ML20198B436

Insp Rept 70-1257/98-204 on 981109-13.Violation Noted. Major Areas Inspected:Open Item Completion,Review of Recent Licensee Criticality Safety Event,Validation & Criticality Safety Analysis

ML20198B436
Person / Time
Site:	Framatome ANP Richland
Issue date:	12/14/1998
From:	NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
To:
Shared Package
ML20198B408	List: ML20198B404 ML20198B419 ML20198B436
References
70-1257-98-204, NUDOCS 9812180165
Download: ML20198B436 (14)

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U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS CRITICALITY SAFETY INSPECTION REPORT i

- Docket No. 70-1257 License No. SNM-1227

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Report No. 70-1257/98-204 4;#=R

-e Licensee: Siemens Power Corporation Location: Richland, Washington 99352 Inspection Dates: November 9 - 13,1998 Inspectors: Dennis Morey Criticality Safety Inspector Santiago Parra Criticality Safety Inspector

' Approved By: Philip Ting, Chief Operations Branch Division of Fuel Cycle Safety and Safeguards T* 1 Enclosure 2 m

7; 9812190165 981214 "l ' 1 POR ADOCK0700g27

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, SIEMENS POWER CORPORATION NRC INSPECTION REPORT ,

70-1257/98-204 EXECUTIVE

SUMMARY

Introduction The Nuclear Regulatory Commission (NRC) performed a routine, announced criticality safety inspection at the Siemens Power Corporation (SPC) located in Richhnd, Washington, from November 9 - 13,1998. The inspection focused on open item completion, review of a recent licensee criticality safety event, validation, and criticality safety analyses. The inspectors reviewed documents, interviewed plant staff, and conducted walkdowns of affected plant areas.

'During the inspection, the inspectors identified a violation concerning the failure to remove fissile material from disassembled plant equipment in accordance with procedures, opened two Inspector Followup Items (IFIs), and closed two open items from a previous inspection.

Results

• The inspectors identified a 6!ation regarding the stacking of removed equipment containing fissile material without a required inspection.

The licensee took immediate and effective corrective action to correct an obscured criticality safety posting.

e The inspectors determined that results reported in nuclear criticality safety evaluations (NCSEs) are conservative, e The inspectors determined that licensee validation is adequate.

e The inspectors also determined that the present criticality alarm system monitors will reliably detect the minimum accident of concern.

e The inspectors noted a weakness associated with licensee training of plant staffin criticality safety requirements.

• The inspectors observed a discrepancy in licensee incident categorization.

i e The licensee control of contractor activities during the decommissioning of ADU l Conversion Line is considered a weakness. j 2

e The inspectors closed two open items from the previous inspection.

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• 1 l REPORT DETAILS 1.0 P_I_ ant Operations

a. Insnection Scope The inspectors performed walkthroughs of the Dry Conversion Facility, Powder Storage Vault, Pellet Manufacturing Area, and Pellet Grinding Area. In these areas, the licensee converts uranium hexafluoride (UF 6) into uranium dioxide (UO )2 Pellets. The inspectors performed a risk based inspection of the operations with a focus on the implementation of administrative controls,

b. Observations and Finding l

The primary control in the areas inspected is moderator exclusion, and the second control

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is commonly a mass or geometry limitation. The inspectors noted that the licensee limits  ;

mass in many areas through administrative controls such as spacing limits. Correct i implementation of these administrative controls is essential to dorSle contingency, and therefore, to assurance of safety for the operation.

The inspectors noted that a key Criticality Posting, located at the southern end of Room 100 in the Tray Loading and Stacking Area, was covered by a crane. The temporary storage of the crane is located in front of the criticality posting associated with the Finished Pellet Storage Racks. The inspectors observed that the posting contained instructions for the spacing of the carts, a key administrative control associated with limiting the geometry that can be achieved in a cart anay. The inspectors notified licensee personnel concerning the obscured posting. Licensee personnel immediately moved the criticality posting to a different location where it would be visible without any obstructions.

After observing the obscured posting, the inspectors interviewed operators in the Tray Loading and Stacking Area and noted that the operators did not know or understand moderation controls as listed in the criticality posting on the pellet carts. The cart postings are located underneath the carts in a location that is hard to see and read. The postings contain excessive information such as different load configurations and moderation control requirements. The operator did not appear to understand what moderation was or how to apply the moderator controls. The questions asked by the inspectors are related to general knowledge of plant operations by personnel which is considered a key safety control. These observations were brought to the licensee's attention for evaluation and appropriate corrective acticas. Licensee management acknowledged the possibility of a weakness in the area of training which is discussed below in Section 6.0.

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The inspectors observed an additional criticality control for the carts which was an administrative requirement to maintain a 12-inch edge-to-edge spacing between carts and other fissile materials. A red line on the floor was used to mark a 12-inch distance

, between the Finished Pellet Storage Area and the carts. The inspectors observed a cart l inside the red line in violation of the 12-inch spacing requirement. A review of the i criticality safety analysis report U400, Rev. O, " Pellet Storage Area," states that the 12-inch spacing requirement is not necessary for nuclear criticality control since the carts had been detemiined to be safe in an infinite array without the 12-inch spacing. Licensee j staffindicated that transfer of pellets to the storage area requires the carts to be moved i across the line which was the normal procedure for pellet transfer. The 12-inch administrative control was an additional administrative requirement intended to prevent i challenges to safety limits throughout the plant through a uniform control. The inspectors questioned the 12-inch requirement in areas where it was not needed or complied with. I The licensee responded that the issue would be investigated. The safety significance of a requirement that is not enforced and cannot or will not be complied with is that criticality i limits and controls throughout the plant are undermined. The inspectors determined that the posted criticality safety limit of a 12-inch edge-to-edge spacing should be removed from the nuclear criticality safety posting ifit will not be enforced. Licensee actions to achieve uniform implementation of facility-wide criticality safety limits will be tracked as IFI 70-1257/98-204-01.

c. Conclusions i

The inspectors identified a generic administrative criticality safety control that was not i uniformly enforced by the licensee. The safety significance of the issue is considered minor since specific criticality controls in the subject cases were determined to be enforced. The inspectors noted that the licensee took immediate and effective corrective action to correct an obscured criticality safety posting.

2.0 Criticality Analysis 1

a. Inspection Scope j l

Criticality safety analysis forms the basis for undeatanding and accepting the safety margin associated with fissile material operations. Based on issues raised recently at the ,

facility, the inspectors elected to review dry operations. The inspectors reviewed the i following licensee NCSEs:

U400, Rev. O, " Pellet Storage Area", and a

D830, Rev. 3," Dry Conversion Powder Preparation" J l

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b. . Observations and Firdmus I

! Determination of bias is safety significant because it forms the basis of the safety margin when criticality computer codes are used to analyze processes, equipment, or facilities. i l The bias used to determine the safety margin in the pellet storage area analysis was taken I l from NUREG/CR-0073, an NRC sponsored validation report. No independent l determination of bias was observed in the analysis and there was no relation between the licensee validation report and the bias used in the analysis. This situation is not anticipated l by the license or any licensee procedures. The inspectors determined that the bias had been developed independently by the analyst using NUREG data and a licensee statistical code. The inspectors considered that the procedure was adequate and reproducible with a weakness in licensee record keeping and documentation.

The inspectors determined that analysis for the handling and storage of pellets was conservative. The analysis was performed with the assumption that the system is fully moderated under its most reactive conditions.

The Dry Conversion Powder Preparation Area is a moderation control area. Possible accidental moderation intrusion into the powder preparation areas was examined. The criticality safety analysis for this area was reviewed to confirm that examination of possible upset conditions was adequate, limits and controls on moderation and enrichment were reasonable, and important equipment necessary for criticality control had been identified.

Bias for the Dry Conversion Powder Preparation Area criticality analysis was taken directly from the licensee validation report. The inspectors considered that the NCSE was adequate.

c. Conclusions The inspectors determined that results reported in NCSEs are conservative.

3.0 Material Accumulation Event

a. Inspection Scope At the request of Region IV, the inspectors reviewed a licensee reported event regarding improperly handled fissile material by subcontractors (URI 70-1257/98-05-01). In areas of the facility with dry processes, the licensee relies heavily on administrative controls.

Assurance of safety is contingent upon confidence that the licensee will maintain control of ongoing fissile material operations. Therefore, any loss of control of these operations is l

l safety significant. The inspectors reviewed the Shift Supervisor's Abnormai Event Log l

dated September 21,1998, the Incident Investigation Board (IIB) review of the criticality safety incident dated September 23,1998, the Siemens 30-day followup report to the Bulletin 91-01 dated October 20,1998, and Criticality Safety Specification UO50.

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b, Observations and Findings On September 22,1998, the licensee issued " Event Description for Bulletin 91-01 24-hr Report" for an event that occurred on September 21,1998. The report described the disassembly of the plant's ammonium diurinate (ADU) Conversion Line 1. During a routine plant inspection conducted while decommissioning operations were underway, a licensee criticality safety specialist noticed that a stacked section of removed ductwork had visible accumulations of solids. This was contrary to the plant's Criticality Safety Specification UO501 B.2. which requires that the disassembly crew have a process operator remove visible solids from removed equipment prior to placing the removed equipment in an accumulation area.

A contractor was engaged by the licensee to remove ductwork from the ADU Conversion L.ine 1. The licensee suspected that the ductwork would contain some fissile material and i longstanding licensee criticality safety requirements called for removed equipment to be l checked for fissile material before release for storage or disposal. Tne contractors involved i had been employed by the licensee on numerous occasions fbr many years and had been provided with a work package that contained the criticality safety requirements.

During a routine walkthrough of the Line I area, a licensee criticality safety engineer noticed approximately 100 feet of ductwork stacked in a corner of the Line 1 area with  ;

visible solids which were later shown to contain approximately 34 kilograms oflow I enriched uranium (LEU) mixed with ammonium nitrate. The total amount of LEU was later determined to be 9.2 kilograms. The subsequent licensee investigation revealed that the contractor had been advised that licensee fissile material handlers were not available to support the ductwork removal and the contractor had c!ccted to continue the operation by bagging and stacking the ductwork pending arrival oflicensee staff who could inspect it.

Licensee immediate corrective actions included shutting down the removal operations until fissile material was removed from the auctwork and affected employees had been trained on criticality safety requirements. IIB short term corrective actions for ADU included:

Training / retraining appropriate Manufacturing Engineering staff and contractors to the applicable criticality safety specifications for the work; Retraining Plant Support Operators that perform the inspection / cleaning; Conducting pre-job briefings of Plant Support supervision, operators, and contractors by Manufacturing Engineering personnel, Inspecting ADU Line 2 ductwork inspection ports for accumulations of uranium

, before t rtup of Line 2.

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,Long term corrective actions included: )

l Evaluation of the relocation of Line 2 HEPA filters closer to the scrubbers and ,

replacing the ductwork using favorable geometry-  !

Evaluation of a better method to provide ductwork inspections, 1

A quarterly training program for Manufacturing Engineering personnel working on processes covered by Criticality Safety Specifications. l The ADU Conversion areas were not moderator controlled. The ADU process lines are j wet operation areas from which the moderator cannot be excluded. Double contingency j for thisjob relied on mass and geometry controls. By stacking the ductwork without first inspecting for and removing fissile material, the contractor defeated the geometry control.

The licensee appropriately identified the event as reportable under NRC Bulletin 91-01.

This event is safety significant in that only one criticality control remained and the loss of I control went undetected for approximately five days. The failure to control decommissioning operations to assure proper implementation of NCS controls is Violation 70-1257/98-204-02.

The licensee cited an inspection of the ductwork prior to removal as a mitigating factor in the safety significance of the event. The inspection of the ductwork wa i done through inspection ports with a borescope using a procedure that called for the operator to identify deposits of one inch or more. The licensee indicated that this inspection limited the amount of mass that could have been involved in the event. The inspectors determined that this inspection was not intended to prevent an accumulation of a critical mass at a specific location in the plant following removal of the ductwork. Rather, the inspection was intended to preclude an accumulation of a critical mass in the ductwork in its original 100 foot long installed configuration. The inspectors determined that the borescope inspection ecceptance criteria was not suflicient to preclude the formation of a critical configuration without the proper implementation of the specified geometry controls.

While investigating the ductwork event, the inspectors learned that a contractor had been removed from the licensee facility for moving a fissile material storage rack in contravention of configuration controls. An electrical contractor performing work in the ADU Conversion Line 1 area approximately two weeks afler the ductwork event needed access behind fissile material storage racks in the area. Without informing licensee staff, the contractor moved one of the storage racks containing several buckets of fissile material out ofits designed position. Upon discovery by licensee stafT, the rack was returned to its position and additional corrective actions were taken including conducting operator pre-job briefings at the work-site. While double contingency remained in efTect for the fissile material bucket storage because the movement of the rack had increased the spacing, this incident occurred only two weeks after the ductwork incident.

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The inspectors determined that there was a nexus between the two events in that control of licensed operations was lost. The two events raised concerns about the effectiveness of the licensee control of contractor activities for a large decommissioning operation involving multiple contractors and large quantities of fissile material over long periods of time. The licensee performance in this area is considered a weakness.

c. Conclusions The inspectors determined that the failure oflicensee staff to remove fissile material from removed ductwork was safety significant and that licensee corrective actions were not effective so that the event did not meet the criteria for a non-cited violation even though it was licensee identified. The licensee control of contractor activities during the decommissioning of ADU Conversion Line 1 is considered a weakness.

4.0 Validation

a. Inspection Scone The use of appropriate bias in NCS calculations is important to safety because it assures an adequate margin of safety with respect to system reactivity. The inspectors reviewed licensee document EMF-94-175, the licensee Validation Report, to verify that the bias used to support the safety margin based on computer calculations was technically adequate and that the validation report met license requirements.

b. Observations and Finding License Section 4.2.7.3 commits the licensee to perform a validation of computer calculation methods in accordance with American National Standards Institute (ANSI)/American Nuclear Society (ANS)-8.1 " Nuclear Criticality Safety in Operations with Fissile Materials Outside Reactors." ANS-8.1, Section 4.3.6 requires the preparation of a written report of the validation and requires that the validation report explicitly identify the area of applicability and the bias. Area of applicability is defined in ANS-8.1 as "The ranges of material compositions and geometric arrangements within which the bias of a calculational method is established." Bias is defined as "A measure of the systematic disagreement between the results calculated by a method and experimental data."

The inspectors did not observe an explicit description of the area of applicability in the licensee validation report. The licensee responded that the validation report is divided into homogeneous and heterogeneous sections and that the area of applicability is everything that fits within these categories that concerns LEU light water reactor fuel. The inspectors determined that the licensee had an area of applicability in mind in preparing the report but that it is too broad and is not explicitly stated. The significance of explicitly stating the area of applicability is that possible confusion over allowed enrichments, compositions, and other variables is eliminated.

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! The licensee validation report establishes bias only for homogeneous situations.

Heterogeneous situations must be validated on a case-by-case basis. License Section 4.2.1 allows the determination of bias on a case-by-case basis and allows for the use of positive bias (i.e. reducing the margin when the code is shown to systematically i over-predict km). These license conditions allow the analyst to determine the bias by a suitable method in each NCSE. The inspectors determined that the bias developed in the i licensee validation report is based on a reasonable method and is adequate. The inspectors  !

l also noted that bias used in NCSEs is not always taken from the validation report, I

particularly when heterogeneous calculations are involved.

L The inspectors determined that the licensee must document bias determination in the NCSE i in accordance with ANS-8.1 in every case where the validation report is not referenced. l This requirement would not preclude use of bias from a reasonable source such as an NRC 1 sponsored validation report but does preclude use of bias from any source other than the licensee validation repcrt without additional documentation. l The licensee agreed that the area of applicability and method for determining bias could be better stated in the validation report and agreed that the new validation report for SCALE 4.4 would improve in these areas. This new report is expected to be initiated within one 2 year. Initial development of the new licensee validation report will be tracked as IFI 70-1257/98-204-03. i l

While reviewing the license conditions associated with validation, the inspectors noted that j

' Chapter 4 of the license references Rqslatory Guide 3.14, which does not exist. The l licensee indicated that this was intended to refer to withdrawn Regulatory Guide 3.41 and I that the error would be corrected with the next license amendment. i i

c. Conclusions .

The inspectors determined that while the licensee does not always adequately document validation, no inadequate calculation or safety limit determination resulted.

5.0 Criticality Accident Alarm System

a. Inspection Scope The inspectors conducted a review to determine whether a criticality accident in the Uranium Oxide and Rod Handling and Bundle Assembly Areas could be detected.

Detection of an inadvertent criticality accident in these areas is important because of the number of worker normally in the areas during routine operations. The inspection included L a review of preventive maintenance and testing activities as well as the physical condition l of the criticahty accident alarm monitors in areas 131,108C and 189.

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, b. Observations and Findings l

The inspector observed that monitor placements are based on conservative nuclear i criticality accident assumptions Placement meets the minimum accident of concern criteria l

set forth in ANSI /ANS 8.3," Criticality Accident Alarm System"(CAAS) and 10 CFR 70.24. Licensee calculations performed on possible criticality accident scenarios in the Uranium Dioxide Building and in the Rod Handling and Bundle Assembly Building demonstrate that the monitors will respond as required. Monitors are placed such that at least two detectors cover the same area. All significant concrete and equipment stnmtures have been adequately considered in the placement of the monitors.

Because the monitors are BF3 neutron detectors and are susceptible to deterioration, the inspector reviewed preventive maintenance and calibration records. Records show that the monitors are tested with an AmBe neutron source every six months and monitor components are tested every two years.

c. Conclusions The inspector determined that the criticality alarm system (CAS) monitors in the Uranium Dioxide Building and Rod Handling and Bundle Assembly Building were adequately located and maintained. The inspectors also determined that the present criticality alarm system monitors will reliably detect an accident meeting the minimum accident of concern criteria.

6.0 Trainine a Insoection Scope The inspectors conducted a performance -based review of the licensee staff s knowledge of generic criticality safety controls used throughout the facility. Proper implementation of these controls had been identified by the licensee as a contributing factor in several recent events. Selected members of both the Operations and Engineering Departments were interviewed to determine their understanding of the controls used in their areas of responsibility.

b. Observations and Findings The inspectors performed walkthroughs of the ADU line #1 and #2 areas and asked operators, supervisors, and an operations engineer several questions regarding generic criticality safety controls in the plant. An example of a generic control in the licensee facility is a checkered line on the floor denoting a restriction on moderator material. An operator indicated that he did not understand the purpose of the line. Another example of a generic controlis a red line on the floor of the plant denoting a restriction on fissile material. An operations engineer was unsure of the purpose of these lines. The inspectors 10

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L i noted that the operations engineer in question was responsible for the shift briefing of

. ' contractor personnel prior to commencing decommissioning work in the facility.

i j . Licensee managers indicated that all of the generic items are covered in general plant training for criticality safety. The inability oflicensee operators and technical staff to recognize or understand facility-wide criticality controls is an indicator of a weakness with licensee criticality safety training. These observations were brought to the licensee's attention for evaluation and appropriate corrective actions. Licensee management i acknowledged the possibility of a weakness in the area of training.

- c. Conclusions The inspectors identified a potential weakness in the training program regarding plant employee understanding of the generic criticality safety requirements in effect for their i areas of responsibility.

7.0 Licensee Reporting Reauirements j

a. Inspection Scope The inspectors reviewed the licensee basis for determination of NRC Bulletin 91-01 reportability. I b, . Observations and Findings During review of a licensee event that was reported under Bulletin 91-01, the inspectors

- observed that licensee evaluation criteria included a determination that " moderation is the -

!- only control." The inspectors noted that the requirement in the Bulletin 91-01,

. Supplement 1, is really " moderation is the primary control." The determination is a subjective judgement whether moderation exclusion is the primary control over the 1 l process, equipment, or facility. j The inspectors determined that license procedure EMF-30, " Reporting Criticality Safety Violations, Incidents, and Emergencies," Appendix 11," Conditions that require Reporting to the NRC," under the immediate reporting criteria stated moderation as the only control as a criteria. The licensee acknowledged the discrepancy and indicated that the procedure would be upgraded. Licensee correction of reporting criteria will be tracked as

. IFI 70-1257/98-204-04.

c. Conclusions l The inspectors observed a discrepancy in licensee incident categorization which the licensee

. agreed to correct.

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8.0 Open Item Review

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i NCV 70-1257/98-201-01 i

1' Concerned the presence of unapproved work procedures, operator aids and an improperly

' documented parameter sheet in work areas. The plant supervisor communicated to staff

via weekly " Plant Operations Communications" dated Ma ch 23,1998, that operations are 4

to be performed only to approved procedures. The inspectors verified that affected operators had read the Plant Operations Communications weekly announcements.

Supervisors were reminded that it is their responsibility to perform weekly walk through

inspections of their areas and take immediate corrective action when discrepant activities or j postings are observed. This item is considered closed.

i l' IFI 70-1257/98-201-02 i

Concerned licensee action to upgrade older CSAs to the more rigorous standards of recent

_ CSAs. This item will remain open pending the completion of CSAs for all Category III i systems.

i IFI 70-1257/98-201-03 I

Concerned a weakness in the requirements for criticality safety review of field changes to engineering change notices (ECNs). After a thorough review of the ECN procedures, the inspectors determined that the current ECN procedures provide adequate safety. The two field changes that had been identified as not receiving adequate review by criticality safety specialists, had indeed been reviewed and approved by criticality safety specialists. This item is considered closed.

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l ITEMS OPENED. CLOSED AND DISCUSSED

- Opened IFI 70-1257/98-204-01 Tracks licensee actions to achieve uniform implementation of facility-wide criticality safety limits .

VIO 70-1257/98-204-02 Concerns the licensee failure to adequately control decommissioning operations.

IFI 70-1257/98-204-03 Tracks licensee development of a new validation report.

IFI 70-1257/98-204-04 Tracks licensee correction of NRC Bulletin 91-01 reporting criteria.

Closed l NCV 70-1257/98-201-01 Concerned the presence of an unapproved work procedures, operator aids and an improperly documented parameter sheet in a work areas.

1 IFI 70-1257/98-201-03 Concerned a weakness in the requirements for criticality safety '

review of field changes to ECNs. j Discussed IFI 70-1257/98-201-02 Concerned licensee action to upgrade older CSAs to the more rigorous standards of recent CSAs.

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l MANAGEMENT MEETINGS l l

The inspector met with SPC management periodically during the inspecuon. The inspector presented the inspection scope and findings to members oflicensee staff at the conclusion of the inspection on November 13,1998. The licensee acknowledged the fimdings presented.

j PARTIAL L.lST OF PERSONS CONTACTED Siemens Power Corporation L

Bernard Femreite Vice President of Manufacturing Bernie Bently Manager, Operations Loren Maas Manager, Regulatory Compliance  !'

JeffDiest Criticality Safety Nak Urza Manager, Manufacturing Technology l Cal Manning Criticality Safety j Andy McGehee Criticality Safety l Doug Kilian Manager, Manufacturing Engineering l

Nuclear Regulatory Commission Dennis C. Morey, Criticality Safety inspect r, NRC Headquaners Santiago A. Parra, Criticality Safety Inspector, NRC Headquarters j 3

ACRONYMS USED '

ANS American Nuclear Society ANSI American National Standards Institute ADU Ammonium Diurinate CAAS Criticality Accident Alarm System CSA Criticality Safety Analysis ECN Engineering Change Notice IFI Inspector Followup Item IIB Incident Investigation Board LEU Low Enriched Uranium NCSE Nuclear Criticality Safety Evaluations NRC Nuclear Regulatory Commission SPC Siemens Power Corporation UF6 Uranium Hexafluoride UO 2 Uranium Dioxide I

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