05000346/LER-2011-004, Regarding Direct Current System Design Issues
| ML11271A007 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 09/26/2011 |
| From: | Allen B FirstEnergy Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-11-272 LER 11-004-00 | |
| Download: ML11271A007 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown |
| 3462011004R00 - NRC Website | |
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5501 North State Route 2 FirstEnergy Nuclear Operating Company Oak Harbor. Ohio 43449 Barry S. Allen 419-321-7676 Vice President - Nuclear Fax: 419-321-7582 September 26, 2011 L-1 1-272 10 CFR 50.73 ATTN: Document Control Desk United States Nuclear Regulatory Commission Washington, D.C. 20555-0001
SUBJECT:
Davis-Besse Nuclear Power Station Docket Number 50-346, License Number NPF-3 Licensee Event Report 2011-004 Enclosed is Licensee Event Report (LER) 2011-004, "Direct Current System Design Issues." This LER is being reported in accordance with 10 CFR 50.73(a)(2)(i)(B),
10 CFR 50.73(a)(2)(ii)(B), and 10 CFR 50.73(a)(2)(v).
There are no regulatory commitments contained in this letter or its enclosure. The actions described represent intended or planned actions and are described for information only. If there are any questions or if additional information is required, please contact Mr. Patrick J. McCloskey, Manager, Site Regulatory Compliance, at (419) 321-7274.
Sincerely, Barry S.AAIlen GMW Enclosure: LER 2011-004-00 cc:
NRC Region III Administrator NRC Resident Inspector NRR Project Manager Utility Radiological Safety Board
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES 10/31/2013 (10-2010)
, the NRC may (See reverse for required number of not conduct or sponsor, and a person is not required to respond to, the digits/characters for each block) information collection.
- 3. PAGE Davis-Besse Nuclear Power Station 05000346 1 OF 5
- 4. TITLE Direct Current System Design Issues
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE I
- 8. OTHER FACILITIES INVOLVED SEQUENTIAL REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NO.
MONTH DAY YEAR 05000 FACILITY NAME DOCKET NUMBER-07 26 12011 2011 004 00 ]
09 26 2011 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply) 1E 20.2201(b)
E]
20.2203(a)(3)(i)
[I 50.73(a)(2)(i)(C)
[I 50.73(a)(2)(vii)
E]
20.2201 (d)
[] 20.2203(a)(3)(ii)
E]
50.73(a)(2)(ii)(A)
E]
50.73(a)(2)(viii)(A)
E]
20.2203(a)(1)
[] 20.2203(a)(4)
[] 50.73(a)(2)(ii)(B)
El 50.73(a)(2)(viii)(B)
E]
20.2203(a)(2)(i)
L 50.36(c)(1)(i)(A)
E]
50.73(a)(2)(iii)
El 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL EI 20.2203(a)(2)(ii)
E] 50.36(c)(1Xii)(A)
LI 50.73(a)(2Xiv)(A)
E] 50.73(a)(2)(x)
E] 20.2203(a)(2)(iii)
[] 50.36(c)(2)
[
50.73(a)(2)(v)(A)
LI 73.71 (a)(4) 100 E] 20.2203(a)(2)(iv)
E]
50.46(a)(3)(ii)
Z 50.73(a)(2)(v)(B)
LI 73.71(a)(5)
E] 20.2203(a)(2)(v)
E]
50.73(a)(2)(i)(A)
[
50.73(a)(2)(v)(C)
LI OTHER Specify in Abstract below E] 20.2203(a)(2)(vi)
Z 50.73(a)(2)(i)(B)
Z 50.73(a)(2)(v)(D) or in
DESCRIPTION OF EVENT
During a Component Design Basis Inspection, NRC inspectors raised concerns regarding the safety-related battery design basis. One aspect of this concern was that non-safety related DC loads, which are powered from the safety-related DC system, could become grounded and impose added loads on the DC buses from which they are powered. In particular, some non-safety related loads such as the reactor coolant pump back-up oil lift pumps [AB-P] and the emergency feed for containment lighting [FH] could be subjected to High Energy Line Break (HELB) and Loss of Coolant Accident (LOCA) environments.
A second aspect of this concern was that automatic transfer switches [EE-ASU] were installed to transfer their non-safety related loads between the two non-safety related inverters. These loads include the station annunciator [IB], the plant computer [ID], non-nuclear instrumentation [JG] and the integrated control system [JA]. Because the inverters are powered by the safety related station batteries, faults on the automatic transfer switches could be transferred from one DC power source to its redundant DC power source, potentially impacting the ability of both safety related battery divisions to perform their safety function. These two concerns were captured as NRC Unresolved Item (URI) 2007-007-05.
As part of the resolution to this URI, NRC Region III personnel requested information from the NRC Office of Nuclear Reactor Regulation (NRR) regarding the electrical separation design and licensing basis of the safety-related batteries. On July 26, 2011, the NRR staff completed their assessment of the issue and provided a copy of their evaluation to the DBNPS staff via the NRC Senior Resident Inspector. Based upon this evaluation from NRR to Region Ill, the following issues were identified:
- 1. The Updated Final Safety Analysis Report (UFSAR) states that non-safety related electrical equipment, whose failure under postulated environmental conditions could prevent satisfactory accomplishment of the specified safety-related electrical equipment required safety functions, is qualified as required. However, the Reactor Coolant Pump (RCP) backup oil lift pump motors and Containment Emergency Lighting Panel L49E1 are located inside containment and are not environmentally qualified. This could challenge the adequacy of electrical separation between the potentially grounded non-safety related equipment and the safety related batteries, and is inconsistent with the UFSAR.
- 2.
Six automatic transfer switches are installed to automatically transfer non-safety related loads such as Non-Nuclear Instrumentation channels X and Y, the Station Annunciators, the Plant Computer, and the Integrated Control System between Uninterruptible Distribution Panels YAU and YBU.
These panels normally receive power from inverters YVA and YVB, which receive power from the safety related DC power system. If a ground fault existed on one of these loads, the fault could be transferred from one power source to the redundant source, potentially impacting the ability of both safety related DC power sources to perform their required functions. This type of transfer is not permitted by Safety Guide 6, which is referenced by the UFSAR.
Based on this information from the NRC, on July 26, 2011, with the plant operating in Mode 1 at approximately 100 percent power, the breakers for the four RCP backup oil lift pump motors and for the emergency power supply to the Containment Lighting Panel were opened, and one train of instrumentation power was placed on its alternate power source from the AC system, eliminating the potential to impact both trains of the DC power system.
- 6. LER NUMBER
- 3. PAGE Y
SEQUENTIAL REVISION YEAR NUMBER NUMBER 2011 004 00
CAUSE OF EVENT
The plant's design basis is that non-safety related electrical equipment, whose failure under postulated environmental conditions could prevent satisfactory accomplishment of the specified safety-related electrical equipment required safety functions, is qualified as required. However, as stated previously, the RCP backup oil lift pump motors and an emergency supply to a Containment Lighting Panel are located inside containment and are not environmentally qualified. This condition has existed since the original design of the DBNPS in the 1970s. The cause analysis of this issue is not complete, therefore; the cause of this issue is indeterminate at this time. Results of the cause analysis will be provided in a revision to this Licensee Event Report.
The automatic transfer switches were installed in the early 1980s in response to NRC Bulletin 79-27, "Loss of Non-Class-l-E Instrumentation and Control Power System Bus During Operation." The intent of Bulletin 79-27 was to ensure that the loss of power to any bus in the plant electric distribution system would not result in control system actions that would cause a plant upset or transient condition requiring operator action concurrent with the loss of control room information upon which these actions would be based. The cause analysis of this issue is also not complete, therefore; the cause of this issue is also indeterminate at this time. As stated previously, the results of the cause analysis will be provided in a revision to this Licensee Event Report.
ANALYSIS OF EVENT
Industry guidance (NEI 00-01, Revision 2, "Guidance for Post Fire Safe Shutdown Circuit Analysis",
endorsed by Regulatory Guide 1.189) on high impedance faults notes that any such fault will likely progress rapidly and either blow the protective device (fuse or breaker), cause an open circuit (fail the faulted wire), or weld itself. For the non-qualified equipment in containment, any fault generated as a result of a harsh environment would therefore either clear in a very short period of time, or be of low enough current to not adversely affect the safety related DC System.
An analysis was performed to determine the incremental conditional core damage probability due to the specified non-safety related loads in containment failing in a harsh environment in a way that would adversely affect all trains of the safety related DC system. This analysis also determined the incremental conditional core damage probability due to one or more automatic transfer switches failing in a manner that would also adversely affect all trains of the safety related DC system. The conclusion of this analysis determined that the combined issues were of very low safety significance.
Reportability Discussion:
These issues with the design of the safety related DC System resulted in potential challenges to the electrical separation between non-safety related equipment and the safety-related batteries/DC System, and to the separation between the two trains of the DC System. Based on these potential challenges, the DC System may not have been able to meet the single failure criterion. Therefore, per the guidance of NUREG-1022, this condition represents an unanalyzed condition that significantly degraded plant safety, and is being reported per 10 CFR 50.73(a)(2)(ii)(B). Similarly, because this condition potentially affected both trains of the DC System, this condition being reported per 10 CFR 50.73(a)(2)(v) as a condition that could have prevented the fulfillment of the safety function of a system needed to: (A) shutdown the reactor and maintain it in a safe shutdown condition; (B) remove residual heat; (C) control
Reportability Discussion: (Continued) the release of radioactive material; and (D) mitigate the consequences of an accident. Also, because the DC System did not meet the plant licensing basis for operability, and since the plant operated with the subject equipment in this condition, this issue represents operation of the plant in a condition prohibited by the Technical Specifications, and is being reported per 10 CFR 50.73(a)(2)(i)(B). Verbal reporting of these issues as required by 10 CFR 50.72(b)(3)(ii)B) and 10 CFR 50.72(b)(3)(v)(A-D) was completed on July 26, 2011, via Event Number 47096.
CORRECTIVE ACTIONS
On July 26, 2011, the breakers for the four RCP backup oil lift pump motors and for the emergency power supply to the Containment Lighting Panel were opened to eliminate the possibility of a fault occurring on this equipment because of a harsh environment inside Containment.
Also on July 26, 2011, one train of instrumentation power was placed on its alternate power source from the AC system, eliminating the potential to impact both trains of the DC power system.
On August 3, 2011, the alternate power supply to each of the six automatic transfer switches was isolated, and both trains of instrumentation power were placed on their normal power source from the DC system. This was done to improve the reliability and power quality for the loads by powering them all from the station inverters instead of power some of the loads from their alternate power supply from the AC power system. It was determined the risk of an event that would result in a reactor trip was higher with one train of instrumentation power on its alternate power source than having the alternate power supply to the automatic transfer switches isolated.
Additional modifications are being planned to restore compliance with NRC Bulletin 79-27 for the six loads powered by the automatic transfer switches while maintaining separation of the DC power system trains. Modifications are also being planned to restore power to the non-qualified equipment in containment.
Additional corrective actions identified following completion of the cause evaluation for this event will be provided in a revision to this Licensee Event Report.
PREVIOUS SIMILAR EVENTS
DBNPS Licensee Event Report 2010-003 documented the inoperability of an Auxiliary Feedwater discharge control solenoid valve due to a ground in the non-essential DC System. The ground of 0.38 milliamps, which induced a voltage greater than the design capacity of the position controller board for the valve, had no adverse impact on the DC System. Therefore, the corrective actions taken for this 2010 event could not have been expected to address the design issues with the DC System described above. There have been no other Licensee Event Reports submitted for the DBNPS in the past three years regarding the DC System.