the basin is released.

Sanitary wastes are treated, and the waste waters are sprinklered on site.

i LicenseNo. SNM-960 Docket No.70-754 Sect. No.

3. 0 Page 3-4 Date 8/23/99 Amends Sect.(s)

3. S.1

r 3.6 EMERGENCY EQUIPMENT A vehicle is available to Radiation Safety and can be equipped quickly with a supply of protective clothing, first aid equipment, respiratory protection equipment, and portable instrumentation and sampling equipment for use during emergencies. Emergency equipment also is stored in selected areas on site, 3.7 INDUSTRIAL SAFETY EQUIPMENT In conjunction with the radiation safety program at VNC, industrial health and safety of VNC personnel also are emphasized. Some of the protection facilities and equipment which are available include portable extinguishers, sprinkler systems, and a wide range of typical industrial safety equipment.

l 3.8 CRITICALITY ALARM SYSTEMS In any Vallecitos Nuclear Center area in which licensed material containing more than 500 grams l

of U-235 is used or stored and does not otherwise qualify as a "subcritical area" as defined in Section 3.23 of Appendix A, a monitoring system, including gamma-or neutron-sensing devices l

which will energize an audible alarm in the event of criticality, is installed and maintained.

Areas which are excepted from criticality alarm monitoring are described in Section 3.8.2 of this Appendix. The system in use on site is described in the following paragraphs.

3.8.1 Gamma Detection System This monitoring system consists of three commercially designed and manufactured gamma detectors which monitor each designated area. Two of the three detectors which are subjected to a dose rate in excess of preset alarm points will cause an alarm condition. Failure of any detection circuit component which would prevent criticality detection activates a warning light on the unit. Failure of any signal-producing component is detected during the monthly test.

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The system is capable of energizing the alarm when the radiation level at a distance of 2 meters j

from the special nuclear material is equivalent to 20 rads of combined neutron and gamma radiation within one minute.

License No. SNM-960 Docket No.70-754 Sect. No.

3. 0 Page 3-5 Date 8/23/99 Amends Sect.(s)

3. 8

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Sensing devices are positioned within 120 fer(, air equivalent, of every required location where special nuclear material is handled, used or stored. The system is tested by exposing the detectors to appropriate sources and sounding the alarm monthly. The alarm system is designed so that me alarm cont nues to sound until reset. The alarm is clearly audible in all locations i

where radiation exposure may result from an accidental criticality incident. If a facility does not have emergency backup power, all movements of SNM are suspended dudng a power failure.

3.8.2 Excepted Areas Criticality sensors are not provided below the surface of the water in the Building 102 storage pool or within the RML cells, nor in the horizontal waste storage facility. Shielding surrounding the special nuclear material is as follows:

Building 102 Storage Pool 16 feet of water l

RML Hot Cells 1.5 to 3.0 feet of high-density concrete Horizontal Waste Storage 5.5 feet of compacted earth m

Licente No. SNM-960 Docket No., 70-7 i?

Sect. No 30 Page 3-6 Date R/?L/9_9 Amends Sect.(s) 3R2

4.0 RADIATION PROTECTION PRGCEDURES A system of Vallecitos Safety Standards establishes the site radiation and criticality protection and regulatory compliance programs. The manager of the Nuclear Safety component issues the standards with review and comment of the managers of the major organizational components located on the site.

l Curr ntly, there are about 40 st ndards dealing with radiation protection matters. The principal features of these are summarized below.

4.1 PERSONNEL WORK RULES l

Requirements are established to prevent or minimize the hazards of radioactivity and radioactive materials. Food storage and consumption (including candy or beverages), the use of cigarettes, or the application of cosmetics are prohibited in posted ra<lioactive materials areas where removable contamination is present on surfaces of floors, walls, plant equipment, or furniture.

Approval by persons responsible for radiation protection may be granted for these activities in a l

posted radiation area which is shown by survey to be free from general removable contamination and conditions are unchanging. Food containers may not be used for storing or handhng radioactive material.

General Electric furnishes protective clothing for service in areas where contamination is likely to contact personal clothing. Protective clothing standards are set by the site Radiation Safety function to atsure effective quality, positive identification, and to avoid use for other than its intended purpoe. The amount and type of clothing for any specific activity are assessed on the basis of potential for personnel contamination.

4.2 LIMITS OF RADIATION IN CONTROLLED WORK AREAS All Vallecitos locations where there is a potential for radiation exposure are classified (radiation area, high radiation area, etc.) in accordance with the definitions of 10CFR20, Sections 20.1003, 20.1502,20.1601, and 20.1602. General Electric's philosophy of protection is to keep radiation exposure at the lowest reasonably achievable level in.11 cases.

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l License No. SNM 960 Docket No.70-754 Sect. No.

4. 0 Page 4-1 Date 8/23/99 Amends Se.1 4.1 L

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With resp;ct to operations which could pri.Jnce airborne radioactive contamination, managers of facilities are responsible for providing ventilattu ' equipment to meet the concentration limits of 10CFR20.1502 without the necessity, or credit, for personal respiratory devices during routine operations. In certain nonroutine situations where adequate ventilation equipment could not ensure control of airbome material, respiratory protection of demonstrated integrity is utilized.

In such cases, individual respiratory protection equipment shall be used in accordance with 10CFR20.1703.

The respiratory equipment currently in use at VNC is approved by the National Institute of Occupational Safety & Health (NIOSH) and, as such, achieves compliance with 10CFR20.

U.S. Divers Company's Survivair self-contained breathing apparatus (SCBA) or other NIOSH-approved SCBA having a backpack air supply, hose, harness, pressure-demand regulator, and a facepiece may be wom in emergency situations.

VNC has adopted only equipment that is approved by NIOSH.

No individual will be permitted to work more than 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months per week under conditions requiring masks. No individual will be permitted to work in a mask until he has received a medical clearance for respirator use and has been thoroughly instructed in methods of proper use, fitting, and field testing of respirators.

4.3 PERSONNEL MONITORING Instructions for the use of film and TLD badges, finger TLD dosimeters, and pocket dosimeters include the proper part of the body on which the device is to be worn and procedures to prevent spurious readings. In addition, personnpl are instructed to use monitoring instruments upon leaving a radioactive materials area.

Hand-and-shoe counters are also provided at some locations.

I License No. SNM-960 Docket No.70-754 Sect. No.

4. 0 Page 4-2 Date 8/23/99 Amends Sect.(s)

4. 2

I Neutron instrument calibrations are performed by measuring a response to a pulse c.

generator and to an Am-Be neutron source. Altematively, calibration using a moderated Cf-252 source will be perfonned.

d.

Gamma calibrations using Cobalt-60 sources standardized with a meter which, in turn, was calibrated with traceability to the National Bureau of Standards.

All radiation monitoring instruments are calibrated as frequently as deemed necessary to assure reliability during use. Portable radiation monitoring instruments are calibrated on an annual basis, before initial use, and after repair..

Stack particulate monitor systems are calibrated by placing a uniformly distributed radioactive source in the same geometry as the filter pape used for collecting particulates. Stack gas monitor systems are checked routinely by placing a reference gamma source on the side of the Kanne chamber and observing whether the response falls within prescribed limits. Calibration with a known radiogas standard has been performed to verify this procedure. Iodine monitors also are source checked routinely.

1 4.5 POSTING AND LABELING i

Instructions are established implementing the posting and labeling requirements of 10 CFR Part

20. Additional precautionary signs may be utilized to meet special requirements and detailed procedures. The area supervisor is responsible for maintaining the proper posting and labeling.

License No. SNM-960 Docket No.70-754 Sect. No.

4. 0 Page 4-5 Date 6/7/99 Amends Sect.(s)

4. 5

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7.3.9 Servicine Areas and Eauinment The Radioactive Materials Laboratory section of Building 102 also contains several shielded cells used in non-SNM work, a waste analyzing area, e equitment decontamination room, a machine shop, equipment storage areas, t,nd a manipulator repair room.

A hot shop facility now occupies the former Plutonium Analytical Laboratory C.R.) next to the radiochemistry laboratory. The PAL was surveyed completely (except for the area oenid a metal security wall) and decontaminated following the discontinuance of plutonium operations.

The area now is used for the repair of contaminated equipment. Only traces of SNM are present in the contamination.

7.3.10 Criticality Alarm Sensors The Radioactive Materials Laboratory is monitored for criticality accidents by three detectors located in the Building 102 storage pool area (Figure 6.2).

l 7.4 RML CRITICALITY CONTROL SYSTEM Special nuclear materials used in connection with RML operations are principally in the form of oxides in irradiated fuel elements and experimental capsules. The spectrum of material and activity types may be quite broad. This required flexibility has been taken into account in the establishment of criticality controls and is reflected in considerably larger safety margins than might be appropriate to more routine or repetitive situations. For example, each CLA in each fuel exa nination cell is limited to 45 percent of a critical number of units (fuel rods, assemblies, etc.), ar,3 each fuel examination cell is limited further so that criticality is not possible if all of the fissine material in the cell comes together simultaneously under conditions of optimum water moderation and full water reflection; but normal activities preclude moderation to any degree and i

License No. SNM-960 Docket No.70-751 Sect. No.

7. 0 Page 7-6 Date 8/23/99 Amends Sect.(s) 7.3.10 l

13.2.1 Solid Waste Accumulation Solid wastes are accumulated at each location wherc radioactive materials are handled. The majority of wastes fits the Low Specific Activity (LSA) category as defined in the Department of Transportation regulations.

For each waste accumulaiion, the generating component is responsible for maintaining a listing of all material in the accumulation. Each accumulation with its listing is forwarded to the WHF for final inspection and/or repackaging.

he WHF imsponsible for implementing waste vohune reduction methods, as appropritte.

13.2.2 du'S Was_te Stigge Solid waste materials are stored in the site radiocctive materials storage facility described in Section 3.5.1 and shown on Figure 13.2.

Waste materials that include or are associated with significant quantities of special nuclear materials are placed in containers called waste liners. Waste liners are stored in the horizontal waste storage facility bunker. Fifty-five-gallon drums and boxes containing lower level wastes l are stored in a covered facility.

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Limits on the maximum quantity of special nuclear material that may be loaded into any containers for puiposes of waste storage have been established by Nuclear Safety. For 55-gallon drum storage, calculations were made assuming optimum water moderation and spherical geometry for the individual masses within each drum. No credit was taken for neutron absorption by the materials between the individual units. On this basis, limits were established which provide criticality safety for an essentially infinite array of these drums. This limit is set l

in this fashion since it is possible to store drums above grade in almost any geometry. However, normal storage arrangement would be a planar array one drum high.

i License No. SNM-960 Docket No.70-754 Sect. No.

13.0 Page 13-5 Date 8/23/99 Amends Sect 4) 13 2.2

v t e Letter Sequence Response to RAI
TAC:L31006, (Approved, Closed)
Initiation Acceptance... Administration Meeting Questions Answers 1 September 1999 Results Approval, Approval Other: ML20217Q036
MONTHYEAR1999-07-26 [Table View]

ML20211N739

Text

References:

Dear Sir:

Subject:

Response

2.1 U-235

2.2 U-235

2.3 Plutonium

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