IR 05000315/2006009
| ML061920589 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 07/07/2006 |
| From: | Dave Hills NRC/RGN-III/DRS/EB1 |
| To: | Nazar M Indiana Michigan Power Co |
| References | |
| IR-06-009 | |
| Download: ML061920589 (24) | |
Text
uly 7, 2006
SUBJECT:
D.C. COOK NUCLEAR POWER PLANT, UNITS 1 AND 2, NRC EVALUATION OF CHANGES, TESTS, OR EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000315/2006009(DRS)
316/2006009(DRS)
Dear Mr. Nazar:
On May 26, 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 D. C. Cook Nuclear Power Plant. Inspectors conducted subsequent in-office review prior to report issuance to resolve several issues identified during the onsite inspection. The enclosed report documents the results of the inspection, which were discussed and others of your staff at the completion of the inspection on May 26, 2006 and on July 6, 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, 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 D. C. Cook Nuclear Power Plant.
In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letter, its enclosure, and your response (if any), 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), 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 05000315/2006009(DRS); 05000316/2006009(DRS)
M. Peifer, Site Vice President L. Weber, Plant Manager G. White, Michigan Public Service Commission L. Brandon, Michigan Department of Environmental Quality -
Waste and Hazardous Materials Division Emergency Management Division MI Department of State Police State Liaison Officer, State of Michigan D. Lochbaum, Union of Concerned Scientists
In a
REGION III==
Docket No: 50-315, 50-316 DRP-58; DRP-74 Report No: 05000315/2006009(DRS); 05000316/2006009(DRS)
Licensee: American Electric Power Company Facility: D.C. Cook Nuclear Power Plant, Units 1 and 2 Location: 7700 Red Arrow Hwy Stevensville, MI 49127-9726 Dates: May 8, 2006 through May 26, 2006 Inspectors: R. Daley, Senior Reactor Inspector C. Acosta, Reactor Inspector Approved by: D. Hills, Chief Engineering Branch 1 Division of Reactor Safety (DRS)
SUMMARY OF FINDINGS
IR 05000315/2006009(DRS); 05000316/2006009(DRS); 05/08/2006 through 05/26/2006; D.C.
Cook Nuclear Power Plant, Units 1 and 2; 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. 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 Severity Level IV Non-Cited Violation of 10 CFR 50.59(d)(1) for the licensees failure to perform a safety evaluation for the deletion of four sections of the Technical Requirements Manual (TRM). Specifically, the licensee deleted Sections 8.4.7, Tavg Lower Limit, 8.6.1, Ice Bed Temperature Monitoring System, and 8.6.2, Inlet Door Position Monitoring System, and 8.3.7,
Post Accident Monitoring (PAM) Instrumentation, Table 8.3.7-1 without evaluating these changes per the requirements of 10 CFR 50.59.
Because the issue potentially impacted the NRCs ability to perform its regulatory function, this finding was evaluated using the traditional enforcement process. The finding was determined to be more than minor because the inspectors, at the time of the inspection, could not reasonably determine that the Updated Final Safety Analysis Report (UFSAR) change, which adversely affected equipment important to safety, would not have ultimately required NRC approval. The inspectors completed a significance determination of the underlying technical issue using NRCs inspection manual chapter (IMC) 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, and answered no to the Mitigating Systems screening questions in the Phase 1 Screening Worksheet. Specifically, even though these TRM sections along with their associated surveillance requirements were deleted, the licensee was able to show that all deleted surveillance requirements had been performed satisfactorily and within their prescribed frequency in spite of the deletion. This issue was entered into the licensees corrective action program. (Section 1R02.1.b.1)
- Green.
The inspectors identified a Severity Level IV Non-Cited Violation of 10 CFR 50.59(d)(1) for the licensees failure to perform a safety evaluation for the modification of the 2-East Centrifugal Charging Pump (CCP). Specifically, the licensee performed modifications to the 2-East Centrifugal Charging Pump that required more restrictive frequency requirements to be established than were already in the Technical Specifications. Had a 10 CFR 50.59 evaluation been performed, as required, the evaluation should have shown that a change to the Technical Specifications (TS) was required so that the new required frequency value could be incorporated into the applicable TS Surveillance Requirements. This issue was entered into the licensees corrective action program.
Because the issue potentially impacted the NRCs ability to perform its regulatory function, this finding was evaluated using the traditional enforcement process. The finding was determined to be more than minor because the inspectors could not reasonably determine that the modification of the 2-East Centrifugal Charging Pump would not have ultimately required NRC approval. The inspectors evaluated the finding using IMC 0609, Appendix A, Phase 1 screening for the mitigating systems cornerstone and determined that the finding was of very low safety significance because they were able to answer no to the Mitigating Systems screening questions in the Phase 1 Screening Worksheet. Specifically, while the 10 CFR 50.59 evaluation, and ultimately the required license amendment, were not performed as required, administrative controls were put into place after the modification was performed such that the CCP would always be able to perform its function. (Section 1R17.1.b.1)
- Green.
The inspections identified a Non-Cited Violation of 10 CFR Part 50, Appendix B,
Criterion III, Design Control, that was of very low safety significance. Specifically, verification of containment lower compartment average temperature per Surveillance Requirement (SR) 3.6.5.2 was being performed using temperature readings that were not representative (and non-conservative) of the true average temperature in the lower containment. The issue was entered into the licensees corrective action program.
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 methodology for determining lower containment average temperature was non-conservative and did not account for the heightened temperatures that were experienced in the Steam Generator (SG)
Rooms. Had average temperature been above the TS limits, temperatures during a Design Basis Accident could have exceeded the ratings of safety related mitigating equipment thereby challenging the functionality of the equipment. This finding was of very low safety significance, because the inspectors answered no to all five questions under the Mitigating Systems Cornerstone column of the Phase 1 worksheet. Specifically, after performing a calculation that included the SG Rooms, the licensee determined that under worst case historical conditions, average air temperature was 119.5 degrees which was still less than the TS requirement of 120 degrees F. (Section 1R17.1.b.2)
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 May 8 through May 26, 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 14 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 Inappropriate Deletion of Technical Requirements Manual Sections
Introduction:
The inspectors identified a Non-Cited Violation of 10 CFR 50.59, Changes, Tests, and Experiments, having very low safety significance (Green) for the deletion of portions of the TRM without performing an evaluation in accordance with 10 CFR 50.59.
Specifically, the licensee deleted four TRM sections. While portions of these sections may have been able to be removed under the provisions of 10 CFR 50.59, certain portions, particularly the surveillance requirements, would have needed NRC approval prior to being removed.
Description:
During implementation of the Improved Technical Specifications (ITS) at D.C. Cook Nuclear Power Plant, certain requirements that were originally a part of the Current Technical Specifications (CTS) were removed from the Technical Specifications and relocated to the Technical Requirements Manual. After relocation, these requirements were subject to the controls of 10 CFR 50.59. For purposes of future changes to the TRM, these requirements were considered to be incorporated by reference into the UFSAR.
During this process in June 2005, 10 CFR 50.59 Screens 2005-0139-00, 2005-0334-00, and 2005-0421-00 were performed to delete certain of these requirements from the TRM prior to implementation of TRM Revision 0. These Screens deleted TRM 8.3.7, Post Accident Monitoring (PAM) Instrumentation, 8.4.7, Tavg Lower Limit, 8.6.1, Ice Bed Temperature Monitoring System, and 8.6.2, Inlet Door Position Monitoring System. The licensee treated the changes as simple procedure changes rather than an actual deletion of UFSAR requirements. Consequently, the licensee incorrectly determined that the changes did not adversely affect the UFSAR. As a result, the required 10 CFR 50.59 evaluation was not performed for these changes. Additionally, since the Screens also deleted the surveillance requirements in those TRM sections, portions of the deletions would not have successfully passed a 10 CFR 50.59 evaluation.
Deletion of portions of these testing requirements for equipment important to safety would result in more than a minimal increase in the likelihood of a malfunction of a structure, system, or component important to safety since surveillance testing minimizes the probability of failure of equipment.
Based upon the inspectors findings, the licensee issued Condition Report (CR) 06129041. As a result of the CR, the licensee verified that all deleted surveillance requirements were performed within their previous required frequencies. Additionally, the licensee planned to re-evaluate the TRM sections that were deleted and put the portions that could not be justified for removal under 10 CFR 50.59 back into the TRM.
Analysis:
The inspectors determined that this issue was a performance deficiency since, in June 2005, the licensee failed to perform a required safety evaluation in accordance with 10 CFR 50.59. Specifically, the licensee deleted TRM Sections 8.4.7, Tavg Lower Limit, 8.6.1, Ice Bed Temperature Monitoring System, and 8.6.2, Inlet Door Position Monitoring System, and 8.3.7, Post Accident Monitoring (PAM) Instrumentation without evaluating this change per the requirements of 10 CFR 50.59. The finding was determined to be more than minor because the inspectors could not reasonably determine that these TRM deletions would not have ultimately required NRC prior approval.
Because violations of 10 CFR 50.59 are considered to be violations that potentially impede or impact the regulatory process, they are dispositioned using the traditional enforcement process instead of the significance determination process (SDP). However, if possible, the underlying technical issue is evaluated under the SDP to determine the severity of the violation. In this case, the inspectors completed a significance determination of the underlying technical issue using NRCs inspection manual chapter (IMC) 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, and answered no to the Mitigating Systems screening questions in the Phase 1 Screening Worksheet. Specifically, while the TRM deletions may not have been able to be done without a license amendment, the licensee was able to show that all deleted surveillance requirements had been performed satisfactorily and within their prescribed frequency in spite of the deletion. Based upon this Phase 1 screening, the inspectors concluded that the issue was of very low safety significance (Green). In accordance with the Enforcement Policy, the violation was therefore classified as a Severity Level IV violation.
Enforcement:
Title 10 CFR 50.59(d)(1) states, in part, that the licensee shall maintain records of changes in the facility, of changes in procedures, and of tests and experiments as described in the UFSAR. These records must include a written evaluation which provides a basis for the determination that the change, test, or experiment does not require a license amendment.
Contrary to the above, the licensee, in June 2005, deleted portions (Sections 8.3.7, 8.4.7, 8.6.2, and 8.6.1) of the TRM, constituting a change in procedures as described in the UFSAR, without providing the required written evaluation. In addition, the inspection team could not reasonably determine that these changes would not have required a license amendment, because the deletion of the associated surveillance requirements could have resulted in more than a minimal increase in the likelihood of a malfunction of a structure, system, or component important to safety. In accordance with the Enforcement Policy, this violation of the requirements of 10 CFR 50.59 was classified as a Severity Level IV Violation because the underlying technical issue was of very low safety significance. Because this non-willful violation was non-repetitive, and was captured in the licensees corrective action program (CR 06129041), it is considered a Non-Cited Violation consistent with VI.A.1 of the NRC Enforcement Policy (NCV).
(NCV 05000315/2006009-01; 05000316/2006009-01 (DRS))
1R17 Permanent Plant Modifications
.1 Review of Permanent Plant Modifications
a. Inspection Scope
From May 8 through May 26, 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 Failure to Perform 10 CFR 50.59 Evaluation for Modification to the 2-East Centrifugal Charging Pump
Introduction:
The inspectors identified a Non-Cited Violation of 10 CFR 50.59, Changes, Tests, and Experiments, having very low safety significance (Green) for the modification of the 2-East Centrifugal Charging Pump. Specifically, the licensee implemented this modification with a 10 CFR 50.59 Screen even though it required a Technical Specification change to implement.
Description:
In the first quarter of 2006, D.C. Cook Nuclear Plant staff initiated and approved EE-2005-0436 allowing for the replacement of the CCP impeller and shaft.
Because of the different operating characteristics of the new impeller, the modified pump would draw a higher Brake Horsepower (BHP) than the original pump. While there was a recognition that this higher BHP would exist, the licensee never addressed the effects of this higher horsepower on the pump motor.
Prior to implementation of the modification, the licensee discovered that with the higher BHP, under the worst case high frequency of 61.2 Hz, the CCP motor would exceed the motor nameplate rating of 690 horsepower (HP) (600 HP with a 15 percent service factor). Because of this discovery, the licensee issued CR 06089052 on March 30, 2006.
At this time, the design modification had still not been field implemented.
Because of the loading problems with the motor, the licensee performed an evaluation to show that the motor would still be able to operate within its design rating as long as the frequency was held below 60.5 Hz. As stated in the CR, On this bases, several Technical Specification Surveillance Requirements, which have an acceptance criteria of 61.2 Hz on the high end, are non-conservative. At this point, instead of delaying the modification, the licensee decided to perform the change; however, more stringent administrative controls were to be established to address the non-conservative TS issues. The licensee believed that this approach was allowable because of the guidance provided in NRC Administrative Letter 98-10, Dispositioning of Technical Specifications That Are Insufficient to Assure Plant Safety. Based upon this decision, the licensee implemented changes to the plant TRM to ensure that Emergency Diesel Generators (EDG) frequency would always be below 60.5 Hz. Additionally, the licensee ensured that all TS surveillances performed that required an upper band of 61.2 Hz met the more stringent criteria of 60.5 Hz. This showed that the EDGs would always start and load at a frequency lower than 60.5 Hz.
While the licensee had adequately addressed any operability concerns by imposing these more restrictive administrative controls, the licensees use of Administrative Letter 98-10 as a tool to knowingly enter a condition (by performing the modification)involving non-conservative Technical Specification Surveillance Requirements was not appropriate. Specifically, Administrative Letter 98-10 was to be used if plant staff found non-conservative Technical Specification values. It was not a justification for creating a non-conservative Technical Specification condition. Additionally, the team further noted that when it was discovered that the CCP motor would exceed the motor nameplate rating, the licensee should have realized that the scope and the effects of the modification had changed. At this point, had the licensee re-evaluated the modification and performed a 10 CFR 50.59 review and evaluation, the results should have shown that a change to the Technical Specifications was required so that the new required frequency value of 60.5 Hz could be incorporated into the applicable TS Surveillance Requirements. The inspectors determined that this change would have required a license amendment prior to being performed.
Based upon the inspectors concerns, the licensee issued CR 06143098. Because, the licensee had already adequately addressed administrative controls for this condition, there were no immediate operability concerns.
Analysis:
The inspectors determined that this failure to perform a safety evaluation in accordance with 10 CFR 50.59 for changes made to the UFSAR was a performance deficiency warranting a significance determination. Specifically, the licensee performed modifications to the 2-East Centrifugal Charging Pump that required more restrictive frequency requirements to be established than were already in the Technical Specifications. Had a 10 CFR 50.59 evaluation been performed, as required, the evaluation should have shown that a change to the Technical Specifications was required so that the new required frequency value could be incorporated into the applicable TS Surveillance Requirements. The finding was determined to be more than minor because the required technical specification change for the CCP modification would have required NRC prior approval.
Because violations of 10 CFR 50.59 are considered to be violations that potentially impede or impact the regulatory process, they are dispositioned using the traditional enforcement process instead of the significance determination process (SDP).
However, if possible, the underlying technical issue is evaluated under the SDP to determine the severity of the violation. In this case, the inspectors completed a significance determination of the underlying technical issue using NRCs inspection manual chapter (IMC) 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, and answered no to the Mitigating Systems screening questions in the Phase 1 Screening Worksheet. Specifically, while the 10 CFR 50.59 evaluation, and ultimately the required license amendment, were not performed as required, administrative controls were put into place prior to plant operations such that the CCP would always be able to perform its function. Based upon this Phase 1 screening, the inspectors concluded that the issue was of very low safety significance (Green). In accordance with the Enforcement Policy, the violation was therefore classified as a Severity Level IV Violation.
Enforcement:
Title 10 CFR 50.59(d)(1) states, in part, that the licensee shall maintain records of changes in the facility, of changes in procedures, and of tests and experiments. These records must include a written evaluation which provides a basis for the determination that the change, test, or experiment does not require a license amendment. Additionally, 10 CFR 50.59(c)(I) states, in part, that the licensee may make changes to the facility provided the change does not require a change to the technical specifications incorporated in the license.
Contrary to the above, the licensee, in 2006, performed a modification to the 2-East CCP without performing the required safety evaluation. Additionally, because the change caused the high frequency band of 61.2 Hz to no longer be adhered to, the change should have involved a TS amendment to change the frequency bands for the applicable TS surveillance requirements prior to implementation of the modification. In accordance with the Enforcement Policy, this violation of the requirements of 10 CFR 50.59 was classified as a Severity Level IV Violation because the underlying technical issue was of very low safety significance. Because this non-willful violation was non-repetitive, and was captured in the licensees corrective action program (CR 06143098), it is considered a Non-Cited Violation consistent with VI.A.1 of the NRC Enforcement Policy (NCV). (NCV 05000315/2006009-02; 05000316/2006009-02 (DRS))
b.2 Non-Conservative Verification of Containment Average Air Temperature
Introduction:
The inspectors identified a Non-Cited Violation having a very low safety significance (Green) of 10 CFR Part 50, Appendix B, Criterion III, Design Control.
Specifically, the licensee was verifying containment lower compartment average temperature per Surveillance Requirement (SR) 3.6.5.2 by using temperature readings that were not representative of the true average temperature in the lower containment.
Description:
In ICP-01045, Revision 00, Unit 1 and Unit 2 Steam Generator Compartment High Temperature Alarm Change, the licensee documented a setpoint change to the temperature alarm from 140 degrees to 180 degrees Fahrenheit (F). This change was necessary since the Steam Generator Enclosure areas, which were located in the containment lower compartment, were operating at temperatures in excess of 140 degrees Fahrenheit.
During the inspection, the team questioned how TS 3.6.5, Containment Air Temperature, was addressed considering these higher temperatures. Specifically, Limiting Condition for Operability (LCO) 3.6.5 requires that average lower containment air temperature be less than or equal to 120 degrees F. Additionally, Surveillance Requirement 3.6.5.2 requires that the containment lower compartment air temperature be verified within limits (120 degrees F). The Bases for SR 3.6.5.2 states, Verifying that containment average air temperature is within the LCO limits ensures that containment operation remains within the limits assumed for the containment analyses. In order to determine the containment average air temperature, an arithmetic average is calculated using measurement taken at locations within the containment selected to provide a representative sample of the overall containment atmosphere . . . In the lower compartment, the locations at nominal elevations 626 ft 6 inches, 624 ft 10 1/2 inches, and 624 ft 0 inches are used. While the location of these temperature instruments were representative of the majority of the lower containment, they were not representative of the SG Enclosure Rooms which were essentially isolated from the other areas in the lower containment. Because the SG Enclosure Rooms contained a significant amount of the air volume for the lower containment (Preliminarily calculated at approximately 12 percent of total volume), the inspectors were concerned that SR 3.6.5.2 was no longer conservative. The SG Enclosure Rooms were historically operating at temperatures in excess of 140 degrees F; however, the temperatures used in determining acceptability of SR 3.6.5.2 did not account for these higher temperatures at all.
Based upon the inspectors concerns, the licensee initiated CR 06145114. Because of the present outside ambient and lake temperatures, the licensee was able to determine that the lower containment was operable. Additionally, the licensee performed a past operability evaluation. In that evaluation, the licensee determined that temperatures in the SG Enclosure Rooms had historically reached as high as 164.3 degrees F.
However, after performing a lower containment temperature calculation for the worst case days in the recent past, the licensee determined that the highest average air temperature was still within TS limits. After including the SG Enclosure Rooms, the licensee determined a worst case average air temperature of 119.5 degrees. While this temperature was within the TS limit of 120 degrees F, it was significantly closer to the TS limit than the original method had determined.
Analysis:
The inspectors determined that this failure to provide conservative acceptance criteria for SR 3.6.5.2 was a performance deficiency warranting a significance determination. 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 methodology for determining lower containment average temperature was non-conservative and did not account for the heightened temperatures that were experienced in the SG Enclosure Rooms. Had average temperature been above the TS limits, temperatures during a Design Basis Accident could have exceeded the ratings of safety related mitigating equipment thereby challenging the functionality of the equipment.
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. Specifically, after performing a calculation that included the SG Enclosure Rooms, the licensee determined that under worst case historical conditions, average air temperature was 119.5 degrees which was still less than the TS requirement of 120 degrees F.
Enforcement:
10 CFR Part 50, Appendix B, Criterion III, Design Control states, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Contrary to the above, no established design basis existed to support the use of the temperature measurements that were being used as a representative sample for the average temperature for the lower containment that was delineated in SR 3.6.5.2.
Because this failure to provide conservative temperature measurements for the lower containment average air temperature was determined to be of very low safety significance and because it was entered in the licensees corrective action program as CR 06145114, this violation is being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy. (NCV 05000315/2006009-03; 05000316/2006009-03 (DRS))
OTHER ACTIVITIES (OA)
4OA2 Identification and Resolution of Problems
.1 Routine Review of Condition Reports
a. Inspection Scope
From May 8 through May 26, 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 Mr. J. Jensen and others of the licensees staff, on May 26, 2006 and on . 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
- D. Baker, Configuration Programs Supervisor
- D. Fadel, Vice President Engineering
- R. Kohrt, Safety Analysis Supervisor
- R. Meister, NRA Senior Specialist
- R. Neuendorf, Configurations Programs
- M. Scarpello, NRA Supervisor
Nuclear Regulatory Commission
- B. Kemker, Senior Resident Inspector
- J. Lennartz, Resident Inspector
- D. Hills, EB1 Branch Chief
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
None.
Opened and Closed
- 05000315/2006006-01; NCV Inappropriate Deletion of Technical Requirements
- 05000316/2006006-01 Manual Sections
- 05000315/2006009-02; NCV Failure to Perform 10 CFR 50.59 Evaluation for
- 05000316/2006009-02 Modification to the 2-East Centrifugal Charging Pump
- 05000315/2006009-03; NCV Non-Conservative Verification of Containment Average
- 05000316/2006009-03 Air Temperature
Discussed
None.
Attachment