ML19253A461

Forwards IE Bulletin 79-21, Temp Effects on Level Measurements. No Action Required

ML19253A461
Person / Time
Site:	Catawba, Perkins, Cherokee
Issue date:	08/13/1979
From:	James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	Dail L DUKE POWER CO.
References
NUDOCS 7909100059
Download: ML19253A461 (2)

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UNITED ST ATES

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'o, NUCLEAR REGULATORY COMMISSION y

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101 MARIETT A ST., N W., SUITE 3100 ATLANTA. GEORGIA 30303

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AUG 131979 In Reply Refer To:

RII:JP0 50-L13, 50-L1h 50 L88, -50 L89 50-L90,150 L91'S (502.L92', 50-L93 i

Duke Power Compary Attn:

L. C. Dail, Vice President Design Engineering P. O. Box 33189 Charlotte, North Carolina 28242 Gentlemen:

The enclosed Bulletin 79-21 is forwarded to you for information.

No written response is required.

If you desire additional infomation regardirg this matter, please contact this office.

Sincerely, 9

'T C4 Js - P. O'Reilly Direuor

Enclosures:

1.

IE Bulletin No. 79-21 v/ encl.

2.

List of IE Bulletins Issued in the Last 6 Months 333100 7900100059

AUC 131979

. Duke Power Company we w/encI:

D. G. Beam, Project Manager Post Office Box 223 29710 Clover, South Carolina J. T. Moore, Project Manager Post Office Box 422 29340 Gaffney, South Carolina I

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Accession No:

7908090193 SSINS No: 6820 UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMEh~f WASHINGTON, D.C.

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

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

1, addresses the effect of increased containment The report, Enclosure No.

h temperature on the reference leg water column and the resultant effect on level.

indicated steam generator water steam generator level to be higher than the actual level and could delay or prevent protection signals and could, also, provide erroneous information d Enclosure No. I addresses only a Westinghouse steam post-accident monitoring. generator reference leg water column; however, measuring systems utilized on other steam generators and reactor coolant systems could be affected in a similar manner.

Actions To Be Taken By Licensees:

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

Review the liquid level measuring systems within containment to determine if the signals are used to initiate safety actions or are used to provide 1.

Provide a description of systems post-accident monitoring information.that are so employed; a descrip be included, i.e., open column or sealed reference leg.

On those systems described in Item I above, evaluate the effect of post-acci ambient temperatures on the indicated water level to determine any change 2.

This evaluation must in indicated level relative to actual water level.

include other sources of error including the effects of varying fluid pressure and flashing of reference leg to steam on the water level measu The results of this evaluation should be presented in a tabular form similar to Tables 1 and 2 of Enclosure 1.

Review all safety and control setpoints derived fr m level signals to verify that the setpoints will initiate the action requiren by the plant safety 3.

analyses throughout the range of ambient temperatures encountered by the Provide a listing of instrumentation, including accident temperatures.

these setpoints.

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

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August 13, 1979 IE Bulletin No. 79-21 Page 2 of 2 to ensure If the above reviews and evaluations require a revision of setpoint:

safe operation, provide a description of the corrective action and the date If any corrective action is temporary, submit a

the action was completed.

description of the proposed final corrective action and a timetable for implementation.

include specific Review and revise, as necessary, emergency procedures to information obtained from the review and evaluation of Items 1, 2 and 4.

3 to ensure that the operators are instructed on the potential for and All tables, curves, or correction magnitude of erroneous level signals. factors that would be applied to If revisions to procedures are required, provide available to the operator.

a completion date for the revisions and a completion date for operator training on the revisions.

A report of the above actions shall be sube'tted 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 water reactors with an operating license and all power reacto with a construction permit, written response is required.

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

given under a blanket clearance specifically for identified generic problems.

Enclosure:

Memo Westinghouse Electric Corp.

to Victor Stello dated June 22, 1979 a~csn y O d = >;, J

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

Dear Mr. Stello:

Srbject:

Steam Generator Weter Level 21, 1979 with Mr.

This is to confim my telephone conversation of June Norman C. Mcseley, 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 l

steam generator water level and recomended that they incorporate those j

cc-recticns in the steam generator low water level prot (Ction system setpoints and emergency operating procedures for operating plants as l

appropriate.

High energy line breaks inside containment can result in heatup of the Increased reference j

steam generator level Tneasurement reference leg.

1eg water column temperature will result in a decrease of the water i

column density with a consecuent apparent increase in the indicated l

stea generator water level (i.e., apparent level exceeding actual l

This potential level bias could result in delayed protection level).

signals (reactor trip and auxiliary feedwater initiation) which areIn the case based on low-low steam generator water level.

rupture, this adverse environment could be present and could delay or l

prevent the primary signal arising from declining steam generator w j

1evel (low-low steam generator level).

signals avcilable in those Westinghouse plants which take credit in l

Finel Safety Analysis Reports for steam generator water level trh with overtemperature delta T; high an adverse containment environment:

For pressurizer pressure; containment pressure and safety injection.

other high energy line breaks which could introduce a similar positive bias to the steam generator water level rneasurem j

would not interfere with needed protective system actuation.

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Westinghcuse has advised all customers with af fected operating plants tha d

, the potential temperature-induced bias in indi.cated level can be compense For g

..for by raising the steem generetor icw-low wa ter icvel setsoint.

imediate action Westinghouse has reco rended a change in the allowable I

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water level setpoint sufficient to acco-:nodate the bias (uo to 10% of level l

Containment which could result from containment ter:peratures up to 230'T.

analyses following a secondary high energy line tren on tycical olents ha shown thtt a containment high pressure signal woule. te gene:ated conteinrent temperature reaches 28? F.

e>etsuremtnt errers occurring simultaneously in the advera *irection res l

in the centainment high pressure signal becor-ir,; the primary prctective function following som'e feedline rupture events, i,e,, for those cases in t

low-which the containment tenperature exceeds 280*F before a steam genera or I

low water level trip is actuated, the high conte'nment pressure signal p The combination of the revised low ',0w water level setpoint and the high contair.9ent pressure signal will provide reacte-trip and aux protecticn.

l feedwater initiation following a feedline rupture and will ensure that the feedline break criteria stated in the Safety Analyds P*3crts centirse to b i

j Sea applicants may choose to use plant-stteific contchmnt analyses, possibly combined with changes in the contairin met.

i lete'l SEtcoint.

j acco nodated in the steam generator low-low wate?

The potential steam generator level measurement bics also has Since the tost-accide.t envirormnt I

post-accident ntnitoring consideratf or.s.for high energy line b i

A cositive 10% limit which must be considered for prctection syster actuation.

bias of up to 20% can be anticipated for en extreme envircn ental co The appropriate bias must be coupled with instrumentaticn a*.d other p errors, to determine the required range of indicated level to be main i

during post-accident monitoring to ensure that tne stear generator tub Westinghouse has fully covered and the steam generator is not water solic.

l provided all of its cust'omers with operating plants with inferrat l

them to codify their emergency operating procedures to ensure tna steam generator level temperature bias allowance is made.

,jg in steam generator level may In c related area, it has been found that a bias D

also be introduced by changes in steam generator press gg steam generator fluid densities.

Westin; house has notified all all of its custoners with operating plants.

customers with operating plants that such a bias will exist in r.ation of all steam generators and that the operator shculd be in ti monitor steem generator pressure, as well as level, to ensure that b3 bias is reflected in his post-accident recovery actions.

G Also, following depressurization of any steam generator, boilin

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occur in the reference leg and cause a major M

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be erronecus due to the indication in the depressurized steam generators may potential boiling in the reference leg.

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For plants under construction, customers have been advised of the above affe 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.

They have been informed i

The attached tables have been supplied to all customers.that w I

under 10CFR21 in operating plants and as a significant deficiency under 8

10CFR50.55(e) for plants under construction.

Should ycu have any questions on tb1s material, please contact W. K. R. J I

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j (412/373-4795).

Very truly yours.

Westinghouse Electric Corooration W

w T. M. Andersen, Manager Nac1 ear Safety i

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. JPC:kk Mr. Norr.an C. Moseley cc:

Director, DRO&I Mr. Samuel E. Bryan Asst. Director, DRO&I I

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.l' TABLE 1 i

. Correction to indicated steam generator water level.for Reference Leg Heatup effects due to post-accident cor.tainmer,t temperature (before reactor tric)

Correction to S/G Level, Maximum containment ter.perature

of Span reached before reactor trip,'F 0%

90*

4%

200*

10%

i 280*

13%

320*

20t 400

' BASIS:

Leve[ Calibration Pressure < 1000 psia i

Referer.ce Leg Calibration Temperature > 90*F Height of Reference Leg 5 1x Level Span 1

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TABLE 2 l

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Corrections to allowable indicated steam generator water level for Reference Leg Heatup cad Pressure changes fo11orting a hign-energy line break, to assure tnat true level is between the level ta::s t

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Corrections to I

Cortection To Maximum Allowed Mininum Allowed Indicated Lesci, Contai nment Indicated Level,

of Span Temperature

% of Span

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-4

• 1 90'

-4

+6 200'

-4

+11 ~

280-

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+14 320'

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+21 400*

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1000 psia t

I Level Calibration Pressure 5 Reference Leg Calibration Temperature 190*F I

1.1 x Level Span 1

Height of Reference leg 3 Pressure y,50 psia

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Pressure 5 200 psi + Calibration Pressure l

Boiling in the Referen:e Leg is not assumed.

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Enclosure IE Bulletin No. 79-21 Page 1 of 3 August 13, 1979 LISTIhc 0F IE BULLETINS ISSUED IN LAST SIX MONTHS Date Issued Issued To Bulletin Subject No.

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

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

Designs Using Concrete OL or a CP Expansion Anchor Bolts 79-12 Short Period Scrams at 5/31/79 All GE BWR Facilities with an OL BWR Facilities mJtgsJi.3.< ; r <, r3 (s

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

79-11 F3ulty Overcurrent Trip 5/22/79 All Power Reactor Facilities with an Device in Circuit Breakers OL or a CP for Engineered Safety Systems 79-10 Requalification Training 5/11/79 All Power Reactor Facilities with an OL Program Statistics 79-09 Failures of GE Type AK-2 4/17/79 All Power Reactor Facilities with an Circuit Breaker in Safety OL or CP Related Systems 79-uS Events Relevant to BWR 4/14/79 All BWR Power Reactor Facilities with an OL Reactors Identified During Three Mile Island Incident 79-07 Seismic Stress Analysis 4/14/79 All Power Reactor Facilities with an of Safety-Related Piping OL or CP 79-05C&O6C Nuclear Incident at Three 7/26/79 To all PWR Power Reactor Facilities Mile Island - Supplement with an OL 79-06B Re"iew of Operational 4/14/79 All Combustion Engineer-ing Designed Pressurized Errors and System Mis-Water Power Reactor alignments Identified Facilities with an During the Three Mile Operating License Island Incident 79-06A Review of Operational 4/18/79 All Pressurized Water Power Reactor Facilities (Rev 1)

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

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

79 06 Review of Operational 4/11/79 All Pressurized Water Power Reactors with an Errors and System Mis-OL except B&W facilities alignments Identified During the Three Mile Island Incident 5/21/79 All B&W Power Reactor 79-05B Nuclear Incident at Facilities with an OL Three Mile Island All B&W Powe r Reactor 4/5/79 79-05A Nuclear Incident at Facilities with an OL Three Mile Island 4/2/79 All Power Reactor 79-05 Nuclear Incident at Facilities with an Three Mile Island OL and CP 79-04 Incorrect Weights for 3/30/79 All Power Reactor Facilities with an Swing Check Valves OL or CP Manufactured by Velan Engineering Corporation 3/19/79 All Power Reactor 78-12B Atypical Weld Material Facilities with an in Reactor Pressure OL or CP Vessel Welds Longitudinal Welds Defects 3/12/79 All Power Reactor Facilities with an 79-03 In ASME SA-312 Type 304 OL or CP Stainless Steel Pipe Spools Manufactured by Youngstown Welding and Engineering Co.

Environmental Qualification 6/6/79 All Power Reactor Facilities with an 79-01A of Class IE Equipment OL or CP (Deficiencies in the Envi-ronmental Qualification of ASCO Solenoid Valves)

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