ML20236N191

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Summary of 871105 Meeting W/Util in Bethesda,Md to Discuss Recent Instances of Water Intrusion Into Instrument Air Sys at Facility.List of Attendees & Presentation Handouts Encl
ML20236N191
Person / Time
Site: Fort Calhoun Omaha Public Power District icon.png
Issue date: 11/09/1987
From: Bournia A
Office of Nuclear Reactor Regulation
To:
Office of Nuclear Reactor Regulation
References
NUDOCS 8711160054
Download: ML20236N191 (62)
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f Iog UNITED STATES y -

g NUCLE AR REGULATORY COMMISSION '

g 'y WASHINGTON. O. C 20555 November 9,1987

/ l l DocTeYNo.50-285 l

l LICENSEE: Omaha Public Power District  ;

l l FACILITY: Fort Calhoun Station, Unit 1

SUBJECT:

MEETING

SUMMARY

OF MEETING HELD ON NOVEMBER 5, 1987 IN BETHESDA, MARYLAND A meeting was held in Bethesda, Maryland on November 5,1987 between Omaha Public Power District (0 PPD) and the NRC staff to discuss the recent instances of water intrusion into the instrument air system at Ft. Calhoun. This meeting followed an enforcement conference held at Region IV Headquarters in Arlington, Texas on October 29, 1987. A list of attendees and their affilia-tion is provided as Enclosure 1.

l The purpose of the meeting was to give OPPD the opportunity to provide the l staff with information that would allow the staff to determine the safety j significance and subsequent appropriate course of action regarding water in the instrument air system. On two separate occasions, July 6 and August 25, 1987, OPPD experienced water intrusion events in the instrument air system at their Fort Calhoun Station, Unit 1, one of which led to failure of their diesel l

generator No. 2 on September 23, 1987. It was expected that sufficient techni-cal detail would be presented to allow staff evaluation of the following topics: chronology of events; immediate response; design of related systems and components; root cause determination; safety implication; engineering evalua-tion; and corrective actions (both short term and long term, including proposed

! programmatic changes and lessons learned).

The licensee provided the staff with a presentation c. hat touched on many of the topics listed above. Enclosure 2 contains the hindouts at the meeting that were also provided to the staff. It is the staff's judgement that OPPD failed to present sufficient information to detertaine that OPPD adequately

! understood the details, scope, or safety significance of the problems or had provided an adequate response concerning the resolution of the problem over the short and long term. The staff concluded that it was appropriate to allow OPPD an additional week in which to assemble the information needed to address the staffs' technical, programmatic and safety concerns. A follow-up meeting, to be held in Bethesda was scheduled for November 13, 1987. The staff's judgement that it was appropriate to allow an additional week to assemble the required information is based on several factors:

1) Immediate actions by OPPD to remove the water from the system were successful, i 2) OPPD has modified the interfaces between the instrumt air system and sources of water, either by removing the connectien or closing the related valves and tagging them out.

8711160054 871109 PDR ADOCK 05000285 P PDR

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3) The plant's Technical Specifications are being met. )

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4) Region IV has recently inspected the surveillance tests and -

corrective action on the instrument air system. j

5) OPPD has presented some programmatic plans to address this concern including: design basis reconciliation; a comprehensive test program; .j improved procedures; and an enhanced preventive maintenance program.

The staff also concluded that the issue of seismic qualification of certain ,

components of the instrument air system is a more long-term concern and it ]

would be appropriate to pursue resolution on a separate, if not parallel, path.

, Finally, OPPD was informed that if the plant went down for any reason'in the future, that, as a minimum, testing of all accumulator check valves and I actuator valves on safety-related equipment and components associated with the j instrument air system that can not be tested while at power will be accomplished.  !

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Anthony Bournia, Project Manager Project Directorate - IV l Division of Reactor Projects - III, IV, V and Special Projects Office of Nuclear Reactor Regulation

Enclosures:

f As stated cc w/ enclosures:

See next page ,

I DISTRIBUTION.. . .

Docket File; NRC PDR Local PDR PD4 Reading ,

J. "Cahlo' A. Bournia 0GC-Bethesda E. Jordan  !

! J. Partlow D. Crutchfield J. E. Richardson J. W. Craig  :

l L. B. Marsh V. Hodge S. R. Stein A. Lee l J. N. Ridgely R. Kirkwood L. J. Callan T. F. Westerman  !

T. O. Martin J. H Wilson ACRS (10) l l

l I PD4/P Ad PD4/D/H6 ABourn i(sy JCalvo l

11/7/87 6

11/4/87 i

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, . .1 Fort Calhoun Station- 1 iMr.!R. L. Andrews-Omaha Public Power District ' Unit ~No '1. -

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.CC Harry H. Voigt, Esq. .'

LeBoeuf, Lamb, Leiby &'MacRae i 1333'New Hampshire. Avenue, NW Washington, D.C.- 20036 .

Mr. ' Jack' Jensen, Chairman '

. J Washington County Board ,

of Supervisors J Blair, Nebraska '68008: ,

q Mr. Phillip Harrell, Resident Inspector: j U.S. Nuclear Regulatory Commission o P; 0.: Box 309 .

Fort Calhoun, Nebraska 68023 '

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Mr. Charles B. Brinkman, Manager..

Washington Nuclear Operations 4 1

C-E Power Systems 7910 Woodmont Avenue Bethesda, Maryland '20814 1

Regional Administrator, Region IV .

I U.S. Nuclear Regulatory Commission 1 i

Office of Executive Director for Operations i n a h1 Harold Borchert, Director Division of Radiological Health Department of Health 301 Centennial Mall, South .)'

P.O. Box 95007 Lincoln, Nebraska 68509 j W. G. Gates, Manager.

Fort Calhoun Station  ;

P. O. Box 399 1 Fort Calhoun, Nebraska 68023 H

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Enclosure 1-List of Attendees at Meeting Between Staff and OPPD'

.1 Held in Bethesda, Maryland on November 5, 1987 NRC D. M. Crutchfield J. A. Calvo i J. E. Richardson 4 J. W. Craig L. B. Marsh V. Hodge S. R. Stein A. Lee J. N. Ridgely i R. Kirkwood j L. J. Callan 1 J

T. F. Westerman T. O. Martin J. H. Wilson s

i OPPD l i

R. L. Andrews  !

J. J. Fisicaro T. J. McIvor l S. K. Gambhir M. E. Eidem l

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Enclosure 2 NOVEMBER 5, 1987 OPENING REMARKS l I. Background /0verview

-Description of Instrument Air System

-Design Description /VSAR

-Fire Protection Modification Installed 1985

-Standard Review Plan Comparison II. July 6,1987 Fire Protection Surveillance Test Performed

-Describe Surveillance Test / Maintenance Procedure

-Sequence of Events

-Investigation'(whathappened)

-What immediate actions taken

-What long term actions taken/ planned

-Results of actions

-Conclusions / Adequacy of corrective actions III. August 25, 1987 DW-CV-86

-Sequence of Events

-Results of Actions

-Mitigating circumstances

-Conclusions IV. September 23,1987 DG-2 Survei'llance Test

-Tech. Spec. Section 2.7

-Diesel Generator Testing 7/6/87 - 9/22/87

-Sequence of Events

-Investigation

-What immediate actions taken

-Results of actions Mitigating circumstances

-Conclusions

-Justification for continued operations V. AE0D/IEN Review VI. Communications with NRC

-Day to Day

-Reportable item.s VII. Lessons learned

-Operation and Maintenance Perspective

-Generic Implications VIII. Lessons Learned

-Engineering Perspective

-Generic Implications

SUMMARY

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Starting in 1979,,in response to NRC IE Bulletins numbered 79-02,! '

V} 79-04, 79-07, 79-14, and 81-01 and NRC Generic letterh No. 81-14, seismic verification of all Critical Quality Element (CQt) piping system defined in the fort Calhoun Station's CQE Itst (except those portion ,3 j

system that were inaccessible), was initiated. i t d to cation process consisted of identifying where adequate documentatio demonstrate the seismic qualification 1 with aof each CQ newer, reanalyzed using the stress criteria identified in Section ibed. The F.2. net tems more sophisticated dynamic. loading model than previously descr result of the reverification program has been thst insome cases, piJI f were determined to be in the resonance group; these were modified by i tl stops and dampers to adequately prevent large a ' j and 18 bz vertical by adding addi.tional restraints.

Special.s sefsmic restraints -(either rigid or snubbers) were provided control vain mechanisins to. preven'tgeverstre.ss when the control.  ;

a. mass center outside .the'ptpe center line and.p- gererates ove'r]

~ strest o'n the piping system due t'o earthquake G loading, The waste disposal tanks are fitted with horizontal restraints at t

, to avoid overloading the equipment support under seism upper and lower a loading.  ;

Special seismic restraints were (nstalled on the electrical cable x The cable trays were supported vertically and horizontally h ks so as to m p ) trays.

the stress criteria under all conditions including the postulated eart qua e .

Q Spacing.of vertical suppcrts is sufficiently close to maintain th cal dominant natural frequent,y;of theWhere cablemultiple traystray above arran'ge- 18 hz , wh the dominant . natural frequency of the building. has ments make determination of the natural frequencies unfeasible the system The 3 been analyzed for a resonant condition, using a time-history' approach.- l '

cable trays are braced horizontally at a spacing to ensure a-minirnu frequency of 6 hz., double the' dominant' horizontal n building.

i within the allowable limits, In order to avoid resonance between any.of the elements of the' rea coolant system and the reactor building, additional seismic restraints Four are ,

installed at the steam generators and at the reactor coolut' pumps.

columns are installed 3 at each steam . generator between the und

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support frame and the foundation mat. Similar columns are provided to permit thennal movement.of the generators. In a6ditibn,4 ort-unde'r'.thefthree support .pa,ds'.cf.,each' react,or coolant pump. '

zontalshydr.aulic.sngbbeys.are #

installed This system ofat'appr.oximately restraints. stiff.pnsyth.e the vertic gravity, of.each reactor. c6clast guinp.and moves the 1dwest dominant frequet)cy of' entire reactor coolaht'sysp(tor coolant pumps e peak away o frord .th.ha ,

the steam ge'neratorsl:ahdReacable flooe response spectrb >

reactor coolant system are within alicwable limits.

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i i EXAMPLES OF CHECK VALVES USED FOR ISOLATION ANSI 51.1 1983 112.1 SAFETY CLASS INITRFACES FOR i

PAR PRESSURE INTBGR11Y OF FLURD SYSTEMS  !

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The Reliabte Model O Ory Pipe Valve in its closed and open specifies that Oulck Opening Devices are required lo ory ys. }

tems having capccities of more than 500 gallons.

positens is shown in Figures 1 and 2. The closed position is rnaintained as long as the air pressure in the system )lping Reliable's Model B Accelerator is an approved Oulck Opening i above the Ory Pipe Vabe is sufficient to exert a gfeater force Device thafis eastty attacited to Rebeble4 liry P6pe Valves.

on the top side of the clapper than is exerted on its underside The accelerator is sertsitive to small drops in system pressure by the pressure of the water supply Since the area included by and its quick operation allows the system air to pass through the or Seat is approximately six times that of the Water Seat to the alarm outlet and the underside of the clapper. Here the 3 (The Differential Principle) the air pressure neoded to keep the air pressure build-up, coupled,with the pressure of the water Ory Pipe Valve closed is only a fraction of that of the water supply, opens the Dry Pipe Valve clapper permitting the water Suppey. Table 1 gives the recommended air pressure for various to flow instantly to the operated ;prinkler(s). (. i water supply pressures for both sizee of Model O Ory Pipe iM r Mil h Vatves. separate bulletins.

Watkr Pressu'e Air Pressure To be Purnped into System VALVE DESCRIPT)ON I sn Supply Une 1. Rated working pressure 175 psl(12.1 bar)

Maximum Not less than Not more than 2. Factory hydrostaticlest pressure 350 psl(24.1 bar) 3 End and trim connections-Three valve connection styles are available.

a. US Standard Flanged inlet and Outlet 75 20 30 25 35

100 125 30 40 .

150 35 45 U.S. Ft.ANGE DIMENSIONS IN INCHES Valve Bolt Bolt Flange Flange No.

Table 1 Size Circle Hole Outside Thickness ' Bolts Dia. Dia. Dia. ,

When one or more Automatic Sprinklers operate due to fire, the w pressure in the system piping above the Dry Pipe Valve 4 7% g g isp, ;g l

Is reduced. The presstare of the water supply, now by exerting 6 9%  % 11 1 8 the greater force,,alses.and r totates'the Clapper to its.open -

position, permitting the water to flow .to the' operated SprirF

  • kierts). Water piso flodrs 'from the Atarrn Outlet of theDry Pipe '. .W P "9.* E"#".? B 2A
  • Reilable's standard trim. sets are compatible with US Vatre througt) a strainer to the'Rellebte Mechanical Sprinkler Flatiged Valees..

Alarm and Elec'tric Alarm Switch to autoniatic'atly actuate mechanical and electrical alarms,

' Color-Light Gray

b. US Standard Flanged inlet and Grooved Outlet (Fig. 3)

Reh.able Model O Ory Pipe Valve can be reset quickly and easity without the use of special tools.

  • Inlet flange mates with ANSI B 161(125 tb.) Flange OUKK OPENING DEVICES U.S. GROOVE DIMENSIONS IN INCHES i When an automatic sprinkler operates, the resulting air pres. . Valve . Outlet Groove Groove Outlet Face (A, sure decay is inversely proportional to the size of the Dry Size - Ok. Ola. ' Width ' - to Groove System.the larger the system, the slower the alt pressure 4 4.500 4.334 3tg Sfs decays which slows the operation of the Dry Pipe Valve. NFPA 6 '6.625 6.455 3/8 SIB Pamphlet No.13 titled " Installation of Sprinkler Systems"

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11. METRIC CONVERSIONS ' Available m SubAssembhes only

! 1/16' = 1.59 mm 35 psi = 241,0 kPa 508.A55tM5 Lits 1/4" = 6 35 mm 40 psi = 276.0 kPa l PARY No.

I 5 ps. = 34.5 kPa 45 psi = 310,0 kPa 50 psi = 345.0 kPa ITIM NO. DESCRIPTION k t Q'D.

10 psi = 68,9 kPa 15 psi = 103,0 kPa 75 psi = 517,0 kPa 10.11.13 20 psi = 138,0 kPa 100 psi = 689,0 kPa 150 psi = 1034,0 kPa eru 21 1,2,4 02269B 02272A Housing Assembly Adjustment Screw Assby.

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25 psi = 172.0 kPa 8&9 01792A Onphragm Assembly 1 30 psi = 207,0 kPa 175 psi = 1206,0 kPa filter Kit 19 & 21 0%07A 1 Part No 04909 F Printedin U 5 A.

Form No 2114 l /

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FIRE PROTECTION M00lffCATIONS USING INSTRUMENT AIR DRY PIPE SPRINKLER SYSTEM FOR DIESEL GENERATOR SQGNS (let-FC-83-1(?).

SYSTEM INSTAttED To REDUCE THE POTENTIAL FOR FREEZHIG SPRINKLER PIPING WEN THE DIESEL GENERATORS ARE OPERATED DURING COLD.WEATMER. 1 SYSTEM WTILIZES A 4-INCM DRY PIPE VALM THAT TEEDS B0ld D.G. ROOMS VIA C0m001 K ADER. ALL SYSTEN PIPNIG .I.S PRESSURIND TO APPROXIMATELY 44 PSI WITN INSTRIMENI AIR. is4EW AHt PAESSURE IN MIE SYSTEM IS RE00CED TO A PREDETERMINES POIST, THE DRY PLPE VALW OPENS ALLOW 11eG TWE PIPING SYSTEM TO FItt- SITN WATER.

INSTRlNENT AIR USE0 F8R SEVERAL REASONS:

MINIM 1ZE A00iTIONAL COMP 0NENTS THAT REQUIRE PERIODIC MAINTENANCE (ADDING DEDICATED COMPRESSOR WOULD ADD ADDITIONAL MAINTENANCE).

PROVIDE A CLEAN, DRY SOURCE OF COMPRESSED AIR, THUS MINIMIZING THE POTENTIAL FOR MOISTURE COLLECTION AND POSSIBLE FREEZING WITHIN THE.

SYSTEM.

THE FOLLOWING CHECK MECHANISMS PROVIDE ASSURANCE AGAINST WATER IN THE INSTRUNENT AIR SYSTEM ,

1) DRY PIPE CLAPPER
2) CHECK VALVES IA-575 & IA-576
3) AIR MAINTENANCE DEVICE IA-6 THE DRY PIPE VALVE, IF THE CLAPPER IS RESET WILL NOT ALLOW WATER UP TO 1A 575, THEREFORE, THE ONLY TIM.E WATER WILL.BE AT IA.576.IS DURING SURVEILLANCE TESTING ~ AND AN ACTUAL FIRE CONDITIO'N.

VEND 0R RECOMMENDS USE OF ONLY ONE CHECK VALVE.

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CQE 1A SYSTEM STATUS (34 VALVES )

NOTES:

I

1. THESE-VALVES WERE FUNCTIONALLY TESTED, UNDER LAL 860269, DUR.ING THE 1986 OUTAGE. THEY ALL l PERFORMED AS REQUIRED.

i 1

2. THESE VALVES WILL BE SEISMICALLY EVALUATED UNDER~

LAL 860269.

3. THE SIZE OF THE ACCUMULATION WILL BE EVALUATED AND CONFIRMED BY FUNCTIONAL TESTING UNDER LAL 860269.

THE DESIGN ~ BASIS FOR EACH MUST FIRST BE DETERMINED.

4. COMPLETION OF LAL 860269 IS EXPECTED AT THE END i 0F THE NEXT REFUELING OUTAGE OF SUFFICIENT DURATION TO PERFORM NECESSARY TESTING'AND/OR REQUIRED MODIFICATIONS.

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- J CQE 1A SYSTEM STATUS (34 VALVES )

ACC.U MULATOR/ PIPING DESIGN  :

DESCRIPTION VALVE TESTEDJCD REMARKS l

SEISMIC 1 YCV-1045A MS TO AFWP TURBINE ES YES MR-FC-83-158-HCV-304 HPSI HDR ISOL SEE NOTE 2 YES SEE NOTE.1 l HCV-385 SIRWT RECIRC ISO YES YES MR-FC-87-24 HCV-400A CCWTO. CONT HVAC VA-1A SEE NOTE 2, ,SEE, NOTE 3 YES SEE NOTE 3 l

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s CQE IA SYSTEM ',rATUS (34 VALVES)  !

l Accumulator / Piping Design

. i Valve Description Seismic Tested JC0 Remarks ]

HCV-4010 CCW To Cont HVAC VA-1B See Note 2= See Note 3 Yes- d )

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HCV-402A CCW To Cont HVAC VA-8A See Note 2 See Note 3 Yes.

HCV-402B CCW To Cont HVAC VA-8A See Note 2 See Note 3- Yes HCV-402C CCW To Cont HVAC VA-8A See Note 2 See Note 3 Yes OK-UNTILL HCV-4020 CCW To Cont HVAC VA-8A See Note 2 See Note 3' Yes completion HCV-403A CCW To Cont HVAC VA-8B See Note 2 See Note 3 Yes of Program, q HCV-403B CCW To Cont HVAC VA-8B See Note 2 See Note 3 Yes . CAL 860269, HCV-403C CCW To Cont HVAC VA-8B See Note 2 See Note 3 'Yes See Note 3 i

HCV-4030 CCW To Cont HVAC VA-8B See Note 2 See Note 3 Yes j LCV-383-1 SIRWT Hdr. Iso. See Note 2 See Note 3- Yes LCV-383-2 SIRWT Hdr. Iso. See Note 2 See Note 3 Yes

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HCV-238 Loop Injection from CH.P. See Note 2 Yes See Note 1 I

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l CQE IA SYSTEN STATUS (34 VALVES)

.4 Accumulator / Piping Design j Valve Description Seismic Tested JC0 Remarks

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l HCV-240 Aux. Spray Iso. Viv. T#&2 Yes See Note 1

]

'YCV-1045A NS to AFWP Turbine Yes Yesw M-FC-83-158 YCV-10458 MS to AFWP Turbine Yes Yes~ ' M-FC-83-158 ]

HCV-239 Loop Injection from CH.P. See Note 2 Yes See Note 1 HCV-712A SFP Char. F1tr. Bypass See Note 2 Yes See Note 1 -

1 See Note 1 I NCV-304 HPSI 2R Isol See Hote 2 Yes N0V-305 1951 2R Isol See Note 2' Yes See Note 1 i HCV-306 HPSI HDR ! sol See Note 2 Yes See Note 1-HCV-307 HPSI HDR Isol See Note 2 Yes M-FC-87-24 HCV-385 SIRVT Rectre Iso Yes "es NR-FC-87-24 HCV-386 SIRWT Recirc Iso Yes Yes NR-FC-87-24 HCV-864 SFP Char Fitr Spray See Note 2 See Note 3 Yes n l

1 HCV-865 SFP Char Fitr Spray See Note 2 See Note 3 Yes HCV-344 Cont. Spry Hdr Iso See Note 2 See Note 3 Yes HCV-345 Cont. Spry Ndr Iso See Note 2 See Note 3 'Yes OK-UNTILL ,

HCV-400A CCW To Cont NVAC VA-1A See pote 2 See Note 3- Yes Completion HCV-4006 CCW To Cont HVAC VA-1A 'See Hote 2 See Note 3 Yes of Program,-

HCV-400C CCW To Cont HVAC VA-1A See Note 2 See Note 3 Yes . CAL 860269, HCV-4000 CCW To Cont HVAC VA-1A See Note 2 See Note 3 Yes See Note 3 HCV-401A CCW To Cont HVAC VA-'1B See Note 2 See Note 3- Yes HCV-401B CCW To Cont HVAC VA-1B See Note 2 See Note 3 Yes U

HCV-401C CCW To cont HVAC VA-1B See Note 2 See Note 3 Yes f

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.i July 6, 1987, Water in Instrument Air Event A. Instrument Air System

1. Important to station operation
2. Historically - only minor problems; a few instances of plugged orifices, etc.

B. ST-FP Includes Test Actuation of Dry Pipe Sprinkler System for DG Rooms

1. Fire protection water supply is screened river water
2. FP header. pressure maintained at approximately 125 psig by jockey pump, using service water 1 3. Dry pipe system normally maintained at approximately 40 1 psig air pressure from IA system i
4. Test actuation initiated by bleeding down the air pressure

,! 5. Following test actuation, dry pipe valve is reset by 4 performing MP-FP-7 a

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C. Sequence of Events - Based on Discussions with Personnel  ;

Involved-

1. Test initiated-per ST-FP-5
2. ST says to reset per applicable steps of MP-FP-7; not. clear'  !

as to where.to begin-

3. _ClosedLFP-514, but did not' reset clapper
4. Entered reset procedure at point of priming FP-513.E Operator thought.that.FP-513 was like the pre-actionJdeluge valves; self-reset when~ pressure re=oved.- Actually, 1' one-of-a-kind dry pipe valve. requiring manual l reset.

4 5. Opened FP-516 and poured priming water into body'of' dry.

, pipe valve

[j 6. Closed IA-570; isolated air 1 maintenance device ,

I j 7. Opened IA-569; noted water blowing out of FP-516 l

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" 8.- Opened-FP-514; noted that both-air-and water pressure

(.: gauges'on valve were at' fire main pressure; knew this was J abnormal I

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9. Opened IA-574; stream of water; knew at this point that there.may be water in the IA' system'
j. 10. Icolated system; discussed with ' shift supervisor; MO

~

written to perform maintenance cn1 check valves and reset I

, dry pipe valve  !

11. Some component problems over the next hour pointed'to water i in the IA system; FC-269X, bubbler for DG fuel: oil tank,

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0. Evaluated Cause of Event }

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1. Check valves IA-575 and IA-576 were prevented from closing i fully by foreign material

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2. When FP-514 was opened, water entered the IA system because the clapper in FP 513 was not reset, IA-569 was open and the water pressure was approximately 30 psi greater than the air pressure
3. Procedural inadequacies were a major contributing factor E. Immediate Actions
1. Performed maintenance on check valves IA-575 and IA-576; restored j

\

2. Reset the dry pipe valve, FP-513 j l
3. Began blowdowns of IA system in the vicinity of FP-513, j extending into areas where we discovered water and/or l operational problems
4. By the end of the day, we felt that substantially all water had been removed from the system and full operability  ;

restored '

5. Gegan an engineering investigation of the problem; scope of i blowdowns required, etc.  !

F. Blowdown Program - Determine Extent of Water Intrusion and Document Its Removal

1. Began drawing review Monday PM, July 6 l
2. Walkdowns performed Tuesday, July 7, to determine possible paths of water, etc.
3. Procedure and data sheets prepared Wednesday, July 8 l
4. I&C began detailed blowdowns and component cycling on
Thursday, July 9

)

i G. Phase I - July 1987

1. 515 components blown down; valves cycled and components actuated except where not possible during operation
2. 49 components from 11 risers had water

?

3. Water confined to lower two levels of the auxiliary building; none in turbine building or intake structure; did j not reach level of IA penetration in containment

4 H. Phase II - August 1987 , ,

1. 11 risers from Phase I blown down complete;7
2. 4 locations on 4 risers showed mist I. Phase III - September 1987 z
1. Planned to blow down the 4 risers showing mitt in August
2. Data sheets issued to maintenance but work not' completed when DG-2 event of September occurred 1

J. Air Accumulators  ;

1. Researched P&ID's to determine which CQE valves in af fected l areas are equipped with accumulators
2. Drained or blew down accumulators; one had 12.5% water, three others had a small amount; all in lower level of auxiliary building
3. DG exhaust. damper accumulators were missed because they are  ;

not shown on the P&ID; different type of accumulator application

4. I&C technician who blew down .IX3 headers believed tnat his actions blew down the accumulators as well; this was later determined to not he the case K. Procedural Inadequacies - Procedure Changes Have Been Prepared to Correct L. Long Term Actions Planned as a Result of the July 6 Event (Taken from Report to PRC Dated August 3, 1987)
1. EEAR FC-87-32 written to reevaluate / remove IA from FP-513
2. Plan developed to disassemble sample of valves and instruments and inspect for adverse effects of moisture intrusion l

l 3. Consider cleaning / system flush during 1988 refueling outage l

1

M. Summary

1. Focus - restoration of system to full operability as rapidly as possible
2. Plant conditions maintained as normal as possible during this process
3. Subsequent re-evaluation has concluded that the event should have been reported 1

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i Water-Plant Event Auguist 25, 1987'  :

j A. Discovery and Immediate Actions

1. Sluggish valve operations
2. Blew down components'en upper level-of water plant - DW trains, sand and carbon filters,.etc.
3. Traced down possible' sources of water - suspect air line to

, clarifier booster pump recirculation valve-

4. This' air connection has been isolated modification is under consideration 1t .i j B. Mitigating Circumstances  ;
4. 1. Localized problem on non-safety-related equipment '
2. Check valve IA-3046 prevents backflow'into turbine buil' ding lj ring header
3. Air receiver in water-plant was blown down l lI 4. Physically and functionally removed from-safety' systems

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l C. Conclusions i

l. Event is not related to July or. September events.

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2. Further followup is required - factor in ' water . plant" refurbishment i

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'DG-2. Failure' Sequence of' Events  ;

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' September 22, 1987 DG-1 started to, verify operability prior.to. fj commencing DG-2. maintenance - '

q September 23, 1987 0906 -' began'ST-ESF-6 on DG-2Lto' meet Tech. ,

. Spec. requirements: for monthly full: load test'

)

1 0911 - 5 minutes .af ter. startup, L commenced '

' loading.DG t 0920 - 14 minutes after starting, DG-2 t shutdown on high cooling temperature l Initiated-immediate investigation'that-  ;

included call to EMD .!)

y About three' hours after trip,.the damper?to ^

the radiator failing to-full open'was thought-to be most probable.cause 'I

-1610 - after c'onfirmingldamperJproblem on i DG-2, blocked open damper.on DG-l' L Commenced detailed investigation i

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1 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _. )

+

Root Csu'se for DG-2-Failure:

~he most probable root'cause for'the DG-2 exhaust radiator.dSmper failure was the presence of water or moisture in the instrument air lines / valve actuator. The inadequacy of the blowdown of the.IA" lines performed in July and again in August is considered'a primary contributor to the eventual failure of the damper.

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Corr'ective Measures 1

I Reinitiated Phase I blowdown / cycling program (538 components to j date) .!

Original scope plus 4 valves from containment and new components identified during walkdown Water / moisture found only in HCV-871F and the PASS (12 oz.

and 1 qt.,.respectively)

Performed JCO for CQE valves not cycled during any of the three phases (29 valves total)

Scope included only CQE valves in the scope of the original l test program that could not be cycled to verify operability l>

It was assumed that valves in containment, turbine building, I intake structure or above the 1025' level were operable f 20 of 29 raw water interface valves j No credit is taken for the 9 remaining valves to perform a I 1

j safety function following any design basis accident )

a i i f Isolated IA from wetted / pressurized tie-ins j j  !

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3 locations isolated; one by jumpering. to PA, the other two )

l' by tagging shut associated isolation' valves  !

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Repaired YCV-871E and YCV-871F

l. -

Primarily a cleaning effort j J

Commenced walkdown of IA system to verify completeness of  ;

blowdown / cycling program Walked down every line in the IA system (except containment)

Purpose was to verify that any components attached to IA were l identified and to find all low points that may not have been '

blown down 1

i Found no new accumulators or connections to IA or pressurized l wetted systems 1 Found a number of components (50-60) that were not previously .

in the P&ID's 1 Letter issued to GSE to correct P&ID's  !

l l

9

Initiated dew point sampling program - initial results indicate need to replace desiccant First dew points about 10/5/87 Initial dew points were 10-20*F. atmospheric (instrument problem suspected)

Desiccant replaced 10/30/87 Latest dew point is -60*F. pressurized-Developing program for inspecting / verifying operability:of valves that depend on accumulators Procedures are near completion for performing accumulator

' verification of 11 of 18 valves that need air to perform DBA function i

l Reviewed IST program stroke time results/ increased frequency of j testing to monthly Review of IST stroke time testing to date shows no significant change (two full cycles now complete) 4 To improve quality of data base, the quarterly IST program l will be conducted monthly I

i Commenced development of stroke time monitoring program for j

g valves not included in the IST program (11/15/87) f -

A separate stroke time testing program for 38 additional valves that includes all valves known to be wetted and can be

( cycled at power plus an additional 15 randomly selected 5

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These valves will also be tested monthly 4

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  • Initiated program to verify air operated valve design bases

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Basic charter is to look at every valve that relies on IA for i any function to ensure we have a clear understanding of valve function Expediting modification to permanently separate IA from pressurized wetted systems The two connections to FP system will be removed and a separate air maintenance device will be installed DW connection is tagged closed

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1 Complete rewrite of AOP-17.from a' technical.and human factors 1

-perspective j

AOP-17 revised in early October to eliminate any steps /

i statements that could be confusing. JAOP-17 will be further j revised from a human factors : standpoint to reflect the guidance in our new Procedures Writers Guide.

, }j

  • I Will evaluate existing PM program for IA components and revise as j needed to improve confidence on operability' 11 u \

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Consultant ~has.been brought in to evaluate.the various-

?! aspects of operation and maintenance of the IA system and I

develop recommendations for' improved operation, monitoring j

and maintenance )

Continue to investigate methods for determining the presence ~ of

, remaining water a

1 I -

Evaluating potential for using NDT methods to conclusively  !

prove no moisture exists at IA low points and accumulators }

i that could-not be blown down j 4

Mo is in progress to make test assemblies that will'then be l taken to an NDT test lab to validate our ability. to test for water l

Issued Mo's to disassemble and. inspect two valves known to have-been exposed to water (waiting on parts)

Two non-CQE valves with operators ,significantly wetted are being disassembled to assess.the potential of fouling from water intrusion Spare parts have been received, hope to have this task

, complete 6 by week's end Will expand this program if inspection finds significant residue 2@1\

Increased evaluation of potentially reportable events

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Revised internal event report system E

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Deportability analysis training for STA's 3

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-- . Serious, sensitive sit'uat' ion .:

- Extensive actions have:been taken to restore' systems to fulli

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- Additional? actions are in progress to prevent-recurrence-ofc .;

i this'or similar situation i

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p Review of AEOD R'e'ortion Air System Problems y

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.l A. 'Done by offsite technical support organization. 'Same time frame '

as recovery efforts from July 6 event.

l 'l B. Evaluation y i- l

1. Each recommendation of report evaluated'for applicability to i Fort Calhoun..

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! 2. Major conclusion - need to perform regular periodic dew' point j l- sampling

3. Weakness - concluded procedures were adequate .

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1 l C. Subsequent actions

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1. Revised AOP-17, Loss of Instrument Air, to resolve errors / ' {

inconsistencies 1

2. Further review / upgrade of procedure is in progress 1
3. Retained c5nsultant for system. evaluation- 'J

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4. Dew point sampling initiated' l
5. Further evaluation in progress c i

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C0tHJNICATION WITH NRC

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I

-3 DAY TO DAY CON 1UNICAT10NS ilITH NRC-

-1 FREQUENT CONVERSATIONS WITH RESIDENT' INSPECTOR ~ .

]

-Plant Manager, Department Heads, Supervisors,sothers.

-Assignment of Direct Site Licensing Person Interface

' 1 COMPLETELY OPEN PHILOSOPHY a

Problems l

Future Plans ]

Normal Operation-FREQUENT CONVERSATIONS WITH NRC/ REGION IV

-Project Manager

-Region IV Personnel  ;

-Regional Vtility Group .;

-Meetings between President /Vice President - Regional Administrator

-Meeting in Future to Rev'iew Planned Programs

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-Detailed Procedure

-Lower Levels Report Problems.to. Supervisory Personne1'

-Plant Review Committee. Involved in Determination. q

-Offsite Personnel Frequently Involved  :

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  • POLICY  !

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-Comply with Requirement-

-Discuss Issues that are Borderline with~ Resident / Project Manager

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.i COMINICATION PROBLDtS ASSOCIATED Willi I.A. EVENTS

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' HANDLED AS AN OPERATION PROBLEM

'NON-INVOLVEMENT 0F 0FFSITELPEOPLE.

' CONTACT WITH RESIDENT NOT UP.-TO NORMAL STANDARDS  :

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CORRECTIVE ACTIONS l

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  • INCIDENT REPORTING PROGRAM

-New Computerized Reporting Mechanism Developed

-Computerized Trending 1

-Computerized Tracking-

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-Review other Depar: ents for Like Reporting: Areas 1

  • INVESTIGATION

-STA's are Assigned Task

-Timely l

-Accuracy

-More Detailed j

' REPORTING

-Independent Review

-Increased Training of Personnel

-Continued Involvement of Offsite Personnel l

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c_______-_-_-_---__--___________- _ _ _ _ _ _ _ _ _ _ _ - _ _ _ - - _ _ _ . _ _ - . - _ _ _ - _ ~

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, 1. The instranent air system will be installed and maintained at or ahve the

. USAR standard.

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2. _ Modifications to plant systees will .take into consideraties the potential for introducing' common mode failures. This will include those introduced through a combination of operator errors, procedural errors'and equipment

- . failures. -

1

3. Events at Fort Calhoun Station will be evaluated promptly for safety l significance and deportability and will be promptly communicated within l OPPD and the NRC.
4. The design besis of the instrument air systes will reflect the proper safety implications and will be analyzed for the installed configuration-to assure it is not detrimental to plant operation.
5. The safety conscience of OPPD personnel will be broadened.
6. Components and systems that are relied upon to perform safety functions are
included in ongoing test programs.

l 7. Procedures shall be clear an concise to preclude the possibility _of an operator error.

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LESSONS LEARNED

- ENGINEERING PERSPECTIVE

- GENERIC ISSUES l

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1 i NOVENBER 5, 198/

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Performance Analysis ,

When the problem being dealt with involves the performance of a component or system, it is desirable to provide a l I

comparison of the relative performance of the original installation tp that of the proposed modification.

Alternately, it may be appropriate to identify expected failure rates of the components being compared.

In replacement of a component, NPRDS data is available on l l the relative failure rates of various corqponents. Where this 1

data is not available, the predominate failure mechanisms of the original components should be identified and contrasted with the design features of the replacement component.

In cornparing the impact of a component failure between pre-

& post modification configurations, the following question should be addressed where appropriate.

o How could the modified configuration / component fail that is dif ferent from the original configuration?

l o What mechanism would produce this failure & what is the probability of it happening?

o What is the worst case outcome of the failure?

o Is the failure in a safe or unsafe direction?

I o How would the failure be detected?

o What inherent provisions in the design to compensate for the failure?

When the modification involves a system type change to any CQE systems or other major systems impacting plant availability, a relativistic Probabilistic Risk Assessment (PRA) calculation can be prepared. This includes modeling of GEG-3.5 Rev.0 10/87

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ENGINEERING GUIDELINES GEG__3_

CONTENTS OF MODIFICATION PACKAGE .

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1.0 Document Control 1 l

1.1 - ' Table of Contents / Revision Index 1.2 Modification Approval Documentation ]

2.0 Scope or Problem Evaluation i I

2.1 Statement of Problem -

2.1.1 Root Cause Analysis 2.1.2 . Impact on Unit ,

2.2 Alternative Analysis 2.2.1 Performance Analysis 2.2.2 Cost / Benefit Analysis 2.3 Recommended Solution ,

3.0 Regulatory Requirements 3.1 Codes aad Standards 3.2 CQE Designation ,

3.3 Code Classification 3.4- USAR Impact 3.5 Tech Spec Impact i

3.6 Licensing Commitments 3.7 Regulatory / Industry Notices IV. Design Input Regtitrements i 4.1 System Functional Requirements l- 4.2 System Performance Requirements l 4.3 System Design Conditions 4.3.1 Environmental 4.3.2 -Scismic l 4.3.3 Lo'ading 4.3.4 Materials 4.3.5 Electrical Power 4.4 Interfaces with Other Systems

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CONTENTS OF MODIFICATION PACKAGE  !

5.0 Design Analysis-5.1 System Design Analysis 5.2 Specifications {

5.3 Drawing List d 6.0 Systems interaction Analysis 1

6.1 Fire Protection j 6.2 Environmental Qualification impact '

j 6.3 High Energy Pipe Break  ;

6.4 Seismic Interaction 6.5 Electrical System interaction 6.6 Human Factors Review - ,

6.7 Security Requirements '

6.8 Environmental Radiological Release .

6.9 Materials Compatibility 6.10 Containment Integrity 6.11 Control Room Hab.

6.12 Missile Protection 6.13 Internal Flooding I 6.14 Separation Criteria j 6.15 Single Failure Criteria l 6.16 Possibility of Operator Error 6.17 Heavy Loads 6.18 Impact on HVAC l 7.0. 10CFR50.59 Analysis 7.1 Design (Operating) l 7.2 Construction 7.3 Testing 8.0. Operating impact 8.1 ALARA Analysis 8.1.1 Construction 8.1.2 Operation 8.2 Constructibility, Operability, Maintainability Review (COM) 8.3 Special Training Requirements 8.4 Special Testing Requirements 8.5 Special Maintenance Requirement 9

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CONTENTS OF MODIFICATION PACKAGE 6 9.0 ' Installation Requirements '~

9.1 - Installation Summary 9.2 . Testing Summary -

9.3 Work Sequence and Scope -]

Breakdown Summary ,

lt 9.4 . Installation Procedures 9.5 Testing Procedures 9.6 Installation & Testing Completion U Summary l

[0.0 Document Revisions - 1 1

Resource Requirements ' j 11.0 l

11.1 - Materials List l

' 4

.11.2 Materials Estimate .

l 11.3 Engineering &. Design Estimate l

11.4 Construction Labor Estimate -

i 11.5 Work Order - 12.0 Schedule 12.1 Design 12.2 Procurement 12.3 C nstruction

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  • LOCATE AND SORT POTENTIAL DESIGN BASIS AND CONSTRUCTION OA RECORDS.

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  • RESOLVE CONSTRUCTION OA RECORDS ISSUE.

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  • GENERATE A SET OF. PLANT LEVEL DESIGN BASIS DOCUMENTS.
  • GENERATE DESIGN BASIS DOCUMENTS ~FOR SELECTED. PLANT SYSTEMS.
  • PHYSICAL VERIFICATION.

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  • RECREATE SELECTED DESIGN BASIS RECORDS.

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., November 3, 1947 DESIGN BASIS RECONSTI'1VIION PROJECT i

PRIORITY LIST Mechanical SDBD-FV-AFW-117 Auxiliary Feedwater-SDBD-DG-112 Emergency Diesel Generators SDBD-CA-105 Instrument-Air-SDBD-SI-CS-131 Containment Spray SDBD-SI-LP-133 Low Pressure Safety Injection SDBD-SI-HP-132 High Pressure Safety Injection-SDBD-SI-130 Shutdown Cooling SDBD-RC-128 Reactor Coolant SDBD-CH-108 -Chemical and Volume Control System SDBD-FW-116 Feedwater SDBD-MS-125 Main Steam and Tbrbine Extraction Steam SDBD-AC-SFP-102 Spent Fuel Pool Cooling.

SDBD-HC-122 Containment Hydrogen Control. --

l SDBD-FW-BD-118 Steam Generator Blowdown' SDBD-AC-CCW-100 Component, Cooling SDF,D-WD-G-144 Gaseous Waste

  • DBD-WD-L-145 Liquid Waste SDBD-WD-S-146 Solid Waste SDBD-AC-RW-101 Raw Water SDBD-RX-401 Reactor SDBD-SL-PRI-135 Primary Sampling SDBD-SL-SEC-136 Secondary Sampling SDBD-SL-PAS-134 Post Accident Sampling i PLDBD-MA-12 Welding PLDBD-MA-13 Painting and Costings Facilities SDBD-FP-115 Fire Protection SDBD-VA-CR-140 Control Room Habitability SDBD-VA-CON-139 Containment HVAC-SDSD-VA-AUX-138 Auxiliary Building HVAC' SDBD-VA-INT 141' Intake Structue HVAC Electrical PLDBD-EE-20 Elec'trical Separation l

PLDBD-EQ-22 Equipment Qualification, SDBD-ES-200 Electrical Grid p SDBD-EE-201 AC Distribution 3

SDBD-EE-202 -j DC Distribution-PLDBD-EE-21 Electrical Equipment '

SDBD-EE-203 CathodicLProtection l

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i November 3, 1987 .j l<

Instrumentation and Controls PLDBD-IC-30 Instrumentation and Control i SDBD-AI-306 Nuclear Instrumentation ]

i SbaD-PC-572 Plant Computer ,

SDBD-PC-ERF-303 DUF Computer l SDBD-PC-304 Security Computer .j QSPDS Computer 1 SDBD-PC-QSD-305 SDBD-CRIC-301 Control Room . Instrumentation and Control Alternate Shutdown PLDBD-IC-31 Human Factors l Structural i

PLDBD-CS-51 Seismic Criteria SDBD-AUX-502 Auxiliary Building Containment SDBD-CONT-501 External Missiles L PLDBD-CS-50 SDBD-STRUC-503 Intake Structure .j SDBD-STRUC-504 Security Building ,

PLDBD-CS-52 Heavy Loads J PLDBD-CS-53 Transportability I 1

Geotechnical PLDBD-CS-54 Geotechnical l

Engineering Hechanics l PLDBD-ME-10 Pipe Stress and Supports 'l l PLDBD-ME-11 Internal Missiles l

Licensing j l PLDBD-NU-61 Regulations, Codes and Standards PLDBD-NU-62 Emergency Preparedness-Radiation Protection SDBD-RM-601 Radiation Monitoring PLDBD-NU-63 Personnel Protection

! Safeguards PLDBD-NU-60 Accident Analysis i

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November 3. 1987 Environmental PLDBL-EV-70 Environmental Security SDBD-PS-701 Plant Security Records PLDBD-RR-71 Records Retention i

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3 GENERIC ISSUES FROM ENGINEERING PERSPECTIVE ,

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- Design Change Program

- Review of Pre-SSONI Modifications

i i

.i; November 5, 1987 j l

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l DESIGN CHANGE PROGRAM REVIEW COMMITTEE REPORT I

INDEX I EXECUTIVE

SUMMARY

~

1.0 BACKGROUND

1.1 Design Inspection i 1.2 Vendor Inspections i 1

1.3 Installation and Test Inspection 2.0 PURPOSE 3.0 SCOPE

4.0 CONCLUSION

S AND RECOMMENDATIONS 4.1 As'sessment of Present Program . ;

4.2 Program Weaknesses u

4.3 Recom.endations o

5.0 REVIEV PROCESS 5.1 General

( ,

5.2 5.3 Regulatory Commitments Program. Weaknesses Identified by SSOMI Reports J 5.4 User. Input 1 5.5 Summary of Major Issues

.i 6.0 DISCUSSION OF MAJOR ISSUES 6.1 Design Basis and Construction QA Records 6.2 Updated Safety Analysis Report (USAR) i y6.3 Docurmntation of Design Assumptions, Design Inputs, and Engineering Judgements 6.4 Commitment Tracking ,

46.5 Design Change Program A 6.6 Procurement 6.7 Safety Evaluations-6.8 Reviews and Approvals ,

6.9 Quality Assurance Program 6.10 Quality Control 9 6.11 Post-Modification Testing 6.12 Installation Procedures.and Field Changes 6.13 System Acceptance 6.14 Emergency Modifications 6.15 Minor Modifications 6.16 Pre-Outage Planning.(Integrated Living Schedule) 6.17 Training 6.18 System Engineer

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