05000275/LER-2014-001, Regarding Technical Specification 3.4.3, Reactor Coolant System Pressure Limit Violation During Vacuum Refill Due to Human Error
| ML14090A450 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 03/31/2014 |
| From: | Allen B Pacific Gas & Electric Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| DCL-14-026 LER 14-001-00 | |
| Download: ML14090A450 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) 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)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 2752014001R00 - NRC Website | |
text
Pacific Gas and Electric Company March 31, 2014 PG&E Letter DCL-14-026 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Power Plant, Units 1 and 2 Barry S. Allen Site Vice President 10 CFR 50.73 Diablo Canyon Power Plant Mail Code 104/6 P. 0. Box 56 Avila Beach, CA 93424 805.545.4888 Internal: 691.4888 Fax: 805.545.6445 Licensee Event Report 1-2014-001-00, Technical Specification 3.4.3, RCS Pressure Limit Violation During Vacuum Refill Due to Human Error Dear Commissioners and Staff; Pacific Gas and Electric Company (PG&E) submits the eQclosed Licensee Event Report in accordance with 10 CFR 50.73(a)(2)(i)(B).
PG&E makes no new or revised regulatory commitments (as defined by NEI 99-04) in this report. All the corrective actions identified in this letter will be implemented in accordance with the Diablo Canyon Power Plant Corrective Action Program.
This event did not adversely affect the health and safety of the public.
Sincerely,
/J~ 5. /Itt-Barry S. Allen lmp/3386/50609672 Enclosure cc/enc:
Peter J. Bamford, NRC Project Manager Marc L. Dapas, NRC Region IV Administrator Thomas R. Hipschman, NRC Senior Resident Inspector IN PO Diablo Distribution A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway
- Comanche Peak
- Diablo Canyon
- Palo Verde
- Wolf Creek
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 01/31/2017 (01-2014)
Estirrated l:l.Jr"001 per re5j:X)IlSe to oorrply wth this rrerdatay oolerlion re:juest: 80 hoors. Repcrted lessons leamed are inoorpJrated into the lia:nsirYJ process and fed back to irx:lustry. Serd cx:mrents regardirYJ burden estirrate to the FaA, PriVCK¥ and lnforrration Cdlerlions Branch LICENSEE EVENT REPORT (LER)
(T-5 F53), U.S. Nl.dear Regulatay Coorrissioo, Wlslit'YJ!on, OC 20555-0001, or Oj internet e-rrail to lnfoa:lle:ts.Pesoora@1rc.gov, and to the D:lsk CXfirer, CXfice of lnforrration and Regulatay (See Page 2 for required number of Affairs, NEOB-10202, (3150-0104), CXficeoffvlcrngerrentand Budget, \\1\\Bshit'YJ!on, OC 20503. If a rreans used to ifYIX)Se an infooretion oolerlioo dces net display a a.urently valid aviE digits/characters for each block) oontrd nurrber, the NRC rray net cord.d cr srmscr. and a person is net required to respcn:l to, the infcrrration oolerlion.
- 3. PAGE Diablo Canyon Power Plant, Unit 1 05000 275 1 OF 4
- 4. TITLE Technical Specification 3.4.3, Reactor Coolant System Pressure Limit Violation During Vacuum Refill Due to Human Error
- 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 NUMBER NO.
MONTH DAY YEAR Diablo Canyon Unit 2 05000 323 01 28 2014 2014 - 001 - 00 03 31 2014 FACILITY NAME DOCKET NUMBER 05000
- 9. OPERATING MODE
- 11. THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR §: (Check all that apply)
D 2o.2201(b) 0 20.2203(a)(3)(i)
[] 50.73(a)(2)(i)(C) n 50. 73(a)(2)(vii) 1 D 20.2201 (d) n 20.2203(a)(3)(ii) n_ 50.73(a)(2)(ii)(A) 0 50.73(a)(2)(viii)(A)
D 20.2203(a)(1)
D 20.2203(a)(4) 0 50. 73(a)(2)(ii)(B) 0 50.73(a)(2)(viii)(B) n 20.2203(a)(2)(i) n 50.36(c)(1)(i)(A) n 50.73(a)(2)(iii) 0 50. 73(a)(2)(ix)(A)
- 10. POWER LEVEL 0 20.2203(a)(2)(ii) n 50.36(c)(1)(ii)(A) n 50. 73(a)(2)(iv)(A) n 50.73(a)(2)(x) 100 D 20.2203(a)(2)(iii)
D 50.36(c)(2) 0 50.73(a)(2)(v)(A) 0 73.71(a)(4) n 20.2203(a)(2)(iv) 0 50.46(a)(3)(ii) n 50. 73(a)(2)(v)(B) n 73.71 (a)(5) 0 20.2203(a)(2)(v)
D 50.73(a)(2)(i)(A) 0 50. 73(a)(2)(v)(C)
DOTHER 0 20.2203(a)(2)(vi)
~
50.73(a)(2)(i)(B) 0 50.73(a)(2)(v)(D)
Specify in Abstract below or in I.
Plant Conditions
At the time of discovery, both units were in Mode 1 (Power Operation) at 100 percent power.
II.
Description of Event
A. Background Technical Specification (TS) 3.4.3, "RCS Pressure and Temperature (PIT) Limits," requires the reactor coolant system (RCS) [AB] pressure, RCS temperature, and RCS heatup and cooldown rates shall be maintained within the limits specified in the pressure temperature limits report (PTLR) at all times.
All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. This limiting condition of operation (LCO) limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.
The PTLR contains PIT limit curves for heatup, cooldown, inservice leak and hydrostatic testing, and data for the maximum rate of change of reactor coolant temperature. Each PIT limit curve defines an acceptable region for normal operation. The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.
The LCO references the PTLR, which establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB). The vessel is the component most subject to brittle failure, and the PTLR limits therefore assure low temperature overpressure protection of the RCS.
The PIT limits are not derived from design basis accident (DBA) analyses. They are prescribed during normal operation to avoid encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, an unanalyzed condition. Administrative Controls Section 5.6.6, identifies the NRC reviewed and approved methodology for determining the PIT limits. Although the PIT limits are not derived from any DBA, the PIT limits are acceptance limits since they preclude operation in an unanalyzed condition. RCS PIT limits satisfy Criterion 2 of 10 CFR 50.36 (c)(2)(ii).
2014 - 001 - 00 Returning from a refueling outage involves refilling and venting noncondensable gas from the RCS to achieve loops filled, which is a condition where natural circulation can be used as a backup means of decay heat removal if forced circulation via residual heat removal (RHR) [BP]
is lost. Prior to 2001, refilling the RCS was an iterative, time and dose-intensive process to ensure each loop, including steam generator U-tubes are filled and capable of natural circulation.
B: Event Description In May of 2001, Pacific Gas and Electric (PG&E) implemented Design Change DCP N-049532 for Unit 1 and DCP N-050532 for Unit 2, "Reactor Coolant System Vacuum Refill (RCSVR)." The design change was implemented via Diablo Canyon Power Plant (DCPP) Operating Procedure OP A-2:1X, "Reactor Vessel-Vacuum Refill of the RCS," starting with Unit 2 tenth refueling outage in May 2001 and followed by Unit 1 eleventh refueling outage in May 2002. This procedure significantly reduced the time needed to reach "loops filled" and capable of natural circulation. It also reduced dose to workers, and improved initial primary water chemistry.
During the last three years, RCS pressure was reduced below 0 PSIG during the vacuum refill process in DCPP Units 1 and 2. Specifically, vacuum refill was performed during the Spring Unit 2 sixteenth refueling outage in 2011, the Spring Unit 1 seventeenth refueling outage in 2012, and Spring Unit 2 seventeenth refueling outage in 2013.
C. Status of Inoperable Structures, Systems, or Components that Contributed to the Event N/A
D. Other Systems or Secondary Functions Affected
N/A
E. Method of Discovery
In accordance with Procedure XI1.DC2, "Regulatory Operating Experience," PG&E Regulatory Services personnel screen various industry regulatory operating experience (ROE) sources. On January 28, 2014, PG&E recognized the generic implications of a unique violation in NRC Inspection Report 05000440/2013007 dated January 3, 2014, and created SAPN 50606432, "ROE: Review NRC Finding (Perry) PTLR."
F. Operator Actions
N/A G. Safety System Responses N/A Ill.
Cause of the Event
A. Apparent Cause The apparent cause of not maintaining RCS pressure within the limits described in the PTLR was a legacy design error. PG&E interpreted the PTLR to only apply below the upper pressure limits, and therefore did not consider it in conflict with the adoption of the vacuum refill analysis and procedure.
B. Contributory Cause Vacuum refill was a vendor-supported process being implemented throughout the industry without prior NRC approval or expanded PIT curves down to 0 PSIA.
IV.
Assessment of Safety Consequences
The design change implementing RCSVR evaluated physical effects of the sub-atmospheric RCS pressure and determined there were no adverse impact on the reactor vessel and RCPB structural integrity.
OP A-2: IX ensures sufficient RCS pressure to assure adequate net positive suction head for the RHR system, over an operating band of temperatures, pressures and RHR flows.
Therefore, the explicit PTLR compliance error during vacuum refill did not adversely affect the health and safety of the public or station personnel.
V.
Corrective Actions
On February 26, 2014, the PTLR for DCPP, "PTLR-1,"Revision 14, became effective, which includes figures showing acceptable operation down to 0 PSIA.
VI.
Additional Information
There were no prior similar events at DCPP.