ML20044C422
| ML20044C422 | |
| Person / Time | |
|---|---|
| Site: | 07000734 |
| Issue date: | 02/04/1993 |
| From: | Hooker C, Reese J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML20044C414 | List: |
| References | |
| 70-0734-93-01, 70-734-93-1, NUDOCS 9303230015 | |
| Download: ML20044C422 (17) | |
Text
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U.15. NUCLEAR REGULATORY COMMISSION REGION Y Report No. 70-734/93-01 Docket No.70-734 License No. SNM-696 Licensee: General Atomics P. O. Box 85608 San Diego, California 92186-9784 Facility Name: Torrey Pines Mesa and Sorrento Valley Sites Inspection at: San Diego, California Inspection Conducted:
January 11-15 and in-office 21-22, 1993
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WWP.3 W
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Inspector:
C.
.} Hooker,Fue F cilities Inspector Date Signed 2+f43 Approved by:
J00twL JA ~ --
e#H.Reese, Chief' Dat'e Signed Facilities Radiological Protection Branch Summary:
Areas Inspected: This was a routine unannounced inspection of criticality safety, management and organization, and followup on violations from a The inspection activities in these areas primarily previous inspection.
focused on the circumstances surrounding a licensee reportable event that.
occurred on January 11, 1993, involving the loss of a criticality control parameter. Inspection' procedures 30703, 93702, 88015, 83822, 88005, and 92702 were addressed.
In the areas inspected, four apparent violations were identified Results:
(1) failure to maintain records of CSAs (Section 2.1.3), (2) that involved:
failure to evaluate all credible accidents and show the K,,, value in CSAs (Section 2.1.3), (3) failure to follow procedures relative to reporting requirements (Section 2.1.4), and (4) one non-cited violation (NCV) for failure to provide test questions for annual nuclear safety retraining (Section 2.1.5).
D 9303230015 930218 PDR ADOCK 07000734 PDR C
DETAILS 1.
Persons Contacted General Atomics (GA)
- R. N. Rademacher, Vice President, Human Resources
- K. E. Asmussen, Director, Licensing, Safety and Nuclear Compliance
- W. L. Whittemore, Chairman, Criticality and Radiation Safety Committee
- R. C.'Noren, Director, Nuclear Fuel Fabrication
- B. E. Thurgood, Director TRIGA Group
- V. Malakhof, Manager, Nuclear Safety
- R. W. Schlicht, Manager, TRIGA Programs
- R. K. Kruger, Manager, TRIGA Fuel Fabrication
- B. J. Laney, Manager Statistical Measurements Control
- C. L. Wisham, Manager, Nuclear Materials Accountability
- S. P. Massey, Senior Engineer, Quality Assurance
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- Denotes those attending the exit interview on January 15, 1993, e
In addition to the individuals noted above, the inspector met and held discussions with other members of the licensee's staff.
2.0 Criticality Safety and Manacement Controls (88005 and 88015)
The licensee's program was reviewed for compliance with the requirements of 10 CFR Part 70, License Conditions, licensee procedures, and recommendations outlined in various industry standards and to verify that operations were being conducted to ensure the safety. of the general l
l public and facility workers.
There have been no new processing operations involving special nuclear material (SNM) requiring a nuclear criticality safety evaluation during the past year. Routine inspections of SNM areas by the Manager, Nuclear Safety since the last inspection of this area were reviewed. No significant nuclear safety' findings were identified from the licensee's inspections. The primary areas where SNM is used and/or maintained is the TRIGA Fuel Fabrication Facility (TFFF) and the' storage vault in Building 41.
With the exception of some limited activities that may be conducted within the next three months, the Sorrento Valley Building 39 (SVB) New Production Reactor test fuel project has-apparently been cancelled.
There were no activities involving the use of SNM-for this project during this inspection. The entire operation was currently limited to a safe batch' limit of 350 grams (93 % U-235).
j SNM activities in the Thermionics Fuel Fabrication Lab ~ also continue to be limited. There had been no SNM processing activities in this laboratory scale operation since the last inspection of this area.
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2 2.1 Licensee NRC Bulletin 91-01 Reoortable Event - TFFF At about 1:45 pm on January 12, 1992, during the onsite inspection, the licensee informed the inspector of an event involving the loss of a criticality control parameter at the TFFF that occurred while casting uranium-zirconium (UZr) alloy fuel rods at about 1:10 pm on January 11, 1993.
2.1.1 fackcround The manufacture of TRIGA fuel involves the mixing of uranium metal with zirconium metal using melting, casting, and standard machining processes. The alloyed uranium zirconium is machined, the machined meats are processed through a hydride furnace, ground to size and welded into metal tubes which serve as cladding for the fuel. The normal enrichment of TRIGA fuel is slightly less than 20% U-235.
The operating staff at this facility consists of a facility manager, two fabrication technicians, and a quality control (QC) inspector. The QC inspector verifies certain process steps and inspects product quality.
All of these individuals have several years experience at this facility.
Two types of fuel rods are manufactured by the casting process. One is the standard (about 1.45-inch diameter) TRIGA fuel rod and a smaller fuel rod of about 0.5 inches in diameter. The melting and casting of UZr alloy and other additives is accomplished using a stainless steel vacuum induction furnace defined as Station 6.
The inside diameter (ID) of the furnace is about 30 inches and the open mold well at the bottom of the furnace has an ID of about 15 inches and a depth of about 12-inches. The exterior surfaces of the furnace and copper induction heating coils for the crucible and mold are water cooled. The water cooling system is a limited volume, closed recirculating system.
Pre-measured UZr alloy is melted in a nominal 5-inch ID and 9-inch long (volume of about 3.2 liters) graphite crucible in the upper portion of the furnace and poured into a heated graphite mold. Holds for UZr ingots or fuel rods of various sizes may be used, but are restricted to a 5-inch ID and a 17-inch length.
The crucible's induction heating assembly consists of fiberglass cloth insulated coiled copper tubing with two attached support shafts. One shaft (conductive / cooling) is made to supply the electrical current and cooling water to the copper coils. This shaft extends to the outside of the furnace through an "0" ring sealed penetration when the. induction heating assembly is placed in the furnace. The preparation of the crucible / heating coil assembly consists of (1) an added epoxied fiberglass. cloth liner next to the coils, (2) centering the crucible on a bed of lampblack, (3) tamping lampblack between the crucible and fiberglass liner to a level below the top of the coils, and (4) adding a-graphite cement and heat transfer cement' cap between the wall of the crucible and coil assembly. After the cement cap has been dried, several small holes are drilled through the cap into the lampblack to
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allow for heat expansion and centering the crucible in 'the induction.
coil assembly. When placed in the furnace, the shaft with the conductive / cooling extension is positioned through a. furnace penetration and the other shaft is placed on a hub.
1 The mold is also heated by an induction copper coil assembly, but it is not prepared with the fiberglass and lampblack like the crucible..After-the mold has been placed in its heating assembly, splash guards are-placed around t'ip mold to protect the copper coils from any spillage of molten alloy. A funnel is also installed, at the top of the mold to channel the molten alloy.
Depending on the quantity of material to be melted and-the mix of the UZr alloy needed, some material is added to the crucible before the melt and the remainder through'a feed hopper at the top of the furnace during the melt. Prior to the melting operation, the furnace door and feed hopper lid are sealed. The water supply and electrical! systems are connected to the crucible / coil and mold assemblies. A pouring handle is attached to the crucible's conductive / cooling shaft on the outside of the furnace..
f After the water connections have been inspected for leaks, a vacuum is drawn on the furnace. The melting and molding operation is performed under a vacuum in an inerted atmosphere. The furnace is equipped with.
'several glass viewing / inspection ports to allow the technicians and QC inspector to visually inspect various steps of the operation. Once the proper vacuum has been achieved, the power generator is turned on to begin the melting process. The power supply system is equipped with a protective ground default interrupter (GFI) device which.is designed to shut-off the power to the furnace in the event of a water leak in the induction coils. Once the proper melt and desired temperature have been achieved, one of the operators manually isolates the GFI so that the power cannot be turned off to ensure the alloy maintains the desired temperature while it is being poured into the heated mold. The pour takes approximately three second.. The mold is allowed to cool for
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about 45 minutes before the furnace door is opened..The mold is subsequently taken to another work station where it is ultimately destroyed to remove the casted alloy.
2.1.2 Time Seouence of Events January 11. 1993 The licensee was in the process of making a second casting of TRIGA fuel with an alloy weight of about 9900 grams containing about 30% U with an enrichment of about 19.77% U-235 (587 grams U-235).
At about 1:10 pm, the technician reached to turn off the GFI on the 3
power supply to the furnace prior to making a pour of melted UZr ' alloy to the mold. As he reached for the GFI, he and the assisting QC inspector heard a loud noise from within the furnace. Both individuals inspected the inside of the furnace through the viewing ports.
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4 According to these individuals, the inside surface of the inspection ports were coated with a dark substance (subsequently determined to be lampblack), and the only thing visible was the glowing hot' crucible lying on the bottom of the furnace. These individuals conjectured that the crucible had been ejected from the induction coil assembly due to some type of equipment malfunction. The power to the furnace was secured at this time. The QC inspector and technician did not observe any physical damage on the outside of the furnace.
At about 1:30 pm the technician and QC inspector informed the TFFF's manager upon his returning from another site location. According to the TFFF's manager, based on his review and inspection of the matter, he assumed that the crucible had been ejected from its induction coil assembly due to a rapid buildup of steam from a water leak coming into contact with the crucible which was at about 1900 degrees Celsius (3452 degrees Fahrenheit). The TFFF'.s manager was aware that the quantity of F
U-235 in use at the time of the event was much less than the station limit of 1600 grams U-235. The manager made a decision to leave the cooling water to the furnace running overnight to cool-down the furnace before performing an internal inspection of the unit the next day.
January 12. 1993 At about 7:15 am the involved technician and QC inspector observed water leaking from the furnace door as it was being opened. About 15 gallons of water was subsequently removed from the furnace by draining and pumping to an unfavorable geometry 3-gallon bucket, and transferring it to an unfavorable geometry 55-gallon waste collection drum within the facility. According to the operators, the removed water was-very clear and did not appear to contain any uranium metal. No sample analysis was performed of the water.
The technician and QC inspector checked that the exhaust ventilation system was lined up to the furnace, and papered the floor area in front of the furnace in preparation for ultimate cleaning out of the furnace.
When the furnace was fully opened, UZr metal was observed to have splashed around the top and bottom of the mold well, some on the crucible, and some inside of the crucible's induction coil heating assembly.
At about 9:00 am the Health Physics Technician (HPT) for this facility was summoned by telephone to survey the furnace and the surrounding At about 9:10 am the HPT arrived at the TFFF and was informed of area.
the event by the QC inspector. Although a considerable amount of lampblack dust was present, wipe tests from inside and outside of the furnace, and the surrounding floor area indicated no measurable activity using portable instrumentation. Wipe tests from the furnace and floor, and nasal smears from the operator doing the cleaning were obtained for <
laboratory counting.
At about 9:30 am the TFFF's manager arrived onsite and was informed of the as found conditions and current status of the furnace.
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At about~ 9:40 am the TFFF's HPT informed the Health Physics manager l
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l (HPM)-oftheevent. The.HPM. questioned-the HPT-as to the quantity of.
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SNM that had been involved in the incident.. The HPT;was not cognizant of the SNM involved since he had devoted his efforts only to the radiological safety aspects of the event. - The HPT subsequently returned.
to the TFFF to change air samples in the facility. Based on the.
conversation with the HPf, the HPM believed that the event had occurred that morning (January 12th).
i At about 9:40 the HPM relayed details of the event to the Director, i
Licensing, Safety, and Nuclear Compliance (DLS&NC) who reported.the j
matter to the Vice President, Human Resources (VPHR). Between 10:00.and.
l 10:15, the DLS&NC and VPHR visited the TFFF to inspect the scene and discuss details of the event with the TFFF's manager.
j At about 11:30 am the DLS&NC discussed the matter with the Manager,.
Nuclear Safety (MNS) who also visited the TFFF.to review the matter.-
i Based on the MNS's review of the safety analysis file for Station 6 and Part I, Demonstration,' Volume 1 of the -license application, he concluded -
l that the upset condition involving the crucible (simultaneous spillage -
of SNM and flooding at the same time)_ appeared to be a' condition that -.
had not been analyzed. The available data (CSA of the mold and. general-description of Station 6 operation in the licensee's Demonstration Volume 1) indicated that the Station 6 limits (1600 grams U-235) were solely based on geometry and mass. The' event with the crucible represented a loss of' geometry. The MNS reduced and posted the Station 6 SNM limit to 457 grams of U-235, which is_ considered by the licensee to be an ever-safe limit for 205; enriched uranium.
At 1:04 pm the licensee reported the event to the NRC Operations l
j Officer. At about 1:40. pm the licensee informed the NRC inspector, who was touring the licensee's Sorrento Valley facilities.
2.1.3 Criticality Safety Analyses /CSAs)
The inspector examined the induction furnace and associated equipment,
j and reviewed the CSA file for Station 6 to di termine whether adequate.
1 limits, barriers, and controls had been estab'ished and were being l
implemented. The only information in the CSA file was three evaluations:
with attached drawings of molds:
j (1)
" Casting Molds," dated September 8,1976.
f (2)
" Casting Mold," dated October -16,1975.
(3)
" Buckling Calculation for TRIGA Mold," dated October 15, 1975 These evaluations only provided hand-performed buckling calculations -
1 which defended the mass limit and geometry of the mold. There was'no evaluation relative to the crucible, discussion of accident scenarios',-
or details of the induction furnace. However, Section 3.7.5.3, " Melting j
and Casting, _ Station 6 (Mass Limit 1.6 Kg U-235)," of Part 1, i
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6 Demonstration Volume I, of the licensee's license application provided a general summary of the operations, criticality controls, and accident scenarios associated with this station, This section under " Nuclear Safety Basis" also included a statement that the crucible had been analyzed according to a similar argument (Figure 1.A.3 of " Handbook of Criticality," Volume 1, UKAEA, AHSB(S),1965) as the' mold, and that the furnace had been found to be safe under two accident conditions: (1) a water leak, and (2) a spill of metal out of the mold. The licensee could not locate the referenced CSA for the crucible.
The licensee believed that the apparent missing CSA for _the crucible and/or any other documentation to support the mass limit for Station 6, had been mis-filed or inadvertently archived. The feasibility of retrieving such documentation was discussed during the inspection and at the exit interview on January 15, 1993. On January 21, 1993, by facsimile, the licensee provided the inspector with a CSA, " Melting Furnace," dated December 27, 1972, that they located in their license correspondence files. The first paragraph of this CSA stated:
"The following is a nuclear safety analysis of a furnace to melt U-Zr alloy fuel where you intend to limit yourself to a 1.8 kg of U or 1.68 kg of U-235 and to answer your question on possible design limitations...."
The second paragraph stated, in part, that the results presented in this CSA were based on data taken from TID-7028 (" Critical Dimensions of Systems Containing U-235, Pu-239, and U-233," June 1964) and doing buckling conversion calculations, and these results are given in an internal memo by... (DCP:72:25)." The CSA did not contain a description as to the type or operation of the furnace.
The December 27, 1972, CSA for a furnace did provide data to support a safe diameter (six inches) of any length of a furnace for two postulated accident conditions (water flooding and double batching), but did not evaluate simultaneous flooding and spilling of UZr melted metal or provide the technical basis as to why such an accident was not credible.
Also, the CSA did not describe the use of a mold.
i As discussed above, Demonstration Volume I, Section 3.7.5.3, stated that the crucible had been analyzed according to a similar argument (Figure 1.A.3 of " Handbook of Criticality," Volume 1, UKAEA, AHSB(S),1965) as the mold. This information differed from the December 27, 1972, CSA that referenced technical document TID-7028, " Critical Dimensions of Systems Containing U-235 Pu-239, and U-233," June 1954.
Based on the review of the above referenced CSAs, the inspector determined that the CSAs conservatively used uranium enriched with 93S; U-235, and the most credible geometry (a sphere) was assumed in the evaluation. However, the CSAs failed to adequately evaluate or discuss the melt / mold operation as one process. Also, the inspector determined that the CSAs lacked the basis used to exclude contingencies (simultaneous flooding and spilling of melted UZr alloy at the same
7 time)' not considered-credible. - The CSAs did not 'contain sufficient detail of the induction furnace and associated equipment to_ allow a
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technical review so that the operatior, of the _ furnace and related nuclear. safety parameters were comprehensible. The December 27, 1972,
-CSA appeared to be 'a preliminary analysis for _some type of furnace that-could melt UZr alloy. - The CSA did not discuss the use of a crucible or :
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mold, or how the UZr was to be melted.
The inspector noted that a licensee letter (AEC material ' License 696; Docket 70-734 Application for Amendment; UZrH Fuel Fabrication) dated-November 23, 1973, requested an amendment to modify the license to permit the process development of the TRIGA fuels by a casting method.
Appendix Page B5-2A, items IIIA and IVA, and the Nuclear Safety Considerations of this appendix, provided the same accident scenarios as; delineated in the licensee's current Demonstration Volume 1 for Station -
6.
There were no accident scenarios described for the crucible.
The Inspector reviewed the license and. commitments contained'in the-license specifications and made the following observations:
1)
Condition No.11 of License No. SNM-696 states, in part:
" Records'of all safety-related reports and analyses shall be retained as' follows:
a.
Copies of criticality and radiation safety analyses shall. be retained for at least 2 ' years or 6' months after a project is terminated, whichever is longer."
Section 3.7.5.3, Demonstration Volume I, stated that the crucible -
had been analyzed according to a similar argument (Figure 1.A.3 of
" Handbook of Criticality," Volume 1, UKAEA.- AHSB(S),1965) as the mold. No such analysis for, the crucible could be found~by the licensee. Failure to retain the CSA for the crucible was identified as an apparent violation ~of License Condition No.11; (70-734/93-01-01).
2)
Condition No. 9 of License No. SNM-696 authorizes the use of.
licensed materials in accordance with the statements, representations, and conditions contained in Part II, " License Specifications," dated July 24, 1981, and supplements dated March:
16, 1982, through June 19, 1992.
Section 5.0, " Nuclear Safety - Technical Requirements," Part II of the license specifications states:
"The continued nuclear safety of the licensee's opera'tions shall be assured.by limits and procedures documented-in accordance with the specifications contained in the-following sections:"
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i Section 5.2, " Basic Assumptions," of the specifications states, in part:
"The basic assumptions that shall be utilized in arriving at particular criticality limits are as follows:"
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" Accident - At least two unlikely, accidental, concurrent, and independent events must occur before a criticality hazard could exist."
i Section 5.4.3, Item d states, in part:
r "The geometrical safe limits shall apply to units meeting the following criteria:
"i.... ii. The unit shall be evaluated for all possible credible accident conditions to assure that the t
safety margin (difference between critical and safelimits)isnotexceeded."
Section 5.4.4, Item b. of the specifications states:
"All geometries analyzed by hand calculations shall have_ a K,,, of 0.90 or less.
Based on the above observations, the inspector concluded that' the CSAs for Station 6 did not consider all credible accident conditions.
Specifically, spilling of molten enriched UZr alloy and flooding of the of the induction furnace from its single jacketed water cooling system t
at the same time was a credible accident condition that had not been analyzed.
In addition, the K ' value was not in these analyses that were performed by hand calculations.
Failure of the CSAs to adequately address credible accident scenarios that could impact criticality safety was identified as an apparent violation of License Condition No. 9 (70-734/93-01-02).
I Failure of the CSAs to provide the calculated K,ff for geometries of the cylinders analyzed by hand calculations to demonstrate that they were critically safe vessels was also identified as an apparent violation of License Condition No. 9(70-734/93-01-03).
Two apparent violation were identified in this area.
2.1.4 Licensee Performance as Related to the January 11. 1993 Event The inspector reviewed the sequence of events related to the event as
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summarized in Section 2.2 above. The inspector evaluated operator and management performance of activities during and after the event to determine. if their performance was consistent with that required by applicable operating procedures and regulatory requirements.
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g As stated in Section 2.2 above, the initial event was observed by an operator and QC inspector at about 1:10 pm on January 11, 1993. These individuals assumed that the crucible had been ejected from the induction coil assembly from a water leak in the system. The operator immediately turned off the power to the induction furnace. There was no apparent specific physical damage to the furnace and no apparent loss of its integrity. The cooling water was left on to cool down the furnace.
These individuals informed the inspector that they believed that the molten UZr metal had spilled out of the crucible, since it was observed to be lying on the bottom portion of the furnace. The manager of the TFFF became cognizant of the event at about 1:30 pm on January 11, 1993.
Condition No. 9 of License No. SNM-696 authorizes the use of licensed materials in accordance with the statements, representations, and conditions contained in Part II, " License Specifications," dated July 24, 1981, and supplements dated March 16, 1982, through June 19, 1992.
Section 3.2.2, " Compliance Functions," Part II of the license specifications, states:
"All functions responsible for assuring compliance with applicable license requirements and controlling the radiological and nuclear safety and safeguards of licensed material are part of the Human Resources Organization of General Atomics. Namely, these functions are: Nuclear Safety, Licensing, Safety and Nuclear Compliance, Nuclear Material Accountability, Statistics &
Measurement Control, Security, and Health Physics."
"The Director of Human Resources, or his designee, will establish the necessary policies of operation, cause'them to be published in company-wide guides and manuals, and coordinate related activities with operating groups to assure compliance with related policies, procedures, regulations and license conditions...."
The review of the operating procedure, " Induction Melting and Pouring of i
U/Zr and U/Zr/Er Alloy," and interviews with the operator and QC inspector did not reveal any deviations from the operating procedure by these individuals. The inspector also noted that the procedure did not provide any actions for process upsets. The Nuclear _ Safety instructions in the procedure only delineated the operators' responsibility for not exceeding the station limit of 1600 grams of U-235, and the limits as-provided for the other stations associated with this operation.
Regarding reporting of the incident, in response to NRC Bulletin 91-01,
- Reporting Loss of Criticality Controls," dated October 18, 1991, the licensee established its NRC reporting requirements and implementing procedures in their Nuclear Safety Guide.
Section 3.4.1, " Definitions," of the licensee's Nuclear Safety Guide, defines " Parameter" as-
"A characteristic of a system which must be controlled for the
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10 purposes of criticality safety, e.g., mass, geometry, concentration and moderation."
Section 3.4.2, " Criteria for Reporting," of the licensee's Nuclear Safety Guide, states in part-
" Situations,- events or items such' as the following are to be reported to management and Nuclear Safety for evaluation when
[the] quantity of SNM is greater than an ever-safe mass:"
Item 1 of Il situations states:
"Procedura cannot be followed as written" Section 3.4.3, " Criteria for Action," of the licensee's Nuclear Safety Guide, states in part:
"Upon becoming aware of a situation, event, or item of such as above, Nuclear Safety, with the assistance from additional specialists as necessary, will evaluate and assess the safety significance of the occurrence. Nuclear Safety will also assess for adequacy, the corrective actions taken, and/or planned."
"Further, Nuclear Safety will consult witn and advise, Licensing, Safety and Nuclear Compliance with regard to proper. reporting, both internally and externally. Reports to the NRC~ Operations Center will also be reported to the appropriate Regional Administrator."
"The criticality safety event reporting criteria and appropriate response actions are given in Table 3.4-1."
Condition 2 of Table 3.4-1 states:
" Substantial degradation or_ loss of a control parameter resulting in only one parameter remaining under control" Action "a" for Condition 2 states:
" Notification of Nuclear Safety and or Licensing, Safety and f
Nuclear Compliance" Action "b" for Condition 2 states:
" Notify NRC as soon as possible within four (4) hours of discovery" The inspector made the following observations related to this event:
1)
At about 1:10 pm on January 11, 1993, while melting low enriched uranium-zirconium alloy of more than an ever-safe mass of U-235, an operator and the TFFF's QC inspector were aware of a significant process upset that resulted in shutting down the i
p 11 operation in progress. Shortly after the event, at about 1:30 pm, the TFFF's manager became cognizant of the event. The event resulted in an unusual condition for which the process could not be completed in accordance with the established operating-procedure, " Induction Melting and Pouring of U/Zr_ and U/ZR/Er-Alloy." Spetifically, (1) 'this event (simultaneous flooding and spilling melted uranium-zirconium alloy to an. unfavorable geometry condition) appeared not to have previously occurred during plant operations, and (2) this. condition was not covered in the operating procedure.
According to the TFFF's manager, he knew that the quantity of U-235(587 grams) being processed was less than the critical mass of U-235 (about 805 grams with optimum moderation and reflection).
He also assumed that the Station 6 (induction furnace)-limit of 1600 grams of U-235 was based on accidents that included spilling of the UZr alloy and flooding at the same time since the furnace Therefore, in his judgement, he' did not have was cooled by water.
an unsafe condition and did not believe that the matter required reporting to Nuclear Safety.
Fortunately, at about 10:00 am on January 12, 1993, through the 2)
HPM, the DLS&NC became aware of the event and subsequently-Based on the MNS's review of the available CSA informed the MNS.
and Demonstration Volume, he concluded that the event resulted in the loss of one of the two control parameters (geometry) that assured criticality safety for Station 6.
This event occurred at about 1:10 pm on January 11, 1993, but was not reported to the NRC until 1:05 pm on January 12,1993. Failure of the TFFF's staff to report an event involving more than an ever-safe mass of U-235 to Nuclear Safety and report the event to the NRC within four hours from the time the event was known was identified as an apparent violation of License Condition No. 9 (70-734/93-01-04).
Regarding the removal of the water from the furnace using the-unfavorable geometry bucket, the MNS stated that due to the nature of material involved (UZr alloy), he would have only recommended the use of a fine screen and would not have required an analysis of the water before its removal from the furnace. He also stated that such events warrant Nuclear Safety's involvement to assure they are properly The inspector noted that the weighing of the UZr alloy as it mitigated.
was removed from the furnace, indicated that the initial 587 grams of U-235 was accounted for.
One apparent violation was identified in this area.
2.1.5 Trainina The inspector made the following ooservations related to general nuclear safety training and training related to NRC Bulletin 91-01:
1)
In review of the licensee's training relative to NRC Bulletin 91-
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01 reporting requirements, the inspector noted that on May 15, 1992, the Licensing, Safety, and Nuclear Compliance group had -
distributed their new reporting requirements and implementing procedures to selected upper site management. During interviews with the HPM who reports to the DLS&NS and the TFFF's manager, these individuals stated that they had not received a copy of (and
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were not cognizant of) the nu NRC reporting requirements.
In addition to distributing these new procedures to upper management, they were also preserved in the Licensing, Safety and Nuclear Compliance group's Nuclear Safety Guide, which has limited distribution.
2)
There was no initial site wide specialized training of site personnel relative to NRC Bulletin 91-01 requirements. The training of site personnel was being performed through' the licensee's annual refresher training for radiological safety and nuclear safety controls.
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The annual refresher training consists of two to four hours' of classroom instruction which includes a nominal 30-40 minute H
session on nuclear safety controls. Following retraining, employees are given a test to evaluate their knowledge of the training provided. As determined from an interview with the MNS who normally provides the nuclear safety training, only about 10 minutes of this training was devoted to the licensee's internal and NRC reporting requirements related to NRC Bulletin 91-01 reporting requirements. This portion of training consisted of a view graph of the licensee's internal reporting requirements:
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" Situations, events or items such as the' following are to be reported to management and Nuclear Safety for evaluation when [the] quantity of SNM is greater than an ever-safe mass:"
The view graph also included the same eleven internal reporting situations delineated in the licensee's Nuclear Safety Guide. No handouts of the view graphs were provided to the individuals attending the training.
Tne inspector presented a copy of the view graph to the TFFF's staff who had received their refresher training in September 1992. None of these i
individuals, including the TFFF's manager, could recall the view graph.
Based on the other subject matter covered (safe mass of U-235, favorable geometries, storage of SNM on racks, process. station posting, process station operation, liquid waste disposal, lab versus full scale operation, and function of Nuclear Safety), the inspector determined that the time allowed for all of the subject matter appeared less than adequate to assure proper retention of the subject matter covered, especially the time devoted to reporting requirements.
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13 During the review of training records, the inspector noted that the retraining tests given to the TFFF's staff in September 1992, did not include questions relative to nuclear safety, which differed from tests Normally one test is given that includes given in previous years.
Based on questions related to radiological and nuclear safety.
discussions with the HPM and MNS, it appeared that a new test for the radiological safety portion of the training had been developed, and due to a lack of comunication between these two components, the nuclear safety questions had been inadvertently omitted.
Section 3.3.5, " Retraining," of the license specifications states:
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" Retraining on selected subjects will be done annually. Subjects The will include both radiological and nuclear safety controls.
understanding of the employees and the program's effectiveness will be evaluated with an examination which must be satisfactorily Records of the training completed with a score of 70 percent.
will be kept which will include the date held, subject matter covered, attendees, instruction and test results. The Manager, Health Physics or other qualified designee is responsible for this training."
When this matter was discussed with the Licensing, Safety, and Nuclear Compliance staff, retraining tests specific to nuclear safety controls 14, 1993. The were immediately given to the Tfff's staff on January inspector noted that all of the individuals passed their test with a The inspector also noted that the two true and score of 100 percent.
false and four multiple choice questions provided to these individuals, were the same questions that had been on the tests given the previous None of the questions involved internal reporting of events two years.
to Nuclear Safety.
The inspector discussed the above observations and the need for the licensee to enhance and evaluate the effectiveness of its nuclear safety retraining program during the inspection and at the exit interview on The inspector's concerns were acknowledged by the January 15, 1993.
The MNS stated that a more comprehensive licensee's management staff.
retraining class and new test questions would be provided to applicable employees in the near future.
Failure to include test questions to evaluate the employees understanding of the training provided for nuclear safety controls was identified as an apparent violation of License Condition No. 9.
However, the violation is not being cited because the criteria specified in Section VII.B(1) of the Enforcement Policy were satisfied (NCV 70-734/93-01-05).
In response to the inspector's observations regarding the limited distribution and training of employees relative to the reporting requirements delineated in NRC Bulletin 91-01, the DLS&NC issued a memorandum, dated January 15, 1993, to SNM users and all site management personnel involved in SNM activities that described the importance of
i 14 reporting any failure or degradation c/ a criticality control to the appropriate parties and management levels, and noted the importance of GA's response to the doing so in a timely and expeditious manner. bulletin which included the to the memo.
One apparent HCV was identified in this area.
2.1.6 Licensee's Investiaation The licensee initiated an investigation to determine the cause of the 1
event so that corrective actions could be taken to prevent recurrence.
The licensee initially pressurized the crucible's induction heating coil assembly with about 10 pounds per square inch gage (psig) of helium with At about 20 psig some leakage was noted on the no observed leakage.
inside lower half of the assembly. When the fiberglass coating was removed from the area of the leakage, three small blemishes were These blemishes had the observed on the surface of the copper coils.
appearance of that typically observed from an electrical arc (melted-The inspection of the intact carbon crucible revealed a copper).
discoloration that would be adjacent to the blemishes on the copper coils with the crucible sitting in the assembly. The blemishes on the copper coils and discoloration on the crucible were also observed by the inspector.
The licensee planned further inspection of the induction coil assembly, which included the removal of the damaged section to perform a more-technical examination in determining the cause.of the. event.
The inspector also examined the radiological surveys taken as a. result i
The results of data' from wipe tests, area air samples, of this event.
and stack effluent samples indicated'that there was no detectable Nasal smears and lung release of radioactive material from the furnace.
counts (the mobile lung counting contractor was onsite performing this routine service) indicated no intakes of radioactive material by the individuals present during the event, 3.0 Manacement and Oroanization Controls (880051 In addition to the management controls discussed in Section 2.0, above, the inspector noted that the management organizational structure and defined responsibilities were consistent with those described in Section The HP group had hired a 3.0, Part II, of the license specifications.
technical professional and several full time HPTs to augment its staff, i
The licensee's Criticality and Radiation Safety Committee's annual audits of the various safety related programs for 1992 were reviewed.
The audit findings appeared to be administrative in nature and did not i
represent any significant impact on safety. Selected procedures and The procedures and work-work authorizations were reviewed.
authorizations included the appropriate reviews and approvals consistent r
i
15 with the requirements specified in Section 3.0, Part II, of the license specifications.
No violations or deviations were identified.
4.0 Followup on Violations (92702)
Item 70-734/92-07-02 (Closed): This item involved the failure to review emergency procedures associated with Building 39 operations. The inspector verified the corrective actions taken to correct the violation and those taken to prevent recurrence as described in the licensee's timely response dated November 12,.1992. The inspector noted that on October 19, 1992, the licensee had completed its review and approval of -
new emergency procedures for Building 39. The inspector also noted that the check list of requirements contained in the licensee's procedure for work authorization approvals'had been revised to ensure that the emergency procedures for each authorization had been reviewed within the appropriate time period. The inspector had no further questions regarding this matter.
5.0 Exit Interview (30703)
The inspection scope and findings were summarized with the individuals denoted in Section I, on January 15, 1993. The following findings were discussed:
1)
The apparent deficiency of the CSA and supporting documentation to clearly define the accident conditions analyzed for TFFF's induction furnace.
2)
The apparent lack of records of the CSA performed for the use of the crucible in the TFFF's induction furnace.
3)
The limited distribution and limited training regarding NRC reporting requirements related to NRC Bulletin 91-01.
4)
The concern that the January 11, 1993, induction furnace event was not promptly reported to Nuclear Safety, and was not reported to the NRC within the specified requirement time.
5)
The apparent inadvertent deletion of annual retraining test questions related to nuclear safety controls, and the reuse of test questions on annual retraining exams.
The inspector also stated that the above observations appeared to represent a lack of management oversight and a lack of attention to detail related to its nuclear safety program. The inspector also expressed a need for the licensee to review the adequacy of all active CSAs relative to license requirements and current standards.
The Director, Licensing, Safety, and Nuclear Compliance committed to develop and implement a program to update all active CSAs as
16.
The Vice President, Human Resources also' acknowledged this appropriate.
commitment.
Also, the licensee informed the inspector that they would continue to search their files for the apparently mis-located CSA for the crucible, and would forward such information for the inspector's in-office review.
21, 1993, the licensee forwarded related records By facsimile on January found. The review of these records on January 21 and 22, 1993, is included in the body of this report (70-734/93-01). The licensee's action relative to a CSA update program will be reviewed in a future inspection. This matter is considered as an inspector followup item (70-734/93-01-06).
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