ML20206B565

Notice of Violation from Insp on 990322-26.Violation Noted: on 981209,Building X-326 Operator Did Not Stop Cell Motors from Fastest Location Upon Unexplained Rises in Motor Amp Load Which Lead to Excessive Overload in Stage 2

ML20206B565
Person / Time
Site:	Portsmouth Gaseous Diffusion Plant
Issue date:	04/22/1999
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20206B564	List: ML20206B561 ML20206B565 ML20206B570
References
70-7002-99-06, 70-7002-99-6, NUDOCS 9904290294
Download: ML20206B565 (4)

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NOTICE OF VIOLATION l l

United States Enrichment Corporation Docket No. 70-7002 Portsmouth Gaseous Diffusion Plant Certificate No. GDP-2 During an NRC special inspection conducted from March 22 through 26,1999, violations of NRC requirements were identified. In accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions," NUREG-1600, Revision 1, the violations are listed below:

1. Technical Safety Requirement 3.9.1 requires, in part, that written procedures shall be implemented for activities listed in Appendix A to Safety Analysis Report, Section 6.11.

Appendix A listed the startup, operation, and shutdown of cascade cells; nuclear criticality safety; and investigations and reporting as activities requiring written procedures. In addition, A. Procedure XP4-CO-CA3900C, Revision 0, " Control of Damaged Centrifugal Compressors," effective date of December 24,1997, Section 2.0, "Immediate Actions," Step 2.1 required, in part, an operator to stop affected cell motors from the fastest location upon excessive stage overload or an unexplained rise in motor amp load. Step 2.2 required the operator to take the cell offstream

[ isolate} in accordance with Procedure XP4-CO-CN2102C. Procedure XP4-CO-CN2102C referred the operator to Procedure XP4-CO-CN2410 which contained guidance on how to isolate Cell 25-7-2 in Building X-326 from the rest of the cascade. j B. Procedure XP4-EG-NS1-25, 'NCS Nuclear Criticality Safety Response to Anomalous Conditions," Revisions 0, Change B, effective date November 30, { '

1998, required, in part, that the nuclear criticality safety staff shall: 1) determine if an anomalous condition involved an unanalyzed condition; 2) assess the safety

. significance of an as-found condition; 3) identify the nuclear criticality safety controls affected by the anomalous condition; and 4) complete the anomalous I

) condition report within the reportability time-frame for the condition. I C. Procedure UE2-RA-RE1030,

• Nuclear Regulatory Event Reporting," Revision 2, Change C, effective date March 3,1997, required, in part, the Plant Shift Superintendent to determine whether an event or condition was reportable to the NRC according to the criteria listed in Appendix D of the procedure. Appendix D, criteria A.3.a., c(1), and c(3) specified, in part, that a report shall be made to the NRC, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from initial observation, for operations that do not comply with the double contingency principle [i.e., operations that are singly contingent),

for which moderation is used as the primary criticality control and that involve:

1) the occurrence of any unanalyzed event for which the safety significance of the event or corrective actions to re-establish the approved controls are not readily identifiable; or 2) the controlled parameter and the control on the parameter cannot be re-established within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the initial observation of the event.

9904290294 990422 PDR ADOCK 07007002  ;

C PDR

V. .

Notice of Violation I D. Procedure XP4-CO-CA2182, " Control d Large Air inleakage," Revision 1, Change C, effective date February 10, il 97, and XP2-SH-IS1034, " Accident l

I Prevention \ Equipment Control Tags," Revision 0, Change A, effective date July 7,1997, required, in part, that the plant staff shall take specific actions to control (tagging and logging) recirculating cooling water to shutdown cells and shall implement specific temporary repairs for events which result in large inleakages to the cascade.

Contrary to the above, A. On December 9,1998, a Building X-326 operator did not stop cell motors from the fastest location upon unexplained rises in motor amp load which lead to an excessive stage overload in Stage 2 of Cell 25-7-2. In addition, the operator did not refer to or immediately isolate the cell from the rest of the cascade in accordance with Procedures XP4-CO-CA3900C, XP4-CO-CN2102C and XP4-CO-CN2410. (VIO 70-7002/99006-01a)

B. On December 9,1998, the nuclear criticality safety staff did not: 1) identify that an anomalous condition resulting from the Cell 25-7-2 fire involved an unanalyzed condition; 2) incorporate into the safety evaluation the potential presence and impact of a deposit within the cell; 3) properly identify the nuclear criticality safety controls affected by the anomalous condition; and 4) complete the anomalous cordition report within the reportability time-frame (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) for the conditions present. (VIO 70-7002/99006-01b)

C. On December 9,1998, the Plant Shift Superintendent did not make a notification to the NRC within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of initial observation of tne loss of moderation control for a singly contingent activity, operation of cascade Cell 25-7-2, as a result of holes in the cascade piping and components which precluded the implementation of moderation control using a dry gas buffer, an unanalyzed condition, for which corrective actions were not readily identifiable.

(VIO 70-7002/99006-01c)

D. On December 9,1998, the plant staff did not tag and log the status of Cell 25-7-2-related recirculating cooling water valves and pipe spool pieces, controlled for nuclear criticality safety-related purposes, and did not implement temporary repairs to the cascade as specified in Procedures XP4-CO-CA2182 and XP2-SH-IS1034 for a cell that was shut down as a result of a fire which caused a large inleakage of air into the cascade. (VIO 70-7002/99006-01d)

This is a Severity Level IV violation (Supplement VI). (VIO 70-7002/99006 01)

Notice of Violation 2. Technical Safety Requirement 3.9.1 requires, in part, that written procedures shall be prepared, reviewed, approved, and implemented to cover activities listed in Appendix A to Safety Analysis Report, Section 6.11. Appendix A listed cellload alarms and cell

. coolant alarms as examples of cascade area control room alarms requiring written procedures for " abnormal operation / alarm response."

Contrary to the above, as of December 9,1998, the certificatee had not prepared, reviewed, approved, and implemented alarm response procedures for alarms such as cell load and cell coolant alarms in the cascade area control rooms.

This is a Severity Level IV violation (Supplement VI). (VIO 70-7002/99006 02)

3. Title 10 of the Code of Federal Regulations, Part 76.93, " Quality Assurance," requires, in part, that the Corporation shall establish and execute a Quality Assurance Program.

Section 2.16, Appendix A, Section 1, item 1.16, and Appendix A, Section 2, item 2.16 of the Quality Assurance Program required, in part, that conditions adverse to quality are identified and corrected as soon as practical.

Contrary to the above, as of March 22,1999, A. The plant staff did not identify and correct, through an August 1998 condition adverse to quality corrective action plan, a failure by some staff to implement portions of Procedure XP2-GP-GP1040, " Equipment History Program."

Specifically, the plant staff, in August 1998, developed and implemented a corrective action plan to ensure the full implementation of Procedure XP2-GP-GP1040 which did not identify the need for or include corrective actions to ensure that responsibilities assigned to the Reliability Engineering Manager were implemented. These responsibilities included the pre-implementation review of work packages and the identification of equipment failures requiring evaluation and failure cause analysis. (VIO 70-7002/99006-03a)

B. ' The plant staff did not identify and correct, through a March 1999 significant condition adverse to quality corrective action plan, inconsistencies between the Emergency Plan and Emergency Plan implementing Procedure XP2-EP-EP1050. Specifically, the plant staff, in March 1999, developed and implemented a corrective action plan to ensure consistency between the Emergency Plan and the Emergency Plan implementing Procedures which did not resolve inconsistencies. The unresolved inconsistencies between the Emergency Plan and Emergency Plan Implementing Procedure included the classification level and initiating conditions for severe wind and security emergencies. - (VIO 70-7002/99006-03b).

This is a Severity Level IV violation (Supplement VI). (VIO 70-7002/99006-03)

Notice of Violation -4 Pursuant to the provisions of 10 CFR 76.70, United States Enrichment Corporation is hereby required to submit a written statement or explanation for items 1 through 3 above to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, D.C.

20555,' with a copy to the Regional Administrator, Region ill, and a copy to the NRC Resident inspector at Portsmouth, within 30 days of the date of the letter transmitting this Notice of Violation (Notice). This reply should be clearly marked as a " Reply to a Notice of Violation" and should include for each violation: (1) the reason for the violation, or, if contested, the basis for disputing the violation, (2) the corrective steps that have been taken and the results achieved, (3) the corrective steps that will be taken to avoid further violations, and (4) the date when full compliance will be achieved. _Your response may reference or include previously docketed correspondence, if the correspondence adequately addresses the required response.

If an adequate reply is not received within the time specified in this Notice, an order or a Demand for Information may be issued as to why the Certificate should not be modified, suspended, or revoked, or why such other action as may be proper should not be taken.

Where good cause is shown, consideration will be given to extending the response time.

if you contest this enforcement action, you should also provide a copy of your response to the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C.

20555-0001.

Because your response will be placed in the NRC Public Document Room (PDR), to the extent possible, it should not include any personal privacy, proprietary, or safeguards information so that it can be placed in the PDR without redaction if personal privacy or proprietary information is necessary to provide an acceptable responts, then please provide a bracketed copy of your response that identifies the information that should be protected and a redacted copy of your response that deletes such information. If you request withholding of such material, you mW11 specifically identify the portions of your response that you seek to have withhold and provide in detail the bases for your claim of withholding (for example, explain why the disclosure of information will create an unwarranted invasion of personal privacy or provide the information required by 10 CFR 2.790(b) to support a request for withholding confidential commercial or financialinformation). If safeguards information is necessary to provide an acceptable response, please provide the level of protection described in 10 CFR 73.21.

In accordance with 10 CFR 19.11, you may be required to post this Notice within two working days.

Dated this 22nd day of April 1999 i