ML20210P379
| ML20210P379 | |
| Person / Time | |
|---|---|
| Site: | 07000734 |
| Issue date: | 02/09/1987 |
| From: | Ketzlach N NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | Crow W NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| NUDOCS 8702130375 | |
| Download: ML20210P379 (10) | |
Text
MY FEB 0 91987 w5 MEMORANDUM FOR: W. T. Crow, Acting Chief Uranium Fuel Licensing Branch Division of Fuel-Cycle and Material Safety, NMSS FROM: Norman Ketzlach Uranium Process Licensing Section Uranium Fuel Licensing Branch Division of Fuel Cycle and Material Safety, NMSS
SUBJECT:
OPERATIONAL SAFETY ASSESSMENT OF GA TECHNOLOGIES, INC.
(GA, DOCKET NO.70-734), JANUARY 12-16, 1987 I. Purpose The purpose of the in-depth assessment is to review all safety aspects of the operations of licensed fuel fabrication facilities and to determine the opera-tional effectiveness. Based on " lessons learned" from the January 4, 1986, rupture of the uranium hexafluoride cylinder at the Sequoyah Fuels Corporation Facility near Gore, Oklahoma, a recommendation was made to review the non-nuclear (e.g., fire protection and process) as well as the nuclear (e.g., radiation,
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nuclear criticality, and process) safety of the operations at uranium fuel cycle.
facilities.
II. Review Team Members of the assessment team were:
A. Region V - NRC
- 1. Buddy L. Brock, Team Leader
- 2. Joseph F. Pang, Radiological Safety
- 3. Phillip M. Qualls, Fire Protection
- 4. Andrew L. Hon, Electrical Engineering B. Headquarters - NRC
- 1. Glen L. Sjoblom, Chief, Safeguards and Materials Program Branch (ManagementControls)
- 2. Norman Ketzlach, Nuclear Criticality Safety C. EPA David Duncan, Radiological Safety 8702130375 870209 PDR ADOCK 07000734 C PDR f
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-1 D. California OSHA
- 1. Isaac Chae, Industrial Hygiene ,
- 2. Silvia Shattuck, Industrial Hygiene
- 3. Don Nielson, Industrial Safety s
E. FEMA
- 1. Joe Dominguez, Emergency Preparedness
- 2. William Patterson, Hazardous Materials F. California Office of Emergency Services (0ES)
As a team, we met with GA management as well as those more.directly involved in the operations. This report will deal with those contacts that are more directly related to the nuclear criticality safety assessment.
III. Personnel Contacted Keith E. Asmussen, Manager, Licensing, Safety, and Nuclear Compliance Vladi Malakhof, Manager, Nuclear Safety Allen Baxter, Criticality and Radiation Safety Committee, Independent Nuclear Criticality Safety Reviewer Robert Rucker, Nuclear Criticality Safety Analyst R. K. Krueger, Supervisor, TRIGA Fuel Fabrication R. Vanek, Supervisor, HTGR Fuel Fabrication Reuben De Valasco, Supervisor, North Warning System Fuel Fabrication Laura R. Quintana, Manager, Health Physics Chet Wisham, Nuclear Material Accountability IV. Plant Operations A tour was made of the areas covered by the SNM-696 license. These included the TRIGA and HTGR fuel fabrication facilities, the R&D facilities in which the i principal process steps for the North Warning System test fuel elements were l performed, the Hot Cell Facility, and the Liquid Waste Treatment Facility.
- None of the operations were being performed during tours, however, the
[ processes used were described.
l l All operations were identified with work and storage station limit signs, and procedures for each step of the operation were available to the operators. .
- Several poor practices were noted
- (1) minimum qualifications for independent nuclear criticality safety evaluation reviewers were not always met, (2) the nuclear criticality safety group does not check the work station limit signs
' prepared by the manufacturing personnel after the related work authorization is. approved, (3) traceability and/or availability of nuclear safety analyses
, (evaluation) for each process step were not demonstrated, (4) SNM storage areas are used for non-SNM material without signs indicating such use, (5) distribu-tion of CRSC annual audit reports should be standardized, and (6) type of work station limit signs should be standardized. These practices will be discussed below under Nuclear Criticality Safety (functional assessment of programmatic operational controls).
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W. T. Crow 3 V. Nuclear Criticality Safety A. Analyses The process for approval of. procedures was reviewed. The-following are the principal steps in the~ review process:
- 1. Informal meeting with Manager, Nuclear Safety, to discuss proposed operation and applicable nuclear criticality safety requirements.
- 2. Manufacturing (user) personnel prepares draft of planned procedures.
- 3. Review and evaluation (includes detailed computer analysis if required) made by Manager of Nuclear Safety.or staff meeting minimum qualifications of Manager, Nuclear Safety (license requirement). If the original analyst does not meet these minimum qualifications, he performs the required analysis under the guidance of a qualified reviewer.
- 4. After the qualified reviewer approves the analysis, the qualified nuclear criticality safety representative on the CRSC performs the final independent review. His recommendations may require further analysis as.
in Step 3 above.
- 5. If the proposed procedure meets the nuclear criticality safety requirements, it is reviewed by other groups as necessary for radiation safety and other related requirements (e.g., accountability), including the CRSC.
- 6. The manufacturing group then proposes a Work Authorization (WA) based on the criteria and procedure approved.
- 7. The work authorization is then reviewed and approved (when all requirements are met) by Health Physics, Nuclear Safety, Material Control, Licensing, and the CRSC.
8 The WA is then released by the appropriate manufacturing group for imple-mentation.
- 9. The work station limit signs are then prepared by the manufacturing group.
It was learned that nuclear safety personnel do not routinely review the work station signs after they are prepared.
A spot check of the nuclear safety evaluations.and their reviews indicated the following:
- 1. The first independent reviewer does not always meet the required minimum
. qualifications. Independent reviews made by R. Rucker and E. Hettergott may have been made by personnel who do not meet the minimum qualifications.
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4 2.-. ~ The : nuclear safety evaluation and their GA independent reviews of the "10-way splitter" were reviewed in detail and discussed with Allsn Baxter,. '
Vladi Malakhof, and Robert Rucker. The following assumptions made in the evaluation and.their apparent' agreement by both independent reviewers appear
- to be non-conservative for the following. reasons:
'a . - Water was assumed to have no. greater effect on neutron reflection than concrete.
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- b. -10 cm of water.was. assumed to ,be equivalent. to that of an infinite water reflector. ,
- c. Although the height of feach of the 10-sample collectors is 8 inches, it was assumed theyfcould-be: filled to a height of only approximately 3.7 inches. .
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- d. The top of the assembly'was assumed to be bare (apparently no consideration was given to reflectors'in -the vicinity of .the.
. equipment). The license requires full reflection be assumed unless the reflectors can-be specifica1.ly excluded. As a~ result of the meeting,' Baxter requested the GA nuclear: safety group perform additional.' analyses. Although the safety of the equipment remains in doubt, based on the above,.GA had earlier determined.the equipment-would_not be required. The NRC. reviewer, however, emphasized the-importance of "outside of reactor" experience needed by the reviewers to perform an appropriate nuclear safety review.
It is' recommended.that (1) the personnel involved in nuclear criticality safety.
= analyses maintain knowledge of up-to-date nuclear criticality safety criteria, and (2) they become more familiar with the SNM-696 license requirements.
B. Work Station Postings A review of the adequacy, maintenance, and surveillance of the nuclear safety i controls indicated the following:
- 1. Work authorizations are kept current, documents are controlled,.and out-dated work authorizations are removed from circulation, and proc'edures are available to the workers. Manufacturing, safety, licensing, and materials control personnel are aware of WA. changes in a timely manner.
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- 2. TRIGA~ Fuel Fabrication- Area'
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- a. -At TRIGA Work Station 14, theclimit sign allows'the ha'ndling,of.one F 1
- " meat" with a maximum of 1.0 kg U-235, including chipsiatithe station. ' '"
Since the meat may contain:as.little as 10 gm U-235,;the sign would indicate 990 gm U-235...in,the form of. chips could be at the' station... "
The WA restricts'the ~ quantity of U-235 at the. station :in the form of , ' O chips to 350 gm.' In addition, there was a 5-gallon bucket for:pla'stic' #: 'A shoe covers at,the station. The bucket and 990 gm U-235lin the.fo m _ N of chips plus water could possibly be made criticalmSeveral otherl A , 4; l
machinina stations were simi.arly misleading. A recommendation"was-
- VM-made- to]1) replace' the signs at' these stations with ones limiting the
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quantity of chips at the stations to 350,gm contained U-235'in the s ,
form of chips,-(2) remove the bucket from Station:14,.(3) require a' '
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, member of the nuclear safety group to review all work station signs - '
before they.are:placedlat the station to confirm they meet the -
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criterion in. the. nuclear safety analysis and.in ;the WA, and (4)' prohibit. /
'the use of waste containers of unsafe. geometry at work stations.',,The. , e above-recommendations were made during one morning of the assessment %
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period and recommendations-(1) and (2) were. implemented thersamel . .
afternoon.
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- b. The end section of shelves in the SNM storage vault (next to the .
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concrete wall farthest from the entry door) was used for non-SNM ,
, materials. Although the containers were of a different~ size than . -
those containing SNM and were labelled, it is good practice to;
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, a.E identify this section of shelves limiting the'use of the shelves to c. '
non-SNM materials. In addition, it was indicated that 55-gallon .
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. drums containing SNM are stored periodically between,this'section of; '
shelves and the concrete' wall. Since this.is only authorized when '
the adjacent section of shelves contain no SNM, the sign limiting the use of the shelves to non-SNM materials would improve the: '
r-administrative nuclear criticality safety controls for the storage 1 of the 55-gallon drums. . ,,
- 3. SV-A Area (HTGR Fuel Fabrication and Scrap Recovery)
Several work station limit signs were noted in the SV-A area that'did not '
conform to the standard formal signs used at practically all work stations. W ~'.
Station'A17-2B (Inert Box, Miscellaneous) contained a limit sign that was hand-written (black ink on yellow background). A similarzlimit sign was observed at _ s .
Station F4-3. It was recommended that use of such non-standard signs be revieweda .-
-If their use cannot be justified, they should be replaced by the standard -type -
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'C; Audits and Inspections -
3The. routine. inspections by the health physics -technician,:several- quarteEly and' special inspections by the Manager, Nuclear Safety, and annual audits made by:
the CRSC were reviewed., The health' physics: technician makes.the routine nuclear ' "
safety inspections. If he finds an item which nuclear safety he. questions or-is _in _ violation of nuclear criticality safety limits (has a " potential nuclear . -
- safety deficiency"), he fills out a special form indicating his findings and.
reports _ it to Manager, Nuclear Safety, for action and to the Manager, Health Physics.
Besides the quarterly inspections required by the license, the Manager, Nudlear .
= Safety, makes special inspections.before a new process ~ step is to be implemented-an'd at other times he -deems necessary. All inspection reports byithe Manager,
- Nuclear Safety, are addressed to the Manager, Licensing, Safety, and Nuclear Compliance. A special inspection made on_ November 12, 1986, is an' example of
-the need for inspecting new operations after approval of the WA but prior to implementation. During this particular inspection, it was found the boats
. planned for a low temperature furnace operation were'aifferent from those of'a" ;
high-temperature furnace operation (contrary to that assumed in the nuclear criticality safety analysis). -A.new analysis was.made modelled with the actual-furnace designs planned and the WA corrected reflecting the safe operating.
limits for the' furnaces. The need for also checking the work-station'
-limit signs prior to start of a new operation was discussed above.
The' audits performed by the CRSC during the past couple years were reviewed.
The audit report is addressed.to the responsible vice president with-copies "
generally sent to the cognizant area manager, Managers of Licensing, Nuclear-Safety, Health Physics, and members of the CRSC. The audits appear to be- -
thorough, documented, and recommended action'taken, and follow-ups made of-corrective action taken. It is recommended that greater uniformity be made in the distribution of the CRSC audit reports. Copies of only some of them appear to'be sent to all members of the CRSC, and the Manager, Nuclear Safety, is on distribution for only some of them. The same is true for the Manager, ~
Health Physics.
~D.- Documentation of Nuclear Safety Analyses
_ LDuring an earlier site visit (June 14-18,1982), the dumping of waste solution ~
- in the dump sink in the SV-A area was reviewed. Since the operation involved the intentional transfer of U-235 solution from a safe geometry vessel tof a
. non-safe geometry vessel, I requested a review of the nuclear safety analysis-l' (evaluation) for the operation. Since a documented evaluation could not be found, I recommended that-an analysis be made, documented, and retained for at i
!. least 6 months after the WA has been terminated. GA management concurred in the desirability of having such an analysis available and made a commitment to.
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, perform the analysis, document it, and retain it. The need for such 3
documentation for each operation is apparent and good practice. The renewal-
- application was modified to incorporate the requirements. The renewed license -
= requires " Work Authorization Records, including supporting criticality .
! . analysis shall be retained for at least 6 months following the expiration date-k.-
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W. T. Crow 7 of the authorization." During this assessment, a request was made to see the supporting criticality analysis. Since the 1982 site visit, an index of all analyses has been prepared. The requested analysis could not be located from the index based on the titles assigned. The present Manager, Nuclear Safety, contacted one of the former Managers, Nuclear Safety, .who had knowledge of the operation. The latter suggested the heading in the index under which the analysis could be found. Malakhof is searching for the analysis and plans to have it available for review by the Region V Fuel Facility Inspector during the next site visit.
During the assessment visit, it became apparent the present Manager, Nuclear Safety, makes a distinction between a nuclear safety analysis and an evaluation.
To him, the former means a computation analysis; the latter, an " analysis" based on a previously approved one, the use of approved nuclear criticality safety tables, or validated documentation. It is only the former that he documents.
Good practices include the documentation of all nuclear criticality safety analyses used, which includes evaluations made. ANSI /ANS-8.19-1984,
" Administrative Practices for Nuclear Criticality Safety" (Section 8.2) states that "the nuclear criticality safety evaluation shall determine and explicitly identify the controlled parameters and their associated limits upon which nuclear criticality safety depends," and (Section 8.3) "the nuclear criticality safety evaluation shall be documented with sufficient detail, clarity, and lack of ambiguity to allow independent judgement of results."
Based on the above, the following recommendation: are made:
- 1. Communications between successive Managers of Nuclear Safety should be improved so that there is a better understanding of license requirements-
, and good nuclear criticality safety practices.
j 2. GA should again review approved operation: and be sure there is adequate documentation to support work station limits and controls.
- 3. Review the nuclear criticality safety standards, guides, and developments to ensure that good practices are followed and the latest criticality data utilized in establishing safe operating limits.
- 4. The Manager, Nuclear Safety, should have all applicable standards and guides readily available.
E. Accident Scenarios A range of postulated accidents has been analyzed by the licensee. The one with the greatest potential offsite radiological consequences but lowest proba-bility is a criticality in the Fuel Fabrication Facility (SV-A Building) at the process liquid dumping site (see discussion above). The liquid dump room is designed to mitigate the results of a criticality accident and has 24-inch thick concrete walls and a 16-inch thick concrete ceiling so that the expected dose to anyone at the site boundary for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> following the accident would be 0.27 rem whole body, 0.26 rem thyroid, 0.015 rem bone, and 0.023 rem lung dose.
These doses are well within the Protective Action Guide (PAG) and EPA limits (1 rem whole body, 5 rem thyroid, and 3 rem other critical organs).
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'W. T. Crow 8 F. GA Planned Response to Criticality Accidents The Chairman of the CRSC reports to the Emergency Coordinator at the scene of
.the emergency. Other members of_ the CRSC report to the Emergency Support Center, pending instructions from the Chairman. The Manager, Nuclear Safety, also reports to the Emergency Coordinator at the scene of the emergency, if the emergency involves special nuclear material.. These personnel, together with those from -
Health Physics who also report to the Emergency Coordinator, provide the neces-sary experience to cope with a criticality accident and restore the facility to normal operations. The assessment team member responsible for evaluating the GA Radiological Contingency Plan will discuss its adequacy in greater detail.
VI. Recommendations
- 1. Personnel involved in nuclear criticality safety analyses maintain knowledge of up-to-date nuclear criticality safety criteria and become more familiar with license conditions.
2.- . Work station limit signs that do not adequately reflect nuclear criticality safety limits should be replaced.
- 3. All non-safe geometry containers not required for a work station operation should be removed from the station.
- 4. The Manager, Nuclear Safety, or his designee should review all work station signs before they are authorized at a station. All current signs should be reviewed to confirm they reflect station limits.
- 5. The type of work station sign at all stations should be standardized.
- 6. These should be greater' uniformity in the distribution of the annual CRSC audit reports.
- 7. Communications between successive Managers of Nuclear Safety should be improved so that there is a better understanding of license requirements and good nuclear criticality safety practices.
- 8. GA should again review approved operations and be sure there is adequate L docurrentation to support work station limits and controls.
, 9. GA should review the nuclear criticality safety standards, guides, and developments to ensure that good practices are followed and the latest criticality data utilized in establishing safe operating limits.
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- 10. The Manager, Nuclear Safety, sh0uld' have all' applicable standards land ;
guides'readily available.'
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Original Signed Bys Norman Ketzlach .-
Uranium Process Licensing Section
- Uranium Fuel Licensing Branch .
Division of Fuel Cycle'and - ,.
Material ~ Safety, NMSS -
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