05000271/LER-1997-012-01, :on 970502,determined That RHR SW Flow Rate During LOCA Could Potentially Have Been Below Required Design Flow.Event Under Investigation.Event Rept to Document Event Was Written
| ML20148J316 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 06/06/1997 |
| From: | Maret G VERMONT YANKEE NUCLEAR POWER CORP. |
| To: | |
| Shared Package | |
| ML20148J300 | List: |
| References | |
| LER-97-012-01, LER-97-12-1, NUDOCS 9706170103 | |
| Download: ML20148J316 (3) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(viii) 10 CFR 50.73(a)(2) 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 2711997012R01 - NRC Website | |
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NR Fcru 366 U.S. NUCLEAR REGULATORY COMMISSION APPROYED BY OMB No. 3150-0104 (4 95)
EXPIRES 04/30/98 ESTIMATED BURDEN PEk RESPCfSE TO COMPLY WITH THIS MANDATO:tY INFORMATION COLLECTION REQUEST: 50.0 HRS. REPORTED LESSGMS LEARNED LICENSEE EVENT REPORT (LER)
ARE INCORPC*ATED INTO THE LICENSING PROCESS AND FED BACK TO INDUSTRY. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE j
INFORMATION AND RECORDS MANAGEMENT BRANCH (T-6 F33), U.S. NUCLEAR REGULATORY COMMIS$10N, WASHINGTON, DC 20566-0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE Of MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.
FACILITY NAME (1)
VERMONT YANKEE NUCLEAR POWER STATION DOCET MlaeER ( )
PARI (3) 05000271 01 0F 03 TITLE (4) Residual Heat Removal Service Water Flow Could Be Potentially Less than the Design Basis Flow due to Instrunent inaccuracles EVENT DATE (5) l LER WlaqBER (6)
REPORT DATE (7)
OTHER FACILITIES INVOLVED (8)
McWTH DAY YEAR YEAR SEQUENTIAL REVISION MONTH DAY YEAR
' FACILITY NAME
' DOCKET No.(S)
NUMBER HUMBER 05000 05 02 97 97 012 00 06 02 97 N/A i
OPERATING THIS REPdRT IS SUOMITTED PURSUANT TO THE REQUIREKhTS OF 10 CFR i: CHECK ONE OR MORE (11) i MODE (9)
N i
i 20.2201(b) 20.2203(a)(2)(v) 50.73(a)(2)(i) 50.73(a)(2)(viii) l POWER 0
20.2203(a)(1) 20.2203(a)(3)(1)
X 50.73(a)(2)(li) 50.73(a)(2)(x)
LEVEL (10) 20.2203(a)(2)(1) 20.2203(a)(3)(iI) 50.73(a)(2)(lii) 73.71 20.2203(a)(2)(li) 20.2203(a)(4) 50.73(a)(2)(lv)
OTHER j
20.2203(a)(2)(iii) 50.36(c)(1) 50.73(a)(2)(v)
(Specify in Abstract below or in NRC 20.2203(a)(2)(iv) 50.36(e)(2) 50.73(a)(2)(vil)
Form 366A) j LICENSEE CONTACT FOR THIS LER (12)
NAME TELEPHONE NO. (Include Area Code)
GREGORY A. MARET, L NT MANAGER 802 257-7711 CupFLETE ONE LINE FOR EACH CCBFONENT FAILURE DESCRIBED IN THIS REPORT (13) l
CAUSE
SYSTEM COMPOF NT MANUFACTURER REPORTABLE
CAUSE
SYSTEM COMPONENT MANUFACTURER REPORTABLE TO NPRDS TO NPRDS NA NA i
NA NA SUPPLEMENTAL REPORT EXPECTED (14)
EXPECTED MO DAY YEAR i
SUBMIS$10N X
YES NO DATE (15) 08 29 97 (If yes, cornplete EXPECTED SUBMISSION DATE)
ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)
On 5/2/97 at 1300 hours0.015 days <br />0.361 hours <br />0.00215 weeks <br />4.9465e-4 months <br />, during the AE inspection preparation it was determined that the Residual Heat Removal (RHR) Service Watsr(SW) flow rate during a LOCA could potentially have been below the required design flow. The design flow of the BHRSW heat exchanger and required flow for a LOCA is 2700 Gallons Per Minute (GPM). Instrument accuracy for flow indication is +/- 200 GPM, which could have resulted in an actual RHRSW flow rate as low as 2500 GPM. This condition has been evaluated and it was determined that the RHRSW system can meet its design cooling capacity provided that the plant is only operated if river water temperature (primary cooling medium) is equal to or less than 70 degrees F. Additionally, if flow for l
the RHRSW system was diverted from other loads such that the actual heat exchanger flow was 2900 GPM, with instrument t
inaccuracy in the conservative direction, it was determined that the remaining loads would still have the required amount of I
coo!ing with the river water temperature restriction in use.
The root cause investigation is in progress. A supplemental LER will be submitted once the root cause has been determined.
l Immediate corrective actions included the initiation of an Event Report to document the concern and notify the Nuclear Regulatory Commission (NRC), the initiation of a Basis for Maintaining Operation (BMO) document with a mandatory read and sign form for the Operations on-shift crews, and the establishment of a river water temperature administrative limit of 70 degrees F. Since plant start-up there has been no accident or LOCA conditions that would have required the use of the RHRSW system in the accident mode. Analysis of the event shows that the potential low flow can be augmented using the opposite RHR loop. There was no threat to the health and safety of the public and no safety consequences resulted from this event.
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9706170103 970606 PDR ADOCK 05000271 j
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PDR l
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l NRC For 366 U.S. r/UCLEAR REGULATORY COMMISSION APPROVED BY OM8 NO. 3150 0104 (4 95) -
CXP!RES 04/30/98 ESTIMATED BLEDEN PER RESPONSE TO COMPLY WITH THIS MANDATORY i
INFORMA110N COLLCCTION REQUEST: 50.0 HRS. REPORTED LESSONS LEARNED LICENSEE EVENT REPORT r.LER)
ARE INCORPORATED INTO THE LICENSING PROCESS AND FED BACK TO INDUSTRY. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (T-6 F33), U.S. WUCLEAR REGULATORY COMMISS10N, WASHINGTON, DC 20566-0001, AND TO THE FAPERWORK ltEDUCTION PROJEC1 '3150-0104), OFFICE OF MANAGEMENT AND BUDGT:T, WASHINGTON, M 20503.
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FACILITY EAME (1)
DOCKET NL14BER (2) I LER NUMBER (6)
PAGE (3)
YEAR SEQUENTIAL NUMBER REV #
VERMONT YANKEE NUCLEAR POWER CORPORATION 05000271 97 012 00 02 0F 03 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
DESCRIPTION OF EVENT
On 5/2/97 at 1300 hours0.015 days <br />0.361 hours <br />0.00215 weeks <br />4.9465e-4 months <br />, during an AE inspection preparation, it was determined that the Residual Heat R moval(RHR)(Ells = BO) Service Water (SW)(Ells = BI) flow rate during a LOCA could potentially have been below the required d: sign flow. The design flow of the RHRSW heat exchanger is 2700 Gallons Per minute (GPM), the same flow rate which is r: quired by the LOCA analysis. To prevent exceeding the RHRSW heat exchanger design flow rate, plant procedures direct the operators to limit the RHRSW heat exchanger flow rate to the design limit of 2700 GPM. This limit would be imposed using Evailable indication; however, instrument accuracy for f!ow indication is +/- 200 GPM which could have resulted in an actual RHRSW flow rate as low as 2500 GPM.
This condition has been evaluated and it was determined by engineering analysis that the RHRSW system can meet its design cooling capacity, with the reduced flow of 2500 GPM, provided that plant operation only continue if river water temperature (primary cooling medium) is equal to or less than 70 degrees F. Additionally, if flow for the RHRSW system was diverted from oth:r loads such that the actual flow was 2900 GPM, to account for instrument inaccuracy in the conservative direction, it was d-trrmined that the other loads would still have the required amount of cooling provided the river water temperature restriction was rnaintained. There would be no adverse impact on the heat exchanger as a result of the increased flow.
CAUSE OF EVENT
The root cause of this event is under investigation. A supplemental L%ense Event Report will be submitted once the root cause has been determined.
ANALYSIS OF_ EVENT The RHRSW System provides a dynamic heat sink for the RHR Sy:; tem by supplying sufficient cooling capacity during a design basis accident (DBA) and minimizes the probability of a release of radioactive contaminants to the environs.
Th3 design basis LOCA parameters for the RHRSW system are: a flow of 2700 GPM, initial torus temperature of 90 degrees F, and a maximum river temperature of 85 degrees F. This allows the torus water to be maintained at or below 176 degrees F following a LOCA. The maximum Torus water temperature of 176 degrees F ensures that there is sufficient Net Positive Suction Head (NPSH) for the Emergency Core Cooling System (ECCS) pumps and prevents degradation of Environmentally Qudified equipment. During a LOCA, the Core Spray system would be used for injecting water into the reactor vessel and the one or both RHR loops could be used for Torus cooling.
Instrument accuracies were not initially considered when determining design basis parameters. The flow indication accuracy is
+ /- 200 GPM. Subsequently, if the RHRSW flow, to cool the Torus during an accident, was set at 2700 GPM the potential cxists for the actual flow to be 2500 GPM. This is less flow than assumed in the analysis of record which demonstrates Vermont Yankee's ability to maintain tho Torus water at or below the required temperature with river water temperature at or near 85 degrees F. li this had happened, operators would have been alerted and given appropriate direction by plant Emtrgency Operating Procedures (EOP's) using the parameters available in the Control Room which would have indicated the cdv rse trend in Torus water temperature. Options available to the operators would be to place both RHR loops on Torus cooling, us:ng the Core Spray System to fill and maintain the reactor vessel cooling. Engineering evaluations would have been provided by the Technical Support Center.
Therefore, adequate cooling was available for the Torus and there was no threat to the health and safety of the pubhc.(5-92)
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3 rs 366 U.S. NUCLEAR REGULATORY C0kt115S10N APPROVE OM8 50 0104 ESTIMATED SURDE2 PER RESPONSE TO COMPLY UlTH THIS MANDATCAY INFORMATION COLLECTION REQUEST: 50.0 HRS. REPORTED LESSONS LEARNED LICENSEE EVENT REPORT (LER)
ARE INCORPORATED INTO THE LICENSING PROCESS AND FED BACK TO INDUSTRY. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND RECORDS MANAGEMENT BRANCH (T 6 F33), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20566 0001, AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON, DC 20503.
FACILITY NAME (1)
DOCKET NLSWER (2)
LER IR8eER (6)
PAGE (3)
YEAR SEQUENTIAL NUMBER REV #
VERMONT YANKEE NUCLEAR POWER CORPORATION 05000271 97 012 00 03 0F 03 TEXT (If more space is required, use additional copies of NRC Form 366A) (17)
CORRECTIVE ACTIONS
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immediate Corrective Actions
1.
An Event Report to document the event was written.
2.
The Nuclear Regulatory Commission (NRC) was notified in accordance with 10 CFR 50.72.
3.
A Basis for Maintaining Operation (BMO) document was written.
4.
A mandatory read and sign form requiring the Operations on shift crews to read and understand the BMO was initiated.
5.
An administrative river water temperature limit of 70 degrees F was established.
Lona Term Corrective Actions 1.
The long term corrective action will evaluate the current Service Water flow model at the upper limits for river temperature and flow under the conditions assumed for the RHR System operation. This will be completed by 7/1/97.
ADDITIONAL INFORM ATION During the past five years similar events involving original design specifications have been reported as LER's 93-13, 95 02, 97-02, and 97-06.
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