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iJuly 6, 1994 ;

NG-94-2549 |!

Mr. John B. Martin |Regional Administrator |Region III iU. S. Nuclear Regulatory Commission !

801 Warrenville Road iLisle, IL 60532 !

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Subject: Duane Arnold Energy Center ;

Docket No: 50-331 !Op. License DPR-49 |Licensee Event Report #94-009 |

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Gentlemen:

In accordance with 10 CFR 50.73 please find attached a copy of the |subject Licensee Event Report. ;

The following new commitments are made in this letter: |i

* Appropriate maintenance procedures will be revised to inform j

technicians of this type of condition. These procedural |

enhancements will be implemented no later than October 31, 1994. !

i* Training will be conducted for mechanical, electrical, |

instrumentation & control, operations, health physics, !engineering, and QA/QC personnel covering the specifics of :this event. This action will be completed no later than |December 31, 1994. j

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Very truly ours,.

hDavid L. Wilson iPlant Superintendent - Nuclear i

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DLW/TW/eah t

cc: Director of Nuclear Reactor Regulation |Document Control Desk |U.S. Nuclear Regulatory Commission iMail Station PI-137 i

Washington, D. C. 20555 9407140004 940706PDR ADOCK 05000331 // p!

1CCCI[NRCResidentInspector-DAEC PDR |

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Duane Arnold Energy Center * 3277 DAEC Road a Palo. Iowa 52324 * 319/851-7611 0

An IES INDUSTRIES Company

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|NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104g en EXPlRES 5/31/95 '

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LICENSEE EVENT REPORT (LER) COMMus HEcARo,Na ouRotN EstuuE TO 1HE ,NronMAnONAND RF CORDS MANAGEMENT BRANCH (MNBB 7710. U E. NUCLE ARHEGULATOAV COMMtSSION W ASHINGTON, DC 20555-0001, AND TOTHE rAPEpwonk HEDUCTON PROJECT (3150 0104, OrFICE Or3

g$ee reverse for requered number Of digits / characters for each block) MANAGEMENT AND BUDGET. W ASHINGTON, DC 20503. |

F ACILITY NAME (1) DOCKET NUMBER (2) FAGE (3)

Duane Arnold Eneray Center 05000 331 1 OF 7llTLE|4)

Control Rod Drive Differential Pressure Transmitter Leak IntO The Reactor Building i

EVENT DATE (5) LER NUMBER (61 REPORT NUMBER (7) OTHER FACILITIES INVOLVED (8) |

AGIUT Y N AME DOCKET NUMBEHLE QUE NT;At REvrSIONMONTH oAv vtAR < EAR uGNTH oA< vEAR,, ,, gg

F ACILITY NAME DOCKET NUMBER

06 09 94 94"

009~

00 07 06 94 05000

OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 1: (Check one or morel (11)MODE (9) 1 20 402(b) 20.405(c) )( 50.73(a)(2)(iv) 73.71(b)

POWER 20 405fas(1)(i) 50.36(c)(1) 50.73(a)(2)(v) 73.71(c)

LEVEL (10) 100 20 405ta)(1)(n) 50.3 sic)(2) 50 73(a)(2)(vu) OTHER

20 405(a)(1)(ud 50.73(a)(2)(i) SC.731a)(2)(vni)(A) (SP'c#V 'a AM"acibelow and in Test. NRC

20.4051al(1)(iv) 50.73(a)(2)(n) 50.73(a)(2)(vtu)(B) ponn x,c,5) j

20 405(a)(1)(v) 50.73(a)(2)(ni) 50.73(a)(2){x) ||

LICENSEE CONTACT FOR THIS LER (12)NAML TELLittONE NUMbtR pnclude Area Gooe) |

Thomas K. VanWyen, Licensing Specialist (319) 851-7465 i'

COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE SYSTEU COMPONE NT MANUF ACTURE R CAUSE EYSTEM COMPONENT MANUFACTURERRD

X AA PDT 1204 NO -- -

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SUPPLEMENTAL REPORT EXPECTED (14) EXPECTED MONTH DAV DEAR

vas SUBMISSIONgp yn. comr-e ExPEcTtc EueMismON DATEl )( DATE (15)

ABSTRACT (Limit to 1400 spaces, i e., approomately 15 s:ngle-spaced typewntten knes) (16) |

At 0109 hours on June 9,1994 with the plant operating at 100% power, an ExcessFlow Check Valves System B Hi Flow alarm was received on control room panel IC03due to a failed differential pressure transmitter (PDT). The respective excessflow check valve immediately isolated, thereby preventing the release of reactorcoolant to the reactor building. This event was considered an Emergency SafetyFeatures (ESF) actuation.

The cause of this event was a packing leak that dripped condensate water on tothe PDT housing. Over time, two bolts that coupled the PDT housing components i

failed, causing subsequent failure of the PDT and the resultant leak. jl

Corrective actions include replacement of the PDT with a model that is notconducive to the entrapment of water in its housing, procedure revision, andperformance of a detailed walkdown of like components to ensure there are nosimilar concerns. There was no ef fect nn plant safety, personnel safety, orplant availability as a result of this event.

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I. DESCRIPTION OF EVENT: i

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On June 9, 1994 the plant was operating at 100% power. There were two ,

limiting conditions for operations (LCOs) in effect on the Residual HeatRemoval (RHR) system; day 3 of 30 for the "B" RHR pump due to pre plannedmaintenance, and day 2 of 7 for the "B" loop of the low pressure coolant '

injection (LPCI) mode of the RHR system due to pre planned valve testing.There were no other operational problems at the time of this event. |

At 0109, control room annunciator IC03C(D-2), " EXCESS FLOW CHECK VALVESSYSTEM B HI FLOW" alarmed. A control room operator immediately performedthe prescribed operator actions dictated by the annunciator responseprocedure (APsP). Upon arriving at control room back panel 1C29, the ;

operator observed that " PRESSURE ABOVE CORE PLATE EXCESS FLOW CHECK VALVE"XFV-4528 indicated that the valve was fully shut. At the same time, whileconducting routine watchstanding duties in the reactor building, anon-licensed operator heard a loud noise and immediately investigated. .

This investigation revealed that control rod drive (CRD) differential i

pressure transmitter PDT-1825 was leaking. The non-licensed operator !'immediately contacted the control room and informed the OSS of this leak.

Subsequent followup by the control room crew revealed the followingindications:

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* CRD Cooling Water Differential Pressure Indicator 1832Afailed high.

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* CRD Drive Water Differential Pressure Indicator 1825A failedlow.

* Core Plate Differential Pressure Recorder 4528 failed high.

* Core Spray Sparger Differential Pressure Indicators 2119 !'

and 2139 indicating higher than normal.

IAt 0118, the non-licensed operator isolated the failed CRD differentialpressure transmitter (PDT-1825) by shutting isolation valves on the iow ,

pressure and high pressure legs of the device. This action stopped the |leakage of CRD system process fluid (i.e. condensate water) into thereactor building. It should be noted that the leakage of reactor coolant ,

into the reactor building was stopped at the onset of this event (time ,

0109) by the successful closure of excess flow check valve XFV-4528. !

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At 0120, XFV-4528 was reset by a control room operator. Upon this valve.

opening, CRD Cooling Water, Core Plate Differential Pressure, and Core !

Spray Sparger Dif ferential Pressure indications were restored. As a ;

precaution, Operations initiated maintenance actions on all restored I

instrumentation to inspect for damage and vent all affected !

instrumentation. No damage was found to have occurred on any of theaffected instruments, and all instruments were successfully vented.

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Health Physics, assisted by Instrument & Control Technicians, contained I

the leak to the area immediately adjacent to the CRD Flow Control Stationand the Traversing In-Core Probe (TIP) mezzanine area. A floor drain wasdirectly below the leaking transmitter, therefore the majority of water i

idischarged was sent to Rad Waste, which quantified the volume atapproximately 50 gallons. Subsequent surveys conducted by Health Physics i

personnel revealed that neither the non-licensed operator or the affecte<Jareas were contaminated. The area was released for general access at 0550.

This event was determined to be reportable pursuant to 10 CFR i

50.72(b)(2)(ii) and 10 CFR 50.73(a)(2)(iv). j

II. CAUSE OF THE EVENT: |.

Two of the four bolts that secure the low and high pressure retaining |housing of the ITT Barton model 368 differential pressure transmitter ;failed. This failure resulted in a loss of pressure integrity at the itransmitter and the subsequent leak into the reactor building. !

IThe test results of the bolts on IIT Barton model 368 conclude that the !

!bolts are of hardened carbon steel and therefore susceptible to crackingin a moist environment. The root cause of the bolt failure was a previously junidentified packing leak of approximately 1 drop every 10 minutes on a i

manual isolation valve for the CRD Stabilizing Flow Outlet Indicator. |The location of this valve is directly above PDT-1825. This packing leak idripped water on top of the transmitter, and collected into a " natural ;reservoir" that is created when the low and high pressure housing are

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bolted together (Figure 1). This condensate water collected around the |hardened carbon steel bolts. Failure analysis identified that the

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submergence of these bolts in condensate over an extended period of time !

caused corrosion assisted cracking and could have been assisted to some jdegree by hydrogen embrittlement. This resulted in ultimate failure on jJune 9, 1994 at 0109.

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III. ANALYSIS OF THE EVENT:

CRD Drive Water Differential Pressure Transmitter PDT-1825 senses themfferential pressure between CRD drive water header pressure and reactor i

pressure. The PDT provides local indication and remote indication to the ,

control room at panel 1C05. This indication enables the operator to !maintain CRD drive water header pressure approximately 260 psig greater i

than reactor pressure via the operation of Drive Water Differential'

Pressure Control Valve MD-1830. The high pressure leg of the transmitteris fed from the CRD system, while the low pressure leg is fed by aninstrument line from the reactor vessel. !

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Upon failure of the transmitter bolts, the transmitter's low and high !

pressure housings decoupled, resulting in both the low and high pressure '

legs leaking into the reactor building. The high pressure leg, which is }connected to the CRD system, was manually isolated by the non-licensed ioperator in order to isolate the leaking CRD process fluid. Condensatewater is the source of water to the CRD system.

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The low pressure leg, as stated earlier, is an instrument line thatpenetrates the drywell and is part of the reactor coolant pressureboundary. As such, it is provided with an excess flow check valve :

(XFV-4528). This valve will automatically shut and thereby prevent an'uncontrolled release of reactor coolant in the event of a rupture

downstream of it. This occurred upon a differential pressure of 10 psidacross the XFV.

The indications that XFV-4528 performed its design function wereimmediately verified by control room operators on panel 1C29 and also by :the remote indications of shared differential pressure transmitters that !

are also downstream of XFV-4528. Further evidence that the XFV operatedcorrectly is substantiated by the fact that neither the non-licensed ioperator or CRD Flow Control station area were found to be contaminated ,

upon subsequent surveys conducted by Health Physics personnel. -

There was no effect on plant safety, personnel safety, or continued i

operation as a result of this event. The excess flow check valve respondedas designed upon receipt of excess flow. '

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IV. CORRECTIVE ACTIONS:i

1. A Rosemont differential pressure transmitter was installed jin place of the failed ITT Barton transmitter.

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2. Operations has performed manual valve packing adjustments on !

the isolation valve for the CR0 Stabilizing Flow Outlet Indicator. |'3. Quality Control performed a walkdown of other ITT Barton

instruments with the same pressure housing design and assured !'

standing water is not present on top of any housing. :

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4. Long term corrective actions are: |

a. Appropriate maintenance procedures will be revised to finform technicians of this type of condition. These i

procedural enhancements will be implemented no later !than October 31, 1994. !

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b. Training will be conducted for mechanical, electrical, jinstrumentation & control, operations, health physics, [engineering, and QA/QC personnel covering the specifics of !

i this event. This action will be completed no later than |December 31, 1994. |

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V. ADDITIONAL INFORMATION: j

A. Previous Similar Evants:i

lhere are no similar events, however, LER 92-008 involved :two bowl retaining bolts that broke when subjected to i

less than required torque requirements. The failures were i

found to be due to hydrogen induced cracking due to |an anomaly in the heat treatment process. |

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B. EIIS System and Component Codes: ,

Control Rod Drive System - AA. ;

Engineered Safety features Actuation System - JE.Transmitter, Differential Pressure - PDT.Valve, Isolation - ISV. -

C. Equipment Information:i

The differential pressure transmitter installed was ITT.

Barton No. 368 Electronic Differential Pressure Transmitter. .

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