ML20153D531

Responds to NRC Re Violations Noted in Insp Rept 50-346/85-25.Corrective Actions:Surveillance Rept 85-118 Written to Document Inaccurate Dimensions for Limitorque-operated Valves

ML20153D531
Person / Time
Site:	Davis Besse
Issue date:	01/21/1986
From:	Williams J TOLEDO EDISON CO.
To:	Norelius C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
References
1-608, NUDOCS 8602240172
Download: ML20153D531 (6)

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TOLEDO Docket No. 50-346 License No. NPF-3 JOE VVitu4Ms. Jn.

Sernor Vce Presders-Nudear (419}249 2300 Serial No. 1-608 PH9)249 5223 January 21, 1985 Mr. C. E. Norelius, Director Division of Reactor Projects United States Nuclear Regulatory Commission Region III 799 Rcosevelt Road Glen Ellyn, IL 60137

Dear Mr. Norelius:

Toledo Edison acknowledges receipt of your December 17, 1985 letter (Log No. 1-1296), Notice of Violation, and Inspection Report No. 50-346/

85025.

Following an examination of the items of concern, Toledo Edison herein offers information regarding these items.

Violation:

10 CFR 50, Appendix B, Criterion V, as implemented by the Toledo Edison Company Nuclear Quality Assurance Manual (NQAM) requires activities affecting quality be prescribed by documented instructions and those activitics be accomp-lished in accordance with these instructions. The Toledo Edison NQAM identifies the Nuclear Facility Engineering Procedures as providing specific implementing instructi;ons for performing activities affecting quality. Nuclear Facility Engineering Procedure (NFEP) 050, Processing Surveillance Reports, requires the identification portion of a survelllance report be completed within one or two days when_a conditign adverse to quality is identified by engineering personnel.

Contrary to the above, licensee did not utilize Proced.re NFEP-050 in a timely manner in identifying two conditions adverse to quality as required by Section A of that proced-ure.

Those two conditions were:

(a) Erroneous infornation on pitch and stem diameter of Limitorque valves. This information is critical in determining the correct torque switch settings and important to the operability of the valve. The licensee was aware of the erroneous inform-ation for over a year but did not initiate a condition 8602240172 860 PDR ADOCK O PDR.

G THE TOLEDO EDISON COMPANY ED! SON PLAZA 300 MADISON AVENUE TOLEDO, OHIO 43652 JAN 2 91986 L

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-adverse to quality report'until the-Fall"of 1985.'

(b) The erroneous interpretation that the water-cooled subsystem of the control room emergency ventilation system (CREVS) is not required for operability of the CREVS. On OctoberJ11, 1985, the licensee was informed that-the both the water-

-cooled and the air-cooled subsystems were required for operability of the CREVS. However, a condition adverse-to quality report was not made until November 4, 1985 (85025-03A, B)

Response: (1)' Corrective-action'taken and results achieved..

a.

Surveillance Report:85-118 was written to' document the inaccurate dimensions'for Limitorque-operated valves which were utilized-by Torrey Pines Technology.

Subsequent to the June 9, 1985 event, the Motor Operated Valve Analysis and Test System (MOVATS) has been used to set the torque switches'on Limitorque motor-operated valves. For all Limitorque operated.

safety-related, valves with wedge seating, the torque switches will be set in-theLo'en direction to the p

maximum value that_will still preclude-valve damage and-will be set in the closed direction to the manufacturers recommended value utilizing MOVATS.

b.

On September 11, 1985, Toledo Edison received Bechtel letter.BT-15669 which clarified.the system design of the Control Room EmergencylVentilation System

-(CREVS) air-cooled:and water-cooledicondensers.

Deviation. Report (DVR); No.85-156 was written on October 4~;.1985 to'do'cumentithe concerns with the operebility requirements for'both'thetair-cooled and water-cooled subsystems.-

(2) Corrective Action taken to avoid'further noncompliance.

Nuclear Engineering Department personnel will receive specific training'on the Conditions Adverse to Quality.

reporting process, including the requirement-for promptness.

(3) Date when full compliance will be'. achieved.

Full compliance will be achieved prior'to restart from thei

' June 9,-1985 event.

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' Violation:'

Technical' Specification 3.7.2.1 LNiting' Condition forf

Operation does not allow steam generator pressure-in y

excess of-237 psig when the-secondary coolant-is'less than-

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110= degrees Fahrenheit.

If the conditions stated above are exceeded, the-pressure must be-reduced to less:than 237 psig within 30 minutes and an engineering analysis-of'the steam generator must be conducted prior to repres-surf zing the ateam generator :above 237-psig.

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Contrary to the-above,~on September 6, 1985,' No. 1 steam-

. generator-was pressurized to approximately-1050 psig,-

depressurized to below'237=psig and repressurized to.

approximately.1050 psig with secondary coolant at less l

than 110 degrees Fahrenheit. -Prior to the secend 1050 psig pressurization of the steam generator, an engineering.

analysis of the steam. generator-was not performed.

Response: (1) Corre'etive action taken and results' achieved.

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During. auxiliary steam testing of Auxiliary. Fee _dwater Pump (AFP) No. 1-1, Steam Generator 1-1 was inadvertently

. pressurized to greater than 237 psig'.with secondary coolant temperature less than 110*F in, violation'of Technical i

Specification 3.7.2.1.

The circumstances which lead to.the overpressurization are' as follows:

Auxiliary steam testing was'first performed on AFP 1 without any significant problems:noted. When-the:first'-

phase testing.of of.AFP 1-2 was complete, the Control Room was contacted to start the.second phase of'the: test, which-y involved running both feedwater pumps in parallel,'at below:

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full cpeed.

As-AFP 1--I was brought. up to - speed, th'e~

Control Room. received a Steam and(Feedwater Rupture Control I

System (SFRCS). full trip' alarm and a' Steam Generator'l-1 low pressure < alarm. Due to several other activities being L

performed on the SFRCS at the: time lwhich were causing

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expected alarms, other alarms occurring at.this' tin:e'were not evaluated and compared to the SFAS and.. Steam Ge,erator

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1-1 alarm. It'was decided.to-try to duplicate the.StRCS full trip incident by bring the pump speed down-and then_to increase the pump. speed again. This attempt produced ~the same resultsLin-that the Control' Room again. received an'.

SFRCS and Steam Generator 1-1.l low pressure < alarm, at which time the pump' speed was'again decreased by the' i'

. operator.

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. Grounds:on SFRCS in the;pastlhad caused similar type..

' alarms.. Control Room personnel, along;with Instrument.and JControl.' personnel' investigated;this possibility and repeated

'theLpump' speed change a-thirdi imei which once_again t

produced the same-alarms.;'At,this-time, a Reactor; Operator arrived in the.. Control Room to-relieve the watch.jHe noted

-a-computer: alarm on Steam Generator pressure which'was

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at approximately 800 psig and de' creasing. The~AFPs~were-shutdown and the system depressurized.

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A: review of' the data showed that on 'three occasions the Steam Generator was pressurized above'1000 psig with.the' highest being 1058.8 psig.~_ Steam Generator shell' temper-ature was.about.101*F. 'The' Technical Specifications limit.

pressure to'237 pisg when temperatures are.less than 110*F.

The pressurization was caused by'the Auxiliary Feedwater Pump's' discharge valve AF3870 being left p'rtially.open a

thereby subjecting the Steam Generator to full pump discharge pressure'of approximately 1000 psig.

An engineering analysis of the Steam Generator was not conducted before repressurizing because the operators'did not realize that.the Action' Statement of Technical Specific-ation 3.7.2.1 was entered.

The secondary side pressurization of Steam Generator 1-1 was later evaluated by the NSSS vendor, Babcock & Wilcox.

Analysis shows no adverse effect on'the Steam Generator.-

Licensee Event Report (LER)85-017, dated 0ctober 4, 1985, was submitted in accordance with-10 CFR 50.73.

The two operators involved with checking the, Auxiliary.

Feedwater Pump.(AFP) discharge valve clostre, the Control--

Room Operator and the duty Shift Supervisor who-were in charge of the test, and the~on-shift operators _were

' formally counseled to more fully analyze abnormal plant.

events with all available indications.

Modifications were made to.the Safety Tagging Procedure-AD_1803.00.and a Temporary Modification-T-9555, was written for. Administrative Proc'edure AD 1839.00, Station Operations. Those; modifications were written to empha' size the need to physically check: valves with sufficient force' to assure that theJvalve'is:in-its proper position..

Temporary Modification T-9588,-was-written'for Surveillance-Test 5071.01, Auxiliary Feedwater System, Monthly Test'.-

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This modification provides for-a vent path whenever the; test is run with Steam Generators'in a wer;1ayup condition.

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Page15 Meetings.were-conducted with Operations Department personnal to' discuss the event, event findings, and'the corrective actions taken..The Shift Supervisors andilicensedl personnel werecinstructed to-limit future activities'in'the Control Room during critical initial' stages of any special testing) i Lof plant equipment.

-(2) Co'rrective' action to be taken to avoid further noncompliance.

In addition to the corrective actions discussed'in-(3) above, the Operations. Department will review applicable procedures that have a potential for overpressurizing;the Steam Generators'to assure that-these procedures include.

provisions to provide a vent path.-ERevisions'will be made.

to the Administrative Procedures for training to ensure that.all operators.are properly. trained on valve positioning."

All. operators will receive hands-on training to demonstrate

-the requirements for physically checking Limitorque valve operators.

(3) The date when full compliance'will be achieved.

Procedure review and revision', along with training..wille be completed prior to restart from the current = outage.

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.-Violation)

' h Technical-- Spesification 6.8.1' requires ' that ' written.

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' procedures be established, implemented and maintained;

,. covering the activities specified in-Appendix A of..

Regulatory Gu'fde 1.33, November, < 1972. Section H of'

• Regulatory Guide 1.33, Procedures and Calibrations,-

, states in part: 9 'Equipuent 'to be - calibrated -by appropriate i

procedures are readour instruments,vinterlock permissive

.and prohibit' circuits, alarm devices, sensors,-signal, '

conditioners, controls and protective circuits."

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. Contrary to'the above, an~ instrument channel inputting--

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to the secondary heat balance and performing.a' Technical Specification requirement was calibrated without the use of a procedure.

(85025-22)-.

Response: -(1).Correctiv'e action taken'and'results achieved.

A generic procedure,. Instrument Strin'g Check-Calibration, cf a

IC 4001.00, was approved on December 3, 1985. This procedura-became the controllingMocument' for checking. the calibratioft

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i of all nuclear safety related and non-nuclear safety.

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w (2) -Corrective action to be taken;to avoid further noncompliance.-

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Toledo Edison has initiated a major procedure writing i

i effort including Instrumentation and Control ~ procedures.

In addition, Toledo Edison has initiated the development 4

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of data packages which v4.11 contain specific informa'thon' l

such as range, tolerance, etc.'for each instrument:or b

control device.

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'6 (3) The date when full compliance'will be achieved.

i Full compliance was achieved on-December 3,:1985.

s Very truly yours, pa l.UN ^ A -.s )r fi u

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cc: lDB-1 NRC Resident Inspector F

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