IR 05000237/2005009
| ML052590556 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 09/14/2005 |
| From: | Ann Marie Stone Division of Reactor Safety III |
| To: | Crane C Exelon Generation Co, Exelon Nuclear |
| References | |
| IR-05-009 | |
| Download: ML052590556 (33) | |
Text
September 14, 2005
SUBJECT:
DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 NRC SAFETY SYSTEM DESIGN AND PERFORMANCE CAPABILITY INSPECTION REPORT 05000237/2005009(DRS); 05000249/2005009(DRS)
Dear Mr. Crane:
On August 12, 2005, the U.S. Nuclear Regulatory Commission (NRC) completed a baseline inspection at your Dresden Nuclear Power Station. The enclosed report documents the inspection findings which were discussed on August 12, 2005, with Mr. D. Wozniak and other members of your staff.
The inspection examined activities conducted under your license as they relate to safety and to 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. Specifically, this inspection focused on the design and performance capability of the emergency diesel generator system and its associated support systems.
Based on the results of this inspection, two NRC-identified findings of very low safety significance, which involved violations of NRC requirements were identified. 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 Dresden Nuclear Power 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/
Ann Marie Stone, Chief Engineering Branch 2 Division of Reactor Safety Docket Nos. 50-237; 50-249 License Nos. DPR-19; DPR-25 Enclosure:
Inspection Report 05000237/2005009(DRS); 05000249/2005009(DRS)
w/Attachment: Supplemental Information cc w/encl:
Site Vice President - Dresden Nuclear Power Station Dresden Nuclear Power Station Plant Manager Regulatory Assurance Manager - Dresden Chief Operating Officer Senior Vice President - Nuclear Services Senior Vice President - Mid-West Regional Operating Group Vice President - Mid-West Operations Support Vice President - Licensing and Regulatory Affairs Director Licensing - Mid-West Regional Operating Group Manager Licensing - Dresden and Quad Cities Senior Counsel, Nuclear, Mid-West Regional Operating Group Document Control Desk - Licensing Assistant Attorney General Illinois Emergency Management Agency State Liaison Officer Chairman, Illinois Commerce Commission
SUMMARY OF FINDINGS
IR 05000237/2005009(DRS); 05000249/2005009(DRS); 07/25/2005 - 08/12/2005; Dresden
Nuclear Power Station, Units 2 and 3; Safety System Design and Performance Capability.
The inspection was a 3-week baseline inspection of the design and performance capability of the emergency diesel generator system and its associated support systems. The inspection was conducted by regional engineering inspectors. Two Green Non-Cited Violations 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
C
- Green.
The inspectors identified a Non-Cited Violation of Technical Specification Surveillance Requirement 3.7.2.1 regarding the failure to periodically verify the position of manual valves. Specifically, the licensee did not verify the correct position of 11 manual valves that were not locked, sealed, or otherwise secured in position in the diesel generator cooling water (DGCW) subsystem flow path associated with the DGCW pump motor coolers. The licensees corrective actions included verifying and then locking the affected valves in the open position and revising operating procedures to reflect that the affected valves are locked in the open position.
This finding was more than minor because it was associated with the mitigating systems attribute of configuration control, which affected the mitigating systems cornerstone objective of ensuring the availability and reliability of the DGCW system to respond to initiating events to prevent undesirable consequences. The finding was of very low safety significance based on the licensee verifying the valves were in their correct position and screened as Green using the SDP Phase 1 screening worksheet.
(Section 1R21.2.b.1)
- Green.
The inspectors identified a Non-Cited Violation of 10 CFR Part 50, Appendix B,
Criterion III, Design Control, due to the design basis emergency diesel generator (EDG) loading sequence during a loss of coolant accident/loss of offsite power not being correctly translated into procedures or instructions. Specifically, the loss of power procedure provided guidance to operate the plant outside the analyzed EDG loading sequence. The licensees corrective actions included evaluating the effect of the procedures unanalyzed load sequence and concluded that the EDG would have been capable of performing its safety function.
This finding was more than minor because it was associated with the attribute of procedure quality, which could have affected the mitigating systems cornerstone objective of ensuring the availability and reliability of the EDGs to respond to initiating events to prevent undesirable consequences. The finding was of very low safety significance based on the results of the licensees analysis and screened as Green using the SDP Phase 1 screening worksheet. (Section 1R21.2.b.2)
Licensee-Identified Violations
None.
REPORT DETAILS
REACTOR SAFETY
Cornerstone: Mitigating Systems and Barrier Integrity
1R21 Safety System Design and Performance Capability
Introduction:
Inspection of safety system design and performance verifies the initial design and subsequent modifications and provides monitoring of the capability of the selected systems to perform design bases functions. As plants age, the design bases may be lost and important design features may be altered or disabled. The plant risk assessment model is based on the capability of the as-built safety system to perform the intended safety functions successfully. This inspectable area verifies aspects of the mitigating systems cornerstone for which there are no indicators to measure performance.
The objective of the safety system design and performance capability inspection is to assess the adequacy of calculations, analyses, other engineering documents, and operational and testing practices that were used to support the performance of the selected systems during normal, abnormal, and accident conditions. Specific documents reviewed during the inspection are listed in the attachment to the report.
The systems and components selected were the emergency diesel generator (EDG)system and its associated support systems (one sample). This system was selected for review based upon:
- having high probabilistic risk analysis ranking;
- considered high safety significant maintenance rule system; and
- not having received recent NRC review.
The criteria used to determine the acceptability of the systems performance was found in documents such as:
- licensee Technical Specifications (TS);
- applicable Updated Final Safety Analysis Report (UFSAR) sections; and
- the systems' design documents.
.1 System Requirements
a. Inspection Scope
The inspectors reviewed the UFSAR, TS, system design basis documents, drawings, and other available design basis information, to determine the performance requirements of EDG system, and its associated support systems. The reviewed system attributes included process medium, energy sources, control systems, operator actions, and heat removal. The rationale for reviewing each of the attributes was:
Process Medium: This attribute needed to be reviewed to ensure that the EDGs would supply the required electrical loading under the design basis events of loss of offsite power and loss of offsite power concurrent with a loss of coolant accident.
Energy Sources: This attribute needed to be reviewed to ensure that the EDGs would start when called upon. In order to ensure that the EDGs would start, the following three subsystems were included in this review: direct-current control power, starting air, combustion air, and diesel fuel.
Controls: This attribute required review to ensure that the trips of the EDGs functioned as specified. This included review of trips bypassed during design basis events to ensure that the trips would not erroneously actuate and impact EDG operation.
Additionally, review of instrumentation, alarms, and indicators was necessary to ensure that operator actions would be accomplished in accordance with the design.
Operations: This attribute was reviewed because the emergency operating procedures permitted the operators to manually load components onto the EDGs during events.
Therefore, operator actions played an important role in the ability of the EDGs to achieve their function.
Heat Removal: This attribute required review to ensure that the heat generated while the EDGs were running can be effectively removed. In order to ensure that the heat generated while running the EDGs could be effectively removed, the following three subsystems were included in this review: ventilation air, diesel generator cooling water (DGCW), and lubrication oil cooling systems.
b. Findings
No findings of significance were identified.
.2 System Condition and Capability
a. Inspection Scope
The inspectors reviewed design basis documents and plant drawings, abnormal and emergency operating procedures, requirements, and commitments identified in the UFSAR and TS. The inspectors compared the information in these documents to applicable electrical, instrumentation and control, mechanical calculations, setpoint changes, and plant modifications. The inspectors used applicable industry standards, such as the American Society of Mechanical Engineers (ASME) Code and the Institute of Electrical and Electronics Engineers (IEEE) Standards, to evaluate acceptability of the systems design. Select operating experience was reviewed to ensure the issue was adequately evaluated and corrective actions implemented, as necessary. The inspectors also reviewed operational procedures to verify that instructions to operators were consistent with design assumptions.
The inspectors reviewed information to verify that the actual system condition and tested capability were consistent with the identified design bases. Specifically, the inspectors reviewed the installed configuration, the system operation, the detailed design, and the system testing, as described below.
Installed Configuration: The inspectors confirmed that the installed configuration of the EDG and its associated support systems met the design basis by performing detailed system walkdowns. The walkdowns focused on the installation and configuration of piping, components, and instruments; the placement of protective barriers and systems; the susceptibility to flooding, fire, or other environmental concerns; physical separation; provisions for seismic and other pressure transient concerns; and the conformance of the currently installed configuration of the systems with the design and licensing bases.
Operation: The inspectors performed a procedure walk-through of selected manual operator actions to confirm that the operators had the knowledge and tools necessary to accomplish actions credited in the design basis.
Design: The inspectors reviewed the mechanical, electrical, and instrumentation design of the EDGs to verify that the system and subsystems would function as required under design conditions. This included a review of the design basis, design changes, design assumptions, calculations, boundary conditions, and models as well as a review of selected modification packages. Instrumentation was reviewed to verify appropriateness of applications and setpoints based on the required equipment function.
Additionally, the inspectors performed limited analyses in several areas to verify the appropriateness of the design values.
Testing: The inspectors reviewed records of selected periodic testing and calibration procedures and results to verify that the design requirements of calculations, drawings, and procedures were incorporated in the system and were adequately demonstrated by test results. Test results were also reviewed to ensure automatic initiations occurred within required times and that testing was consistent with design basis information.
b. Findings
Two findings of very low safety significance associated with Non-Cited Violations (NCVs) were identified.
b.1 Technical Specification Requirements for Position Verification Not Met
Introduction:
The inspectors identified an NCV of TS Surveillance Requirement (SR) 3.7.2.1 having very low safety significance (Green) for the failure to periodically verify the position of 11 manual valves in the DGCW subsystem flow paths associated with the DGCW pump motor coolers.
Description:
On July 26, 2005, the inspectors identified 11 manual valves in the DGCW subsystem flow path associated with the DGCW pump motor coolers that were not locked, sealed, or otherwise secured in their required open position. Per TS SR 3.7.2.1, these valves were required to be periodically verified in their correct position every 31 days while in Mode 1, 2, or 3. The valves were not included in any surveillance to meet this requirement. The valves were in the flow path that provided cooling water to the motor bearings. Without bearing cooling, the DGCW pumps would not be able to perform their safety function. The licensee initiated an issue report (IR) to address the concern.
As a result, the licensee verified the valves were open and then locked the valves in the open position. In addition, the licensee revised the following procedures associated with system valve lineups and locked valve checklists to reflect that the affected valves were locked in the open position:
- (1) Unit 2 DOP 6600-M1, Unit 2 Standby Diesel Generator;
- (2) Unit 3 DOP 6600-M1, Unit 3 Standby Diesel Generator;
- (3) Unit 2/3 DOP 6600-M2, Standby Diesel Generator;
- (4) Unit 2 DOP 0040-M3, Unit 2 Accessible Locked Valve Checklist; and
- (5) Unit 3 DOP 0040-M4, Unit 3 Accessible Locked Valve Checklist.
Analysis:
The inspectors determined that the failure to perform TS SR 3.7.2.1 to ensure that manual valves in the DGCW subsystem flow paths associated with the DGCW motor coolers were in their correct position was a performance deficiency and a finding.
The inspectors determined that the finding was more than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Dispositioning Screening, because it was associated with the attribute of configuration control, which affected the mitigating systems cornerstone objective of ensuring the availability and reliability of the DGCW system to respond to initiating events to prevent undesirable consequences. A potentially mispositioned valve in the DGCW system flow path could render the DGCW pump and its associated EDG incapable of performing their required safety function.
The inspectors evaluated the finding using IMC 0609, Significance Determination Process, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, Phase 1 screening, and determined that the finding screened as Green because it was not a design issue resulting in loss of function per Generic Letter (GL) 91-18, did not represent an actual loss of a systems safety function, did not result in exceeding a TS allowed outage time, and did not affect external event mitigation. The affected valves were verified to be in their correct open position.
Enforcement:
Technical Specification SR 3.7.2.1 required, in part, that each DGCW subsystem manual valve in the flow path, that was not locked, sealed, or otherwise secured in position, be periodically verified in their correct position every 31 days while in Mode 1, 2, or 3.
Contrary to these requirements, on July 26, 2005, it was identified that since January 1997 (following implementation of License Amendment Numbers 150 and 145 for Unit 2 and Unit 3, respectively), the licensee did not periodically verify the position of 11 manual valves in the DGCW subsystem flow paths that were not locked, sealed, or otherwise secured in position, every 31 days while in Mode 1, 2, or 3. Because this finding is of very low safety significance and has been entered into the licensees corrective action program (IR00356996), it is being treated as an NCV, consistent with Section VI.A of the NRCs Enforcement Policy (NCV 05000237/2005009-01; 05000249/2005009-01). As part of its corrective action, the licensee verified and then locked the applicable valves in the correct open position and revised the operating procedures to reflect that the affected valves were locked in the correct position.
b.2 Unanalyzed Diesel Loading Sequence in Operating Procedures
Introduction:
The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, having very low safety significance (Green). Specifically, it was identified that the design basis EDG loading sequence during a loss of coolant accident/loss of offsite power (LOCA/LOOP) was not correctly translated into operating procedures.
Description:
The inspectors reviewed DGA-12, Partial or Complete Loss of AC Power, and identified a minor concern with the guidance for allowable loading levels on the EDGs. The licensee initiated IR00360160 to follow up on the issue. Based on the inspectors questions, during a subsequent review of DGA-12, the licensee determined that the procedural guidance did not implement the manually connected load sequence for LOCA/LOOP scenarios that was analyzed in the EDG loading calculation.
The EDG loading calculations analyzed all auto-connected and manually connected loads required for LOCA/LOOP conditions. Those loads manually connected included two containment cooling service water (CCSW) pumps that were started 10 minutes into an accident scenario. The EDG load calculation assumed that one low pressure coolant injection (LPCI) pump was tripped before the first CCSW pump was loaded onto the EDG, at which point the EDG was supplying one core spray pump, one LPCI, and one CCSW pump. The load sequence analyzed in the EDG loading calculation was also described in Chapter 8 of the UFSAR.
In contrast, procedure DGA-12, which implemented the manual load additions for LOCA/LOOP scenarios, instructed operators to start the first CCSW pump without tripping a LPCI pump. The procedure directed removal of a LPCI pump from the EDG only before starting the second CCSW pump.
The licensee initiated an IR and subsequently evaluated the effect of starting a CCSW pump before tripping a LPCI pump and concluded that the system would have performed its safety function.
Analysis:
The inspectors determined that the failure to provide operating procedure guidance in accordance with the design and licensing basis analyses was a performance deficiency and a finding. The inspectors determined that the finding was more than minor in accordance with IMC 0612, Appendix B, Issue Disposition Screening, in that the finding was associated with the attribute of procedure quality, which affected the mitigating systems cornerstone objective of ensuring the availability and reliability of the EDG to respond to initiating events to prevent undesirable consequences. Specifically, the EDGs could have been potentially overloaded based on the loading sequence, which could potentially render the safety-related EDG incapable of performing its required safety function.
The inspectors evaluated the finding using IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, Phase 1 screening, and determined that the finding screened as Green because it was not a design issue resulting in loss of function per GL 91-18, did not represent an actual loss of a systems safety function, did not result in exceeding a TS allowed outage time, and did not affect external event mitigation. The inspectors agreed with the licensee's analysis that, even with the increased loading as a result of the different sequencing of equipment, the EDG system would perform its safety function.
Enforcement:
Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, required, in part, that measures 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, as of August 8, 2005, the design basis EDG loading sequence for manually connected loads was not correctly translated into specifications, drawings, procedures, or instructions. Specifically, DGA-12 provided guidance to operate the plant outside the analyzed EDG loading sequence. Because this finding is of very low safety significance and has been entered into the licensees corrective action program (IR00361558), it is being treated as an NCV, consistent with Section VI.A of the NRCs Enforcement Policy (NCV 05000237/2005009-02; 05000249/2005009-02). The licensees initial corrective action included analyzing the procedure guidance load sequence and concluded that even with the increased loading on the EDG, the EDG system would perform its safety function.
.3 Components
a. Inspection Scope
The inspectors examined the EDG and its associated support systems to ensure that component level attributes were satisfied. Specifically, the following attributes of the EDG and associated support systems were reviewed:
Equipment/Environmental Qualification: This attribute verifies that the equipment is qualified to operate under the environment in which it is expected to be subjected to under normal and accident conditions. The inspectors reviewed design information, specifications, and documentation to ensure that the EDG and its associated support system components were qualified to operate within the temperatures specified in the environmental qualification documentation.
Equipment Protection: This attribute verifies that the EDG and its associated support systems were adequately protected from natural phenomenon and other hazards, such as high energy line breaks, floods, or missiles. The inspectors reviewed design information, specifications, and documentation to ensure that the EDG and its associated support systems were adequately protected from those hazards identified in the UFSAR which could impact their ability to perform their safety function.
Operating Experience: The inspectors reviewed condition reports, problem identification forms, and other documents to confirm that the licensee adequately evaluated industry information regarding EDG problems.
b. Findings
No findings of significance were identified.
OTHER ACTIVITIES (OA)
4OA2 Problem Identification and Resolution
.1 Review of Condition Reports
a. Inspection Scope
The inspectors reviewed a sample of EDG and its associated support systems problems that were identified by the licensee and entered into the corrective action program. The inspectors reviewed these issues to verify an appropriate threshold for identifying issues and to evaluate the effectiveness of corrective actions related to design issues. In addition, condition reports written on issues identified during the inspection were reviewed to verify adequate problem identification and incorporation of the problem into the corrective action program. 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 of significance were identified.
4OA5 Other Activities
.1 (Closed) Unresolved Item (URI) 05000237/2004006-03:
Potential Common Mode Failure Due to Hardened Lubricant on Safety-Related Merlin Gerin 4kV Electrical Breakers About 10 years ago, the licensee experienced several breaker failures due to hardened grease on the breaker mechanisms. The licensee replaced these breakers; however, in 2004, additional breaker failures occurred. The failures of the these breakers, which appeared to have a common cause, was considered unresolved pending completion of a cause determination and implementation of corrective actions. The inspectors reviewed the cause determination and concluded the recent failures were not similar to previous breaker failures, resulting from hardened grease on the breaker mechanism.
The recent failures were due to sluggish operation of the breaker release mechanism, which prevented an immediate re-closing of the affected breakers. The sluggish operation was caused by an oily substance on the release mechanism pivot points. The extent of condition investigation found no evidence of widespread stray lubrication and the source of the substance could not be positively determined, although a test breaker from the original Merlin Gerin breaker purchase had the same substance on it.
Based on the licensees cause determination, the previous corrective actions implemented to prevent breaker failures due to hardened grease would not have been expected to prevent these recent failures. Therefore, no performance deficiency or violation of NRC requirements were identified. The corrective actions taken were considered acceptable. Based on our review, URI 05000237/2004006-03 will be closed.
.2 (Open) Unresolved Item (URI) 50-237/02-06-02; 50-249/02-06-02:
Emergency Diesel Generator Testing During a previous inspection, the inspectors identified that calculated design basis loads for a LOCA/LOOP event exceeded the continuous rating of the EDGs. The inspectors noted that TS SR 3.8.1.3, SR 3.8.1.11, and SR 3.8.1.15 used a load band of 2340 to 2600 kilowatts (kW) based on 90 to 100 percent of the EDGs continuous rating of 2600 kW as basis for acceptability. However, the inspectors questioned whether the TS surveillance requirements should have been revised to reflect the design basis loads, which exceeded the EDG continuous ratings. The inspectors opened a URI to track the issue.
During this inspection, the inspectors reviewed information related to the previously described URI and identified additional concerns related to the level of compliance with SR 3.8.1.15 and the associated bases.
The EDGs at Dresden were rated as shown in the following table, along with the predicted post accident loads on the unit 2 EDG from the design basis loading calculation. The short term loads were the automatically connected loads required during core flooding (less than 10 minutes). The long term loads were manually connected and were required to ensure containment integrity.
EDG Ratings Unit 2 EDG Calculated Loads Continuous 10% Overload
- 2000 Hour Short term
(<10 minutes)
Long term
(>10 minutes)
Kilovolt Amps (kVA)3250 3575 2510 3249 kW 2600 2860 2228 2851 Kilovolt Amps Reactive (kVAR)1950 2145 1155 1557 Power Factor (pf)0.8 0.8 0.88 0.88 Predicted Loads Exceeding Surveillance Requirements The original concern of the URI regarding the lack of a surveillance requirement which enveloped the design basis loads remains open. The table above and chart below depict where the Dresden EDGs would operate within the 10 percent overload rating beginning at 10 minutes after a design basis event and for an extended period of time after manual loads were added onto the EDG. However, the current TS surveillance requirements, and licensee testing practices, only demonstrated the ability of the EDG to carry a load near this level for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
EDG Loads and TS SR Load 1500 2000 2500 3000 500 1000 1500 Time (min)
Load (kW)
Predicted Loads TS SR lower bound TS SR Upper bound Dresden load calculation limits appeared to have been established in 1981 based upon documents from the Systematic Evaluation Program (SEP) Topic VIII-2, which evaluated licensees against criteria from Regulatory Guide (RG) 1.9, Selection, Design, Qualification, and Testing of Emergency Diesel Generator Units Used as Class 1E Onsite Electric Power Systems at Nuclear Power Plants, Revision 2. Paragraph C.2 of RG 1.9 stated At the operating license stage of review, the predicted loads should not exceed the short-time rating (as defined in Section 3.7.2 of IEEE Std 387-1977
[Standard Criteria for Diesel-Generator Units Applied as Standby Power Supplies for Nuclear Power Generating Stations] of the diesel-generator unit). The licensee complied with this requirement.
As the licensee converted to the Improved Technical Specifications (ITS), the EDG surveillance requirements were taken from NUREG-1433, Standard Technical Specifications General Electric Plants, BWR/4, Volume 1, Revision 2, which referenced RG 1.9, Revision 3. A basic premise of this guide was that design basis loads do not exceed the continuous rating of the EDG. Paragraph C.1.3 of RG 1.9, Revision 3, stated At the operating license stage of review, the predicted loads should not exceed the continuous rating (as defined in Section 3.7.1 of IEEE Std 387-1984) of the diesel-generator unit. While the licensee asserts that they were not committed to RG 1.9, Revision 3, it was referenced in the Dresden TS.
This created a situation in which the licensees surveillance requirements did not envelope the design basis accident loads. The current 24-hour surveillance test procedure specified 2730 to 2860 kW for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. This was potentially less than the design load requirement of 2851 kW for an extended period. The licensee maintained that there was no commitment on their part to envelope the design basis loads during this surveillance test. However, the inspectors questioned whether this was an adequate demonstration of the EDGs capability to carry design basis loads.
Requirements for Testing at Power Factor The inspectors identified a closely related concern during this inspection regarding the licensees compliance with the requirements of TS SR 3.8.1.15, which was the 24-hour endurance run. SR 3.8.1.15 required the licensee to:
Verify each DG operating within the power factor limit operates for >24 hours:
a.
For 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> loaded $ 2730 kW and # 2860 kW; and b.
For the remaining hours of the test loaded $ 2340 kW and # 2600 kW.
The TS SR Bases established the power factor limit as # 0.85. Note 2 of SR 3.8.1.15 stated If grid conditions do not permit, the power factor limit is not required to be met.
Under this condition, the power factor shall be maintained as close to the limit as practicable.
The licensee developed surveillance procedure DOS 6600-12, Diesel Generator Tests Endurance and Margin/Full Load Rejection/ECCS/Hot Restart, to demonstrate compliance with SR 3.8.1.15. In this surveillance test, the EDG was connected to the grid and operated for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at a load between 2730 and 2860 kW and approximately unity power factor (+/- 300 kVARS). The EDG load was then lowered to between 2340 and 2600 kW for the remaining 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> of the test. Sometime during this 22-hour period, the power factor was adjusted by increasing to a band of 1550 to 1600 kVARS (0.83 - 0.86 pf) if possible, keeping the voltage on the emergency bus less that 4300 volts. This was held for only 10 minutes before returning to the +/- 300 kVAR band.
During a surveillance conducted on March 22, 2005, the licensee limited load to 1100 kVARS (0.91 pf) to stay within the voltage limits.
Prior to EDG testing, the licensee does not perform any evaluation as to the condition of the grid, with respect to whether or not the power factor limit can be achieved. Rather, regardless of whether the grid conditions may support testing at the power factor limit, the licensee has established a testing practice that only tests at this limit for 10 minutes.
The licensee asserted that this method has been approved by the NRC. In the licensees transition to the ITS format of NUREG-1433, they submitted to the NRC an amendment, Technical Specifications Changes for Dresden Nuclear Power Station, Units 2 and 3, dated March 3, 2000, which contained the following statement in the justification for deviation from ITS 3.8.1:
Therefore, it is not practicable to operate the generator in droop mode at the anticipated worst case accident power factor for long periods.
The inductive load will vary during the accident. VAR demand is dependent on the connected loads, starting of induction motors and system impedance. Raising the voltage regulator for an output of 1600 kVAR (equal to approximately 0.85 power factor at rated kW output), maintaining this output for a short period, then returning output to near unity power factor is more representative of system requirements.
The licensee maintained that since this amendment was approved with no exception taken to the above statement, they were complying with their surveillance requirements.
However, the surveillance procedure requiring a reactive load for only 10 minutes of the 24-hour run was not in literal compliance with the requirements and would appear to violate the intent of the surveillance requirements.
Summary Since these two issues were inter-related and could not be resolved with the licensee, the two issues will be combined into URI 50-237/02-06-02; 50-249/02-06-02. This concern will remain open pending resolution of the issues through a Task Interface Agreement (TIA) with the Office of Nuclear Reactor Regulation. In general, the resolution of these issues need to address whether the TS surveillance provides reasonable assurance of the EDGs capability to carry design basis loads and whether operating the EDG at the reactive load for only 10 minutes of the 24-hour run meets the supporting regulatory analysis and intent of the surveillance requirement.
4OA6 Meetings, Including Exits
.1 Exit Meeting
The inspectors presented the inspection results to Mr. D. Wozniak and other members of licensee management at the conclusion of the inspection on August 12, 2005.
Proprietary information reviewed by the inspectors was returned to the licensee at the conclusion of the inspection.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- C. Barajas, Shift Operations Superintendent
- J. Bashor, Work Management Director
- D. Blackwell, System Engineer
- G. Bockholdt, Maintenance Director
- J. Fox, Design Engineer
- D. Galanis, Design Engineering Manager
- G. Graff, Training
- J. Griffin, NRC Coordinator
- M. Kluge, Design Engineering
- D. Knox, Design Engineering
- G. Kusnik, Nuclear Oversight
- D. ORourke, Operations Support Manager
- R. Ruffin, Operations
- P. Salas, Regulatory Assurance Manager
- J. Strasser, Design Engineering
- J. Strmec, Chemistry, Environmental and Radwaste Manager
- M. Wegner, Nuclear Oversight
- G. Wilhelm, Nuclear Oversight
- D. Wozniak, Plant Manager
Nuclear Regulatory Commission
- C. Pederson, Director, Division of Reactor Safety
- A. M. Stone, Chief, Engineering Branch 2, DRS
- D. Smith, Senior Resident Inspector
- M. Sheikh, Resident Inspector
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
- 05000249/2005009-01 NCV Technical Specification Requirements for Position Verification Not Met (Section 1R21.2.b.1)
- 05000249/2005009-02 NCV Unanalyzed Diesel Loading Sequence in Operating Procedures (Section 1R21.2.b.2)
Closed
- 05000237/2004006-03 URI Potential Common Mode Failure Due to Hardened Lubricant on Safety-Related Merlin Gerin 4KV Electrical Breakers (Section 4OA5.1)
- 05000249/2005009-01 NCV Technical Specification Requirements for Position Verification Not Met (Section 1R21.2.b.1)
- 05000249/2005009-02 NCV Unanalyzed Diesel Loading Sequence in Operating Procedures (Section 1R21.2.b.2)
Discussed
- 05000249/2002006-03 URI Non-Conservative Emergency Diesel Generator Testing (Section 4OA5.2)