05000261/LER-2016-002, Unit No. 2, Regarding Unanalyzed Condition Related to Main Steam Line Break Inside Containment
| ML16165A311 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 06/13/2016 |
| From: | Glover R Duke Energy Progress |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RNP-RA/16-0035 LER 16-002-00 | |
| Download: ML16165A311 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 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)(v), Loss of Safety Function 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability |
| 2612016002R00 - NRC Website | |
text
(~ DUKE ENERGY Serial:J ~iJ..~fiJ035 United States Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261/RENEWED LICENSE NO. DPR-23 LICENSEE EVENT REPORT NO. 2016-002-00:
R. Michael Glover H. B. Robinson Steam Electric Plant Unit 2 Site Vice President Duke Energy Progress 3581 West Entrance Road Hartsville, SC 29550 0. 843 857 1704 F: 843 857 1319 Mike.Glovet@d11ke-e11ergy.com 10 CFR 50.73 UNANALYZED CONDTION RELATED TO MAIN STEAM LINE BREAK INSIDE CONTAINMENT Ladies and Gentlemen:
Pursuant to 10 CFR 50. 73, Duke Energy Progress, Inc. is submitting the attached Licensee Event Report. Administrative controls have been implemented to ensure full compliance with NRC regulations. Should you have any questions regarding this matter, please contact Mr. S. Connelly, Acting Manager - Nuclear Regulatory Affairs at (843) 857-1569.
This document contains no new regulatory commitments.
Sincerely, qJt~,~
~o~
R. Michael Glover Site Vice President RMG/jmw Attachment c:
Regional Administrator, NRC, Region II NRC Resident Inspector, HBRSEP D. Galvin, NRR
United States Nuclear Regulatory Commission Attachment to Serial: RNP-RN16-0035 5 Pages (including this cover page)
H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 UNANALYZED CONDTION RELATED TO MAIN STEAM LINE BREAK INSIDE CONTAINMENT
NRC FORM 366 U.S. NUCLEAR REGULA TORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 1013112018 (11-2015)
Estimated burde~ per response to comply with this mandatory collection request: BO hours.
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Reported lessons learned are incorporated into the licensing process and fed back to industry.
Send comments regarding burden estimate lo the FOIA, Privacy and Information Collections
~..
~) LICENSEE EVENT REPORT (LER)
Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by internet e-mail to lnfocollects.Resource@nrc.gov, and lo the Desk Officer, Office of Information and (See Page 2 for required number of Regulatory Affairs, NEOB-10202, (3150-0104), Office of Management and Budget, Washington, digits/characters for each block)
DC 20503. If a means used to impose an information collection does not display a currenUy valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.
- 3. PAGE H. B. Robinson Steam Electric Plant, Unit No. 2 05000 261 1 OF 4
- 4. TITLE Unanalyzed Condition Related to Main Steam Line Break Inside Containment
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED I
SEQUENTIAL I REV FACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR MONTH DAY YEAR NUMBER NO.
05000 FACILITY NAME DOCKET NUMBER 04 13 2016 2016 -
002 - 00 06 13 2016 05000
- 9. OPERA TING MODE
- 11. THIS REPORT IS SUBMITIED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
D 20.2201 (b}
D 20.2203(a)(3)(i)
D 50.73(a)(2)(ii)(A)
D 50.73(a)(2)(viii)(A) 1 D 20.2201 (d)
D 20.2203(a)(3)(ii)
[{] 50.73(a)(2)(ii)(B)
D 50.73(a)(2)(viii)(B)
D 20.2203(a)(1)
D 20.2203(a)(4)
D 5o.73(a)(2)(iii)
D 50.73(a)(2)(ix)(A)
D 20.2203(a)(2)(i)
D 50.36(c)(1)(i)(A)
D 50. 73(a)(2)(iv)(A)
D 50.73(a)(2)(x)
- 10. POWER LEVEL D 20.2203(a)(2)(ii)
D 50.36(c)(1)(ii)(A)
D 50.73(a)(2)(v)(A)
D 13.11 (a}(4)
D 20.2203(a}(2)(iii)
D 5o.3s<c><2>
D 50.73(a)(2)(v)(B)
D 13.11 (a}(5)
D 20.2203(a)(2)(iv)
D 5o.4s(a)(3)(ii)
D 50.73(a)(2)(v)(C)
D 13.77(a)(1) 100 D 20.2203(a)(2)(v)
D 50.73(a)(2)(i)(A)
D 50. 73(a)(2}(v)(D)
D 13.77(a)(2)(i)
D 20.2203(a)(2)(vi)
D 50.73(a)(2)(i)(B)
D 50.73(a)(2)(vii)
D 73.77(a)(2)(ii)
D 50.73(a)(2}(i)(C)
D OTHER Specify in Abstract below or in EVENT DESCRIPTION (Continued)
An active failure ofa FRBV whereby the valve fails to close will increase the secondary mass available for release to the containment as well as increase the heat transferred to the secondary fluid. During an ESF AS initiation or Feedwater Isolation signal, the condensate pumps[P] and the heater drain pumps[P] (until low level trip) would continue to provide feedwater to the failed open FRBV. Increased secondary mass available for release to the containment will result in a higher peak containment pressure that would exceed the containment design pressure of 42 pounds per square inch gauge (PSIG), and can alter the offsite and control room total effective dose equivalent (TEDE) doses.
The relevant analyses that appear to be unanalyzed for this scenario are Updated Final Safety Analysis Report (UFSAR) Chapter 6.2.1.4, "Containment Analysis for Postulated Secondary System Pipe Ruptures," (WCAP-15305) related to containment design, and UFSAR Chapter 15.1.5, "Main Steam line Break Event," (ANP-2560) related to radiological consequences for both I 02% power and Hot Zero Power conditions.
WCAP-15305 and ANP-2560 do not address a single failure ofa feedwater bypass valve, and based on the estimated flow rates through the feedwater bypass valves following a feedwater pump trip. The resulting increase in secondary mass available for release to the containment under this postulated single failure is considered an unanalyzed condition.
A Prompt Determination of Operability (PDO) was prepared that recommended the Containment be considered operable but in a non-conforming condition while the plant was in Modes 1, 2, and 3, and that the FRBVs should remain closed and/or isolated while operating in these Modes. Administrative controls and corrective actions have been established to ensure this configuration.
CAUSAL FACTORS The cause analysis resulting from this event has determined that during the establishment of the plant's current licensing basis for MSLB inside containment, the postulated single failure of the feedwater regulating bypass valve to close was not considered. This is a combination of original design issues and plant engineering decisions and activities that took place during the 1980s.
CORRECTIVE ACTIONS
Corrective actions in response to the condition and to restore compliance with regulations are listed below.
Immediate:
- 1. CR 2012658 generated to capture conditions in Corrective Action Program.
- 2. CR 2020495 generated to capture extent of condition in Corrective Action Program.
- 3. CR 2018710 generated to capture incorrect response to NRC IE Bulletin 80-04.
- 4. Standing Instruction 16-0003 implemented and Caution Tags placed on the FRBVs.
Planned:
- 1. Revise plant containment analysis input documents to complete a single failure analysis to address all the secondary side assumptions and make provisions for implementing further corrective actions as necessary.
- 2. Complete engineering change to update single failures to containment analysis inputs and licensing basis for MSLB to incorporate the feedwater isolation single failures as defined in related calculation into the current licensing basis.
SAFETY ANALYSIS
Failure of the feedwater bypass valves to automatically close following a main steamline break inside containment was found to be an issue as a single failure may lead to containment overpressure. This scenario was assessed utilizing probabilistic risk assessment methodology to determine the likelihood of this scenario occurring. Using conservative modeling techniques and conservative failure frequencies for the postulated events, this scenario was assessed and determined to have a very low safety significance. Two different failure modes were postulated for the valves: failure to close on demand and spuriously open during operation (and relevant common cause failures). These failure modes were evaluated for time frames determined from plant operation data and limited operator action was credited. The assessment determined that the scenario has a low safety significance for the periods with the valve(s) open. In addition this analysis did not credit any other potentially mitigating system to avert core damage in order to bound the analysis.
ADDITIONAL INFORMATION
A review of the industry Operating Experience (OE) was performed for a five (5) year span from 2011 to 20 I 6. The search produced 433 reports. Several reports were reviewed for applicability based on subject titles, however no OE reviewed had the same circumstances as the subject condition, where an inadequate single failure analysis of the system was utilized in containment structure design pressure determination.
Energy Industry Identification System (EIIS) codes for systems and components relevant to this event are identified in the text of this document within brackets [ ].