IR 05000254/2005007
| ML053110524 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 11/07/2005 |
| From: | Dave Hills NRC/RGN-III/DRS/EB1 |
| To: | Crane C Exelon Generation Co, Exelon Nuclear |
| References | |
| IR-05-007 | |
| Download: ML053110524 (19) | |
Text
ber 7, 2005
SUBJECT:
QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 NRC EVALUATION OF CHANGES, TESTS, OR EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000254/2005007; 05000265/2005007(DRS)
Dear Mr. Crane:
On October 7, 2005, 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 Quad Cities Nuclear Power Station, Units 1 and 2. The enclosed report documents the results of the inspection, which were discussed with Mr. M. Perito and others of your staff at the completion of the inspection on October 7, 2005.
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 no findings were identified.
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 Docket Nos. 50-254; 50-265 License Nos. DPR-29; DPR-30 Enclosure: Inspection Report 05000263/2005011(DRS)
See Attached Distribution
SUMMARY OF FINDINGS
IR 05000254/2005007; 05000265/2005007(DRS); 10/03/2005 - 10/07/2005; Quad Cities
Nuclear Power Station, Units 1 and 2; Evaluation of Changes, Tests, or Experiments (10 CFR 50.59) and Permanent Plant Modifications.
The inspection covered a one-week announced baseline inspection on evaluations of changes, tests or experiments and permanent plant modifications. The inspection was conducted by three regional based engineering inspectors. Two Unresolved Items 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 No findings of significance were identified.
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 October 3 through 7, 2005, 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 17 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 Downgrade of Relief Valves from Category I Environmental Qualification (EQ) to Category II EQ Components During the inspection, the team identified that in 2004 the licensee had replaced the Unit 2 reactor pressure vessel relief valves, Target Rock Power Operated Relief Valves (PORVs) qualified Category I in accordance with the EQ requirements in 10 CFR 50.49, with Dresser Electromatic Relief Valves (ERVs) qualified as Category II EQ Components. The inspectors determined that this EQ downgrading of the reactor pressure vessel relief valves appeared to be in violation of the requirements in 10 CFR 50.49, Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants.
The licensee originally replaced the Unit 2 reactor pressure vessel relief valves in 1995.
The old Dresser ERVs - the original design for the reactor pressure vessel relief valves -
were replaced by Target Rock PORVs. Consistent with the provisions of 10 CFR 50.49, the replacement valves were EQ upgraded in accordance with the Category I requirements specified in NUREG-0588, Interim Staff Position on Environmental Qualification of Safety-Related Electric Equipment.
After several years of operation, the licensee decided to replace the PORVs because of operational performance issues with these replacement valves. In 2004, the licensee replaced the PORVs in Unit 2 with the ERVs that were in the plants initial design. The licensee justified the change because the ERVs were within the plants original license and design basis. However, even though the licensee purchased new ERVs to replace the PORVs, these valves still were not EQ qualified to Category 1 unlike the valves that they replaced.
The licensee justified replacing the PORVs with the non-Category 1 ERVs by performing a Sound Reasons to the Contrary evaluation. 10 CFR 50.49(l) requires that replacement equipment be qualified in accordance with the provisions of 10 CFR 50.49 unless there are sound reasons to the contrary. Regulatory Guide 1.89, Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants, contains guidance on the justifications that would qualify as sound reasons to the contrary. Based upon this guidance, the licensee, in EC 345004, Revision 0, reasoned that a suitable replacement design, qualified in accordance with 10 CFR 50.49, would require significant plant modifications to accommodate its use. This reasoning was primarily based upon the cost of the next cheapest alternative (a Crosby direct acting valve). The cost of these replacement valves would be greater than the cost of reverting to the original ERV design.
The inspectors questioned whether this reversion to the original design was allowable under 10 CFR 50.49. Specifically, since the PORVs were already established as part of the design and licensing basis of the plant, the inspection team questioned if it was allowable under 10 CFR 50.49 and whether the licensees Sound Reasons to the Contrary were adequate. This issue is being treated as an Unresolved Item pending further evaluation of the requirements for qualification of replacement equipment in 10 CFR 50.49. (URI 05000254/2005007-01; URI 05000265/2005007-01)
1R17 Permanent Plant Modifications
a. Inspection Scope
From October 3 through 7, 2005, 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 Lack of a Design Analysis Evaluating Secondary Fire Effects of Non-Fused 120 VAC Control Circuitry on the Plants Fire Protection Safe Shutdown Analysis The inspection team identified that 120 VAC circuitry did not have fusing to isolate potential faults in the circuitry, nor did the licensee have a design analysis to support this configuration during fire scenarios. The inspectors were concerned that these unfused circuits could cause secondary fires at the station should faults occur in the control cabling.
During inspection related to modification EC 336482, Reconfigure the Motor Control Circuit for the 2B SBLC Pump, the team discovered that in certain configurations of 120 VAC circuitry, no fuse isolation existed. The primary configuration of concern was 120 VAC ungrounded circuits powered from Control Power Transformers (CPTs)(480 - 120 VAC). Based upon this information, the inspectors asked if the licensees engineering staff had considered the potential adverse effects on the plants fire protection safe shutdown analysis. Specifically, the inspectors were concerned that if a fire were to occur in a fire area and cause a fault on one of these unfused 120 VAC circuits, a concurrent fire could occur somewhere else in the circuitry due to the high amperage caused by the faulted condition. This could adversely affect the safe shutdown functions during a fire, because it is an implicit assumption that only one fire, in one fire area, can occur at one time. This type of condition invalidates this assumption.
While the licensee was able to produce an analysis that addressed possible grounding and/or shorting out of the CPTs, the licensee was not able to produce any analysis which addressed the possible faulting of the control cabling in other areas. The engineering staff was able to produce an interoffice memorandum between their Architect Engineering firm, Sargent & Lundy, and the licensee dated February 11, 1985 that pertained to this issue; however, this document appeared to only address potential fires from the CPTs.
While it is expected that in most circuits, the licensee will be able to show that the CPTs would fail before the cabling, it is uncertain if, and how many, cases exist where the cabling has the potential to fault in other areas prior to the CPT failing. At the time of the inspection, the licensee could not provide such an evaluation. To address this issue, the licensee initiated corrective action document AR 00382847. Since the licensee had still not been able to conclusively determine whether this issue adversely affected the licensees fire protection safe shutdown analysis, this issue was considered an Unresolved Item pending a full evaluation by the licensee of this circuitry configuration.
(URI 05000254/2005007-02; URI 05000265/2005007-02)
OTHER ACTIVITIES (OA)
4OA2 Identification and Resolution of Problems
.1 Routine Review of Condition Reports
a. Inspection Scope
From October 3 through 7, 2005, the inspectors reviewed nine Corrective Action Process documents (CAPs) that identified or were related to 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 Mr. M. Perito and others of the licensees staff, on October 7, 2005. 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
- W. Beck, Regulatory Assurance Manager
- B. Edmark, Design Engineer
- R. Gideon, Plant Manager
- K. Moser, Site Engineering Director
- M. Perito, Operations Director
- W. Porter, Design Engineering Manager
- J. Taft, Electrical Engineering Supervisor
- M. Wagner, Regulatory Assurance Engineer
Nuclear Regulatory Commission
- D. Hills, Chief, Engineering Branch 1
- M. Kurth, Resident Inspector
- M. Ring, Chief, Reactor Projects Branch 1
Attachment
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
- 05000254/2005007-01; URI Downgrade of Relief Valves from Category I
- 05000265/2005007-01 Environmental Qualification (EQ) to Category II EQ Components
- 05000254/2005007-02; URI Lack of a Design Analysis Evaluating Secondary Fire
- 05000265/2005007-02 Effects of Non-Fused 120 VAC Control Circuitry on the Plants Fire Protection Safe Shutdown Analysis
Opened and Closed
None.
Discussed
None.
Attachment