ML17331A639

Safety Evaluation Supporting Amends 43 & 25 to Licenses DPR-58 & DPR-74,respectively

ML17331A639
Person / Time
Site:	Cook
Issue date:	01/28/1981
From:	Office of Nuclear Reactor Regulation
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Shared Package
ML17326A839	List: ML17326A838 ML17331A638 ML17331A639
References
NUDOCS 8102190172
Download: ML17331A639 (14)

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATIOH RELATED TO AMENDMEHT NO.

43 TO FACILITY OPERATING LICENSE t(0.

DPR-58 AND AMENDMENT t<0. 25 TO FACILITY OPERATING LICENSE NO. DPR-74 INDIANA AHD HICHIGAH ELECTRIC COMPANY DONALD C.

COOK NUCLEAR PLANT UtIIT NOS.

1 AHD 2 DOCKET NOS.

50-315 AND 50-316 I.

Introduction By letter dated February 22, 1980 the licensee requested Technical Specifi-'ation changes to the setpoint for steam generators low-'1ow levels settings for D.

C.

Cook Plant Units Nos.

1 and 2.

These new steam generator low-low level setpoi nts were provided in the licensee's letter of November 5, 1979 in response to IE Bulletin 79-21 concerning effects of containment tempera-ture on safety-related level monitoring systems.

Two level. systems inside containment impacted by temperature are-pressurizer level and steam genera-tor level.

This submittal was assighed to EGSG Idaho, Inc. (our consultants) for review and evaluation under our technical assistance program.

II.

Discussion Enclosure 1, Technical Evaluation Report, "The Effect of Containment TempeIature of Liquid Level Measurements for D.

C.

Cook Units 1

and 2" was prepared for us by EG8G Idaho.

EG&G concluded that the steam generator low-low setpoints for 0.

C.

Cook Plant Unit Nos.

1 and 2 listed below are acceptable.

Although no credit has been taken for the pressurizer level instruments in the safety analysis, the range of indicated pressurizer level already established is acceptable and will assure that the heaters remain covered and that the pressurizer does not go water solid.

This acceptable range was established taking into consideration temperature effects on level instruments.

III. 'nstrument Tri Set pints

-Steam Generator low-low level Unit,l Unit 2 Tri Value

> 17K

> 23%

Allowable Values

> 16K

> 20'

0 5 ~

IV.

Evaluation Based on our review of the consultant's technical evaluation, as depcribed

above, we have concluded that the level instruments inside containment have been properly evaluated for temperature effects.

The higher setpoint values for steam generator low-low for D.

C.

Cook Unit Nos. land 2 are therefore acceptable.

These values were discussed with the licensee and he agrees to them.

V.

Environmental Consideration Me have determined that the amendments do not authorize a change in effl.uent types of total amounts nor an increase in power level and wi,ll not result in any significant environmental impact.

Having made this determination, we have further concluded that the amendments involve an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 551.5(d)(4), that an environmental'impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with th'e 'issuance of'.these. amendments.-

Conclusion Me have concluded, based on the considerations discussed above, that:

(1) because the amendments do not involve a significant increase in the proba)lity or consequences of accidents previously considered and do not invplve a significant decrease in a safety margin, the amendments do, not involve a significant hazards consideration, (2) there is reasonable assurance that the health and.safety of the public will not be endangered by operation in the proposed

manner, and (3) such activities will be con-ducted in compliance with the Commission's regulations and the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the pdblic.

Date:

January 28, 1981

-3" REFERENCES 2.

3.

4 NRC Letter (K. Kniel) April 11, 1977 to Indiana

& Michigan Power Company.

NRC (E. Case)..

Indiana

& Michigan Power Company Letter (J. Tillin hast)

Ma 17 1977 as ay to NRC Letter (K. Kniel) July 8, 197? to Indiana

& Michigan Power Company.

Case).

westinghouse Electric Corp. Letter (T. Anderson)

June 21 1978 t NRC (E.

e, o

NRC (E.

Case).

5.

Indiana & Michigan Power Company Letter (J. Tillinghast) J

,22 1978 s

une, to to NRC (H. Qenton).

Indiana

& Michigan Electric Company Letter (J.

Dolan) February 22 1980

1671F Enclosure 1

TECHNICAL EVALUATION REPORT THE EFFECT OF CONTAIQZNT TBfPERATURE ON LIQUID LEVEL MEASUREMENTS DONALD C.

COOK NUCLEAR STATION, UNIT NOS ~

1 AND 2 Docket iNos. 50-315 and 50-316 August 19SO A. C. Udy Reliability and Statistics Branch'ngineering Analysis Division EG6G Idaho, Inc.

TAC Nos.

129S1 and 12982

ABSTRACT As an indirect resul t of an assumed loss-of-coolant accident, there is an apparent increase in the indicated water level for those systems whose sensors and rererence legs'are exposed to the elevated containment temper-ature.

This report examines the effected

systems, the effect on the initi" ation of safety systems, and the effect on 'the information displayed to the reacto r operator.-

4 CONTENTS ABSTRACT

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1. 0 INTRODUCZION 2.0 EVALUATION OF THE COOK NUCLEAR STATION) UNITS 1

AND 2

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2. 1 Review Guidelines 2.2 Description of Liquid Level Measurement Systems 2.3 Evaluation of Liquid Level Measurement Systems

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3.0

SUMMARY

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TECHNICAL EVALUATION REPO THE EFFECT OF CONTAIRKNT TEMPERATURE ON LIQUID LEVEL MEASUREMENTS DONALD C.

COOK NUCLEAR STATION, UNIT NOS+

1 AND 2 1.0 INTRODUCZION Based on the information supplied by Indiana

& Michigan Electric Com-pany (I&MECo), this report addresses the effect of the containment temper-ature on the steam generator and pressurizer water level detectors.

In June 1979, the Power Systems Division of the Westinghouse Electric Corporation (W) notified the Nuclear Regulatory Commission (NRC)1 and W

utility customers of corrections that should be applied to indicated steam generator water level and associated low water level protec" ion system setpoints and emergency operating procedures.

'Ehe problem identified was

that, as the temperature of the level measurement reference leg increased due to a highmnergy line break, the water column density deqrgased.

".his appears as an apparent increase in the ind'icated ~ater level, which could result in delayed., protection (reactor trip and auxiliary'eedwater) signals.

On August 13,

1979, the NRC sent an IE bulletin '(879-21) to all reactor facilities.

Boiling water reactors and those facilities wit) con-struction permits were notified for information.

All operating pressurized water reactor licensees were directed to take steps to evaluate the problem, take corrective actions

needed, and to notify the,NRC of all actions taken as a result of the evaluation.

I&MECo responded in a letter of November 5, 1979 and requested tech-nical specification changes to Tables 2.2-1 and 3.3-4 on Feburary 22, 1980".

This report is a technical evaluation of the material submitted by, and actions taken by, I&MECo for the Cook units.

2.0 EVALUATION OF THE COOK NUCLEAR STATION, UNITS 1 AND 2 2,.1 Review Guidelines.

Ii Bulletin No. 79-21 provided the pressuri "ed water power reactor licensees with the follow ng NRC guidelines:

~ l.

'Guideli

- The licensee is to review l.iquid level measurement systems within containment.

Ef the signals axe used to initiate safety actions or to pro-vide post-accident monitoring information, a description of the system is to be suhaitted.

2.

Guideline 2 - For those systems identified by guide-line l, the licensee is to evaluate the effect of post-accident temperature on the indicated water level (i.n comparison to the actual water level, including all sources of error).

3.

Guideline 3 The licensee is to provide a listing of all safety and control setpoints used with the level instrumentation, and veri,fy proper setpoint actuation throughout the range of ambient temperature (including accident temperatures) 4.

Guideline 4 Ef a change of setpoints is necessary to ensure safe action, the licensee is to describe the corrective action and state when the action was taken.

5.

Guideline 5 " The licensee is to ensure that the opera-tors are instructed on the potential for, and the poten" tial magnitude of, erroneous level signals.

The com-pletion date for procedure changes and operator training is to be identified.

2.2 Description of Li uid Level Heasurement Systems.

E&iKCo has identified the steam generator narrow-range water level (SG level) and pressurizer level (PZR level) systems as inside containment, and used to ini,tiate safety functions or to provide post-accident monitoring informa-3 tion.

They provided the following descript'n:

"The SG level ference leg is a conventional condensing pot open column system contained entirely within the lower volume of the containment.

The PZR level high side reference leg is a sealed bellows type filled with distilled water and is contained partially in the lower volume and partially in the upper volume of the containment.

The PZR level low side reference leg is a conventional open column system entirely contained in the lower volume of the containment."

The FSAR indicates that the SG low-low level trip, on two out of three coincidence in any one steam generator, will cause a reactor trip.

High pressurizer level serves as a backup to high pressurizer pressure on two out of three coincidence, to cause a reactor trip; however, the FSAR does not take credit for this function Both the SG'level and PZR level

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3 provide post-accident monitoring information in the control room.

Table h,.5-2 of the FSAR shows that the PZR level is designed to operate for l/2 hour after an accident.

Section 7.5.2 of the FSAR indicates that pres-surizer pressure and level are the only transmitters inside containment that are required to actuate the Engineered Safety Features.

As a result of the W evaluation, Z&14ZCo raised, on June 30,

'979, the SG low-low setpoints to 17/ and 21/ of the narrow range instrument span for Units l and 2

res e

p ctively.

The Increase was chosen as recommended bv W, to correspond the temperature limit at which the containment pressure-high setpoint would cause a safety injection.

Z&MECo dismissed leve'evel bias associated with high SG water level at elevated containment temper-f atures as insignificant.

2 3.3 Evaluation of Liquid Level 'measurement S stems.

Guideline 1

requested the 1'censee to revie~ all liquid measurement systems within containment.

I&1KCo has provided this review and descriptions oi the pressurizer level and steam generator level transmitters.

Guideline quested the licensee to compar ndicated and actual water levels in a post-accident environment.

Reference 3 provides the requested comparison, and states that reference leg boiling will not occur.

Guideline 3 requested that the setpoints used on safety and control circuits be identified with verification of proper setpoint actuation thoughout the temperature range.

Guideline 4 requested documentation of any corrective action taken.

'Ihe SG low-low setpoints were increased at both units.

The revised setpoints of 17X and 21K of narrow range instru-mentation (Units 1 and 2, respectively) were incorporated in June 1979 The increased setpoint is based on the M recommendation and analysis, and on the containment temperature expected before the high containment pressure reaches its setpoint for safety injection (which vill cause a reactor trip).

Since no credit is taken in the safety analysis for the pressurizer level high trip, no changes were made to this setpoint.

The FSAR (Sec-tion 7.5.3) states that the reference leg of the pressuri er level transmitter will not exceed 140 F.

A corresponding temperature for the Steam Generator level transmitters is not established by the FSAR.

The signals identified and setpoints revised for the temperature range specified by l&HECo are vithin the analyzed limits of the FSAR and redundant instrument trip signals.

To ensure that the pressurizer does not become water solid, or that the pressurizer heaters do not become uncovered, I&MLCo has established procedural limits to the allowable pressurizer level that takes the temper-ature effect on the level sensor into account.

The actions taken and 3

the information documented satisfies guidelines 3 and 4.

Guideline 5 requires any procedure changes and operator training needed as a result of lZ Bulletin No. 79-21 be scheduled.

Reference 3 indicates that operator training and needed procedure changes have been completed.

0 The maternal submitted by I&MECo for this review has been evaluated to the guidelines of IE Bulletin No. 79-21.

The changes in operator training and procedural changes concerning the steam generator level and the pres-surizer level signals satisfy the guidelines of the IE bulletin.

Changes to the steam generator low-low level trip setpoints satisfy the guidelines bf the IE bulletin and the westinghouse recommendations.

The pressurizer level high trip setpoint was not changed since this was not credited with a safety function in the FSAR.

This is acceptable as the trip is within the analyzed limits.

The NgC should approve the proposed technical specification changes for the steam generator Low-low level setpoints of 17/ for Unit 1 and 21K for Unit 2, with allowable Limits of 16K and 20K, respectively.

4. 0 REFERENCES 1.

t letter T. M. And An erson, to U ~ S

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NRC, Victor Stello, "Steam Generator Mater Level," NS'-TMA-2104, June 22, 1979

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NRC letter to allll power reactor licensees and construction permit holders, "IE Bulletin No. 79-21 " Augus t 13, 1979.

3 ~

I&HECo le tter R.

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S ~ Hunter, to U.S ~

NRC, James G ~ Keppler, "Response

.to IE Bulletin No. 79-21," AEP:NRC:00271, November 5, 1979.

I6MECo letter, John E. Dolan, to U ~ S ~

NRC, Haro Ld R ~ Denton 1

AEP: NRC:00313, February 22, 1980.