United States Nuclear Regulatory Commission - Protecting People and the Environment

EA-03-077 - River Bend 1 (Entergy Operations, Inc.)

December 29, 2003


Paul D. Hinnenkamp
Vice President - Operations
River Bend Station
Entergy Operations, Inc.
P.O. Box 220
St. Francisville, LA 70775


Dear Mr. Hinnenkamp:

The purpose of this letter is to provide you the final results of our significance determination of the preliminary White finding identified in the subject inspection report. The inspection finding was assessed using the Significance Determination Process and was preliminarily characterized as White, a finding with low to moderate increased importance to safety that may require additional NRC inspections. This White finding involved a failure to properly lock open River Bend Station Condensate Prefilter Vessel Bypass Flow Control Valve CNM-FCV200 in May 2002. This performance deficiency resulted in a loss of feedwater flow to the reactor on September 18, 2002, when Valve CNM-FCV200 unexpectedly closed following a reactor scram.

At your request, a Regulatory Conference was held on June 23, 2003, to further discuss your evaluation of this issue. During the meeting, your staff acknowledged the performance deficiency and described your assessment of the risk significance of the finding. In a supplemental letter dated July 9, 2003, you provided additional information regarding your risk evaluation of this event. In your July 9, 2003, letter, you agreed that the failure to control the position and properly lock Valve CNM-FCV200 was a performance deficiency and a violation of your Technical Specifications; however, you took exception to certain aspects of NRC's evaluation of risk associated with this event. After considering all of the information available, as explained further in the attached enclosures, the NRC has concluded that the finding is appropriately characterized as White.

In the supplemental information provided on July 9, 2003, you restated your assertion, presented during the Regulatory Conference, that the risk associated with this event would be very low, making this a finding characterized at a Green level. This assertion was based on your belief that it is inappropriate for NRC to use the Individual Plant Examination of External Events (IPEEE), in concert with "best effort" estimations, for the purpose of determining risk for inspection findings in today's regulatory environment without more detailed analyses to improve precision.

Specifically, you asserted that the overall risk associated with this event, including the change in core damage frequency from fire, was very low because: (1) the safety systems in the plant were functional, including the control rod drive system, which would have provided a high pressure injection source after the first 6 hours; (2) Valve CNM-FCV200 would have failed only during a plant scram and not during a controlled manual shutdown, as evidenced by the July 2002 plant shutdown; and (3) the fire risk from a fire area is nonexistent for evaluation of this event, if there is no plant scram caused by a fire in that area. While we took into account the first two considerations in our independent assessment of the risk of this event, we disagree with your assertion that fire risk from a fire area is nonexistent for the evaluation of this event. The basis for our position is discussed in greater detail below and in Enclosure 2 to this letter.

In your supplemental response, you indicated that the NRC's use of your IPEEE results, together with "best effort" estimations, was not appropriate for the purpose of evaluating the risk of inspection findings. However, NRC Manual Chapter 0609, "Significance Determination Process," Appendix A, Attachment 1, step 2.5, "Screening for the Potential Risk Contribution Due to External Initiating Events," states that the impact of external initiators should be evaluated and could increase the risk significance of a finding by as much as one order of magnitude. Step 2.5 also states that the evaluation may be qualitative or quantitative in nature. Qualitative evaluations of external events should, as a minimum, provide the logic and basis for conclusion and should reference all the documents reviewed. The NRC has qualitatively assessed the significance of the external events contribution to the risk of this finding. Additionally, quantitative methods used by the staff indicate that external factors would increase the risk significance of the subject finding by at least a factor of two over risk caused by internal initiators alone. More detail regarding our evaluation is contained in Enclosure 2 to this letter.

Your supplemental response indicated that it is inappropriate to import the results of the IPEEE screening method into the Significance Determination Process without fully appreciating the context in which they were developed. We agree that IPEEE data should be used carefully and that importing results directly from the IPEEE for those items that were screened in the process would result in significant overestimation of the risk. However, the results of the IPEEE were not directly imported for use in our preliminary evaluation. We reviewed your IPEEE to identify those fire areas in which feedwater was important to risk. In evaluating the change in risk from fire initiators in those 18 areas, our preliminary evaluation utilized your model of record to obtain quantitative results as opposed to directly importing the results from the IPEEE evaluation. Additionally, while industry and NRC tools for evaluating the risk associated with external initiators are not fully developed, the Significance Determination Process requires that we evaluate the total risk associated with a finding using the best available information.

For the risk determination under consideration, you contend that the external event contribution from various potential risk initiators, as well as numerous specific areas within the plant, screened out as being insignificant based on the IPEEE screening criteria. As a result, you believe these potential risk initiators and areas should not be used to adjust the risk of a specific internal event, such as the one in question. We have determined that the contribution to risk of selected external events, such as high winds, tornados, and hurricanes; transportation hazards; severe weather storms; and lightning, should not be excluded from consideration simply because they screened out during the initial development of your IPEEE.

In your assessment of the risk of the subject event, your staff chose to refine the assumptions used in the IPEEE for the fire areas that we specifically evaluated in our preliminary assessment. Your stated purpose was to demonstrate that the original screening criteria were correct and that these events should be screened as not significant to the risk analysis for this event. While we agree that increasing the precision of the analyses is appropriate, we have concluded, as described in our preliminary risk assessment, that the affected areas still contribute to increasing the overall risk of the event as described in our preliminary risk assessment. Additionally, your analysis of the impact of fires within the plant did not fully analyze the potential for fires to cause indirect reactor scrams. Your analysis used assumptions for fire severity factors, fire sizing, ignition frequencies, and fire modeling that were not fully supported by the information provided. Also, the increased risk to the plant from increased probability of human error as a result of the fires was not evaluated. Of the fire areas at the River Bend Station, only 32 were actually analyzed. The remaining areas were either quantified using generic industry data, assuming similarities to the 32 analyzed or, in one instance, was not assessed. Therefore, we conclude that you have provided an insufficient basis for determining that the increase in risk associated with fires was insignificant. A more detailed description of our evaluation of your risk assessment is included in Enclosure 2 to this letter.

After considering the information developed during the inspection, the information presented at the regulatory conference on June 23, 2003, and the additional information you provided in your letter dated July 9, 2003, the NRC has concluded that the risk significance of the subject inspection finding should be based on our preliminary risk assessment further supported by our assessment described in Enclosure 2. The assessment in Enclosure 2 is intended to address each of the points presented by Entergy Operations, Inc. during the regulatory conference and provide our position that those points did not provide a basis for concluding that the issue should be characterized as Green. Accordingly, NRC has concluded that the finding is appropriately characterized as White, an issue with low to moderate increased importance to risk, which may require additional NRC inspections or other NRC actions.

You have 30 calendar days from the date of this letter to appeal the staff's determination of significance for the identified White finding. Such appeals will be considered to have merit only if they meet the criteria given in NRC Inspection Manual Chapter 0609, Attachment 2.

The NRC has also determined that the failure to lock open Valve CNM-FCV200 properly is a violation of Technical Specification 5.4.1.a, as cited in the enclosed Notice of Violation (Notice). The circumstances surrounding the violation are described in detail in the subject inspection report. In accordance with the NRC Enforcement Policy, NUREG-1600, the Notice of Violation is considered escalated enforcement action because it is associated with a White finding.

You are required to respond to this letter and should follow the instructions specified in the enclosed Notice when preparing your response.

Because plant performance for this issue has been determined to be in the regulatory response band, we will use the NRC Action Matrix to determine the most appropriate NRC response for this event. We will notify you, by separate correspondence, of that determination.

In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter, its enclosures, and your response will be made available electronically for public inspection in the NRC Public Document Room or from the NRC's document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html. To the extent possible, your response should not include any personal privacy, proprietary, or safeguards information so that it can be made available to the Public without redaction. The NRC also includes significant enforcement actions on its Web site at www.nrc.gov; select What We Do, Enforcement, then Significant Enforcement Actions.



Bruce S. Mallett
Regional Administrator

Docket: 50-458
License: NPF-47

1. Notice of Violation
2. NRC Evaluation of Inadequately Secured Condensate Valve

cc w/enclosure:

Senior Vice President and
   Chief Operating Officer
Entergy Operations, Inc.
P.O. Box 31995
Jackson, MS 39286-1995

Vice President
Operations Support
Entergy Operations, Inc.
P.O. Box 31995
Jackson, MS 39286-1995

General Manager
Plant Operations
River Bend Station
Entergy Operations, Inc.
P.O. Box 220
St. Francisville, LA 70775

Director - Nuclear Safety
River Bend Station
Entergy Operations, Inc.
P.O. Box 220
St. Francisville, LA 70775

Wise, Carter, Child & Caraway
P.O. Box 651
Jackson, MS 39205

Mark J. Wetterhahn, Esq.
Winston & Strawn
1401 L Street, N.W.
Washington, DC 20005-3502

Manager - Licensing
River Bend Station
Entergy Operations, Inc.
P.O. Box 220
St. Francisville, LA 70775

The Honorable Richard P. Ieyoub
Attorney General
Department of Justice
State of Louisiana
P.O. Box 94005
Baton Rouge, LA 70804-9005

H. Anne Plettinger
3456 Villa Rose Drive
Baton Rouge, LA 70806

West Feliciana Parish Police Jury
P.O. Box 1921
St. Francisville, LA 70775

Michael E. Henry, State Liaison Officer
Department of Environmental Quality
Permits Division
P.O. Box 4313
Baton Rouge, LA 70821-4313

Brian Almon
Public Utility Commission
William B. Travis Building
P.O. Box 13326
1701 North Congress Avenue
Austin, TX 78701-3326



Entergy Operations, Inc.
River Bend Station
  Docket:  50-458
License: NPF-47

During an NRC inspection concluded on November 14, 2002, a violation of NRC requirements was identified. In accordance with the "General Statement of Policy and Procedure for NRC Enforcement Actions," NUREG-1600, the violation is listed below:

Technical Specification 5.4.1.a requires that written procedures be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978.
Regulatory Guide 1.33, Revision 2, Appendix A4, "Procedures for Startup, Operation, and Shutdown of Safety-Related BWR Systems," Item n., lists "Condensate System (hotwell to feedwater pumps, including demineralizers and resin regeneration)."
System Operating Procedure SOP-0007, "Condensate System," Revision 21, required Condensate Prefilter Vessel Bypass Flow Control Valve CNM-FCV200 to be locked open.
Contrary to the above, on September 18, 2002, Valve CNM-FCV200 failed closed as a result of not having been properly locked open, as required by System Operating Procedure SOP-0007, "Condensate System." As a result, the feedwater flow transient resulting from a reactor scram on September 18, 2002, caused Valve CNM-FCV200 to close unexpectedly, causing a complete loss of feedwater flow to the reactor pressure vessel.
This violation is associated with a White Significance Determination Process finding.

Pursuant to the provisions of 10 CFR 2.201, Entergy Operations Inc. is hereby required to submit a written statement or explanation to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555 with a copy to the Regional Administrator, Region IV, and a copy to the NRC Resident Inspector at the facility that is the subject of this Notice, 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; EA-03-077" and should include for each violation: (1) the reason for the violation, or, if contested, the basis for disputing the violation or severity level, (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 previous 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 license should not be modified, suspended, or revoked, or why such other action as may be proper should not be taken. Where good cause is shown, consideration will be given to extending the response time.

If you contest this enforcement action, you should also provide a copy of your response, with the basis for your denial, to the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, DC 20555-0001.

Because your response will be made available electronically for public inspection in the NRC Public Document Room or from the NRC's document system (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html, to the extent possible, it should not include any personal privacy, proprietary, or safeguards information so that it can be made available to the public without redaction. If personal privacy or proprietary information is necessary to provide an acceptable response, then please provide a bracketed copy of your response that identifies the information that should be protected and a redacted copy of your response that deletes such information. If you request withholding of such material, you must specifically identify the portions of your response that you seek to have withheld and provide in detail the bases for your claim of withholding (e.g., explain why the disclosure of information will create an unwarranted invasion of personal privacy or provide the information required by 10 CFR 2.790(b) to support a request for withholding confidential commercial or financial information). If safeguards information is necessary to provide an acceptable response, please provide the level of protection described in 10 CFR 73.21.

Dated this 29th day of December 2003

Page Last Reviewed/Updated Monday, July 06, 2015