ML18130A540

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

ML18130A540
Person / Time
Site:	Surry, North Anna
Issue date:	08/13/1979
From:	James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	Proffitt W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
References
NUDOCS 7909100106
Download: ML18130A540 (12)

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UNITED STATES NUCLEAR REGULATORY COMMISSION REGION II 101 MARIETTA ST., N.W., SUITE 3100 ATLANTA, GEORGIA 30303 In Reply Refer To:

RII:JPO~*

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60-281 Virginia Electric and Power Company Attn:

W, L, Proffitt Senior Vice President, Power P.O. Box 26666 Richmond, Virginia Gentlemen:

23261 AUG 13 1979 Enclosed is IE Bulletin 79-21 which requires action by you with regard to your PWR power reactor facility(ies) with an operating license.

Should you have any questions regarding this Bulletin or the actions required by you, please contact this office.

Sincerely,

~~~~

Enclosures:

~ James P, 0 'Reilly Director

1.

IE Bulletin No. 79-21 w/encl.

2.

List of IE Bulletins Issued in the Last 6 Months 7 909 I OO(Cf,

~

Virginia Electric and Power Company cc w/encl:

e W.R. Cartwright, Station Manager Post Office Box 402 Mineral, Virginia 23117 P. G. Perry Senior Resident Engineer Post Office Box 38 Mineral, Virginia 23117

~. L. Stewart, Manager Post Office Box 315 Surry, Virginia 23883

.e AUG 1 3 1979 \\' ' i

I UNITED STATES Accession No:

7908090193 SSINS No:

6820 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. 1 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 ~e 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 described in Item 1 above, evaluate the effect of post-acciden1 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 o~ the water level measurements 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.

e 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 inc}ude 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 completion date for the revisions and a completion date for operator training on the revisions.

A report of the above actions shal'l 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 water 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

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Mr. Victor Stello

• Director, Office of Inspection and Enforcement u.s,. Nuc1~ar Regulatory Con1T1is~ion East West Towers Building 4350 East West Highway Bethesda, Maryland 20014

Subject:

Steam Generator Weter Level

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June 22, 1979 IIS-TMA-2104 This is to confirm my telephone conversation of June 21, 1979 with Mr.

Horman C. Moseley, Director, Division of Reactor Operction and Insoec-tion and Mr. Sawuel ~- Bryan, Assistant Director for-Field Coordin!tion.

In that conversation, I reported that Westinghouse hed ~nfonned its utility customers of corrections that should be applied to indicated steam generator water level and reconrnended that they incorporate :hos!

==~~e:!i,ns in the,steam generator low water level protection system setpoints and emergency operating procedures for operating plants as appropriate.

High energy line breaks inside containment can resu1t in heatup of the steam generator level -measurement reference leg.

Increesed reference leg water column temperature wi11 result in a decrease of the water column density with a consequent apparent increase in the indicated steam generator water 1eve1 (i.e., apparent level exceeding actual level).* This potential level bias could result in delayed protection signals (reactor trip and auxiliary feedwater initi?tion) which are based on low-low steam generator water level.

In the case of a feed1ine rupture, this adverse environment could be present and could delay or.

prevent the primary signa1 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 Fina1 Safety Analysis Reports for steam generator water level trip \\r.'ith 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 simi1ar positive bias to the steam generator water le,vel measurement, steam generator level does not provide the primary trip function and the potential bias would not ir.terfere with needed protective system actuation

• e e West1nghcuse has advised al1 customers with affect~d o~erating plants thet

, the potential temperature-induced bi as in in di.ca ted 1evel cen be canpenseted

,.for by raising the steam generator low-low water levfl set:,oint. For 1anediate action, Westinghouse has reco!TITlended a change in the allowable wa~er level setpoint sufficie~t to accommodate the b;as (u~. to 101 of 1eve1) wtnch could result from conta,nment temperatures up to 2ao~r.

Containrner.t analyses following a secondary high energy line break on tytica1 plants have

.shown.that a containment high pressure signal wou1~ =e g=n~rated before the

• conti1n~nt te.~perature reaches 28DcF.

Thus, post~latio~ ~f a11 weter-l~v:1 me~surem:nt errors occurring simu1~aneous:y in* the a~verse cirection results in the containment high pressure signal becoming the pri7iary 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-1ow water level trip is actuated, the high contai.ni.'lent pressure signal provides protecticn. The combination of the revised 1ow-1ow water level setpoir.t and the high contair.:nent pressure signal will provide re!ctc~ trip and auxilia~y feedwater initiation fo11owing a feedline rupture and \\<Jill ensure that the feedline break criteria stated in the Safety Ana1ysis Re~cr:s continue to be met.

Scx:ie app1ic!nts may choose to use plant-specific contfinment ana1ys,s.

~ossibly corroi ned with changes in the containment high-;:re-ssure.setpoint, to Justify redu~ing the bias introduced due to reference 1e; hfatu? which must be ecco:inodited in the steam generator low-low water level set:oint.

The potential steam generator 1eve1 measurement bias also hes i~plicat~ons for.

post-acc1dent monitoring consideratior.s. Since the jOS:-accide~t envirol"lm!nt for high energy line breaks can exceed 280°F, the 1e1el ~ias can exceed the 101 limit which must be considered for protection syster actuation. A positiv!

bias of up to 20: can be anticipated for tn extreme env~rcn.-ental condition.

The eppropriate bi as *must be coupled with i nstru::ient: ti en a!1d other process errors. !o determine the required range of indicated 1eve1 !a be maintained during post-accident monitoring to ensure that the stea~ senera~or tubes are fully covered and the steam generator is not water solic. iiiestinghouse has provided a1i of its customers with operating plants with ~nforrr~tion to enable them to modify their emergency operating procedures to ensure that suitable stear.i ge~erator level temperature b1as allowance 1s made.

• Ina re1ated area. it has been found that a bias in steam g!nerator level may also be introduced by changes in steam generator pressure. due to changes in stea~ gener~tor fluid densities.

Westinghouse has q~an:ifie~ this effect for a11 of its customers with operating plants. Westinghouse h?s 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 shculd be instructed to monitor steim g!nerator pressure. as we11 as level, to ensure *that the poten~iq bias is reflected in his post-accident recovery actions.

A1so. fc11owing depressurization of any steam generator,*~oi11ng coul_d conceiv; occur ir. the reference leg and cause a ~jor bias fer a sho!"t time period.

Westinghouse has notified all customers with operating ?1ants that the wate~ 1 indication in the depressurized steam generators r.:ay be er~necus due to *the pot~ntit1 boiling in the reference leg.

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• For p1ants under construction, customers have been advised of the abcve,ffects, and the options open to them for corrective a~tion wi11 be r~viewed in a timely manner.

The NRC will ~e advised of proposed resolutions for the~e ~1~nts.

The attached tables he*e been supplied*to a11 customers.

They have been informed that we are reporting this to you as a potential substantia1 safety h~zard under 10CFR21 in operating p1ints and as a significant deficiency under

_10CFRS0.55(e) for p1ants under construction.

Should ycu have any questions on this materia1*, p1ease contact ~~r. K. R. Jordan (412/373-4795).

• JPC: kk cc:.Mr. Norman C. Moseley Director. DRO&l Mr. Samuel E. Bryan Asst. Di rector, ORO& I Very truly yours, Westinghouse E1ectric Cor;,ora.tion (2;/~------

T. M. Anderson, Manager Nuc1ear Safety

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TABLE l

. Correction to indicated steam generator

• water level. for Reference leg Heatup effects due to post-accident containment temperature (before reactor trip)

'*ximut:i containr.ient ternper!ture reached before reactor trip. °F Correction to S/G Level,

~ of Span 90° 200° 280° 320° 400° 0~

4%

BASIS:

Level Calibration Pressure!. 1000 psia Reference Leg Ca1ibration Temper~ture ?_ 90°F Height of Reference Leg~ 1. lx Level Span

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t I I I TABLE 2 Corrections to allowable indicated steam generator water level for Reference Leg Heatup and Pressure changes follo~ing a high-energy line break.

to assure that true level is betv,een the level taps Correction To Corrections to Con ta i nm!nt Mini num A 11 owed Maximum A 11 owed Ter.ipera ture Indicated Level, lndiceted Level,

• f i of Sp!n

~ of Span goo

'+ l 200°

+ 6 280:

+11.

320°

+14 400°

+21 BASIS:

Level Calibration Pressure.! JOOO psia Reference Leg Calibration Temperatur~*~ 90°F Height of Reference Leg~ 1.1 x Level Span Pressure> SO psia Pressure.!290 psi+ talib~ation Pressure Boiling in the Reference Leg is not assumed.

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IE Bulletin No. 79-21 August 13, 1979 Bulletin No.

79-20 79-19 79-18 79-17 79-16 79-15 79-14 79-13 79-02 (Rev. 1) 79-12 LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS Subject Date Issued Packaging Low-Level 8/10/79 Radioactive Waste for Transport and Burial Packaging Low-Level 8/10/79 Radioactive Waste for Transport and Burial Audibility Problems 8/7/79 Encountered on Evacuation Pipe Cracks in Stagnant 7/26/79 Borated Water Systems at PWR Plants Vital Area Access Controls 7/26/79 Deep Draft Pump 7/11/79 Deficiencies Seismic Analyses for 6/2/79 As-Built Safety-Related Piping System Cracking In Feedwater 6/25/79 System Piping Pipe Support Base Plate Designs Using Concrete Expansion Anchor Bolts Short Period Scrams at BWR Facilities 6/21/79 5/31/79 Enclosure Page 1 of 3 Issued To Materials Licensees who did not receive Bulletin No. 79-19 All Power and Research Reactors with OLs, fuel facilities except uranium mills, and certain materials licensees All Power Reactor Facilities with an Operating License All PWR' s with operating license All Holders of and applicants for Power Reactor Operating Licenses who anticipate loading fue prior to 1981 All Power Reactor Licensees with a CP and/or OL All Power Reactor facilities with an OL or a CP All PWRs with an OL for action. All BWRs with a CP for information.

All Power Reactor Facilities with an OL or a CP All GE BWR Facilities with an OL

e IE Bulletin No. 79-21 August 13, 1979 Bulletin No.

79-11 79-10

• 79-09 79-08 79-07 79-05C&06C 79-06B 79-06A (Rev 1)79-06A LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS Subject Faulty Overcurrent Trip Device in Circuit Breakers for 'Engineered Safety Systems Requalification Training Program Statistics Failures of GE Type AK-2 Circuit Breaker in Safety Related Systems Events Relevant to BWR Reactors Identified During Three Mile Island Incident Seismic Stress Analysis of Safety-Related Piping Nuclear Incident at Three Hile Island - Supplement Review of Operational Errors and System Mis-alignments Identified During the Three Mile Island Incident Review of Operational Errors and System Mis-alignments Identified During the Three Mile Island Incident Review of Operational Errors and System Mis-alignments Identified During the Three Hile Island Incident Date Issued 5/22/79 5/11/79 4/17 /79 4/14/79 4/14/79 7/26/79 4/14/79 4/18/79 4/14/79 Enclosure Page 2 of 3 Issued To All Power Reactor Yacilities with an 01 or a CP All Power Reactor Facilities with an 01 All Power Reactor Facilities with an 01 or CP All BWR Power Reactor Facilities with an OL All Power Reactor Facilities with an OL or CP

• To all PWR Power Reactor Facilities with an OL All Combustion Engineer-ing Designed Pressurized Water Power Reactor Facilities with an Operating License All Pressurized Water Power Reactor Facilitie~

of Westinghouse Design with an OL All Pressurized Water Power Reactor Facilities of Westinghouse Design with an OL

IE Bulletin No. 79-21 August 13, 1979 Bulletin No.

79-06 79-0SB 79-0SA 79-05 79-04 78-12B 79-03 79-0lA LISTING OF IE BULLETINS ISSUED IN LAST SIX MONTHS Subject Date Issued Review of Operational 4/11/79 Errors and System Mis-alignments Identified During the Three Mile Island Incident Nuclear Incident at 5/21/79 Three Mile Island Nuclear Incident at 4/5/79 Three Mile Island Nuclear Incident at 4/2/79 Three Mile Island Incorrect Weights for 3/30/79 Swing Check Valves Manufactured by Velan Engineering Corporation Atypical Weld Material 3/19/79 in Reactor Pressure Vessel Welds Longitudinal Welds Defects 3/12/79 In ASME SA-312 Type 304 Stainless Steel Pipe Spools Manufactured by Youngstown Welding and Engineering Co.

Environmental Qualification 6/6/79 of Class lE Equipment (Deficiencies in the Envi-ronmental Qualification of ASCO Solenoid Valves)

Enclosure Page 3 of 3 Issued To All Pressurized Water Power Reactors with an OL except B&W facilities All B&W Power Reactor Facilities with an OL All B&W Power Reactor Facilities with an OL All Power Reactor Facilities with an OL and CP All Power Reactor Facilities with an OL or CP All Power Reactor Facilities with an OL or CP All Power Reactor Facilities with an OL or CP All Power Reactor Facilities with an OL or CP