ML19207B656

Notifies That IE Bulletin 79-21, Temp Effects on Level Measurements, Was Sent 790813 to Licensees Listed

ML19207B656
Person / Time
Issue date:	08/15/1979
From:	Pappas H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:	Wright G ILLINOIS, STATE OF
References
NUDOCS 7909050048
Download: ML19207B656 (1)

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'3, - 'h NUCLEAR REGULATORY COMMISSION E

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GLEN ELLYN, ILLINOIS 60137 AUG 151979 State of Illinois Department of Public Health ATTN:

Mr. Gary N. Wright, Chief Division of Nuclear Safety 535 West Jefferson Street Springfield, IL 62761 Gentlemen:

The enclosed IE Bulletin No. 79-21 titled " Temperature Effects On Level Measurements" was sent to the following licensees on on August 13, 1979, for action and information:

ACTION American Electric Power Service Corporation Indiana and Michigan Power Company D. C. Cook 1, 2 (50-315, 50-316)

Commonwealth Edison Company Zion 1, 2 (50-295, 50-304)

Consumers Power Company Palisades (50-255)

Northern States Power Company Prairie Island 1, 2 (50-282, 50-306)

Toledo Edison Company Davis-Besse 1 (50-346)

Wisconsin Electric Power Company Point Beach 1, 2 (50-266, 50-301)

Wisconsin Public Service Corporation Kewaunee (50-305)

INFORMATION Cincinnati Gas & Electric Company Zimmer (50-358)

Cleveland Electric Illuminating Company Perry 1, 2 (50-440, 50-441)

Commonwealth Edison Company Braidwood 1, 2 (50-456, 50-457)

Byron 1, 2 (50-454, 50-455)

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Dresden 1, 2, 3 (50-10, 50-237, 50-249)

La Salle 1, 2 (50-373, 50-374)

Quad-Cities 1, 2 (50-254, 50-265)

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State of Illinois Consumers Power Company Big Rock Point (50-155)

Midland 1, 2 (50-329, 50-330)

Dairyland Power Cooperative LACBWR (50-409)

Detroit Edison Company Fermi 2 (50-341)

Illinois Power Company Clinton 1, 2 (50-461, 50-462)

Iowa Electric Light & Power Company Duane Arnold (50-331)

Northern Indiana Public Service Company Bailly (50-367)

Northern States Power Company Monticello (50-263)

Tyrone Energy Park 1 (50-484)

Ohio Edison Company Erie 1, 2 (Pre-CPP)

Public Service of Indiana Marble Hill 1, 2 (50-546, 50-547)

Union Electric Company Callaway 1, 2 (50-483, 50-486)

Sincerely,

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Helen Pappas, Chief Administrative Branch Enclo;ures:

1.

IE Bulletin No. 79-21 2.

List of IE Bulletins Issued in the Last 6 Months cc w/encls:

Mr. D. W. Kane, Sargent & Lundy Central Files Reproduction Unit NRC 20b Local PDR NSIC TIC 3 >'),.

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Accession No:

7908090193 SSINS No:

6820

• J UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.

20555 August 13, 1979 IE Bulletin No. 79-21 TEMPERATURE EFFECTS ON LEVEL MEASUREMENTS Description of Circumstances:

On June 22, 1979, Westinghouse Electric Corporation reported, to NRC, a potential substantial safety hazard under 10 CFR 21.

The report, Enclosure No.1, addresses the effect of increased containment temperature on the reference leg water column and the resultant effect on the indicated steam generator water level.

This effect would cause the indicated steam generator level to be higher than the actual level and could delay or prevent protection signals and could, also, provide erroneous information during post accident monitoring.

Enclosure No. I addresses only a Westinghouse steam generator reference leg water column; however, safety related liquid level measuring systems utilized on other steam generators and reactor coolant systems could be affected in a similar manner.

Actions To Be Taken By Licensees:

For all pressurized water power reactor facilities with an operating license:*

1.

Review the liquid level measuring systems within containment to determine if the signals are used to initiate safety actions or are used to provide post-accident monitoring information.

Provide a description of systems that are so employed; a description of the type of reference leg shall be included, i.e., open column or sealed reference leg.

2.

On those systems Gescribed in Item 1 above, evaluate the effect of post-accident ambient temperatures on the indicated water level to determine any change in indicated level relative to actual water level.

This evaluation must include other sources of error including the effects of varying fluid pressure and flashing of reference leg to steam on the water level measurenants.

The results of this evaluation should be presented in a tabular form similar to Tables 1 and 2 of Enclosure 1.

3.

Review all safety and control setpoints derived from level signals to verify that the setpoints will initiate the action required by the plant safety analyses throughout the range of ambient temperatures encountered by the instrumentation, including accident temperatures.

Provide a listing of these setpoints.

" Boiling water reactors have been requested by a July generic letter from the NRC to provide similar information.

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IE Bulletin No. 79-21 August 13, 1979 Page 2 of 2 If the above reviews and evaluations require a revision of setpoints to ensure safe operation, provide a description of the corrective action and the date the action was completed.

If any corrective action is temporary, submit a description of the proposed final corrective action and a timetable for implementation.

4.

Review and revise, as necessary, emergency procedures to include specific

. information obtained from the review and evaluation of Items 1, 2 and 3 to ensure that the operators are instructed on the potential for and magnitude of erroneous level signals.

All tables, curves, or correction factors that would be applied to post accident monitors should be readily available to the operator.

If revisions to procedures are required, provide a complation date for the revisions and a completion date for operator training on the revisions.

A report of the above actions shall be submitted within 30 days of the receipt of this Bulletin.

Reports should be submitted to the Director of the appropriate NRC Regional Office and a copy should be forwarded to the NRC Office of Inspection and Enforcement, Division of Reactor Operations Inspection, Washington, D. C.

20555.

For boiling w-ter reactors with an operating license and all power reactors with a construction permit, this Bulletin is for information purposes and no written response is required.

Approved by GAO, B180225 (R0072); clearance expires 7/31/80.

Approval was given under a blanket clearance specifically for identified generic problems.

Enclosure:

Memo Westinghouse Electric Corp.

to Victor Stello dated June 22, 1979 3 %(W' ' -

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IE Bulletin No. 79-21 Enclosure August' 13, 1979 Page 1 of 3 LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS Bulletin Subject Date Issued Issued.To No.

79-20 Packaging Low-Level 8/10/79 Materials Licensees Radioactive Waste for who did not receive Transport and Burial Bulletin No. 79-19 79-19 Packaging Low-level 8/10/79 All Power and Research Radioactive Waste for Reactors with OLs, Transport and Burial fuel facilities except uranium mills, and certain materials licensees 79-18 Audibility Problems 8/7/79 All Power Reactor Encountered on Evacuation Facilities with an Operating License 79-17 Pipe Cracks in Stagnant 7/26/79 All PWR's with Borated Water Systems at operating license PWR Plants 79-16 Vital Area Access Controls 7/26/79 All Holders of and applicants for Power Reactor Operating Licenses who anticipate loading fuel prior to 1981 79-15 Deep Draft Pump 7/11/79 All Power Reactor Deficiencies Licensees with a CP and/or OL 79-14 Seismic Analyses for 6/2/79 All Power Reactor As-Built Safety-Related facilities with an Piping System OL or a CP 79-13 Cracking In Feedwater 6/25/79 All PWRs with an System Piping OL for action. All BWRs with a CP for information.

79-02 Pipe Support Base Plate 6/21/79 All Power Reactor (Rev. 1)

Designs Using Concrete F1cilities with an Expansion Anchor Bolts OL or a CP 79-12 Short Period Scrams at 5/31/79 All GE BVR Facilities BWR Facilities with an OL n'jG

IE Bulletin No. 79-21 Enclosure August'13, 1979 Page 2 of 3 LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS Bulletin Subject Date Issued Issued To No.

79-11 Faulty Overcurrent Trip 5/22/79 All Power Reactor Device in Circuit Breakers Facilities with an for Engineered Safety OL or a CP Systems 79-10 Requalification Training 5/11/79 All Power Reactor Program Statistics Facilities with an OL 79-09 Failures of GE Type AK-2 4/17/79 All Power Reactor Circuit Breaker in Safety Facilities with an Related Systems OL or CP 79-08 Events Relevant to BWR 4/14/79 All BWR Power Reactor Reactors Identified During Facilities with an OL Three Mile Island Incident 79-07 Seismic Stress Analysis 4/14/79 All Power Reactor of Safety-Related Piping Facilities with an

.c OL or CP 79-OSC&O6C Nuclear Incident at Three 7/26/79 To all PWR Power Mile Island - Supplement Reactor Facilities with an OL 79-06B Review of Operational 4/14/79 All Combustion Engineer-Errors and System Mis-ing Designed Pressurized alignments Identified Water Power Reactor During the Three Mile Facilities with an Island Incident Operating License 79-06A Review of Operational 4/18/79 All Pressurized Water (Rev 1)

Errors and System Mis-Power Reactor Facilities alignments Identified of Westinghouse Design During the Three Mile with an OL Island Incident 79-06A Review of Operational 4/14/79 All Pressurized Water Errors and System Mis-Power Reactor Facilities alignments Identified of Westinghouse Design During the Three Mile with an OL Island Incident

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IE Bulletin No. 79-21 Enclosure August 13, 1979 Page 3 of 3 LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS Bulletin Subject Date Issued Issued.To No.

79-06 Review of Operational 4/11/79 All Pressurized Water Errors and System Mis-Power Reactors with an alignments Identified OL except B&W facilities During the Three Mile Island Incident 79-05B Nuclear Incident at 5/21/79 All B&W Power Reactor Three Mile Island Facilities with an OL 79-05A Nuclear Incident at 4/5/79 All B&W Power Reactor Three Mile Island Facilities with an OL 79-05 Nuclear Incident at 4/2/79 All Power Reactor Three Mile Island Facilities with an OL and CP 79-04 Incorrect Weights for 3/30/79 All Power Reactor Swing Check Valves Facilities with an Manufactured by Velan OL or CP Engineering Corporation 78-128 Atypical Weld Material 3/19/73 All Power Reactor in Reactor Pressure Facilities with an Vessel Welds OL or CP 79-03 Longitudinal Welds Defects 3/12/79 All Power Reactor In ASME SA-312 Type 304 Facilities with an Stainless Steel Pipe Spools OL or CP Manut.ctured by Youngstown Welding and Engineering Co.79-01A Environmental Qualification 6/6/79 All Power Reactor of Class 1E Equipment Facilities with an (Deficiencies in the Envi-OL or CP ronmental Qualification of ASCO Solenoid Valves) d l-s m:

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June 22,1979 RS-TMA-2104 Mr. Victor Stello Director, Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission East West Towers Building 4350 East West Highway Bethesda, Maryland 20014

Subject:

Steam Generator Water Level This is to confirm my telephone conversation of June 21. 1979 with Mr.

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Norman C. Moseley, Director, Division of Reactor Operation and Inscec-tion and Mr. Samuel E. Bryan, Assistant Director for Field Coordination.

In that conversation, I reported that Westinghouse had informed its utility customers of corrections that should be applied to indicated steam generator water level and recocmended that they incorporate those ccrrecti:ns in the steam generator low water level protection system

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setpoints and emergency operati_ng procedures for operating plants as appropriate.

High energy line breaks inside containment can result in heatup of the steam generator level measurement reference leg.

Increased reference leg water column temperature will result in a decrease of the water column density with a consequent inparent increase in the indicated steam generator water level (i.e., apparent level exceeding actual level). This potential level bias could result in delayed protection signals (reactor trip and auxiliary feedwater initiation) which are based on low-low steam generator water level.

In the case of a feedline rupture, this adverse environment could be present and could delay or prevent the primary signal arising from declining steam generator water level (low-low steam generator level).

The following is a list of backup signals available in those Westinghouse plants which take credit in their Final Safety Analysis Reports for steam generator water level trip with an adverse containment environment:

overtemperature delta T; high

, pressurizer pressure; containment pressure and safety injection.

For other high energy line breaks which could introduce a similar positive bias to the steam generator water level measurement, steam generator level does not provide the primary trip function and the potential bias would not interfere with needed protective system actuation.

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Westinghcuse has advised all customers with affected operating lants that the potential temperature-induced bias in indi.cated level can be compensated for by raising the steam generator low-low water level setooint.

For innediate action, Westinghouse has recommended a change in the allowable water level setpoint sufficient to accommodate the bias (uo to 10% of level) which could result from containment temperatures up to 2309.

Containment analyses following a secondary high energy line break on tytical plants have

.shown that a containment high pressure signal would be generated before the containment te:nperature reaches 280*F.

Thus, postulation of all water-level measurement errors occurring simultaneously in' the afverse cirection results

'in the centainment high pressure signal becoming the pri.ary protective function following some feedline rupture events, i.e., for those cases in which the containment temperature exceeds 280*F before a steam generator low-low water level trip is actuated, the high containment pressure signal provides protection.

The combination of the revised low-low water level setpoint and the high containment pressure signal will provide reactor trip and auxiliary feedwater initiation following a feedline rupture and will ensure that the feedline break criteria stated in the Safety Analysis Reports continue to be met. Scae applicants may choose to use plant-specific containment analyses, possibly combined with changes in the containment high-pressure.setpoint, to justify reducing the bias introduced due to reference le2 heatup which must be accor4nodated in the steam generator low-low water level setpoint.

The potential steam generator level measurement bias also has implications for post-accident monitoring considerations.

Since the cost-accident environment for high energy line breaks can exceed 280'F, the level bias can exceed the 10% limit which must be considered for protection syster actuation. A positive bias of up to 20% can be anticipated for en extrem. enviren. ental condition.

The appropriate bias must be coupled with instrumentaticn and other process errors, to determine the required range of indicated level to be maintained during post-accident monitoring to ensure that the steam generator tubes are fully covered and the steam generator is not water solic.

a'estinghouse has provided all of its cust'omers with operating plants with information to enable them to modify their emergency operating procedures to ensure that suitable steam generator level temperature bias allowance is made.

In a related area, it has been found that a bias in steam generator level may also be introduced by changes in steam generator pressure, due to changes in steam generator fluid densities.

Westinghouse has quantified this effect for al.1 of its customers with operating plants.

Westinghouse has notified all customers with operating plants that such a bias will exist in the level indi-cation of all steam generators and that the operator should be instructed to conitor steam generator pressure, as well as level, to ensure that the potential bias is raeflected in his post-accident recovery actions.

Also, following depressurization of any steam generator,' boiling could conceivably occur in the reference leg and cause a major bias fer a short time pe'riod.

Westinghouse has notified all customers with operating plants that the water level indication in the depressurized steam generators may be erronecus due to the pote_ntial boiling in the reference leg.

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For plants under construction, customers have been advised of the above affects, and the options open to them for corrective ac. tion will be reviewed in a timely The NRC will-be advised of proposed resolutions for these. plants.

manner.

The attached tables have been supplied to all customers.

They have been informed that we are reporting this to you as a potential substantial safety hazard under 10CFR21 in operating plants and as a significant deficiency under 10CFR50.55(e) for plants under construction.

Should ycu have any questions on this material, please contact Mr. K. R. Jordan (412/373-4795).

e ne Very truly yours.

Westinghouse Electric Corocration i

d T. M. Anderson, Manager Nuclear Safety JPC:kk ccc Mr. Norman C. Moseley Director DRO&I Mr. Samuel E. Bryan Asst. Director, DRO&I e

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TABLE 1

. Correction to indicated steam generator water level for Reference Leg Heatup effects due to post-accident containment temperature (before reactor trip)

Maximen contain=ent temperature Correction to S/G Level, reached before reactor trip, 'F

" of Span 90' 0%

200*

4%

280' 10 320' 13:

400' 20t

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1 Level Calibration Pressure 5 000 psia Reference Leg Calibration Temperature 190'F Height of Reference Leg 5 1.1x Le"a1 Span 6

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.s s-TABLE 2 Corrections to allowable indicated steam generator water level for Reference Leg Heatup and Pressure changes following a high-energy line break, to assure that true level is between the level taos, Correction To Corrections to Containnent Mininum Allowed Maximum Aliowed Temperature Indicated Level.

Indicated Level,

'F 1 of Span

" of Span 90'

+1

-4 200*

+6

-4 280*

+11 *

-4 320'

+14

-4 400*

+21

-4 BASIS:

Level Calibration Pressure 5 1000 psia Referance Leg Calibration Temperature 1 90*F Height of Reference Leg - 1.1 x Level Span Pressure 1 50 psia Pressure 5 200 psi + Calibration Pressure Boiling in the Reference leg is not assumed.

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