05000266/LER-1997-001, :on 970108,safety Injection Delay Times Exceeded Design Basis Values.Caused by Degraded Voltage Conditions.Licensee Engineers Will Prepare FSAR Change Requests to Reflect LBLOCA Evaluation

:on 970108,safety Injection Delay Times Exceeded Design Basis Values.Caused by Degraded Voltage Conditions.Licensee Engineers Will Prepare FSAR Change Requests to Reflect LBLOCA Evaluation

ML20134H539
Person / Time
Site:	Point Beach
Issue date:	02/04/1997
From:	Dante Johnson, Kos J WISCONSIN ELECTRIC POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-97-001, LER-97-1, PBL-97-0036, PBL-97-36, NUDOCS 9702110350
Download: ML20134H539 (6)

LER-1997-001, on 970108,safety Injection Delay Times Exceeded Design Basis Values.Caused by Degraded Voltage Conditions.Licensee Engineers Will Prepare FSAR Change Requests to Reflect LBLOCA Evaluation

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition
2661997001R00 - NRC Website

text

1 Wisconsin Electnc POWER COMPANY Point Beach Nuclear Plant (414) 755-2321 6610 Nuclear Rd., Two Rivers. WI 54241 PBL 97-0036 February 4,1997 Document Control Desk US NUCLEAR REGULATORY COMMISSION Mail Station Pl-137 Washington, DC 20555 Ladies / Gentlemen:

DOCKET 50-266 AND 50-301 LICENSEE EVENT REPORT 97-001-00 SAFETY INJECTION DELAY TIMES IIXCEED DESIGN B ASIS VALUES POINT BEACH NUCLEAR PLANT. UNITS 1 AND 2 Enclosed is Licensee Event Report 97-001-00 for Point Peach Nuclear Plant, Units I and 2. This repon is j

provided in accordance with 10 CFR 50.73(a)(2)(ii)(B),"a condition that was outside the design basis of

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the plant." This report describes a condition where delay times assumed for the high and low head safety i

injection flow in the Large Break Loss of Coolant Accident analysis were not conservative.

If you require additional information, please contact us.

Sincerely,

/hrma WY wd.

b ouglas F. Johnson D

Manager-Regulatory Sersices & Licensing JAK f[

Enclosure

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cc:

NRC Resident Inspector NRC Regional Administrator, Region 111 9702110350 970204 DR ADOCK 05000266

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. Nic FORM 366 U.S. NUCLEAR REOULATORY COMMISSION APPROVED BY OMB NO. 3150 0104 (4 95)

EXPtRES 04/30/98 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THt$ INFORMATION COLLECTION REQUEST: 50.0 HRS.

LICENSEE EVENT REPORT (LER)

REPORTED LESSONS LEARNED ARE INCORPORATED INTO THE LICENSING PROCESS AND FED BACK TO INDUSTRY.

FORWARD COMMENTS REGARDING BURDEN ESTIMATE (See reverse for required nurnbor of TO THE INFORMATION AND RECORDS MANAGEMENT digits / characters for each block)

BRANCH (T-6 F33L U.S.

NUCLEAR REGULATORY COMMIS$10N, WASHINGTON, DC 20555-0001, AND TO THE PAPERWORK RFDUCTION PROJECT FActLITY NAME til DOCKET NUMBER (2)

P AGE 13)

Point Beach Nuclear Plant, Unit 1 05000266 1 OF 5 TITLE 14)

Safety Injection Delay Times Exceed Design Basis Values EVENT DATE (5) l LER NUMBER (El REPORT D ATE 17)

OTHER FACILITIES INVOLVED 18) s SEQUENTIAL REVISION F ACILITY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER NUMDER MONTH DAY YEAR PBNP Unit 2 05000301 l

08 l 97 l 97 l

FACILITY NAME DOCKET NUMBER

'01 001 00 02 04 97 05000 l THIS REPORT IS SUBMITTED PUSISUANT TO THE REQUIREMENTS OF 10 CFR 8: { Check one or morel (11)

OPERATING MODE 19)

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20.2203(all2Hiv) 50.36tcH2) 50.73(aH2Hvil) er in NRC Form 388A LICENSEE CONTACT FOR THis LER (12)

NAME TELEPHONE NUMBER linclude Area Codel Jeff Kos, Design Basis Engineer (414) 221-4917 COMPLETE ONE LINE FOR E ACH COMPONENT FAILURE DESCRIBED IN TH18 REPORT (13)

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CAUSE

SYSTEM COMPr"^T l MANUFACRJHER REPORTABLE

CAUSE

SYSTEM COMPONENT MANUFACTURER REPORTABLE TO NPROS TO NPROS i

SUPPLEMENTAL REPORT EXPECTED 114)

EXPECTED MONTH DAY YEAR YES SUSMISSION (If yes, complete EXPECTED SUBMIS$lON DATE).

X NO DATE (15)

ABSTRACT lumrt to 1400 epaces. Lo., opprossmately 15 emg6e-spaced typewrreten hnes) (15)

On January 8,1997, with Unit 1 operating at 90% power and Unit 2 in a refueling shutdown, licensee engineers determined that me delay times assumed for high and low head safety injection (SI) flow in the Large Break Loss of Coolant Accident (LBLOCA) analysis were not conservative. The LBLOCA licensing basis analysis assumed. hat the high and low head SI systems were capable of providing full flow within five and ten seconds respecively. A conservative licensee evaluation concluded that the total delay times may be as high as 8.0 seconds for high head SI and 23.7 seconds for low head SI. The delay time assumptions for the licensing basis analysis did not account for time delays associated with SI signal processing, sequencer delay time uncertainty, or an increased time for pump acceleration to full speed due to degraded voltage conditions. A Westinghouse safety assessment concludes that the increased safety injxtion delay times do not result in exceeding any design or regulatory limit for Point Beach Units 1 and 2.

NRC FORM 366 (495)

NRC FORM 306A U.S. NUCLEGR REGULATORY COMMISSION 84-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (U DOCKET NUMBER 12)

LER NUMBER 16)

PAGE (3)

YEAR SEQUENTIAL REVISION

)

Point Beach Nuclear Plant, Unit 1 05000266 NUMBER NUMBER 2 OF 5 i

97 001 00 TEXY tif more space is required. use additional copoes ofIVRC Form 366A) (17}

Event Description

On January 8,1997, with Unit 1 operating at 90% power and Unit 2 in a refueling shutdown, licensee engineers determined that the delay times assumed for the high and low head safety injection (SI) flow in the existing LBLOCA analysis were not conservative. This condition was discovered during a review of the j

stroke time performance requirements for the SI system valves SI-852A,B The existing LBLOCA analysis assumes that the high and low head SI systems are capable of providing full flow within five and ten seconds respectively. These assumptions are based on a 5 second delay for the high head SI pump to come up to speed and a 10 second delay for the low head SI pumps to load on the sequencer and come up to speed. This delay time represents the time from when the SI setpoint is reached to the time when the pumps are capable of providing full flow. The delay time assumptions for the existing licensing basis analysis do not account for (1) time delays associated with SI signal processing, (2)

{ sequencer delay time uncertainty, or (3) an increased time for pump acceleration to full speed due to degraded voltage conditions. When these delays were combined, the licensee evaluation concluded that the total delay times may be as high as 8.0 seconds for high head SI and 23.7 seconds for low head SI. The impact of the increased delay times on the LBLOCA analysis is described below.

The applicable acceptance criteria for the LBLOCA analysis is a peak cladding temperature (PCT) of 2200*F, as identified in 10 CFR 50.46. The most recent submittal (which does not include the results of the additional time delay) to the NRC on ECCS Evaluation Model changes reflects a 109'F penalty in margin allocutions to the licensing basis analysis LBLOCA PCT of 2028 F, resulting in a PCT cf 2137 F.

Westinghouse formally esaluated the increase in the safety injection delay times on the LBLOCA analysis.

The revised cumulative LBLOCA PCT is 2181*F, based on a 44*F penalty due to the longer SI delay times. This evaluation concludes that the calculated maximum fuel element cladding temperature does not exceed 2200 F, the calculated total local oxidation of the cladding nowhere exceeds 0.17 times the total cladding thickness before oxidation, and the calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam does not exceed 0.01 times the hypothetical amount that would be generated if all the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react.

Conformance to the criteria above demonstrates that the core geometry is maintained such that the core remains amenable to cooling. The increase in SI system delay will not impact the long-term ability to maintain cora temperature at an acceptably low value and to remove decay 1. eat for an extended period of time required by the long-lived radioactivity remaining in the core. Therefore, the specific safety limits defined by the ECCS Acceptance Criteria of 10 CFR 50.46 for the LBLOCA licensing basis analysis j

including penalties are met.

NQC FORM 366A 14-95)

NRC FOR;.1368A U.S. NUCLEAM REOULATORY COMMISSION 84 95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FActLITY NAME (1)

DOCKET NUMBER f 2)

LER NUMBER IS)

PAOE (3)

Point Beach Nuclear Plant, Unit 1 05000266 NU ER NU R

3 OF 5 97 001 00 iExi ut more sosce is reewroo. use soonsonal copues or tvRc Form 366A) (1D The IEEE Standard 803A-1983 component identifier for this report is:

Pump (p)

Component and System Description:

The SI system shall deliver borated cooling water to the reactor coolant system during the injection phase of SI to support core cooling and to ensure adequate shutdown margin in the event of a main steam line break.

The licensing basis accident analyses, as described in Chapter 14 of the FSAR, ensure that the SI system is capable of performing these safety-related functions. The LBLOCA analysis (FSAR Section 14.3.2),

SBLOCA (FSAR Section 14.3.1), MSLB (FSAR Section 14.2.5), SGTR (FSAR Section 14.2.4) and the Containment Integrity Evaluation (FSAR Section 14.3.4) assume SI actuation resulting in safety injection flow to the reactor coolant system.

PBNP Technical Specification 15.4.6.A.2 requires that a test be performed to demonstrate the ability of a diesel generator to automatically start, shed load, and restore particular vital equipment to operation following an actualinterruption of normal AC station service power supply to associated engineered safety systems busses together with a simulated safety injection signal. The Technical Specifications require that the test be conducted to assure that the diesel generator will start and assume required load in accordance with the timing sequence listed in FSAR Section 8.2 after the initial starting signal. These acceptance criteria are also listed in Operations Refueling Test (ORT) 3, Appendix C, Attachment 1. The acceptance criteria from this test are used in evaluating the affects on the LBLOCA analysis.

Ccuse:

The existing LBLOCA analysis provides inadequate margin in the SI system delay time assumption to accommodate for time delays associated with SI signal processing, sequencer delay time uncertainty, and an increased time for pump acceleration to full speed due to degraded voltage conditions. The basis for the delay times are a result of typical vendor methodology which may result in a non-conservative LBLOCA analysis with respect to SI system actuation ('elay times. Also, the acceptance cri;eria for the SI sequence test of ORT 3 as described in FSAR sectian 8.2 and ORT 3 Appendix C Attachment 1 should have been thoroughly reviewed to assure tat the times remain within the LBLOCA analysis assumptions.

e N'.C FOLM 36;A U.S. NUCLEAR f.EGULATOAY COMMISSION 14 951 LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME til DOCKET NUMBER (2)

LER NUMBER 16)

PAGEtt j

YEAR SEQUENTIAL REVISION Point Beach Nuclear Plant, Unit 1 05000266 NUMBER NUMBER 4 0F 5 97 001 00 TEXT IIf more space is required. use additional copies of IVRC Form 366A) (17)

Safety Assessment

The specific safety limits defined by the ECCS Acceptance Criteria of 10 CFR 50.46 are met. Therefore, the health and safety of the public is not compromised by the conditions described herem.

1 I

Corrective Actions

The following corrective actions have been taken or are planned to address this event.

1.

Westinghouse has formally evaluated the increase in the safety injection delay times on the LBLOCA analysis. This evaluation determined that the specific safety limits defined by the ECCS Acceptance Criteria of 10 CFR 50.46 impacted by an increase in safety injection delay time for the LBLOCA analysis are met.

2.

Licensee engineers will prepare FSAR change requests to reflect this LBLOCA evaluation.

3.

Licensee engineers will review the SBLOCA (FSAR Section 14.3.1), MSLB (FSAR Section 14.2.5),

SGTR (FSAR Section 14.2.4) and the Containment Integrity Evaluation (FSAR Section 14.3.4) accident analyses. These analyses assume SI actuation resulting in safety injection flow to the reactor coolant system. This review will verify the assumptions associated with SI system delay times are appropriately conservative.

4.

The acceptance criteria for SI sequence test of ORT 3 as described in FSAR sectios 8.2 and ORT 3, Appendix C, Attachment 1, will be reviewed to assure that the delay times associtted with sequenced safeguards components remain within the FSAR Chapter 14 accident ar.alysis assumptions.

Reportability

A 1-hour prompt notification per 10 CFR 50.72(a)(2)(ii)(B) was reported to the NRC duty officer at 1344 CST on January 8,1997. This Licensee Event Report is being submitted in accordance with the requirements of 10 CFR 50.73(a)(2)(ii)(B), "A condition that was outside the design basis of the plant."

NaC FORM 366A 14-95)

i.

e NAC FO;.M 3144 U.S. NUCLEAR REGULATORY COMMISSION 44-95)

+

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER 12)

LER NUMBER (6)

PAGE(3)

YEAR SEQUENTIAL REVISION Point Beach Nuclear Plant, Unit 1 05000266 NUMBER NUMBER 5 OF 5 97 001 00 TEXT (11 more space is requend, use additenal copes of NRC form 366Al (17)

Similar Occurrences:

A search of the LER database discovered a prior occurrence of non-conservative accident analysis assumptions resulting in a system being declared in a condition that is outside the design basis.

LER Iitle 266/96-015-00 Main Steam Safety Valve Lift Setpoints Exceed Design Basis Values i

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