ML20247H858
| ML20247H858 | |
| Person / Time | |
|---|---|
| Site: | 07001113 |
| Issue date: | 05/13/1998 |
| From: | NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| To: | |
| Shared Package | |
| ML20247H842 | List: |
| References | |
| 70-1113-98-203, NUDOCS 9805210271 | |
| Download: ML20247H858 (13) | |
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e U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS CRITICALITY SAFETY INSPECTION REPORT i
Docket No.:
70-1113 License No.:
SNM-1097 Report No:
70-1113/98-203
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Licensee:
General Electric Company General Electric Nuclear Energy Production Location:
Wilmington, NC Dates:
May 4 - 8,1998 l
Inspector:
Dennis Morey, Criticality Safety inspector 1
i Approved by:
Philip Ting, Chief Operations Branch Division of Fuel Cycle Safety and Safeguards, NMSS l
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Enclosure 1
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9905210271 990513 PDR ADOCK 07001113 C
GENERAL ELECTRIC NUCLEAR ENERGY PRODUCTION NRC INSPECTION REPORT l
70-1113/98-203 J
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EXECUTIVE
SUMMARY
Intruductiun The NRC performed a routine, unannounced criticality safety inspection at the General Electric Nuclear Energy Production facility (GENE) in Wilmington, North Carolina, from May 4 - 8, 1998. The objective of the inspection was to review the adequacy of the licensee's criticality safety program. The inspector reviewed plant activities, reliability of controls, verification of analytical assumptions, operating pro'cedures, internal audits and inspections, and configuration
,I management.
During this inspection, the inspector identified one non-cited violation (NCV) concerning an out-of-date posting, and two inspector followup items (IFI) concerning reliability of controls and
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verification of analytical assumptions. The licensee was determined to have a strong program in the areas of configuration management and emphasis on engineered controls.
Results The licensee emphasis on engineered safety features was found to be a strength.
A non-cited violation was identified concerning an out-of-date nuclear safety requirement document found posted in a work area.
An isolated incident was identified where the licensee had not established the reliability and availability of an engineered safety feature required by criticality safety analysis.
A weakness was identified in the licensee pre-operational inspection to determine that as-built conditions matched analytical assumptions where the licensee assumed a specific material composition in order to establish safe geometry.
e Internal audit items are not formally trended unless they are classified as Potential Non-Compliances (PNC). The failure to trend or reanalyze frequently violated nuclear safety requirements ic a minor weakness in the internal audit program.
e Uncontrolled posting of safety documents in the plant was found to be a weakness.
The licensee configuration management program is adequate and was found to be a strength.
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REPORT DETAILS e
1.0 Plant Onerations a.
Inspection Scope License Section 3.9.1 requires that licensed material activities be conducted in accordance with management control programs described in administrative and general plant practices approved and issued by cognizant management at a level appropriate to the scope of the practice.
During the course of the inspection, the inspector conducted walkdowns of plant operating areas and observed compliance with criticality safety limits and controls. Areas reviewed included ammonium di-urinate processing, powder preparation, uranium recovery, radwaste processing, fuel assembly, and the dry conversion process.
b.
Observations and Findings The inspector observed that the licensee has placed an emphasis on engineered safety features and has minimized administrative controls. Where administrative controls are required to be posted, they are specified in a Nuclear Safety Release / Requirements (NSR/R) document. Most criticality safety postings were generic and identified either critical valves and their position or moderator controls. The inspector determined that the licensee emphasis on engineered safety and the simplified criticality safety postings were adequate and were good practices.
Licensee Practice and Procedure (P/P) 40-04, Nuclear Safety Design Criteria, defines two levels of control on moderation as Moderation Control Area (MCA) and Moderation Restricted Area (MRA). In an MCA, the introduction of sufficient moderator coupled with failure of another, independent control, may result in criticality and, in an MRA, the introduction of sufficient moderator alone may result in a criticality. The actual limits on moderation for panicular MCAs and MRAs vary and are described in the NSR/R for the facility, equipment, or operation. Design requirements are also part of MCA and MRA controls and are listed in NSR/R for the specific facility. The inspector determined that use of these zones has the positive effect of greatly reducing the number of posted criticality safety requirements in the plant.
The inspector conducted walkdowns in selected MRAs and MCAs and observed that it was difficult to gauge compliance with specific Requirements in an area without having the NSR/Rs affecting that area immediately available. The inspector questioned 3
employees in the areas regarding what the specific criticality controls for the area were and determined that operators were sufficiently familiar with criticality requirements in I
their areas.
1 The licensee does not formally control postings in the plant when the posting is not I
specifically mandated by an NSR/R. The inspectors noted that NSR/Rs are posted in various locations in the plant and some individual criticality controls are posted informally. The licensee stated that the NSR/Rs th'.; mc, rt:d in the plant are controlled through the document control system and are tracked according to the area where they are posted. The inspector reviewed several NSR/Rs tha, were observed to be posted in the plant and identified one example of a posted NSR/R with an outdated revision. The NSR/R 02.13.14 had been posted for over nine years and was over two years out of date.
The NSR/R was removed immediately, and the licensee stated that the correct version was in the area in a notebook. The inspector noted that none of the safety requirements on the NSR/R had been changed by the revision, and the unauthorized posting had low safety significance. The licensee is currently in the process of computerizing all NSR/R documents and has made a significant effort to remove outdated documents from the facility.
This failure constitutes a violation of minor significance and is being treated as a non-cited violation, consistent with Section IV of the NRC Enforcement Policy. The presence of an unauthorized NSR/R in a work area is NCY No. 70-1113/98-203-01.
c.
Conclusions The licensee emphasis on engineered safety features is a strength. A non-cited violation was identified due to the failure to remove an out-of-date NSIUR that was posted in a work area of the plant.
1 2.0 Reliability of Controls a.
Inspection Scone l
License Section 6.2.1 requires, in part, that " Prior to use in any process, nuclear criticality safety controls are verified against criticality safety analysis criteria."
The inspector reviewed controls described in selected NSR/Rs against their associated plant operation or equipment to ensure that the controls were reliable, available, and would function as required or assumed in the analysis.
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Observations and Findings The inspector reviewed the NSR/R requirements for the exhaust scrubber and compared NSR/R requirements to analysis. The exhaust scrubber removes uranium from the air by passing the air through a water saturated filter. The exhaust scrubber is covered by analysis for the roof scrubber dated March 21,1989. The analysis defines two controls as
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required for safety under the analysis. Control No.1 is Geometry / Mass which is met l
because the ground / plane area of the scrubber will accommodate the entire contents of a l
UF. cylinder. Control No. 2 is Mass which is met because the 2 g:dlon per minute bleed l
rate (the rate at which water is introduced into and removed from the scrubber) results in a complete change of scrubber water in approximately 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> thereby assuring that a i
critical mass will not accumulate in the body of the scrubber.
The NSR/R for the exhaust scrubber requires a flowrate of 120 gallons per hour (gph) or two gallons per minute (gpm). Requirement No. 2 states that the total drain liquid flow rate must be approximately 120 gph. The licensee stated that the flow rate was confirmed by a flowmeter in the system which had to read 120 gph. The licensee noted that the meter was calibrated when received, but was never checked or re-calibrated after receipt l
except to see ifit was reading out a value. If the meter appeared to be broken,it would be replaced. The licensee stated that it was not necessary to calibrate the meter because small changes in the flow rate would not impact the safety of the system; however, the inspector noted that this was not reflected in the analysis. The inspector noted that the safety requirements in paragraph VI at the end of the analysis differ slightly from the requirements in the NSR/R in that the analysis discusses the flow rate in the paragraph III controls section and leaves it out of paragraph VI potentially creating confusion about the significance of the meters.
The licensee perfonned a flow test of the equipment and determined that the flow rate was actually two gpm. The licensee also developed and implemented a monthly 1
maintenance requirement to check the flow rate on this equipment in both the Fuel Manufacturing Operation and Fuel Manufacturing Operation Expansion facilities. The requirement to periodically check the flow rate on the exhaust scrubber will be tracked as IFI No. 70-1113/98-203-02.-
c.
Conclusions Generally, required maintenance, calibration, and surveillance requirements ensure the reliability and availability of engineered safety features in the GENE plant. An isolated l
incident oflow safety significance was identified where the licensee had not established the reliability and availability of an engineered safety feature, a flowmeter, required by criticality safety analysis. The licensee has established surveillance requirements to assure the availability of the feattre.
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1 3.0 Criticality Analysis e
a.
Inspection Scone License Section 2.2.1.4 requires the criticality safety function to perform neutronics calculations, write criticality safety analyses, approve proposed changes in process conditions or equipment involving fissionable material, and specify criticality safety i
control requirements and functionality.
The inspector reviewed analyses for the gadolinia vacuum pump, the exhaust scrubber, j
and the fuel assembly inspection pit to determine the adequacy ofcriticality safety controls and verify that controls specified in the analyses were implemented in the plant.
b.
Observations and Findings In the gadolinia area, the inspector observed a vacuum system designed to transport UO 2 powder to a mill. The vacuum consisted of a 10-1/2" diameter filter housing and a 10-1/2" diameter silencer. The analysis of the silencer showed it to be slightly larger than a safe geometry. In order to meet the license requirement for k,aless than 0.97 under upset conditions, a 3/16" sheet of stainless steel was wrapped around the outside of the silencer. The stainless steel sheet makes the silencer geometrically safe at 5 wt%
enrichment due to absorption of neutrons in the steel. The analysis for the vacuum pump system, which included the silencer, was reviewed and was determined by the inspector to be of high quality. The analysis included a system description, objective documentation of double contingency, a listing ofimportant assumptions, calculational results including discussion of bias, and a listing of safety requirements that may be compared to implementation documents such as procedures or training requirements.
Because the analysis of the silencer used KENO stainless steel (SS) which is actually 304 stainless steel, the inspector requested documentation that the metal material wrapped around the silencer was actually stainless steel. The licensee stated that the material was obviously stainless steel due to its appearance and magnetic properties. Also, the licensee stated that the procurement process assured that stainless steel or a reasonable substitute were purchased. The licensee did not have any documentation establishing the identity of the material wrapped around the silencer. The material apparently was procured from a vendor and immediately installed without going through any kind of receipt inspection.
A parallel example observed by the inspector was the procurement of annular tanks, known as accumulators, for the uranium recovery area. The analysis for the accumulators was reviewed and was found to be of high quality as noted for the previous analysis. The analysis of the annular tanks uses KENO SS to achieve an upset condition k,y that barely meets the license requirements. In this case, the licensee established, through documentation, that a receipt inspection process verifies the material composition of 6
tanks built by vendors. The licensee indicated that such inspection was not necessary for the silencer, because the silencer was being analyzed for a very remote upset condition a
and was not routinely in contact with uranium solutions. The inspector determined that the metal wrapped around the silencer appeared to be stainless steel; however, the pre-operational verification that analytical assumptions were satisfied was incomplete. This has low safety significance because of the evidence that the metal around the silencer is actually SS, the one percent margin in the calculational result, the licensee's three-sigma margin for safety limits, and the low likelihood of moderator intrusion. The licensee agreed to develop procedural requirements to ensure that material composition of components relied on for safe geometry were as arsumed in the criticality safety analysis.
Licensee procedural requirements for the verification of analytical assumptions will be tracked as IFI No. 70-1113/98-203-03.
c.
Conclusions Licensee criticality safety analyses are generally very thorough and is a strength. A weakness was identified in the licensee pre-operational inspection to determine that as-built conditions matched analytical assumptions where the licensee failed to establish the material composition required for safe geometry.
4.0 Onerating Procedutn a.
Inspection Scone License Section 3.9 requires that licensed material processing or activities will be conducted in accordance with properly issued and approved P/Ps, plant practices or operating procedures.
The inspector reviewed selected operating procedures against th:ir associated NSR/R to determine whether controls from the NSR/R had been incorporated into the working procedure.
b.
Observations and Findings l
The inspector determined that all activities observed were being performed under j
properly issued and approved procedures. Criticality safety requirements are identified by the licensee in a criticality safety analysis which is maintained subsequently in a l
change request file. Criticality safety requirements are then flowed to an NSR/R which lists the engineered and administrative requirements upon which criticality safety is based for the operation or equipment. When necessary, criticality safety requirements are j
flowed into operating procedures. The licensee relies primarily on training of operators i
in procedural requirements for the implementation of administrative controls in the 7
facility. During a plant walkdown, the inspector questioned an operator in a solution o
processing area and found him to be knowledgeable of administrative requirements for the area.
I The licensee operating procedures reference specific NSR/Rs associated with the work that the procedure performed, however, specific NSR/R requirements are not flagged in i
the procedure. The inspector did not identify any criticality safety requirements that were l
not incorporated into the appropriate procedures. The procedures that were reviewed for i
the dry conversion process demonstrated a strong preference for engineered controls for l
criticality safety. The inspector observed few administrative controls in the NSR/Rs.
The inspector noted two examples ofitems in NSR/Rs that had been assigned to operations when there was no action that could be performed by an operator. The first item concerned a requirement to ensure that the overflow line from certain process tanks is "open at all times." The inspector observed the equipment in question and noted that the overflow lincs were really engineered safety features and that the NSR/R requirement could not actually be performed. The second example is a requirement that the dimensions of a discharge hopper on a centrifuge be limited to a specified dimension I
when only one size discharge hopper is actually available. The safety significance of both of these NSR/R requirements is low since they are verified engineered safety features. The licensee indicated that NSR/Rs are occasionally complex, and a decision is made about which organization will be responsible for particular requirements. The licensee stated that review and clarification of requirements is a priority. The licensee agreed to review the requirements in question. All NSR/R requirements reviewed by the inspector were engineered features that were correctly implemented, and no improperly assigned actions were observed that alTected safety.
c.
Conclusions Licensee activities are performed under properly issued and approved procedures.
Organization of procedures and flowdown of criticality safety controls is adequate.
Instances were identified of criticality safety requirements assigned to operations which could not actually be performed by opcrators. The safety significance was low for the instances identified.
5.0 Internal Audits and Inspections a.
Insnection Scone License Section 3.6 requires that planned and scheduled internal and independent audits are performed to evaluate the application and effectiveness of management controls and implementation of programs related to activities significant to plant safety.
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License Section 3.6.1 requires that criticality safety audits be perfonned quarterly by members of the criticality safety function under the direction of the manager of the criticality safety function. Required corrective actions must be documented and tracked to completion.
The inspector reviewed results of past quarterly audits and interviewed a licensee auditor to determine whether the audits were conducted as required, findings identified, and corrective actions established and tracked to completion, b.
Observations and Findings The licensee procedure controlling nuclear safety audits is Nuclear Safety Instruction E-2.0, Intemal Nuclear Safety Audits. The inspector reviewed the procedure and licensee records of the past three quarterly audits. The licensee's intemal audit program involves performing an audit of the entire facility during a quarter with a segment done each week.
Findings are reported to the manager of Environmental llealth and Safety who determines if any finding is a Potential Noncompliance (PNC). PNCs are tracked until closure.
Observations, findings, and PNCs are reported to the affected area manager who must respond to PNCs within 30 days.
The inspector noted that a broad range of findings and observations are reported under the system. Findings are appropriately classified as PNC, and reports are sent to managers as required. All documents verifying closure of findings appeared to contain appropriate corrective actions. The inspector determined that the licensee's internal audit program was adequate.
During walkdowns in operating areas, the inspector noted that shops and operators are able to extract requirements from operating procedures or NSR/Rs and then post the requirements in the plant, if this is necessary to assist in compliance with the pmticular requirement. An example is the requirement not to stack more than five empty three or five gallon buckets. The posting limiting the stacking of buckets appears to be primarily the result of frequent violation of the requirement as identified by licensee's internal audits. The postings are not controlled by the criticality safety function. The licensee's l
criticality safety staffindicated that the practice was acceptable and had resulted from frequent violation of the stacking limit by temporary employees who had been in the area during contractor work. The inspector did not observe any conflict between the postings and analytical or NSR/R requirements but noted that there was no formal policy
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controlling the posting of NSR/Rs or individual criticality safety requirements in the
- plant, it appears to be a weakness that requirements important enough for operators to want posted or that are frequently violated, as determined by internal audits, are not reanalyzed or readdressed in the associated NSR/Rs. Documentation of quarterly audits does not 9
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demonstrate trending in this case. The licensee stated that, although there is no formal i
policy controlling the posting of NSR/Rs or individual criticality safety requirements in the plant, licensee internal inspections and audits ensure that only correct revisions or appropriate requirements are posted.
I c.
Conclusions The licensee's internal audit program is adequate and meets all license requirements although only PNCs are trended. The failure to trend or reanalyze frequently violated nuclear safety requirements is a minor weakness in the internal audit program.
Uncontrolled postings in the plant were also found to be a weakness. No safety issues were identified related to these weaknesses.
6.0 Configuration Management a.
Inspection Scope License Section 3.1.1 requires a formal configuration management process governed by written approved practices.
The inspector reviewed the licensee configuration management program to detennine whether changes to equipment or processes affecting criticality safety received an appropriate review by the criticality safety function.
b.
Observations and Findines I
The licensee configuration management program is controlled by Procedure P/P 10.10, Configuration Management Program - Fuel islanufacturing. The inspector reviewed the controlling procedure, interviewed the manager of the licensee configuration management program, and reviewed selected change request files. All licensee actions involving a change to a process or facility require the initiation of a Change Request Report (CRR).
The licensee procedure requires all CRRs to be reviewed by the criticality safety function to determine whether a criticality safety review of the proposed change is required. The licensee processes as many as 800 of these requests each year. The inspector reviewed selected records of completed CRRs and determined that all had received appropriate criticality safety function review. The inspector determined that the licensee configuration management program was strong and could be expected to preclude any change affecting criticality safety without appropriate review.
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The inspector reviewed procedural requirements controlling the decision by maintenance personnel to initiate a change request. Adequate like-kind replacement control was in l
place, and where vessels are not geometrically safe, a change request is always required.
The inspector determined that configuration controls on maintenance activities were adequate.
c.
Conclusions The licensee configuration management program is adequate and was determined to be a strength.
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l ITEMS OPENED. CLOSED. AND DISCUSSED j
Onened j
l NCV 70-1113/98-203-01 Concerns an out-of-date nuclear safety requirement document found posted in a work area.
l IFI 70-1113/98-203-02 Concems the failure to establish the reliability and availability of an engineered safety feature required by criticality safety analysis.
IFI 70-1113/98-203-03 Concems the failure to determine that as-built conditions matched analytical assumptions where the licensee assumed a specific material composition in order to establish safe geometry and comply with the license.
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M ANAGEMENT MEETINGS The inspector met with GENE management periodically during the inspection. The inspector i
presented the inspection scope and findings to members of the licensee's staff at the conclusion of the inspection on May 8,1998. The licensee acknowledged the findings presented.
i PARTIAL LIST OF PERSONS CONTACTED 1
General Electric Nuclear Energy J.E. Kline Manager, Manufacturing C.C. Tarrer Manager, Configuration Management C.J. Monetta Manager, Environment, Health and Safety L.E. Paulson Manager, Nuclear Safety S.P. Murray Manager, Facility Licensing C.M. Vaughan Facility Licensing J.T. Taylor Criticality Safety P.J. Vescovi Criticality Safety Nuclear Regulatory Commission Dennis Morey, Criticality Safety Inspector, NRC Headquarters 12
l ACRONYMS USED CRR Change Request Report GENE General Electric Nuclear Energy l
gph Gallons per hour I
gpm Gallons per minute l
IFl Inspector Followup Item KENO A criticality analysis computer code MCA Moderator Controlled Area MRA Moderator Restricted Area NCV Non-Cited Violation NSR/R Nuclear Safety Release / Requirement PNC Potential Non-Compliance P/P Practices and Procedures SS Stainless Steel 13 w________________
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