IR 05000237/2015007
| ML15183A063 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 07/01/2015 |
| From: | Dariusz Szwarc Engineering Branch 3 |
| To: | Bryan Hanson Exelon Generation Co, Exelon Nuclear |
| References | |
| IR 2015007 | |
| Download: ML15183A063 (20) | |
Text
UNITED STATES uly 1, 2015
SUBJECT:
DRESDEN NUCLEAR POWER STATION - EVALUATIONS OF CHANGES, TESTS, AND EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000237/2015007; 05000249/2015007
Dear Mr. Hanson:
On May 29, 2015, the U.S. Nuclear Regulatory Commission (NRC) completed an Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications inspection at your Dresden Nuclear Power Station. The enclosed inspection report documents the inspection results, which were discussed on May 29, 2015, with Mr. Shane Marik, and other members of your staff.
The inspection 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.
NRC inspectors documented two NRC-identified findings of very-low safety significance (Green)
in this report. These findings were determined to involve violations of NRC requirements.
However, because of their very-low safety significance, and because these issues were entered into your Corrective Action Program, the NRC is treating these issues as Non-Cited Violations (NCVs) in accordance with Section 2.3.2, of the NRC Enforcement Policy.
If you contest the subject or severity of the NCVs, 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 copies to the Regional Administrator, Region III; the Director, Office of Enforcement, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001; and the NRC Resident Inspector at Dresden Nuclear Power Station. In addition, if you disagree with the cross-cutting aspect assigned to any finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region III, and the NRC Resident Inspector at Dresden Nuclear Power Station. In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public Inspections, Exemptions, Requests for Withholding, 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 NRCs Public Document Room or from the Publicly Available Records (PARS)
component of the NRC's Agencywide Documents Access and Management 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/
Dariusz Szwarc, Acting Chief Engineering Branch 3 Division of Reactor Safety Docket Nos. 50-237, 50-249;72-037 License Nos. DPR-19; DPR-25
Enclosure:
Inspection Report 05000237/2015007; 05000249/2015007
REGION III==
Docket Nos: 50-237; 50-249 License Nos: DPR-19; DPR-25 Report No: 05000237/2015007; 05000249/2015007 Licensee: Exelon Generation Company, LLC Facility: Dresden Nuclear Power Station Location: Morris, IL Dates: May 11 - 29, 2015 Inspectors: George M. Hausman, Senior Engineering Inspector (Lead)
Jorge J. Corujo-Sandin, Engineering Inspector Mark T. Jeffers, Engineering Inspector Observer: Christopher A. Hunt, Reactor Engineer Approved by: Dariusz Szwarc, Acting Chief Engineering Branch 3 Division of Reactor Safety Enclosure
SUMMARY
Inspection Report 05000237/2015007; 05000249/2015007; 05/11/2015 - 05/29/2015; Dresden
Nuclear Power Station; Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications.
This report covers a 2-week announced baseline inspection on evaluations of changes, tests, and experiments, and permanent plant modifications. The inspection was conducted by Region III based engineering inspectors. Two findings of very-low safety significance were identified by the inspectors. The findings were considered Non-Cited Violations of U.S. Nuclear Regulatory Commission (NRC) regulations. The significance of most findings is indicated by their color (i.e., greater than Green, or Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). Cross-cutting aspects were determined using IMC 0310, Aspects within the Cross-Cutting Areas. Findings for which the SDP does not apply may be Green, or be assigned a severity level after NRC management review. All violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy dated July 9, 2013. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 5, dated February 2014.
Cornerstone: Mitigating Systems
- Green.
The inspectors identified a finding of very-low safety significance, and an associated Non-Cited Violation (NCV) of Title 10, Code of Federal Regulations (CFR),
Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to ensure that applicable regulatory requirements and the isolation condensers (ICs) design bases were correctly translated into procedures. Specifically, the licensee added steps to the IC control procedures which directed operators to secure the IC in order to prevent the water level in the shell from going below 3.5 feet. The added steps would result in the IC being shutdown when required to operate per the ICs design bases. The licensee entered the issue into their Corrective Action Program (CAP) as Action Request 02506445, NRC MOD/5059 Inspection: ISCO [Isolation Condenser] Operating Procedures, dated May 28, 2015.
The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of Procedure Quality, and affected the cornerstones objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage).
Specifically, the inadequate procedures would drive the operators to stop the IC during a design bases event and prevent the IC from performing its design function of removing decay heat from the reactor. The finding has a cross-cutting aspect in the area of Human Performance; Teamwork, because the licensee did not communicate and coordinate activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, the Operations Department failed to communicate and coordinate with the Engineering Department when developing the procedural changes.
[H.4] (Section 1R17.1b)
- Green.
The inspectors identified a finding of very-low safety significance, and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to account for increased fuel oil consumption during the development of the Emergency Diesel Generator (EDG) Calculation 10553-CALC-07, Dresden Station
Emergency Diesel Generators Endurance Calculations, Revision 2, which resulted in non-conservative Technical Specifications (TS). Specifically, the licensee failed to account for the increased fuel oil consumption at an EDG frequency of 61.2 Hertz (Hz),
and ensure that the minimum fuel oil level in the EDG day tanks, as required per TS 3.8.1.4, was adequate to support the EDGs mission time at 110 percent for one hour. The licensee entered the issue into their CAP as Action Request 02506869,
NRC MOD/5059 Inspection: Emergency Diesel Generator Fuel Consumption, dated May 28, 2015.
The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstones objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the licensee failed to account for the increased fuel oil consumption resulting from operation at a higher EDG frequency. Therefore, the licensee did not ensure that the minimum fuel oil level in the day tanks, as required per TS 3.8.1.4, was adequate to support the EDGs mission time at 110 percent for one hour. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution; Identification, because the licensee did not did not thoroughly evaluate the EDG fuel oil consumption when considering EDG frequency variation. Specifically, the licensee failed to translate applicable design bases into specifications which resulted in non-conservative TS.
[P.1] (Section 1R17.2b)
REPORT DETAILS
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R17 Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications
.1 Evaluation of Changes, Tests, and Experiments
a. Inspection Scope
The inspectors reviewed 7 evaluations performed pursuant to Title 10, Code of Federal Regulations (CFR), Part 50, Section 59, to determine if the evaluations were adequate, and that prior U.S. Nuclear Regulatory Commission (NRC) approval was obtained as appropriate. The inspectors also reviewed 15 screenings, where licensee personnel had determined that a 10 CFR 50.59 evaluation was not necessary. The inspectors reviewed these documents to determine if:
- the changes, tests, and experiments performed were evaluated in accordance with 10 CFR 50.59, and that sufficient documentation existed to confirm that a license amendment was not required;
- the safety issue requiring the change, tests or experiment was resolved;
- the licensee conclusions for evaluations of changes, tests, and experiments were correct and consistent with 10 CFR 50.59; and
- the design and licensing basis documentation was updated to reflect the change.
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 (RG) 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.
This inspection constituted 7 samples of evaluations, and 15 samples of screenings and/or applicability determinations as defined in Inspection Procedure (IP) 71111.17-04.
b. Findings
Procedure Revisions Resulted in Isolation Condenser Unable to Meet Design Basis
Introduction:
The inspectors identified a finding of very-low safety significance (Green),and an associated Non-Cited Violation (NCV) of 10 CFR, Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to ensure that applicable regulatory requirements and the isolation condensers (ICs) design bases were correctly translated into procedures. Specifically, the licensee added steps to the IC control procedures which directed operators to secure the IC in order to prevent the water level in the shell from going below 3.5 feet. The added steps would result in the IC being shutdown when required to operate per the ICs design bases.
Description:
The safety-related IC system functions as a heat sink for decay heat removal from the reactor vessel following a reactor scram, and isolation from the main condenser. Each IC (one per unit) consists of two tube bundles immersed in a large water storage tank. The IC system operates by natural circulation. During operation, the IC tube side will contain reactor coolant water, and the shell side will contain clean demineralized water.
The inspectors observed that the licensee made changes to three IC control procedures.
The procedure changes directed operators to secure the IC on a low-shell side level condition. Specifically, the IC would be secured if the shell side level could not be maintained above 3.5 feet. At this height, the water level in the shell is just above the top of the tube bundles. The purpose of the procedure change was to prevent uncovering of the tube bundles in order to protect the IC for future availability (i.e., when the shell side level could be restored to ensure the IC would be available to respond to a Beyond Design Basis External Event). The procedures affected were as follows:
- DOP 1300-02, Automatic Operation of Isolation Condenser, Revision 25, Step G.4.a;
- DOP 1300-03, Manual Operation of the Isolation Condenser, Revision 34, Step G.9.a; and
- DAN 902(3)-3D-4, Isolation Condenser Level Hi/Low Annunciator Respond Procedure, Revision 12, Step B.2.a.
In the Updated Final Safety Analysis Report (UFSAR), Section 5.4.6 and the Technical Specifications (TS) Bases 3.5.3, the IC design basis was described as:
- (1) remove 252.5 Million British thermal units (MBtu)/hour, which is equivalent to the decay heat rate 8.8 minutes after the scram; and
- (2) provide sufficient decay heat removal capability for 20 minutes of operation without makeup water to the shell. The TS requires a number of surveillance requirements (SRs) be performed to ensure these bases are met, including the following:
- SR 3.5.3.1: Verify the IC system shellside water level 6 feet, and shellside water temperature 210°F; and
- SR 3.5.3.4: Verify the IC system heat removal capability to remove design heat load [i.e., 252.5MBtu/hour].
The licensee maintains the shell water level above 7 feet via administrative controls and monitors the temperature in an effort to remain well below the temperature limits.
However, the licensing bases of the isolation condenser states that water level of the shell is expected to go below 3.5 feet in order to mitigate a credited event under design bases conditions. As a result, procedural guidance, which would prevent the isolation condenser shell water level from going below 3.5 feet, would preclude the component from performing its design function under design bases conditions.
As part of immediate corrective actions the licensee marked the procedures for review in order to determine required changes. In addition, the licensee demonstrated the shell water temperature was monitored in order to maintain it well below TS requirements and the water level was administratively maintained at or above 7 feet. This provides additional margin to the IC to perform its design function prior to reaching the 3.5 foot limit.
The licensee documented the inspectors concern under AR 02506445, NRC Mod/5059 Inspection: ISCO [Isolation Condenser] Operating Procedures, dated May 28, 2015.
The licensee plans to evaluate the IC procedural changes and determine what modifications are needed. In addition, the licensee is considering performing an Apparent Cause Evaluation to evaluate the concern. The inspectors concluded that procedural changes were performed by the Operations Department without engaging the Engineering Department to ensure there were no adverse impacts to the IC or associated design and licensing bases.
Analysis:
The inspectors determined that the licensee added steps to the IC control procedures which directed operators to secure the IC in order to prevent the water level in the shell from going below 3.5 feet, which was contrary to 10 CFR, Part 50, Appendix B, Criterion III, Design Control, and was a performance deficiency.
Specifically, the added steps would result in the IC being shutdown when required to operate per the ICs design bases.
The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of Procedure Quality, and affected the cornerstone's objective of ensuring the availability, reliability and capability of the systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the inadequate procedures would drive the operators to stop the IC during a design bases event and prevent the IC from performing its design basis function of removing decay heat from the reactor.
In accordance with Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, Table 2, the inspectors determined the finding affected the Mitigating Systems cornerstone. As a result, the inspectors determined the finding could be evaluated using Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, for the Mitigating Systems cornerstone. The performance deficiency affected the design or qualification of mitigating structures, systems, and components (SSC); however, the SSC maintained its operability or functionality as applicable. Specifically, the licensee administratively maintains the shell water level at or above 7 feet and the temperature is maintained below 210 degrees Fahrenheit. This provides additional margin to the IC to perform its design function prior to reaching the 3.5 foot limit. Therefore, the inspectors answered "yes" to the Mitigating Systems Screening Question A.1 in Exhibit 2, and screened the finding as having very-low safety significance (Green).
The finding has a cross-cutting aspect in the area of Human Performance; Teamwork because the licensee did not communicate and coordinate activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, the Operations Department failed to communicate and coordinate with the Engineering Department when developing the procedural changes. [H.4]
Enforcement:
Title 10 CFR, Part 50, Appendix B, Criterion III, Design Control, requires, 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, from April 22, 2013, to May 29, 2015, the licensee failed to ensure that applicable regulatory requirements and the ICs design basis were correctly translated into procedures. Specifically, the licensee added steps to the IC control procedures which directed operators to secure the IC in order to prevent the water level in the shell from going below 3.5 feet. The added steps would result in the IC being shutdown when required to operate per the ICs design bases.
This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy because it was of very-low safety significance and was entered into the licensees CAP as Action Request 02506445, NRC MOD/5059 Inspection: ISCO Operating Procedures, dated May 28, 2015. The licensee will continue to administratively control level of the shell at or above seven feet to provide additional margin to the IC to perform its function. (NCV 05000237/2015007-01; NCV 05000249/2015007-01, Procedure Revisions Resulted in Isolation Condenser Unable to Meet Design Basis)
.2 Permanent Plant Modifications
a. Inspection Scope
The inspectors reviewed nine permanent plant modifications that had been installed in the plant during the last 3 years. This review included in-plant walkdowns for portions of the modified High Pressure Coolant Injection (HPCI) system to assess recent replacement of the auxiliary oil pump. The modifications were selected based upon risk-significance, safety significance, and complexity. The inspectors reviewed the modifications selected to determine if:
- the supporting design and licensing basis documentation was updated;
- the changes were in accordance with the specified design requirements;
- the procedures and training plans affected by the modification have been adequately updated;
- the test documentation as required by the applicable test programs has been updated; and
- post-modification testing adequately verified system operability and/or functionality.
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.
This inspection constituted nine permanent plant modification samples as defined in IP 71111.17-04.
b. Findings
Emergency Diesel Generator Usable Fuel Calculations Failed to Consider Appropriate Emergency Diesel Generator Frequency Variations
Introduction:
The inspectors identified a finding of very low safety significance (Green)and an associated NCV of 10 CFR, Part 50, Appendix B, Criterion III, Design Control, for the failure to account for increased fuel oil consumption during the development of the Emergency Diesel Generator (EDG) endurance calculations which resulted in non-conservative TS. Specifically, the licensee failed to account for the increased fuel oil consumption at an EDG frequency of 61.2 Hertz (Hz), and ensure that the minimum fuel oil level in the EDG day tanks, as required per TS 3.8.1.4, was adequate to support the EDGs mission time at 110 percent for one hour.
Description:
The TSs SR 3.8.1.2 allows an EDG frequency tolerance of +/- 2 percent.
This tolerance was based on RG 1.9, Application and Testing of Safety-Related Diesel Generators in Nuclear Power Plants, Revision 4, requirements that the EDG frequency recover to within +/- 2 percent of 60 Hz (i.e., 58.8 - 61.2 Hz) within a specified period during the sequencing of loads on the bus. Therefore, the EDGs could operate at a frequency of 61.2 Hz, which would be the worst-case scenario for loading of the EDGs.
Additionally, TS SR 3.8.1.4 verifies adequate level of fuel oil in the day tank and the bulk storage tanks. The level selected is to ensure adequate fuel oil for a minimum of one hour of EDG operation at 110 percent of full load for the day tank and approximately two days at 100 percent full load for the bulk storage tanks. The levels identified by TS SR 3.8.1.4 are 205 gallons and 10,000 gallons for the day tank and bulk storage tank, respectively.
During review of Calculation 10553-CALC-07, Dresden Station Emergency Diesel Generators Endurance Calculations, Revision 2, the inspectors questioned why the licensee based their fuel consumption on the EDGs operating at 110 percent at 60 Hz rather than 61.2 Hz as allowed by TS SR 3.8.1.2. The higher frequency would result in higher fuel consumption; therefore, would be more conservative. Specifically, the estimated fuel consumption at 110 percent loading at 61.2 Hz would be approximately 211.3 gallons/hour. The calculation did identify that the 2 percent frequency tolerance would result in higher fuel consumption; however, the more conservative frequency was only applied to calculating EDG loading at 100 percent. The calculation of the day tank level at 110 percent only considered the less conservative 60 Hz frequency.
The inspectors determined that the EDGs could operate at a steady state frequency up to 61.2 Hz according to TS SR 3.8.1.2. This would result in a higher fuel consumption that would exceed the TS 3.8.1.4 minimum volumetric fuel requirements. The TS 3.8.1.4 minimum fuel requirements were based on operating the EDGs at a frequency up to 60 Hz, rather than 61.2 Hz, which resulted in non-conservative TS.
The inspectors discussed the issue with the licensee and identified that the licensee had administrative procedures that would limit the frequency of the EDG to 60.5 Hz and would ensure the day tank level remained greater than 350 gallons. The Dresden Procedure DGA-12, Partial or Complete Loss of AC Power, Revision 73, ensures operators maintain frequency of the EDGs between 59.5 to 60.5 Hz. Additionally, Dresden Surveillance Procedure DOS 6600-14, Diesel Oil Transfer Pump Operation and Fuel Consumption Test, Revision 20, requires operators to maintain at least 350 gallons in the day tank. Therefore, the EDG would remain capable of performing its specified safety function. However, even with the administrative limits, the minimum fuel requirements identified in TS SR 3.8.1.4 would remain non-conservative since the fuel consumption would still be higher at the 60.5 Hz which is not represented in TS SR 3.8.1.4. The licensee captured this issue and entered it into their CAP as Action Request 02506869. The licensee intends to evaluate the effect of the increased frequency on their EDG Calculations.
Analysis:
The inspectors determined that the licensees failure to account for increased fuel oil consumption during the development of the EDG Calculation 10553-CALC-07, Dresden Station Emergency Diesel Generators Endurance Calculations, Revision 2, resulted in non-conservative TS and was contrary to 10 CFR, Part 50, Appendix B, Criterion III, Design Control, and was a performance deficiency. Specifically, the licensee failed to account for the increased fuel oil consumption at an EDG frequency of 61.2 Hz, and ensure that the minimum fuel oil level in the EDG day tanks, as required per TS 3.8.1.4, was adequate to support the EDGs mission time at 110 percent for one hour.
The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstones objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the licensee failed to account for the increased fuel oil consumption resulting from operation at a higher EDG frequency. Therefore, the licensee did not ensure that the minimum fuel oil level in the day tanks, as required per TS 3.8.1.4, was adequate to support the EDGs mission time at 110 percent for one hour.
In accordance with IMC 0609, Significance Determination Process, Attachment 609.04, Initial Characterization of Findings, Table 2, the inspectors determined the finding affected the Mitigating Systems cornerstone. As a result, the inspectors determined the finding could be evaluated using Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, for the Mitigating Systems cornerstone. The performance deficiency affected the design or qualification of a mitigating SSC; however, the SSC maintained its operability or functionality as applicable. Specifically, the licensee was able to demonstrate that adequate fuel in the storage tanks would be available to support the EDGs mission time when operating at the administratively controlled higher frequency limit specified in procedures. Therefore, the inspectors answered "no" to all the Mitigating Systems Screening Questions in Exhibit 2, and screened the finding as having very-low safety significance (Green).
This finding has a cross-cutting aspect in the area of Problem Identification and Resolution; Identification because the licensee did not thoroughly evaluate the EDG fuel oil consumption when considering EDG frequency variation. Specifically, the licensee failed to translate applicable design bases into specifications which resulted in non-conservative TS. [P.1]
Enforcement:
Title 10 CFR, Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established to ensure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions.
Contrary to the above, from May 25, 2011, until May 29, 2015, the licensee failed to ensure that applicable regulatory requirements and the design basis were correctly translated into specifications during development of the EDG Calculation 10553-CALC-07, Dresden Station Emergency Diesel Generators Endurance Calculations, Revision 2, which resulted in non-conservative TS. Specifically, the licensee failed to account for the increased fuel oil consumption at an EDG frequency of 61.2 Hz, and ensure that the minimum fuel oil level in the EDG day tanks, as required per TS 3.8.1.4, was adequate to support the EDGs mission time at 110 percent for one hour.
This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy because it was of very-low safety significance, and was entered into the licensees CAP as Action Request 02506869, NRC MOD/5059 Inspection:
EDG Fuel Consumption, dated May 28, 2015. The licensee is evaluating the effect of the increased frequency on their EDG calculations, and will continue to administratively control level in the fuel oil day tank high enough to ensure the SSC remains operable.
(NCV 05000249/2015007-02; NCV 05000237/2015007-02, EDG Usable Fuel Calculations Failed to Consider Appropriate EDG Frequency Variations)
OTHER ACTIVITIES
4OA2 Problem Identification and Resolution
.1 Routine Review of Condition Reports
a. Inspection Scope
The inspectors reviewed several corrective 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 of changes, tests, and experiments. 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 inspectors are listed in the attachment to this report.
b. Findings
No findings were identified.
4OA6 Management Meetings
.1 Exit Meeting Summary
On May 29, 2015, the inspectors presented the inspection results to Mr. Shane Marik, and other members of the licensee staff. The licensee personnel acknowledged the inspection results presented and did not identify any proprietary content. The inspectors confirmed that all proprietary material reviewed during the inspection was returned to the licensee staff.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- G. Baxa, Regulatory Assurance
- M. Budelier, Senior Manager Design Engineering
- D. Doggett, Regulatory Assurance
- B. Franzen, Regulatory Assurance Manager
- R. Gaston, Licensing Manager - Corporate
- T. Griffith, Senior Licensing Engineer
- B. Kapellas, Maintenance Director
- B. Madderom, Design Electrical Manager
- S. Marik, Site Vice President
- G. Morrow, Operations Director
- P. OBrien, Site Corrective Action Program Manager
- R. Osgood, Senior Nuclear Site Communications Specialist
- E. Rogers, Nuclear Oversight
- R. Schmidt, Chemistry Manager
- B. Surges, Work Control
- D. Walker, NRC Coordinator
- J. Washko, Plant Manager
- D. Wolverton, Design Mechanical Manager
U.S. Nuclear Regulatory Commission
- K. OBrien, Director, Division of Reactor Safety
- G. Roach, Senior Resident Inspector
LIST OF ITEMS
OPENED, CLOSED AND DISCUSSED
Opened and Closed
- 05000237/2015007-01; NCV Procedure Revisions Resulted in Isolation Condenser Unable
- 05000249/2015007-01 to Meet Design Basis (Section 1R17.1b.)
- 05000237/2015007-02; NCV EDG Usable Fuel Calculations Failed to Consider Appropriate
- 05000249/2015007-02 EDG Frequency Variations (Section 1R17.2b.)
Discussed
None