05000324/LER-2005-002, Re Automatic Shutdown Due to Condensate System Transient
| ML051650239 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 06/08/2005 |
| From: | Hinds D Progress Energy Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| BSEP 05-0072 LER 05-002-00 | |
| Download: ML051650239 (7) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 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 |
| 3242005002R00 - NRC Website | |
text
OM Progress Energy JUN 0 8 2005 SERIAL: BSEP 05-0072 10 CFR 50.73 U. S. Nuclear Regulatory Commission ATM N: Document Control Desk Washington, DC 20555-0001
SUBJECT:
Brunswick Steam Electric Plant, Unit No. 2 Docket No. 50-324/License No. DPR-62 Licensee Event Report 2-2005-002 Ladies and Gentlemen:
In accordance with the Code of Federal Regulations, Title 10, Part 50.73, Carolina Power
& Light Company, now doing business as Progress Energy Carolinas, Inc., submits the enclosed Licensee Event Report. This report fulfills the requirement of a written report with sixty days of a reportable occurrence.
Please refer any questions regarding this submittal to Mr. Edward T. O'Neil, Manager - Support Services, at (910) 457-3512.
Sincerely, David H. Hinds Plant General Manager Brunswick Steam Electric Plant GLM/glm
Enclosure:
Licensee Event Report Progress Energy Carolinas. Inc.
Brunswick Nuclear Plant P.O. Box 10429 Southport, NC 28461
Document Control Desk BSEP 05-0072 / Page 2 cc:
U. S. Nuclear Regulatory Commission, Region II ATTN: Dr. William D. Travers, Regional Administrator Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, GA 30303-8931 U. S. Nuclear Regulatory Commissiorn ATTN: Mr. Eugene M. DiPaolo, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 U. S. Nuclear Regulatory Commission (Electronic Copy Only)
A1TN: Ms. Brenda L. Mozafari (Mail Stop OWFN 8G9) 11555 Rockville Pike Rockville, MD 20852-2738 Ms. Jo A. Sanford Chair - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-0510
- - V NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 EXPIRES 06/30/2007 (6-2004)
Esinated Auden per meporse lo cwndly Zl ftis mnwomry Hahonnni cledon redqust 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> Reop~ed lessons Werred ana kcaporaed kib M fluiming process &xt fed back fo LICENSEE EVENT REPORT (LER) ds ed co... aregangbtend O
Branc U-5 F52, U-11 Nudeer RePa Cawerissi Wasbigtn DC 205M, or by (See reverse for required nmaber of Int e-fnl to Weds~flrgov and lo She Desk Officer. OlrdI ftnmon Wnd digits/characters for each block)
Ragdatry Alim, NE0131202 (3150-014 Office o Manageirtl and Budg. WasthntoR DC20503. 1a rrs used tolrnose an hnerou cba Andoes nd sday a unyvald v~ ors atd "rrw tie MIirey n, ctxlc or sHor.
awid a pe isa Istrered to espndLA tie wanton
- 3. PAGE Brunswick Steam Electric Plant (BSEP), Unit 2 05000324 1 OF 5
- 4. TITLE Automatic Shutdown Due to Condensate System Transient
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED MONTH I DAY YEAR YEAR' lSEQUENTIAL I REV MONTH DAY YEAR FILIW NAME DOCKET NUMBER NOEI
___ONTHDAYYEAR 05000 04 l09 2005 2005 002
)
16 08 2005 FACIr Y NAME 050NB REPOTISLJB~rTEPURUANTOTHREOUREMNTS05000
- 9. OPERATING MODE 11.THIS REPORT ISSWAFEI)PURSUANT THEREQUIREMENTSOF1OCFR§: (Checkoneormore) 13 20.2201 (b) 0 20.2203(a)(3)(I)
El 50.73(a)(2) (i) (C) 0 50-73(a)(2)(vii) 0 20.2201 (d)
E 20.2203(a)(3)(ii) 0 50.73(a)(2)(ii)(A) 0 50-73(a)(2)(viii)(A) 0 20.2203(a)(1) 0 20o203(a)(4) 0 50.73(a)(2)(ii)(B) 0 50.73(a)(2)(viii)(B)
Dl 20.2203(a)(2)(i) 0 50.36(c)(1)(1)(A) 0 50.73(a)(2)(fif) 0 50.73(a)(2)Cix)(A)
- 10. POWER LEVEL 0 20.2203(a)(2)(ii) 0 50.36(c)(1)(Ii)(A)
X 50.73(a)(2)(iv)(A) 0 50.73(a)(2)(x) 065 0 20.2203(a)(2)iii)
El 50.36(c)(2) 0 50.73(a)(2)(v)(A) 0 73.71 (a)(4) o 20.2203(a)(2)(iv) 0 50.46(a)(3)(Ii) 0 50.73(a)(2)(v) (B) 0 73.71 (a)(5) o 202203(a)(2)(v) 0 50.73(a)(2)(I(A) 0 50.73(a)(2)(v)(C)
OTHER below o 20.2203(a)(2)(vi)
X 50.73(a)(2)(l)(B) 0 50.73(a)(2)(v)(D) or inNC (If more space Is required, use additional copies of NRC Fonn 366A) (17)With the unit in Mode 3 (i.e., Hot Shutdown) following the shutdown, the cooldown rate at the reactor vessel bottom head and RR loop A exceeded the limit of 100 degrees F in any one hour period as required by TS Surveillance Requirement 3.4.9.1.b. TS 3.4.9 Limiting Condition for Operation (LCO) Required Action A.2, requires a determination, to be performed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, that the Reactor Coolant system (RCS) is acceptable for operation. An engineering evaluation was performed within the required timeframe, and it concluded that the temperature changes in the lower head region of the reactor vessel and RR loop A did not violate any safety margins of the American Society of Mechanical Engineers (ASME) Code,Section XI or 10 CFR 50, Appendix G (i.e., the RCS was acceptable for continued operation.) In two other instances, RCS heatup rates were exceeded but the conditions were not recognized until April 14, 2005, and therefore, an engineering evaluation was not performed in the required timeframe. The first instance occurred on April 9, 2005 at approximately 0520 hours0.00602 days <br />0.144 hours <br />8.597884e-4 weeks <br />1.9786e-4 months <br /> while the operating crew was warming the RHR discharge piping in preparation for the shutdown cooling mode. Based on review of operating data, the heatup rate was found to be 139 degrees F/hr on RR loop A and 146 degrees F/hr on RR loop B. The second instance occurred on April 9, 2005, at approximately 1020 hours0.0118 days <br />0.283 hours <br />0.00169 weeks <br />3.8811e-4 months <br /> while the operating crew was raising and lowering RHR flow to establish shutdown cooling. A thermal stratification had formed in the lower vessel region after the shutdown. When the thermal layer dispersed, the bottom head drain temperature was found to have increased at a rate of approximately 131 degrees F/hr.
Unit 2 entered Mode 4 (i.e., Cold Shutdown) on April 9, 2005, at 1635 hours0.0189 days <br />0.454 hours <br />0.0027 weeks <br />6.221175e-4 months <br /> and entered Mode 2 (i.e., Startup) on April 11, 2005, at 1207 hours0.014 days <br />0.335 hours <br />0.002 weeks <br />4.592635e-4 months <br /> during the restart sequence. Neither heatup rate exceedence was recognized until operating data was reviewed in the corrective action program on April 14, 2005. Because Unit 2 changed operating modes on April 11, 2005, with this condition unrecognized, an engineering evaluation was not performed in accordance with TS 3.4.9 Required Action A.2, until after Unit 2 had changed operating modes to commence restart and power operation. This is prohibited by TS LCO 3.0.4. which states, "When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall not be made... "
Subsequent engineering evaluations determined that temperature changes in the lower head region of the reactor vessel and RR loops did not violate any safety margins of the ASME Code,Section XI or 10 CFR 50, Appendix G.
EVENT CAUSE
The root cause of the shutdown was that the operating strategy of the Condensate system was to use a pump discharge pressure band for control without establishing a flow limit. The Condensate and Feedwater systems had undergone modifications during the spring 2005 refueling outage that changed flow and pressure operating characteristics. The modifications included a detailed hydraulic analysis and a methodical startup testing process, including inspections, hold points, data analysis, and evaluation of anomalies. At the time the automatic RPS actuation occurred, the Condensate system discharge pressure was being controlled within a pressure band of 150 to 190 psig, as required by a plant operating procedure. This was being accomplished by diverting flow to the condenser via operation of a Condensate return valve. Also, with the 2A RFP removed from service, the associated minimum flow valve was opened, which diverted additional flow to the condenser.
(If more space Is required, use additional copies of (If more space Is required, use additional copies of NRC Form 366A) (17)
PREVIOUS SIMILAR EVENTS
LER 1-2003-001: Reactor Feed Pump Trip Results in Specified Systems Actuation On January 12, 2003, a reactor feedwater pump turbine (RFPT) trip, resulted in the actuation of the RPS, as well as PCIS Group 2 and 6 valve closures. At the time of the event, Unit 1 was initially operating at 94 percent of RTP. The cause of the RPS and subsequent equipment actuations was attributed to insufficient lube oil pressure on the RFPT 1B bearing header during an oil pump trip.
The corrective actions in this LER involve equipment failures, and therefore, could not be expected to prevent the event described in LER 2-2005-002.
COMMITMENTS
There are no commitments in this LER. The actions discussed within this report will be implemented in accordance with corrective action program requirements.