IR 05000454/2006006
| ML061220758 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 04/28/2006 |
| From: | Dave Hills NRC/RGN-III/DRS/EB1 |
| To: | Crane C Exelon Generation Co, Exelon Nuclear |
| References | |
| IR-06-006 | |
| Download: ML061220758 (20) | |
Text
April 28, 2006
SUBJECT:
BYRON STATION, UNITS 1 AND 2, NRC EVALUATION OF CHANGES, TESTS, OR EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT (IR) 05000454/2006006; 05000455/2006006 (DRS)
Dear Mr. Crane:
On March 24, 2006, the U.S. Nuclear Regulatory Commission (NRC) completed a combined baseline inspection of the Evaluation of Changes, Tests, or Experiments and Permanent Plant Modifications at the Byron Nuclear Power Station. The enclosed report documents the results of the inspection, which were discussed and others of your staff at the completion of the inspection on March 24, 2006.
The inspectors examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations, and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel. Based on the results of the inspection, two NRC identified findings of very low safety significance were identified, which involved violations of NRC requirements. However, because these violations were of very low safety significance and because they were entered into your corrective action program, the NRC is treating the issues as Non-Cited Violations in accordance with Section VI.A.1 of the NRCs Enforcement Policy.
If you contest the subject or severity of a Non-Cited Violation, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S.
Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with a copy to the Regional Administrator, U.S. Nuclear Regulatory Commission - Region III, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the Resident Inspector Office at the Byron Station.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
David E. Hills, Chief Engineering Branch 1 Division of Reactor Safety
Enclosure:
Inspection Report 05000454/2006006; 05000455/2006006 (DRS)
Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is a
REGION III==
Docket No:
50-454, 50-455 Report No:
05000454/2006006; 05000455/2006006 (DRS)
Licensee:
Facility:
Byron Station, Units 1 and 2 Location:
4448 N. German Church Byron, IL 61010-9750 Dates:
March 6, 2006 through March 24, 2006 Inspectors:
R. Daley, Senior Reactor Inspector, Team Leader A. Klett, Reactor Inspector Approved by:
D. Hills, Chief Engineering Branch 1 Division of Reactor Safety (DRS)
Enclosure
SUMMARY OF FINDINGS
IR 05000454/2006006; 05000455/2006006(DRS); 03/06/2006 - 03/24/2006; Byron Station;
Evaluation of Changes, Tests, or Experiments (10 CFR 50.59) and Permanent Plant Modifications.
The inspection covered a two week announced baseline inspection on evaluations of changes, tests, or experiments and permanent plant modifications. The inspection was conducted by two regional based engineering inspectors. Two Green Non-Cited Violations (NCV) were identified.
The significance of most findings is indicated by their color (Green, White, Yellow, Red), using Inspection Manual Chapter 0609, Significance Determination Process (SDP). Findings for which the SDP does not apply, may be Green, or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 3; dated July 2000.
A.
Inspector-Identified and Self-Revealed Findings
Cornerstone: Mitigating Systems
- Green.
The inspectors identified a Non-Cited Violation of 10 CFR Part 50, Appendix B,
Criterion III, Design Control, having very low safety significance for the licensees failure to correctly translate the design basis into procedures. Specifically, the licensee failed to update operator rounds to verify the revised design basis minimum value for essential service water flow to the component cooling water (CC) heat exchangers. In addition, because the operator rounds were not revised, the design basis minimum flow value was not bounded by the emergency operating procedure used for establishing initial cold leg recirculation in the event of a loss of coolant accident (LOCA). This issue was entered into the licensees corrective action program to revise the operator rounds.
The issue was more than minor because it was associated with the Mitigating System cornerstone attribute of Design Control, and affected the cornerstone objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to have operator rounds verify the design basis minimum service water flow or to have the emergency operating procedures ensure the minimum flow prior to establishing initial cold leg recirculation in the event of a LOCA could potentially have allowed the service water flow to be less than the required value to maintain the design heat load during post LOCA conditions. This finding was of very low safety significance because it screened out as Green using the SDP Phase 1 worksheet.
Even though the licensee did not control their bounding design basis service water flow procedurally, the flow to the CC heat exchangers has historically been well above the bounding design basis flow. (Section 1R02.1.b.1)
- Green.
The inspectors identified a Non-Cited Violation of 10 CFR 50.48(a)(1) having very low safety significance for the licensees failure to provide fire fighting systems of appropriate capacity and capability to minimize the adverse effects of fires on structures, systems, and components important to safety. Specifically, the licensee abandoned standpipes and manual hose stations located near safety-related equipment (essential service water makeup pumps) which reduced the fire suppression capacity and capability to protect such equipment. In addition, the site relied on a local fire department instead of the site fire brigade to manually suppress a fire that could have affected safety-related equipment. This issue was entered into the licensees corrective action program, and compensatory measures were taken This finding was considered more than minor because it was associated with the Mitigating System cornerstone attribute of Protection Against External Factors, and affected the cornerstone objective of ensuring the availability of systems that respond to initiating events to prevent undesirable consequences. Specifically, removing the manual hose stations reduced the fire suppression capacity and capability for protecting the emergency service water cooling tower makeup pumps and their diesels in the event of a fire (Section 1R17.1.b.1)
Licensee-Identified Violations
No findings of significance were identified.
REPORT DETAILS
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R02 Evaluations of Changes, Tests, or Experiments
.1 Review of 10 CFR 50.59 Evaluations and Screenings
a. Inspection Scope
From March 6 through March 24, 2006, the inspectors reviewed six evaluations performed pursuant to 10 CFR 50.59. The inspectors confirmed that the evaluations were thorough and that prior NRC approval was obtained as appropriate. The inspectors also reviewed 13 screenings where licensee personnel had determined that a 10 CFR 50.59 evaluation was not necessary. In regard to the changes reviewed where no 10 CFR 50.59 evaluation was performed, the inspectors verified that the changes did not meet the threshold to require a 10 CFR 50.59 evaluation. The evaluations and screenings were chosen based on risk significance, safety significance, and complexity.
The list of documents reviewed by the inspectors is included as an attachment to this report.
The inspectors used, in part, Nuclear Energy Institute (NEI) 96-07, Guidelines for 10 CFR 50.59 Implementation, Revision 1, to determine acceptability of the completed evaluations and screenings. The NEI document was endorsed by the NRC in Regulatory Guide 1.187, Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments, dated November 2000. The inspectors also consulted Part 9900 of the NRC Inspection Manual, 10 CFR Guidance for 10 CFR 50.59, Changes, Tests, and Experiments.
b. Findings
b.1 Failure to Translate Design Basis Into Procedures for Service Water Flow to the CC Heat Exchangers
Introduction:
The inspectors identified a Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having very low safety significance (Green) for the licensees failure to correctly translate the design basis into procedures. Specifically, the licensee failed to procedurally ensure that required essential service water (SX) flows to the CC heat exchangers, which were established in design basis calculations, would be available during a design basis LOCA event. During a LOCA, this could have resulted in unanticipated operator actions to restore flow, or in the worst case, exceeding temperature limits for the CC heat exchanger.
Description:
10 CFR 50.59 screening, 6E-04-0115, Revise Updated Final Safety Analysis Report (UFSAR) Table 9.2-11 (DRP 10-071), evaluated an activity in June 2004 which revised UFSAR Table 9.2-11 to reflect the increase of the lower limit for SX flow through the CC heat exchangers from 5000 gallons per minute (gpm) to 5400 gpm. The basis for the increase in flow was described in calculation BYR2000-014, Byron/Braidwood Uprate Project - Post LOCA Component Cooling Water System Temperature
Analysis.
This calculation evaluated the effects on the CC system temperature from a five percent power uprate for both units. Higher core power levels increased the CC heat loads for power operation, residual heat removal (RHR) cooldown, and post accident cooling. The calculation also determined the required SX flow (5400 gpm) at 100 °F to maintain the design heat load during post LOCA conditions.
Calculation BYR99-010, Documentation of the Basis of the Emergency Operating Procedure (EOP) Setpoints established the basis for the 6000 gpm minimum SX flow to the CC heat exchangers. This calculation stated, A 5400 gpm flow rate will remove the post LOCA heat load at power uprate conditions. At this minimum flow rate, the CC temperature will exceed the 120 °F limit briefly during the initial recirculation phase.
However, it will not exceed 130 °F, which is acceptable to the components being cooled.
The margin between the 6000 gpm setpoint and 5400 gpm limit provides 600 gpm margin for addressing instrument uncertainty for this function.
The inspectors discovered that SX flow to the CC heat exchangers could only be verified by local indication during daily operator rounds; however, the operator rounds did not reflect the new minimum SX flow of 5400 gpm. Instead, the operators were verifying a minimum flow of 5000 gpm. In addition, the operator rounds also did not take into account instrument uncertainty of the flow meter. Consequently, a reading of 5000 gpm would have been acceptable per the operator rounds; however, the actual flow could have been as much as 600 gpm less. The actual flow in this worst case would have been almost a 1000 gpm below the design basis minimum value for SX flow.
In the event of a LOCA, the licensees emergency operating procedure, 2BEP ES-1.3, Transfer to Cold Leg Recirculation, would be entered after the reactor water storage tank level reached the low-low level switchover point. This procedure provided the necessary instructions for transferring the emergency core cooling system and containment spray system to the recirculation mode. The procedure also established CC flow to the RHR heat exchangers; however, this flow was not verified until after operation in the recirculation mode had been established. Because the operator rounds did not verify that SX flow to the CC heat exchangers was greater than 6000 gpm, the potential existed for the flow to be approximately 1000 gpm less than the design basis minimum value required for the initial recirculation phase, as established in the stations calculations. Insufficient SX flow to the CC heat exchanger could have resulted in unanticipated operator actions to restore flow, or in the worst case, exceeding temperature limits for the CC heat exchanger for an extended period of time.
The licensee entered this issue into the stations corrective action program as Issue Report 463574. The licensees corrective action was to update the operator rounds to verify 6000 gpm on the SX flow to CC heat exchanger flow meter to ensure that the design basis minimum flow of 5400 gpm would be present.
Analysis:
The inspectors determined that the failure to correctly translate the revised minimum SX flow to CC heat exchangers into procedures was a performance deficiency warranting a significance evaluation. This finding was considered more than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports because it was associated with the Mitigating System Cornerstone attribute of Design Control and affected the cornerstone objective of ensuring the capability of systems needed to respond to initiating events to prevent undesirable consequences.
Specifically, the failure to have operator rounds verify the design basis minimum or to have the EOPs ensure the minimum flow prior to establishing initial cold leg recirculation in the event of a LOCA could potentially have allowed the SX flow to be less than the required value to maintain the design heat load during post LOCA conditions.
In accordance with IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations, Attachment 1, the inspectors performed an SDP Phase 1 screening. The finding screened as having very low significance (Green)using IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for the At-Power Situations, because the inspectors answered no to all five questions under the Mitigating Systems Cornerstone column of the Phase 1 worksheet. Even though the licensee did not control their bounding design basis service water flow procedurally, the flow to the CC heat exchangers has historically been well above the bounding design basis flow.
Enforcement:
10 CFR Part 50, Appendix B, Criterion III, Design Control, requires that measures be established to assure that the design basis is correctly translated into procedures. The stations calculations established the minimum SX flow to the CC heat exchangers at 5400 gpm. Contrary to 10 CFR Part 50, Appendix B, Criterion III, the licensee failed to procedurally ensure that required essential service water flows to the CC heat exchangers, which were established in design basis calculations, would be available during a design basis LOCA event. Since this finding is of very low safety significance and was entered into the licensees corrective action program (Issue Report 463574), it is considered an NCV consistent with Section VI.A.1 of the NRC Enforcement Policy. (NCV 05000454/2006006-01; 05000455/2006006-01 (DRS))
1R17 Permanent Plant Modifications
.1 Review of Permanent Plant Modifications
a. Inspection Scope
From March 6 through March 24, 2006, the inspectors reviewed six permanent plant modifications that had been installed in the plant during the last two years. The modifications were chosen based upon risk significance, safety significance, and complexity. As per inspection procedure 71111.17B, one modification was chosen that affected the barrier integrity cornerstone. The inspectors reviewed the modifications to verify that the completed design changes were in accordance with the specified design requirements and the licensing bases and to confirm that the changes did not adversely affect any systems' safety function. Design and post-modification testing aspects were verified to ensure the functionality of the modification, its associated system, and any support systems. The inspectors also verified that the modifications performed did not place the plant in an increased risk configuration.
The inspectors also used applicable industry standards to evaluate acceptability of the modifications. The list of modifications and other documents reviewed by the inspectors is included as an attachment to this report.
b. Findings
b.1 Reduction of Fire Suppression Capacity and Capability
Introduction:
The inspectors identified a Non-Cited Violation of 10 CFR 50.48(a)(1)having very low safety significance (Green) for the licensees failure to provide fire fighting systems of appropriate capacity and capability to minimize the adverse effects of fires on structures, systems, and components important to safety. Specifically, the licensee abandoned standpipes and manual hose stations located near safety-related equipment which reduced the fire suppression capacity and capability to protect such equipment. In addition, the site relied on a local fire department instead of the site fire brigade to extinguish a fire that could affect safety-related equipment.
Description:
Modification EC 351113, Abandonment of Fire Protection Ring Header, dated August 2005, abandoned in-place the fire protection water ring header in the River Screen House (RSH). The RSH contains safety-related SX Cooling Tower Makeup Pumps and their respective diesel fuel oil storage tanks. The reason for the modification was to prevent main control room alarms that were being caused by a leak in the fire protection header line. The modification disabled the alarms and installed positive isolation (blank plate) between the circulation water supply header and the fire protection ring header. The modification also resulted in the abandonment of standpipes and six manual hose stations in the RSH.
The inspectors questioned the adequacy of abandoning the manual hose stations since this modification adversely affected the fire suppression capacity and capability for equipment important to safety (SX Makeup Pumps). However, when evaluating the change, the licensee determined that the change did not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire. Furthermore, the evaluation stated that the hose stations were not the primary means of suppression protecting safety-related equipment (a localized carbon dioxide (CO2) system was provided),portable fire extinguishers were available, and the local fire department would have the primary responsibility for suppressing the fire using their own equipment.
Based upon the licensees Fire Protection Report (FPR), the inspectors determined that the licensees conclusion that the level of protection for the safety-related equipment in the area would not be diminished was not adequate. The stations FPR, Section 3.0, Guidelines of Branch Technical Position (BTP) CMEB 9.5-1 listed the stations fire protection program requirements. BTP CMEB 9.5-1, Section 3.1.c.(1) stated NRCs position on the fire suppression system design basis, Total reliance should not be placed on a single fire suppression system. Appropriate backup fire suppression capability should be provided. The stations response was, Comply. Backup fire suppression equipment is provided in the form of manual hose stations and portable fire extinguishers at or near where automatic fire suppression systems are installed as well as at other locations throughout the plant. BTP CMEB 9.5-1, Section 3.7.k stated NRCs position on guidelines for specific plant areas, specifically for safety-related pumps, Hose stations and portable extinguishers should be readily accessible [for pump houses and rooms housing redundant safety-related pump trains]. Again, the licensee reaffirmed their compliance with this portion of the BTP.
The inspectors were also concerned that the station depended on the local fire department instead of the sites fire brigade to extinguish a fire that could affect safety-related equipment. BTP CMEB 9.5-1, Section 3.3.b stated the NRCs position on fire brigades, A site fire brigade trained and equipped for fire fighting should be established to ensure adequate manual fire fighting capability for all areas of the plant containing structures, systems, or components important to safety. The licensees response was that the station complied with this section.
Based upon the licensees repsonse to the BTP, as contained in the Byron FPR, the inspectors determined that the hose stations and fire brigade were critical fire protection features to protect the safety related equipment in the RSH. Consequently, the inspectors concluded that relying on an offsite fire department in addition to abandoning the manual hose stations reduced the fire suppression capacity and capability for protecting the SX cooling tower makeup pumps and their diesels, therefore adversely impacting on equipment important to safety. The licensee entered this issue into the stations corrective action program as Issue Report 469894.
Analysis:
The inspectors determined that the failure to provide fire fighting systems of appropriate capacity and capability to minimize the adverse effects of fires on structures, systems, and components important to safety was a performance deficiency warranting a significance evaluation. This finding was considered more than minor in accordance with IMC 0612, Power Reactor Inspection Reports because it was associated with the Mitigating System cornerstone attribute of Protection Against External Factors, and affected the cornerstone objective of ensuring the availability of systems that respond to initiating events to prevent undesirable consequences. Abandoning the manual hose stations reduced the fire suppression capacity and capability for protecting the SX cooling tower makeup pumps and their diesels, therefore having an adverse impact on equipment important to safety. The manual hose stations would have been needed when the concentration of the localized CO2 suppression dissipated in the event of a fire.
The loss of the SX cooling tower makeup pumps could have resulted in the degradation of the safe shutdown function to provide essential mechanical support for hot standby.
In accordance with IMC 0609, Appendix A, the inspectors performed an SDP Phase 1 screening and determined that the finding degraded the Fire protection (FP) portion of the Mitigating Systems Cornerstone. Therefore, screening under IMC 0609. Appendix F, Fire Protection Significance Determination Process, was required. The finding was determined to affect the element of Fixed Fire Protection Systems in accordance with Table 1.1-1 in Appendix Because the finding involved the manual suppression system, no credit for suppression of the fire before it involved both trains of SX makeup was considered. Each diesel has a localized automatic carbon dioxide suppression system which would likely be effective in suppressing smaller fires. However, the carbon dioxide suppression was not explicitly credited in the significance determination. The SRA calculated a conditional core damage probability using the NRCs Standardized Plant Analysis Risk (SPAR) model for Byron assuming that the fire results in a transient and affects both trains of SX makeup. The risk model includes the deep well pumps which can perform the same function as the SX makeup pumps and are unaffected by the fire.
The calculated conditional core damage probability was 6.7E-5. As a result, the delta CDF for this finding was calculated to be approximately 7.5E-7. Because this estimate is below 1.0E-6, this finding was characterized as having very low safety significance (Green).
Enforcement:
10 CFR 50.48(a)(1) stated, Each operating nuclear power plant must have a fire protection plan that satisfies Criterion to minimize the adverse effects of fires on structures, systems, and components important to safety. Contrary to this requirement, the licensee failed to provide fire fighting systems of appropriate capacity and capability to minimize the adverse effects of fires on structures, systems, and components important to safety. Specifically, the licensee abandoned standpipes and manual hose stations located near safety-related equipment which reduced the fire suppression capacity and capability to protect such equipment. In addition, the site relied on a local fire department instead of the site fire brigade to extinguish a fire that could affect safety-related equipment. Since this finding is of very low safety significance and was entered into the licensees corrective action program, it is considered an NCV consistent with Section VI.A.1 of the NRC Enforcement Policy. (NCV 05000454/2006006-02; 05000455/2006006-02 (DRS))
OTHER ACTIVITIES (OA)
4OA2 Identification and Resolution of Problems
.1 Routine Review of Condition Reports
a. Inspection Scope
From March 6 through March 24, 2006, the inspectors Action Process documents that identified or were related to 10 CFR 50.59 evaluations and permanent plant modifications. The inspectors reviewed these documents to evaluate the effectiveness of corrective actions related to permanent plant modifications and evaluations for changes, tests, or experiments issues. In addition, corrective action documents written on issues identified during the inspection were reviewed to verify adequate problem identification and incorporation of the problems into the corrective action system. The specific corrective action documents that were sampled and reviewed by the team are listed in the attachment to this report.
b. Findings
No findings of significance were identified.
OTHER ACTIVITIES
4OA6 Meetings
.1 Exit Meeting
The inspectors presented the inspection results to Ms. M. Snow and others of the licensees staff, on March 24, 2006. Licensee personnel acknowledged the inspection results presented. Licensee personnel were asked to identify any documents, materials, or information provided during the inspection that were considered proprietary. No proprietary information was identified.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
E.Blondin, Mechanical/Structural Design Manager
- T. Fluck, Regulatory Assurance Specialist
- J. Langan, Regulatory Compliance
- V. Naschansky, Electrical/Iand C Design Manager
- W. Perchazzi, Engineering Response Manager
- R. Randels, Sr. Manager Design Engineering
Nuclear Regulatory Commission
- B. Bartlett, Senior Resident Inspector
- R. NG, Resident Inspector
- D. Hills, EB1 Branch Chief
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
None.
Opened and Closed
- 05000455/2006006-01 NCV Failure to Translate Design Basis Into Procedures for Service Water Flow to the CC Heat Exchangers
- 05000455/2006006-02 NCV Reduction of Fire Suppression Capacity and Capability
Discussed
None.