05000413/LER-2014-002, Regarding Unanalyzed Condition Due to Deviations from Fire Protection Current Licensing Basis Identified During NFPA 805 Transition
| ML14155A128 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 06/02/2014 |
| From: | Henderson K Duke Energy Carolinas, Duke Energy Corp |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| CNS-14-062 LER 14-002-00 | |
| Download: ML14155A128 (8) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(1), Submit an LER, Invalid Actuation 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2) |
| 4132014002R00 - NRC Website | |
text
Kelvin Henderson DUKE We President ENERGY.CaabNula tin CNOIVP 4800 Concord Road York, SC 29745 o: 803.701.4251 f: 803.701.3221 CNS-14-062 June 2, 2014 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001
Subject:
Duke Energy Carolinas, LLC (Duke Energy)
Catawba Nuclear Station, Units 1 and 2 Docket Nos. 50-413 and 50-414 Licensee Event Report (LER) 413/2014-002-00 Pursuant to 10 CFR 50.73(a)(1) and (d), attached is LER 413/2014-002-00, entitled "Unanalyzed Condition Due to Deviations from Fire Protection Current Licensing Basis Identified During NFPA 805 Transition".
This report is being submitted in accordance with 10 CFR 50.73(a)(2)(ii)(B).
There are no regulatory commitments contained in this letter or its attachment.
This event is considered to be of no significance with respect to the health and safety of the public.
If there are any questions on this report, please contact Tolani Owusu at (803) 701-5385.
Sincerely, Kelvin Henderson Vice President, Catawba Nuclear Station TEO/s Attachment: Licensee Event Report
U. S. Nuclear Regulatory Commission June 2, 2014 Page 2 xc (with attachment):
V.M. McCree Regional Administrator U.S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, GA 30303-1257 G.E. Miller (addressee only)
NRC Project Manager U.S. Nuclear Regulatory Commission Mail Stop 8 G9A 11555 Rockville Pike Rockville, MD 20852-2738 G.A. Hutto, III NRC Senior Resident Inspector Catawba Nuclear Station INPO Records Center 700 Galleria Place Atlanta, GA 30339-5957
NRC FORM 386 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 0113112017 (02-2014)
Estimated burden per response to comply with this mandatory collection request 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br />.
Reported lessons learned are Incorporated into the licensing process and fed back to industry.
Send comments regading burden estimate to the FOIA, Piac end Information Collections LICENSEE EVENT REPORT (LER)
Branch (T-5 F53), U.S. Nucear Regulatory Comrnisson, Washington, DC 20555-0001, or by LICENSE e-mEll to lnfocohects.Resource@nrc.gov, and to the Desk Officer, Office of Information and (See Page 2 for required number of Regulatory Alars, NEOB-10202, (3150-0104), Ofte of Management and Budget Washington, DC 20503. If a means used to Impose an Information collection does not display a currently vid OMB digits/characters for each block) control number, the NRC may not conduct or sponsor, and a person Is not required to respond to, the information collection.
- 3. PAGE Catawba Nuclear Station, Unit 1 05000413 Page 1 of 6
- 4. TITLE Unanalyzed Condition Due to Deviations from Fire Protection Current Licensing Basis Identified During NFPA 805 Transition
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH DAY YE SEQUENTIAL REV FACILITY NAME DOCKET NURSER MONT O^
Y*
YEAR SE UNTUMBER NO.
Mr oY Catawba Nuclear Station, Unit 2 05000414 FACILITY NAME DOCKET NUMBER 04 02 2014 2014 -
002
- - 00 06 02 2014
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
El 20.2201(b)
El 20.2203(a)(3Xi)
[] 50.73(aX2)(i)(C)
El 50.73(a)(2XvWi)
El 20.2201(d)
El 20.2203(aX3Xii)
[] 50.73(aX2XiiXA)
[] 50.73(aX2XvAIiXA)
El 20.2203(aXl)
El 20.2203(aX4)
[
50.73(aX2XiixB)
El 50.73(aX2XvIiiXB)
El 20.2203(a)(2Xi)
[] 50.36(cX1 )(iXA)
E] 50.73(a)(2Xiii)
El 50.73(aX2Xix)(A)
- 10. POWER LEVEL El 20.2203(aX2Xii)
El 50.36(cX1)(iiXA)
El 50.73(a)(2)(ivXA)
El 50.73(a)(2Xx)
El 20.2203(aX2Xiii)
El 50.36(c)(2)
El 50.73(aX2XvXA)
El 73.71(a)(4) 10 20.2203(a)(2Xiv)
El 50.46(a)(3)ii)
(0 50.73(a)(2)(vXB)
El 73.71(aX5) 1E 20.2203(a)(2Xv)
El 50.73(aX2XiXA)
El 50.73(aX2XvXC)
El OTHER El 20.2203(aX2)(vi)
El 50.73(a)(2XiXB)
El 50.73(aX2)(vXD)
Specify In Abstact below or in
BACKGROUND This event is being reported in accordance with 10 CFR 50.73(a)(2)(ii)(B) as an event or condition that resulted in the nuclear power plant being in an unanalyzed condition that significantly degraded plant safety.
Energy Industry Identification System (EIIS) codes are identified in the text as [EIIS: XX]. Catawba Nuclear Station (CNS), Units 1 and 2 are Westinghouse four-loop Pressurized Water Reactors (PWR) [EIIS: RCT].
As part of the transition to National Fire Protection Association (NFPA) 805 and associated License Amendment Request (LAR) development, the NFPA 805 project team evaluated the CNS Fire Protection Program against the NFPA 805 Nuclear Safety Performance Criteria (NFPA 805, 2001 edition, Section 1.5.1). As part of this review, 19 of the CNS original 53 Fire Areas were evaluated for a different Safe Shutdown (SSD) train. Although the assured SSD train changed for these 19 areas and the NFPA 805 Nuclear Safety Performance Criteria is different from the existing CNS Current Licensing Basis (CLB)
(NUREG-0800), there are multiple circuit issues identified that present a risk to the CNS ability to safely shut down in the event of a fire. Furthermore, as stated in NEI 00-01, "Guidance For Post Fire Safe Shutdown Circuit Analysis", Revision 1, dated January 2005, there is industry recognition that "determining whether an issue is beyond the licensing basis may be difficult because the licensing basis is not well understood or documented". As a result, NEI 00-01 provides guidance for addressing "potential risk-significant issues regardless of compliance with the licensing basis".
EVENT DESCRIPTION
On April 2, 2014, when this issue was determined to be LER reportable, Units 1 and 2 were operating in Mode 1 at 100% power. However, both units operated in all modes throughout the existence of this issue.
In May 2013, the analysis performed in support of the NFPA 805 LAR development identified 802 Variances From Deterministic Requirements (VFDRs). All of the 802 VFDRs were evaluated during the Fire Risk Evaluation (FRE) process as part of the NFPA 805 LAR development. The risk insights from the FRE process were used to determine which VFDRs were potentially risk significant. Those VFDRs that required a plant modification to ensure an acceptable risk were evaluated as a risk-significant deficiency and a compensatory measure was implemented. Most of the risk-significant VFDRs that required a plant modification were associated with a Non-Coordinated Load (NCL) issue.
The following table identifies Fire Areas that involve either of the following: 1) VFDRs that require a plant modification, or 2) a compensatory measure that was implemented where a potential concern existed with respect to NRC Information Notice (IN) 92-18, "Potential for Loss of Remote Shutdown Capability During a Control Room Fire", dated February 28, 1992. This conservative approach was implemented to ensure that further analysis could be performed for the individual motor-operated valves that could have the potential for a IN 92-18 type of circuit failure. There were 160 potential IN 92-18 issues, 32 of which were related to the ability to achieve hot standby (HSB). The potential IN 92-18 issues associated with HSB
were evaluated in the FREs and are not risk significant. The balance of the potential IN 92-18 issues are related to achieving cold shutdown (CSD).
Fire Areas that require plant modifications as part of the FRE and have potential IN 92-18 issues were evaluated as having safe shutdown separation deficiencies.
The total affected Fire Areas with VFDRs that require plant modifications and/or IN 92-18 issues are summarized as follows:
Removal and Containment Spray Pumps Room 2
Unit 2 Auxiliary N (0)
Y 02-VFDR-08 Feedwater Pump Room 3
Unit 1 Auxiliary N (0)
Y 03-VFDR-07 Feedwater Pump Room 4
Auxiliary Building Y (65)
N None Common Area El 543 5
Unit 2 Electrical Y (4)
N None Penetration Room El 560 6
Unit I Electrical Y (3)
N None Penetration Room El 560 7
Unit 2 B Train Essential Y (1)
Y 07-VFDR-02, Switchgear Room 03 8
Unit 1 B Train Essential Y (1)
N None Switchgear Room 9
Unit 2 Battery Room Y (4)
Y 09-VFDR-02 10 Unit I Battery Room Y (1)
N None 11 Auxiliary Building Y (43)
N None Common Area El 560 14 Unit 2 A Train Essential Y (4)
N None Switchgear Room
lit 1 A I rain _ssent Switchgear Room 17 Unit I Cable Room Y (1)
N None 18 Auxiliary Building Y (13)
N None Common Area El 577 20 Unit I Electrical N (0)
N
- Cable route Penetration Room El deficiency 594 associated with 1KSI 45 Unit I Cable Room Y (5)
Y 45-VFDR-06 Corridor 46 Unit 2 Cable Room Y (2)
N None Corridor 48 Unit 2 Interior Main N (0)
Y 48-VFDR-01, Steam Doghouse 03, 07 49 Unit I Interior Main N (0)
Y 49-VFDR-01, Steam Doghouse 03, 07, 08, 09, 10 RB1 Unit 1 Reactor Building Y (3)
Y RB1-VFDR-02, 03, 05, 06, 11, 12, 13, 14, 34, 35, 36, 37, 45, 46 RB2 Unit 2 Reactor Building Y (2)
Y RB2-VFDR-02, 03, 05, 06, 12, 13, 14,15, 16, 36, 37, 38, 39, 46, 47, 53
- This is not a VFDR because the success path for FA 20 changed from the Standby Shutdown System to Train A.
From the extent of condition review of the above issues, several cold shutdown issues were discovered in three more Fire Areas that require compensatory actions. The ability of the plant to safely achieve and maintain cold shut down could potentially be adversely impacted by fire induced cable failures to the listed components in the following locations:
- 1. FA 16 (Unit 2 Cable Room) 0 Nuclear Service Water System (RN) strainer 2B and its backwash valve 2RN-40B
Refueling Water Storage Tank (RWST) isolation valves to the spent fuel pool 2KF-1 01 B and 2KF-103A Unit 2 B and C Steam Generator (SG) Power Operated Relief Valve (PORV) isolation valves 2SV-28A and 2SV-26B Component Cooling Water System (KC) recirculation isolation valve 2KC-54B Potential loss of normal supply to 600V MCC 2EMXH
- 2. FA 21 (Main Control Room)
Nuclear Service Water System (RN) strainer 2A and its backwash valve 2RN-30A Unit 2 Refueling Water Storage Tank (RWST) isolation valves to the spent fuel pool 2KF-1 01 B and 2KF-103A; Nuclear Service Water System (RN) strainer 1A and its backwash valve 1RN-30A Unit 1 Refueling Water Storage Tank (RWST) isolation valves to the spent fuel pool 1 KF-1 01 B and 1KF-103A Unit 2 B and C Steam Generator (SG) Power Operated Relief Valve (PORV) isolation valves 2SV-28A and 2SV-26B Unit 1 B and C Steam Generator (SG) Power Operated Relief Valve (PORV) isolation valves 1 SV-28A and 1 SV-26B Unit 1 Component Cooling Water System (KC) recirculation isolation valve 1KC-51A Unit 2 Component Cooling Water System (KC) recirculation isolation valve 2KC-51A
- 3. FA 22 (Auxiliary Building Common Area El 594)
Power supplies associated with 600V MCCs 1 EMXG, 1 EKPG, 2EKPH, 2EMXH affecting Control Room Ventilation System (VC) and Nuclear Service Water System (RN)
Control Room Ventilation System (VC) outside isolation valves 1VC-5B, 1VC-6A, 2VC-5B, 2VC-6A Unit 1 B Steam Generator (SG) Power Operated Relief Valve (PORV) isolation valve 1 SV-28A Unit 2 B Steam Generator (SG) Power Operated Relief Valve (PORV) isolation valve 2SV-28A CAUSAL FACTORS The issues identified in this LER were attributed to latent design deficiencies. Therefore, no additional analysis was conducted for this LER in an attempt to evaluate the cause of these design deficiencies.
CORRECTIVE ACTIONS
Completed 1.
Fire watches and/or transient control of combustible materials were established in the Fire Areas associated with the circuit and cable route issues that present risk-significant concerns or safe shutdown non-compliances.
Planned
- 1.
As part of the transition to NFPA-805, the CNS current licensing basis will be revised to maintain a Fire Protection Program that complies with this standard. Furthermore, the required plant modifications committed to as a result of the NFPA 805 transition will be implemented as part of the transition plan.
SAFETY ANALYSIS
The fire related non-compliance conditions identified in this LER are based on postulated fire scenarios that have not occurred at CNS. Upon discovery of the identified conditions, appropriate fire watches and transient control of combustible materials were implemented for the affected Fire Areas. Additionally, fire detection and suppression equipment in the affected Fire Areas was generally functional throughout this event and at the time of discovery. Exceptions were logged in the CNS fire protection impairment log and appropriate compensatory measures were implemented. Therefore, there is no present safety concern associated with the identified conditions. Probabilistic Risk Analysis is presently conducting a detailed analytical review of the identified conditions to assess the past safety significance of this event. When completed, the results of this review will be submitted to the NRC in a supplement to this LER. The identified non-compliances will be fully remediated as part of the CNS transition to NFPA 805.
ADDITIONAL INFORMATION
The issue described in this LER was discovered as part of the transition to NFPA 805. No fire protection related LERs have been reported to the NRC during the most recent three-year period. Therefore, the issue described in this LER has been determined to be non-recurring in nature.
The issue described in this LER is not considered to be reportable to the INPO Consolidated Event System (ICES) (formerly called the Equipment Performance and Information Exchange (EPIX) program).
The issue described in this LER is not considered to constitute a Safety System Functional Failure. There was no release of radioactive material, radiation overexposure, or personnel injury associated with the issue described in this LER.