ML20217N728
| ML20217N728 | |
| Person / Time | |
|---|---|
| Issue date: | 09/01/1995 |
| From: | Dan Collins NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Mallett B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| Shared Package | |
| ML20217N558 | List: |
| References | |
| FOIA-97-246 NUDOCS 9708260306 | |
| Download: ML20217N728 (10) | |
Text
, Fromi Dw glas M. Collins O' Date: Friday,BSMI . -bc u(DMC) /MAde/h T2 To:
September 1, 1995- 7:18 am
Subject:
GE Meeting I talked to Bob Pierson about changing the date of the GE DCF meeting from 9/27-28, since this was during our Retreat and I now heard that the Ms. Ten Eyk will attend. We would like a 01 vision level person to accompany her. He said that she-is_ going to the meeting (if she goes, on her calendar but something may come up) more as an opportunity to get a good look at the facility in her new cipacity. The issues to be discussed are not of such substance that NHSS t1ought-they needed a Division Director to lead the meeting. He said thac at this point he really did not want to change the meeting, since he had a difficult-time setting a date and we needed to get licensing issues moving. At this point, we will be represented by McAlpine and Troup.
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-I MEETING AGENDA GE/NRC MEETING SEPTEMBER 27-28, 1995 '
'l. Site and Facility Orientation:
general tour of the facility (GE) discuss facility changes due to DCP (GE) 2, license Format & Content:
discuss general SRP draft NUREG-1520/ Licensee-format (GE/NRC) discuss specific SRP areas to be used in GE application (GE) discuss environmental infonnation needs (NRC) discuss possiMe 6E cow.ntments to complete certain parts of application after renewal (GE/NRC)
, - discuss contents of Process Os scription Document (NRC) 3,1.icense review schedule:
- discuss NRC's review schedule (NRC)
- pre-submittal efforts
- submittal and issuance dates
-- contingency plan for DCP approval
~
- AE vs-categorical exclusion
- technical review meetings
-. discuss Region II's construction inspection schedula (RG II)-
4 W
REVIEW TEAM 3
FOR
, GE WILMINGTON LICENSE APPLICATION
, SRP (NUREG 1520) SI'CTIONS:
1,0 General Information - Keith McDaniel
- 2.0 Management Organization - Keith McDaniel 3.0 Conduct of Operations - Keith McDaniel 4.0 Integrated Safety Analysis - Keith McDaniel 5.0 Radiation Safety - Elaine Keegan/Ed Flack 6,0 Nuclear Criticality Safety - Marc Klasky-7.0 Chemical Safety - Lidia Roche
- 8.0 Fire Safety - Amar Datta 4
9,0 Emergency Management - Ed Flack
). 10.0 Environmental Protection - Elaine Keegan 11.0 Decommissioning - Keith McDaniel/WM 1
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September 28, 1995 t General Electric Company
' ATTH: Mr. C. P. Kipp, General Manager GE Nuclear Energy Production P. O. Box 780 Wilmington, NC 28402
SUBJECT:
NRC INSPECTION REPORT NO. 70-1113/9508
- Gentlemen:
This refers to the inspection conducted by G. L. Troup of this office on September 11-15, 1995. The inspection included a review of activities authorized for your Wilmington facility. At the conclusion of the inspection,
- the findings were discussed with those members of your staff identified ,n the report..
Areas examined during:the inspection are identified in the report. Within these areas, the inspection consisted of selective examinations of procedures and representative records, interviews with personnel, and observation of activities in progress. The purpose of the inspection was to determine
' whether activities authorized by the license were conducted safely and in accordance with NRC requirements.
Within the scope of the inspection, violations or deviations were not identified.
In accordance with 10 CFR 2.790' of the NRC's " Rules of Practice," a copy of this letter. and its enclosure will be placed in the NRC Public Document Room.
Should you have any questions concerning this letter, please contact us.
Sincerely, (original signed by D. H. Collins)
Douglas H. Collins, Chief Nuclear Materials Safety and Safeguards Branch Division of Radiation Safety and Safeguards Docket No. 70-Ill3 License No. SNM-1097 l
Enclosure:
NRC Inspection Report cc w/ enc 1: (See Page 2) '
em .
4 GE 2 cc w/ enc 1:
J. F. Klapproth, Manager Fuels and Facility Licensing General Electric Company P. 0. Box 780. Mail Code J26 Wilmington, NC 28402 Dayne H. Brown, Director Division of Radiation Protection N. C. Department of Environment.
Health & Natural Resources P. O. Box 27687 Raleigh, NC - 27611-7687 -
Distribution w/ encl:
E. McAlpine, Ril _
G. Troup R. Bellamy, RI G. Shear, RIII C. Cain, RIV F. Wenslawski, RIV PUBLIC Distribution w/o enc 1:
License Fee Management Branch sENr' To puntic Boet9MEMf ROOM 7 fYES) NO OFFICE Ril:0RSS Rit:DRSS SIGNATURE NAME LTro EJMcAlpine DATE OSMM195 0977396 09 / /96 09 / / 95 09 / /95 09i /95 COPY 7 [ YES M O YES h YES' NO YES NO YES NO YES NO OFFICI W CORD COPY _ 00CUMEh! ttAME: G:\RPIS\R$PS\GE9508.GLI
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, Report No.: 70-1113/95-08.
Licensee: General Electric Company
- Wilmington, NC 28401 4
Docket No.: 70-1113 License No.: SNM-1097 l Facility Name: Nuclear. Energy Production Inspection Conducted: September 11-15, 1995
/ s Inspector: I n. -
G. L. Troup, Sr. fuel pacility-iDate2f!V Signed Project Inspector y I
Approved by: '
MM i ia E. J. McAlpine, Chief Date Signed i Radiation Safety Projects Section Nuclear Materials Safety and Safeguards Branch
- Division of Radiation Safety and Safeguards 1
SUMMARY
Scope:
1 This routine, unannounced inspection was conducted in the are&s of management and organization, nuclear criticality safety, facility operations, facility 3 changes,-training and fire protection. In addition, previous inspection findings were-reviewed.
- Results:
Within the scope of the inspection, no violations or deviations were identified. One previously identified violation was closed based on the licensee's corrective actions (paragraph 5.a).
Several changes.in the organization had been made. Many of-these changes do not require license changes. Two: changes awaiting internal action are the designation of Area Managers based on current assignments and the designation
.of the chairman of the WSRC.
Enclosure q /*/'
l REPORT DETAILS
- 1. Persons Contacted l
- 0. Barbour, Coordinator - Radiation Protection
- R. Foleck, Sr. Licensing Specialist
- B. Kaiser, Manager, Chemical Product Line
- J. Klapproth, Manager, Fuels & Facility Licensing
- R. McGowan, Team Leader, URLS/WT '
- S. Murray, Manager, Nuclear Safety
- L. Paulson, Program Manager, Criticality Safety
- S. Selby, Team Leader, U0, P:oduction
- W. Sependa, Manager, GE-NE EH&S G. Smith, Team Leader, FM0 Maintenance Support
- H Strickler, Manager, Environmental Protection & Industrial Safety
- J. Taylor, Principal Engineer, Criticality Safety C. Tarrer, Team Leader, Configuration Management C. Vaughan, Project Manager, New Facility Licensing / Safety F. Welfare, Program Manager, Criticality Safety, DCP The inspector also interviewed other licensee employees including area coordinators, engineers, operators and technicians.
- Attended exit interview on September 15, 1995.
Acronyms and initialisms used throughout this report are listed in Appendix A.
- 2. Management Controls and Organization (88005, 88015)
- a. On July 24, 1995, the Vice President, GE-NE announced an organization change in that the EH&S & NQA organization was divided into two organizations: Quality and EHS. The former manager of EH&S & NQA was appointed Manager, Quality and the former Manager, Chemical Product Line was appointed Manager, EH&S.
Neither of these positions are described in Part I of the license application and can be changed as desired by management,
- b. In August,1994 the Manager, Regulatory Compliance was assigned to the DCP. The Manager, EH&S & NQA issued a memorandum on August 8, 1994 stating that the responsibilities of the Manager, Regulatory Compliance would be carried out by his office. With the organizational change, this was no longer effective. On September 1,1995, the new Manager, EH&S issued a memorandum stating that the responsibilities of the Manager, Regulatory Compliance would be performed by his office. The inspector determined that the qualifications of the new Manager, EH&S met the qualifications specified in Part I, Chapter 2, Section 2.5.6 of the license application.
i 2
- c. With the changes in GE NE management and the DCP, several other management changes, including the assignment of " acting" managers, were made.
While these are not defti.ed in tha license application, the designation of Area Managers is defined in the license application. P/P 40-16, Nuclear Safety - Area Manager Responsibilities and Assignments, defines the assignment of Area Managers. Specific assignments are made by the Manager, GE-NEP.
The inspector reviewed the current memorandum making the assignment and noted that several designated Area Managers are no longer in positions associated with the assignment. The inspector discussed these aseignments with two individuals. Both stated that they were aware of the difference and were continuing to carry out the functions of the Area Manager as appropriate. The inspector also reviewed documents which have to be signed / approved by the Area Manager and verified that these individuals had
' properly signed the documents. Because the qualifications of an Area Manager specify a minimum experience at the facility will be made when the replacements achieve the necessary , changes qualifications. The inspector noted that the documentation included the initials of the prospective Area Manager, indicating "
that they were involved in current changes. A representative advised the inspector that changes had been identified and were awaiting action on the designating memorandum,
- d. Part I, Chapter 2, Section 2.3.1 of the license application defines the membership and responsibilities of the WSRC. WSRC activities are prescribed in P/P 40-1, Wilmington Safety Review Committee. The inspector reviewed the P/P and the minutes of the meetings June 9 and July 28 meetings and determined that the scope was in accordance with the WSRC charter. The membership of the WSRC was changed to include the Manager of the Fuel Examination Technology facility (which is under a State of North Carolina by-product material license rather than the NRC SNM license) because many of the responsibilities of the WSRC concern the FET operation as well. The inspector noted that the chairman of the WSRC was the Manager, EH&S & NQA. With the organization change (paragraph 2.a. above) the Manager, EH&S will assume this position. A change to make this assignment is being incorporated into P/P 40-1 as part of a current revision.
Within the scope of the inspection, violations or deviations were not identified.
- 3. Nuclear Criticality Safety (88010, 88015, 88025) a.
Part I, Chapter 4 Section 4.2.2.5 of the license application !
states that calculational methods used in nuclear criticality safety analyses shall include GEKEN0 and GEMER Monte Carlo codes, i and that newly developed codes may also be used when they have l been' validated and benchmarked. The GEKENO and GEMER codes had been converted for use on personal corputers. Review of the ,
t 3
validation 70 1113/94-09, of therespectively.
codes was documented in irs 70 1113/93 06 and The inspector discussed the status of the codes, such as any new codes or major modifications to existing codes, with licensee representatives. They stated that no new Gdet were being introduced nor were there any changes currently planned for the codes. While nr currently used, the mainframe computer is still available to run the codes, but may be removed in the future.
b.
The inspector reviewed two Change Requests for the installation and modification was approved toof the new install the dissolver new filter.filterIncluded in URU. CR 95-003 in this change was a change to the control system logic, which was implemented by an SSR andoflater completion verified to have been satisfactorily changed by the FTI. The nuclear criticality safety analysis was confirmed to have been done with approved codes by qualified analysts and demonstrated to meet the license requirements. An interaction analysis was performed for the dissolver room and demonstrated that the license requirements were met under accident cenditions. All of the Nuclear Safety requirements were verif ted to be complete and the CR was released by the Area Manager.
Subsequently, another CR was generated to modify the filter after operations problems were encountered. This CR was also reviewed and determined to have been reviewed and approved in accordance l with the license requirements,
- c. The Criticality Safety group developed a training program for engineers as part of their qualification to use the computer codes (and as a refresher for experienced users). At present the training is being conducted with two sessions per week for -
approximately 10 weeks. The inspector reviewed the approved '
training module and the associated materials. These reflect the background information necessary for an individual to become an approved user of the codes. A separate module has been approved for the indoctrination of new employees in Criticality Safety Engineering.
- d. The inspector reviewed the monthly source check records for the criticality monitors for the period May - September 1995 and determined that the monitors were respondia as requ, ired and the alarms were activated when two detectors in an area reached the alarm point. The monthly records indicated that when a detector did not respond properly, a work order was written and the detector replaced; the replacement detector was verified to have been calibrated and source checked after installation with a two point check. The inspector also reviewed the annual calibration records for the test performed December 5-7, 1994.
Within the scope of the inspection, violations or deviations were not identified.
l 1
l 4
- 4. Plant Operations (88015, 88020, 88055) l
- a. During the inspection, the inspector toured various plant areas to observe conditions and operations in progress. Items observed included storage of materials in authorized locations, enrichment of materials in accordance with posted limits, proper storage containers for flammable liquids, and general plant house keeping. l The inspector also observed that fire extinguishers, hoses and '
emergency kitt were unobstructed,
- b. The inspector reviewed the following items related to the fire protection program:
(1) monthly test for September and semi-annual test results in June for the diesel fire pump (2) monthly checks for fire water valves in July and August (3) monthly fire water flow alarm tests for July and August (some areas were secured during shutdown in July and were not tested but were tested in August.
Because of construction activities for the DCF, water flow tests could not be conducted for some sub-systems (this would discharge water into construction areas). However, other portions of the system were tested which indicated flow in the system. Also, during the August monthly fire inspection, a Factory Mutual Insurers representative accompanied the licensee personnel,
- c. During tours of the facility, the inspector observed fire extinguishers and hoses and verified from the inspection tags that the inspection was current. However, the inspector observed one fire extinguisher located in the REDCAP area which had been inspected in January and February but not since, as evidenced by the markings on the inspection tag. When brought to the attention of licensee personnel, action was taken to correct the situation.
The issue of how the fire extinguisher was not inspected for six months when it should have showed up in the monthly inspection records as " missed" was being reviewed by licensee representatives,
- d. On September 12, 1995, a fire broke out in the motor of a dryer in the laundry. When a operator pulled the handle on the fire call box, nothing happened. The fire was extinguished using portable equipmt y . An Unusual Incident critique was initiated to discover why the alarm did not sound when pulled. It was determined that the electrical contacts in the box were corroded so no contact could be made. The contacts were cleaned and checked satisfactorily. In addition, the approximately 160 pull stations around the plant site were tested and none failed to activate. As part of the corrective actions, a review of the PM program for the pull' stations is being made.
i 5
- e. On September 14, 1995 the inspector observed five wooden pallets containing cardboard cartons and/or wrapped in plastic stored in the spent solvent warehouse between two sets of solvent drums.
The NSR/R for the building allows for the storage of the spent solvent but prohibits "any other flammables." While the wood and plastic are not " flammable", the inspector questioned why ;
l
" combustibles" were stored in the location, since the material had '
nothing to do with the solvent. Licensee representatives stated that the pallets had been moved in out of the weather
" temporarily" but acknowledged that " temporary" had become too long. The Area Manager directed that the material be moved to another, suitable storage location, which was accomplished that evening.
- f. A licensee representative advised the inspector that a problem had been encountered with the liner in one of the waste lagoons. This i
was discussed further with the Team Leader. The situation was that the inner liner of a double-lined lagoon had apparently been torn by the dredge. This lagoon is used intermittently for temporary storage of calcium fluoride when-part of Waste Treatment was shutdown. There had been no evidence of leakage from the lagoon. Two alternatives to correct the situation were being developed. On September 15, the Manager, EH&S informed the inspector that a decision had been made and work would be starting to implement the plan.
Within the scope of the inspection, violations or deviations were not identified.
- 5. Review of Previous Inspection Findings (88005, 88015, 88025)
- a. (Closed) Violation 95-02-01 This violation dealt with the use of trash bags and bag holders which did not comply with the nuclear criticality safety requirement that the bag separate from the holder or fail if a safe batch mass were placed inside. The licensee's response dated April 7, 1995 stated that new trash bags would be procured which have a perforated edge which will tear before the safe batch mass is exceeded.
The inspector reviewed CR 95-0274 and the nuclear safety analysis for the use of the new bags. The CR was approved and released by the Area Manager on June 9 1995. During tours of plant controlled areas, the inspector observed that new, perforated bags
- were in use and the holders had been modified to prevent sealing the bag in the holder. Licensee representatives also demonstrated that the bag would tear free when a 25 kilogram weight was placed inside. The inspector had no further questions.
o l
- b. (0 pen) IFl 95 02 02 This IFl dealt with the use of various stamps on drawings to I indicate that the drawing was a "controllto copy." There were several different stamps which had been used and drawings in one location had different stamps, creating confusion as to their status.
The licensee initiated the use of a new stamp and replaced the drawings in the control room with a set of drawings all stamped the same. Howevar, in the ECC, the drawing files contained ;
drawings with the different " controlled" stamps and also had some drawings stamped " uncontrolled copy". The cognizant leader stated that the ECC drawing file would be updated by November 1 to contain only controlled orawings, in the meantime, the licensee '
is developing a method to control drawings not developed or revised by the Configuration Management Center, such as facilities or utilities drawings. This item will be reviewed during subsequent inspections,
- 6. Exit interview On September 15, 1996, the scope of the inspection and the findings were discussed with those persons identified in Paragraph 1.
The inspector informed licensee management representatives that Violation 95 02 01 was closed based on the use of new trash bags.
IFl 95 02 02 remained open pending the resolution of " controlled' drawings in the ECC.
While not specifically addressed as IFIs, the ins)ector stated that the licensee's actions on PH for the fire alarm pull soxes and the repair of the waste lagoon liner would be reviewed during future inspections.
Although proprietary documents were reviewed during the ins)ection, the proprietary nature of the documents has been deleted from tiis report.
r _ _ _
APPEND 17.A Acronyms and IniticLtima i
DCP Dry Conversion Project ECC Emergency Control Center EH&S Environmental, Health & Safety FET Fuel Examination Technology FM0 Fuel Manufacturing Operation FT! Functional Test Instruction GE NE General Electric Nuclear Energy-GE NEP General Electric Nuclear Energy Production IR inspection Report NQii Nuclear Quality Assurance PM Preventive Maintenance
- P/P Policy / Procedure !
PROD Process Requirements and Operator Documents i
- SNM - Special Nuclear Material SSR Software Service Request URLS/WT Uranium Recovery from Lagoon Sludge / Waste Treatment URV Uranium Recovery Unit WSRC Wilmington Safety Review Committee f
l May 16, 1996 l
General Electric Company ATTN: Mr. C. P. Kipp, General Manager GE Nuclear Energy Production P. O. Box 780 Wilmington, NC 28402
SUBJECT:
NRC_ INSPECTION REPORT NO, 70 1113/96 05
Dear Mr. Kipp:
This refers to the inspection conducted on April 15 19. 1996, at the Wilmington facility. The >urpose of the inspection was to determine whether activities authorized by t1e license were conducted safely and in accordance with NRC requirements. At the-conclusion of the inspection, the findings were discussed with those members of your staff identified in the report.
Areas examined during the inspection are identified in the report. Within these areas, the inspection consisted of selective examinations of procedures and representative records, interviews with personnel, and observation of activities in progress.
Within the scope of the inspection, violations or deviations were not identified. .
In accordance with 10 CFR 2.790, of the NRC's " Rules of Practice," a copy of this letter will be placed in the .C Public Document Room.
Should you have any questions concerning this-letter, please contact us.
Sincerely,
/s/
Edward J. McAlpine, Chief Fuel Facilities Branch Division of Nuclear Materials Safety Docket No. 70 1113 License No. SNH 1097
Enclosure:
NRC Inspection Report cc w/ enc 1: (See page 2)_ l I
<KJO 070029 A b 13 C PDR- gq
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l GE 2 cc w/ enc 1:
Dr. Ralph Reda, Hanager Fuels and Facility Licensing General Electric Cc,maany P. O. Box 780. Hail Code J26 Wilmington, NC 28402 Dayne H. Brown. Director Division of Radiation Protection N. C Department of Environment, Health & Natural Resources P. O. Box 27687 Raleigh, NC 27611 7687 Distribution w/ enc 1:
E. McAlpine, Ril G. Troup R. Bellam G. Shear,y, RI RIII C. Cain, RIV .
F. Wenslawski, RIV PUBLIC D'stribution w/o enc 1:
L' cense fee Managew;nt Branch MF1ff P f l @% ttf M
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Nyt AGooden Gir o DATE 05 /dv / 96 05 / C4 / 96 05 / / 96 0$ / / 96 05 / / 96 05 / / 96 COPY? [Y[$ NO YE5 NO Y[$ NO Y[$ NO Y[$ NO Vf5 NO Of FETALW)kD u COP 00Qtthl kW:15:\fA55\0EETAG
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Report No.: 70 1113/96 05 Licensco: General Electric Company l
Wilmington, NC 28401 l Docket No.: 70 1113 License No.: SNM 1097 Facility Name: Nuclear Energy Production Inspection Conducted: April 15 19, 1996 Inspector: O. ethb ._. ~p 6 /ff, A. Gooden, Radiation Specialist 5fgned Approved By: DW M E. McAlpine, Chief '
M6[96 Date Signed Fuel Facilities Branch Division of Nuclear Materials Safety
SUMMARY
Scope:
This routine, announced inspection was conducted to assess the operational readiness of the licensee's emergency preparedness program. Areas reviewed included the coordination and interface with offsite authorities on emergency planning; training for onsite and offsite response personnel; facilities and equi) ment maintenance: independent audits including the corrective action tracting in response to drills, exercises, audits, and inspection findings:
the administrative and procedural program for notification of emergency response personnel; and program changes since the June 1994 inspection.
Results:
Within the areas reviewed, a non cited violation was identified for failure to certify an Emergency Response Team member was physically able to use respiratory protection equipment (Paragraph 5). An Inspector Followup Item was identified to assure the timely activation and steffing of the li .ansee's emergency response organization (Paragraph 4). With the exception of the aforementioned items, the licensee's program was adequately maintained.
Contact with two of the offsite support agencies disclosed the licensee was maintaining a very effective relationship with frequent contact to discuss matters of a mutual concern.
Enclosure
-960s070rs r vsos m PDR iI/('
ADOCK 07001113 C PDR
i REPORT DETAILS
- 1. Persons Contacted Licensee Employees l
- J. Bradberry, Senior Program Hanager, Regulatory Team
- R. Foleck, Senior Licensing Specialist
- P Godwin, Coordinator Fire Safety and Emergency Response
- J. Harmon, P4ef Technologist and Emergency Director
- R, Keenam Hup. Site Security and Emergency Preparedness i
- J. Kline, hwga , Chemical _ Production Line l A. Habry, hcH# Nuclear Safety Engineer
- S. Hurray, Manager Nuclear Safety
- R. Reda, Manager, feels and Facility Licensing
- B. Robinson, Principal Nuclear Safety Engineer
-*W. Sependa Manager Environment. Health and Safety
- H. Strickler Manager Environmental Protection and Industrial- Safety
- R. Torres, Auditor, Compliance Auditing
- T. Winslow, Manager, Haterial Control and Accountability Otherlicenseeem!ians, engineers, techni administrative,andsecuritypersonnel.loyees contacted durin Other Organizations L. Ray, Director New Hanover County Emergency Medical Services D. Summers Director New Hanover County Department of Emergency Management
- Attended Exit Interview An index of abbreviations used throughout this report will be found in the last paragraph.
- 2. Offsite Support Agencies (88050)
Discussions were held with a member of the licensee's staff regarding the coordination of emergency planning with offsite support agencies.
The inspector reviewed the status of agreement letters with offsite support agencies, and noted that with one exception, current agreements were maintained on file by the Site Security and Emergency Preparedness Manager. The exception involved the absence of a current agreement with REACTS. When questioned regarding the availability of an agreement letter, the inspector was informed by the licensee contact that an agreement was no longer on file: however, during the next Plan or agreement letter update, an agreement letter will be executed with REACTS, The absence of an agreement letter was also identified by the licensee during an independent audit conducted in September 1995, and was assigned to the licensee's tracking system for corrective actions.
The adequacy of the licensee's interface with offsite support agencies was determined via telephonic discussion with representatives of two
l l
2 offsite support organizations (see Paragraph 1). The contacts described i
a very close working relationship with excellent communications between the licensee and county. County representatives cited as examples the licensee's involvement in an offsite planning partnership, and the frequent contact to discuss matters of a mutual concern such as l
organizational changes, training, and periodic site tours. Section 7.2 of the RC&EP required the licensee to offer periodic training to offsite groups. The inspector reviewed documentation to show that during March 1995, 21 members of the VFD (Castle Haynes, Wrightsboro, and Ogden) participated in a site tour to discuss site chinges and planned or future changes. During the period of the inspection, the licensee conducted training at the New Hanover Regional Medical Center.
i Gased on interviews with offsite contacts and documentation to show periodic t "ining and site familiarization tours, the inspector concludei r *he licensee's interface with offsite support agencies was very & ive in assuring that the offsite authorities were familiari F *.1eir role, responsibilities, and the potential impact of an inciden. ne interface with the offsite authorities was considered a program t,trt ;,th.
No violations or deviations were identified.
- 3. Emergency Plan and Implementing Procedures (88050)
The review, ap)roval, and distribution of Plan changes was examined to determine whetler significant changes were made since the June 1994 inspection, to assess the impact of any changes on the overall state of emergency preparedness at the facility, and determine if the Plan implementing procedures (referred to as Emergency Procedures) were revised to reflect those chan are found in 10 CFR 70.22(1).ges, Requirements applicable to this area Since the June 1994 inspection, two plan changes (Revs. O and 1) were submitted for NRC review. The licensee had submitted Rev. 0 (dated December 2, 1994) to the RC&EP, to address numerous concerns that were identified by NRC in a letter dated January 5, 1994, and inconsistencies identified during the June 1994 inspection (see IR 70 1113/94 07).
Rev. I was merely administrative changes (e.g. page numbers and titles).
The aforementioned changes were ap) roved by NRC via letters dated January 10, 1995 (Rev. 0) and Marc 1 1, 1995 (Rev. 1). The RC&EP in effect at the time of the inspection was Rev. I dated February 1,1995.
The licensee maintained nine EPs for implementing the requirements of the Plan. All procedures were revised and reissued during March 1996.
Current copies of both the RC&EP and EPs were available in the ECC, staging areas, and security. The inspector reviewed documentation to confirm that Rev. 1 to the RC&EP was distributed to both onsite and offsite copyholders in a timely manner. According to the documentation.
Rev. 1 changes were issued to copyholders on March 8, 1995.
3 .
In addition to the changes regarding the RC&EP and EPs, the emergency preparedness program was in a state of transition. Effective January 4. 1996, a personnel change resulted in the reassignment of the responsibility for emergency preparedness. Previous personnel was assigned responsibility as Manager, Emergency Preparedness Security, Material Control and Accountability. Recent change resulted in a newly appointed individual with other collateral duties and the title HSSEP.
.l Currently, in addition to emergency preparedness and security, l management responsibilities include all other aspects of emergency preparedness such as the ERT (also known as fire brigade), EHT, and fire safety, Previously, the re>orting chain for the ERT and fire safety was l the Manager, Environmental lealth and Safety. There were no changes to l the individuals assigned the day to day responsibilities for those additional areas. The ins)ector determined via interview and documentation that the MSS:P training, and experience in emergency
>reparedness, was quite extensive due to involvement as a Radiological
)lanning Officer for New Hanover County, and experience in conducting independent audits of the RC&EP as a member of the licensee's compliance audit staff. Regarding the ERO, the impact of this change should be minimal, with no affect on the licensee's capability to respond in light of the following: no changes were made to personnel aerforming day to-day support activities to emergency preparedness: altlough reassigned, the previous Managers of Emergency Preparedness both remains onsite and are available for assistance /consultatien; and the licensee's ERO staff is one of experience and stability with no recent personnel changes in key positions, No violations or deviations were identified.
- 4. Emergency Facilities and Equipment (88050)
Facilities and equipment were inspected to determine whether the licensee's ECC, emergency response equipment, instrumentation, and supplies were maintained in a state of operational readiness, and to assess the impact of any changes on the emergency preparedness program.
Requirements applicable to this area are found in Section 6 of the RC&EP.
The inspector reviewed the operational readiness status of equi) ment and facilities via facility walk down, inventory, and operability c1ecks.
With very minor exceptions, equipment, supplies, and instrumentation stored at the ECC and staging areas were not only operational but adequate inventories were available for response personnel, The exceptions involved an inoperable telephone, an outdated phone listing, an inoperable survey instrument, and a bull horn, in response to the inoperable ecuipment, the licensee took immediate actions to determine the cause anc return equipment to an operable status. It was determined that the survey instrument and bull horn batteries were in need of replacement. Instruments were calibrated and responded properly to both battery and radioactive source checks. Air tanks were examined for SCBA units stored onboard the ERT vehicle and determined to be full and ready for use. One additional aspect of assessment equipment observed by the
o 4
inspector as operational and within calibration was the onsite meteorological monitoring station located in the ECC. When records were reviewed for equipment surveillance performed in the ECC during the period September 1994 to June 1995, no problems were noted.
The inspector reviewed the licensee's backup power provisions in the event a loss of power occurred. According to Section 5.3.4 of the RC&EP, emergency power is provided for a supervised alarm system and essential equipment. The inspector observed a weekly test of the 150 KW diesel operated generator which provides an automatic startup of the emergency generator in the event of a power failure. No problems were noted, the diesel was operated for several minutes.
According to the licensee contact, the following facility or equipment changes had been made since the last inspection of this area:
e A new site alarm system was being installed at the time of this inspection. The anticipated date for completion of installation, testing, and turnover to operations is May 6, 1996. An interview with security personnel disclosed that training on the new system was in progress. .
- A cellular phone was maintained for use by the On Scene Incident Commander for communicating details from the scene of the incident -
to the ECC without influence or cross talk resulting from other response personnel operating on analogous radio frequency.
e Two (2) TV monitors were installed within the ECC and ECC bay area providing the capability to observe activity involving decision-making (ECC) or team deployment (ECC bay area) during an emergency.
e- An onsite capability for refilling SCBA equipment was procured during CY 95, e One staging area was relocated due to access during severe weather, and a new staging area was built to accommodate the FET and DCP buildings, o The capability for accessing the GE Information System (RDMS.
plant drawings, etc.) was recently installed in the ECC.
The licensee's notification system for activating the ERO during off.
hours was reviewed for effectiveness in staffing the ECC. During off-hours, ERO personnel are contacted by security >ersonnel via pager-and/or telephone. The inspector discussed wit 1 the licensee contact and reviewed documentation to show that on a quarterly basis, notification drills were conducted for determining the availability of ERO staff during off hours. Drills were conducted as follows: 1) security activates group pagers: 2) when paged, the ERO member calls the number listed within 15 minutes and provide their name and the
- anticipated travel time to arrive onsite: 3) fifteen minutes after
e 5
activating pagers, security check the recorded messages for res)onders ;
providing ETA: 4) in the event no response provided, security pione the member's residence: and 5) if the primary is unavailable, security contacts the backup or alternate. The inspector reviewed records covering the period August 1994 to March 1996, and noted that three of seven tests disclosed poor results. The test dates were August 16, 1994, February 1, 1995, and most recently March 28, 1996. ERO personnel delayed contacting security for more than thirty (30) minutes to provide an estimated time of arrival. During the February drill, 59 minutes elapsed before one individual contacted security with an ETA: and for one position, the primary nor alternate responded to the pager test and I security was unable to contact the primary or alternate via telephone.
The most recent drill results (March 1996) disclosed from 17 individuals paged, five responded to security page within 34 35 minutes after notification: no response was obtained from three individuals (phone /pager), and nine members of the ERO were not contacted via phone (or pager) by security for almost an hour. Based on the documentation review, the inspector discussed with the licensee contact assigned res)onsibility in this area, the adequacy of administrative and >hysical mec1anisms to ensure the timely and appropriate staffing of the ECC during off. hours, In response to the inspector's comments and the recent drill results, the licensee conducted an unannounced off hours pager drill on April 17, 1996 and several problems were noted by the licensee and NRC inspector: several pagers failed to activate; actions by
' security personnel were not consistent with procedures and ERO personnel were in some instances not following pager guidelines, In response to the aforementioned observations -the licensee reviewed all aspects of the administrative and )hysical mechanism for off hours staffing and activation and took tie following immediate actions:
- 1) paper listing was reviewed and updated with current information:
- 2) lead security personnel for evening shift was given informal training
-on activation procedures: and 3) an unannounced remedial drill was conducted on April IB,1996, during off hours. The inspector and licensee contact noted significant improvements in the res>onse times during the remedial drill from the previous drill (which t1e licensee attributed to an outdated pager listing), but problems were again noted in the areas of human errors and procedure (e.g. security personnel entered an incorrect pager signal). Consequently, the licensee discussed the following corrective actions to resolve the various problems in the area of timely staffing and activation of the ERO:
- All personnel assigned to the group pagers were informed of the above problems via memo on April 19, 1996, e A discussion with site staff regarding pager problems, lessons learned, and pager responsibilities was planned for April 23, 1996, e Revise and simplify security procedures regarding the
-implementation of group pager notifications and tests, e
Train all security personnel on revised procedures when completed.
+
6
- Rewrite and redistribute pager responsibilities to all ERO personnel.
Periodically verify as current the computer printout of the group Call list from the paging company.
The inspector informed the licensee that the corrective actions to ensure the timely and effective activation and staffing of the ECC was considered an IFI for review during a subsequent inspection.
IFl 70 1113/96 05 01: Verify the adequacy of corrective actions to ensure timely activation and staffing of the ECC.
During the facility walkdown discussed above, the inspector accompanied by a licensee representative during backshift hours, examined the '
evacuation routes leading to the new staging area for accessibility during evacuation, the amount and/or condition of lighting, and the operability of equipment in all onsite staging areas. The site tour disclosed no areas of impediment to evacuation. However, in light of construction activity and equipment associated with the DCP facility, areas were noted by the licensee and inspector where the installation of evacuation signs would ensure evacuees travel via the most expedient route to the staging areas. According to the licensee contact, as site characterization changes due to the construction activity, the adequacy of existing signs and evacuation routes to the new staging area will be further evaluated.
Based on the records reviewed, interviews, and facility walkdown, the inspector determined that the licensee was maintaining equipment in a state of readiness, No violations or deviations were identified.
- 5. Oriils and Training (88050)
Emergency response training was reviewed to determine if the licensee was providing training in accordance with the RC&EP. The requirements for training are found in Table 7.1 of the RC&EP. Section 7.3 of the RC&EP required periodic drills to maintain proficiency in emergency response.
The inspector reviewed Table 7.1 for a description of the training 3rogram for Emergency Directors Emergency Staff Advisors, Site Emergency Support Groups (Building Managers), and the Emergency Response Team. The inspector requested training documentation for randomly selected personnel listed on the emergency notification roster (dated April 1. 1996) and the ERT roster. Based on the review of documentation and interviews with personnel assigned the responsibility for training and/or tracking training, with one exception, selected personnel training was current and up to date. The exception involved personnel assigned to the ERT. Ten names were randomly selected from the ERT roster for training verification, included as part of the training
7 verification, the inspector also confirmed that the individuals received a medical examination and were respirator qualified as described in procedures P/P 40 22 and NSI 0 1.0. The referenced procedures also implemented the requirements in 10 CFR 20.1703 (a)(3)(iv), and 20.1703 (a)(3)(v). Three of ten names selacted had not been medically certified for respirator use. Two of the individuals were recent volunteers to the ERT (late 1995 and early 1996). However, one of the three individuals joined the ERT approximately three years ago. According to documentation and an interview with medical staff, the last medical review and certification was performed during 1993. Further, documentation and an interview with a licensee contact disclosed during the period 1993 to present, the aforementioned individual used SCBA equipment during drills and in response to an actual emergency (non-radioactive) on April 21, 1995, involving a compressor fire in the FC0 building. Procedural requirements previously stated that " respirator users must be trained in the proper use of respirators specified for their job assignment; and must have respirator medical clearance reevaluated annually." 10 CFR 20.1703 (a)(3)(v), prior to March 1995, ;
required " determination by a physician arior to initial fitting of respirators, and at least every 12 montis thereafter, that the individual user is physically able to use the respiratory protection equi ment." Effective March 13, 1995. NRC amended the rules in 10 CrR 20.1703 to require " determination by a physician prior to initial fitting of respirators, and either every 12 months thereafter or periodically at a frecuency determined by a physician, that the individual user is mecically fit to use the respiratory, protection equipment." In response to the amendment, the licensee s procedures and license (Section 3.2.4.5.2) were revised to reflect the reduced frequency. Irrespective of the referenced revisions in calendar year 1995, personnel failed to obtain medical certification for more than two years. The licensee was informed that failure to medically certify ERT members for respiratory protection equipment was a violation of 10 CFR 20.1703(a)(3)(v) and site procedures which implements the respiratory prctection program. The licensee acknowledged the finding and indicated that the lack of certification resulted due to the administrative system used for tracking the training of non radiation workers (e g. ERT members from Aircraft Engines) in comparison to radiation workers (tracked via RDMS). In response, the licensee took immediate action as follows:
- Responder was administered a vital capacity test and certified as medically fit for respirator use.
- The remaining two ERT members were notified that a vital capacity test was required during their annual physical, o Medical records for all ERT members were reviewed to verify that members had a vital capacity test and certified as medically fit for respirator use.
- Hedical records were compared to the RDMS tracking system data base and the tracking system updated as necessary.
l I
- 8 e All ERT members not previously tracked via RDHS, will be added to RDMS under the training requirement of RP 201 to require medical staff perform a vital capacity test during the annual physical, e Personnel with responsibility for ERT will be granted access to RDHS tracking system to periodically review ERT certification status.
In light of the above actions, this NRC identified violation is not being cited because criteria specified in the NRC Enforcement Policy were satisfied. The licensee was informed that this finding was considered a NCV, NCV 70 1113/96 05 02: Failure to certify a member of the ERT in accordance with 10 CFR 20.1703 and site implementing procedures that the ,
individual is medically fit to use respiratory protection equipment, l The inspector reviewed dor:umentation to confirm ERO training was provided and attended by key personnel assigned to the ERO. The inspector compared the training attendance roster with the current emergency organization listing dated April 1,1996..for verification of training. No problems were noted. In addition to the records review, to assess the effectiveness of the licensee's training program, the inspector interviewed two members of the licensee's staff with i
1 responsibility in the area of accident assessment and dose calculation.
Interviewees were very familiar with their role ar.d responsibility during an event. As a tool for reviewing the performance of interviewees, a postulated accident with offsite consequences was validated by a member of the licensee's= staff and presented to interviewees for taking the a)propriato actions. Members of the licensee's staff along with t1e inspector observed the interviewee's performance in manual dose calculation and protective actions. No significant problems were noted. Personnel demonstrated an excellent use of procedures, and understanding of various assumptions and defaults in the procedure for manual dose calculations.
One NCV was identified.
- 6. Independent Review / Audits This' area was reviewed to determine if the licensee had performed an independent review or audit of the emergency preparedness program.
An independent audit was conducted during September 1995 by personnel from the Nuclear Quality Assurance group, and documented as Com)liance Audit No. 36/95 1. The referenced audit included a review of tie NRC ap) roved RC&EP (Rev. O, dated December 2, 1994, and Rev. 1. dated Fe>ruary 1, 1995), and EPs. One concern was identified regarding the adequacy of the RC&EP in addressing the OSHA regulations for Emergency Plans in the non nuclear buildings onsite. The audit program was
l l 9 l effective in the identification of areas requiring improvements and/or corrective actions. Audits were detailed and compliance oriented to ensure license commitments in the RC&EP were satisfied.
The licensee utilized the plant wide tracking system (known as Safety Compliance and EHS Activity Items) to monitor open items and issues in emergency preparedness resulting from NRC inspections, independent audits, and actual incidents. A review of the computer print out showed that the system was appropriately detailed, and indicated for each item the responsible individual and status of completion. The inspector determined that the licenses was effectively using the referenced tracking system as a tool to ensure the completion of corrective action for identified aroblems in emergency preparedness. In addition to the referenced tracting system, the Training Officer for the ERT was tracking items resulting from ERT activities and providing a weekly status update to plant management.
No violations or deviations were identified. !
- 7. Exit Interview The inspection scope and results were summarized on April 19, 1996, with those persons indicated in Paragraph 1. The inspector described the areas inspected and discussed in detail the inspection results listed below. Dissenting comments were not received from the licensee.
Proprietary information is not contained in this report.
Item Number 5111m Descriotion and Reference 70 1113/96 05 01 Open IFI Verify the adequacy of corrective actions to ensure timely activation and staffing of the ECC (Paragraph 4).
70 1113/96 05 02 Closed NCV Failure to certify a member of the ERT in accordance with 10 CFR 20.1703 and site implementing procedures that the individual is medically fit to use respiratory protection equipment (Paragraph 5).
- 8. Index of- Abbreviations Used In This Report CFR Code of Federal Regulation DCP Dry Conversion Process ECC Emergency Control Center ED Emergency Director EHA Emergency Hanagement Agency EHT Emergency Medical Team EP Emergency Procedure EPA Environmental Protection Agency ERO Emergency Response Organization
10 ERT Emergency Response Team ETA Estimated Time of Arrival FC0 Fuel Components facility FET Fuel Examination Technology l GE General Electric IFI Inspector followup Item l IR Inspection Report KW Kilowatt HSSEP Hanager Site Security and Emergency Preparedness NSI Nuclear Safety Instruction OSHA Occupational Safety and Health Administration P/P Practices and Procedures PAG Protective Action Guide PAR Protective Action Recommendations RC&EP Radiological Contingency and Emergency Plan RDHS Radiation Data Management System REACTS Radiation Emergency Assistance Center Training Site REV. Revision SAE Site Area Emergency '
SCBA Self Contained Breathing Apparatus VfD Volunteer Fire Departmer,t ,
9 6
~
, June 13, 1996 i
General Electric Company ATTN: Mr. C. P. Kipp, General Manager GE Nuclear Energy Production P. O. Box 780 l Wilmington, NC 28402
SUBJECT:
NRC INSPECTION REPORT NO. 70 1113/96 06
Dear Hr. Kipp:
This refers to the inspection conducted on May 20 24, 1996, at the Wilmington facility. The >urpose of the inspection was to determine whether activities authorized by tie license were conducted safely and in accordance with NRC requirements. At the conclusion of the inspection, the findings were discussed with those members of your staff identified in the report.
Areas examined during the inspection are identified in the report. Within these areas, the inspection consisted of selective examinations of procedures and representative records, interviews with personnel, and observation of activities in progress.
Within the scope of the inspection, violations or deviations were not identified, In accordance with 10 CFR 2.790, of the NRC's " Rules of Practice," a copy of this letter will be placed iin the NRC Public Document Room.
Should you have any questions concerning this letter, please contact us.
Sincerely, Original Signed by D. Collins for:
Edward J. McAlpine, Chief Fuel Facilities Branch Division of Nuclear Materials Safety Docket No. 70 1113 License No. SNM 1097
Enclosure:
NRC Inspection Report cc w/ encl: (See page 2) e g>l QW) 2fI'
GE 2 cc w/ enc 1:
Dr. Ralph Reda. Manager Fuels and Facility Licensing General Electric Company P. O. Box 780. Hail Code J26 l
Wilmington, NC 28402 Dayne H. Brown. Director i
Division of Radiation Protection N. C. Department of Environment, r Health & Natural Resources P. O. Box 27687 Raleigh NC 27611 7687 Distributionw/ enc 1:
E. McAlpine. Rll G. Troup. RIl R. Bellamy. RI G. Shear, Rill C. Cain. RIV .
F. Wenslawski. RIV PUBLIC Distribution w/o enc 1:
License fee Management Branch
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EMc41016e DATE 06 /\I' / 96 06//I/96 06 / / 96 06 / / 96 06 / / 96 06 / / 96 COPYT Vt5 NO Yt1 NO v[5 NO VE5 NO YES NO Yt1 NO Of f AllAL hlC(40 C0i'1 U0CUMENI h4%: 4 : \f f Bl1\EiMI5\GEit600.GL)
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.....f Report No.: 70 1113/96 06 Licensee: General Electric Conpany Wilmington, NC 28401 i
Docket No.: 70 1113 License No.: SNH 1097 Facility home: Nuclear Energy Production Inspection Conducted: May 20 24, 1996 Inspector:
6J G. L. Troup, Sr. Fuelpfacility Project Inspector Date Signed k/ b[dWA--
~
Approved by:
I, //3/94 j E. J. McAlpine, Chief
(/p4 Fuel Facilities Branch Efate'51gned Division of Nuclear Materials Safety
SUMMARY
Scope:
This unannounced inspection was focused on the inspection of the installation of the final membrane of the roof of the Direct Conversion Facility (DCF),
which is intended to be a Moderation Exclusion Area. In addition, the inspection included organization changes, nuclear criticality safety, facility operations and review of the corrective actions for previously identified inspection findings.
Results:
The final stage of roof construction for the DCF was in accordance with the design criteria and associated construction documents. Organization changes and appointment of new managers was in accordance with license requirements.
Two previously identified inspector follow up items (IFIs) were closed. One violation a3 proved (non cited) relating to the failure to follow the requirements of 3rocedures and nuclear criticality safety requirements was identified
()aragrap1 B).
Enclosure
[]Qf G u " Y!
a REPORT DETAILS
- 1. Persons Contacted
-Licensee Employees
- H. Chilton, Manager, Joint Conversion Project
- A. Del Priore, Engineer. Environmental Programs
- 0, Dowker, Team Leader, Fuel Support
- R Foleck, Sr. Licensing S>ecialist
- R. Keenan. Manager, Site Security & Emergency Preparedness
- H. Lamb, Team Leader, Powder Pre)aration and Packaging
- A. Lehmann. Principal Engineer. 'JRU Team
- A. Habry Principal En !
- 0, McCaughey Manager.gineer Nuclear SafetyFuel and ChemHet Lab Quality
- R. Mciver, Manager. Facility Construction, JCP S. Hurra
- R. Pace,yProject Manager NuclearFacility Manager. SafetyIntegration
- L. Quintana Mana
- R, Reda, Manager,ger Fabrication Fuel and Product Line facility Licensing
- S. Selby. Team Leader. VO Production
- W. Sependa. Manager GE.Nk EH&S .
- G. Smith. Team Leader, FHO Haintenance Support
- H. Strickler, Manager. Environmental Protection and Industrial Safety
- C Tarrer
. *J. Taylor ,Team Leader, Principal Configuration Engineer. NuclearManagement Criticality Safety
- K Theriault. Team Leader URV
- C, Vaughan. Project Manager. New Facility Licensing / Safety C. Williams, Team Leader, URLS/WT
- T. Winslow, Manager, Material Control & Accountability The inspector also interviewed other licensee employees including area coordinators, engineers, operators, and technicians.
- Attended exit interview on May 24, 1996.
Acronyms and initialisms used in this report are listed in Appendix A.
- 2. Facility Status (88020)
The GE Wilmington site (covering NEP and Aircraft Engine facilities) was designated a STAR facility by North Carolina OSHA, under the OSHA Voluntary Protection Program (VPP). VPP endeavors to promote and recognize effective safety and health management programs. It reduces OSHA inspections and emphasizes the ability of the facility to conduct
-its safety programs.
- 3. - Organization and Staffing (8800S)
Since the last inspection, three changes were made in the management organization. On lay 6.- 1996. L. Quintana assumed the position of Manager Fuel Fabrication Product Line. T. Flaherty was appointed as
Start up Manager for the Direct Conversion facility.
Team t.eader for Waste Treatment /VRLS. These positionsC.areWilliams became not specifically described in Part I Chapter 2 of the license ap No specific qualifications are sp,ecified for these positions. plication.
Within the scope of the inspection, violations or deviations were not identified.
- 4. Direct Conversion Facility Roof Construction (88020)
- a. The facility is designed as a " moderation exclusion area where no moderating materials will be permitted, exce)t under specifically s powder is produced designated or conditions, handled. A principal feature in those of the v eas where :J0, tor "modera exclusion" f
\
principle is the construction of a roof which precludes any leaks -
of rain into the facility. Details of the roof construction are described in JCC drawings in the A21 ARXX series of drawings and the roofing material supplier project description. The DCF roof is designed to provide multiple barriers to any leakage or seepage of external moisture into the DCF. Details of the roof design and construction are discussed in IR 70 1113/96 02. Paragraph 4.
- b. During the inspection, installation of the final (top) roof membrane began. The inspector observed the preparation of the surfaces, installation of the membrane, and preparation of the edges for sealing. The inspector also observed the condition of sealed joints which had already been made. Through direct observation and discussions with personnel wrforming the installation, the inspector verified that tie membrane installation was being performed in accordance with the requirements specified in the E/C roof insta11atien documents.
There were several spots in the installed membrane which appeared to be air bubbles or foreign material inclusions. A licensee representative stated that the E/C roof cxpert wo'u ld be brought in to inspect the membrane and determine those areas which required repair. Repairs will be in accordance with the E/C requirements.
- c. The inspector also observed that the " bent plates" had been installed around the facility to hold the lower roof membrane in place. The edge trim. metal eave trim and drain gutter have to be installed.
Within the scope of the inspection, violations or deviations were not identified.
- 5. DCF Integration (88020)
- a. The DCF will produce VO powder, which will then be transferred to existingworkstationsinFH0/FH0Xfortheproductionofpellets.
Because the existing ADV lines will remain in production during the DCF start up and qualification, a plan for the installation of new facilities, changes in equipment and construction of moderation exclusion ~ areas is required,
3 b.
l The inspector discussed the plan for accomplishing the change.over l while maintaining both existing and new nuclear safety criteria. A plan has been drafted and is currently being reviewed by NEP, DCF, and E/C personnel. Approval of the plan is presently projected for June. Once the plan is a l schedule will be developed. p3 roved, apersonnel
.icensee detailed integration stated that the NRC would h kept informed of the status of the integration plan,
- c. A staffing plan has been developed for the DCF and candidates have applied for the openings. Based on the persons selected, then positions in other areas, such as Chemical Conversion, URV or maintenance, will be filled by promotions and new hires. The-licensee is including the plan-for filling vacant positions and training persons for the new positions in the integration plan.
Training of the DCF operators will start after replacements have been trained. Some personnel have already been selected and are in training at the equipment supplier's facility.
, The inspector had no further questions at this time.
- 6. Facility Changes (88020) d
- a. As discussed in IR 70 1113/96 203, Paragraph 5. the licensee is installin leacher")g a system to recover uranium from incinerator ash (" ash The CSAs for the leacher system and the interaction with surrounding equipment were reviewed in that report and found to be adequate,
- b. The inspector reviewed the Change Request (CR #95 371) file for the ash leacher and discussed the schedule with the cognizant engineers. The inspector noted that the leacher s
" approved for installation at risk" on October 2 ystem1995,had and been had been "a> proved for installation" on March 20, 1996. The Technical Report lad been approved by the Area Manager on April 2,1996.
The system tests and the AEC functional tests had not been completed but were being planned. Final release for operations with licensed material will require final approval of the NSR/R.
Within the scope of the inspection, violations or deviations were not identified.
- 7. Operations (88020)
- a. In IR 70 1113/95 08 when a fire alarm pu. Paragraph 4.d a situation was identifiedll box did small fire in the contaminated clothing laundry. Part of the corrective action was to review the preventive maintenance program for the pull boxes. During the intervening period since that inspection, a new alarm system has been installed. This system includes a circuit which periodically automatically tests the operability of the circuit. While the old pull boxes are still installed in the plant areas, they are covered with adhesi'le tags
l 4
stating they are inoperable. Other signs tell personnel to call a specific telephone number if the box does function. The new system boxes were observed throughout the plant and were observed to have the function lines flashing. The inspector discussed the status of removal of the old pull boxes and alarm system with the cognizant supervisor. As the new system is functional, the issue of a preventive maintenance program for the old system is moot, l b. The inspector toured various plant areas during the inspection to observe operations, plant conditions and general conditions.
During a tour of the construction areas of the DCF, the inspector observed a fire extinguisher which had not been inspected in two months, as evidenced by the inspection tag attached to the handle.
When this was brought to the attention of a licensee representative, action was taken immediately to replace the fire extinguisher.
Cleanliness and housekeeping in plant areas toured were good. SNH containers were stored in authorized locations and in approved arrays. Flammable materials were stored in approved containers and there were no accumulations of rags. Construction areas in FH0/FH0X were unclutered and materials staged to prevent tripping hazards. Floors in liquid processing areas were dry with no visible accumulations of dried solids.
Within the scope of the inspection, violations or deviations were not identified.
- 8. Reportable Event Review (88020)
- a. On November 26. 1995, the licensee made an report under the requirements of NRC Bulletin 9101 (NRC'EN 29641) because U0 powder had been found under the muffle and on the sides of tile muffle in the oxidation furnace in the REDCAP area. A total of-22 kgs, of zwder were found. This was determined to be from cracks in t1e welds in the muffle. The lic-e initiatri a root cause investigation of the powder spill.
The inspection of the muffle found a " yellow" .naterial. l..ch was presumed to be ADU. ADU processing is not allowed in the REDCAP furnace. X ray diffraction analysis of the material determined that the material was produced from UNH rather than ADV sludge so moder4 tion control had not been lost.
The investigation determined that the root cause was less than adequate preventive maintenance of the muffle welds, in that the aired. Corrective welds actionshad previously included failed and adding periodic had been inspection rep / examination of the muffle to the operating procedure-and including inspection of all
1 l
6 The licensee was informed that this licensee identified violation will not be subject to enforcement action because the licensee's efforts in identifying and correcting the violation meet the criteria specified in Section VII.B of the Enforcement Policy (NCV 70 1113/96 06 01).
- 9. Review of Previous Inspection Findings (92701)
- a. (Closed) IFI 95 06 01 This IFI was opened as the result of an inspection.into a reportable event concerning the loss of operability of the in line monitors in the fluoride waste discharge line in the Chemical-Area. As discussed in IR 70 1113/95 06, this IFI was opened to track the licensee's identification of long term corrective actions-and lessons learned which might come from the investigation of this event.
The inspector reviewed the investigation report for this event.
Both short term and long term actions were identified and responsibility assigned for each action.
This IFl is closed as he long term actions were identified and assigned for action.
- b. (0 pen) IFl 95 06 02 This IFI was opened to revbw the completion of the long term actions identified by the investigation team.
The inspector reviewed the status of the long term actions. Four of the five action items had been completed and documented. One item regarding up grading of the detector software and performance of the necessary functional tests following the software changes were not yet complete.
This item remain open until the detector software change is completed.
- 10. Exit Interview On May 24, 1996, the scope of the inspection and the findings were discussed with those persons identified in Paragraph 1.
The inspector informed management representatives of the findings:
NCV 96 06 01 - Failure to follow procedures for handling boats in the REDCAP-furnace hood.
IFI 95 06 01 - Closed based on the completion.of the investigation report.
7 IFl 95 06 02 - Open pending completion of the software change and l functional testing.
{
No'dissentingviewsorobjectionswerestated. 1 Although proprietary materials were reviewed during the ins >ection, the l
proprietary nature of the documents has been deleted from t11s report.
E
l 1 .
APPENDIX A l -- ACRONYMS AND INITIALISHS l
- ADV Ammonium Diuranate AEC Active Engineering Control CSA
- (Nuclear) Criticality Safety Analysis DCF Direct Conversion Facility FM Emergency Control Center E/C Engineer / Constructor EH&S Environment Health & Safety EN Event Number FCR Facility Change Request FHO Fuel Manufacturing Operation GE General Electric IFI Inspector Follow up Item IR Inspection Report KGS. Kilogram '
JCP sointConversionProject NCS Nuclear Criticality Safety NE Nuclear Energy NEP Nuclear Energy Production l
NRC Nuclear Regulatory Commission !
NSR/R Nuclear Safety Requirement / Release OSHA Occupational Safety and Health Administration OA - Quality Assurance-REDCAP Reclaim Enrichment Degradation and Control Prevention 4
U0 Uranium Dioxide UNIi Uranyl Nitrate Hexahydrate URLS Uranium Recovery from Lagoon Sludge URU Uranium Recovery Unit 4
WT Waste Treatment
i i-
/ 'g NUCLEAR ME T Y COMMISSION YTNA, son NN e...* August 23, 1996 General Electric Company--
ATTN: Mr. C. P. Kipp General Manager GE Nuclear Energy Production P. O. Box 780 Wilmington, NC 28402
SUBJECT:
NRC INSPECTION REPORT NO. 70 1113/96 09
Dear Mr. Kipp:
This refers to the inspection conducted on July 2P. 26,1996, at the Wilmington facility. The purpose of the inspection was to oetermine whether activities authorized by the license were conducted safely and in accordance with NRC requirements. At the conclusion of the inspection, the findings were discussed with those members of your staff identified in the report.
Areas examined during the inspection are identified in the report. Within these areas, the inspection consisted of selective examinations of procedures and representative records, interviews with personnel, and observation of activities in progress.
Within the scope of the inspection, violations or deviations were not identified.
In accordance with 10 CFR 2.790, of the NRC's " Rules of Practice." a copy of this letter will be placed in the NRC Public Document Room.
Should you have any questions concerning this letter, please contact us.
Sincerely.
DY Edward J. McAlpine, hief Fuel Facilities Branch Division of Nuclear Materials Safety Docket No. 70 1113 License No. SNM 1097
Enclosure:
NRC Inspection Report ec w/ enc 1: (Ste page 2) wy w 3 u
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y U.S. NUCLEAR REGULATORY COMMISSION REGION II Report No.: 70 1113/96 09 i
License No.: SNH 1097 l
! Licensee: General Electric Company Wilmington, NC 2B402 l Facility Name: Nuclear Energy Production Inspection Conducted: July 22 26, 1996 Inspectors: D. Kasnicki, Fuel Facility Inspector, Region 11 L. Lessler Senior Operations Research Analyst, NHSS, Fuel Cycle Operations Branch Accompanying Personnel: D. Outlaw. Senior Scientist, SAIC (NRC Contractor)
Approved by: YkM*
E. J. McAlpine, Chief \
u 8!23 /9la Tale Signed-Fuel Facilities Branch Division of Nuclear Materials Safety Inspection Sgnary Areas insoected:
NRC performed a routine, announced criticality safety inspection of General Electric Fuel Fabrication Facility. Wilmington, NC on July 22 26, 1996. This inspection was conducted using staff from NRC Headquarters. Region II, and a contractor. The inspection focussed on the criticality safety program at General Electric and selected areas of the plant that either had recent, updated Criticality Safety Analyses (CSAs), had undergone recent modifications, or otherwise were of special criticality safety interest. The inspection also followed up on the licensee's final corrective actions to address a fire in their incinerator building which occurred on March 27, 1996
[ Inspector Followup Item (IFI) 70 1113/96 203 01],
Major programmatic portions of the criticality safety program reviewed included:
e Management and Administrative Practices for Nuclear Criticality Safety Recent Nuclear Criticality Safety Evaluations Enclosure
e.
4:
2 e - Nuclear-Criticality Safety Configuration Control Program
- e - Maintenance Program for Active Criticality Safety Controls Nuclear Criticality Safety Training Program The inspection was conducted usina the objectives of draft Headquarters Inspection Procedure 88015, as well as the existing inspection proceduras.
68015 and 88020.
, Results:
e For the systems tl:e inspectors selected, the licensee's implementation of their overall criticality safety program was good, i L e L The quality and technical adequacy of the Criticality Safety Analyses- l (CSAs) was acceptable. The analyses were structured and written so that they could be audited by an external auditor or independent reviewer.
e The analyses identified workable controls. Spot checks indicated they <
had been implEdnted through Nuclear Safety Release / Requirements (NSR/Rs) and operational procedures and then functionally tested to ensure that they worked as expected.-
e 1
Inspection of the areas'of the plant covered by these CSAs indicated
- that the required criticality controls were being adequately implemented.-
L e -The overall criticality safety configuration management / control system was acceptable, o
The inspectors reviewed portions of the scheduled maintenance and calibration program and no concerns were' identified, o
The ins W tors reviewed portions of the training program and found that the progrom ensures that managers. - supervisors.- and staff were trained in the nature of and respon!.ioility for nuclear criticality safety. The general nuclear criticality safety training curriculum was also reviewed and raised no concerns.- One concern was identified regarding the training program for " Qualified Reviewers". This program was used to train staff to judge the need for criticality safety function involvement in reviewing new or modified operating procedures. The licensee's plans to address this concern will be tracked as an Inspector Followup' Item (IFI 70 1113/96 09 01: Section 3).
e -The licensee's final corrective actions to address the fire which had occurred in tneir incinerator building on March 27. 1996 were adequate and IFI 70 1113/96 203 01 is closed.
Attachment:
Key Licensee Personnel-Contacted. Exit Meeting. and List of Opened and Closed Items
e:
.Q-3 DETAILS 1.
MANAGEMENT AND ADMINISTRATIVE PRACTICES FOR NUCLEAR CRITICALITY SAF
! 1.1 Discussion The inspectors performed a review of management and administrative practices for nuclear criticality safety to determine if they conformed to the requirements of the License and provided for adequate nuclear criticality safety. This inspection was performed based on observation of plant practicts, document review cnd discussions with criticality safety management, and operations st aff. Topics inspected included:
1.1.1 Plant Poliev for Nuclear Criticality Safety
. Discussions with plant staff indicated that the plant did not have an explicit, written policy that described each employee's authority and responsibilit for nuclear criticality safety.- - However, the License does state that "t e GE Wilmington )olicy is to maintain a safe work place for its employees..."(Section 2.1), T1e organizational authorities and ,
responsibilities for criticality safety for the management organizations were clearly defined in the License and in GE Wilmington Practices and Procedures
- (P/P) 1010 (Configuration Hanagement), 40 04 (Nuclear Safety Design Criteria), and 40 17 (Nuclear Safety Training). Eri'oyees were being taught in the annual training that nuclear safety is the res>onsibility of each emp oyee. The criticality safety manager indicated t1at he was aware of the
-lac of an explicitly written plant policy on criticality safety, has been working on a draft and was committed to-get it issued.-
1.1.2 Plant Manaoer's Resoonsibility for Nuclear Criticality Safety Discussions with plant staff and review of documentation indicated that the General Manager. GE Nuclear Energy Production (i.e., Plant Manager), had emoowered every'emoloyee and plant department with the authority and re'sponsibility to ensure nuclear criticality safety. The General Manager was aware of and involved in criticality safety at the plant and reviewed the findings of Wilmington Safety Review Committee and.the external audits of the criticality safety program. One stated purpose of the Wilmington Safety Review Committee was to provide the General' Hanager "with an overview of the safety of nuclear c ori rations" (P/P 40 01). He also held his staff accountable for icality safety.
1.1.3 Operations Manaoers' Resoonsibilities for Nuclear Criticality Safety ,
Observatien of alant operations, discussions with staff, and review of plant Procedures and 3ractices indicated that the Area Managers accept and carry out
'their NCS responsibilities, including approval of all operating proceduns.
I
y:
4 1.1.4 Ooerations Suoervisors' Resoonsibilities for Nuclear Criticality Safety I Observation of alant operations, discussions with staff, and review of plant Procedures and )ractices indicated that the Area Coordinators. Product Coordinators,-and Shift Operations Leaders understand, accept, and carry out i
their NCS responsibilities, These o development of operating procedures,perations giving on su>ervisors were and tle job training, involved in requiring conformance to NCS specifications, 1.1.5 Nuclear Criticality Safety Function's Resoonsibilities Discussions held with nuclear criticality safety, management, and operations staff, as well as observation of operations and review of plant documentation, indicated that management has established an effective nuclear criticality safety function independent of operations management. This function was effectively proceduralized in GE Wilmington Practices and Procedures documents, including 1010 (Configuration Management), 40 04 (Nuclear Safety Design Criteria), and 4017 (Nuclear Safety Training), and in Nuclear Safety Instructions, including E 1.0 (Nuclear Safety Review Records). E 2.0 (Internal Nuclear Safety Audits). E 3.0 (Nuclear Safety Review Requests), and E 4.0 (Criticality Safety Analysis Hethods & Verification). .
Discussions and review of documentation indicated that the nuclear criticality staff has been involved in and provided technical guidance on all plant changes to SNM operations.
Discussions with nuclear criticality safety staff indicated that they have maintained familiarity with current safety standards, guides, and codes. They have participated in and presented papers at national meetings in their field.
Overall. the inspectors found that the nuclear criticality safety function had been effectively implemented at the plant, met the License requirements, and functioned in accordance with the requirements and expectations of the national consensus standards.
1.1.6 SuDDort Function Manaaement Resoonsibilities for Nuclear Criticality Safety Discussions with staff and review of minutes of the Wilmington Safety Review
-Committee indicated that other support functions, including radiation protection, were aware of and involved in support of criticality safety, All support functions participated in the safety committee meetings.
1.1.7 Hanaaement Involvement in Providina Written Administrative-and Ooeratino Procedures Review of plant procedures and discussions with staff indicated that management has established effective administrative procedures that define the interface between operations, nuclear criticality safety, and other operations support functions.
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Procedures were in place (P/P 10 10, Configuration Hanagement Program) to ensure-that the nuclear criticality safety function would be consulted before a chango or new activity that could affect nuclear criticality safety was initiated. All procedure changes would be screened by a " Qualified Reviewer" to was needed.
determine if further evaluation by the nuclear criticality safety function (Refer to Section 3, Configuration Control Training, para. 2) 1.1.B Nuclear Criticality Safety Trainino Discussions with staff and review of plant documentation, including P/P 4017 e
(Nuclear Safety Training) and criticality safety related training materials, indicated that nuclear criticality safety training was included as a part of annual nuclear safety training required of all fissile material workers, their supervisors and managers, and support function engineers and managers.
1.1.9 thtclear Criticality Safety Advisory Committee Discussions with staff and review of plant documentation, including P/P 40 01 (Wilmington Safety Review Comittee) and minutes of recent safety committee meetings, indicated that management had established an effective, functioning safety comittee. The responsibilities of the comittee included criticality safety. Membership on the comittee was required by procedure to include the program manager for criticality safety as well as various operations and support function managers. Action items and recomendations have been documented and tracked in the committee minutes Based on the discussions and review effectively. of meeting minutes, the committee appeared to be functioning 1.2 Conclusion Overall, the management and administrative practices for nuclear criticality safety in these areas conformed with the standard requirements and practices identified in national consensus standards.
- 2. NUCLEAR CRITICALITY SAFETY EVALUATIONS 2,1 Discussion The detailed criticality safety evaluation documentation onkages for several selected plant changes and systems of special criticality safety interest were reviewed for technical adequacy. Each of these systems was selected by the inspectors because they were important from a criticality safety perspective and because they required detailed technical analysis and review to ensure ,
that the processes could be operated safely. The processes reviewed were:
- 2.1.1 Dry Conversion Process (DCP) t Extensive discussions were-held with the criticality and engineering staff on the status and approaches to criticality analyses for the new dry conversion processes. The analyses have progressed significantly from the parametric
4 i
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studies No. 70 1113/96 reviewed during the previous inspection (Inspection Report 203), i
- however, no new analyses had been finalized and therefore only draft DCP analyses were reviewed during this inspection.
! Initial draft criticality safety analyses (CSAs) for the DCP vaporization I coldtrap and the vaporization autoclave had been completed and were undergoing internal peer review during this inspection. Selected criticality safety controls of these two systems were reviewed by inspection of the draft CSAs, inspection of the systems with a criticality safety engineer, and discussion of the criticality safety issues with the criticality safety staff. The inspectors found no significant issues.
Discussions with the criticality safety staff indicated that CSAs for the other portions of the DCP, particularly those relying principally on-moderation control, were progressing well but were still incomplete. Most of the systems had been modeled and initial runs had been made for a number of accident conditions. Based on these initial calculations, the licensee expected that several basic modeling assumptions for process conditions and parameters would be changed in future runs.
Discussions between criticality safety and engineering staff indicated that additional ex were needed. peritaental data on powder characteristics during upset conditions Licensee representatives stated that they planned to generate the needed data. The licensee planned to use this data to verify that the bounding assumptions of powder densities under upset conditions used in the criticality analyses are conservative.
A rough draft CSA for the DCP cooling hopper was also reviewed with the author and no significant issues were identified.
'2.1.2 Bapheement of the "7B" dumo with a "JNF style" dumo The criticality safety documents supporting realacement of the existing powder dump station from the "7B" rotary press with t1e "JNF style" powder dump station from the "2A" rotary press were reviewed. This change would allow the processing of higher enrichment U 235 powder (5 percent) through the affected process line. The criticality safety analysis sup)orting these modifications was completed in June 1996 and the modifications lad been made. This modified system was inspected with the critice'ity safety engineer and criticality safety issues associated with the-new equi) ment were discussed.
Final approval to operate the line at the higher enricament had not been granted at the time of the inspection pending completion of final pre-operational reviews.
2.1.3 Reolacement of the 421 drum stacker with a drum conveyor system The criticality safety documents supporting replacement of the 421 drum stacker with a conveyor system for drum storage were also reviewed. This modification involved the removal of equipment from the drum stacker and installation of arrays of conveyors for the storage of drums in planar arrays. The criticality safety analysis supporting this modification was completed in April 1996. The inspection of this system with the criticality
v 7
safety engineer indicated that this modification had not been completed but limited drum storage was occurring on the conveyors that had been installed under interim approvals from the criticality safety group, Criticality safety issues associated with the new equipment were also discussed, 2.1.4 Criticality safety and the Uranium Recovery from Lacoon Sludoe (URLS) f acility Criticality safety files and supporting documentation from the existing URLS facility were also reviewed during this inspection in support of the more detailed field inspection of the implementation of the identified criticality safety controls. The inspectors toured the URLS facility interviewed URLS operations personnel, and verified the implementation of a sampling of Nuclear l Safety Release / Requirements determined by the CSA for URLS. Operations l
procedures were discussed, records of operator performance of administrative i
controls and records of the calibration and functional testing of active engineering controls were examined. The control room operator interviewed was very knowledgeable, both of the theoretical and practical aspects of the facility. Implementation of the inspected NSR/Rs at the URLS facility was adequate.
2.2 Conclusion Each of-the analyses listed above had been prepared in accordance with the procedural- requirenients and, if final, had been indeandently reviewed and verified by a senior criticality safety engineer. T1e inspectors found the completed analyses to be technically adequate and no significant issues were raised. In those areas inspected, the implementation of criticality safety requirements determined by the CSAs was adequate,
- 3. NUCLEAR CRITICALITY SAFETY CONFIGURATION CONTROL PROGRAM 3.1 Discussion The inspectors reviewed the NCS aspects of the configuration control program to verify conformance with License conditions and determine the adequacy with whic5 it provided for nuclear criticality safety. This inspection was 3 performed based on observation of plant practices, document review, and discussions with criticality safety, configuration management, engineering, program management, and operations staff. Topics inspected included:
3.1.1 Confiauration Manaaement Proaram The licensee has established a configuration management program that effectively recognizes changes in precedures, maintenance, calibration, facility changes, and preventive maintenance activities with criticality significance and ensures that (1) all such changes are properly documented; and (2) criticality-safety approvals are provided and criticality safety
-analyses are done where necessary. This program was clearly documented in Practices and Procedures 1010 (Configuration Management Program Fuel-.
Manufacturing). That precedure clearly specified responsibilities for concurring with and approving controlled changes.
8 The configuration management program was designed to ensure that controlled documentation would be updated and when necessary, a criticality safety evaluation performed for changes to processes, equipment, or facilities, Included in the system were facility changes other than like kind or identical, procedure changes, drawing changes, and technical report changes.
A system was also in place to control the revision and maintenance of these documents.
3.1.2 Confiauration Control of Nuclear Criticality Safety Evaluations The inspect'rs reviewed the configuration control systems for criticality safety evaluations by retrieving the documentation for several recent facility changes. They were able to locate controlled copies of each of the evaluations and related documents, including Nuclear Safety Release /
Requirements (NSR/Rs), for all changes.
3.1.3 Confiauration Control Chance Procedure The inspectors reviewed aspects of the configuration management program
' lated to the process of making changes in that program and found the process u be well documented in P/P 10 9 (03eration Procedures FHO) and P/P 1010 (Configuration Management Program uel Manufacturing). Those procedures also established a system for retention of records reviewed as a 3 art of the change control process. Spot checks by the inspectors verified tlat the configuration control program was being followed.
I I
3.1.4 Confiouration Control Trainina The inspectors also reviewed the configuration control training and found that this training was included in annual training. The con +ent of training in the configuration management and change control process appeared to be adequate.
The training of those perforning the screening on whether changes in procedures and other documents required additional review by the criticality safety group was also reviewed. The training program and records for
" Qualified Reviewers" who, per P/P 10 09.3.1, are trained to judge the need for additional review by the Environment, Safety & Health department (ES&H) which includes criticality safety, of procedure revisions, temporary changes to operating procedures, and special operating instructions, were specifically reviewed. The training to be a " Qualified Reviewer" consisted of a day's training, including criticality safety, radiological safety, industrial safety, environmental protection, material control and accounting, and quality assurance modules, A written exam was given after each module of the training.
Operating procedures were in the process of being rewritten to incorporate specific criticality safety controls into the procedure itself, rather than have them in separate NSR/Rs and the importance of having persons knowledgeable in criticality safety ]erforming these initial screens of changes to operating procedures was 3ecoming even more important than in the past when criticality safety requirements were in a separate document, the NSR/Rs, While specific criticality steps were being indicated in the new
s' 9
procedure format changes in other procedure steps could also have criticality safety importance that might not be obvious to an untrained reviewer.
j Review of the latest training records and exams indicated that on January 10, 1996, 15 candidates took the " Qualified Reviewer" training, The next day, on January 11, all 15 were authorized to serve as " Qualified Reviewers" despite the following: (1) The exams on the criticality safety module taken were apparently not graded, and (2) at least one of those authorized demonstrated a poor knowledge of criticality safety on the exam, leaving several answer sp:t.es blank, and providing incomplete or wrong answers in a number of other instances.
i The examination process for this training had written questions but did not require written grading and at least one individual listed as a " Qualified Reviewer" was judged to have performed unsatisfactorily by both the inspectors and the GE criticality safety manager. Imediate investigation indicated that this individual had not performed any duties as a " Qualified Reviewer" and he was removed from the list of those deemed qualified, Also, the investigation indicated that this individual had taken the training for informational purposes only, and-it was not intended that he become a qualified reviewer, but the administrative system for this training did not make provisions for this type of attendee at the trainirig course.
During the exit meeting, the manager for GE NE Environment, Health & Safety stated that he:
(1) will make training materials available to attendees for study prior to future " Qualified Reviewer" courses:
(2) will limit the formal qualification training to individuals-who meet minimum qualifications to be " Qualified Reviewers":
(3) will require future exams to be graded, with the criticality safety module graded by a criticality safety engineer:
(4) will allow auditing of-the course for staff not needing to be -
qualified; and (5) will also a) ply these changes made to the criticality safety-module to tie other " Qualified Reviewer" training modules which address the other safety disciplines.
.The results of these actions will be tracked as an Inspector Followup Item:
IFI 70 1113/96 09 01: Follow up on changes to' training program for " Qualified Reviewers" of procedure changes.
3.2 Conclusion Overall, the criticality safety portions of configuration management program aapeared adequate with the exception of the concern over the administration of t1e qualification process for " Qualified Reviewers".
t 10 4.
MAINTENANCE PROGRAM FOR ACTIVE CRITICALITY SAFET( CONTROLS The inspectors-reviewed portions of the scheduled maintenance and calibration program to verify (1) that maintenance schedules were established, adhered to, and were adequate for active criticality safety control systems: (2) that calibration schedules were established for all NCS control systems: and (3) that these schedules were corsistent with license requirements.
' -Discussions-with the maintenance program and criticality safety staff and i review of selected documentation raised no concerns with the maintenance and calibration program. That program was designed to assure that functional i
tests, calibrations, surveillance, etc. required to support the active engineered controls identified in the criticality safety analyses would be
' wrformed on time with appropriate approvals. There was no significant aacklog of criticality related maintenance tasks to be performed.
4.2 Conclusion-The licensee's implementation of their scheduled maintenance program for criticality safety controls appeared adequate.
5,-
GENERAL NUCLEAR CRITICALITY SAFETY TRAINING / RETRAINING PROGRAM The inspectors reviewed portions of-the training program to verify that the program ensures that managers, supervisors, and staff are trained in the
- nature of and responsibility for nuclear criticality safety, and that the training curriculum is adequate. Review of training materials, including a 40 minute video on criticality and radiation safety presented to general stai'f, and discussion of the training program with training and nuclear criticality safety staff, indicated that the program met general expectations of the inspectors. The program emphasized the individual's responsibility for safety. A test was given at the end of the training to confirm that the criticality safety principals were understood. Additional training, including on the job training had been provided to operations staff en an as needed basis,. The general nuclear criticality safety training curricula was also reviewed by the inspectors ar.d raised no concerns.
5.2 Conclusion The content and implementation of the licensee's general nuclear criticality safety training / retraining program appeared adequate.
- 6. HARCH 27, 1996 FIRE AT INCINERATOR BUILDING: FOLLOWUP ON LICENSEE'S FINAL CORRECTIVE ACTIONS A small fire occurred in the licensee's incinerator building when equipment malfunctioned, causing a box of trash to not be fully inserted into the incinerator's burn chambar from its load chamber. The box caught fire while still partially in the load chamber. Operators detected the problem, opened the load chamber door, and extinguished the burning box with a fire extinguisher. However, the force of the discharge from the fire extinguisher blew a few pieces of burning debris onto a second box which was staged to be burned next, burning a few small- holes into the polyethylene wrapping on the
i' ,
i 11 box, Further details of this entire incident are documented in Inspection Report No. 70 1113/96 203. At the time of that inspection, the licensee had 1 performed a thorough investigation of the incident and documented a list of proposed corrective actions, however, from this list, they had not yet had time to choose their final corrective actions and implement them.
During this inspection the inspector determined, through discussions with a cognizant licensee representative, a review of related documentation, and a tour of the incinerator building, that final corrective actions had been chosen and implemented. The licensee's corrective actions taken included the following:
e- A fire suppression system was installed in the incinerator's load
' chamber. If a box were similarly misloaded again, this would eliminate the need to open the load chamber door and manually extinguish the fire, e The queuing of the next box to be burned was changed so that it no longer sits just below the load chamber door. If somehow, in spite of the load chamber's fire suppression system, it became necessary to open the load chamber door, the next box to be burned would be less vulnerable, o The box in the load chamber gets pushed into the burn chamber by a
' hydraulic ram, and a mechanical problem with the position sensing of this ram had caused the mislosding of the box in the incident.
Th^ position sensing device was changed from mechanical to an it... ared reflection / detection device, precluding the same type of -
mechanical malfunction.
Also, the licensee, unrelated to their incident but rather in response to their assessment associated with an incinerator fire at another licensed facility, increased the capacity of the emergency cooling water to their incinerator.
6.2 Conclusion The above described corrective actions appeared to adequately correct the causes of the incident, and IFI 70 1113/96 203401 is closed.
ATTACHMENT
- 1. KEY LICENSEE PERSONNEL CONTACTED i
- H. Chilton, Manager, Joint Conversion Project
- S. Dale, Criticality Safety Engineer
- D. Dowker, Team Leader, Fuel Support
- R. Foleck, Sr. Licensing Specialist
- J. Kline, Manager, Chemical Product Line !
- G, Luciano Environmental Operations Engineer
- D. McCaughey, Manager Fuel and ChemHet Lab Quality
- L. Paulson, Manager, Criticality Safety Engineering,
! Acting Manager, Nuclear Safety
- L. Quintana, Manager, Fabrication Product Line
- R, Reda, Manager, Fuel and Facility Licensing
- G. Robinson, Criticality Safety Engineer ,
- W. Sependa, Manager, GE NE EH&S
- H. Shea, Criticality Safety Engineer
- G. Smith, Team Leader FHO Haintenance Support
- S. Smith, Radiation Safety Monitor
- P. Vescovi, Criticality Safety Engineer -
- C, Williams. Team Leader, URLS/'wT
- T. Winslow, Manager, Material Control & Accountability
- Attended exit meeting Other licensee personnel contacted included engineers, operators, and technicians.
- 2. EXIT HEETING The inspection scope and findings were summarized during a meeting on July 26, 1996 with the licensee personnel indicated above. The inspectors discussed the likely informational content of the inspection report with regard to documents and processes reviewed during the inspection. Although proprietary information was reviewed during this inspection, this information was not included in this report. Within the scope of this inspection, no violations or deviations were identified. However, one Inspector Followup Item, IFI 70 1113/96 09 01 was-opened for further tracking.
- 3. LIST OF OPENED AND CLOSED ITEMS Item Number Statuji Descriotion and Reference 70 1113/96 203 01 Closed- IFI Follow up on final corrective actions for March 27, 1996 incinerator fire. (Section 6) 70 1113/96 09 01 Open IFI - Follow up on changes to training program for " Qualified Reviewers" of procedure changes.
(Section 3)
e-
/p* E80g% UNITED STATES NUCLEAR REGULATORY COMMISSION
- S 101 MARIETTA STRE .W SO4TE 2900 j ATLANTA, GEORGIA 30323@90 A, ...../
August 27, 1996 General Electric Company-ATTN: Mr. C, P Kiap, General Manager GE Nuclear Energy Production P, 0, Box 780 Wilmington, NC 28402 SUB 'ECT: NRC INSPECTION REPORT N0, 70 1113/96 08 "
Dear Mr. Kipp:
This refers to the inspection conducted on July 15 20, and July 29 Agust 2, 1996, at the Wilmington facility. The purpose of the inspection was to determine whether activities authorized by the license were conducted safely and in accordance with NRC requirements. At the conclusion of the inspection, the findings were discussed with those members of your staff identified in the report, Areas examined during the inspection are identified in t5e report, Within these areas, the inspection consisted of selective examinations of procedures and representative records, interviews with personnel, and observation of activities in progress, Within the scope of the inspection, violations or deviations were not identified.
In accordance with 10 CFR 2,790 of the NRC's " Rules of Practice," a copy of this letter and its enclosure will be placed in the NRC Public Document Room, Should you have any questions concerning this letter, please contact us.
Sincerely, 7tfM lll Edward J. McAlpine. Chief
/.cd Fuel Facilities Branch Division of Nuclear Materials Safety Docket No. 70 1113 License No, SNH 1097
Enclosure:
NRC Inspection Report cc w/ encl: (See Page 2) g,
I
-GE- 2 cc w/ encl:
Dr. Ralph Reda, Manager Fuels and Facility Licensing ,
General Electric Company P, O. Box 780, Mail Code J26 Wilmington, NC 28402 Dayne H. Brown Director Division of Radiation Protection N. C. Department of Environment, Health & Natural Resources i i P, 0. Box 27687 l
Raleigh, NC 27611 7687 i
4
U. S. NUCLEAR REGULATORY COMMISSION REGION II Docket No.: 70 1113 License No.- SNH 1097 i
Report No.: 70 1113/96 08 Licensee: General Electric Company Facility: Nuclear Energy Production Location: Wilmington, North Carclina Dates: July 15 20, 1996 .
July 29 August 2, 1996 Inspector: G. L. Troup Sr. Fuel Fecility Inspector Approved by: E. J. McAlpine, Chief Fuel Facilities Branch Division of Nuclear Materials Safety Enclosure W l- I'
EXECUTIVE
SUMMARY
l General Electric , Nuclear Energy Production NRC Inspection Report 70 1113/96 08 This-inspection included aspects of licensee operations, management controls, l
maintenance, fire protection, event review, and review of previous inspection findings.
Ooerations e Preparations and response to potential impact from Hurricane Bertha were(section the facility well coordinated 1.1) and resulted in the safe shutdown of e Criticality monitor alarms due to lighting storms have been a recurring problem but re)lacement of parts in the grounding system are planned to correct tie problem (section 1.2) e The investigation into the failure of the slug press downchute failure was thorough and identified a number of corrective actions to reduce future failures (section 1.3) .
Manaaement Controls e External audits did not identify any major programmatic issues (section 3.1) e A notice was issued to all employees reminding them of their responsibility to report deviations and non compliances in accordance with 10 CFR 21 (section 3.1)
-Maintenance
- e. Up grades to process monitor software were completed and satisfactorily tested (section 4.1.1) e Functional tests of AECs were satisfactorily conducted (section 4.1.2)
Fire Protection o Smoke and fire detectors were satisfactorily inspected and tested by a contractor but the test report needs to be improved (section-5.1.1)
- Fire equipment and systems were satisfactorily inspected and tested (section 5.1.2) e Signs identifying locations of fire alarm pull boxes will be relocated to identify the new pull boxes (section 5.1.4)
2 DSE e TheProjectIntegrationPlanhasbeenapprovedandwillbe reflected in the Integration Schedule (section 6.1.1)
The test program for the process control system is being developed (section 6.1.2)
Resultst No violations or deviations were identified. Two IFIs were closed. l
3 Rgoort Details
- 1. OPERATIONS ISSUES (88020)
M Hurricane Bertha In response to the potential for Hurricane Bertha making landfall in the Wilmington area, the licensee began shutting down operations on July 11.
The ECC was staffed for the monitoring of conditions. All fuel production activities. -as well as other activities on the site, were suspended at 3:00 pm on July 11. Only volunteers remained on site to operate selected waste systems, tonitor equipment which was still being cooled down, and monitor the facilities.
At 11:00 am on July 12, the licensee declared an ALERT in accordance with emergency procedures when sustained winds exceeded specific levels.
The NRC Operations Center and Region II were notified. After the hurricane moved further north without making landfall and winds -
diminished,- the licensee declared that the ALERT was terminated at 8:00 pm and notified the NRC Operations Center accordingly. Manning of the ECC was shutdown at 11:00 pm but personnel rem n ed on site to monitor the facilities.
' Some minor damage occurred to buildings (such as bent siding) and trees were blown down around the site. The loss of an off site power transformer resulted in the loss of power to Waste Treatment and ground water pumps. Power was rerouted from another transformer by plant personnel on July 13 and equipment returned to operation.
Fuel production operations resumed on third shift on July 14, consistent with production schedules. The inspector observed portions of the start up of the chemical conversion lines on July 15.
M Criticality Monitors On June 25 two criticality monitors alarmed during a thunderstorm. As-these detectors were in different plant areas, the plant evacuation alarm did not activate. Personnel manned the ECC until the situation could be evaluated and determined to be false alarms. Coverage of the site areas were monitored by the redundant monitors which did not alarm.
On July 2, two criticality monitors again alarmed during a thunderstorm.
These were in outside areas and again were determined to be false alarms.
On July 16. during a thunderstorm, the inspector observed the read out panel for the criticality monitors. Four monitor sets in outside areas were showing fluctuations in readings but did not reach the alarm setpoint.
W 4
A review of the criticality monitoring system by licensee personnel and discussions with the equi > ment supplier indicated that certain parts (the " power sifter".-whici is a surge suppressor in the grounding system) may be beyond the recommended lifetime. New power sifters were ordered and will be installed during the up coming plant shutdown. The power sifters are part of the grounding system and should not effect the detector unit calibration.
La Powder Soill On June 18, 1996, the licensee made a report under the requirements of NRC Bulletin 91 01 (NRC EN 30638) because on June 17.-a hose fitting on the powder transfer hose had failed and approximately 35 kgs. of U0 powder had been spilled into the enclosure of the "A" hie slug pres,s.
The U0 had an enrichment of 4.95 percent. This event was reported becaus,e the two nuclear safety controls for the slug press are moderation control and geometry control. The spill of the powder resulted in the loss of geometry control. The operator noticed the spill and immediately shut down operations and started removal of the powder into favorable geometry containers.
The transfer hose (downchute) is connected to the mill baghouse discharge transition piece at the top and to the )owder shuttle at the bottom. - The upper flange had separated frem the lose, leaving the flange bolting ring attached to the transition piece and the downchute and part of the flange in the enclosure.
An unusual Incident Report (UIR No. ChPL 9632) was initiated on June 17.
The initial investigation identified that the two flanges on the hose had two different designs for the connection of the flanges to the hose nipple fittings. The joint that failed had apparently been welded and then most of the weld machined off. providing a very weak joint. It was also documented that operators had been " manually assisting" (e.g.,
pounding on the downchute with a mallet to assist the flow of aowder which sometimes'" sticks" in the downchute). This " assisting" 1ad been in progress when the joint-failed. The operation of the slug press shuttle produces vibration in the upper flanged joint. The vibration and pounding apparently caused'the failure of the welded joint. The UIR contained seven actions to be taken. The Manager. Chemical Product Line directed that a higher level critique be conducted and assigned this to the Team Leader. Powder Preparation and Packaging-The critique report concluded that the primary causes of the failure were (1) incomplete and poor quality welds in the flange construction, and (2) operators pounding on the downchute, coupled with the vibration.
weakened the welds. A-review of the maintenance records revealed that these downchutes _have to be replaced at an interval of about six months due to breaking,of the braided jacket and/or breaks in the liner.
=
The downchute on the "B" slug press was examined and found-to be of a different design than that on the "A" press. Both downchutes were replaced with a new part which has a-single piece flange. To assure a
5 that this design of flange is always used, Shop & Purchasing set up a
" single source" vendor and requires inspection for each new downchute when delivered. Operators were instructed that beating directly on the downchute was not an acceptable practice.
Additional actions being evaluated are establishing a mass control limit for the system so that a powder spill does not result in only one l nuclear criticality safety control remaining, This control will be i evaluated for any adverse impact on the production rate. Additionally, other types of connections are being investigated.
The inspector determined that the amount of powder spilled as less than a minimum critical mass under conditions of optimum moderation and geometry, neither of which existed during the spill. For the actual '
moderation condition, the mass was about seven percent of the minimum critical mass for a hemisphere with optimum reflection.
- 2. ORGANIZATION CHANGES (88005)
On June 24, 1996, the former Manager Nuclear Safety was apaointed Team Leader, Chemical Conversion. This team is responsible for 'J0 powder production. An-Acting Manager, Nuclear Safety was designated,pendirg the selection of a permanent replacement. The inspector was informed on August 1 that the position would be posted and employees could apply.
Final selection of the new manager is pending.
- 3. MANAGEMENT CONTROLS (88005)
L1 Discussion The inspector noted that there was one change in the management-organization, as documented in paragraph 2.
The external audit of the nuclear criticality safety program, as required by Cha)ter 2, Section 2.8,3 of the license application, was conducted Decem3er 4 8, 1995. The audit team was approved by the General Manager NEP, as required. No Corrective Action Requirements were identified but the team made several observations. Corrective action on one observation is on going; the remainaer have been resolved.
The report concluded the program is basically sound and that operations are conducted safely from a nuclear criticality safety standpoint.
The licensee's insurance carrier conducted an-ins)ection on February 29, 1996, to assess the insurance risk associated wit 1 activities presently being performed. As a result of this_ inspection, no new recommendations were made.
The WSRC held the two routine quarterly meetings and one special meeting in 1996. Items reviewed during the routine meetings included
.significant UIRs,- PNCs and classified incidents and = violations.
Additionally, the WSRC reviewed such topics as the ISA program, which is being conducted for the DCF and is proposed for the rest of existing operations.
6 On August 1. 1996, an " Update" notice was distributed on the site to remind GE NE employees and anyone conducting business at the Wilmington site, of their responsibility under 10 CFR 21 to report deviations and non. compliances in hardware or software provided to a customer that could craate a potential safety hazard or lead to exceeding a Technical S>ecification Safety Limit. Included with the " Update" was a copy of t1e 10 CFR 21 posting. This " Update" served as a reminder to licensee employees both at the NEP site and at reactor sites of their legal obligation to report deficiencies.
L2 Conclusion The inspector concluded that adequate management oversight of plant i
operations and problems is being maintained. External evaluations have not identified significant problems in safety programs.
l 4. MAINTENANCE (88025)
_4d Discussion 4.1.1 In:Line Uranium Monitors .
, As the result of a power supply problem and a software oroblem, the Fluoride Waste uranium monitor did not isolate the disc 1arge when power was lost and then restored to the system (see IR 70 1113/95 06). The last of the long term corrective actions was to install new software for the monitor system and perform the necessary functional tests.
CR 96.0295 and CR 96.0297 were approved to install the software in the URU and ADO computers, respectively. A previous CR. 95.0582 was prepared to install new software but off line testing revealed that the software was inadequate. The new version of the software was tested
' line under CR 96.0287. The inspector noted that CR 96.0119 was enced in the two CRs as performing the off line test but, on review, the inspector determined that this CR was for the
' ion of an assay meter in another portion of the plant. After e ' view, licensee rearesentatives acknowledged that CR 96.0287 w.,
.orrect document. T1e official co)ies of CR 96.0295 and 96.0297 were currected. The inspector verified tlat 96.0287 satisfactorily performed the off line test.
A faulty detector in the Fluoride Waste monitor was replaced in conjunction witn the up grade. A full calibration of the monitor was performed on July 30. The calibration involved a total of six uranium solutions, a water blank and the monitor internal sources. The data were reduced using linear regression methods to establish the monitor parameters, which were loaded into the computers.
The inspector observed acrtions of the calibration activities and discussed the ste)s wit 1 the 3ersonnel involved. The inspector determined that t1e licensee lad a written procedure for the energy calibration of the various detectors.
7 l 4.1.2 Functional Tests Following the energy calibration of the Fluoride Waste pipe detector,-
the licensee conducted the-functional test of the detector and associated valves on July 31, 1996. This-test was conducted in accordance with OP 1010.77 FTI F2, Several minor problems were noted during the conduct of the test and were properly resolved and approved.
Corrections in the FTI were noted, with revision to be completed by a specified date, l i
i During the same period, the hie skid uranium monitor was tested in accordance with OP 1010.77, FTI F1. This test took longer than anticipated because the prerequisite tests had not been performed before the hie system had been shutdown for the calibration of the piae detector. Licensee representatives stated that the FTI would 3e revised to clearly permit / require these tests to be performed before the system is shutdown.
The inspector observed various portions of the two functional tests and determined that test personnel were using approved procedures, observed the control room computer screens to confirm that actions were occurring as required, and that the tests were properly reviewed and approved.
The inspector also reviewed the record sheets for the daily checks of 1 the parameters for the V0G system. These are defined in OP 1102.08, i
Exhibit 1, and recorded each day. The inspector reviewed the record for four separate weeks and confirmed that all parameters were within specified limits.
4.1.3 Criticality Monitor Calibrations Procedure NSI 0.4.0 specifies the requirements for the annual energy calibration of the nuclear criticality monitoring system detectors. The inspector reviewed the records for the calibration performed November 28 29. 1995, and confirmed that all detectors performed within specified limits. Also, all portable units (SRM 100s) responded within specified limits.
The inspector also reviewed the monthly test records-for the criticality monitoring system for the period January May, 1996, and determined that the system responded within specified limits. While problems with-alarms were encountered during thunderstorms _(paragraph 1.2), the-
- problems were apparently with the grounding system and did not adversely affect the performance of the detectors.
4.2 Conclusions The ?nspector determined that recuired calibrations and functional tests were properly conducted in accorcance with approved procedures, results were satisfactory, and test results were properly reviewed.
8
- 5. FIRE PROTECTION (88055)
Ed Discussion 5.1.1 Smoke and Fire Detectors The inspector reviewed the records for the annual test and inspection of smoke and fire detectors installed in various ventilation ducts. These ;
tests were performed by a contractor on April 29, 1996. The test checked the operation of the detectors, remote monitoring of the detector alarms in the various control rooms, and confirmed that detectors responded to test smoke.
In addition to the tests, the contractor checked wiring. cleaned and tested detectors, and checked the interlock sequence. The inspector noted that the documentation provided by the contractor was not as detailed as in previous years. While the documentation identified all of the tests performed, it only listed a total number for the types of detectors. In previous years, the documentation had listed each detector and described any problems encountered. The cognizant licensee representative agreed that the former documentation provided better information concerning the status of the detectors 'and stated that, in the future, more detailed documentation would be requested, l
5.1.2 Eouioment Tests The inspector reviewed the test results fcr various tests conducted on fire system equipment, During the annual fire hydrant water flow test conducted on July 18, 1996, two hydrants were reported as "hard to open" or "could not open." Work orders were instituted on August 1 to repair these hydrants.
The inspector reviewed the results for the monthly water flow tests, which include the sounding of water flow alarms in the security office.
The test for June was not performed because of a change to a new alarm system was in progress. Remaining tests, including July. were satisfactorily performed although certain streams were not tested because the test fitting discharges into construction areas.
On July 31 the inspector observed the annual fire hose tests, which were performed by a contractor. Six hoses which were tested failed the test, either due to leaks in the hose or failure of the end fitting connection. New hoses were installed in place of the failed hoses.
Hoses tested were from non radiological controlled areas. Hoses from radiological controlled areas are routinely replaced at five year intervals.
The inspector reviewed the records during the period January July, 1996 and verified that weekly and monthly inspections of fire valve positions-were performed and the valves were properly aligned.
M j
9 5.1.3 Ooerations During tours of various plant areas, the inspector observed various factors affecting the fire arotection program. Fire extinguishers and fire hose stations were marced, were unobstructed and ecuipment inspections were current. Flammable liquids were storec in approved containers the liquid. and there were no accumulations of materials saturated with Flammable gas cylinders were capped when stored and oxidizers were stored in separate areas.
' In areas where construction activities were in progress, areas were clear of accumulations of combustibles.
5.1.4 Alarm Stations The licensee is installing a new central alarm system. Most of the new fire alarms are functional but some of the chemical alarms are not com)lete. The old fire pull boxes are still in place but are covered wit 1 paper stickers.
The inspector observed that in many areas of the facility, signs are installed stating " fire alarm" and are set to be visible from some distance, such as in long hallways. The inspector' observed that several of these signs are over the old aull boxes and the new box was installed
- in another location (across the 1allway, on the other side of a door, etc.). An individual trying to sound the fire alarm would have to search for the box because the marked box was out of service. In one location (outside of the welding development laboratory),'the sign was clearly visible in the hall but the pull box was out of service. No new alarm box is installed in that location. In another location, the old box had been reinstalled on top of the new box, making pulling the' alarm lever very difficult, i
licensee representatives stated that the signs would be reviewed and relocated as needed to provide proper icientification of the pull box.
5.2 Conclusions I The inspector determined that the licensee had implemented a satisfactory fire protection program and was operating the facility in a fire safe manner. ;
. 6. DRY CONVERSION FACILITY (88020, 88025)
$_d Discussion 6.1.1 OCF Inteoration The DCF will produce UO powder, which will then be transferred to existing work stationsin FM0/FM0X for-the production of fuel pellets.
Because the existing ADU lines will remain in operation during the DCF l
j
10
)
start up and 3roduct qualification, a plan for the installation of new facilities, clanges in equipment and moderation exclusion areas is required.-
The inspector discussed the plan for accomplishing the chan maintaining both existing and new nuclear safety criteria. geA over while plan was developed for the work and was approved on July 3,1996. A detailed schedule for accomplishing integration activities will be developed from the plan. The plan and schedule will be used by the NRC in the development of the inspection plan.
6.1.2 Process Control System ihe control room for the DCF will include a process control system for the operation of the systems. P test and approve a new process ntrol co/P systems.
120 15 details the steps required to Licensee l
reg esentatives discussed the test process with the~ inspector and stated i
that the test program would be developed and approved by October..
Testing will be performed as part of the start up program.
6.1.3 Construction Activities .
The first (upper) roof membrane had been installed (IR 70 1113/96 06).
Final installation of edge trim eave trim and drain gutter will be
- installed in conjunction with the siding. Repairs to the membrane have been identified and will be performed in accordance with the supplier's instructions.
Major components have been received and landed on the foundations.
Alignment of systems was observed by the inspector. Piping installation was in progress in various rooms.
Work is continuino of the performance of nuclear criticality safety evaluations for equipment and processes. Drafts of four evaluations are being reviewed by the NSE criticality staff. ISAs using the Haz0p methodology are being performed for the operations. ISA findings, conclusions and corrective actions will be reviewed by WSRC. These will be part of the NRC review program.
6.1.4 Staffino Selected managers, engineers and operators have completed training at the system supplier's facility. Knowledge gained during this training is being applied to writing operating procedures and FTIs.
Operator staffing for the DCF and replacement operators for FM0/FM0X is still on going. The licensee is resolving a number of personnel issues involving the: staffing plan.
11 62 Conclusions Work on the facility is progressing. Other activities necessary for the start up and operation of the facility are also progressing. The NRC will continue to monitor progress and establish the inspection program accordingly.
- 7. PREVIOUS INSPECTION FINDINGS (92701)
M (Closed) IFI 95 06 02 l
This IFI was opened as the result of an inspection into a reportable event concerning the loss of operability of the in line uraniun monitors in the Fluoride Waste system in the Chemical Area. The investigation report identified a number of long term corrective actions. Four of the five actions had previously been comaleted and documented (IR 70 1113/96 06). The last item was t1e up grading of the detector software and the performance of the necessary functional tests. As discussed in paragraph 4.1, the new software was installed, functional tests completed and-NSR/Rs revised accordingly. This IFI is closed.
~
M (Closed) IFI 95 09 01 By letter dated July 14, 1995, the licensee transmitted Revision 3 to the Physical Security Plan. Item 5 of the revision addressed a change in the time period that the licensee would use to conduct a trace investigation following the discovery of an ap)arent lost or unaccounted for shipment, and subsequent notification of tie NRC. Item 5 was referred to ONHSS for review, and was identified in IR 70 1113/95 09 as an IFI, By letter dated June 21 1996.- Region Il notified the licensee that, based on the ONHSS review. Item 5 was deemed acceptable for inclusion in the Plan. This IFI is closed based on the ONHSS review.
- 8. EXIT MEETING -
SUMMARY
The inspector presented the inspection results to members of the licera,ee management and staff at the conclusion of the inspection on August 2, 1996. The licensee acknowledged the findings presented.
Proprietary information was reviewed and discussed during the inspection. It has been excluded from the report details.
12 PARTIAL LIST OF PERSONS CONTACTED LICENSEE H. Chilton, Manager, Joint Conversion Project T. Flaherty, Start up Manager JCP R. Foleck, Sr. Licensing Specialist C. Kipp, General-Manager, GE NEP R. Keenan,-Hanager, Site Security & Emergency Preparedness J.-Kline, Manager, Chemical Product Line M. Lamb, Team Leader, Powder Production and Packaging S. Hurray, Team Leader, Chemical Conversion L. Paulson, Acting Manager, Nuclear Safety Engineering L. Quintana, Mana R. Reda, ManagerFuelsger, Fuel- Fabrication and Facility Product Line Licensing W.- Sependa Manager, GE NE EH&S G. Smith Team Leader, FH0 Maintenance Support C. Tarrer Team Leader, Configuration Management & ISA K. Theriault, Team Leader. URV 4
7 9
a
13 INSPECTION PROCEDURES USED IP 88005: Management Organization & Controls IP 88020: Operations Review IP 88025: Haintenance/ Surveillance Testing IP 88055: Fire Protection IP 92701: Follow up on Inspector Problems ITEMS OPENED, CLOSED, AND DISCUSSED Opened None i
Closed 70 1113/95 06 02 IFI Com)letion of long term actions following power pro)lems to Fluoride Waste syster,. monitor 70 1113/95 09 01 IFI Time period to trace apparent lost fuel shipments Discussed None
e 14 LIST OF ACRONYMS USED
-AEC Active Engineered Control ChPL Chemical Product Line CR Change Request DCF Dry Conversion Facility DCP Dry Conversion Project E/C Engineer / Constructor ECC Emergency Control Center EH&S Environmental Health & Safety EN Event Number FHO Fuel Manufacturing Operation FTI Functional Test Instruction GE NE General Electric Nuclear Energy
-GE NEP General Electric Nuclear Energy Production IFI Inspector Follow up Item IP Inspection Procedure
-IR Inspection Report ISA Integrated Safety Analysis KGS Kilograms NRC Nuclear Regulatory Commission NSI Nuclear Safety Instruction NSR/R Nuclear Safety Requirements / Release ONHSS Office of Nuclear Materials Safety and Safeguards OP Operating Procedure PNC Potential Non Compliance P/P Practices & Procedures SNH Special Nuclear Material SSR Software Service Request UIR Unusual Incident Report
. Uranium Dioxide UO[J UR Uranium Recovery Unit V0G Vessel 0ff Gas WSRC Wilmington Safety Review Council I
-i ONE LINERS FOR SEIrmMBERC226
- 1. Research ReaglDLdlatus
- a. Georgia Institute of Technolo y The licensee recently sent a letter to IIcadquarters infcnning the NRC that two SROs had left the facility and that a third will be leaving the end of next month. When that individual leaves, the licensee will have 1 RO and 3 SROs remair.ing on staff 11 Lily Statham, or e of the SROs who is currently serving as the Acting Reactor Supervisor, is available and able to be at the facility about 40% of the time. One of the .emaining SROs, Peter Newby, has been appointed as Assistant Reactor Supervisor.
No work is being performed at the facility other than some maintenance in the liot Cell and liigh Bay area. The licensee is preparing to paint those areas.
The licensee has indicated that no deelslon will be made about obtaining the new LEU fuel until after the ASLB has made their decision concerning the license renewal. .
GE is scheduled to be at the facility the week of September . a remove the Co-60 sources from the storage casks and remove the casks from the storage pool. The empty casks will then be shipped back to California,
- b. North Carolina State University The licensee continues to operate on a routine schedule of 3 to 4 days per week in support of research and experiments. No reportable events have been noted.
NRR is currently working on the license renewal and will be submitting a
' final" set of questions to the licensee by next week. With the proper responses to the questions, the license is expected to be renewed by NovemNr.
The reactor was in a shutdown mode on September 5 and 6 during the time when tropical storm Fran passed through the area,
- c. University of Florida The licensee completed their biennial fuel inspection and found no problems or anomalies with the fuel. Two other surveillances were also completed, including control blade worth, and the reactor is again fully operational. No Attachment 1
9 I
l One Liners 2 reportable events have been no'ed recently.
De reactor will not be operated as much as this next semester as it has been in the past because the former Reactor Supervisor left to return to graduate school on a full time basis. One of the remaining SROs has been named as the Acting Reactor Supervisor,
- d. University of Virginia The reactor is being operated on a routine schedu.a of 3 to 4 days per week now. The licensee continues to parform neutron radiography and to irradiate l seeds for medical use and topaz for the jewelry industry. No reportable events l
l have occurred during the past tuo weeks.
l l
The reactor pool leak, jlscovered after tuting the pool gate in December 1995, remains at approdmately 50 gallons pei day. The 50 gallon total in.ludes the amot.4 of liquid lost due to evaporation as well as to the actual leakage. The liccasce let bids for pool repairs and continues to review the schedule.
Recently a " Reactor Review Committee" was appointed by the University of Virginia to review the, operation of the research reactor, make an appraisal of the operation, and make a recommendat i on to the i.esident about the viability of the program and the feasibility of continued operation. The review is scheduled to be completed by the end of Oc.ober, ne reactor was shutdown during the tropical storm Fran; off site power has been lost, but the facility is secured.
- 2. General Electric _- Nuclear Encrev Production GE declared an ALERT at 7:20 pm on September 5 in anticipation of hurricane force winds. At the time of the declaration, custained winds were about 60 mph. The ALERT remained in effect until it was terminated at 1:33 am on September 6, following the passing of the eye of hurricane Fran.
Initial damage assessment at sunrise showed superficial damage to siding on the Airemft Engine, Emergency Control Center (ECC) and the Fuel Component Operation (FCO) buildings. The exhaust stack on the incinerator was damaged but the incinerator has been down for about two weeks. Numerous trees were blown down or snapped off. Some water was found in a hallway in Fuel Manufacturing Op: ration (FMO) building where the new Dry Conversion Process (DCP) building Attachment 1 '
One Liners 3 transition seal will be. No SNM was in this area. Some minor damage (over pack lids blown off) was identified to some waste boxes on pad 5 but the tntegdty of the protective wrapping does not ap;<ar to be breached. Inspection of the boxes and ~
radiological surveys of the area will be performed on September 6,1996.
The licensee did lose a secondary off site power source, which provides power to some cooling water well pumps and a warehouse. The licensee has the primary off.
site source. Prior to the storm, production and processing of uranium were shutdown. The licensee searched and cleared areas outside buildings of material that might become airbome. All routine operations at the site ended at 3 pm on September 5.
A response crew of about 50 people remained onsite covering the various buildings, j
I The gates were locked and security personnel brought to the Emergency Control Cenwr.
- 3. B&W - NNFD The forecast is for 510 inches of rain in the Lynchburg area as the result of FRAN.
The area is under a tornado watch. The licetisec is continuing to monitor the level of )
the James River. Some storm damage has occurred in the surrounding areas including downed trees and some isolated flooding. The road to the plant was closed due to flooding as of 11:30 am on September 6,1996. No loss of power to the site has been experienced. Flooding could occuring in the Waste Treatment facility.
Prior to the storm, the licensee moved material from this area to areas where it ceuld be affected by flood waters.
At 9:00 am on September 6, B&W started a plant shutdown and started sending people home. A crew is being selected to maintain site coverage ( ecurity, maintenance, etc.).
The Resident inspector cannot get to the site due to downed trees in his neighborhood.
4, Nuclear Fuel Services. Int As a result of inspection fmdings, the licensee confirmed to NRC on September 4.
1996, that actions would be taken to reassess the nuclear criticality safety of vault storage for the Rocky Flats material. This will also include retraining of operators, implementation of revised procedures if necessary, and assurance that vault storage is incorporated into the configuration control program. NRC will acknowledge the Attachment I er,
l
- Ce Lh.es 4 i
licensee's actions and confirm that analysis of movement of material into and out of the vault will also be analyzed in a letter to the licensee.
NRC has not received a response from NFS on the Notice of Violation and Proposed Imposition of Civil penalty in the amount of $12,500 that was issued on August u, 1996 for violations associated with the fire that occu Ted in the incinerator ductirg on April 2,1996.
- 5. Westinchousc At 0900 on September 5,1996, the licensee held an emergency response meeting and it was determined that all preparations had been completed for minimizing the effects of a possible strike from 11urdcane Fran. The preparations included securing all fuel shipping containers stored outside and moving both empty and full 55 gallon low level waste dmms to storage locations within the buildings. The licensee was using a lessons learned approach based on experience gained from IIurricane liugo in 1989.
On September 6,1996, the licensee indicated that it was not necessary to shutdown operations and there was no damage to the facility as a result of the storm. The licensee expected to returri to normal operations on September 6,1996.
The Region has followed up on the August 25,1996, small UF6 leak from the packing in a valve in conversion line 3. This particular valve design has been susceptible to 11F corrosion. During the last 2 3 years, the licensee has attempted re-engineering the valve design, using different metals, and during the last attempt, used a teilon coating on the valve stem. On August 25,1996, the teflon coating also failed. The licensee had checked the valves on the other conversion lines and found no apparent problems.
6.
Framatome Cocema Fuels m&W Fuel Comoans)
The plant is shutdown with no one on site in anticipation of flooding of Mt. Athos Road and isolation of the site.
- 7. U. S. Navy On September 4,1996, the Navy Radiation Safety Committee reported the loss of 9 pCi of americium 241. The material, in the form of two 4.5 pCi foil sources, was in a laser target designator pod on board an Navy F/A 18 alteraft which crashed into the Atlantic Ocean near Puerto Rico on August 9,1996.
- 8. Agirement State issues Attachment 1 I
4 One Liners 5 No significant issues were reported by the Agreement States in Region II during the week.
- 9. Fire on Federal Express Plane
' A fire on a Federal Express plane enroute to Boston, MA., occurred on September 5.
De plan was carrying radioactive materials, including " spent
- radiography sources from Alonso and Caruso, a radiography licensee in Puerto Rico. See attached PN.
- 10. Anplied Radiant Energy Comnany Westinghouse Hanford completed the loading of the final nine WESF sources into the shipping cask and has removed the cask from the irradiator pool on September 5, 1996. The cask will be dewatered and dried over the weekend. The shipment of the sources to Hanford is scheduled to take place on September 11, 1996.
Attachment 1
-1. September 5, 1996 P*ttLIMINARY NOTIFICATION OF EVENT OR UNUSUAL OCCbaRENQ3 PNO-I-96-063
.4 preliminary notification constitutes EARLY notice of events of POSSIBLE safety or public interest significance. The information is as initially received known withoutI verification by Region staff or evaluation, and is basically all that is (King of Prussia, Pennsylvania)n on this date.
Facility Stewart Airport Licensee Emeraency Clammification Stewart Airport Notification of Unusual Event Alert Nawburgh,New York Site Area Emergency General Emergency X Not Applicable
Subject:
FIRE ON FEDERAL EXPRESS AIRPLANE - STEWART AIRPORT, NEWBURGH, NEW YORK Region I was notified by personnel from the New York State Department of Health Bureau of Environmental Radiation Protection that the cabin of a Federal Express airplane enroute from Memphis to Boston made an emergency
-landing at Stewart Airport after it caught fire while in flight. The DC-10 plane was transporting hazardous materials and radioactive materialr..
The Bureau responded with a team of 5 central office (Albany) and regional (New Rochelle) radiological health specialists. Health I physicists from Indian-Point Units 2 and 3 also responded at the request Orange County emergency management personnel.
Autormation obtained by the State from the Federal Express radiation safety officer material indicates that the airplane carried several radioactive packages, including three containers of Xenon-133 (total activity GBq) 73 Gigabecquerels), Iodine-125 (0.28 GBq), Iodine-131 (21.9 two industrial radiography cameras, one containing 29 Ci curies and the other 12 curies Iridium-192, C-14 (50 microcuries), P-32 (250 j microcuries), H-3 (5 mci), and S-35 (2 mci). The fire was reported to have started in a container with hazardous materials only. The radioactive materials were loaded in a separate container close to where the fire started. The type of the hazardous materials involved and condition of their container is not known at this time. The plane was reported to have sustained extensive damage. The flight crew were taken to a local hospital, treated for emoke-inhalation and released. Radiation surveys of firefightcra showed no evidence of contamination.
Region I Public Affairs Office is prepared to respond to media inquiries.
Contact:
-(610)337-5216 a t s. a -
W[
- UNITED STATEal
/"@S E849'o, NUCLEAR REGULATORY COMMISSION I ^
"ffETdbhi!!"J'"
..... September 9. 1996 General Electric Company ATTN: Mr. C. P. Kipp General Manager GE Nuclear Energy Production P. O. Box 780 Wilmington NC 28402 4
SUBJECT:
NRC INSPECTION REPORT NO. 70 1113/96 07 l
Dear Hr. Kipp:
This refers to the inspection conducted on August 5 9, 1996. at the Wilmington facility. The aurpose of the inspection was to determine whether activities
( cuthorized by tie license were conducted safely and in accordance with NRC requirements.: At the conclusion of the ins ction, the findings were discussed with those members of your staff dentified in the report. l Areas examined during the inspection are identified in the report. Within these areas, the inspection consisted of selective examinations of procedures and representative records, interviews with personnel, and observation of activities in progress.
Within the scope of the inspection, violations or deviations were not i identified.
In accordance with 10 CFR 2.790 of the NRC's ' Rules of Practice." a copy of this letter and its enclosure will be placed in the NRC Public Document Room.
Should you have.any questions concerning this letter. please contact us. '
Sincerely.
EDT d( u ._
Edward J. McAlpine, ief Fuel Facilities Branch Division of Nuclear Materials Safety Docket No. 70 1113 License No. SNH 1097
Enclosure:
HRC Inspection Report cc w/ enc 1: (See page 2) m e . c ,,. n yp9
L' GE 2
cc w/ encl:
Dr. Ralph Reda, Hanmer fuels and Facility Licensing General Electric Company P. O. Box 780. Hail Code J26 Wilmington, NC 28402 i Dayne H. Brown. Director Division of Radiation Protection N. C. Department of Environment.
Health & Natural-Resources P. O. Box 27687 l Raleigh, NC 27611 7687 :
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1-c GE 2
cc w/ encl:
Dr. Ralph Reda. Manager .
Fuels and Facility Licensing General Electric Company P. O. Box 780, Hail Code J25 Wilmington,NC 28402 Dayne H. Brown, Director Division of Radiation Protection N.Health C. Dep& artment c' Environment.
Natural Resources P. O. Box 27687 Raleigh, NC- 27611 7687 i l
Jistribution w/enel: i E. McAlpine, RII
& Gr Troup, RII-R. Bellam G. Shear,y. RI RIII C. Cain, RIV .
F. Wenslawski, RIV DUBLIC Distributionw/oengl:
License fee Management Branch tr# Topiatte occiprxte y Ad io 0FFitt Ril:DNMS All:tWM5 RII:DNMS Rll:DNMS
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DAT( 09 / ($ / 96 09 / [p / 96 09 / T / 96 09 / 9 / 96 09 / / 96 09 / / 96 com /fts; _NO , to NO Cfl.d NO (Vih No YE1 N0 YE1 h0 Of fICIAL RECORD COPY 00CUMEhl NAME: I;\f f ElliREKsT5\GE96 07.RPI
U.S. NUCLEAR REGULA10RY COMMISSION REGION 11
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Docket No.: 70 1113 i
License No.: SNH 1097 1 Report No.: 70 1113/96 07 Licensee: General Electric Company Wilmington, NC 28402 4
Facility Name: Nuclear Energy Production Dates: August 5 9, 1996 Inspectors: W. Gloersen. Project Inspector A. Gooden, Radiation Specialist Approved by: E. J. McAlpine. Chief Fuel Facilities Branch Division of Nuclear Materials Safety Enclosure
. . C p off. Q 7 o" = L 7 p
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- 2 Executive Summarv Badiation Protection The inspection of the licensee's radiation protection program involved a review of the following health physics program elements: independent radiation safety audit, external and internal exposure controls, organizational change, respiratory protection program NRC Form 5 reporting, instrument calibrations, and alpha sources leak testing program. Within the areas reviewed, radiation protection management controls and procedures were in ) lace and appeared to be adequate to protect the health and safety of )lant wor (ers. Personnel exposures (internal and external) were less t1an 10 CFR Part 20 requirements and/or GE administrative limits.
Radioactive Solid Waste Manaaement The licensee's controls, procedures, and waste management program appeared acceptable and capable of accomplishing its safety objectives. The CaF relocation project had progressed ahead of schedule, however, implementation of the Final Status Survey and Release Plan, Revision 2.' dated February 28, 1996 had not begun. The present arrangement for the storage of LLRW was less than adequate due to the large volume of waste being stored and due to the fact that the cardboard and wooden waste storage containers were not shielded from the natural elements. Within the Low Level Radioactive Waste management program, the licensee had implemented a waste minimization / prevention program to reduce the amount of solid waste generated onsite. In addition, the licensee was in the process of constructing a waste treatment facility for the decontamination and volume reduction of the waste generated. The radioactive waste management program was managed effectively.
Transportation The transportation activities were managed effectively and the associated directives and procedures which incorporated and im)lemented the applicable provisions of both NRC and 00T regulations were tecinically adequate.
Attachment:
Persons Contacted and Exit Interview List of Items Opened, Closed, and Discussed List of Acronyms Enclosure
- {
l 3
- 1. RADIATION PROTECTION (83822) 1.1 Discussion 10 CFR 20.1101(a) requires each licensee shall develop, document, and implement a radiation protection program commensurate with the scope and extent of licensed activities and sufficient to ensure compliance with the provisions of $$20.100120.2402.
1.1.1 Radiation Protection Oraanization and Staffina The inspector discussed with a licensee representative organizational changes swcific to the radiation protection program since the last MPC assessment of t11s area (January 1996 documented in IR No. 70 1113/96 01). Yhe inspector noted that the structure and reporting chain for the RP function had not changed since the previous inspection. However, the previous Hauser of Nuclear Safety was reassigned during June 1996. As a result of the vacancy, a Criticality Safety Program Manager was acting as Nuclear Safety Manager until :
the vacancy was filled. The inspector discussed with the licensee contact the overall impact of this change on the Nuclear Safety organization. No significant adverse effect had been identified by the organization as a result of the temporary change.
1.1.2 Audits and Agoraisals SNM License 1097 requires audits to be performed in accordance with procedures to determine if actual operations conform to criticality and radiation safety requirements. An independent audit of the radiation safety program was performed during the period November 28 to December 1. 1995. The inspector reviewed documentation to show that the Audit Team qualifications were approved by GE management in accordance with license conditions. The audit was detailed and compliance oriented.- No potential noncompliances were identified. Items identified during the audit as findings were assigned to the regulatory tracking system for tracking corrective-actions. All items opened during the audit had been appropriately addressed and were considered c60 sed. The indeper. dent audit was effective in the identification of findings, recommendations, observations, and best practices.
1.1.3 Exoosure Controls 10 CFR 20.1201 requires each licensee to control the occupational dose to individual adults to established dosa limits. 10 CFR 20.1502(a) and (b) require each licensee to monitor occupational ex supply and require the use of individual monitoring posure devicestofor radiation and to adults likely to receive an annual dose in excess of 10 percent of the limits in 10 CFR 20.1201. For the period January 1,1995 to December 31, 1995, the maximum TEDE was 2.47 rem (compared to the NRC limit of 5 rem) and was assigned to an employee working in the powder preparation area. The maximum assigned SDE was 1.11 rem and was assigned to a ceramic worker. The GE Action Limit for SDE was 40 rem / year while:the NRC limit is 50 rem / year. The average Enclosure
l .
4 SDE axposure for 1995 was 30 mrem / year. During the previous reporting period (1994), the average was 43 mrem / year. The external deep dose maximum ex)osure was 480 mrem / year, and the 1995 average was 18 mrem / year. The maximum D)E exposure was assigned to a worker in the bundle assembly area. The GE action limit was 4 rem / year while the NRC limit was 5 rem / year. According to an exposure graph and a discussion with a licensee contact, the annual DDE average had continually decreased. During 1992, the average was 44 mrem, 1993 was 29 mrem, 1994 was 27 mrem, and 1995 was 18 mrem. The inspector noted that the maximum internal ex>osure (CEDE) during 1995 was 2.413 rem. The GE action limit was 4 rem while t1e NRC limit was 5 rem. The maximum assigned CEDE was l
to a worker in the powder preparation area. The average internal dose for 1995 was 0.27 rem / year. In addition to exposures for 1995, the inspector reviewed assigned doses current as of August 5, 1996 and noted that the maximum assigned SDE was 0.28 rem, the maximum CEDE assigned was 1.247 rem, and the maximum TEDE was 1.247 rem. The maximum DDE (based on TLD resulu. for the first five months of data in 1996) was 0.i' 8 rem.
The inspector reviewed the licensee's bioassay program which was detailed in two documents: P/P 4019 Bioassay Program
- and NSI 0 2.0
- Bioassay (Excreta) Program". The inspector reviewed selected personnel exposure records for individuals assigned work in areas where soluble uranium materials were processed. A review of the urinalysis tracking system printout disclosed that selected workers were providing urine samples for analysis as required by procedures. Further, a review of gra]hs and summary documentation disclosed a significant reduction from CY 95 to t1e corresponding months in 1996 in the number of workers failing to submit urine samples at the end of the work week.
For example, during Harch 1995, total missed were 25 compared to March 1996 total of ten. The overall trend during CY 96 thus far was a reduction in the number of missed samples. The inspector further noted that intakes were calculated as required when urinalysis results exceeded established action limits and that no worker had exceeded the licensee's administrative intake limit of 7.5 mg/wk. The inspector reviewed the licensee's cross check program fcr sample analysis quality assurance. On a monthly basis, urine samples with known amounts of uranium were submitted to the onsite lab and an offsite vendor for analysis. The inspector reviewed documentation for the period February to June 1996 and noted good agreement in the results that were obtained.
As part of the exposure monitoring program, the inspector verified that the licensee had provided NRC Form 5's for 1995, summarizing the occupational radiation exposure to those workers who required exposure monitoring for the year. Documentation was provided the inspector to show that the Form 5's were 3rovided prior !.c April 30, 1996. The inspector determined that the licensee lad adequately monitored personnel exposures and all of the assigned exposures were within regulatory limits.
1.1.4 Facility Tours During the inspection, the inspector toured the chemical product line, pellet production and rod load areas. the VRU, the CheHet laboratory, the laundry Enclosure
- l l 5 area, the URU Pump Control Room, and the Conversion Control Room. During the l tours, the inspector did not identify any significant weaknesses in the licensee's progran to maintain and control radioactive materials, nor with area postings. The inspector did note the )resence of candy or gum wrappers in the chemical product line area and the URU. The inspector discussed during the tour and exit interview the presence of such material ma employee disregard for area postings and exposure controls. y indicateIn response to the inspector's comments, a licensee contact informed the inspector that l evidence of this sort in a controlled area is considered a serious matter and !
GE would continue to monitor this incident for taking actions as appropriate.
The inspector accompanied and observed a Radiation Control Technician perform smear-surveys of the powder transfer area. In addition, the inspector i observed personnel performing one minute counts on smear samples for the presence of contamination. No elevated samples or equipment problems were noted. Both the survey technique and method of analysis performance was adequate for detecting the presence of contamination. Based on a random review, where required, personnel were properly wearin0 respiratory protection and dosimetry.
1.1.5 Instruments and Eauioment Several survey instruments were checked and verified at iarious locations as operational and within calibration dates. In addition, maintenance and/or calibration records were reviewed for an assortment of instruments to ascertain if calibrations were done in accordance with NSI 0 4.0 " Nuclear Safety Instrumentation." No problems were noted. Administrative controls were in place and appeared to be effective for ensuring that instruments were calibrated at the required frequency.
The inspector discussed with a licensee representative the license requirements for conducting periodic _ leak testing of radioactive sources. The licensee's performance in this area was verified via a review of the leak testing program for-selected alpha sources. No problems were noted. The documentation disclosed that selected sources were calibrated at the frequency as specified in the license.
1.1.6 Resoiratory Protection The inspector reviewed procedures governing the respiratory protection training and maintenance program. In addition, the inspector viewed the respiratory protection portion of the video used for initial and refresher training purposes. Topics included types of mask, donning equipment, proper fit test. rules for use, and checking operability of equipment. No written test specific to respiratory protection was required following training, According to the training contact, for initial respiratory training, a
- practical demonstration (mask use) is administered at the end of training session. -Respiratory protection training records was reviewed for several individuals assigned to the controlled area. No problems were noted.
Enclosure
e i
6 The inspector reviewed and discussed the respiratory protection maintenance program with the licensee contact assigned responsibility in this program area. The inspector reviewed documentation for monthly surveillance of SCBA equipment for the period January to July 1996. In addition, SCBA equipment was examined by the inspector at the URV Pump Control Room and the Conversion Control Room. No problems were noted. Tanks indicated full, rasks appeared to have been cleaned and no visual damage was observed to rubber hosing, valves, or face piece. The inspector also observed face masks cleaning and l
laundering by laundry personnel. Although no issues were identified in this I program area, the inspector discussed in detail the housekeeping improvements necessary to alleviate unnecessary delays in obtaining SCBA equipment from storage areas. During the tour, the licensee experienced a delay in excess of five minutes attempting to locate equipment as a result of modifications in the area; and in another instance, equipment was improperly stored and impeded access to the SCBA equipment.
1.2 Conclusion The inspector determined that within the areas selected for review, the licensee's performance in radiation protection complied with regulatory requirements and license conditions, and appeared to be adequate for protecting the health and safety of employees and the public.
- 2. RADI0 ACTIVE SOLID WASTE MANAGEMENT (84850) 2.1 Discussion During the inspection, licensee programs associated with the management of low level radioactive waste (LLR4) and the Calcium Fluoride (CaF ) Relocation Project were reviewed and discussed with cognizant licensee repr,esentatives.
2.1.1 Decommissionina Activities (CaF. Pelocation Pro.iect)
The inspector reviewed the licensee's progress in removing the Caf, from the northwest CaF storage area and relocating the material to an above ground storagefacilityintheControlledAccessArea(CAA). The licensee planned on storing the material until economic recovery of the uranium in the CaF, could be achieved. The northwest storage area consisted of seven shallow trenches located on the high ground in the northwest quadrant of the facility. CaF, sludges were placed in the in ground storage trenches from the time of conversion o>eration commencement in 1968 to the time of Waste Treatment Facility ( W ) operation in 1972. The seven pits cantained approximately 70,000 cubic feet of CaF, material. From February 1996 to May 1996, the licensee excavated the northwest storage area and moved approximately 120,000 cubic feet of soil and CaF, containing 3255 kilograms of uranium to the storage warehouse. From May to July 1996, the licensee screened ap3roximately 69,000 cubic feet of soil between 30 ti/ gram and 200 Pci/ gram so t1at this material could be shipped to the GSX _aidlaw Facility (a Resource Conservation Recovery Act (RCRA) hazardous waste burial facility) in Pinewood, South Carolina. The licensee was authorized in the special authorization section of Enclosure
7 the license application, Section 1.8.5, Disposal of Industrial Waste Treatment Products) to dispose of waste containing small quantities of LEU in accordance with Option 2 of the 1981 Branch Technical Position (BTP), " Dis:osal or Onsite Storage of Thorium or Uranium Waste from Past Operations," whic1 basically limited the uranium concentration in the waste material to 250 pC1/gm (insoluble) and 100 pC1/gm (soluble). As of the end of July 1996, the licensee had essentially completed the excavation.
By letter dated April 16, 1996, the NRC had approved the licensee's Final Area (FSS Plan),
Status Survey Revision 2, datedand Release February 28,Plan 1996.for The the Northwest purpose of CaF,the FSS Plan was to demonstrate that residual radioactivity concentrations at the Northwest CaF, Storage Area satisfy the NRC criteria for future use without licensing restrictions and radiological controls. The licensee was in the process of evaluating the residual radioactivity remaining in the burial trenches. The licensee had identified contaminated soils below the groundwater level. As of i June 24, 1996, soil concentration remaining in the soil, which included soils below the water table ranged from approximately 0.5 pCi/ gram to 129 pCi/ gram.
The inspector also reviewed groundwater monitoring data collected from 15 wells (identified as CAF Series wells) located around the Northwest CaF, Storage Area during the period February 1 to June 25, 1996. The licensee analyzed the well samples for gross alpha, gross beta, total uranium and fluoride. During the excavation, a maximum concentration of 3181 pCi/ liter total uranium was noted at well location 6A. Well 6A was downgradient from the excavation site. On June 25, 1996, the total uranium concentration at well 6A was 706 Pci/ liter. Total uranium concentrations at Well 12A, which was located at the site boundary ranged from 10 pC1/11ter to 1613 pC1/ liter.
The maximum occurred during the excavation activities. On June 25, 1996, the total uranium concentration at Well 12A decreased to 37 pCi/ liter. The licensee indicated to the inspector that the groundwater data collected since June 1996 would continue to be reviewed so that any trends could be identified.
2.1.2 M tc Manacement As of July 1,1995, the State of South Carolina was no longer accepting LLRW generated from licensee located in the State of North Carolina for burial at the Barnwell Facility. Thus, at present, the licensee's General Electric Company Nuclear Energy Production facility in Wilmington, NC had no options for the disposal of LLRW, Various solid wastes were generated from the fuel manufacturing operations (FH0) and field examination technology (FET) operations. FHO solid LLRW ranged in form and type, such as, packaging and construction materials, worn cut tools and equipment, spent process oils and chemicals, uranium sludge, and by product generated hydrofluoric acid. FET solid LLRW consisted primarily of compactible and non compactible dry active wastes (DAW) and cuno filters. FET wastes were volume reduced offsite and then stored onsite in 55 gallon drums at the FET radwaste storage warehouse.
Enclosure 4
l
? .
1
- 8 At
) resent, the licensee temporarily stored radioactive solid wastes pending l
autiorization to resume shipments to an authorized LLRW disposal facility.
The inspector toured the licensee's temporary waste storage locations, which consisted of several outside storage areas or
- pads' located in the controlled area. These pads were secured to prevent the unauthorized removal of the LLRW.
! The licensee had designated Pad Nos. 6 and 8 for waste box storage. As of August 5,1996, approximately 444 metal (50 cubic feet each) waste boxes containing noncombustible wastes were on the storage pads. The licensee was in the process of consolidating the wastes to reduce the number of boxes on the storage pads. In addition, the licensee was negotiating with a waste disposal facility in Utah to accept the noncombustible waste debris.
Contaminated, combustible. solid process wastes were stored on Pad No. 7A.
Combustible wastes were stored either in wooden or cardboard boxes. The combustible wastes, such as rags. mops, shop paper, plastic, and worn out i protective clothing, were designated for burning in an onsite incinerator designed for processing uranium contaminated wastes. Incinerator ash containing sufficient uranium to make recovery economical was processed onsite for the recovery of the uranium.
As of August 5, 1996, there were approximately 730 cardboard and/or wooden boxes on storage pad 7A. The licensee's goal was to have an inventory of approximately 600 boxes on storage pad 7A by the end of 1996. The backlog was due to the shutdown of the incinerator on March 27, 1996 due to a waste box in the incinerator load chamber that burned due to a failure of the loader ram
- limit switch. Extinguishing the waste box resulted in damage to another box queued for incineration on the scissors lift. The incident was a reportable event due to the damage caused by the fire. The licensee resumed incinerator operations in June 1996 after a thorough review of this event and another event which occurred on April 2.1996 at the Nuclear Fuel Services. Inc.
facility which involved a fire in the incinerator ducting. A review of this entire incident was documented in Inspection Report No. 70 1113/96 203 and a review of the licensee's corrective actions was documented in Inspection Report No. 70 1113/96 09.
The ins >ector observed that the licensee had implemented common sense approacles to waste minimization and prevention. This program consisted of (1) minimizing the introduction of non contaminated items into the restricted area: (2) segregation of highly contaminated waste items from low contaminated waste for the recovery and recycle of uranium as well as the decontamination of material for reuse where feasible; and (3) localized separation of wastes into combustible and noncombustible categories. The inspector noted throughout the restricted area the placement of combustible and non combustible waste receptacles. A written list of examples of both types of wastes were placed on each of the receptacles, Enclosure
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2.1.3 Facility Tours The inspector toured the facility where the Decontamination and Volume Reduction Facility (DVRF) was being constructed. The licensee had allocated funds to refurbish the old GECO (dry conversion) facility to be used for the construction of the DVRF. The licensee estimated that the DVRF would be operational by October 1996. The DVRF was designed to make increased use of available technologies for the decontamination ecuipment (ultrasonic cleaning, ultra high pressure washing) so that contaminatec items could be decontaminated to the unrestricted use limits as specified in Section 1.8.2 of the license application and the use of super compaction equipment for waste volume reduction.
Throughout the rest of the facility, the inspector noted proper use of the combustible and non combustible waste receptocles. The inspector toured the licensee's temporary waste storage locations, as noted abovo, which consisted of several outside storage pads located in the controlled area. The present arrangement for the storage of LLRW was less than adequate due to the large volume of waste being stored and due to the fact that the cardboard and wooden waste storage containers were not shielded from the natural elements.-
However, the restart of the incinerator should reduce the volume of combustible waste stored onsite and a successful negotiation with a wasto
-disposal facility in Utah to accept the noncombustible waste debris should reduce the amount of noncombustible waste presently being stored onsite.
2.2 Conclusion The inspector concluded that the licensee's controls, procedures, and waste management program appeared acceptable and capable of accomplishing its safety objectives. The CaF relocation project had progressed ahead of schedule, however, implementation of the FSS Plan, Revision 2.- dated February 28.-1996-had not begun. In addition, the present arrangement for the storage of LLRW was less than adequate.
- 3. TRANSPORTATION ACTIVITIES (86740) 3.1- Discussion 10 CFR 71.5(a) requires each licensee who transports licensed material outside the confines of its plant or other place of use, or who delivers licensed material to a carrier for transport, to comply with the applicable requirements of the regulations ap Department of Transportation (00T)propriate to the in 49 CFR Parts 170mode 189. of transport of the 3.1.1 Shionina Activities During the onsite inspection licensee transportation activities-regarding shipments of unirradiated fuel. UF, heels, and of other radioactive materials were reviewed. Selected records for the following consignments were reviewed in detail Enclosure i 4
( ,
10 RQ, Radioactive Haterial Fissile, NOS, 7. UN2918, for selected separate shipments of fissile material containing fuel elements shipped to selected power reactor facilities from January 1996 to July 1996.
RQ, Radi > active Haterial Fissile, NOS, 7. UN2977, for selected shipments of fiss'le material containing heels of enriched uranium hexafluoride in 30A/30B cylinders from January 1996 to June 1996.
Radion tive Material. NOS, 7, UN2918, for a shipment of 4907 kilograms of uranium dioxide to a fuel facility on March 18, 1996.
The inspector reviewed and discussed in detail the documentation used, and subsequently maintained in the licensee's records for each radioactive material shipment, including, the Bill of Lading, Radioactive Material Shipment Record, Vehicle Inspection Report, Receipt and Loading Verification Checklist, Packing List (Fuel Assemblies / Component Assemblies), Fuel Shipment Information Form Container Log Sheet, and Health Physics Survey Forms.
In general, the shipping re:;ords referenced above were complete and the information supplied on the shipping papers was appropriate. The ins >ector noted isolated examples of shipping paper inconsistencies, however, t1ese problems were not considered significant issues end were~ corrected immediately. Another shipping paper inconsistency was noted in which the licensee had incorrectly referenced a shi> ping container as a IAEA Certificate of Compliance (CoC) USA /9019/AF package wien, technically, the container should have been referenced as a NRC CoC package for a shipment of 4907 kilograms of uranium dioxide to a fuel facility in Lynchburg, VA. The licensee acknowledged the inspector's observation and corrected the inconsistency.
3.1.2 Authorited packaaes 10 CFR 71.12 (Subpart C) requires, in part, that (a) a general license is issued to any licensee of the Commission to deliver to a carrier for transport, licensed material in a package for which a license, certificate of compliance (CoC), or other approval has been issued by the NRC and; and ap)1ies only to a licensee w1o (c)(1) has a copy of the s)ecific CoC, and ot1er approval of the package and has the drawings and otler documents referenced in the approval relating to the use and maintenance of the packaging and to the actions to be taken prior to shipment and (c)(2) complies with the terms and conditions of the license, CoC, or other approval as applicable, and the applicable requirements of Subparts A G, and H of this Part.
49 CFR 173.471 details additional requirements for the shipment of NRC-approved packages.
During the onsite inspection, licensee activities associated with packaging and shipping of radioactive materials were reviewed in detail. The inspector reviewed and discussed with cognizant licensee representatives selected Enclosure
11 aspects of current quality control (QC) program activities associated with Jackaging and transportation o
>etween January and July 1996.perations for fissile material shipments made The ins >ector noted that the fuel rod and assembly shipments were shipped using NRC CoC No, 4986 USA /4986/AF, Hodel Numbers RA 2 and RA 3 packaging.
The inspector verified that the licensee maintained the most recent revision of CoC No. 4986 (Revision 33) and referenced documents in accordance with 10 CFR 71.12(c)(1) requirements. In addition, the inspector verified that the licensee was a registered user of the package in accordance with 71.12(c)(3).
The inspector also verified that the licensee was a registered user of the NRC CoC 9196. USA /9196/AF, Model UX 30, which was an UF, cylinder overpack. The inspector verified that the licensee maintained the most recent revision of CoC No. 9196 (Revision 9) and referenced documents in accordance with 10 CFR 71.12(c)(1) requirements. The inspector noted that the licensee had purchased 15 Hodel UX 30 overpacks in 1995 and ten additional UX 30 overpacks in 1996. The ins)ector verified, with regard to reporting defects and noncompliances, tlat the procurement documents included the statement that the I
provisions of 10 CFR 21 apply as required by 10 CFR 21.31. In addition, the ins >ector verified by record review that the licensee 3erformed inspections of l eac1 overpack before the first use as required by 10 C R 71.85. The inspections ascertained that there were no cracks, pinholes, uncontrolled voids, or other defects that would significantly reduce the effectiveness of the packaging and that the gaskets were in place when the upper half of the package was removed. The licensee also conspicuously and durably marked the ackage packaging with itsassigned identification number modelby number, the NRC serial number, in accordance withgross 10 CFRweight, and a p/1.85.
Before applying the model number, the licensee, through its OA )rogram, verified that the packing had been fabricated in accordance wit 1 the designed approved by the NRC.
I In addition, the inspector verified that the licensee maintained the most recent revision of NRC CoC No. 9019 (Revision 23) for package Model No. BU 7, which was used once for shipping uranium dioxide powder not to exceed five sercent enrichment in the U 235 isotope to a fuel facility in Lynchburg, VA in 4 arch 1996. The inspector also verified that the licensee was a registered user of the Model BU 7 package.
The inspector also reviewed the records of three shipments of eitner powder or pellet samples to various laboratories from April 2 to July 30, 1996. In all cases, the shipping papers indicated that the material was shipped in a strong tight container (00T 12B30). 49 CFR 173.411 requires the use of industrial packagings. Industriel Packages must reet the requirements of 49 CFR 173.410 and 173.411 and are categorized as follows: (1) Industrial package Type 1 (IP 1): (2) Industrial package Type 2 (IP 2): and (3) Industrial package i
Type 3 (IP 3). At the time of this inspection, the licensee was unable to determine if the DOT 12B30 met the requirements for an IP 1 package. During the exit meeting, the inspector identified this issue as an unresolved item (URI) (70 1113/96 07 01 (JRI): Cetermine whether or not the strong tight Enclosure
s 12
+
container identified as DOT 12830 satisfied the packaging requirements specified in 49 CFR 173,410 and 173.411 for an industrial package Type 1),
3.1.3 Procedural Guidance During the onsite inspection the procedural guidance specified for the shipment of radioactive materials was reviewed against the selected criteria sacified in 49 CFR Parts 171 178, Specifically, the inspector verified that t1e necessary procedure revisions were made to implement the changes in the DOT regulations which became effective on April 1, 1996, In addition to the shipping paper requirements specified in 49 CFR Subpart C, procedural details were reviewed against requirements established for packaging (49 CFR Part 173), marking and labeling (49 CFR Part 172, subpart D SS172,400 407 and $$172.436 440), monitoring (49 CFR Part 171, subpart I),
and emergency response information (49 CFR, subpart G).
- TI 14. Exclusive Use Shipment. Revision 5, March 29,1995 and modification dated March 29, 1996
- TI 15 RA Inner Refurbishing, Revision 10 March 23,1996 a TI 16. Radioactive Material Packaging and Shipment Record, Revision 2, March 13, 1995 6nd modification dated March 29, 1996
- TI 18. Empty Radioactive Material Shipping, Revision 4, September 14, 1995 and modification dated March 29, 1996
- TI 19, Radioactive Name and Shipaing Papers, Revision 5, September 14, 1995 and modification dated Marc 1 29, 1996
- TI 20,-Radiation Radioactive Packages, Revision 2 June 8, 1995 and modification dated March 29, 1996
- TI 33, shipment of UF 6 Material and Containers, Revision 0, March 13, 1995 The inspector reviewed the procedures noted above and observed that the procedures incorporated and-implemented the applicable provisions of both NRC and DOT regulations.
In addition, the inspector verified that the appropriate personnel in the traffic department had current copies of the applicable DOT regulations.
The licensee used a vendor service that provided periodic updates to 40 CFR 106 180. This service provided a well organized version of the DOT regulations with an indexing system that allowed quick access to the appropriate regulation.
Enclosure t %
e 4
l 3.2 Conclusion The licensee's performance in this program was acceptable. Transportation activities were managed effectivelv and the associated directives and
>rocedures which incorporated and implemented the applicable provisions of
>oth NRC and 00T regulations were technically adequate. One unresolved item was identified.
4 I
Enclosure
.w , _ _ - .
ATTACHMERT
- 1. PERSONS CONTACTED 1.1 LicenseePersonnel D. Barbour, Coordinator, Radiation Protection
- J. Bradberry, Senior Program Hanager, Regulatory Team D. Brown, Team Leader, Environmental Project W. Croom, Area Coordinator (Bundle Assembly) 1
- N. Dookeran, Team Leader, CheHet Lab D. Dowker Team Leader, Environmental Process Team
- R. Foleck, Senior Licensing Specialist
- G, Fornasiero. Engineer, Packing and Transportation l *P. Godwin, Coordinator Fire Safety and Emergency Response l *H. Lamb, Team Leader, Powder Preparation and Packaging
- A. Habry Principal Nuclear Safety Engineer i
- D, McCaughey, Manager, Fuel and CheHet Lab Quality I G. Hobley, Coordinator, Maintenance
- S. Hurray, Team Leader, Chemical Conversion
- L, Paulson, Acting Manager, Nuclear Safety R. Raynor, Instructer, Training
- R. Reda, Hanager, Fuels and Facility Licensing .
- B. Robinson, Principal Nuclear Safety Engineer C. Rochelle, Lab Analyst, CheHet Lab
- E Rouse, Monitor, Radiation Protection G. Sbraco, Packaging /Trans>ortation Engineer K. Toussaint, Specialist, )rocess Control Engineering
- F. Walker, Manager, Shipping and Traffic
- T. Winslow, Manager, Material Control and Accountability Other licensee employees contacted included engineers, technicians, production staff, security, and office personnel.
- Denotes those present at the exit meeting on August 9, 1996,
- 2. Exit Interview On August 9, 1996, the inspection scope and results were summarized with licensee representatives. The inspectors discussed in detail the routine program areas inspected, and the findings, including the Violation listed below. There were no dissenting comments expressed by the licensee. Licensee management was also informed that an IFI from a previous inspection was reviewed and considered closed, Attachment
2 Item Number Status Descriolion and Discussion 70 1113/96 01 01 Closed IFI Verify the effectiveness of corrective actions in preventing leaks and contamination to laboratory hoods and/or sink drains.
70 1113/96 07 01 Open URI Determine whether or not the strong tight container identified as 00T 12B30 satisfied the packaging l requirements specified in 1 49 CFR 173.410 and 173.411 for an industrial package Type 1
- 3. List of Acronyms CEDE Committed Effective Dose Equivalent CFR Code of Federal Regulation CY Calendar Year .
DDE Deep Dose Equivalent GE General Electric j
IFI Inspector Followup Item IR Inspection Report MG/WK Hil11 grams per week HREM Hillirem NSI Nuclear Saf9ty Instruction P/P Practices i Pro:edures RP Radiation Pr)otection R/YR Rem Per Year SCBA Self Contained Breathing Apparatus SDE Shallow Dose Equivalent SNM Special Nuclear Material TEDE Total Effective Dose Equivalent TLD Thermoluminescent Dosimeter UF6 Uranium Hexafluoride URI Unresolved Item URV Uranium Recovery Unit VIO Viciation Attachment
- 't 3 !
- 4. Previously Identified Inspector followup Item (Closed) IFI 70 1113/96 01 01: Verify the effectiveness of corrective actions in preventing leaks and contamination to laboratory hoods and/or sink drains. 1 The licensee had completed corrective actions in accordance with the commitments detailed in Paragra>h 5 of IR 701113/96 01. The inspector reviewed licensee survey data sleets which documented the drain inspection results for the period January to June 1996. In addition to the records review, the inspector conducted an-independent survey and examination of several drain pipe surfaces and directly underneath the drain pipe in each hood located in the Chemet Lab. No problems were noted. -According to discussions with a licensee representative assigned responsibility in this area, all sink drain piping was replaced (at a cost of approximately $50,000) with a type of pipins known as 'kynar".
One other aspect of the-licensee's corrective actions reviewed by the inspector involved a self assessment of the adequacy of corrective actions taken by the laboratory during the past five years. The referenced audit was very detailed and critical of the laboratory corrective action program: and for those items identified as inadequate, according to documentation the proper corrective actions were being taken. The licensee's actions were considered appropriate for closure of the IFI.
1 Attachment
, +---w-,-.- + ,-m,-., ,--rr n--,v., . . -,h ,,, ew,, y , ,me - - , . - - - -.,vw--,s.r -
,..w*
t a
ONE LINERS FOR OCTOBER 11. lo96
- 1. fuel Facility Status
- a. B&W NNFD ,
Five inspectors from the International Atomic Energy Agency (IAEA),
accompanied by a NMSS representative will be on site the week of October 14 to review activities associated with the down blending of Project Sapphire material. This will include the installation and calibration of process monitors, review of data collected by existing instrumentation and collection of material samples.
The processing of Project Sapphire uranium beryllium metal alloys is progressing in Uranium Recovery. Downblending to low enriched material has started after several start up problems (replacing a PVC vent line which overheated, installing a bigger pump after piping system modifications) but converting the liquid to crystals may start October 11.
The licensee expects to receive the laboratory results for the material which i caused a visible " flame" in a dissolver early next week. Under the conditions !
vf the CAL, they will discuss the results with Region 11. !
Naval fuel and research reactor fuel areas are operating normally with no problems. Uranyl nitrate recovered from zero power fuel is being processed into feedstock for the fabrication of fuel for the High Flux Isotope Reactor. ;
NMSS has drafted a CAL to NNFD confirming their commitment not to ship certain uranium beryllium materials under the exemption (as fissile) under 10 CFR 71.53. NNFD identified that, under certain circumstances involving super moderators, the regulations were not sufficiently conservative for certain-package arrays. NNFD had informed the NRC (NMSS) of this on September 18 and had provided written commitment not to ship on September 20. NNFD also notified the Depanment of Transportation and the Department of Energy.
NMSS has requested RES revise the regulations promptly and is considering an Information Notice, b, Framatome Cogema Fuels (B&W Fuel Company)
Disassembly of bundles and down loading of fuel pelleta under the Nukem contract is continuing. Some difficulties were encountered with the removal of t'.e outer shroud around the assembly but this appears solved. The pellets from the fuel rods will be sent to Siemens for recycle.
Attachment I g3 k
One Liners 2 B&W fuel assembly is shutdown based on production schedules,
- c. General Electric The remnants of Hurricane Josephine passed the Wilmington area on October
- 8. Approximately 5 inches of rain fell at the site but no damage occurred to any of the facilities. All lagoons (chemical and waste water) remained within normal levels.
On October 8, the licensee made a notincation under Bulletin 91-01 when they determined that a bulging tank exceeded the nuclear safety specincation (see enclosed EN). The criticality safety analysis for the equipment essumed full reDection and an innnite slab (very conservative). The licensee reviewed l records of past contents of the tank and could nnd no instance when the tank contained the concentration limit that was one of the controls. The licensee plans to review the tank supports to determine if additional structural supports are needed.
Testing of the modification to Line 5 has been completed and it is now processing recycle material from Uranium Recovery (UR) up to 5%
enrichment. Prior to the modification, the recycle line was limited to 4%
although UR could process up to 5%.
ADU lines 1-4 are all operating as needed for production schedules. Pellet production, rod loading and bundle assembly are running on a 5 day schedule with no problems, The waste incinerator is running normally. All systems in UR are functioning normally except for the oxidation furnace. During the 6 month inspection of the mufne area for cracks or powder accumulations (NCS requirement) several heating elements were damaged and have to be replaced.
Selection of personnel to staff the new Direct Conversion Facility is expected next week. The original intention was that the selections would have been .
done in June and replacements / backfills trained for r.reas such as ADU, UR e and URLS before the staff went to DCF for training. This was not completed for several reasons. However, the start up date for the DCF has not changed and the training programs will have to be mn in the remaining time.
The Team Leader, Ceramic Production (pellet pressing and sintering) resigned to take a position with Kingsford Charcoal. The Team Leader, Powder Preparation left in September. Acting team leaders have been appointed pending selection of new leaders.
Attachment I
h One Liners 3
- d. Nuclear Fuel Services, Inc.
The licensee is continuing to receive shipments of uranyl nitrate (UNH) from Rocky Flats. However, problems with the ll liter solution cylinders meeting the dimensional requirements of the Certificate of Compliance may delay some future shipments, ne licensee is continuing to conduct equipment check-outs and functional tests ,
of portions of the systems which will be used in the UNH processing.
Components are being repaired or replaced as a result of this activity.
During a conference call on October 9, the licensee's progress toward completing all the items specified in their Rocky Flats Operational Readiness Review Plan was discussed. The licensee committed to submit a response concerning their nuclear criticality safety evaluation of the operation on October 10. It now appears that all other " readiness" items, except training, will be completed by October 25. Training is now scheduled for the period from October 28 through November 15. As a result, the start date for the actual processing of Rocky Flats material has been slipped and now is projected for November 18 at the earliest. NMSS and Region 11 are l coordinating a joint inspection which could be conducted the week of November 4.
- e. Westinghouse Conversion line four is shut down for major modincations and upgrades, and will remain shut down for another four to six weeks. UF. is currently being run on three conversion lines. A fourth line is running uranyl nitrate recycle material. Other portions of the plant are operating on a normal schedule, i.e.
pelleting and bundle assembly, IFBA, and recovery / waste processing.
Since the last report, the licensee has experienced no process upsets or other safety signincant events. They are continuing to operate on their new schedule of seven days per week for chemical conversion operations and six days per week for pelleting and recovery.
- 2. Aercement State Issues
- a. Action to Remove Sources at Bankrupt Facility in Tennessee See attached PN and press release. USDOE/ORNL took possession of the three radiography sources on October 8,1996, and they were transported by Attachment 1 I
i
s b
One Liners 4
, the State of Tennessee to ORNL for storage until a decision is made on final disposition.
- 3. Medical Consultant's Renort for Dr. Fernandez's Misadministrations A copy of the medical consultant's report will be transmitted to the Puerto Rico Health Department (PRHD) and Dr. Fernandez. The PRHD will be requested to do follow-up on the patients, in addition, an Order is to be issued this week requiring review of patient records, notification of patients who received misadministrations, and transfer of the sources in 90 days.
4 Herman Strauss.Spran Yard Soureg See attached PN and Morning Report. The EPA and EPA contractors opened the container on site on October 10 to identify the radioisotope involved and to initiate actions to ship the sources off site. They found low lesels of contamination on the >
inside of the container (20,000 cpm). They identified the isotope to be radium by portable spectral analysis. They planned to rescal the source, package it for j shipment, and ship it to a contractor facility in Tennessee.
- 5. New Director. Naval Nuclear Proculsion
( Admiral Bruce Demars, Director, Naval Nuclear Propulsion, retired on October 1.
l Vice Admiral Frank Bowman, currently Chief of Naval Personnel, was selected a.i his replacement.
- 6. Enforcement Discretion for Reciorocity Casu On October 7,1996, enforcement discretion letters were sent to four State of Florida licensees (Okaloosa Asphalt, Inc.; Edward M. Chadboume, Inc.; Baily Engineering and Testing, inc.; and Larry M. Jacobs, Inc.) who had used moisture / density gauges in areas of exclusive federal jurisdiction in Florida without filing for reciprocity (thus without an NRC license). These licensees had been unaware of the need to file for reciprocity.
(
Attachment 1 l
1
S 4
FUEL CYCLE FACILITY EVENT NUMBER: 31113
.CILITY: GE NUCLEAR ENERGY NOTIFICATION DATE: 10/08/96 RXTYPE: URANIUM FUEL FABRICATION NOTIFICATION TIME: 11:20 ( E'.
COMMENTS: LEU CONVERSION (UF6 TO UO2) EVENT DATE: 10/07/96 LEU FABRICATION EVENT tit'E: 16:00(ED'.
LWR COMMERICAL FUEL LAST UPDATE DATE: 1r/08/96 CITY: WILMINGTON REGION: 2 -
COUNTY: NEW HANOVER STATE: NC NOTIFICATIONS LICENSE #: SNM-1097 AGREEMENT: Y DOCKET: 07001113 JOHN POTTER RDO COMBS /TROSKOSKI EO NRC NOTIFIED BY: PAULSON HQ OPS OFFICER: CHAUNCEY GOULD EMERGENCY CLASS: NOT APPLICABLE 10 CFR SECTION:
NBNL RESPONSE-BULLET 1N EVENT TEXT 0 THE 1ULGING.
LICENSEE REPORTED LOSS OF GEOMETRY CONTROL AS THE RESULT L THE BIENNIAL TANK WIDTH MEASUREMENT EVALUATION SHOWED THAT THE GEOMETRY CONTROL HAD BEEN EXCEEDED ON ONE TANK. THE LARGEST ACTUAL MEASUREMENT IN THE LOCALIZED BOWED AREA WAS 4.86 INCHES EXCEEDING THE ALLOWABLE 4.45 INCHES. THE SECOND CONTROL, CONCENTRATION, REMAINED IN EFFECT. THE MAXIMUM CONCENTRATION WAS LESS THAN 1 PPM U (LIMIT +16,000 PPM U). THUS NO UNSAFE CONDITION EXISTED. TOTAL URANIUM INVOLVED WAS 6 GRAMS.
THE LICENSEE IS REPORTING THIS CONDITION PURSUANT TO NRC BULLETIN 91-01.
THIS REPORT IS A 24 HOUR REPORT SINCE GEOMETRY WAS LOST AS AS RESULT OF TANK BULGING. GEOMETRY CONTROL WAS RE-ESTABLISHED WITHIN 4 HOURS BY USING LARGE TIMBERS AND "C" CLAMPS APPLIED TO THE LOWER HALF OF TANK. THE SLAB TANK WILL BE DRAINED AND ADDITIONAL INTERNAL STAYS WILL BE INSTALLED TO PREVENT BOWING ON THE LOWER HALF OF THE VESSEL AS PART OF LONG-TERM CORRECTIVE ACTIONS.
THIS EVENT WAS OF A LOW SIGNIFICANCE SINCE THE CONCENTRATION LIMIT REMAINED INTACT (<1 PPM U IN SCRUBBER WATER INSIDE SLAB TANK) . MULTIPLE FAILURE l MODES WOULD BE REQUIRED BEFORE A CRITICALITY EXCURSION COULD BE POSSIBLE.
i m 1
V i
4' I
ONE UNERS FOR OCTODER 25.1996
- 1. ' Egel Facilliv Slaud
- a. Nuclear Fuel Services, Inc.
At 12:21 p.m. on October 22,1996, the licensee notified the NRC of a malfunction of their Uninterruptible Power Supply (UPS) system (EN31190).
Maintenance had been performed on the backup power generator to install an oil heater in preparation for cold weather. Upon completion of the ,
maintenance activity, the UPS system was tested to verify its operability.
During the testing at 11:
42 a.m. that morning, the system failed to properly switch to the backup power generator causing a loss of power to plant safety and security systems for four minutes.
The systems that were affected included security detection, criticality safety alarm and fire protection detection. The criticality safety alarm system )
provided visual and audible indication of system unavailability, ne plant fire
- safety system, and the security systems (central and secondary alarm stations, perimeter intrusion detection, and security lighting) were also without power l during this time, immediate compensatory actions were taken by the
. licensee's securit; staff la response to the event, ne critic.tlity alarm and fire protection system *ere functionally tested after restoration of power and found to be operating properly.
}; The licensee stated there were no similar problems during previous tests. The cause of the malfunction is currently under investigation by the licensee. No fuel processing was ' occurring at the facility at the time of the malfunction.
The regional staff will followup on the results of the licensee's investigation during the next inspection, and will verify that corrective actions have been taken.
A team inspection consisting of RH/DNMS and HQ/NMSS representatives will -
be onsite during the week of November 11-15,1996 to assess the licensee's operational readiness for the processing Rocky Flats HEIJNH. In addition, a senior management meeting at NFS, Inc. to discuss licensee corrective actions -
after the incinerator incident and the Performance Management Program has been scheduled for November 15, 1996.
Attachment 1
I
_.One Liners 2
- b. MW Naval Nuclear Fuel Division Fuel manufacturing operations in Naval fuel (AFP and MFP) and research reactor elements are operating normally, he fuel reclamation system in MFP
- is shut down because of mechanical problems with filters in the waste line.
De down-blending system for Project Sapphire in Uranium Recovery is shutdown due to a cl**ali=! problem in the columns. De columns apparently were not adequately cleaned before using them for this operation. Also, a . _
1
_ sample valve connection was broken, which drained the column bank, but this
. has been repaired. This was reported by the licensee on October 15 and discussed in last week's report. Hot functional testing of the product dryer is suspended until a product _ quality issue is resolved. Dissolution of U Be metal alloy material is continuing, and will probably run for 12 weeks more.
A force-on force security exercise was conducted on the evening of October j 24. His exercise was observed by ONMSS and Region II security personnel, i
An article in The Washinnton Timen reported that in excess of 120 pounds of l weapons grade uranium was missing from Project Sapphire. This alleged loss is the result of sampling and measurernents differences between the shipper and the receiver, See attached chronology and dmft memo to the Commissioners for further details,
- c. Framatome Conema Fuel (BW Fuel Comogy)
The fuel fabrication and bundle assembly lines have been shut down since late August because of schedule requirements, he lines will start back up the week of November 4 and operate on a two-shift. schedule.
The down loading of the Nuvem fuel is proceeding slowly. They have developed methods to remove the shrouds from the assemblies and are removing rods. One problem they have encountered is the thin wall of the cladding is easily crimped during rod removal. This has crushed some pellets and also makes pellet removal difficult.
' The service equipment facilities (SERF-2,3,4) are operating with no problems.
The licensee determined that the protection factor for respiratory protection equipment used during the down-loading oflhe Watts Bar fuel was less than previously assumed. This was based on information from the manufacturer.
Air sample results are being reevaluated and doses to about 12 workers -
Attachment I s
One Liners 3 recalculated. They will inform Region H of the results in about a week.
- d. Qcceral Electric Company Two management changes have occurred. Jack Sependa, Manager, GE-NE Emironment, Health & Safety (EH&S), accepted a position in a new business venture under GE Motors. Tom Hauser, Manager, GE-NE Quality is acting as Manager, EH&S, pending selection of a new manager. Sam Armijo, Gene.ral Manager, Nuclear Fuel, has taken a position in San Jose, reporting to GE Nuclear Energy Vice President. Mark Savoff, formerly President, Reuter-Stokes, replaced Armijo. l l
All fuel lines, including the recycle line, are operating this week. Only the hie and recycle lines will run over the week-end. All production lines ran last week-end because the lines were down for two days when the DI water system was down. Pellets, rodt and fuel assembly lines are running on a five day schedule. Uranium Recovery, Waste treatment and the incinerator are all nmning.
Personnel selection for the operating staff of the new production facility (DCP) was announced on October 16. Operator training will begin December 1, 1996. The project schedule is still to start hot functional testing with natural uranium about March 1,1997.
A GE design review of the program for the process control computer in DCP raised serious questions about the security and integrity of programmed parameters. GE will require that the program be rewritten to assure that it meets their requirements. Because of schedule problems, they plan to do cold and hot testing with the existing program but will require that the new program is in place and tested before enriched uranium is introduced, which is presently scheduled for early May. Region H will inspect this.
The first shipment (1,488 rods) of the second batch of fuel from the Wurgassen reactor facility in Germany has been received for down-loading. A total of 6,448 rods will constitute this batch. The fuel pellets are being removed frcm the rods and processed through Uranium Recovery for recycle.
A third batch is expected in the Spring.
The annual plant evacuation drill for day shift was satisfactorily conducted on October 24. The drill for the remaining shifts will be conducted next week.
Attachment 1 m _.
4 One Liners 4
e.
Westinchouse_
l Work on Conversion Line 4, which had been shut down for major modifications and upgrades, has been completed and the line is being checked out. The licensee expects to have that line operational by next week. UF. is currently being proccased through the three other conversion lines and a forth line is processing uranyl nitrate recycle material. Other systems in the plant are operating on a normal schedule, i.e., pelleting and bundle assembly, IFBA, and ret.overy/ waste processing. The licensee is continuing to operate i on a new sche 0ule of seven days per week for the conversion area and six days per week for pelleting, bundle assembly, and recovery.
Last week the licensee received a phone call from the State of South Carolina radiation protection personnel concerning some scrap me.1 that the licensee had sold to a local sctap dealer. The scrap metal consisted of metal grids and skeletons from fuel elements from the Vogtle site that had to be re-worked by the licensee due to over pressurization in the IFBA-type fuel The pins.
scrap dealer had in turn sold the scrap metal to a company in Pittsburgh. When the scrap metal arrived in Pittsburgh by rail car, the radiation monitors at the entrance to the site alarmed. The company in Pittsburgh would not accept the shipment and the rail car was shipped back to the scrap dealer in South Carolina. The State and the licensee sent radiation technicians to the scrap
&&r's facility to check on the problem. The technicians found that the background in the area read from 1012 micro-R per hour (pR/hr) and the l highest readings at one or two locations around the rail car were 18-20 pR/hr.
The licensee checked the scrap metal and found no loose surface contaminatior.. The scrap metal (about two dumpsters full) was taken back to the Westinghouse site.
2.
Indsvendent Soent Fuel Storare bstallation Status- Surry Recently, a lightening strike near the Surry site caused a loss of power to the ISFSI; the backup generator at the ISFSI started as designed and carried the load until powe was restored. As of October 18,1996, the licensee had 29 fully loaded dry storage casks (DSCs) for a total inventory of 638 irradiated fuel assemblies (IFAs). 28 DSCs were on Pad #1 and one DSC was on Pad #2. The number of fuel assemblies discharged in the fuel pool vns 874. The margin over full core reserve was 2.
Presently, the licensee has one TN-32 DSC onsite vyhich will be used to perform various dry run tests. The licensee was seeking authorization from the SFPO/NRC to use the TN-32 casks. NRC approval of the license for this cask was on hoM until questions on certain test results were resolved. Four TN-32 casks were in fabrication Attachment 1
One Liners 5
(of tic eight on order from Transnuclear) for expected delivery by the end of 1996.
The TN 32 transport dry run was performed on October 21,1996. The TN-32 k.auing dry run was scheduled for November 12,1996. 'Ihe TN-32 unloading dry run was scheduled for November 18, 1996. The actualloading of TN-32 DSC #1 was scheduled for December 9,1996 subject to the approval of the TN 32 TSAR and a Surry ISFSIlicense amendment by the SFPO.
3.
Pensacola Testine 12boratQdt1 On October 23,1996, mnhwi enforcement action was taken against Pensacola Testing Laboratories, Inc., Pensacola, FL for failure to file for reciprocity with the NRC before working under their Agreement State license in area of exclusive federal jurisdiction. Region II issued a Notice of Violation containing a Severity Level III violation for failure to comply with 10 CFR 30.3 and 10 CFR 150.20. No civil penalty was proposed because Pensacola Testing had not bet:n the subject of eebtrv3 enforcement action in the pa.st two years or two inspections and they took extensive corrective action.
- 4. Dr.Fernandez i
l
} Dr. Fc nand:z received the Order and Press Release, and agreed to preserve and j s arch the patient files of the late Dr. Vazquez, for patients with high risk for inexaocular infection resulting from Sr-90 induced scleromalacia. Region II is confirming this in a letter to Dr. Fernandez to be issued October 25,1996.
Office staff of the Secretary of Health Dept of Puerto Rico confirmed receipt of the NRC/NMSS request to perform followtip studies of high risk eye applicator patients, and issues.
referred the matter to their legal department for advice on liability and funding
- 5. Agreement State Slalus See attached 12 vent Notification Item No. 31188.
Attachment 1
., 4 Chronolooy and History of Sapohire Pro _iect 9/27/93 President Clinton announced po? icy to prevent proliferation of weapons (included purchase of h30 from former Soviet Union).
11/94 DOE acquired. approx. 580 kg of HEU from Kazakhstan under code name ' Project Sapphire". Material sealed in Kazakhstan by Y-12 personnel.
Material transported and stored in Y-12 Plant in Oak Ridge.
2/7/95 USEC issued request for bid for downblending material to B&WNNFD and NFS.
(USEC, as part of Energy Policy Act of 1992 established as responsible for US enrichment activities. Therefore USEC i-l obtained contract with B&W NNFD on behalf of DOE.
7/20/95 HRC held meeting onsite with B&W NNFD to discuss plans for l project.
B/95 - 10/31/95 Material received at B&W NNFD facility.
- Material shipped to B&W NNFD in original containers sealed in Kazakhstan. Upon receipt, R&W NNFD verified seal integrity.
1/10/96 Material received was in 4 main categories (metal, oxides,
.U/Be rods and U/Be scrap. Processing of material other than 1
scrap started on this date.
4 - 5/96 Characterization of U/Be scrap initiated.
5 - 6/96 Calicining of U/Be scrap conducted.
7 - 8/96 Dissolution processing of U/Be scrap started.
8/22/96 NHSS informed Region II that S&W NNFD discovered difference btwn amount of U/Be scrap detected during processing at site and amount DOE said delivered.
9/26/96 Memo / trip report from individual in DOE-ORO MC&A Branch discussing the bias.
-.10/24/96- News article in Washington Post e
+
9 MEMORANDUM FOR:
Chairman Jackson Commissioner Rogers Commissioner Dicus Commissioner Diaz Cosnissioner McGaffigan FROM:
James M. Taylor Exect.tive Director for Operations
SUBJECT:
MEDIA ARTICl.E ON PROJECT SAPPHIRE DISCREPANCIES ,
l The purpose of this memorandum is to provide clarification to the Commission regarding recent media articles concerning an apparent nuclear material discrepancy at the Bebcock and Wilcox (B&W) B&W plant in Lyn high enriched uranium (HEU) removed from Kazakstan under a U.S./Kazakstan cooperative effort referred to as " Project Sapphire."
effort, an apparent bias has been identified during the processiIn the course of this portion of the material identified as uranium / beryllium scrap (ng of a 232 kilograms).
involvement and NRC oversight are described below.The history of the mater -
In view of a concern that the HEU could not be adequately protected in Kazakstan, the U.S. 00E staff went to Kazakstan to package the material for shipment to
, Material types included uranium metal, homogeneous uranium oxides uranium / beryllium reds, heterogeneous uranium / beryllium scrap and mixed low,
! level residues in graphite crucibles and lab salvage. Measurements were performed on a best. effort basis and the containers tampersealed. The i
heterogeneous uranium / beryllium s, crap, which could not be representatively sampled andassay nondestructive is the(NDA).
source of the apparent referenced bias, was measured by uncertainty 100 percent. on the NDA measurements was large, on the order of plus or The Project Sapphire material was placed in storage by 00E at Y-12 in Oak Ridge, Tenneessee where it remained for approximately a year while final contract HEU arrangements were completed by DOE for processing and recovery of the at B&W, that time. No receipt measur6aents were performed by DOE on the material at Also during this period, discussions took place between the U.S.
and the International Atomic Energy Agency on when the material would be subject to IAEA safeguards as had been agreed with the Kazakstan Government.
Due to IAEA resource constraints and the difficulties that would be eccountered in applytr,g safeguards at Y-12, it was decided to defer IAEA safeguards until the caterial was received by B&W. In August 1995, shipments of the material by DOE were initiated. .NRC involvement was initiated prior to the receipt of material at B&W to ensure adequate accountability and facilitate the application of IAEA safeguards. R&W receipt activities included tampersafe seal verification, item identification and weight verification on an item basis prior to placing the material in the vault.
After processing parameters were established and agreement was reacted with the IAEA on where in the process international safeguards would be initiated.
B&W cocnenced sampling of the homogeneous uranium metal and oxides. NRC was present for the sampling and procured a split sample from that taken for the
- n. - . . , .
W
. J IAEA for independent verification. Analyses on samples randomly selected from this population by NRC's contractor laboratory were in excellent agreement with B&W's measureacnts.
the validity of the as-received measurements. Recovery results provided further confirmati i NRC also procured samples of !
the recovery product for independent confirmation and is awaiting results. No samples were taken of the heterogeneous uranium aforementioned difficulty in representatively s/ampling beryllium scrap such due to the material. It was agreed that accountability values would be based on samples taken after dissolution which is a routine practice in the industry.
B&W initiated dissolution of the uranium / beryllium scrap in July 1996. To-date approximately 30 percent of the 232 kilograms total have been processed in a number of process batches. There has been a consistent negative bias relative to what was declared on the as-received material. The current estimate of the magnitude of the bias is on the order of 28 percent. Due to a safety concern that developed in related to the measurement bias,further processing this particular processing has beenscrap material not delayed.
With regard to implication: for IAEA safeguards, the apparent bias will have no impact since the decision was to apply safeguards to the downblending only I which is not impacted by shipper-receiver differences. This decision was -
based on IAEA's position that no statements could be made by the Agency regarding the origin of the material. , b Headquarters and Region Il staff have reviewed the results of safeguards 8 8
inspections perfomed at B&W since the initiation of Project Sapphire material 3pI' processing and determined that there were no abnormalities to indicate a potential loss or diversion.
Staff will continue to monitor the processing of the Project Sapphire material, including the taking of split samples for independent verification and monitoring the magnitude and implications of the apparent bias during the completion of the scrap processing.
O bd9 No im __
O w
TRANSPORTATION EVENT EVENT NUMBER: 31188 L.cENSEE: DUPONT CITY: MEMPHIS NOTIFICATION DATE: 10/22/96 COUNTY: SHELBY REGION: 2 LICENSE #: STATE: TN NOTIFICATION EVENT DATE: TIME: 04:34 (ET}
AGREEMENT: Y 10/22/96 DOCKET: EVENT TIME: 01:12(EDT)
LAST UPDATE DATE: 10/22/96 NOTIFICATIONS ;
CAUDLE JULIAN, REG 2 RD-JOHN LINEHAN, NMSS EO NRC NOTIFIED BY: PETTY OFFICER ERIGHT STUART RUBIN, IRD HQ OPS OFFICER: LEIGH TROCINE AEOD REGGIE JOHNSON DOE EMERGENCY CLASS: NOT APPLICABLE BLAKE WELLING R1DO 10 CFR SECTION:
NTRA TRANSPORTATION EVENT EVENT TEXT
'; NATIONAL RESPONSE CENTER REPORTED A RELEASE OF 0.02 LITERS OF
.LLIUM-201 FROM A FEDERAL EXPRESS PACKAGE IN MEMPHIS, TENNESSEE.
A FEDERAL EXPRESS PACKAGE BEING TRANSPORTED FROM DUPONT IN MASSACHUSETTS, ,
EXPRESS TRUCK ON A RAMP AT A FEDERAL EXPRESS HUB LOCATED TENNESSEE. THE BOX WAS BREECHED, ,
MILLILITERS) AND APPROXIMATELY 0.02 LITERS (19.6 THE ACTIVITY OF THE THALLIUM-201 WAS 725 MEGABECQUERr.L i UP, AND THERE THE EVENT. WERE NO PERSONNEL CONTAMINATIONS OR INJURIES AS FEDERAL EXPRESS NOTIFIED DUPONT OF THE INCIDENT.
CENTER ENVIRONMENTAL NOTIFIED PROTECTIONTHE ENVIRONMENTAL AGENCY REGION 4, PROTECTION AGENCY , THE RADIATI (REFER TO THE HOO LOG FOR A FEDERAL EXPRESS CONTACT AND ADDRESS.)
NAME,AND THE DEPARTME TELEPHONE NUMBER, I
l l
October 25, 1996 NOTE T0: Bruce Mallett Doug Collins FROM: EdMcAlpind'M
SUBJECT:
WEEKLY November STATUS 1, 1996 REPORT AND ACTIVITIES PLANNED FOR OCTOBER 28 to 1.
Planned Insoections and Sionificant Licensino Actions for Next Week Inspector Licensee Target Areas l
None scheduled II. Trainina to be Conducted None scheduled.
l III. Other Sionificant Activities Planned for Next Week Continued preparation for B&W NNFD Licensee Performance Review.
- Preparation of briefing material to support management meeting at NFS on November 15, 1996. This meeting has been rescheduled from October 29.
Continued effort will be placed on closure of allegation cases.
Conduct performance appraisal discussions with branch staff.
Attend meeting with NMSS on the 1997 f scal year inspection schedule on October'28 in Rockville.
IV. Sionificant Accomolishments from This Week Briefing package for Commissioner Rogers' visit to Westinghouse on November 6,1996 was finalized and submitted to the ED0's office
- following coordination with NHSS.
Interviews for the Senior Resident Inspector position at Nuclear Fuel Services have been completed and the proposed selection has-been submitted to the EDO for approval.
Continued inspection report and allegation followup documentation.
Completed writing of performance apprais&ls for branch staff.
Input to NHSS on Sapphire material accountability issue was completed.
Revision of the 1997 fiscal year inspection schedule was completed.
3- Observation of a force-on force exercise at B&W NNFD.
7.-. . - -
October 25, 1996 NOTE T0: Bruce Mallett Doug Collins FROM: C. Hosey 4
SUBJECT:
MATERIALS OCTOBER 28-NovemberLICENSING 1, 1996 / INSPECTION BRANCH 1 ACTIVITIES FOR WEEK OF i
- 1. Planned Inspections for Next Week Parker- Virginia Newport News General Hospital Virginia Geotechnical Bear Island Paper Froehling & Robertson Molins Richmond NS Savannah Other Insoection Activities Issue inspection schedules for November and December,1996.
Complete closeout letter for Veterinary Imaging allegation.
Continue review of materials data in the IFS system and close out items in IFS system that have already been inspected and closed.
II.
' Sionificant Licensina Actions for Next Week Complete review and issue licensing actions for five oldest cases assigned _ to MLIB1 III. Trainirg J. Pelchat to attend Effective Communications Course in region II.-
IV.
Other Stanificant Activities Planned for Next Week Complete giving performance appraisals.
Issue materials licensing and inspection status report for September.
V.
Siunificant Accomclishments from This 6!eek Issued escalated enforcement action for Pensacola Testing.
EventsReported hone s
f
October 25,1998 NOTE TO: Bruce Mellett DM: John Potter
SUBJECT:
MUS2 ACTIVITIE!i FOR CT 28 NOV 1,1996 1, inspection Activities
",A
_ Uoensee -. . Focus J. Henson Abbott Health Products. Vega Alts PR Info Not1/ Serv Bulletins J. Diaz Velez Baxter Healthcare, Albonito, PR info Noti / Serv Bulletins
- 11. Itaining e Acceptance Review Letter for Ucensing Actions (P&G Dir PG 1221 e Allegation Processing Forms and Procedures e IMPLP Followup items, Ucensing Review Documentation, etc.
lit. .Qlht!
9 Finalize Order Prohibiting involvement to J. Maas, formerly with National Circuits, Inc. In Puerto Rico, e Continuing license reviews by D. J. Collins, reciprocity work, IMPEP Report, and response on DAITS and DAFFY ltems, e Continuing LLWDB BTP closeout survey guide review, terminated site reviews and procedures by J. Henson.
9 Finalize letter tout with NMSS to resolve VA Med Center Broad Scope fee category issues.
9 Followup on licensees who are required to notify & plan for timely decommissioning.
. Activities Durina Previous Week (10/2125/961
_ , J.' _ Scope Results Princeton Diagnostic isotopes MC2800 Uc Cond Pending Princeton, WV R.B.S.,Inc, MC2800, Uc Cond Pending White Sulfur Springs, WV Gelen of WV, Inc MC2800, Uc Cond Pending Ronceverte. WV No significant vloietions, as of Friday 10/25/96,11:00 a.m.
- Followup documentation on the escalated cases against Syncor and Diagnostic Photon, Dr. Femandez, and San Juan Coment.
9 Prepared US00E pickup requests for a 300 MCI Am-241 source that cannot be buried at Bamwell, SC and a 50 mci St 90 source that the Puerto Rico Health Department has stored at our request.
9 Prepared another version of replies to certain VA Med Ctts who are interested in teaming hcw to convert their broad scope licenses into limited scope Econses with a lower annual fee.
e issued Syncor de Puerto Rico license, closing uncertainty.pver the controlling interest.
9 Prepared letter to Secretary, Puerto Rico Health Dept and Dean of Medicine. University of Puerto Rico requesting that they conduct loragterm followup studies of 25 misadministered St 90 eye applicator patients.
O Revised AO Report on the Dr. Fernandez misadministrations.
9 Discussed more efficient method for broadcasting event reports with Fred Coombs, using existing technology.
s:tdrss\mlibwk45.jpp
NOTE TO:'
Bruce Mallett, Director, DNHS Doug Collins, Deputy Director, DNMS FROM:
Richard L. Woodruff
SUBJECT:
WEEXLY ACTIVITIES FOR OCT 28-NOV 1, 1996 I.
Elanned Activities for Next Week The completed referrals and close out of allegations will be as appropriate.
The Draft Louisiana IMPEP report will be completed .
A Tennessee State Inspector will be accompanied in the Nashtrille Area in preparation for the December 1996 IMPEP review of the Tennessee program.
II.
Other Gianific3nt Aareement State Issues l
States in Region II during the week.No significant issues we III.
Sionificant Accomolishments from this Week Agreement States. Continued work on the referral of allegations to Provided technical assistance as needed to Region II States.
Worked on the Draft Louisiana IMPEP report .
file: DNMS\BR_ACT96.298
I i
November 13, 1996 I
General Electric Company .
ATTN: Mr. C. P. Kipo, General Manager
- GE Nuclear Energy Production P. O. Box 780 2 Wilmington, NC 28402
SUBJECT:
NRC INSPECTICM REPORT N0. 70 1113/96 11
Dear Mr. Kipp:
This refers to the inspection conducted on October 14 18, 1996, at the Wilmington facility. The aurpose of the inspection-was to determine whether activities authorized by t1e license were conducted safelv and in accordance with NRC requirements.
< At the conclusion of.the inspection, the findings were
. discussed with those members of your staff identified in the report.
Areas examined during the inspection are identified in the report. Withi.n
-these areas, the inspection consisted of selective examinations of procedures ,
and representative records, interviews with personnel, and observation of activities in progress.
Within the scope of the inspection violations or deviations were not identifiea.
r In accordance with 10 CFR '2.790, of the NRC's " Rules of Practice," a copy of this letter will be placed in the NRC Public Document Room.
Should you have any questions concerning this letter, please contact us.
Sincerely.
Original signed by:
Edward J. McAlpine Chief Fuel Facilities Branch Division of Nuclear Materials Safety Docket No. 70 1113 License No. SNM 1097-
Enclosure:
.NRC Inspection Report cc w/ encl: (See page 2) cg%,g
.s .
GE 2
cc w/ encl:
Dr. Ralph Reda, Manager Fuels and Facility Licensing General Electric Company P. D. Box 780, Mail Code J26-
'Wilmington, NC 28402 Dayne H. Brown Director Division of Radiation Protection- .
N.~C. Department of Environment, '
Health & Natural Resources P.-0. Box 27687 '
Raleigh, NC- 27611 7687 pj.sitibution w/ encl:
' E. McAlpine, RII G. Troup
! -G. Shear RIII
-C. Cain. RIV
!- F. Wenslawski, RIV E :
PUBLIC Distribytion w/o encl:
License Fee Management Branch l
M
- M rter s't rws
$1GNAtuRE g.
het Girose DAtt 11 / / 3 / 96 11 / / 96 11 / 't 96 H/ / 96 ' 11 / / 96 11 /
C0Pf7 (ffO NO 115 h0 Yt1 NO 115 ho
/ 96 VES no urriU AL hr,w nD LUFT. LWlDi W4; P; WCfoil .6Li its NO 1
0 U. S. NUCLEAR REGULATORY COMMISSION REGION II j
Docket No.: 70 1113 License No.: SNH 1097 Report No.: 70 1113/96 11 Licensee: General Electric Company Facility: Nuclear Energy Production location: Wilmington North Carolina 1
l l Dates: October 14 18, 1996 '
- Inspector
- G. L. Troup i ,
Sr. Fuel Facility Inspector Approved by: E. J. McAlpine. Chief Fuel Facilities Branch Division of Nuclear Materials Safety Enclosure DU{e~ .
{TGTt6% r2_5 [g ., ,
EXECUTIVE
SUMMARY
General Electric Nuclear Energy Production NRC Inspection Report 70 1113/96 11 This routine inspection was focused on operations, functional testing.
facility changes, status of the new Dry Conversion Facility (DCF) and review l of events reported to the NRC. The inspection was focused on the safe operation of the fac'lity.
Plant Ooerations The licensee has completed the modification of Line 5 in accordance with procedural requirements for change control, nuclear safety analysis and functional testing. Controls have implemented for safe operation of the system.
The licensee had implemented an effective program to control the keys for Active Engineered Controls (AECs), but noted that while the requirement for key control was specified in a Nuclear Safety Requirements / Release (NSR/R) a memo, there was no procedure which specified how the program was to be implemented. ~
Declaring the criticality alarm inoperable system based on the failure of the Public Address (PA) system was a conservative but reasonable approach.
The action to up date the drawings of the alarm system and place it in the configuration control system is a positive effort.
e Providing technicians performing Nuclear Criticality Safety (NCS) measurements with the out of specification conditions for tests for prompt notification of cogtizant personnel strengthened the process.
The bulging of the sump was not adequately addressed when first identified in 1987 or properly reviewed again when identified in 1992.
Manaaement Oraanization and Control Persons appointed to positions described in the license application met the educational and experience requirements. Other persons were qualified for the position.
Maintenance / Surveillance The fixed neutron absorbers were properly evaluated with acceptacle results. Test procedures will be revised and approved based on the new techniques.
New power sifters were installed in the grounding system of outside criticality monitors.
1 a
l l
DCF Pro.iect The major process components of the process systems have been installed.
Supporting piping and instrumentation systems have yet to be installed.
Resolution of identified deficiencies in the process control system software has not been officially confirmed.
Staffing for the DCF is behind schedule. Completion of classroom training and completed. readiness for testing will have to be satisfactorily Within the scope of the inspection, one non cited violation (NCV) and one new Inspector Followup Item (IFI) were identified.
Attachments:
Partial List of Persens Contacted Inspection Procedures Used List of Items Opened, Closed and Discussed List of Acronyms '
REPORT DETAR S Summary of Plant Status This report covers a one week period. During the period, fuel production operattons were running normally. PowJer production was shut down for about two days during the period due to a problem with the water supply to the process systems.
Construction progressing.
activities for the new powder production facility (DCF) were No other NRC inspections occurred during the period.
I. Safety Doerations 03 Plant Operations (88020) 03.02 Facility Modifications and Confiouration Controls
- a. Inspection Scope The inspector reviewed the documentation and modifications associated with up grading Line 5 to process hie materials from the Uranium Conversion (UCON) skid,
- b. Observations and Findings Various plant systems are qualified by Nuclear Safety to process uranium with a nominal maximum enrichment of 4% ("LoE") while.
others are qualified to process uranium with a nominal maximum enrichment of 5% (" hie"). Line 5 consists'of a defluorinator. and associated equipment for processing ammonium diuranate aroduced by processing uranyl nitrate from Uranium Recovery in the JCON skid and producing Uranium Dioxide (U0 powder. Uranium Recovery and UCON were qualified for hie but th)e Line 5 ' system was qualified only for LoE. A project was initiated to ua grade Line 5 to
-handle hie material. A change request had m en approved to accomplish the up grade.
The inspector reviewed the documentation for the up grade of Line 5 and ins modifications.pected (walk 5d down) various portions of the The process description and safety descr13 tion were incorporated into Technical Report No. 3.10.1. hie UC04 Process. Revision (Rev.) 1. dated September, 1995. New AECs were documented in Technical Report No. ADU AEC 1. Chemical Area AECs. Rev. O. The inspector selected 10 AECs from the list and compared the described function with the Technical Report and the Criticality Safety Analysis (CSA). Descriptions and-functions were all consistent. The inspector also-reviewed the associated Functional Test Instructions (FTIs) and determined that the tests confirmed
2 i
proper operation of the controls. The inspector reviewed the o
completed FTI reports and determined that controls performed as {
required.
Two uranium monitors have been installed in the system to measure the uranium concentration in the solution and maintain a safe mass in designated tanks. The inspector reviewed the test and
>: calibration records for the monitors and also discussed the '
modifications made to the sample chamber for uranium concentretion measurement. '
i Using the flow diagram in the Technical Report and the AEC list, '
the inspector walked down the system and visually confirmed that the equipment had been installed and the flow path was as shown.
The inspector also observed that acid flush and drain valves on the monitors were locked in the " closed" position so that positive operator action is required to perform that operation,
- c. Conclusions l
The licensee has completed the modification 6f Line 5 in accordance with procedural requirements for change control, nuclear safety analysis, and functional testing. Controls have implemented for safe operation of the system.
03.03 Imolementation of Proegss Safety Controls ,
- a. Inspection Scope The inspection consisted of a review of the licensee's program for the control of keys and locks which are used to implement positive control of AECs.
- b. Observations and Findings A number of NSR/Rs contain the requirement that valves or equipment be locked and the keys be controlled by Radiation Protection (RP).- In Uranium Recovery Unit (URU) and on Line S.
examples are the acid flush valves for uranium monitors. Another exanple is the lock installed on the timer chamber on pellet presses. The timer require that pellets be removed in a specified period or the press shuts down.
The inspector asked several- o erators in production areis how to
-obtain the keys for various 1 cks installed on pieces of equipment designated as AECs. The response in each case was that only the Shift Technical Resource (STR) could obtain the key and unlock the equipment. The ins Coordinators (ACs).pector
.and RP discussed the situation with STRs, Area controlled. In each case, personnel as to how the keys were the individual stated that the STR could obtain the key from the radiation protection office after signing out the key, and unlock the equipment for work, then lock
3 it again and return the key to the RP office. The STR logs the key in and RP signs the log that the key was returned. In many cases, one key will o>en several locks in a particular process area, because the loc (s must be opened to accomplish a particular system alignment. The inspector concluded that this is an acceptable practice so long as control is maintained over the keys.
The inspector also discussed the situation with Nuclear Safety personnel who described the same procedure for cbtaining the key.
They also discussed the situation for auditing the control of the keys to maintain control. Certain keys can only be obtained by Nuclear Safety personnel or by designated maintenance personnel.
During discussion of the key control program, licensee personnel informed the inspector of a condition which had been identified during an internal audit. Locks had been obtained from stores which were
- keyed same" such that the keys with any lock in a carton would o>en all of the locks in the carton. This case was '
i documented in 'Jnusual Incident Report (UIR) Chemical Product Line l (ChPL) 9659. A number of locks were replaced and the procurement policy was changed to permit establishment of control over keys for locks installed on AECs.
, c. Conclusions The inspector determined that the licensee had implemented an effective program to control the keys for AECs. but noted that while the requirement for key control was specified in a NSR/R a memo, there was no procedure which specified how the program was to be implemented. Licensee personnel acknowledged this comment and stated that the key control program would be formalized in a procedure, such as a Nuclear Safety Instruction (NSI).
03.07 Other Ooerational Issues
- a. Inspection Scope The inspector reviewed two events which had been reported to the NRC since the last inspection. The ins)ection included a review of the circumstances of the event and tie corrective actions taken. Events reviewed were: (1) inoperable criticality warning F system horns (NRC Event Number (EN) 30853), and (2) loss of-geometry control in a slab tank (NRC EN 31113),
- b. Observations-and Findings (1) Inoperable horns Numerous horns are located throughout the plant areas which serve both the PA voice system and the criticality alarm system. On August 8, 1996, three of the horns were
4 -
I determined to be inoperable when a message was announced over the PA system. Operations involving Special Nuclear Material (SNH) were suspended, including movement of material, and personnel were subsequently removed from the i affected areas. Subsequent testing showed that three additional horns were inoperable due to the failure of amplifiers in the circuitry. As these horns are aart of the ;
criticality alarm system, the licensee declared t1e system !
inoperable and reported the condition under the requirement of 10 CFR 70,50(b)(2) when accident mitigation equipment is required and fails, even though the criticality alarm system was checked to determine if the horns responded. The required 30 day written report was submitted on September 4, 1996.
The investigation revealed that in addition to the amplifier failure, three horns were found to be defective or clogged with foreign material. One amplifier was re repaired and the three horns were replaced.The placed, one original incident report, UIR 9649 ChPL, indicated the need for a formal investigation.
- The investigation revealed that the schematic drawings for l
the system was not current. There were horns wired into parts of the system which were under different management teams. There was also a question about the ability to hear horns in adjacent areas if a horn was not functioning. The corrective actions included testing the horns for audibility in adjacent, u> dating and controlling the system drawings and changing tie emergency response procedures to address the loss of horn audibility. These corrective actions are scheduled for completion by January,1997. IFI 96 11 01 is opened to follow up on the long term corrective actions.
(2) Loss of geometry control During the plant shutdown period, various tanks were measured using ultrasound techniques to determine if internal dimensions conformed to NCS requirements as specified in the NSR/R. Test results for the URV Exhaust Scrubber Sump (T 965) showed a bowing-or bulging at the bottom, with the dimensions of 4.59 inches up to 4.86 inches, compared to the NCS limit of 4.45 inches. Once this condition was evaluated by Nuclear Safety, it was identified as a loss of geometry control and reported to the NRC in
-accordance with NRC Bulletin 91 01.
Temporary measures were taken to push the sump walls back to the required dimension and retain that condition. A review of sampling records of the sump showed that the concentration of uranium in the liquid in the sump was
s.
5 significantly lower than the NCS limit. The' limit is 16,000 parts per million (ppm) while the highest sample result in the last thirty months was 0,7 ppm. ;
i The licensee's investigation revealed that when the sump was- l initially installed in 1987 and filled with wate, the sides bulged due to inadequate design (thin walls with no external support), The resolution was the installation two rows of constraints to >ull the upper half in and hold it. The dimensions of tie resulting space apparently were not documented nor was the dimension of the bottom of the tank checked. In 1992, the sump dimension was measured with ultrasound techniques and shown to exceed the 4.45 inch requirement in several locations but no record could be found that this had been evaluated by Nuclear Safety.
The CSA for the sump was calculated for the actual
' dimensions of the sump as shown to be safe for up to a five inch dimension. The original CSA had been based on an infinite length rather than the actual the CSA determined that the bulges were, length. Even acceptable, it wasthough decided to maintain the original _ dimensional requirement.
Additional constraints will be installed in the sump to maintain the dimensions. [ Record note this was accomplished by Oc_tober 30.]
Part I, Chapter 4. Section 4.2.4.3 of the license
[ application states that when criticality control is directly dependent on the integrity of a structure used to retain the geometric form of a fissile material accumulation, the structure shall be designed with an adequate ~ strength factor to assure against failure undar foreseeable loads or accident conditions. The design of the sump was inadequate in that it bulged and deformed under normal operating loads and thus did not meet the license criterion or the criterion of NSR/R 02.01.06. While the inadequate design is a-violation, the operating conditions were such (all inputs to the exhaust ventilation system had been previously processed resulting in uranium concentrations significantly below the limit) that this is of minor significance and is being treatej as an NCV consistent with Section IV of the NRC Enforcement Policy (NCV 96 11 02).
In reviewing the-licensee's files on this condition, the
< inspector noted that over a month passed between the time that the measurements were made (August 28) and the report was evaluated by Nuclear Safety (October 7). The technicians were told which tanks and sumps to measure but were not given an action limit or "go no go" limit which would require an immediate notification if the measurement exceeded the limit. Consequently, there was nothing which caused any concern until the report was reviewed during the
3 6
course of normal work. Licensee represent 1tives acknowledged this comment and stated that in the future, the technicians would be provided with a limit and directed to report immediately if the limit was exceeded.
- c. Conclusions Declaring the criticality alarm system inoperable based on the failure of the PA system was a conservative but reasonable approach.
The action to up date the drawings of the alarm system and place it in the configuration control system is a positive effort.
Providing technicians performing NCS measurements with the out of-specification conditions for tests for prompt notification of i
cognizant personnel strengthened the process, l
The bulging of the sump was not adequately addressed when first -
identified in 1987 or properly reviewed again when identified in 1992,
. One IFI and one NCV were identified. i 05 Hanaaement Oraanization and Controls (88005) 05'01 Oraanizational Structure
- a. Inspection Scope The inspector reviewed changes in the organization since the last-inspection, especiall license application, y as they apply to positions described -in the
- b. Observations and Findings The former General Manager; Nuclear Fuel moved to a new positinn reporting to the Vice Presidelit and General Manager, General Electric Nuclear Energy (GE NE). The former President, Reuter-Stokes was appointed' General Manager; Nuclear Fuel.
The former Manager Environmental Health and Safety (EH&S) was appointed to a position in a different part of GE. The Manager, GE NE Quality is-acting in the Manager. EH&S on a temporary basis pending selection of a replacement.
On 0ctober-18, 1996, a new Manager, Nuclear Safety was appointed.
This individual had been acting in that position. At the same time, a functional manager for the radiation safety function was
- appointed, who will re) ort to the Manager, Nuclear Safety. The inspector determined tlat the Manager. Nuclear Safety meets the qualifications for the criticality safety function and the
7 functional manager meets the qdifitations for the radiation j safety function as specified in Part I. Chapter 2. Section 2.5.3 of the license application. The license application pertits division of the functional area management.
- c. Conclusions l
Persons appointed to positions described in the license application met the educational and experience requirements. Other persons were qu211fied for the position.
IV. Facility Suncort F1 Maintenance / Surveillance (88025)
F1.06 Sg_ysillance Testina
- a. Inspection Scope The inspector reviewed the licensee's actions to comply with the license requirement to verify the integrity of fixed neutron absorbers on a seriodic schedule. The replacement of power suppresors in tie grounding system of criticality moritors was also verified.
- b. Observations and Findings l
- Part I, Chapter 4. Section 4.2.4.4.1 of the license application requires that the integrity of fixed neutron absorber systems must be verified on a periodic schedule. The licensee has determined that a two year cycle is a)propriate for the neutron absorber panels associated with tan (s in URU and Chemical Conversion. This requirement is specified in the applicable NSR/Rs.
Testing of the fixed neutron absorber systems (poison panels) was performed during the shutdown period in August September. A contractor had performed the last test and their approvM procedure was available. Some changes were necessary because some of the poison panels had been repositioned during structural changes in the external supports and the neutron source used had was stronger than that specified by the contractor. These changes were marked in the procedure as field changes. The licensee also increased the number of standards checks for the source and detector in the field, which made the measurements more reliable than the vendor's method. A licensee representative informed the inspector that these changes would be incorporated into the procedure and approved as a revision.
w _
8 The final test report was still being arepared during the inspection. The inspector discussed t1e results and reviewed some of the field data for different tanks. The licensee's test results determined that the effectiveness of the panels had not decreased and all met the minimum safety specifications. 1 Several of the criticality monitors installed in outside areas had experienced power surges resulting in false alarms during electrical storms (IR 70 1113/96 08. Paragraph 1.2).
the system had identified that the surge suppressors (A review of "aower sifters") exceeded the vendor's recommended lifetime. )uring the shutdown period, new power sifters were installed in the six outside units. The inspector reviewed the records in the Maintenance Planning and Control database and verified that new batteries had been installed and the power sifters replaced.
- c. Conclusions The fixed neutron absorbers were properly evaluated with acceptable results.
l based on the new techniques. Test procedures will be revised and approved New power sifters were installed in the grounding system of outside criticality monitors.
V. Special Topics T1 DCF Pro.iect (88020)
T1.01 Encility Construction :
- a. Inspection Scope The inspector reviewed the installation of process equi ment and reviewed the status of the roof construction for the DC:.
- b. Observations and Findings Using the system description in Technical Report S14.1320. Process Description and Design Basis, and various P& ids, the inspector >
walked down each process line and confirmed that all major pieces of equipment for the powder production process from the vaporizers (including cold traps) through the homogenizers were installed.
Much of the inter connecting piping and service piping and associated instrumentation had not been installed as yet.
The DCF :s designed as a " moderation restricted" facility. As such, the roof of the facility has a significant role in preventing the intrusion of moderating material (water) into the facility. The installation of the roof components through the first (outer) roof membrane had previously 3een inspected. The final stages of the roof installation is the installation of the
1 l* .
9 edge seal and trim. The ins trim had not been installed.pector observed Licensee that the edge representatives seal that stated and the schedule for the final sealing had not yet been finalized but would be about two months away. Final inspection of the roof seal will be delayed.
Practices and Procedures (P/P) 12015 requires that a formal design review be conducted for nen process control systems. The process control system for the DCF was programmed by a contractor for the primary supplier. The design review by licensee arsonnel identified problems with the control of AECs and interloccs in that the computer system is able to be changed without the formal change control system required by the licensee. The resolution of this situation has not yet been defined,
- c. Conclusion The major process components of the process systems have been l installed. Supporting piping and instrumentation systems have yet to be installed.
Resolution of identified deficiencies in the process control l system software has not been officially confirmed.
T1.02 Staffino and Training
- a. Inspection Scope s
The inspector reviewed the status of staffing the DCF for operation ard the status of training of the staff,
- b. Observattu1s and Findings The original plan for designating the staff for the DCF was to be completed by September. However, a number of personnel issues delayed the selection of the staff. During the inspection period, offers were made to applicants to staff the DCF.
Training for the staff has been delayed by the identification and selection of the operating staff. Training will begin about December 1. This is due to the need to select and train personnel to replace persons selected for the DCF, The completion of classroom and on.the. job training required to support functional testing is still being scheduled.
- c. Conclusions Staffing for the DCF is behind schedule. Completion of classroom training and readiness for testing will have to be satisfactorily completed.
~
e
=
l l
10 VI. h ement Heetinas H1 Exit Interview Summary On October 18, 1996, the inspection scope and findings were summarized with licensee representatives. The inspector discussed in detail the areas inspected, the findings and concerns which had been identified.
There were no dissenting comments expressed by licensee representatives.
4 e
ATTACHMENT PARTIAL LIST OF PERSONS C04TACTED LicenseePersonnel,
- H. Chilton. Hahager Joint Conversion Project
- T. Flaherty. Start up Manager. JCP
- R. Foleck. Sr. Licensing Specialist
- R. Keenan Manager. Site Security & Emergency Preparedness
- J. Kline Manager Chemical Product Line- 3 D. Landry Leader. Technology Team !
- S. Hurray. Team Leader Chemical Conversion
- L. Paulson Manager Nuclear Safety Engineering
- L. Quintana, Mana
- R. Reda, Hanager.ger.-FuelsFuel and Fabrication Product Line Facility Licensing
- G. Smith. Team Leader. Fuel Manufacturing Operation (FHO) Maintenance Support C. Tarrer. Team Leader Configuration Management & Integrated Safety Analysis (ISA)
- K Theriault. Team Leader. URU
- C Vaughan, Program Manager. Dry Conversion Product (OCP) EH&S and Regulatory
- C. Williams. Team Leader. Waste Treatment *
- P. Winslow. Manager. Material Contrcl and Accountability (HC&A)
- Attended exit interview on October 18, 1996.
INSPECTION PROCEDURES USED IP 88005 Hanagement Organization and Control IP 88020 Operations Review --
IP 88025 Maintenance / Surveillance Testing
-L:ST OF ITEMS OPENED. CLOSED, AND DISCUSSED Opened Item Number Iygg Descriotion and Discussion 96 11 01 IFI Fol'ow up on long term' corrective actions for criticality system warning horns 96 11 02 NCY lack of structural integrity to maintain a nuclear safety limit llgitd None Qiscussgd None
l '
LIST OF ACRONYMS AC Area Coordihator AEC Active Engineered Control CFR Code of Federal Regulations ChPL Chemical Product Line CSA Criticality Safety Analysis DCF Dry Conversion Facility DCP Dry Conversion Project EH&S Environmental EN Event Number , Health & Safety FHO Fuel Manufacturing Operation FT1-Functional Test Instruction-GE NE General Electric -Nuclear Energy I IFI Inspector Follow up Item l IP- Inspection Procedure ISA Integrated Safety Analysis i HC&A Material Control & Accountability NCS Nuclear Criticality Safety NCV Non Cited Violation NRC Nuclear Regulatory Commission NSI Nuclear Safety Instruction NSR/R Nuclear Safety Requirements /Releat.e -
PA Public Address PPM Parts Per Hillion P/P Practices & Procedures
- Rev. Revision RP Radiation Protection SNH Sweial Nuclear Material STR S11ft Technical Resource UCON Uranium Conversion 01P. Unusual Incident Report
-00 Uranium Dioxide 1 Rb -Uranium Recovery Unit
d .
1 l
l ONE LINERS FOR JANUARY 3. 1997 l
l 1. Fuel Facility status
- a. General Electric - Nuclear Enerov Production All calciner units'except unit #3 were shut down for the holiday schedule. Maintenance activities including tube flip was completed on calciner unit #5 on December 29. Unit #3 is currently processing virgih material, unit #5 is processing recycle material.
The Uranium Recovery Unit is currently processing waste stream materials. Dissolution and solvent extraction operations are expected to resume-the week of January 6. Fuel assembly production is also expected to resume full production during the
, week of January 6.
Line 1 of the new dry conversion process is expected to be completed and released by the construction contractor the week of January 6. Lines 2 and-3 are expected to be completed in February.
- b. B&W Naval Nuclear Fuels Olvision Operations have restarted from the recent holiday shutdown.
Recovery operations are proceeding in processing the zero power fuel material and should be completed January 14.
The Sapphire material downblending operation is continuing.
Development runs are underway in the Compact Reactor Fuel operations.
UNH conversion operations are proceeding with precipitation and sintering for HIFER.
- c. Framatone Coacma Fuel Co.
Routine rod loading, fuel assembly and bundling are being accomplished on the day shift only. The licensee anticipates the return of the assembly which was inadvertently shipped to Germany at 10 a.m. this morning to the Lynchburg facility. On December 20, 1996, the licensee furnished ONMSS the route cause analysis of the shipping error, and is under review, 1
I One Liners 2
- d. Westinahouse The facility was shutdown from Christmas through New Years but most people were back at work on January 2. Only minor maintenance and some roof repair work was done during the holidays.
The licensee was beginning plant startup on January 2 and will begin processing UF, during the evening shif t on Lines 2, 3, and
- 4. No problems were noted or reported during the past two week period.
- 2. Jose L. Fernander. M.D. (San Juan /Mavaa0ez. puerto Rico)
The NRC requested the aid of the Puerto Rico Oepartment of Health (PRDH) l in the patient medical followup of misadministrations caused by the licensee. On December 12,_1996, Ms. Feliciano, Secretary of the PRDH, notifled the NRC about the Deprtment's openness to discuss the situation and willingness to " help in any way in which we agree is
. beneficial to the patients and falls under our jurisdiction." Region 11 is drafting a response to the letter, outlining potential followup actions. A followu) telephone call will be conducted'next week to further actions wit 1 the PRDH, and to suggest a meeting to discuss approaches.
On October 21, 1996, the NRC issued an Order to the licensee to obtain a consultant to identify misadministrations. Dr. Fern 6ndez has submitted the credentials of two possible consultants, for NRC review. A followup telephone call will be conducted next week to obtain the final decision by Dr. Fernandez on the consultant selected. Based on the results,_the NRC Will issue a letter to confirm the selection.
- 3. Jaca and Sierra festina Labs (Carrairo Truiillo Alto. PR)
On December 31, 1996, the licensee reported that a car containing a portable moisture density gauge had been stolen. When the car was found, it had been stripped and the gauge was gone. The licensee notified local law enforcement, TV, and radio. The gauge was recovered on December 31 with the source still secured.
- 4. Aareement State Events Pe-238 Pacemaker in Florida On December 28, 1996, Grandview Hospital (Dayton, Ohio) reported that an individual with a Pu-238 pacemaker had died while visiting West Palm Beach, Florida. The family had initially wanted the pacemaker removed by the funeral home before return of the body to Ohio. Eventually, the body was returned _to Ohio with the pacemaker, where the pacemaker will be explanted by Grandview Hospital.-
e .
?
i QNE LINE_R_S.. EOR JANUARY 17.1992
- 1. Fuel Facility Statua
! a. General Electric Nuclear Enerov Production Powder production units 1 through 4 are currently processing {
virgin material, unit #5 is processing recycle material. Fuel asserbly production is already behind schedule for 1997 and will be operating 7 days / week for the near term to reduce backlog. No problems have been noted with operation of the calciner units since the events that occurred in early December 1996.
Approximately 500 3 gallon cans of 3.95% enriched U0 is expected i to crrive from Kazakstan the week of January 20. Malertal l characterization studies will be performed to develop a processing strategy.
I Line 1 of the new integrrted dry conversion process (10P) was completed and released by the construction contractor the week of January 6. The associated warehouse and sand blast / painting facility was also released. Uranium storage in the new warehouse is expected to begin the week of January 20. 10P lines 2 and 3 are expected to be completed in February. Region 11 will be ins)ecting the licensee's preoperational testing of these systems in rebruary March 1997.
Testing of the modifications to the URALS process is proceeding.
Two filter cakes have been produced from sludge in the West lagoon
, and are undergoing analysis. Also, the licer.see's new decontamination and volume reduction unit is expected to be operational the week of January 27.
- b. Westinchouse The licensee began plant startua on January 2 after the holiday shutdown and operated through t ae weekend of January 3 4 with lines 2 through 4 processing UFs. Conversion line #1 was operational after undergoing major refurbishment and started processing uranyl nitrate hexahydrate (UNH). Line 1 was switched to UF6 processing on January 14. No problems were noted or reported upon startup or during the past two week period.
- c. B&W Naval Nuclear Fuels Divisio_q Conventional fuel manufacturing operations are being conducted on a 4 day 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> per day schedule. Testing is being conducted after modifications to Modified Fuel Process (MFP) equi) ment. MFP personnel are working a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day schedule. RTRFE (research reactor fuel production) is operating on a normal 4 day. 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> per day schedule.
h
l One Liners 2 The UNH conversion process, used to produce yellow cake (U 0,) for 3
use as a RTRFE feed stock (HFIR) is operating normally. Routine development work is being conducted in the Compact Reactor Fuel (CRF) facility t
The dissolution of Sapphire UBe scrap materials remains on hold pending the resolution of arocess concerns with NRC. 00E and U.S.
Enrichment Corporation (USEC). Thc downblending system is shutdown to complete modifications to the LEU drum dryer.
IAEA inspectors are scheduled to be on site January 18 21 for their routine monthly visit to monitor Sapphire downblending ;
operations. Brian Horn from H0s NMSS will accompany them.
- d. Framatome Cooema Fuel Comoany Routine fuel loading and production activities are being informed.
Bill Gloersen and Al Gooden were on site this week for a routine inspection of health physics, environmental monitoring and waste management. No significant issues have been identified.
Downloading operations from the NUKEM reactor in Germany remain shutdown pending resolution of shipping issues. A CAL which places restrictions on further processing and receipt of NUKEM fuel remains in effect. An enforcement conference regarding this issue is scheduled for January 27th. The fuel element inadvertently shi by the licensee, pped to Ge.*many has been recently received back
- e. Nuclear Fueijiervices An inspection team composed of personnel from Region 11 and Headquarters, and augmented by personnel from the Licensing Branch in Headquarters, has been on site this week to review the status of the licensee's readiness to begin recovery operations of uranyl hexahydrate from Rocky Flats. No start up have been noted thus far. problems Critical that pathswould are preclude the approval of the licensee's financial assurance plan modifications and of the licensee's program information' submitted in response to License Condition S-2. Phil Ting and Ed McAlpine are also at the site to observe the licensee's preparations review the operation with licensee management, and attend the exit on January 17. 1997.
- 2. Jose L. Fernandez. M D. (San Juan /Mayaa0ez. Puerto Rico)
In October 1996 NRC issued an Order to the licensee. which required them to obtain an independent consultant to identify and notify patients who received'overexposures due to a miscalibrated Sr-90 eye applicator. The license has 45 days to complete the review. DNMS (Region II) and NMSS
One t.iners 3 have completed the review of the proposed consultant's credentials, iMr. Heriberto Torres Castro). A letter to Dr. Fernendez to indicate NRC acceptance of Dr. Torres should be signed on January 17 or January 21. 1997.
- 3. Roberts Construction On January 13, 1997 Region 11 held an enforcement conference with the President of Roberts Construction Company. The Company had an NRC license to o) crate a moisture / density gauge containin Cesium 137 )ut the license expired January 31. 1997.gThe Americium 241 and Company was required to maintain the gauge in storage while pursuing another NRC license. The Conference was held to discuss the findings from a <
November 1996 NRC inspection at the Company's facilities in West Virginia. The findings included failure of the Company to maintain adequate security for the gauge and failure to perform leak tests of the gaugc for a two year period prior to license expiration. The licensee discussed the root causes for the a) parent violations and addressed the NRC's question of why the Company siould remain an NRC licensee in view
, of the failure to abide by NRC regulations.
- 4. Aareement State Events
, There were two reportable events, this week as follows:
North Carolina BASF lost two nickel 63 sources. This was discovered and reported on January 14, 1997.
Florida Paradyne lost a polonium 210 static eliminator.
Reported to NRC on January 16, 1997.
5, Business Process Reenaineerina (BPR)
Region 11 and NHSS are conducting a pilot of the BPR design for materials licensing. The pilot will use electronic versions of applications in a Regional Office setting. The allot is to be conducted in Region 11 the week of January 27 31, 1997. Tiere are nine confirmed entities that are participating as applicants. Next week. NMSS will conduct training for the pilot in Region II,
- 6. Integrated Materials Performance Evaluation Program (IMPEP)
The staff will orief the Commission on the status of IMPEP on January 31, 1997. B. Mallett will be presenting the Regional Assessment on IMPEP effectiveness. There is a pre-brief of Hugh Thompson on January 23, 1997.
i" ' . .s each UNITED STATES
%,, NUCLEAR REGULATORY COMMISSION AT TA, RGIA . 1 January 28, 1997 l
General Electric Company 1 ATTN: Mr. C.P. Kiap, General Manager GE Nuclear Energy Production P. O. Box 780 Wilmington, NC 28402
SUBJECT:
NRP. REGION II SPECIAL INSPECTION TEAM REPORT NO. 70 1113/96 12 AND NOTICE OF VIOLATION j
Dear Mr. Kipp:
1 On December 4 8, 1996, a special NRC team inspection was conducted at the l Wilmington plant-to review the causes and your corrective actions associated with a failure of the Line 3 calciner tube reported to the NRC on Decembe:r 3, 1096. During that inspection, the team reviewed the circumstances surrounding the tube failure, precursors to the event, your response to the event, developed an understanding of the root causes, and verified the initial corrective actions that General Electric (GE) implemented to preclude recurrence.
' The inspection team was composed of Hessrs. E. McAlpine (Team Leader), and D. Ayres of this office; and D. Stout, G. Smith, and C. Tripp of the Office of Nuclear Material Safety and Safeguards (NHSS). The enclosed copy of our report identifies creas examined by the team.
Based on the results of this inspection, the NRC has determined that three violations of NRC requirements occurred. The violations are cited in the enclosed Notice of Violation (Notice) and the circumstances surrounding them are described in detail in the re mrt. Although the violations represent problems noted in your program, w11ch contributed to the calciner tube failure, your response to the failure and corrective actions were excellent.
In accordance with 10 CFR 2.790 of the NRC's " Rules of Practice." a copy of this letter and its enclosures will be placed in the NRC Public Document Room.
Should you have any questions concerning this. letter, please contact us.
Sincerely, Bruce S. Hallett, Director Division of Nuclear Material Safety Docket No. 70 1113 License No SNH 1097
Enclosures:
(See page 2) quum .
'b GE 2
Enclosures:
- 1. Notice of Violation
- 2. NRC Inspection Report No. 79 1113/96 12
( cc w/encls:
Dr. Ralph Reda. Manager Fuels and Facility Licensing General Electric Company P.O. Box 780 Hail Code J26 Wilmington, NC 28402 Dayne H. Brown. Directer Division of Radiation Protection N.C. Department of Environment, Health & Natural Resources P.O. Box 27687 Raleigh, NC 27611 7687 4
9 I
l' l
d GE 2
Enclosures:
- 1. Notice of Violation
- 2. NRC Inspection Report )
i No. 70 1113/96 12 cc w/encis:
Dr. Ralph Reda, Manager Fuels and Facility Licensing General Electric Company P.O. Box 780, Hail Code J26 Wilmington, NC 28402
' I Dayne H. Brown, Director Division of Radiation Protection N.C. Department of Environment.
Health & Natural Resources P.O. Box 27687 Raleigh, NC 27611 7687 Distribution w/encig:
l E. McAlpine, RII -
G.;.Troup
,G. Shear. RIII C. Cain, RIV F. Wenslawski, RIV PUBLIC Distribution w/o encis:
License fee Management Branch
- see previous concurrence N rfff #f f rw=1, P!! Mws eff 4 #ff M #ff M
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CATE 01 / / 97 01 / / 97 01 / / 97 f 97 01 / / 97 jl / 31 / / 97 COPY? Yt3 NO Tis Vis
~ NO NO Vt1 No it$ NO tt$ NO Off]CIAL,RECORDCurt (AA,UFihi Akit.: S nURd5\UEso 12
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I NOTICE OF VIOLATION General Electric Company Wilmington, NC Docket No. 70 1113 License No. SNH 1097 During an NRC inspection conducted on December 4 8,1996, violations of NRC requirements were identified.
l In accordance with the
- General Statement of Policy and Procedures for NRC Enforcement Actions " NUREG 1600, the violations l are listed below: /
!' 1,
! License Condition Number (No.)-S 1 of Special Nuclear Materials (SNH)
License No. SNH 1097 requires that licensed materials be used in accordance with statements, representations, and conditions of Part I of the application dated October 23, 1987, and supplements thereto.
Part I, Chapter 4, Section 4.2, subsection 4.2.11 of License SNH 1097 re uires that, " Engineered controls detect an undesired situation and im lement corrective action without requiring human intervention.
En ineered contrtls must be... capable of performing the criticality safety purpose for which they are specified." 1 Contrary to the above, on November 30, 1996, the engineered control associated with the Line 3 calciner tube failure did not detect the undesired situation and was not capable of performing the criticality
- safety purpose specified (i.e., preventing an accumulation of greater than 25 kilograms of uranium). The Active Engineered Control (AEC) in place relied upon the total breakage of the tube and the stoppage of tube rotation at the discharge end. As a result, material accumulation within the annulus of the calciner of 38.77 kilograms exceeded the analyzed safety basis.
This is a Severity Level IV violation (Supplement VI). I
- 2. License Condition No S 1 of SNH License No. SNH 1097 requires that licensed materials be used in accordance with statements, representations, and conditions of Part I of the application dated October 23, 1987, and supplements thereto.
Part I, Chapter 4. Section 4.2, subsection 4.2.4, Paragraph 4.2.4.3 of License SNM 1097 requires that, "Whenever criticality control is directly dependent on the integrity of a structure used to retain the geometric form of a fissile material accumulation ... the structure shall be designed with an adecuate strength factor to assure against failure under foreseeable logs or accident conditions."
Contrary to tha above, on November 30, 1996, criticality safety in the calciner operations was directly dependent on the integrity of the calciner tubes and the calciner tubes in all units were not designed Enclosure 1 ty)O(w r% 4 Hod 5ah !
Notice of Violation 2 with an adequate strength factor to assure against failure under a foreseeable load. A contributor to this was that there were no formal provisions (e.g., preventive maintenance) in place to adequately assure against such failures.
l This is a Severity Level IV violation (Supplement VI).
- 3. License Condition No. S 1 of SNM License No. SNH 1097 requires that licensed materials be used in accordance with statements, representations and conditions of Part I of the application dated October 23, 1987, and supplements thereto.
Part I, Chapter 4, Section 4.2, subsection 4.2,5, Paragraph 4.2.5,1 of provide criticality safety, the mass of uraniumor(or 0 ULicense is SNM 1097 re administratively controlled based on measurement by one or more of the following tecnniques:
i
.e The mass of uranium (or U"5 or U"') 5 of the volume and the uranium (or U"isordetgrmined as the product 0.) concentration as i measured by qualified counting methods.
~
The mass of uranium (or U"5 or U"') is determined by qualified i counting methods, i 4
The total-mass or change in mass of a system is measured assuming the most reactive credible composition."
j; Contrary to the above, on November 30, 1996, mass was being administrative 1y controlled to limit input to the calciners upon tube failure and measurements were not being performed as required. The ll amount of material present in the calciner during a tube breakage was being limited by an administrative control which involved an operator removing the ADU feed tube from the calciner.
This is a Severity Level IV violation (Supplement VI).
Pursuant to the provisions of 10 CFR 2.201, General Electric Company is hereby required to submit a written statement or explanation to the U.S. Nuclear Regulatory Commission ATTN: Document Control Desk. Washington. 0.C. 20555 with a copy to the Regional Administrator, Region II, and a copy to the Chief, fuel Facilities Branch, Region II, within 30 da transmitting this Notice of Violation (Notice).ysThis of the date reply of the should beletter clearly marked as a " Reply to a Notice of Violation" and should include for each violation: (1) the reason for the violation, or, if contested the basis for 1 5
disputing the violation. (2) the corrective steps that have been taken and the results achieved.-(3) the corrective steps that will be taken to avoid further violations, and (4) the date when full compliance will be achieved. Your response may reference or include previously docketed correspondence, if the correspondence adequately addresses the required res>onse. If an adecuate reply is not received within the time specified in t11s Notice, an orcer or a
Notice of Violation 3 Demand for Information may be' issued as to why the license should not be modified, suspended, or revoked, or why such other action as may be proper should not be taken. Where good cause is shown, consideration will se given to extending the response time.
Because your res)onse will be placed in the NRC Public Document Room (PDR). to I the extent possiale, it should not include any personal privacy, aroprietary, or safeguards information so that it can be placed in the PDR witicut redaction. If personal privacy or proprietary information is necessary to '
provide an acceptable- response, then please provide a bracketed copy of your response that identifies the information that should be protected and a redacted copy of your response that deletes such information. If you request withholding of such material, you E!st s)ecifically identify the portions of your response that you seek to have withield and provide in detail the bases for your claim of withholding (e.g., explain why the disclosure of information will create an unwarranted invasion of personal privacy or provide the information required by 10 CFR 2.790(b) to support a request for withholding ,
confidential commercial or financial information). If safeguards information !
is necessary to provide an acceptable response, please provide the level of l protection described in 10 CFR 73.21. .
Dated at Atlanta, Georgia this 28th day of January,1997 ll 4
(
U. S. NUCLEAR REGULATORY COMMISSION REGION 11 Docket No.: 70 1113 License No.: SNH 1007 Report No.: 70 1113/96 12 l Licensee: General Electric Company i Facility: Nuclear Energy Production Location: Wilinington, North Carolina Dates: December 4 8, 1996 Incpectors:
C. McAlpine, Team Leader D. Ayres, Senior Radiation Specialist D. Stout, Chemical Engineer
" G. Smith, Chemical Engineer C. Tripp, Nuclear Criticality Safety Engineer I
Approved by: Bruce S. Hallett, Director (
Division of Nuclear Materials Safety Enclosure DV%
_O_yoGtio}*I MS
4 TABLE OF CONTENTS
- 1. EXECUTIVE
SUMMARY
...........,.............. 1
- 11. REPORT DETAILS ..................,, ...... 3
- 1. Purpose and Charter of the Special Inspection Team ... ..... 3
- 2. system Description and Process History ............... 3
- 3. Event Summary and Sequence of Events ................ 5 4 Precurtor Events .......................... 5
- 5. Analysis of the As Found Condition ................ 7
- 6. Potential for Greater Accumulation ................. 8
- 7. Worst Case ' safety Margin ........... . . . . . . . . . 10
- 8. Root Cause lovestigation .................. . . . . 12 i
- 9. Adequacy of Safety Controls . . . . . . . . . . . . . . . . . . . . 14
- 10. Licensee Response to the Event ..................17 111. EXIT HEETING . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 i
I. EXECUTIVE
SUMMARY
On November 30, 1996, the failure of the Line 3 calciner tube caused the shutdown vf uranium Production facility. powder production at the General Electric Nuclear Energy The affected equipment was cooled for disassembly in i i
preparation for repairs to begin. When disassembly began on December 3, 1996, an accumulation of special nuclear material was discovered outside its favorable geometry containment. The quantity of material found was in excess of the amount used in the analyzed criticality safety basis of the operation.
The NRC special inspection team concluded that precursors to the event could have alerted the licensee to the deteriorating condition of the calciner tube.
Levels of air activity from the calciner exhaust system were above action levels weeks before the total failure of the tube occurred. The licensee's stack sampling program was.to have caused the notification of responsible personnel if stac; samples exceeded these action levels, but no such notification took place.
The criticality controls in place for the calciner process included the geometry of the calciner tube, and the control of mass in the annular space outside the calciner tube. Both of these controls were defeated with a single event.
control.The-fracture of the tube itself is considered a loss of geometry I
The accumulation of material in the annulus in excess of the analyzed amount in the criticality safety basis is considered a loss of the combined mass / geometry control for the annular space in the calciner.
The NRC team concluded that the root cause for the loss of the geometry control (i.e. the calciner tube fracture itself) was primarily due to the lack '
of a formal preventive maintenance program for rolled Inconel tubes. The team concluded that the root causes for the loss of the combined geometry / mass control failure are the inadequate' design of the Active Engineered Control, and the inadequate analysis and control of the design changes made to the system.
1 Although both criticality controis were defeated with this single event, the calculations performed by the licensee and the NRC inspection team show that there was a wide margin of safety-in the as.found condition, provided that internal moderation is minimal. However, the criticality controls credited to the system in the analyzed criticality safety basis were not the reasons why the system remained subcritical. The system remained subcritical due to the limited total amount of material involved (in the tube and the annulus) and the lack of neutron moderators in and around the system. Upon further review, the inspection team concluded that the potential existed for the accumulation of material quantities.of up to about 6 times the as.found condition.
The licensee's responsiveness in reporting and reacting to the event was adequate for the scope of the problem, The operators within the control room and on the production floor responded quickly and appropriately based on the information available to them.
i e 2
! The set of short term corrective actions developed by the licensee adequately addressed the causal factors of the event by improving the monitoring, control, and robustness of the calcining operation. The long term corrective actions (once fully implemented) were adequate to provide im of the criticulity safety program for the calciner systems. proved reliability 4
s 4
3
- 11. REPORT DETAILS
- 1. Purpose and Charter of the Soecial Inspection Teg This special team inspection was conducted to evaluate the safety
! significance of an event which resulted in the loss of certain criticality safety controls in the uranium powder product line on November l 30, 1996. The NRC was notified of the event on December 3.
1996, when an accumulation of material in excess of the analyzed nuclear j
< criticality safety basis was discovered outside of the favorable geometry process equi ,
-on Occember 4, 1996. pment. The special-inspection team was established l The team was chartered to gather information and make areas:
appropriate findings and conclusions by focusing on the following 1 i
l Determining the safety margin remaining in the as found condition.
Evaluating the potential for the accumulation of U 235 material in excess of the as.found condition.
Determining the adequacy of the criticality controls in place for '
the affected process line and the generic implications for the other processing lines with similar equipment and engineered safety controls, Determining the root causes for the loss of safety controls, and Evaluating the adequacy of the licensee's response.
- 2. System Descriotion and Process History The General Electric Nuclear Energy Production facility uses six (6) rotary calciners as part of its operation to produce uranium oxide powder. The calciners consist of a rotating nickel alloy tube inside a cylindrical, refractory lined heating chamber. The calciner tubes have a nominal inside diameter of 10 inches and are 26 feet long. A 7 inch thick annular space exists inside each calciner between the tube and the heating chamber. A-set of agitator flight inserts are placed in each
!alciner tube to assist the flow of product through the system.
Five of the calciners in this product line can convert ammonium diuranate (ADU) to uranium oxide by heating it to about 750'C (1382*F).
The ADU is injected as a paste through a wand inserted 3 feet into one end of the tube. Conversion to uranium oxide occurs as the material moves down the inside of the rotating, heated tube while reacting with steam and hydrogen from dissociated ammonia. The sixth caiciner only processes recycled uranium oxide powders,
I 4
The annular space outside the tube is heated to the process temperature via combustion of natural gas. The natural gas combustion products are exhausted through a ventilation duct to an outdoor stack that is sampled for radioactivity. Calciner units 1 through 4 have their own exhaust stacks, while units 5 and 6 share a stack.
The operational history of the calciners is such that failures of <
calciner t; des occasionally occur. These failures involve the l
circumferential fracturing of the tube at the ;oint where ADU paste is injected, presumably due to stresses created w1en the cool paste contacted the heated tube. The fractures were normally smooth, complete l
i breaks such that the discharge (non drive) end of the tube would stop p rotating even though the feed (drive) end of the tube continued turning. 4 Based on this failure mechanism, a switch was placed on the discharge end of the tube to detect the stoppage of rotation, and thus a possible tube failure.
Over time, failures were frequent enough that the licensee developed an informal tube replacement schedule based on the total amount of A)U it had processed. The failure rate for the original tubes were established such that they were " flip)ed" after approximately.340,000 kilograms (kg) of accumulated ADU throug1put in order to inject the feed into the end i of the tube that had not experienced the thermal stresses. Once a
" flipped" tube had reached the prescribed accumulated throughput, it was removed from service and replaced with a now tube.
The licensee also contemplated improving the strength of the tubes. '
Records of the manufacture of the calciner tubes showed that the original design specified the construction material to be casted T 63 alloy (35% Ni, 35% Fe, 27% Cr, 0.5% C). In 1987 a design specification change was investigated by installing a prototy>e tube made from rolled and welded Inconel 600 (75% Ni, 9% Fe, 16% Cr, 0.07% C). In b.-der to relieve stresses in the rolled tube created during fabrication, the manufacturer suggested a final annealing step at an elevated temperature which was subsequently performed, The prototype tube was used in the line 3 calciner for five years without failure or "fli) ping" with an accumulated ADV throughput of niore than 922,000 kg. Tie prototype tube was removed from service for further evaluation, but it had performed so well that several more rolled and welded Inconel tubes were purchased for use in the calciners. However, the original fabricator's suggestion concerning a final annealing step was disregarded for these additional rolled tubes.
The non annealed rolled tubes were gradually placed in calciner units 3 (1992), 5 (1994), and 1 (1996). Although no re)lacement schedule had been determined for rolled tubes, the rolled tu>e in unit 3 was replaced with another rolled tube in 1994 after an accumulated ADU throughput of 541,000 kg. The original prototype was returned to service in calciner unit 4 in 1995.
5
- 3. Event Summary and Seam 1rg of Events At approximately 11:50 a.m. on November 30, 1996, the calciner Wet End (feed end) Operator notified the ADU control room o>erator that smoke was filling the calciner area and had the smell of )urning grease. The operator at the discharge end of the calciner noticed an unusual, erratic tube rotation. At 12:00 noon, the ADV feed tube was removed from the Line 3 calciner to stop the introdJCtion of uranium into the system. At 1:00 p.m., the Line 3 calciner was shut down and began cooling.
At 9:00 a.m. on December 3, 1996, the calciner had cooled enough to remove the top of its refractory lineu shell. A significant l
accumulation was observed within the annular space between the calciner tube and the bottom refractory. The material was collected and found to be 38.77 kg of uranium oxide enriched to 4.9% uranium 235. The material was removed and placed into criticality safe storage containers. The i licensee subsequently reported to the NRC Operations Center within four hours ad by 5:45 p.n... all ADV lines were shutdown pending a system-I wide investigation.
{
l The tube failure on November 30, 1996, in calciner unit 3 occurred after an accumulated through experienced with a roiput of 569,000 led Inconel tube.kg ADU. This was the first failure
- 4. Precursor Events
- a. Inspection Scope The NRC inspection team assembled at the licensee's plant site on December 5, 1997. In order to understand the conditions at the facility, the team reviewed the secuence of events, including the available precursory indicators anc the licensee's responses to them,
- b. Observations and Findings According to the licensee's records, the initial indicator that suggested a breach of the calciner tube integrity was the occurrence of elevated and gradually increasing radioactivity measured in the ventilation exhaust for the Line 3 calciner. The calcinerventilationexhaustissampgdweeklyandhasa concentration action limit of 3 x 10 pCi/ce. These measurements were originally intended solely for emissions reporting and have not normally been used to detect calciner tube failures. The radioactivity in the Line 3 calciner exhaust exceeded the licensee's action limit based on samples taken as early as November 14, 1996. Samples taken the following two weeks centinued to exceed the action limit and the radioactivity concentration approximately doubled in magnitude with each sample.
The licensee's own >rocedures require that an Action Level Investigation form >e completed upon discovery of an elevated I
l i
6 calciner exhaust stack samale. However, no such forms were initiated for any'of the t1ree elevated weekly results associated with this event until it was reported to the NRC Operations Center on December 3, 1996.
The next potential indicator of a tube breakage was the presence of a small (approx.1/2") piece of metal -lodged in the Line 3 calciner product discharge rotary valve. This was discovered on November 20, 1996. The calciner was shut down in order to remo w the lodged metal. The discharge end of the calciner was disassembled to visually inspect the internal structure of the-calciner tube. No damage to the internal structure of the tube was apparent-from the discharge end of the calciner, and the decision was made to restart the calciner on November 22, 1996.
No obvious slip) age of the tube was visible during the restart of the calciner, t1us a broken tube was not evident,
- c. Conclusions The precursors to the event could have alerted the licensee to the deteriorating condition of the calciner tube. The licensee's documented stack sampling program and license SNH 1097 both state that personnel responsible for the operation of the exhaust system are not 3 x 10'jfied if a weekly stack sample exceeds a value of
- #C1/cc. The failure to issue an Action Level Investigation form upon the discovery of an elevated calciner exhaust stack sample violated the licensee's Environmental U Protection Instruction No. 0 6.0 and is a violation'of the requirements of Chapter 5. Section 5.1, subsection 5.1.1 Paragraph 5.1.1.2 of the SNM 1097 license application. -However, this violation will not be subject to enforcement action because !
the licensee's efforts _in identifying and coriecting the violation "
meet the criteria specified in Section VII.B of the NRC Enforcement Policy. This problem is thus noted as a Non Cited Violation (NCV).
- 5. Analysis of the As Found Condition
- a. Scone The criticality safety margin of the as found condition of the material in calciner 3 was to be determined by the inspection team. This margin was defined in the team charter as the difference between the actual conditions and the condition necessary to reach criticality. This determination involved review of licensee's calculations and NRC inspection team members independently aerforming criticality safety calculations. This provided a metiod to verify that the geometry and material compositions in-the calculations were correct and that the results were reproducibic.
7 bi Observations and Findings The licensee's nuclear criticality safety staff ran a number of Keno calculations immediately after the tube rupture to determine the margin of safety. The licensee's model assumed that the calciner tube and firebox were concentric cylinders composed of I Inconel 600 and Oak Ridge concrete respectively. For conservatism in the Xeno calculations, the calciner tube was assumed to be half filled with V 03 along its entire length, and the annulus between the calciner, tube and firebox was half filled with U 0 powder. Calculations were performed such that the material $n, the annulus was assumed to either run-the entire length of the calciner or occupy a section four feet in length. The dimensions used in the calculations were found to corres 1 dimensions for the calciner tube and firebox, pond to theincrease with some drawing 1 for dimensional tolerances and an allowance for corrosion.
Several models were run corresponding to different degrees of moderator in the two regions and different reflection conditions.
The geometry and material compositions used in the model were verified by the inspection team. ,
The calculations in which the deposit ran the whole 24' length of H the firebox resulted in only a slightly greater L as the calculationswherethedepositoccusiedafourfoo,flength. The as found condition was modelled as 40 with 5% moisture in the calciner tube and dry U 0, the full len,gth of the firebox, with 3
vacuum boundary conditions. Thisresultedinaby of 0.66, it should be noted this model represents an extremely conservative calculation of the as found geometry: the calciner tube-is -
normally filled to a depth of 12* and the half filled firebox represents a maximum credible accumulation. The as found depth of material in the firebox was approximately to the bottom of the calciner tube.
The NRC inspection team also performed independent Keno calculations to verify the results of the licensee's calculations.
This independent calculation included constructing a more detailed geometrical model of the calciner and firebox as well as a less conservative and more realistic configuration of fissile material.
A geometrical model which explicitly showed the internal structure of the calciner/ firebox assembly was constructed for the independent verification, to determine whether the licensee's calculation was conservative. This more detailed model added an Inconel flange on either side of the calciner tube and included the two internal firebox baffles, or collars of smaller radius that separate the firebox into three sections. These additional
. ~
W 8
features were found to have negligible effect on the results. The maximum credible accumulation was assumed in this model to involve filling the annulus to the half way point in the first of three sections, between the aperture to the firebox and the first baffle.-
The inspection team attempted to model the as found condition as realistically as possible. The assumed composition was V3 0 with 5%moistureinsidethecalcinertube(correspendingto50,d00 pm as an upper bound on the as found moisture content) and dry U 3 outside the tube. The calciner tube was filled to a depth of "
and the annulus filled to a depth of 6%"'so that the uranium just reached the bottom of the calciner tube, along the length of the first firebox section. This resulted in a L of 0.50. This lower value is indicative of the conservatism found in the licensee's calculations,
- c. Conclusions The calculations wrformed by the licensee and the NRC inspection team shows that t1ere is a wide margin of safety in the as found condition. The geometr correct, reproducible, yand used the by the conservative.
model licensee was found to be
- 6. Potential for Greater Accumulation
- a. Scope '
The inspection team reviewed licensee equipment drawings and process parameter specifications to evaluate the potential for accumulation of material -in excess of the as found condition. The inspection team also estimated the maximum possible accumulation of material in the unfavorable geometry configuration,
- b. Observations and Findings
-In order to evaluate the potential for an accumulation of U 235 material in excess of the as found condition and to estimate the maximum accumulation of material possible within the calciner shell, the NRC inspection team reviewed operating procedures, equixnent specifications, and drawings. 'The internal dimensions of tie calciner refractory lined shell was found to be a 24" diameter cylinder, 20 feet long, installed at a slope of 1.5* from horizontal, The_ licensee. indicated that the approximate bulk density of _the discharged product from the calciner is 2 g/cc.
Since air activity data was not used by the licensee to determine a breach in tube integrity,. elevated air activity was not considered in determining a maximum accumulation potential for past events. Also, since the-tube began leaking material 2 3 weeks before its catastrophic failure, yet held together enough to continue to operate the tube rotation switch, the Active
9 l
Engineered Control was found to be unreliable and was not used by the inspection team in determining the maximum accumulation potential.
l- c. Conclusions Based on the event sequence, the process operating procedures and Active Engineered Controls in place at the time of the event, the process equipment configuration, and the product material properties, it can be concluded that the potential existed for the accumulation of U.235 in excess of the as found condition, As an extreme case, material-could have continued to slowly accumulate in the annulus until the fracture in the tube was effectively blocked by the accumulation. The fracture could not be totally blocked until the annulus was nearly full of material,
! corresponding to an accumulation of over 1500 kg of material.
However, abnormal operating conditions (pressures and temperatures) within the calciner, product quality fluctuations, and deviations in material balances would likely be evident long before such an accumulation would occur. .
A more reasonable scenario would involve an accumulation in the annulus underneath and u) to the bottom of the calciner tube and ex1 ally extending from tie ADU injection point to the first baffle d.
in the refractory. This would corresoond to an accumulation amount of 230 kg of material, or abou't six times the amount in the as.found condition.
- 7. Watst Case Safety Haroin a, Scope Once the potential for a larger accumulation of material was deemed plausible, the inspection team focused on the criticality safety margin associated with a worst case scenario, This included reviewing calculations performed by the licensee's staff for a worst case scenario, verifying that the results used by the licensee's staff were reproducible, and that the models were conservative.
The inspection team performed several-different independent calculations to verify the conservatism of the licensee's calculations. This review involved performing calculations while varying the moisture content, geometry, reflection conditions, material compositions (of both the calciner/ firebox and the uranium deposit). The results of these independent NRC calculations were compared to the results of the licensee's calculations for the most reactive credible configuration.
4
4 10
- b. Observations and Findit.gs The licensee attempted to model the worst case condition by assuming optimal moderation inside the calciner tube (assumed to be 25 w/o moisture) and 5 w/o moisture outside. When the licensee assumed that a four foot section of the firebox was filled with U3 0 and vacuum boundary conditions existed, L,, was found to be 0.02. When the entire length of the firebox was assumed to be filled L was 0.96. When a four foot section was assumed to be filled and,the system assumed to be under full water reflection.
L,,wasfoundtobe0.94. Full water reflection was imposed to ensure the calculational model accounted for all reflective and moderating materials outside the firebox.
The inspection team's calculations assuud the same geometry as was used for its as found condition calculations. The composition of the fissile material was assumed g be a homogeneous mixture of water and U 0 with a 5% 0 enrichment.
The density of the material was calc,ulated assuming the theoretical densities of water and U 0. The ammonium diuranate j a bu,lk density of 2.5 g/cm .(0 0 Several (NP ),)were iterations feed material gas assumed performed l with various moisture contents and geometrical configurations.
During normal operating conditions. where the moisture content is essentially zero and there is no tube breakage, the value for L,,
i was found to be less than 0.27. Since the calciner operates at 750'C and the ADU feed rate is 60 kg/hr. the ammonium diuranate dries almost instantaneously, reacts to form U 30,, and then reduces to UO,. The calciner takes several days to cool to sufficient temperature to remove the top of the firebox. so that significant loss of moderation control is nearly incredible. When only geometry control is lost the system was still found to have of less than 0.6, despite the conservative a value for L , filling both the calciner tube and the firebox assumptionsof halfway. Thus the system has a wide margin of safety when the moderation is minimal.
The inspection team modelled the effect of moderation by continuously varying the moisture content of the system. The licensee's assumption of optimum moderation at 25 w/o moisture was found to be valid. At optimally moderated conditions. L, for a half filled calciner for a completely filled tubetube was found was found to be to be 0.90. 0.60.
The annuand L,ius was assumed to have a moisture content of only 5 w/o in these models.
The occurrence of this amount of moderator is unlikely; a U3 0 H,0 mixture that contains St moisture by weight would equate to an accumulation of water of over 71 liters. The system was found to exceed a L, value of 0.90 only when the moistura content of the system is between 10 w/o and 55 w/o. This stresses that system reactivity is sensitive to moisture content. Clearly, in the high temperature environment of an operating calciner. this amount of l
l 11 l
moisture is unlikely. However, a calciner that has been shut down i
and cooled with its shell removed may be susceptible to higher moisture levels. It should also be noted that the area external to the firebox is maintained as a moderation control area.
In addition to testing the sensitivity of moisture content of the system to criticality safety, the composition of various fissile and structural materials were varied to determine the conservatism of the materials used in the calculations. Various different Inconel compounds were substituted for Inconel 600 from the SCALE standard materials library, which differed primarily in the amount of nickel, iron, and chromium they contained. This made a difference in the results of no more than 2 percent. Several different standard concrete compositions were substituted for the Oak Ridge concrete and a similar dependence was found. The difference between filling the calciner tube and annulus with U 30, was negligible. If the calciner tube is assumed to be or U0,d with ammonium diuranate, the results are larger to the fille order of several percent: however, the normal-high tem arature environment precludes the existence of ADV except at tle extreme feed end. y Finally, the variations in the internal agitator flights and the feed tube assemblies were modelled by replacing them with
- cylinders composed of Inconel 600 and stainless steel.
respectively, of equivalent masses. This was found to result in a slight overall reduction in L,, upon displacement of the fissile material. The geometric model was thus found to be conservative and the-addi'lon of structural material inside the tube is not expected to increase the overall (,,.
- c. Conclusions Tne licensee's calculations were verified independently by modelling the same composition of uranium and moderator mixture as the licensee. The geometry and material compositions of uranium
.and structural materials used in the problem were found to be conservative, with a wide margin of safety.provided that-internal moderation is minimal. The licensee's assumption of optimal moderation was found to be valid and conservative.
- 8. Root Cause Investiaation
- a. Scope The NRC inspectiori team investigated the root causes for the loss of criticality safety controls. The team observed equipment damage resulting from the tube failure. The team reviewed licensee policies for change control. The team discussed with the-licensee the formulation of the calculated criticality safety I
4 i
1 12 basis for the calciner operations. The team also reviewed the licensee's Taproot investigation of the causal factors of the ;
event.
- b. Observations and Pedings The team observed the tube which had been removed from calciner unit 3 and found the break to be an irregular, jagged, cog like fracture. The team also observed photogra>hs and reviewed written descriptions of prior fractures of cast tuaes. These suggested that a much smoother treak was associated with the failure of cast tubes. The team consulted with an NRC metallurgist and found that these two types of failures are consistent with the differences in
' properties of' the cast and rolled tubes due to alloy composition
- and methods of fabrication. The lack of stress relief on the rolled Incorel tube purchased after the prototype was also found l to be a likely contributing factor to the failure in unit 3.
Although the calciner rotational drive mechanism was at the feed end of the tube, the cog like " teeth" associated with the fracture of the rolled Inconel tube helped enable the discharge end of the i
tube to continue to rotate after the breaka'ge occurred. This, in turn, helped to defeat the Active Engineered Control which was
}
dependent upon the rotational switch at the discharge end of the calciner. This situation was not considered when the change in tube fabrication was evaluated.
i The team observed an agitator flight which spans the length of each tube. These inserts are fabricated in sections which have structural support rings at each end. The inserts are normally installed by tack welding the sections together as they are being placed inside the tube. The agitator flight insert sections for the tube which fractured in calciner unit 3 were found to have been welded together with a continuous bead instead of tack welded. This made the structure of the tube inserts stronger than normal. The stronger internal structure of the tube could have helped the discharge end of the calciner tube to continue rotating after the breakage occurred. The licensee's Taproot investigation states that this welding technique had not been evaluated for its impact on equipment reliability,
- c. Conclusions The cause of the material accumulation in the annulus was the ci?cumferential failure of the calciner tube at the point where relatively cool ADU paste is injected into the heated tube. The failure created a 3/81/2" wide gapp in the tube wall around its entire circumference through which the spillage occurred.
4 l
13 The root causes for the loss of criticality safety controls stem
- from the tube failure itself and the material accumulation being greater than the analyzed criticality safety basis. The licensee's Taproot investigation listed three causal factors for the reported event.
- 1. Failure of active engineered control to prevent accumulation of material in the annulus due to changes in the tube i material and design indirectly impacting the integrity of the AEC.
l
- 2. Inadequate design control and preventive maintenance for rolled Inconel tubes.
- 3. Delayed notification of the Line 3 calciner exhaust ventilation stack results.
Since calciner tut;es are now known to break regardless of whether they ne fabricated via a cast or rolled process, the loss of the
, geometry safety control can not be traced back to the tube's material of construction or method of fabrication. Instead, the cause of the tube break can be derived from~the way it was maintained while in use. An informal preventive maintenance program had been successfully implemented for the cast tubes. The NRC team concluded that the root cause for the calciner tube fracture itself was primarily due to the lack of a formal preventive maintenance program for rolled Inconel. tubes.
The root cause for the accumulation of material in excess-of the analyzed safety basis is more complex. One cause of this safety system failure is due to the way the tube failure occurred. The unexpected mode of fracture and the strengthened tube inserts kept the discharge end of the tube rotating, even after the catastrophic failure of the tube. Thus, one root cause for this_
- failure is concluded to be the inadequate analysis and control of the design changes made to the system.
However,- the loss of this second criticality control could have been prevented if the control had been better designed. The control in place relied upon the total breakage of the tube and the. stoppage of its rotation at the discharge end. The tube rotation switch by itself did not and could not ensure that the material accumulation within the annulus of the calciner was limited to less than'the analyzed safety basis. - Chapter 4
- Section 4.2, subsection 4.2.11 of the license' application requires that, " Engineered controls detect an undesired situation and implement corrective action without requiring human intervention.
Engineered controls must be...ca>able of performing the criticality safety purpose for w11ch they are specified." Despite
14 the above requirement, the engineered control did not detect the undesi,4;d situation and was not ca)able of performing its criticality safety purpose under t1e event scenario, and is thus a violation of the requirements of license SNM 1097. i
- 9. Adeouacy of Safety Controls
- a. Inspection Scope The inspection team reviewed the criticality controls in place for the Line 3 calciner process to deduce their adequacy and the generic implications of this event for the other processing lines.
b, Observations and Findings In addition to the calculated calciner criticality safety basis, the team reviewed the licensee's criticality design s)ecifications and criticality safety requirements for operation. T1e team observed the o)eration of the Active Engineered Control (tube rotation switc1) on an empty calciner and observed the passive engineered control (tube geometry) for a typical calciner. The team also used the results of their review on the history of the manufacture and maintenance of calciner tubes to assess the adequacy of the licensee's criticality controls, and reviewed the licensee's evaluations performed as a result of previous tube fractures, The dual criticality controls for the calciner operation consisted of geometry control (based on the dimensions and-integrity of the tubes) and a geometry / mass combination control based on limiting the amount of material in the annular space between the calciner tube and the refractory lined shell.
At the time of the event, the operational and automatie engineered criticality safety requirements included the following:
material enrichment limits (5.00% on lines 3, 5, and 6:
4.025% on lines 1, 2, and 4):
termination of the uranium and steam inputs to the calciner upon tube rotation stoppage, low calciner temperature, or pluggage of the discharge end for over 90 minutes:
the control system was to automatically stop the steam flow if tube rotation-stopped or if a low calciner temperature occurred (lines 3, 5, and 6 only), and automatically stop hydrolyzed uranyl fluoride-(HUF) flow to the precipitation tank (line 3 only):
the calciner tube rotation timer switch was set to indicate t be rotation stoppage if the switch was not actuated at least once per 30 seconds.
i 15 The Active Engineered Control was observed to be mountad on the discharge end of the calciner adjacent to a rotating portion of the tube. A cam was observed to be welded on the rotating equipment such that when a revolution of the discharge end of the tube was completed, a rocker arm on the switch was actuated by the cam. The licensee informed the team that the actuation of the switch would reset the system's 30 second timer. The passive engineered control (tube geometry) was observed by inspection of new and used calciner tubes, with all tubes being of smooth bore and consistent diameter.
l c. Conclusions The criticality controls in place for the calciner process included the geometry of the calciner tube, and the control. of mass in the annular space outside the calciner tube. Both of these controls were defeated with a single event. The fracture of the tube itself is considered a loss of geometry control. The accumulation of material in the annulus in excess of the 25 kg (as analyzed in the criticality safety basis) is considered a loss of the combined mass /geoinetry control for the. annular space in the calciner. .
Even though both criticality controls were defeated with this single event, the safety margin in the as found condition was adequate to ensure the system remained well into the suberitical range. However, the criticality controls credited to the system in the analyzed criticality safety basis are not the reasons why t the system remained subcritical. The system remained subcritical due to the limited total amount of material involved (in the tube and the annulus) and the lack of neutron moderators in.and around the system. -This causes the following concerns to' surface relating to the overall philosophy used to assure nuclear criticality safety:
The criticality safety basis was based on two controls-that could be defeated by a single event.
The ahilosophy of double. contingency requires at least two unli ceiv. independent and concurrent changes in process conditions to occur before a criticality accident is possible. The history of tube failures in the calciners is such'that this loss of geometry control may not be unlikely, and thus.may not be able to be credited toward the double contingency standard.
Chapter 4. Section 4.2. subsection 4.2.4. Paragraph 4.2.4.3 of the license application states that, "Whenever criticality control is directly dependent on the integrity of a structure used to retain the geometric form of a fissile material accumulation...the structure shall be designed with an adequate strength factor to assure against
16 failure undar foreseeable loads or accident conditions."
l The design of the calciner tubes was adequate to assure against failure in some cases until they have been used for a certain length of time or for a certain amount of throughput, in the case of the rolled tubes, however, the design was not adequate to assure against failure. For example, the design did not include formal provisions in place to assure against failures. Since a failure could occur and was known to occur under foreseeable loads or accident conditions, and no formal provisions were in place to detect and prevent such failures, this is a violation of License SNH 1097. Chapter 4, Section 4.2, subsection 4.2.4, Paragraph 4.2.4.3 of the license application.
The control system (tube rotation switch) was designed to automatically stop hydrolyzed uranyl fluoride (HUF) flow to the precipitation tank for the Line 3 calciner upon loss of l tube rotation or low calciner tem)erature. However, the output from the precipitation tant consisted of the ADU which was fed into ihe calciner, and there were no automatic a shut offs for this flow. In order to. limit the amount of material present in the calciner during a tube breakage, an administrative control was implemented which involved an operator. removing the ADU feed tube from the calciner.
Since this is an administrctive control of mass in support of nuclear criticality safety, the provisions of license SNM 1097. Section 4.2.5 apply. In particular, Paragraph 4.2.5.1 of license SNM 1097 states,
'"Where safety, thecontrol mass ofofuranium mass is(orused U 2 orto Uprovidg)eriticality 2
is administratively controlled based on measurement by one or more of the following techniques:
e The mass of uranium (or U* or U") is determined as23 }he product of the volume and the uranium (or U or U*) concentration as measured by qualified counting methods, o The mass of uranium (or U* or US) is determined by qualified counting methods, e The total mass or change in mass of a system is composition."
Since the mass within the calciner was under administrative control within the calciner, and mass controi was one of the parameters used in the criticality safety basis, the mass within the calciner was subject to the requirements of JWL
1,:
e i
17
' License SNM 1097, Paragraph 4.2.5.1. The required measurements were not being performed and this is a violation of Chapter 4 Section 4.2, subsection 4.2.5, Paragraph 4.2.5.1 of License SNH 1097.
- 10. Licensee Resoonse to the Event ,
a,- !
Scope
)
The inspection team evaluated the adequacy of the licensee's actions during and following the event, including their response ,
to the event and corrective actions implemented. The team reviewed the information available to the licensee to use as future indicators of a tube failure, the timeliness of reporting to the NRC, and the licensee's response to criticality and radiological safety issues, b, Observations and Findings The team reviewed documentation of the event and found that the licensee notified NRC within the four hour. time limit after discovery of the reportable condition and clearly stated the facts of the situation. The inspection team also found that the licensee clearly understood the magnitude of the situation and 4
responded with the proper amount of management involvement.
Yeam members interviewed the control room operators that were on duty at the time of the event and found that operations personnel reacted quickly to the report of smoke in the-area, found the problem unit, and shut it down, It was also found that the licensee assembled investigatory teams promptly with a mixture of technical expertise in order-to fully analyze the event.
The team was )rovided with a set of licensee corrective actions in response to t1e event.- The team thoroughly discussed the short term and long term corrective actions with licensee management.
The short term corrective actions identified were:
Tighten the tolerance of the timing on the calcir.er tube
. rotation indicator switch. The timer setting at the time of the event was nearly twice the tube rotation time. The tightening of the timer tolerance would effectively shut down the calciner if the tube rotation time was~more than
-2 seconds beyond the normal value. This would help limit the accumulation of material in the annulus in situations where the discharge end of a broken tube continues to turn erratically due .to support from the tube's agitator flight.
Automate the valving of the ADV paste feed pum) so that the administrative control of removing the feed tuae is not necessary for stopping ADU feed into the calciner. This 3-way valve was manually operated to either allow ADU paste A
'4 18 flow to the calciner or to recirculate the ADV paste in the ADU feed ho>per. Automating the valve would cause it to switch to tie recirculation mode upon indication of tube rotation stoppage.
l Implement a control system program change to correct weaknesses in the calciner stack monitoring and reporting system. This would include a arogram change that requires calciner stack information to >e input within a certain time frame. If stack results were not entered into the system at a prescribed frequency, the calciners would be automatically shut down. If stack results were abnve a prescribed limit, an Action Level Investigation would be automatically '
produced. These programming changes would require this readily available indicator to be used to detect calciner tubes in the process of failing.
Evaluate and formalize a Preventive Maintenance program for cast and rolled calciner tubes. This program would formalize an accumulated through calciner tube would be " flipped"put level at which a or replaced.
The long term corrective actions that were proposed by the licensee were also reviewed by the inspection team and discussed 4 with licensee management. They included:
Revising the Criticality Safety Analysis for the calciner operation in order to improve on the existing neutronic model of the postulated " tube break" accident condition.
This corrective action would analyze the accident condition beyond the 25 kilogram accumulation postulated in the previous analysis. This corrective action would also -
evaluate adding a new, independent mass / moderation parameter control _to the existing basis for safety.
Revising the vendor specification for Inconel 600 rolled alloy calciner tubes to include annealing (stress relieving) of rolled and welded tube. This would likely improve-the strength and lengthen the service life of calciner tubes.
Revising the licensee's change control process to emphasize the impacts of material property changes of process equipment. This would be followed by sensitivity training to all fuel manufacturing process engineers.
Evaluating new technology to determine mass accumulation inside the calciner annulus. Gamma scanning techniques are being considered, but their applicability is uncertain.
In order to verify its usefulness as an indicator of material accumulation in the calciner annulus, the-inspectors constructed a nathematical model to deduce the mass of material as a function of 1
~
l 19 the calciner stack radioactivity data. Inis model assumed that the-measured stack activity was directly )roportional to the surface area of the uranium deposit and t1e total air flow, allowing an extrapolation of mass at the times of the previous stack samples.
The uranium was assumed to be in the form of UO, with a density of 2 g/cc. The shape of the uranium deposit was assumed to be a l
normal distribution in the transverse and axial directions. The standard deviation in the transverse direction was taken as the radius of the calciner tube and the standard deviation in the axial direction was taken as one foot. The uncertainties in the actual density of the U0, and the distribution shape dominated the model systematics.
The surface area of the deposit in the annulus was computed numerically, and the volume was multiplied by the densit obtain the total mass for different sized distributions.y to This resulted in a surface area vs. mass curve that was related to the stack emissions by using the surface area corresponding to the as-found mass. The mass corresponding to the previous three stack emissions was estimated using this curve. The results shown below give the extrapolated accumulated mass for each of the three previous weeks.
Sample Stack Activi Surfage - Estimated Date (pC1/cc x 10'}y) Area (cm ) Mass (kg) 11/27/96 18.90 4600 38.77-11/21/96 7.40 1801 1.25 11/14/96 3.40 828 -0.375-11/7/96 - : 0,43 105 0.1 When the short term corrective-actions had been initiated.
ins >ection team members observed functional testing of the meclanical and electronic improvements. Team members then observed training of operators on the new procedural requirements stemming from these corrective actions,
- c. Conclusions The licensee's responsiveness in reporting and reacting to the event was adequate for the scope of the problem. The operators within the control room and on the production floor responded
_ quickly and appropriately based on the information available to
~
them.
I 20 The set of short term corrective actions adequately addressed the causal factors of the event by improving the monitoring, control, and robustness of the calcining operation. The tighter tolerance on the calciner tube rotational switch would tend to detect nearly any slippage of the tube rotation. The improvements in the-calciner exhaust stack monitoring system would help provide early indication of material leakage into the annulus. The areventive maintenance should help minimize the occurrences of tu>e fractures.
The long term corrective actions (once fully implemented) appear adequate to provide improved reliability of the criticality safety '
program for the calciner systems. The revising of the Criticality Safety Analysis should-provide a more realistic representation of the actual system parameters in ) lace for preventing a criticality accident. The revising of the c1ange control process would I provide more insight into the many faceted considerations which must be reviewed as part of any change to a system with nuclear safety concerns.
From the air activity vs. material accumulation model, it was concluded that stack sampling data could be used as ar. early warning of impending tube rupture for this mode of failure. Based on limited data, such improvements in reaction to air sampling
' data could result in detection about two weeks prior to complete !
tube failure. However the frequency of the stack sampling
(,1kly) makes it difficult to affirm that a large accumulation mid not occur between samples.
I i
a
21 III. EXIT HEETING The team leader presented the inspection results to licensee representatives by phone on January 24, 1997, and summarized the purpose of the team inspection, the inspection team charter, the inspection findings, and the three violations listed in the NOV. The licensee had no dissention to the NOV. The licensee did not indicate that the report would be considered proprietary.
4
ATTACHMENT 1.
PARTIAL' LIST OF PERSONS CONTACTED Licensee C. Kipp,-General Manager
- J. Kline, Manager, Chemical Product Line
- L. Paulson, Manager, Nuclear Safety
- S. Hurray, Team Leader, Chemical Conversion
- R. Reda, Manager Licensing
- C. Monetta, Manager, Environmental Health and Safety
-T. Hauser, Manager, Quality Assurance
- Denotes personnel present at the exit briefing conducted by phone on January 24, 1997.
EC N. Economos, Inspector Division of Reactor Projects, Region II
, INSPECTION PROCEDURES USED- .
I IP 88020 Operations Review ITEMS OPENE0; CLOSED. AND DISCUSSED Ooened 70 1113/9612 01 Violation Failure of an engineered control being capable of performing its specified criticality safety purpose.
70 1113/9612 02~
Violation Failure to provide assurance of the strength of a structure upon which criticality control is directly, dependent.
70 1113/9612 03 Violation. Failure to perform required measurement-techniques when administratively controlling mass for criticality safety purposes.
LIST OF ACRONYMS USED ADU ammonium diuranate AEC' ~ Active Engineered Control em centimeter 9/cc grams-per cubic centimeter GE General Electric
~
HUF hydrolyzed uranyl fluoride kg- kilograms NCV Non Cited Violation NOV- Notice of Violation PDR Public Document Room !
pCi/cc- microcuries per cubic centimeter
')
l l
__q
?
4 ONE LINERS FOR FEBRUARY 14. 1997
- 1. Fuel Facilities Status
- a. Babcock and Wilcox (B&W) NNFD Routine manufacturing and maintenance activities are being conducted in the Modified Fuel Process (MFP), Advanced fuel Process (AFP) and the Research and Test Reactor Fuel Element (RTRFE) portions of the plant. The uranyl nitrate hexahydrate (UNH) conversion process, used to produce U3 0 for use as RTRFE L feedstock,isshutdownduetomecnanicalproblemswithsintering ovens.
Routine development work is being conducted in the Compact l Reactor Fuel (CRF) facility.
A routine Naval Reactors recovery campaign is being conducted in the Uranium Recovery Facility. Modifications to Sapphire downblending equipment are nearing completion.
A three person team was onsite February 10 13 from headquarters for a routine inspection of the nuclear criticality safety program.
Senior lesident Inspector and a regional inspector !
coordinated their inspection with the team.
IAEA inspectors are scheduled to be onsite February 15 18 for a routine monthly visit to monitor Sapphire downblending activities.
They will be accompanied by Bryan Horn from NHSS. The dissolution of Sapphire materials remains on hold.
On February 10, 1997, Region II and NHSS staff completed the first phase of the-licensee activities at the B&W 3erformance review (LPR) for the licensed 4NFD facility near Lynchburg, Virginia. The LPR was conducted in accordance with NRC Inspection Manual Chapter No. 2604. The results will be presented to the Regiunal Administrator with a subsequent meeting with the licensee,
Routine rod loading and production activities are being performed.
Routine decontemination an'i refurbishment activities are being conducted in the SERF facilities.
A CAL preventing the further downloading of NUKEH fuel assemblies remains in effect. A second shipment of NUKEH fuel assemblies remains at a port facility. -The CAL had to do with the loss of a fuel assembly.
S. Wilkerson has been named as Manufacturing Manager replacing J.
Ford.
Three inspectors from ONMSS/FCSS were onsite February 13 and 14 for a routine inspection of the nuclear criticality safety program, p
l
9 One Liners 2
- c. General Electric Nuclear Enerav Production Several utilities (such as Niagara Mohawk Nine Mile Point) have slipped outage schedules either to later in 1997 or into 1998.
This has pushed the GE production schedules back, with the result that the plant is ahead of schedule rather than behind. Pewder production is being slowed as a result. Pellet pressing will be down for 3 days over the week end.
Two weeks ago approximately 15 tonnes of powder were quarantined because of problems with " inclusions" which resulted from a bad screen in #3 granulator. Approximately one half of this material (powder) was re screened and processed with no problems. Some pellets and rods will have to be recycled oxidized and then reduced to 00, powder.
I On February 12, while withdrawing the loading ram on the incinerator, the shaft bent. The incinerator is down until the shaft is replaced. A design review will be conducted because this shaft has bent in the past. .
The Dry Conversion Facility (DCF) system check outs are
{ progressing. Loop checks on the Line 1 instruments and controls will be completed this week. Line 2 checks may start this week.
Checks of t1e HF facility will start next week. Engineers from FBFC and the equipment manufacturer di;l be on site next week for the first heat up of the conversion kiln on Line 1. The start of functional testing has been pushed back to mid March to permi completion of reprogramming of the process control system.
- d. Nuclear Fuel Services On February 5,1997, the Regional Administrator and the Director of the Division of Nuclear Materials Safety visited the site. The purpose of the meeting was for a plant tour, operational overview, and discussion of issues of mutual interest.
On February 12, 1997, A. Maxin Vice President, Safety and Regulatory Management, was replaced by Dr. Thomas R. Baer. A.
Maxin will report to Dr. Baer (see attachment).
The licensee continues to process the HEU material shipaed from the DOE facility in Rocky Flats with no significant pro 31 ems. The licensee is preparing to receive 26 ty3e SB UF cylinders from the Portsmouth Gaseous Diffusion Plant. T1e licensee will recover the residual uranium from these cylinders. As of February 14, 1997, the licensee had not received any Type 5B cylinder shipments.
Apparently, Portsmouth is awaiting an approval from DOE before the first shipment is made.
e One Liners 3
- e. Westinohouse This week Line 1.was used to process recycle material. UF was being processed through Lines 2, 3 and 4. Line5.normallyused to process uranyl nitrate recycle material, was shutdown for clean out. Processing on all lines has been shutdown for a limited amount of time due the licensee's new " Time Based Management" concept. This concept dictates that only a certain amount of powder be produced and then the conversion lines are shutdown until the pelleting area uses what is on hand and needs more.
This program is designed to limit the amount of powder inventory on hand. There have been no process upsets or reportable incidents during the past two weeks.
Last week regional inspectors were on site performing a health physics and environmental protection inspection. This week an inspector from head safety inspection, quarters is on site performing a chemical
- 2. , Business Process Reenaineerina (BPR)
On February 18, 1997, the staff in NHSS and Region II will brief the Commission on the status of the BPR for materials licensing. B. Hallett will be participating in the briefing for Region II. The brief will
- include the results of the pilot, which was held in Region II during January 27 31, 1997. The pilot tested the receipt of electronic applications from licensees, the use of the new electronic / computer assisted review checklist, the use of the electronic / computer database for all the guidance and the generation of the materials license via the computer assisted database, Participants in the pilot included fo~ur licensees, four portable gauge manufacturers, NRC, and representatives from the Agreement States of Georgia, Illinois and-Texas.
- 3. Jose L. Fernandez. H.D.
On October 21, 1996, the NRC issued an Order to the licensee to obtain a consultant to identify and notify patients who received overexposures due to a miscalibrated Sr 90 calibrator. The licensee is progressing with the review, which must be complete by February 7, 1997 (45 days after NRC approval of the consultant on January 21, 1997),
The Order also required the licensee to dispose of the two applicators-
. licensed. The licensee has shipped one of the applicators t11s week and plans to ship the second to an authorized recipient in New York upon return of the shipping container from the first shipment.
Region 11 is continuing to pursue the review of records from another physician, Dr, Vazquez, who also used the miscalibrated applicator to
--treat patients. On February 12, 1997, the Region issued a letter to the corporation who holds the records for the deceased Dr. Vazquez. The plan is to send Region II and Headquarters staff to perform the review
-during the first week in March.
One Liners 4
- 4. DeDartment of the Army. Rock Island. II, In response to a request from NMSS, Region II provided copies of Army licenses in Region II along with the results of inspections performed over the last n /e years. Information regarding inspections of-Navy-facilities was also provided. This information was requested in preparation for a meeting between the Director, NHSS, and the Department l_
i of the Army concerning the control and accountability of radioactive commodities in the possession of the Army, Region III (responsible region for the Army commodity 11ccises) has identified a number of issues with the Army's control and accountability of radioactive commodities (i.e. compasses, range finders, etc.) and failure to re) ort l the loss of commodities. Region II identified similiar problems wit 1 the Navy's control and accountability of radioactive commodities in 1996 and the Navy is in the process of fixing their system.
- 5. Navy On February 8, 1996, the Navy informed Region II that in December, 1995 and January,1996, a Marine aircraft expended 1,520 rounds of 25 mm depleted uranium (DU) ammuntion at an Air Force gunnery range on a small, uninhabited island a> proximately 50 miles from Okinawa, The Navy's license does not autlorize the firing of DU munitions at this range. The aircraft had been mistakenly loeded with the DU rounds because of a supply system error. The Air Force is responsible for maintaining control of the island and subsequently had the range surveyed and recovered approximately 192 rounds, In December, 1996, representatives of the NHK, a Japanese television network, contacted Region II and requested information.on the incident, Region II staff provided what information they had on the incident and referred the NHK representatives to the Navy's Radiation Safety Committee contact for additional information.- On February 10 and 11, '1997, articles about this incident and the reported failure of the U.S. government to officially report this incident to the Japanese government appeared in the Washington Times and Washington Post. Region II staff have provided infermation regarding this incident to the NRC Office of Public Affairs and Office of International Programs so these offices could respond to inquiries from the Japanese Embassy,
FEB-14-97 FRI 09:48 NFS FAX NO. 4237439025 P.02 Huckcv Fwl Swket. bc.
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NUCLEAR FUEL SERVICES ERWIN TENNESSEE NOTICE TO ALL EMPLOYEES February 11,1997 Vice President of Safety and Regulatory, offective Februa Dr. Baer received his B.S. from the U.S. Naval Academy, M.A. at the U S Naval ..
College, Cincinnati. and his M.S. and Ph.D. In Nuclear Engineering from the Univers D
waste and hazardous waste industries. His Navy bac ,
l service and he retired as a Captain in the Navy Reserves. He previously held such as industry. Chief Operating Officer and Vice President of various companies in the Addit!onally, Regulatory organization: I wish to also announce the following changes in the Safety an Andrew M. Maxin will assume the position of Director of Safety and Regula Compliance, locations. which will encompass responsibilities at the Erwin facility and off. s and Regulatory as of February 12,1997. James R. Clark will conclude a consultant on an as needed basis for various projects.However, he will continue I know on tha challenges that you willthem.
that face continue to provide ycur support as each of these employee A e Dwight'B. Kerguson
' M)W O
President and Chief Operating Officer u -
e GE Nuclear Energy Crer E re c C:~cer ro esaso a - unx:sez 910675 5XC
\
February 19,1997 Mr. M. F. Weber, Licensing Branch, NMSS '
U.S. Nuclear Regulatory Commission Mail Stop T 8 D 14 Washington, DC 20555 0001
Subject:
License Renewal - ISt. for DCP Summary tieference: (1) NRC License SNM 1097, Docket 701113 (2) License Renewal Application,4/5/96 (3) Submittal, RJ Reda to ED Flack,5/6/96 (4) Submittal, RJ Reda to RC Pierson,5/14/96 (5) Letter, RC Pierson to RJ Reda,7/18/96 (6) Submittal, RJ Reda to RC Pierson,8/30/96 (7) Submittal, RJ Reda to ED Flack,9/26/96 (8) Letter, MA Lamastra to RJ Reda,10/2/96 l (9) Submittal, RJ Reda to MA Lamastra, 11/22/96 l
(10) Application, RJ Reda to MF Weber,12/16/96
' (11) Letter, MA Lamastra to RJ Reda, 12/17/96 (12) Letter, RJ Reda to MF Weber,2/5/97 (13) Letter, MA Lamastra to RJ Reda,2/10/97
Dear Mr. Weber:
As part of our facility modifications for the Dry Conversion Process (DCP), and as part of our site license renewal for NRC license SNM 1097, GE committed to complete an Integrated Safety Analysis (ISA) f . the DCP and submit a summary of that work to the NRC. In accord with this commitment, GE has completed the ISA and has summarized the results of this work as an attachment to this letter.
Six copies of this submittal are being provided with a dated tab for insertion into the binders we previously sent.
Please contact Charlie Vaughan on (910) 675-5656 or me on (910) 675 5889, if you have any questiens or would like to discuss this matter further.
Sincerely, GE NUCLEAR ENERGY L
Ralph J. Reda, Manager Fuels & Facility Licensing g
Attachment '
cc: RJR-97-014 W
_cj q g Q o L-9 k(ui ,f'
- Mr. M. F. Weber February 19,1997 Page1ofI s
ATTACHMENT TO LETTER FROM RJ REDA TO MF WEBER DATED FEBRUARY 19,1997 l
l i
l ISA
SUMMARY
GE WILMINGTON DRY CONVERSION PROCESS DATED FEBRUARY 19,1997 l
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ISA Summary GE Wilmington Dry Conversion Process GE Nuclear Energy Wilmington, NC February 19,1997 SupplementalInformation to License Renewal:
SNM-1097, Docket 70-1113 buh* ~
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ISA Summary GE Wilmington Dry Conver ion Process Table of Contents Title &
Table of Contents i,11, iii introduction 1 ISA Teams 1 Procedures, Techniques and Tools 1 l Process Organization 2 Methodology 2
, Building Design Basis 5 General DCP Arcas 5 Loss of moderation control in the DCP Facility due to water ingress 5 Fire / Explosion within the DCP Facility 6 Radiological exposure of personnel to airbome uranium 6 Asphyxiation of worker from excessive concentrate of nitrogen 7 Failure of the distributed process control system 7 Cylinder Storage sud Handling 7 Loss of contained external to DCP leading to release of UF6 to the 7 environment Vaporization 8 UF6 release due to a loss of UF6 cylinder integrity within the 8 autoclave i ofiii I
.z Table of Contents
- Illit Essa
- Vaporization (cont'd) 8 )
UF6 release to the room due to failure cf UF6 piping 8 4
Cold trap rupture leading to release of UF6 9 Criticality due to backflow of steam moderator into the cylinder 9 Conversion 9 Loss of hydrogen containment leading to fire or explosion including 9 loss of HF containment and uranium containment 4 Loss of moderation control leading to criticality in the kiln and 10 associated reactor and outlet hopper Criticality in HF areas caused by release of particulate uranium to 10 unfavorable geometry liquid system
. Criticality in the HF area caused by a release of unreacted UF6 10 aqueous uranium to s.nsafe geometry -
Loss of moderation control due to backflow ofliquid from the HF 11 area Exposure of an operator to HF, steam, or radiological hazards from 11 backflow of reactor contents during recycle operations Personnel injury cause by moving machinery and rotating parts 11 Powder Outlet 12 Excess moderation in the cooling hopper 12 Homogenization 12 Excess moderation by introduction of hydrogenous material such as 12 pore former, die lubricant, water or oil Radiological exposurc of personnel to airbome uranium caused by 13 loss of containment Personnel injury caused by rotating parts on the sifter and 13 homogenizer ii oflii s
Table of Contents 1.1111 - f. Bit Blending, Slugging, Granulation and Tumbling 13 j Excess moderation by introduction of too much hydrogenous 13 material, for example pore former or die lubricant, as a result of errors in weight, additive or mixing Excess moderation as a result of wet powder 14 Excess moderation as a result of oil 14 Radiological exposure of personnel to airbome uranium caused by 14 loss of containment Personnel injury caused by moving machinery and rotating parts 15 Powder Pack 15 L Radiological exposure of personnel to airbome uranium caused by 15 L- loss of containment I
Personnel injury caused by moving machinery and rotating parts l$
Container Storage And Handling 15 Exceed moderation restricted area limits by overfilling storage 16 containers, introduction of too much hydrogenous material, or violating spacer requirements for containers Large powder container falls on a person 16 HF Area 16 Criticality in HF area caused by excess uranium in the off-gas 16 collected in unfavorable geometry liquid system Release of fluorides through the stack to the environment 17 Employee exposure to excessive HF 18 Environmental spill ofliquid HF 18 Fire or explosion in the HF building or off gas line 18 iii oflii
ISA Summary GE Wilmington Dry Conversion Process Introduction An integrated safety analysis (ISA) was performed to support the new dry conversion process at the GE Wilmington site. Le analysis was conducted in accordance with Chapter 4, Integrated Safety Analysis, of SNM License 1097, submitted April 5,1996, as amended, ne purpose of the analysis was to determine potential accident scenarios and risk, to ensure that controls are in place to prevent and/or mitigate accidents, and to rank the controls in relation to the risks which they mitigate so that the proper level of assurance measures can be applied to each. The broad scope of the analysis included criticality safety, radiological safety, environmental protection and industrial safety, including chemical safety and fire protection. This report summarizes the important results of the safety study for the dry conversion facility. Detailed records are retained on site.
I f
I ISA Teams The ISA was performed by teams of people of different expertise who systematically analyzed the hazards in a focused meeting environment. Each team included expertise in criticality safety, radiological safety, industrial safety and environmental protection. Also inc uded were process and controls engineers. Because the process had never been opented at the Wilmington site, experts in the operation of existing plants utilizing this technology participated along with GE personnel.
Procedures, Techniaues and Tools ne ISA team used the Hazop method and limited use of the What if method to complete the analysis. He hazop analysis was facilitated by and documented in a software package known as HazopPC, deve.oped by Primatech, Inc. This software was customized for GE to specifically capture nuclear hazz-ds. Consultants from Primatech provided training to team participants early in the project, and facilitated the preliminary sessions. Guidance for performing the analysis came from Primatech and from Page 1 of18
Guidelines for Hazard Evaluation Procedures'. As the project neared completion, an internal procedure, P/P 10-20, Integrated Safety Analysis, was developed to define how future ISAs would be performed and how the results ofISAs would be used to truage risk in the opcrating factory.
Process Oreanization The DCP ISA was broken into the following segments and analyzed using the method sbown:
Cylinder Movement - Hazop Vaporization - Hazop l Conversion - Hazop l Powder Outlet - Hazop Homogenization - Hazop Blending, Precompaction, and Granulation and Tumbling - Hazop HF Treatment - Hazop Containers / Storage - Hazop Powder Pack - What if Methodology Hazards were identified using historical input from a variety of sources including safety information from other dry conversion facilities and public domain documents from other sites in the fuel cycle. The relationships between the hazards and DCP process deviations were developed through guided brainstorming and event visualization techniques.
Where the hazop methodology was used, each process segment was divided into logical i nodes. For each node, the team first considered the intended operating conditions. Next they considered the guible deviations from the design intention using a list of parameters and guidewords. Typical parameters included:
Flow Temperature I Pressure Composition Level l l
Weight
' Guidelinesfor Ha:ard Evaluation Procedures. Second Edition with Worked Damples, Center for Chemical Process Safety of the American instinne of Chemical Engineers, New York,1992.
Page 2 of18
Guidewords'were used to identify possible deviations from normal. Example guidewords included: 1 i
No/None More bss As Well as Part of Reverse Other than Once possible deviations were identified, possible causes of the deviations were hypothesized. The consequence of the deviation was estimated through experience or technical evaluation. The severity of each consequence was rated on a three point scale, shown in Exhibit 1. The likelihood of the accident occurring with no controls in place was estimated from experience on a three-point scale, shown in Exhibit 2. Safeguards or controls that prevent or mitigate the consequence were identified. Outside of the team -
l environment, the controls were ranked in importance by the unmitigated risk, where:
(Risk) - _ > = (Severity) x (Likelihood),w The level of unmitigated risk was used to determine the level of assurances that would be applied to the controls. Risk ranking is summarized in Exhibit 3.
Finally, the likelihood of the consequence occurring with the controls in place was estimated. Final risk was used by the team to judge the adequacy of the controls, where (Risk)w = (Severity) x (Likelihood) w if the final risk was judged by the team to be too high, recommendations for improvem'ent were captured and reviewed with management.--Recommendations that were adopted were tracked to completion.
Consequences associated with UF. and HF were estimated using past experiences of OE Wilmington, FBFC, and other uranium-processing facilities and experiences in the .
chemical industry. Criticality consequences were _ estimated using standard analytical, imethods Radiological consequences were estimated by extrapolating our experience with the existing processes. Environmental, industrial,- and chemical cousequences, including fire and explosion, were estimated with the aid of material safety data sheets, chemical interaction information, and various modeling techniques, including emission calculations and dispersion models for fence-line concentration.
Page 3 of 18
Exhibit 1 Severity of Consecuences Severity RadiologicaU EnvironmentaU Ranking Criticality industriaVChemical e exposure to an individual rnemoet of e fatality 3 the pubhc off-site ($ rem,30 mg antake e medica: assenerd for a member of the of U) public off-site e severe exposure to an employee (400 e permanent disabihty rem intemal plus external dose or 230 e off-site contamination above regulatory mg intake of U) standards e exceed regulatory hmets for employee e senous hjury 2 exposure (5 rem,10 mg U intemal) e exceed permit hmits or regulatory nmits e lost time injury reportable release e exceed administrative hmits on daily air
- OSHA recorcable iniury i samples, lung counts, bioassays, e first aid contamination, TLDs e exceed intemal limits e 10% of annual exposure hmit e spillinside containment Unusualincident (tar)
Exhibit 2 Likelihood Level Frequency Likelihood 3 more frequent than once Likely to occur in the immediate future every two years 2 every two to fifty years Likely to occur dunng the hfe of the fac$ity 1
less frequent than once every Not likely to occur dunng the hfe of the facihty 50 years.
O incredible Likehhood '4 indistinguishable from zero.
Exhibit 3 Risk Matrix C 3 Mid-level o S Risk n e s y 2 Low Risk Mid-level e e Risk q r u i e t 1 Low Risk Low Risk Mid-level n y Risk C
1 2 3 e
Likelihood Page 4 of 18
Jgildina Desian Basis We DCP facility has been designed to meet the relevant building and fire protection standardsc ne design criteria were taken as bounding assumptions by the ISA teams with respect to extemal events, of which heavy rain, lightning strikes, and high wind are considered credible. The DCP facility is designed with a structural integrity providing a safety factor of two. It is built to withstand a sustained wind speed of 120 mph. Although the seismic activity is not considered credible, the DCP facility has been designed to meet the lateral acceleration requirements for Wilmington, North Carolina, as contained on the contour maps in ASCE 7-9.f. He electrical classification is Class I, Division. 2, Group B in close proximity of the conversion kiins, where there is some risk of H escape.
Elsewhere within DCP, the electrical hazard classification is non-hazardous, ne GE Wilmington site is located above the 100 year flood plain and thus is not considered susceptible to storm surge or flash floods.
What follows is a summary of the more significant potential accidents identified and i
discussion of the controls that are in place to protect against them.
, GeneralDCP Areas l
PURPOSE: To provide a facility where moderation is restricted, fire or explosion is unlikely, and radiological exposure or contamination is minimi=A ACCIDENT: Loss of moderation control in the DCP Facility due to water ingress.
EROTECTION: ne DCP process areas are designated as Moderation Restriction Areas (MRA).- ne building is designed to be water tight. ne roof of the DCP building is designed to prevent _ water leaks and to provide an early_waming of potential degradation.
'Its special multi layered construction incorporates a reinforced concrete slab roof deck, an EPDM membrane, a drainage system of extruded polystyrene insulation, a concrete slab over geotextile fabric, and a fully adhered butyl sheet. Degradation of the external barrier
- is detected by a visual observation of water in the drainage system. Also, there are no penetrations on the roof of the building, and the roof slopes from east to west to convey rain water away to a gutter and downspout system on the west side of the building.
Air handling equipment is designed to minimize the possibility of water entering the Moderation Restricted Areas in= DCP, ' Moisture detectors- and alarms are fitted-downstream of the last coil on the air intake to DCP. While small amounts of water on the floor do not present a safety problem, the ground floor of the DCP building is raised in relation to the existing FMOX ground floor to prevent in-leakage of water from rising
_ water. _
Page5of18
1 ACCIDENT: Fire / Explosion within the DCP Facility PROTECTION: ne DCP building is designed to comply with relevant fire codes.
Interior doors and walls provide fire resistance. A comprehensive fire alarm and detection system includes smoke and heat detectors as well as H2 and HF detectors. Fire alarm homs annunciate in the DCP building and in the HF building. Alarm signals are relayed to both the DCP control room and to the Site Emergency Control Center.
A CO 2fire suppression system protects the HVAC ductwork and the HEPA filters.
Consistent with the Moderation Restriction Area, only carbon dioxide fire extinguishers are installed within the DCP building. A water sprinkler system protects the HF building.
Operating and traintenance personnel receive fire fighting emergency training and are periodically tested by unannounced emergency evacuation exercises.
Ha is handled and processed within closed systems. Ignitable concentrations of H2 are prevented by positive mechanical ventilation. Hydrogen sensors are located within each kiln room, and continuously monitor atmosphere for unacceptable concentrations of hydrogen.
He DCP kiln generates a low specific surface area product, which is resistant to spontaneous oxidation in storage. Further protection against oxidation includes powder cooling, nitrogen blankets, and interrupted mixing cycles.
1 ACCIDENT: Radiological exposure of personnel to airbome uranium.
PROTECTION: A general feature of the design of the DCP facility is the containment and ventilation philosophy to provide defense in depth against the escape of airbome uranium into areas in which personnel would not nonnally be required to wear personal protective equipment. Specifically, airbome uranium is contained by nitrogen seals at the connections between the processing equipment and storage / transfer containers. Uranium processing activities are housed within ventilated rooms, designed so that the lowest pressure is within the rooms themselves and air flows from outside to 'mside the rooms.
Each room has HEPA filtration to trap particulates within the room, and the air intake and extraction ducts are arranged to draw air downward and away from personnel work areas and provide adequate air changes to avoid static air pockets. Ultunately, the extracted air is filtered through another HEPA filter bank before discharge to the environment.
Redundant fan capacity and fan failure alarms are provided on the DCP HVAC system.
Operating procedures require floor operators to inspect the plant areas on a routine basis.
Existing GENE practices require operators to clean up visible contamination as soon as it is discovered. Tight fitting connections equipped with inflatable seals are 'mstalled at all container filling and dump stations. Where flexibility is required in the powder feed pipework, rubber boots and intemal metal sleeves are used to limit powder spillage in the Page 6 of18
l cvent of a rubber boot failure. Operating and maintenance penonnel are required to wear the designated personal protective equipment when breaking primary containment on uranium processing equipment.
A large powder spill within a containment room are evident when the operator views the operations through a window or enters the room to remove the container.
ACCIDENT: Asphyxiation of a worker from excessive concentration of nitrogen.
PROTECTION: Nitrogen piping systems were designed to exceed the antfcipated operating pressures, and are pressure test,xi during insthilation. In the event of a leak in a process room, positive ventilation provided by the liVAC system is designed to supply adequate air changes. The HVAC system is equipped with redundant fans to assure reliability. Plant procedures for entering confined spaces protect workers during maintenance of the large vessels such as the homogenization blender.
I ACCIDENT: Failure of the distributed process control system PROTECTION: All valves are designed to fail in a safe position in the event of a total i control system outage. Active safety controls are implemented separately from functions
. that may be needed for nonnal operating functions and access, resulting in a more robt+.
software operatinC system. Changes to the software are subject to strict version control per procedure. Important controls are functionally tested before start up and periodically thereaAer.
Cylinder Storare and Handline PURPOSE: Safe movement of UFs feed material into and out of the facility.
ACCIDENT: Loss of containment external to DCP leading to release of UF6 to the environment.
PROTECTION: UFs cylinder integrity complies with the 30B type specification dermed in USEC 651 Rev. 7. Jan 1995. The process design ensures that cylinders are handled only when the UFs inside them is in the solid phase. Cylinder lifting equipment has been designed and tested to handle the load. Operator training in UF6 cylinder handling includes the use of forklin trucks, cranes and_ hoists, Cylinders are moved with the cylinder valve cover in place and lifted to the minimt.m lift height consistent with the terrain. Overpressurization of a cylinder due to fire is prevented by strict control of the inventory of hydrocarbon fuels and combustible materials in the UF6 cylinder storage area as described in NUREG 1491. Existing administrative controls for the receipt and release Page 7 of18
of UFs cylinders ernure that overweight cylinders or enrichments grtater than 5% *U are not processed in DCP.
I Vaporiration ;
PURPOSE: Supply UF, gas to the reactor where it is converted to UO2 powder.
l ACCIDENT: UF6 release due to a loss of UFe cylinder integrity within the autoclave.
l l PROTECTION: ne autoclave has been manufactured to meet the ASME pressure I
vessel code, and is pressure. tested annually, ne rated pressure exceeds the nonnal operating pressure of the UF cylinders, ne integrity of the door seal is tested each time a cylinder is loaded into the autoclave. A door latch proximity switch is interlocked with the cylinder heating cycle Loss of seal pressure disables the autoclave heating sequence.
The intemal pressure within each cylinder is verified to be below atmospheric pressure before heating as a safeguard against overpressurization during the heating cycle, ne UF6 cylinder pressure is controlled at the desired set point by monitoring the UF6 line pressure downstream of the cylinder, and uses feedback control to regulate the current sepplied to the autoclave heaters. Active controls prevent overpressurization of the cylinder by de-energizing the autoclave heaters on high temperature at the cylinder surface, or high temperature inside the autoclave.
UFs leak detection is installed as part of each autoclave system, and draws a sample from the autoclave annular space. The annular space is swept with heated dry nitrogen to prevent reaction with air in the event of a UF6 leak. His nhrogen may be fed at high pressure to sunpress potentialleakage of UF6. As a result, a leaking cylinder can continue feeding safely to the reactor until the cylinder is empty and the hazard is reduced.
To protect workers from exposure during cylinder installation, personnel are required to wear appropriate personal protective equipment when connecting and disconnecting the UF6 line pigtail to the cylinder.
ACCIDENT: UF release to the room due to failure of UF6 piping.
PROTECT 10Mt The piping integrity is assured by 100% radiographed welds on all UF6 lines. UF rnsistant 6 gaskets are used at flanges, and bellows sealed valves minimize stem leaks. He UF6 line is further protected against failure due to hydraulic rupture by automatic temperature control of the trace heating with high and low temperature alarms and high temperature interlocks that disable the alTected section of trace heating.
1 Page 8 of 18
UF6 leak detectors are installed in each autoclave room a the event of a UF6 leak in an j autoclave room, the ventilation to that room is shutdown to contain the UF6 inside, i
ACCIDENT: Cold trap rupture leading to a release of UF6 PROTECTION: The cold trap has been manufactured and tested in compliance with ASME and CODAP pressure vessel codes. The cold trap heaters are disabled on high I temperature, high pressure, or high weight. Valve line ups are interlocked to prevent the inadvertent transfer of UF6 to unintended destinations, for example, cylinder to-cylinder or cylinder to-cold trap during conversion.
l ACCIDENT: Criticality due to backflow of steam moderator into the cylinder.
PROTECTION: During operation of the plant, a positive flow of gas is maintained from the UF6 line to the reactor. Active controls automatically isolate the cylinder from the reactor in the event that positive pressure differential between the cylinder and reactor is lost. The UF6 cylinder skin temperature must be above a predetermined minimum value before the UF feed valve can be opened.
1 Conversion PURPOSE: To convert UF6 into ceramic grade UO powder suitable for pelletizing.
This process includes the recycle of discrepant UO powder to reduce fluoride content.
ACCIDENT: Loss of hydrogen containment leading to fire or explosion. (Also loss of HF containment leading to worker injury or loss of uranium contaimnent leading to personnel exposure.)
' PROTECTION: The primary protection against loss of containment is the kiln itself.
The kiln seals are oflabyrinth construction with the interspace pressurized with nitrogen.
The kiln is protected from overpressure by an automatic pressure control system.
Pluggage of the sintered metal filters at the gas discharge of the kiln cou'd result in a pressure higher than desired. The filters are kept clean by a computer-controlled sequence that blows back the metal filters with nitrogen. During operation the [nssure drop across the sintered metal filters is continuously monitored. Deviations from normal result in audible / visual alarms in the control room. Furthermore a high kiln pressure interlock stops the UF6 feed, followed by an orderly shutdown of the hydrolysis and pyrohydrolysis steam and Ha feeds.
3 '
The kiln and associated vessels are housed inside a ventilated containment area. Ha detectors in the rooms automatically stop the H and UF6 feeds to the affected conversion Page 9 of I8
kiln. HF detectors send audible / visual alarms to the control room, and the operators are trained to manually shutdown the gas feeds to the affected kiln.
ACCIDENT: Loss of moderation control leading to criticality in the kiln and associated reactor and outlet hopper.
PROTECTION: To ensure that the steam fed to the reactor and kiln is dry, saturated steam is superheated using electric heaters equipped with automatic temperature controls regulated by the computer control system. All steam lines are electrically trace heated and insulated. A low temperature condition activates audible / visual alarms in the control room. The heat tracing is equipped with redundant heating circuits. Low steam ,
temperature fed at the superheater or downstream of the superheater is sensed and interlocked to the steam feed control to the reactor. Both hydrolysis steam and pyrohydrolysis steam supplies are protected by these controls.
A low temperature inside the reactor or on the reactor wall is sensed and interlocked to the steam and UFs feed control A low temperature interlock on the exterior of the kiln barrel also stops the steam and UF4 feeds, Similar alarms and interlocks are activated by thermocouples located inside and on the exterior of the outlet hopper.
ACCIDENT: Criticality in HF area caused by release of particulate uranium to unfavorable geometry liquid system.
PROTECTION: 'Ihe primary protection agalast entry of particulate uranium wo the offgas system are the sintered metal filters housed in the top of the reactor.- Further protection is provided by a back up sintered metal filter unit installed in the offgas line downstream of the reactor. 'Ihe integrity of both sets of filters is assured by continuous monitoring of the pressure drop across the filters. The kiln filters are periodically backpurged with nitrogen during operation to minimize powder build up.
Passive and active controls in the HF area that prevent criticality are described in the HF section.
ACCIDENT: Criticality in the HF area caused by a release of unreacted UFs aqueous uranium to unsafe geometry.
PROTECTION: An excess of hydrolysis steam (normally a factor of two) is fed to the reactor to provide complete conversion of UFs to the intermediate UO2F2. Furthermore, sufficient pyrohydrolysis steam !. provided to convert all the UF6t o UO2F2 and to UO 2 in the event ofloss of the hydrolysis steam flow. A high UFs flow interlock closes the UF6 feed valve and starts sweeping nitrogen to the injector nozzle. Similarly, low steam flow Page 10 of18
to the reactor stops UFs feed. Rese controls prevent less than stoichiometric flow of steam to the kiin and thus a loss of UFs in the off gas.
Other potential causes of a loss of steam flow to the reactor, for example, an open manual vent valve in steam line, an open pressure relief valve on steam superheater, are mitigated by the low now interlock described above, operating and maintenance procedures
! including a stanup checklist, routine inspections, and vendor recommended periodic maintenance.
4 i Passive and active controls in the HF area that prevent criticality are described in the llF section.
ACCIDENT: less of moderation control due to backflow ofliquid from the HF area.
EROTECTION: The off gas flow created by the condenser eductor and exhaust fans I prevents backflow. Condensation of HF offgas at the condenser creates an additional favorable draft. A passive barometric leg, by virtue of the height difference between the reactor offgas line and the vapor liquid separator downstream of the liF condenser, prevents siphoning ofliquid to the reactor. Specific engineered controls are summarized in the HF section.
During shutdowns, protection against backflow is afforded by administrative controls 2
which require physical isolation of the reactor from the IIF system.
ACCIDENT: Exposure of an operator to HF, steam, or radiological hazards from backflow of reactor contents during recycle operations.
PROTECTION: In the event of a low pressure alann at the inflatable seal at the recycle feed station, the recycle feed valve closes, preventing altbome contamination at the recycle port. He recycle station is located inside ar melosed kiln room. Operators are required to wear the designated personal protective eqtJpment when making or breaking powder feed connections. A double valve arrangement equipped with a nitrogen purge in between prevents backflow of steam and hydrogen into the recycle container. A low pressure alarm on the recycle feed station alerts the operator to a condition which could result in off gas flowing to the unicone.
ACCIDENT: Personnel r.ury caused by moving machinery and rotating parts.
PROTECTION: ne kiln drive units, reactor and recycle screw motors are protected adequately by machine guards. De start up checklist includes a step to check that these guards have been refitted following maintenance. A site wide procedure dermes lock-out/ tag-out practices.
Page 11 of 18
Powder Outlet PURPOSE: To cool the UO powder and ensure that it meets moisture criteria applicable to the Moderation Restricted Area, and to transfer the powder into transfer containers.
ACCIDENT: Excess moderator in the cooling hopper.
PROTECTION: Controls in conversion that ensure dry powder in the kiln also limit the availability of moisture to condense on the powder. To keep steam from flowing into the cooling hoppers, a valve lock system is installed at the outlet of the kiln. 'nds lock consists of two valves with a dry nitrogen purge between them. The cooling hoppers are further protected by continuous redundant dew point analyzers which measure the free moisture in the carrier nitrogen gas. A high moisture interlock stops the UFs feed to the kiln and initiates a controlled shutdown of the steam flow. The control system automatically isoletes the powder in the cooling hopper until the powder can be transferred under conditions favorable to criticality control requirements.
, To prevent overfilling of the unicone on discharge from the cooling hopper, the computer l system monitors the weight dispensed into the unicone via the process weigh scales, and l terminates the sequence when the set weight is reached. in addition, there is a high weight interlock on the scale at the container filling station that will closes the cooling
! hopper discharge valve.
I I
Homogenization PURPOSE: To produce a UO2 powder that is physically and chemically uniform for subsequent pressing into pellets and sintering. Homogenization breaks down soft agglomerates, removes hard particulate material, and mixes the sifted powder to produce a uniform batch.
ACCIDENT: Excess moderation by introduction of hydrogenous ma' rial such as pore former, die lubricant, water or oil.
PROTECTION: Upstream controls in the conversion area limit the moisture content in the powder. Upstream controls in the recycle area similarly limit the moisture content of the recycled U30s. A high mass interlock closes the powder feed valve and limits the amount of surface adsorbed water available for redistribution from the surface of the powder to one spot within the homogenizer.
A computer assisted administrative control prevents the introduction of feed material containing pore former or die lubricant.
Page 12 of 18
ne gearbox for the screw and arm assembly inside the homogenizer contains a special non hydrogenous low moderation oil, ne procurement and use of this oil is controlled via the maintenance planning and control system.
ACCIDENT: Radiological exposure of personnel to airbome uranium caused by loss of containment, t
PROTECTION: To prevent airbome contamination, the siRer shan seals are purged with nitrogen. The siRer and homogenizer are housed within a ventilated containment area.
For additional protection, a dedicated containment hood surrounds the siner and magnetic ,
separator. - A high powder level alarm is fitted on the oversize line from the slaer to alert i the operator that the oversize bottle requires changing. Administrative controls including procedures and operator training ensure that the oversize outlet valve is closed before the bottle is removed.
To prevent overfilling of the unicone on discharge from the homogenizer, the computer system monitors the weight dispensed into the unicone via the process weigh scales, and terminates the sequence when the set weight is reached. In sddition, there is a high weight interlock on the scale at the container filling station that will closes the homogenizer discharge valve.
ACCIDENT: Personnel injury caused by rotating parts on the siRer and homogenizer.
PROTECT 10M nere is an interlock that stops the rotation of the homogenizer screw and arm when the inspection hatches are opened on the top or bottom of the homogenizer.
Also, the siRer rotation stops whenever the top inspection cover is opened or the end cover is removed. A site wide procedure defines lock out/ tag out practices.
Blendine, Slunnior, Granulation and Turnbline PURPOSE: He blending operation creates a powder which is of a specific, uniform enrichment. Additives improve compatibility. Slugging followed by granulation, forms a flowable, pressable, granular solid suitable for high quality pressing and sintering.
Tumbling improves flowability of the powder .
ACCIDENT: Excess moceration by introduction of too much hydrogenous material, for example pore former or die lubricant, as a result of errors in weight, additive, or mixing.
PROTECTION: Additives are prepared in separate equipment to prevent mix ups during processing. Additives are weighed on an accountability scale and labeled, ne additive bottle size limits the mass of moderator that can be added at one time. Unique couplings on the additive bottles and the additive ports assist in the control of the addition.
Page 13 of 18
He blend plan defines and documents the quantity and type of additive required for each blend, ne computerized traceability system tracks the concentrations of additives and moisture in the powder fed to the blender. Computer systems ensure that each additive addition complies with the blend plan. Additive bottles are barcode labeled, and the computer systems prevent the feed valve from opening if the bottle does not match the expected barcode. The computer system provides a final comparison of blend weight to that expected from the blend plan. Local prograrnmable logic controllers lockout the feed valve on high weight.
Rotation of the blender is assured by a sensor to improve the unifonnity and reliability of the additive mixing step.
In the tumbling operstion, additive preparation is protected by similar controls. Only one bottle is available in a tumbling room, and that bottle is admin'stratively controlled so that it is assigned to the correct turnbler, ne additive bottle size is limited, ensuring the necessary safety margin for criticality safety control.
offlDENT: Excess moderation as a result of wet powder.
PROTECTIONtUpstream controls in conversion limit moisture content in the powder.
I Similarly, upstream controls in the recycle area limit moisture in the U3 0s. A high mass interlock closes the powder feed valve and limits the amount of surface adsorbed water available for redistribution from the surface of the powder to one spot within the homogenizer.
ACCIDENT: Excess moderation as a result of oil.
PROTECTION: Limited volumes of oils and greases are used on the slug press. He gearbox for the screw and arm assembly inside the blender contains a special non-hydrogenous low moderation oil. The procurement and use of this oil is contoned in the maintenance planning and control system. Grease is extemal to the granulator and tumbler and is not considered a significant hazard because ofits physical separation from the powder and the small quantity used.
ACCIDENT: Radiological exposure of personnel to airborne uranium caused by loss of l containment.
PROTECTION: The blend / slug / granulate and tumbling processes are designed to protect workers from radiological exposure with the same protection described in the General DCP Area section of this report. Additional controls prevent overfilling of the unicone at Page 14 of 18
I l
the blender filling station or the bicone at the granulator filling station: the computer system monitors the weight dispensed into the container via the process weighscales, and terminates the sequence whea the set weight is reached. In addition, there is a high weight interlock that closes the appropriate powder filling valve. ;
l ACCIDENJ; Personnel injury caused by rnoving machinery and rotating farts.
PROTECTION: There are interlocks that stop the rotation of the screw and arm when the inspection hatches at the top or bottom of the blender anc opened. Containment around the rotating turret of the slug press limits access to the moving parts. The inspection hatch on top of the granulator is fitted with a hard wired interlock which stops the motor when the hatch is opened. A hard wired door lock on the tumbler room prevents ,
operation of the tumbler unless the door is locked.
Powder Pac.h PURPOSE: To dispense UO powder or granules from DCP storage containers to approved containers for shipment to customers.
ACCIDENT: Radiological exposure of personnel to airbome uranium caused by loss of containment. .
PROTECTION: ne con'.alner integrity is assured administratively by obser/ing the operedng procedures and recommended maintenance. Operators are required to wear appropriate personnel protective equipment when connecting or disconnecting powder feed connections, An inflatable seal assembly is fitted at the container feed station.
Powder packing operations are carried out within a ventilated hood, so that the operator can handle and seal the packed product through arm slits. The slits have adequate face velocity to reduce the risk of escape of airbome contamination.
ACCIDENT: Pers wel injury caused by moving machinery and rotating parts.
PROTECTION: The powder packing facility is housed in a separate room adjacent to the main storage area on the first floor of the DCP building. He powder packing equipment includes machine guards which protect operators from moving parts.-
Container Storage And Handline P_URPOSE: To store large containers of powder containing various enrichments and additives until the powder is needed at pellet press operations.
Page 15 of 18
i I
} ACCIDENT: Exceed moderation restricted area limits by overfilling storage containers, introduction of too much hydrogenous material, or violating spacing requirements for
[ containers.
PROTECTION: Powder is stored in high integrity cicsed containers which are routinely inspected and maintained. Container gross, tare and net weights are tracked on the computerited uranium material accountability system. A significant deviation from the official tare weight triggers a discrepancy alarm and halts any subsequent material transactions involving the affected container.
Upstream process controls provide assurance that the uranium product meets the criteria for hydrogen content within the storage area.
A fixed support grid assists operating personnel to maintain the required minimum spacing between containers. Additional storage is authorized provided it takes place in approved locations.
ACCIDENT: Large powder container falls on a person.
PROTECTION: The use of cranes and hoists for lifting and transporting containers is covered in a site wide safety procedure that spells out safe practices and periodic inspection requirements. Operators are trained in the safe use of this equipment.
HF Area PURPOSE: To produce nominal 50% aqueous hydrofluoric acid from the hydrogen fluoride off gas generated during convenion of UF, to UO 2and load the 11F product into tank trucks for shipment to customers. De facility also scrubs the condenser off gas to remove residual HF fumes before release to the environment. Dilute HF scrubbing liquir' is typically loaded into a truck for delivery to the site waste treatment facility to remove the HF before the water is released to the environment.
ACCIDENT: Criticality in HF area caused by excess uranium in the off gas collected in unfavorable geometry liquid system.
PROTECTION: Dual inline sintered metal filters upstream in the conversion process reduce the risk of uranium canyover to the HF area, in the unlikely event that both filters fail, each conversion line liquid HF stream is equipped with an individual uranium detector. If uranium is detected, the liquid is diverted to a favorable geometry tank known as the polluted tank. If the uranium concentration exceeds a higher trip level, the UF6 feed is stopped, followed by an orderly shutdown of the steam feeds to the affected kiln. De three HF streams combine in a common stream which passes through another uranium detector. If uranium is detected here, the UF6 and steam feeds to all three kilns Page 16 of 18
6 are stopped, and the liF produced during shutdown is diverted to a favorable geometry tank, ne liF condensers are of favorable geometry, and the cooling capacity of the condensers is great enough to trap the UFs in the event of a catastrophic r(lease from the conversion i kiln. ne chilled water supply is protected by a low flow interlock at the condenser j which stops the UFs flow to the kiln. The chilled water supply is also protected by r high temperature alarm and administrative controls on the set up of the system.
Key manual valves that are opened to prepare the system for maintenance are included in pre startup checklists, operating procedures, and training to prevent inadvertent by pass of safety controls during operation.
i A high liquid level in the vapor liquid sepantor stops UFs feed to the affected kiln, and protects against unanalyzed liF condensate backflowing to the unfavorsble geometry washing colurnn. If the high level is coincidem with uranium concentration detected by I the uranium detector, both UFs feed and steam flow to the kiln are stopped. Another protection against uranium in the washing column is a high liquid level interlock on the polluted tank that stops UFs flow to all kilns.
There is a high level sump alarm that annunciates locally in the 11F building and remotely in the DCP control room. Administrative controls require that any leakage of unanalyzed IIF from a piping failure prior to the uranium detectors ccilected in the dike be sampled and analyzed for uranium, his material is pumped back to the IIF storage tanks or cut to the mobile tar 6er only if the uranium concentration meets established release limits.
l ACCIDENT: Release of fluorides through the stack to the environment.
PROTECTION:- ne kiln off gas is scrubbed to remove fluorides from the off gas before it is released to the environment. His process equipment is located inside a building, protected from freezing and severe weather. De temperature and pressure at the top and bottom of the wash column are continuously monitored and alarms locally in the liF building and also in the DCP control room. A low flow of water in the wash column stops the UFs feed to the kiln (s). In the event of a failure in the water supply, an emergency feed water tank is maintained in a ready state to continue sembbing until the reraaining fluorides in the system are purged from the kiln lines. A high liquid level in the base of the wash column stops UFs feed to the idlns. De liF stack emissions are administratively monitored to provide a feedback loop so that corrections can be made if the scrubbing process is insuflicient to meet requirements.
. Page 17 ofl8
ACCIDENT: Employee exposure to excessive HF.
EROTECTION: The HF equipment is des lgned to maintain integrity. The HF offgas line is PTFE lined and rated for 250 deg C. Equipment relisbility is enhanced by periodic maintenance. The integnty of the air-driven HF pump is protected by a interlock that shuts off the air supply in the event of high discharge pressme. Preventative maintenance of the pump diaphragms is performed in accordance with vendor recommendations.
He building is equipped with safety showers in case of personnel exposure to liquid liF.
Use of personal protective equipment is required when making or breaking containment.
Operators are trained in safe use of personal protective equipment. Operators are required to wear appropriate personal protective equipment during sampling. This requirement is including in operating procedures and training.
IIF fumes in excess of the alarm limit ntivates an alarm in the control room which is audible in the HF building. Additional waming lights are activated at the entrances to liF building. Periodic inspection of the remote HF building is required by the operators. An emergency HF scrubber is available for scrubbing the room air.
All tanks are vented to the wash column. Tank trucks are vented to the wash column.
He tanker loading lines are washed with D1 water into the tank truck before the lines are disconnected. A vacuum break valve facilitates complete draining of the tanker feed line before breaking containment and minimizes the potential for spills.
ACCIDENT: Environmental spill ofliquid HF.
PROTECTION: ne HF storage tanks are installed within a dike designed to contain liquid HF spills. The volume of the diked area is sufficient to contain a spill of the entire contents of the largest tank. The dike comprises two concrete slabs with the top slab coated with an acid resistant medium. A grid is installed between the two concrete slabs which gravity drains to a sample point. Periodic sampling allows for early detection of a breach in the top slab, ne tanker loading bay also drains into the main dike and is similarly contained with a concrete slab coated with an acid resistant coating.
ACCIDENT: Fire or explosion in the HF building or off gas line.
PROTECTION: ne eductor that controls the kiln pressure is opemted with nitrogen instead of air. He off gas from the wash column is diluted to below the lower explosive limit of hydrogen by use of an air blower.
Page 18 cf l8
february 27, 1997 General Electric Company ATTN: Mr. C. P. Kio), General Manager GE Nuclear ~ Energy Production -
-P. O. Box 780 Wilmington, NC 28402 3UBJECT: NRC INSPECTION REPORT N0. 70 1113/97 01
Dear Hr. Kipp:
-This refers--to the inspection conducted on January 2'/ 31, 1997, at the Wilmington facility. The >urpose of the inspection was to determine whether -
-activities authorized by t1e license were conducted safely and in accordance with NRC requirements. At the-conclusion of the inspection, the findings were discussed with those members of your staff identified in the report.
I Areas examined during:the inspection are identified in the report. Within l these areas, the inspection consisted of selective examinations of procedures '
and representative records, interviews with personnel, and observation of activities in progress.
scope of the inspection, violations or deviations were not i
In accordance with 10 CFR 2.790, <>f the NRC's " Rules of Practice," a copy of
' this letter will be placed in the NRC Public Document Room.
Should you have any questions concerning this letter, please contact us.
Sincerely, '
<S Edward J. McAlpine, Chief fuel Facilities Branch-Division of Nuclear. Materials Safety Docket No. 70 1113 License No. SNM 1097-
Enclosure:
NRC Inspection Report-cc w/encli (See page 2)
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4 GE 2 cc w/ encl:
Dr. Ralph Reda, Manager Fuels and Facility Licensing General Electric Comsany P. O. Box 780, Mail Code J26 Wilmington, NC 28402 Dayne H. Brown, Director Division of Radiation Protection N. C. Department of Environment, Health and Natural Resources P. O. Box 27687 Raleigh, NC 27611 7687 Distribution w/ encl:
E. McAlpine, RII G. Troup G. Shear RIII C. Cain RIV F. Wenslawski, RIV PUBLIC t
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