ML20206E437

Insp Rept 70-7002/99-03 on 990222-0406.Violations Noted. Major Areas Inspected:Aspects of Plant Operations,Maint, Engineering & Plant Support

ML20206E437
Person / Time
Site:	Portsmouth Gaseous Diffusion Plant
Issue date:	04/29/1999
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20206E431	List: ML20206E428 ML20206E432 ML20206E437
References
70-7002-99-03, 70-7002-99-3, NUDOCS 9905050128
Download: ML20206E437 (15)

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U.S. NUCLEAR REGULATORY COMMISSION REGIONlli Docket No: 70-7002 Certificate No: GDP-2  ;

Report No: 70-7002/99003(DNMS)

Facility Operator: United States Enrichment Corporation Facility: Portsmouth Gaseous Diffusion Plant Location: 3930 U.S. Route 23 South P.O. Box 628 Piketon, OH 45661 Dates: February 22 through April 6,1999 Inspectors: D. J. Hartland, Senior Resident inspector C. A. Blanchard, Resident inspector i Approved By: Patrick L. Hiland, Chief Fuel Cycle Branch l Division of Nuclear Materlats Safety l

9905050128 990429 PDR ADOCK 07007002 I l

PDR

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l EXECUTIVE

SUMMARY

United States Enrichment Corporation Portsmouth Gaseous Diffusion Plant l

NRC Inspection Report 70 7002/99003(DNMS)  !

l This inspection report includes aspects of plant operations, maintenance, engineering, and '

plant support. The report covers a six week period of routine resident inspections. i l

Plant Operations l e The inspectors identified two issues associated with Technical Safety Requirement implementation. One previously identified issue with the smoke detector logic at the withdrawal station was not adequately resolved to address all circuit failure scenarios. l The second issue regarded a 1997 analysis that was not formally evaluated and '

supported by nuclear criticality safety approvals. (Section 01.1)

Maintenance and Surveillance e The inspectors identified that ir.strument mechanics and associated first line managers did not clearly understand the actions necessary to perform independent verification activities. The certificatee's immediate corrective actions appeared adequate.

(Section M1.1) e The inspectors identified a violation, in that work activities were not performed in accordance with applicable confined space entry and radiation work permits.

(Section M1.2)

Enaineerino e The inspectors determined that the certificatee did not always evaluate the impact of revised nuclear criticality safety approval analysis and controls on existing nuclear criticality safety approvals. The certificatee provided interim guidance to nuclear criticality safety staff regarding identification and implementation of more conservative controls during the revision of existing or the development of new nuclear criticality safety approvals. (Section E1.1)

• The inspectors determined that the certificatee systematically evaluated the design and installation of process piping hangers and structures in accordance with a methodology developed in the 1980s for the seismic evaluation at nuclear power plants. The plant staff satisfactorily self-corrected identified piping structure concems while conducting the seismic evaluation. (Section E4.1) 2

Report Details I. Operations 1

01 Conduct of Operations 01.1 Technical Safety Reauirement imolementation Problems

a. insoection Scooe (88100) i The inspectors reviewed operations to ensure compliance with Technical Safety Requirements (TSRs).

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a. Observations and Findinas During the inspection period, the inspectors identified two issues regarding TSR implomentation: I e On March 5, a certificatee quality assurance (QA) inspector documented in ,

- Problem Report (PR)-PTS-99-01315 an inadequate statement in the Safety Analysis Report (SAR) regarding the description of the smoke detection systems located at the withdrawal stations. The SAR stated that the block and safety valves would automatically close to isolate the pigtail when any two smoke heads actuated in a coverage area. The QA inspector noted that the SAR statement was inadequate because the automatic valve closure would occur only when two detectors over the same withdrawal position alarmed upon detecting a uranium hexafluoride (UF.) release, in response, the certificatee agreed to revise the wording in the SAR.-

On March 10, during review of the PR, the NRC inspectors noted the actuation logic for the smoke heads was not consistent with the actions required by TSR 2.5.3.4.

Specifically, TSR 2.5.3.4A allowed the certificatee to complete filling a cylinder if one smoke detector for the withdrawal position became inoperable after initiating the l evolution. However, if the smoke head did not fail safe, i.e., the failure mode was an

- open circuit, the actuation of the second smoke head would not result in actuation of the pigtailisolation system.

After discussions with plant staff, the certificatee took immediate action to issue guidance to isolate the cylinder and withdrawal manifold within 15 minutes if one smoke head was rendered inoperable. This was consistent with actions required by TSR 2.5.3.4B when both smoke detectors for a withdrawal position were inoperable.

The certificatee also intended to change the logic so that the actuation of a single smoke head would actuate the system. During followup, the inspectors noted that the system engineer identified the same issue in March 1998, as documented on PR 98-01933, but no actions were apparently taken to address the issue.

The certificatee reviewed maintenance records since March 1998 and identified an example where cylinder filling operations were allowed to continue after a smoke hesd at Low Assay Withdrawal (LAW) was declared inoperable. As a result, on March 24, 1998, the certificatee made a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> notification to the NRC due to a TSR safety system not being available and no redundant equipment being available to perform the 3

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required safety function.' Upon further review of the maintenance records, the certificatee determined that the condition would not have prevented the system from operating as required. The inoperable smoke head failed in the alarm mode providing .

one of the two logic signals required for actuation.

Therefore, the failure to take corrective actions when the problem with the system logic i was identified in March 1998 is a violation of minor safety significance and is not subject l to formal enforcement action.

o On March 23, the inspectors identified an apparent discrepancy between XP2-CO-CA2030," Operation of the Criticality Accident Alarm System," and the applicable TSRs for Criticality Accident Alarm System (CAAS)in the cascade buildings. Specifically, the procedure allowed for movement of cylinders, vessels, wastes, or equipment containing up to 350 grams of uranium enriched to greater than 1 percent when the CAAS system was inoperable. The TSRs restricted handling or moving waste containing uranium 235 enriched to 1 percent or greater.

The basis for the procedural latitude was an informal TSR analysis documented in an internal memo dated September 16,1997. The analysis addressed the need to perform sampling operations during CAAS outages. The analysis identified that CAAS coverage was required by nuclear criticality safety approval (NCSA) for sampling activities, but the certificatee did nc identify the need to revise the NCSAs to implement the analysis.

Following discu< .1 with plant staff, the certificatee immediately put a hold on the mcvement of trx tial greater than 15 grams, the minimum designated as a fissile operation, duri' CAAS outages until the evaluations were completed.

The inspectors determined that the failure to document and approve in an NCSA the safety bases for the movement of a limited quantity (less than 350 grams) of fissile material is of minor safety significance and is not subject to formal enforcement action.

The inspectors also determined that the acceptability of moving limited quantities of .

materials, under the current TSR, may require further discussions with the NRC's l

licensing staff. ~

c. Conclusions The inspectors identified two issues associated with TSR implementation. One previously identified issue with the smoke detector system logic at the withdrawal station was not adequately resolved to address all circuit failure scenarios. The second issue involved a 1997 analysis that was not formally evaluated and supported by existing NCSAs.

08 Miscellaneous Operations issues 08.1 Certificatee Event Reports (90712)

The certificatee made the following operations-related event reports during the inspection period. The inspectors reviewed any immediate safety concerns indicated at the time of the initial verbal notification. The inspectors will evaluate the associated written reports for each of the events following submittal, as applicable.

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Number Qate Status Title 35403 02/24/99 Open Safety System Actuation in Building X-330 Tails Mezzanine smoke head SSWE.

i 35507 03/25/99 Closed Safety System Failure, LAW 1 A Pyrotronics I smoke detector was inoperable on July 8, l 1998, with cylinder filling operations in )

progress. The event notification was retracted on April 6,1999. The inspectors determined that the retraction was l warranted. (Section 01.1) 35513 03/25/99 Open Safety System Actuation in Building X-330 Cascade Automatic Data Processing (CADP) smokehead at Cell 31-3-3.

08.2 Bulletin 91-01 Reoorts (97012)

The certificatee made the following reports pursuant to Bulletin 91-01 during the inspection period. The inspectors reviewed any immediate NCSA concerns associated )

with the report at the time of the initial verbal notification. Any significant issues emerging from these reviews are discussed in separate sections of this report or in future inspection reports.

Number Qalg Tjlig 35458 03/11/99 4-Hour Report - NCSA-0705-038 dealing with the I disassembly of equipment was determined to be deficient.

35516 03/26/99 24-Hour Report - Buliding X-330, ACR No.2, F-can and GP can were discovered to be spaced less than the l required 23 inches center-to-center in violation of NCSA requirements.

08.3 (Closed) VIO 70-7002/98007-02: With one uranium hexafluoride (UF.) cylinder high temperature channel inoperable, the certificatee transitioned Autoclave No. 5 from Mode ill to Mode IV which was a TSR violation. The certificatee determined that the root cause of the violation was inattention to detail by the front line manager (FLM).

Specifically, when the operators notified the FLM of the safety system actuation, the FLM incorrectly concluded that Autoclave No. 5 was in Mode IV and directed the operators to perform sampling operations rather than review Procedure XP4-TE-FD2705, " Liquid Sampling in X-342 and X-343." On June 21,1998, the certificatee performed a "Use of Proadures," stand-down to ensure all employees understood management's expectation to conducted safety related activities in accordance with approved procedures. The inspectors consider the violation closed.

08.4 (Closed) CER 70-7002/98-03: Cascade Automatic Data Processing (CADP) smoke detector S64 alarmed in Building X-330 Area Control Room (ACR) No. 2. The certificatee determined that the safety system actuated from an outgassing of UF, caused by a leak from a brazed joint on the buffer line supplying air to the X-165 5

expansion joint on Cell 31-2-6, Stage 9. The certificatee's root cause investigation determined that a poor braze joint (the joint was brazed between 1978 and 1983) caused the failure. Currently, the plant operators were required to satisfactorily pass a certification test prior to performing instrument line brazing. The inspectors identified that the line was repaired and considered this item closed.

08.5 (Closed) Certificatee Event Reoort (CER) 70-7002/98-06: CADP smoke detectors SSWA and SSWE alarmed in Building X-330 ACR No. 2. The certificatee determined that the direct cause was an outgassing of UF, from valve SMDS-2, a 1-inch control valve located within the enclosed housing adjacent to withdrawal position No. 4. The certificatee determined that two aluminum gaskets, rather than the required one, were installed between the bonnet and valve body which allowed a path for the flow of the UF, out of the valve body into the atmosphere. The certificatee determined that an error in the valve installation work instruction was the root cause for the installation of two gaskets. The work instruction was revised and proceduralized. The inspectors noted that the procedure clearly addressed installing one gasket between the bonnet and valve body. The inspectors verified that the actions had been effective in preventing recurrence of the event and the item was closed.

08.6 (Closed) CER 70-7002/98-12: During a routine safety and fuel stop, the truck driver noted that overpack locking pins were not in the container holes of the Model UX-30 packages. The certificatee determined that the Portsmouth Traffic Department inspector had visually checked the trailer to confirm that the overpack locking pins were installed before the shipment left Portsmouth Gaseous Diffusion Plant as the procedure required. However, the certificatee determined that a visual verification failed to ensure that the locking pins were fully engaged and functioning as designed when installed in the overpack. As a corrective action, the shipping procedure was changed to require the Traffic Department inspector to physically ensure that the overpack ball locking pins were locked into position. The inspectors consider the item closed.

II. Maintenance and Surveillance 1

M1 Conduct of Maintenance and Surveillance M1.1 Imolementation of Indeoendent Verification 1

a.' Insoection Scope (88103) l The inspectors observed instrument mechanics perform a TSR surveillance.

b. Observations and Findinas On March 18, the inspectors observed instrument mechanics (IMs) perform a TSR surveillance on autoclave (A/C) No. 2 in Building X-344 per Procedure XP4-TE-UH6716,

" Autoclave Semi-Annual Testing." The inspectors noted that the work instruction included a lifted lead and jumper sheet which invoked an independent verification activity per Procedure XP2-GP-GP1033," Lifted Leads and Jumpers." In discussions with the inspectors, the instrument mechanics explained that observing another IM restore a circuit was acceptable as an independent verification. Subsequent y, the inspectors determined that the IM FLMs did not clearly understand what constituted an adequate independent verification. Following the inspectors discussions with the IM 6

management, operations management stopped the A/C No. 2 surveillance activity and A/C No. 2 was declared inoperable pending a resolution to the procedure and the independent verification issues.

The inspectors held discussions with plant management in order to understand their expectations for independent verification. Plant management explained that an acceptable independent verification did not include observing the actions of an operator performing the initial alignment, instal!ation, or verification to determine the correct component identification, position, or condition. As immediate corrective action, the L l Maintenance Manager stopped all maintenance activities until a crew briefing was conducted to clarify the requirements for an independent verification, as required in maintenance Procedure M-19, " Independent Verification." A daily operating instruction (DOI) was also issued which clearly articulated the requirements for an independent verification. As a followup, the inspectors leamed that the certificatee was evaluating all procedures that required an independent verification and was developing )

a plant wide procedure, an XP2 level procedure, to further commursicate the  ;

requirements for an independent verification. The inspectors will track certificatee's l action regarding independent verification guidance as an inspection Followup Item j (IFl 70-7002/99003-01). i

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Conclusions:

The inspectors identified some plant staff did not fully understand the required actions necessary to perform an independent verification. The certificatee's immediate corrective actions appeared adequate. The inspectors will track the certificatee's actions to address the independent verification issue as an IFl.

M1.2 Observation of Maintenance Activities in Buildina X-330

a. Insoection Scooe (88103)

The inspectors observed operations staff conduct troubleshooting activities for a cell in Building X-330.

b. Qbservations and Findinas On March 25, the inspectors observed operations staff preparing to inspect Cell 31-3-3 for a UF, process leak. During the job briefing, the building management (BM) reviewed selected aspects of the confined space permit with the entry team (three operators and an attendant). Areas reviewed included the following: 1) cell air sample test results;

2) required personnel protective equipment (PPE); 3) emergency notification process;

4) posting and controlling the confined space entry location; and 5) the expected scope 1 of the inspection. The inspectors noted that the BM only briefly discussed the radiation work permit (RWP) requirements and did not address the method that would be used to ensure continuous communication between the attendant and operators as specified in l the confined space entry permit. The inspectors noted that the entry team signed the confined space entry permit at the conclusion of the job briefing.

I The inspectors observed the operators satisfactorily donned PPEs, as required by the i confined space entry permit, and the attendant established the confined space entry location in a suitable location. The confined space entry permit requirements included:

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1) eye protection; 2) safety shoes; 3) respirator; 4) full anti-contamination clothing; and !

5) the performance of the required respirator fit test. The inspectors noted that the attendant displayed the required entry postings and barriers. The inspectors also noted the cell door was only partially open, hindering visual contact between the attendee and operators in the cell housing.

Prior to the operators' entry into the cell housing, the inspectors noted poor lighting and poor general housekeeping in the cell housing. In discussions with the inspectors, an operator explained that although the lighting was poor in cell housings; the operators were expected to use a flashlight to navigate within the cell housing. Subsequently, inspectors and operators observed, through cell windows (small covers on the side of the ceil housing for observation), that three of the six celllights were inoperable. In addition, the inspectors identified a bag of debris and numerous instrument hanger felt liners scattered on the cell floor. Plant management later informed the inspectors that the plant staff planned to address the noted deficiencies.

The inspectors reviewed a copy of the confined space entry permit. The confined space  !

entry permit required that the attendant maintain voice and visual communications with the operators in the cell housing. However, the inspectors observed that the attendant failed to maintain voice or visual communications with the operators in the cell housing i and did not immediately raise the issue regarding this limitation to plant management.

As previously discussed, the cell housing door was only partly opened; cell housing lighting was limited; and, the configuration of the process equipment within the cell i housing did not allow for visual or voice communications between the attendant and operators in the cell housing. In discussions with the inspectors, the attendant agreed ~

that visual and voice communication between the attendant and operators were not maintained due to the identified deficiencies.

Technical Safety Requirement 3.9.1, requires that written procedures shall be prepared, reviewed, approved, implemented, and maintained to cover activities described in SAR, Section 6.11.4.1, and listed in Appendix A to SAR Section 6.11.

2 Procedure XP2-SH-lH1032, " Confined Space Program," required, in part, that for work activities conducted in confined spaces, a confined space entry permit must be issued. l The procedure listed the requirements needed to conduct work activities in confined '

space areas. One such requirement was that the attendant maintain voice and visual communication with the operators inside the cell housing. The failure of the attendee to maintain voice and visual communication with the operators inside the cell housing was {

1 identified as a Violation (VIO 070 07001/99003-02a).  !

The inspectors noted that operators entering Cell 31-3-3 also failed to follow the  !

instructions specified in general RWP No. 99-330-0008-3-G. Specifically, PR-99-01898 identified that RWP No. 99-330-0008-3-G, "X-330 Cell Housing Entry," required air samples and contamination surveys prior to entry into a cell housing. Procedure XP2-XP-HO1032, " Radiological Work Permit," established the method for implementing, using, and documenting the RWP program which was used to control radiological work activities. The RWP program requirements applied to all plant staff requiring access to  !

radiological areas. The certificatee classified the cell housing as a radiological area.  !

RWP No. 99-330-0008-3-G included two hold points that required health physics (HP) to

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conduct specific actions to ensure that plant staff entering the cell housing were not {

exposed to radiological hazards. The two hold points required an air sample and I i

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contamination survey prior to entry into the cell housing. The inspectors noted that the operators entered the cell housing without HPs performing the required air sample and contamination surveys. The failure to perform required air sample and contamination surveys, in accordance with RWP No. 99-330-0008-3-G, was identified as a Violation (VIO 070-07001/99003 02b).

In addition, the inspectors identified that the process building RWPs viere inconsistent l and had limited guidance. Specifically, the specialinstruction section of RWP No. 99-330-0008-3-G, used for cell house entry in Building X-330, indicated that a, "Special RWP" may be required for cell housing entrance with confined space. The inspectors noted that the BM did not request a special RWP for entry into Cell 31-3-3 on March 25. Rather, the BM indicated to the inspectors that the general RWP covered entry into a cell housing for all inspection activities which included inspecting a cell with a known UF, process leak after the cell was shut down and below atmospheric pressure (the status of Cell 31-3-3), in addition, the inspectors noted that RWPs for the same activities in other process buildings had different requirements. Specifically, the hold points for entry into a cell housing in Building X-330 required air sample and contamination surveys before entry but no surveys were required in the RWP for Building X-333. In response to the inconsistent and limited guidance concerns, the i HP Supervisor closed all RWPs in the process buildings (X-326, X-330, and X-333) and directed HPs to develop new RWPs that required the BM to contact HP management before initiating a confined space entry.

I As an immediate corrective action, plant management temporarily halted all confined space work, and initiated crew briefings and required reading to address the confined i

space and RWP issues. The crew briefings and required reading addressed the l circumstances surrounding this incident, along with a review of the plant's confined space entry program, {

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Conclusions:

1 During the observation of ongoing activities, the inspectors concluded that work j

activities were not performed in accordance with applicable confined space entry and i RWPs. One violation was identified with two examples.

M8 Miscellaneous Maintenance issues M8.1 (Closed) CER 070-7002/99-03: On February 19,1999, in response to a CAAS trouble alarm in Building X-700, the certificatee discovered the nitrogen cylinder was valved off to the hom in that building. With the valve closed, the system would have been unable to perform the intended safety function. The certificatee was not able to determine how the valve was shut. Maintenance was performed on the system 3 days earlier; however, records indicated that the valve was independently verified open at the completion of the maintenance activity.

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o As an immediate action, the certificatee verified that all other CAAS nitrogen cylinder valves in the plant were open. As followup action, the certificatee installed tamper indicating devices on all of the valves. Subsequent testing determined that the alarm overlay from adjacent buildings provided adequate and audible detection and warning-coverage. This non-repetitive, license-identified and corrected violation is being treated as a Non-Cited Violation (NCV), consistent with Section Vll.B.1 of the NRC Enforcement Policy, lit. Enaineerina E1 Conduct of Engineering E1.1 Nuclear Criticality Safety Reauirement imolementation Evaluation

a. -Insoection Scooe (88020)

The inspectors reviewed the certificatee's implementation of the NCS corrective action plan.

b. Observations and Findinas On March 1, during the moming meeting, the certificatee discussed an NCSA prepared for an abandoned piece of equipment. The inspectors noted that one additional control was to be implemented by the NCSA beyond the controls currently in place. The additional controlinvolved the erection of a 10-foot boundary around any equipment discovered to be abandoned and not covered by an NCSA until a non-destructive analysis (NDA) of the equipment could be performed.

After the meeting, the inspectors questioned the NCS staff why a similar boundary was not required for existing abandoned equipment. NCS staff replied that the existing equipment was determined to be either geometrically safe or the plant staff had performed an NDA to qualify the uranium deposit. In response to questions regarding the handling of additional abandoned equipment discovered before the new NCSA was implemented, NCS staff issued a DOI that provided interim guidance for addressing NCS issues with abandoned equipment.

In response to the broader issue regarding the identification of more conservative controls during the revision of an existing or a new NCSA, an additional dol was issued i to the NCS staff. The guidance required that if a new control was the result of a newly l Identified and previously unanalyzed contingency, then a PR wou!d be written. The operation then would be stopped or else appropriate compensatory action would be l taken. The certificatee intended to revise plant procedures to include the guidance l contained in the DOI.

c. Conclusion As followup to the inspectors' issues, the certificatee provided interim guidance to the NCS staff regarding the identification of more conservative controls during the revision of an existing or the creation of a new NCSA.

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I E4 Engineering Staff Knowledge and Performance E4.1 Pioina. Structure. and Hanaer Deslan and Condition

a. Insoection Scooe (88Q2Q)

The inspectors reviewed the certificatee's method to evaluate, ensure, and maintain piping, structures, and hangers. The inspectors compared, for accuracy, the installation of piping, structures, and hangers with required design parameters and as-built

~ installation drawings for selected piping systems. The condition of selected piping supports and hangers was examined,

b. Observation and Findinos The inspectors noted that the certificatee reviewed ' process piping hangers and supports throughout the plant site. The certificatee performed the review as part of the SAR '

upgrade effort. The review was conducted using the seismic qualification utility group (SQUG) methodology developed in the 1980s for seismic evaluations of nuclear power ,

plants. The SQUG methodology relied on screening techniques such as experience '

data and seismic qualification protocol. As part of pipe seismic evaluations,

- representative piping systems in the process buildings and support facilities were walked-down to verify piping supports and hanger integrity. The effort identified that some equipment and pipe did not meet the seismic requirement of 0.05 g peak ground acceleration assigned to the plant based on a 250-year return earthquake. For example, the review found some heating, ventilation, and air conditioning (HVAC) ducts mounted with _C-clamps directly above a withdrawal UF, cylinder pigtail and a  !

UF. accumulator hanger in Building X-326 that was inadequately designed to meet the assigned seismic load. The inspectors noted that the certificatee took the appropriate actions to correct the seismic concems identified in the review. The certificatee's corrective actions were completed on February 13,1999, in addition to reviewing the certificatee's plant seismic evaluations, the inspectors observed the condition, placement, size, and types of supports and hangers for recycle cooling water (RCW), steam, air, Freon, UF process, and sprinkler piping in Buildings X-330, X-333, X-326, and X-343. The selected pipe hangers and supports observed were in satisfactory condition and were installed per the installation drawings  !

except for steam condensate hangers. The inspectors noted some corroded steam condensate pipe hangers that appeared structurally degraded. Through a review of prs, the inspectors determined that the degraded steam condensate pipe hangers were fixed in a timely manner. In addition, the inspectors noted a maintenance practice which could potentially fatigue an RCW Winch vent pipe. The Winch RCW vent pipe ran from the Freon condenser to the 8-inch RCW retum pipe. During maintenance activities, the inspectors ~ observed that a spool piece, connecting the Sinch RCW pipe on the Freon condenser with the return 8-inch RCW pipe, was removed. When the spool piece was

~ removed, the inspectors observed that the Sinch vent pipe was required to absorb the energy produced between the vibrating 8-inch retum RCW pipe and Freon condenser, in response to that observation, the engineering organization was assigned to evaluate the stresses imparted on the Sinch vent pipe by the inherent building vibrations. In addition, the inspectors noted that the certificatee was developing work instructions to remove the Winch RCW vent pipe when the 8-inch spool piece was removed. The inspectors noted that the remaining pipe systems observed were supported in 11

accordance with the American Society of Mechanical Engineers (ASME) Sections B31.1 and B31.3 and the National Fire Protection Association Code, Section 13.

The inspectors reviewed an engineering report, POEF-2038, " Cell Floor Vibration Evaluations and Remedial Actions Committee,"lasued on November 18,1992, that was

. performed, in part, to determine the effect building vibrations had on process building equipment and structures. The report identified that equipment vibrations produced structural stress levels well below the allowable limits for structural members (building

. structure, piping hangers,' and structures) and indicated no long term structural fatigue effects due to equipment vibrations. The inspectors reviewed select engineering report i calculations and determined 1 hat the certificatee systematically analyzed stresses imposed on structural members. The vibration amplitudes were transmitted into worst '

case structural deflections. The engineering staff then calculated the structural member

. stress imposed by maximum deflections. The engineering report determined that no  ;

short or long-term' safety concems existed for the structural members (building structure, piping hangers, and structures) from building vibrations.

The inspectors noted that the certificatee's configuration control process included ensuring that equipment and piping modifications were evaluated for seismic considerations prior to approval. The evaluation process included ensuring that pipe hangers and supports were designed and installed in accordance with ASME Sections B31.1 and B31.3 and the National Fire Protection Association Code, Section 13.

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Conclusion:

The inspectors noted that the certificatee systematically evaluated the design and installation of process piping hangers and structures in accordance with a methodology developed in the 1980s for seismic evaluations of nuclear plants. The inspectors found that the responsible lead seismic engineer was knowledgeable of requirements for hanger and structure designs and had thoroughly documented design calculations. The inspectors identified maintenance work practices that had the potential to fatigue a RCW vent line. Engineering staff evaluated the issue and initiated work instructions to resolve this specific issue. The inspectors noted that the plant staff satisfactorily corrected identified piping structures concerns when identified, IV. Plant Suonort S8 Miscellaneous Security Issues S8.1 - Certificatee Security Reports (90712)

The certificatee made the following security-related,1-hour reports pursuant to 10 CFR 95 during the inspection period. The inspectors reviewed any immediate security concerns associated with the reports at the time of the initial verbal notification.

Rgtg Title 02/26/99 Classified matter discovered on an unclassified computer system that is shared between Portsmouth and Paducah and potentially accessible by uncleared personnel.

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03/25/99 Classified information found in unclassified environmenta! bulletin in Building X-100.

03/25/99 Building X-100 engineering service order was discovered to contain classified information.

04/06/99 Building X-100 classified document was left unattended on desk.

V. Manaaement Meetinas X1 Exit Meeting Summary The inspectors presented the inspection results to members of the facility management on April 6,1999. The plant staff acknowledged the findings presented. The inspectors asked the plant staff whether any materials examined during the inspection should be considered proprietary. No proprietary information was identified.

PARTIAL LIST OF PERSONS CONTACTED 1

Lockheed Martin Utility Services

• S. Casto, Work Control Manager

• S. Fout, Operations Manager

• R. Helme, Engineering Manager

• P. Musser, Enrichment Plant Manager United States Enrichment Corooration

• J. Brown, General Manager

• L. Fink, Safety, Safeguards & Quality Manager

• P. Miner, Regulatory Affairs Manager l

• Denotes those present at the exit meetir.g on April 3,1999.

INSPECTION PROCEDURES USED IP 88100: Plant Operations IP 88020: Nuclear Criticality Safety IP 88103: Maintenance Observations IP 97012: In-office Reviews of Written Reports on Nonroutine Events

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ITEMS OPENED, CLOSED, AND DISCUSSED Ooened 070-7002/99003-01 IFl ' Follow actions to correct independent verification guidance.

070-7002/99003-02_ VIO The plant staff failed to implement confined space and radiation work permit requirements Sr worker safety.

35403 CER Safety System Actuation in Building X-330 Tails Mezzanine smoke head SSWE.

35507 CER Safety System Failure, LAW 1 A Pyrotronics smoke detector was inoperable on July 8,1998 with cylinder filling operations in progress. The event notification was retracted on April 6,1999. The inspectors determined that the retraction was warranted. (Section 01.1) 35513 CER Safety System Actuation in Building X-330 Cascade Automatic Data Processing (CADP) smokehead at Cell 31-3-3.

Closed 70-7002/98007-02 VIO TSR Violation During Sampling Evolution.

70-7002/98-03 CER Cascade Automatic Data Processing UF, smoke detector actua'ed at the tails withdrawal station in Building X-330.

70-7002/98-06 CER Small outgassing of ur'nium hexafluoride occurred in the tails withdrawal area in Lailding X-330.- l 70-7002/98-12 CER Overpack locking pins on GE UX-30 packages were not properly installed during shipment. l 070-7002/m' M CER Criticality accident alarm system nitrogen operated hom, a slaved system, was discovered with nitrogen cylinder valved off in Building X-700.

35507 CER Safety System Failure, LAW 1 A Pyrotronics smoke detector was inoperable on July 8,1998 with cylinder filling operations in progress. The event notification was retracted on April 6,1999. The inspectors determined that the retraction was warranted. (Section 01.1)

Discussed None 14

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l LIST OF ACRONYMS USED A/C Autoclave ACR Area Control Room ASME American Society of Mechanical Engineers BM Building Management CAAS Criticality Accident Alarm System CADP' Cascade Automatic Data Processing  ;

CER Certificate Lvent Report CFR Code of Federal Regulations DNMS Division of Nuclear Material Safety ,

DOI Daily Operating Instruction 1 FLM First Line Manager HP Health Physics l

HVAC Heating, Ventilation, and Air Conditioning IFl Inspection Followup Item IM instrument Mechanic LAW Low Assay Withdrawal NCS Nuclear Criticality Safety NCSA Nuclear Criticality Safety Approval NCV Non-Cited Violation NDA Non-destructive Analysis NRC Nuclear Regulatory Commission PDR Public Document Room PPE Personnel Protective Equipment l PR Problem Report QA Quality Assurance RCW Recycle Cooling Water RWP Radiation Work Permit SAR Safety Analysis Report SQUG Seismic Qualification Utility Group TSR Technical Safety Requirement UF, Uranium Hexafluoride USEC United States Enrichment Corporation VIO Violation s

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