05000266/LER-1997-025, :on 970520,pressurizer Level Was Controlled Higher than Assumed in Accident Analysis.Caused by Inappropriately Changing Procedures W/O Adequate Consideration.Listed Affected Procedures Will Be Revised

:on 970520,pressurizer Level Was Controlled Higher than Assumed in Accident Analysis.Caused by Inappropriately Changing Procedures W/O Adequate Consideration.Listed Affected Procedures Will Be Revised

ML20148N327
Person / Time
Site:	Point Beach
Issue date:	06/19/1997
From:	Castell C WISCONSIN ELECTRIC POWER CO.
To:
Shared Package
ML20148N293	List: 05000266/LER-1997-025, :on 970520,pressurizer Level Was Controlled Higher than Assumed in Accident Analysis.Caused by Inappropriately Changing Procedures W/O Adequate Consideration.Listed Affected Procedures Will Be Revised
References
LER-97-025, LER-97-25, NUDOCS 9706270045
Download: ML20148N327 (4)

LER-1997-025, on 970520,pressurizer Level Was Controlled Higher than Assumed in Accident Analysis.Caused by Inappropriately Changing Procedures W/O Adequate Consideration.Listed Affected Procedures Will Be Revised

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(ii)
2661997025R00 - NRC Website

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o NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 (4 45)

EXPIRES 04/30/98 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECT!ON REQUEST: 50.0 HRS.

REPORTED LESSONS LEARNED ARE INCORPORATED LICENSEE EVENT REPORT (LER) iNTO THE LICENSING PROCESS AND FED BACK TO INDUSTRY.

FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATION AND (See reverse for required number of RECORDS MANAGEMENT BRANCH (T 6 F 33),

U.S.

digits / characters for each block)

NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555 0001, AND TO THE PAPERWORK REDUCTION PROJECT ECluTY NAME (1)

DOCKET NUMBER (2)

PAGE (3)

Point Beach Nuclear Plant, Unit 1 05000266 1 OF 4 I TITLE (4)

Pressurizer Level Controlled Higher than Assumed in Accident Analysis EVENT DATE (5)

LER NUMBER (6)

REPORT DATE (7)

OTHER FACILITIES INVOLVED (8) l YEAR SEQUENTIAL REVISION FACluTY NAME DOCKET NUMBER MONTH DAY YEAR NUMBER NUMBER MONTH DAY YEAR Unit 2 05000301 l05 201 97 I 97

.. 025..

00 l06 FACILITY NAME DOCKET NUMBER 19 97 05000 I

OPERATING THIS REPORT IS FUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 5: (Check one or more)(11)

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MODE (9)

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20.220iio) 20.2203(.H2Hv) 50.73i.H2Ho 30.73i3n2Hvm)

POWER 20.2203(aH1) 20.2203(aH3Hi>

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50.73(aH2Hei) 50.73(aH2Hx)

LEVEL (10) 000 20.2203(aH2Ho 20.2203iaH3Hio 30 73caH2Hm) 73 7i 20.2203(aH2Hii) 20.2203(aH4) 50.73(aH2Hiv)

OTHER 20.2203(aH2Hm) 60.36(cH1) 50.73(aH2Hv)

Speedy in Absusct 1

below

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20.2203(aH2Hiv) 50.36(cH2)

50. 73(aH 2 Hvn) or in NRC Form 366A

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LICENSEONTACT FOR THIS LER (12) ' '

lMAME TELEPHONE NUMBER (include Aree Code)

I Curtis A.

Castell (414) 221-2019 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE

SYSTEM COMPON5NT MANUFACTUNER REPORTABLE

CAUSE

SY3TE COMPONENT MANUF ACT URER REPORTABLE TO NPPDS M

TO NPRDS SUPPLEMENTAL REPORT EXPECTED (14) l PECTED MONTH I

(YES SUBMISSION if yes, complete EXPECTED SUBMISSION DATE).

X NO CATE (15)

ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 sergle-spaced typewritten lines) 1161 On May 20, 1997, while Point Beach Nuclear Plant (PBNP) Unit 1 was in a cold shutdown condition and Unit 2 was in a defueled condition, it was discovered that three operating Procedures (0P) allow manual control of pressurizer level 10% higher than assumed in the PBNP FSAR section 14.2.5 rupture of steam pipe analysis assumption of 20%.

This condition was discovered by Reactor Engineering personnel while reviewing operations and reactor engineering procedures in anticipation of the impending restart of Unit 2.

These procedures were changed in the mid-1980's t0 allow manual control of pressurizer level at 30% in lieu of the automatic program level of 20% at zero power operation.

This condition was caused by inappropriately changing the procedure without adequate consideration of potential affects on the PBNP accident analyses. The affected procedures will be revised prior to restart of Unit 2.

It was determined that the consequences of the PBNP FSAR section 14.2.5 " Rupture of a Steam Pipe" analysis would not exceed 10 CFR 100 limits, even if some fuel damage could occur based on the use of a higher pressurizer level.

NRC FOHM 366 (4-95) 9706270045 970619 PDR ADOCK 050002 6 S

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1 NRC FORM 386A U.S. NUCLEAR REGULATORY COMMISSION (4 95)

LICENSEE EVENT REPORT (LER) i TEXT CONTINUATION l

i FACILITY NAME (U DOCKET NUMSER (2)

LER NUMBER (6)

PAGE (3) l YEAR SEQUENTIAL REYiSION Point Beach Nuclear Plant, Unit 1 05000266 NUMBER NUMBER 2 QF 4 97 025 00 TEXT tit more space os requsred, use additoonalcopes of tJRC form 366Al (11)

Event Description

l l

On May 20, 1997, while Point Beach Nuclear Plant (PBNP) Unit 1 was in a l

cold shutdown condition and Unit 2 was in a defueled condition, it was discovered that Operating Procedures OP-1A, " Cold Shutdown to Low Power l

Operation," OP-1C, " Low Power to Normal Power Operation," and OP-3A,

" Normal Power.to Low Power Operation" allow manual control of pressurizer level 10% higher than assumed in the PBNP FSAR section 14.2.5 rupture of j

steam pipe analysis assumption of 20%.

This condition was discovered by Reactor Engineering personnel whila reviewing operations and reactor engineering procedures in anticipation of the impending restart of Unit 2.

These procedures were changed in the mid-1980's to allow manual control of pressurizer level at 30% in lieu of the automatic program level of 20% at zero power operation.

This change was implemented to provide more operational margin between the chemical and volume control system letdown l

isolation level setpoint at 12% and the actual level being maintained in l

the pressurizer.

Cause

This condition was caused by inappropriately changing the procedure I

without adequate consideration of potential affects on the PBNP accident analyses.

l

Corrective Actions

l The affected procedures, OP-1A, " Cold Shutdown to Low Power Opera tion,"

l l

OP-lC, " Low Power to Normal Power Opera tion," and OP-3A, " Normal Power to l

. Low Power Operationi " will be revised to discontinue the practice of manually controlling pressurizer level at 30% prior to restart of Unit 2.

Recent improvements to the procedure change and 10 CFR 50.59 review processes used at PBNP are expected to minimize the possibility of this j

condition recurring.

I A root cause evalua tion is being completed.

Additional corrective actions

l will be taken, as appropria te, from recommendations containe( in the root l

cause evalua tion.

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l U.S. NUCLEAR REGULATORY COMMISSION i4-ssi LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILIT f NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL FtEVISION Point Beach Nuclear Plant, Unit 1 05000266 NUMBER NUMBER 3OF4 97 025 00 TEXT lit more space a requked, use additoonalcopes of NRC Fonn 366M (11)

R3 portability:

This Licensee Event Report is being submitted in accordance with the requirements of 10 CFR 50.73 (a) (2) (ii) ( A), "Any event or condition that resulted in the nuclear power plant being in an unanalyzed condition that significantly compromised plant safety."

Component and System Description:

l Pressurizer level is automatically controlled using a linear program from 20% at the zero power average temperature condition of 547 F to 45.8% at the full power average temperature condition of 570 F.

Below 547 F level is limited to 20% and above 570 F level is limited to 45.8%.

Level deviations from the program setpoint causes increasing or decreasing charging pump flow based on actual pressurizer level below or above the program level.

Safety Assessment

l The PBNP FSAR section 14.2.5 " Rupture of a Steam Pipe" analysis assumes a 30% pressurizer level, which is based on 20% program level plus 10%

allowance for conservatism.

By changing the program level to 30% and continuing to assume the 10% allowance, the pressurizer level could be assumed to be 40%.

If 40% was used in the analysis, the consequences of the steam line rupture may increase.

In particular. higher pressurizer level could delay depressurization which reduces th+ safety injection flow and delays the onset of accumulator injection.

Arm;.yses to quantify the i

l effect of the higher pressurizer level have not been performed.

It was determined that the consequences of the PBNP FSAR section 14.2.5,

" Rupture of a Steam Pipe" analysis would net exceed 10 CFR 100 limits, even if some fuel damage could occur based on the use of a higher pressurizer level.

This is based on the judgment that the consequences of the large break loss of coolant accident (PBNP FSAR section 14.3.5), which i

is based on the release of all volatile fission products into the

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containment and subsequent release of 0.4 weight %/ day for the first 1

24 hou:s and 0.2 weight %/ day for the next 29 days, would be greater.

The results of the large break loss of coolant analysis show compliance with l-10 CFR 100.

Therefore, it is not expected that operation with increased NRC FOHM 366A (4-95)

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.U.S. NUCLEAR REGULATORY COMMISSION (4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACluTY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL REVISION Point Beach Nuclear Plant, Unit 1 05000266 NUMBER NUMBER 4OF4 97 025 00 1

TEXT tti more space is requwed, use adatterwl copies of JVRC form 366Al (11) pressurizer level during the shutdown condition would cause consequences of the PBNP FSAR section 14.2.5 " Rupture of a Steam Pipe" analysis to exceed 10 CFR 100 limits, even if some fuel damage could occur based on the higher pressurizer level.

System and Component Identifiers The Energy Industry Identification System component function identifier for each component / system referred to in this report are as follows:

Component / System Identifier Pressurizer PZR Charging pump P

Level Controller LC Main Steam System SB

)

Reactor Coolant System AB Similar Occurrences:

A search was conducted of previously submitted licensee event reports similar to this situation for PBNP.

The specific criterion used was based on a search for licensee event reports that were submitted due to plant procedures that allowed or caused the plant to be not in accordance with accident analysis assumptions.

LER 266/301-84-005-00 identified a condition that pilowed operation of the units such that the conditions of FSAR section 14p.l.1, " Uncontrolled RCCA Withdrawal from a Subcritical Condition," analysi's could be invalidated.

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NRC FORM 36bA (4-9M I