ML17229A349

LER 97-001-00:on 970423,containment Isolation Actuation Occurred.Caused by Increased Radiation Levels During Removal of Upper Guide Structure.Proper Actuation of Containment Isolation Components Was verified.W/970521 Ltr

ML17229A349
Person / Time
Site:	Saint Lucie
Issue date:	05/21/1997
From:	Benken E, Stall J FLORIDA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
L-97-138, LER-97-001-03, LER-97-1-3, NUDOCS 9705280064
Download: ML17229A349 (14)

Text

~ CATEGORY REGULATORY INFORMATION DISTRIBUTION SYSTEM (RZDS)

ACCESSION NB%". 9705280064 DOC. DATE: 97/05/21 NOTARIZED: NO DOCKET FACIL:50-389 St. Z,ucie Plant, Unit 2, Florida Power s Light Co. 05000389 AUTH. NAME AUTHOR AFFZLZAT*ION BENKEN,E.J. Florida Power & Light Co.

STALL,J.A. Florida Power & Light Co; RECZP.NAME RECIPIENT AFFIZ ZATZON

SUBJECT:

LER 97-001-00:on 970423,containment isolation actuation occurred. Caused by increased radiation levels during removal of upper guide structure, Proper actuation of containment isolation components was verified.W/970521 ltr.

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR ENCL SIZE:

TITLE: 50.73/50'.9 Licensee Event Report (ZER), Incident Rpt, etc.

NOTES:

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-3 PD 1 1 WIENS, L. 1 1 INTERNAL: ACRS 1 1 A S AB 2 2 AEOD/SPD/RRAB 1 1 IL 1 1 NRR/DE/ECGB 1 1 1 1 NRR/DE/EMEB 1 1 NRR/DRCH/HHFB 1 1 NRR/DRCH/H ICB 1 1, NRR/DRCH/HOLB 1 1 NRR/DRCH/HQMB 1 1 NRR/DRPM/PECB 1 1 NRR/DSSA/SPLB 1 1 NRR/DSSA/SRXB 1 1 RES/DET/EIB 1 1 RGN2 FILE 01 1 1 EXTERNAL: L ST LOBBY WARD 1 1 LITCO BRYCE,J g 1 1 NOAC POORE,W. 1 1 NOAC QUEENER,DS 1 1 NRC PDR 1 1 NUDOCS FULL TXT 1 1 NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE CONTACT THE DOCUMENT CONTROL DESK)

ROOM OWFN SD-S(EXT ~ 415-2083) TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEEDt FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 25 ENCL 25

C Florida Power 5 Light Company, 6501 South Ocean Orive, Jensen Beach, FL 34957 May 21, 1997 L-97-138 10 CFR 50.73 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555

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Re: St. Lucie Unit 2 Docket No. 50-389 Reportable Event: 97-001

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Date of Event: April 23, 1997 Containment Isolation Actuation due to Increased Radiation Q

The attached Licensee Event Report is being submitted pursuant to the requirements of 10 CFR 50.73 to provide notification of the subject event.

Very truly yours, J. A. Stall Vice President St. Lucie Plant JAS/EJB Attachment cc: Regional Administrator, USNRC Region II Senior Resident Inspector, USNRC, St. Lucie Plant IIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIIII

'rr705280064 'ir7052i PDR ADOCK 0500038'rr 8 PDR an FPL Group company

APPROVED BY OMB No. 316041104 NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION EXRRES 04/30lSS (4-95) ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS MANDATOR I

INFORMATION COLLECTION REQUEST: 60.0 HRS. REPOR'rED LESSON LEARNED ARE INCORPORATED INTO THE UCENSING PROCESS AND FE BACK TO INDUSTRY. FORWARD COMMENTS REGARDING BURDEN ES'TIMAT LICENSEE EVENT REPORT (LER) TO THE INFORMATION ANO RECORDS MANAGEMENT BRANCH fT.B F33I U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON. OC 20666~1 AND TO THE PAPERWORK REDUCTION PROJECT I3160%104), OFRCE Of (See reverse for required number of MANAGEMENTAND BUDGET, WASHINGTON, OC 20603.

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FACIUTY NAME 11) DOCKET NWIISER 12I PAGE (3I ST LUCIE UNIT 2 05000389 1 QF7 TITLE I>>

Containment Isolation Actuation due to Increased Radiation Levels during Removal of Upper Guide Structure FACIUTY NAME DOCKETNUMBER SEOUENTIAL REVISION MONTH DAY YEAR MONTH DAY YEAfl NUMBER NUMBER N/A FACIUTYNAME DOCKET NUMBER 04 23 97 97 001 00 05 21 97 N/A OPERATING MODE (9) 20. 2201 (b) 20. 2203(a) (2) (v) 50.73(a) (2) (I) 50.73(a)(2)(viii)

POWER LEVEL (10) 000 20.2203(a) (2) (I) 20.2203(a)(3)(ii) 50.73(a)(2)(iii) 73.71 OTHER 20.2203(a)(2)(m) 50.36(c) (1) 50.73(a) (2) (v) Specify ln Abstract below or in NRC Form 366A 20.2203(a)(2)(iv) 50.36(c)(2) 50.73 (0) (2) (vii)

NAME TELEPHONE NUMBER Snolude Area Code)

Edwin J. Benken, Licensing Engineer (561) 476 - 7156 f.'QSL,I?.g., REPORTABLE REPORTABLE CAUSE SYSTEM COMPONENT MANUFACTURER CAUSE SYSTEM COMPONENT. MANUFACTURER To NPRDS To NPRDS

$ '.:('Nef; je MONTH OAY YEAR EXPECTED YES SUBMISSION (If yes, complete EXPECTED SUBMISSION DATE).

X No DATE (15)

ABSTRACT (Umit to 1400 spaces, i.e., approximately 15 single. spaced typewritten lines) (16)

On April 23, 1997, St. Lucie Unit 2 was in Mode 6 during a scheduled refueling outage. At 1758, a containment isolation actuation signal (CIAS) was received during the removal of the upper guide structure (UGS) in preparation for fuel movement. The CIAS was anticipated based on previously identified high radiation levels in the area of the upper incore instrumentation (ICI) guide tubes. Non-essential personnel had been removed from the area prior to UGS movement and the evolution was completed without further incident. The CIAS was reset following verification of component actuation and the completion of the UGS relocation to its refueling storage location.

The CIAS was generated as a result of high radiation as measured by two-out-of-four containment isolation radiation monitoring instruments. The monitor set points had been adjusted to 90 mR/hr as required by plant Technical Specifications during the refueling mode. The elevated radiation levels were caused by the presence of irradiated ICI segments which had broken during removal and remained in the upper section of the ICI guide tubes contained in the UGS.

Corrective Actions include: 1) Proper actuation of containment isolation components was verified. 2) ICI material found in the guide tubes was removed or relocated within the guide tube to ensure adequate shielding during UGS movement. 3) An engineering safety evaluation was performed to verify the acceptability of operation with simulated incore assemblies and ICI remnants. 4) Procedure changes were made to improve the ICI removal process. 5) FPL is continuing to evaluate the ICI failure mechanism and the need for additional corrective actions. 6) Radiological survey practices are being reviewed for improvement.

NRC FOAM 366 (4.95)

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSIO I4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION YEAR SEQUENTIAL REVISION ST. LUCIE UNIT 2 05000389 2 OF 7 97 001 00 TEXT Iffmore speceis required, use edditionel copies of NRC Farm 3GSAJ I171 On April 23, 1997, St. Lucie Unit 2 was in Mode 6 during a scheduled refueling outage. At 1147, the upper guide structure (UGS) [EIIS:AB] was being removed from the reactor vessel in accordance with General Maintenance Procedures, as part of the refueling sequence. The reactor cavity water level was being maintained at approximately 56.1 feet, as required, and containment integrity was established.

Four channels of containment isolation radiation monitoring [EIIS:IL] were in se'rvice with alarm set points adjusted at 90 millirem per hour (mR/hr) in accordance with Technical Specifications (TS).

As the UGS was raised approximately eight feet, and the incore instrumentation (ICI) [EIIS:IG] guide tube cluster assembly (refer to Figure 1) cleared the surface of the water, measured radiation levels in containment rapidly increased. Containment Isolation monitor channel MD [EIIS:IL] alarmed at a measured radiation level of 90 mR/hr and an automatic containment evacuation alarm was sounded.

Three other containment isolation radiation monitoring channels (MA, MB and MC) [EIIS:IL] also indicated elevated radiation levels, of up to 30 mR/hr, however, only the MD channel exceeded the CIAS set point of 90 mR/hr and therefore a CIAS (requiring 2-out-of-4 channel logic) was not initiated.

The UGS was lowered into the reactor vessel at 1154 and radiation levels decreased to normal values when the ICI guide tube cluster assemblies were submerged in water. Following the containment evacuation alarm, an orderly evacuation of the containment was conducted.

An inspection of the UGS was initiated using a radiation monitor and extended boom. Radiation levels in excess of 500 REM were detected in the area of the incore detector guide tubes, which indicated that activated ICI material had become lodged in the upper portion of the tubes during instrument removal. The incore detectors had been removed by FPL Instrument and Control (IRC) personnel prior to the UGS lift, for replacement with new detectors and an ICI flange modification.

Based upon the radiological conditions associated with the UGS, a meeting was held with plant management to assess available options and determine the appropriate actions to facilitate removing the broken detector material. It was determined that the UGS would be moved to its refueling storage position in the lower reactor cavity prior to performing any additional inspection or material retrieval.

Relocating the UGS to this position would preclude potential foreign material exclusion issues in the area of the reactor vessel.

On April 23, 1997, at 1705, operations conducted a crew briefing to review moving the UGS. A CIAS was anticipated to occur during the UGS lift and was discussed with operating crews prior to performance of the evolution. All non-essential personnel were then cleared from the containment prior to the start of the evolution. At 1750, the UGS lift was commenced, and a valid CIAS was received at 1758. Following receipt of the CIAS, operators verified the proper actuation of required components.

At 1810, on April 23, 1997, the UGS was placed in its refueling storage position in the lower reactor cavity, and the CIAS was reset at 1812.

Subsequent inspection of the UGS, including boroscopic examination of the ICI positions, identified that broken incore detector segments were present in three incore instrumentation (ICI) guide tube locations.

The affected detector locations were L-16, R-9 and T-13 (refer to Figure 2).

NRC FORM 368A I4-95I

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSIO I4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION YEAR SEQUENTIAL REVISION ST. LUCIE UNIT 2 05000389 3 OF 7 97 001 00 TEXT Iffmore spece is required, use edditionel copies of NRC Form 366AI I17)

Following a detailed inspection of the UGS, efforts to retrieve the incore detector material from the guide tubes were initiated. The ICI detector debris located at core position L-16 was successfully removed using a specifically designed extraction tool and transferred to the spent fuel storage area.

Detector debris found in the ICI guide tubes at positions R-9 and T-13 could not be extracted due to the condition of the material. The detector segments at these locations were subsequently inserted toward the bottom of the instrument tube thimble assemblies to allow adequate shielding during movement of the UGS back to its core location. Partial length, simulated ICI assemblies were installed at detector locations R-9 and T-13 to simulate the physical rigidity of a normal ICI in the curved guide tube region.

When the above work was completed, the UGS was temporarily raised to normal travel height to verify that radiation levels remained acceptable and then reset in the lower reactor cavity. This test was satisfacto(ily completed on April 30, 1997.

The Containment Isolation Actuation Signal (CIAS) which was actuated at 1758, on April 23, 1997, was a valid actuation based on elevated radiation levels measured in containment during the removal of the UGS from the core to its refueling position. The CIAS was anticipated and planned, and containment isolation system (CIS) components were verified to have operated in accordance with their design following the actuation. The removal of the UGS to its refueling storage location was completed without further incident.

The increase in containment radiation levels encountered during the UGS lift were caused by the presence of irradiated ICI detector segments located in three of the incore detector guide tubes.

Following boroscopic examination of the incore instrumentation flanges, it was determined that incore detectors at core positions L-16, R-9, and T-13 had broken while being removed from their associated guide tubes. The detectors were being removed in preparation for the installation of an ICI flange modification, and were scheduled for replacement. The incore detectors at the above three locations were broken when the irradiated portion of the detector assembly passed through the "S" bends in the upper portion of the guide tubes located above the ICI support plate. The exact mechanism for the failure of the'CIs during their removal has not yet been determined.

The containment isolation radiation monitoring instrumentation at St. Lucie Unit 2 was operable during this event with the alarm/trip set point adjusted to less than or equal to 90 mR/hr as required by Technical Specification (TS) 3.3.3.1 Additionally, the CIS was operable as required by Technical

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Specification 3.9.9 for core alterations and fuel movement. During the UGS relocation from the reactor to its refueling storage position in the lower reactor cavity, an anticipated CIAS was received when 2-out-of-4 containment area radiation measurement channels exceeded 90 rnR/hr. This event is therefore reportable under 10 CFR 50.73(a) (2) (iv), as "Any event or condition that resulted in a manual or "

automatic actuation of any engineered safety feature (ESF), including the reactor protection system...

NRC FORM 366A I4.95I

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSIO I4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION YEAR SEOUENTIAL REVISION ST. LUCIE UNIT 2 05000389 4 OF 7 97 001 00 TEXT Iifmore spaceis required, use additional copies of NRC Form 366AI I17I Operability of radiation monitoring channels, per TS 3.3.3.1, ensures that radiation levels are continuall measured in the'areas served by the individual channels and the alarm or automatic function is initiated when the radiation level trip set point is exceeded. For this event, the containment isolation radiation monitoring instrumentation was operable and functioned as designed during the initial UGS lift to automatically actuate a containment evacuation alarm upon measurement of radiation levels at the prescribed set point of 90 mR/hr. Additionally, continuous coverage was provided by Health Physics personnel while lifting the UGS to ensure that only essential personnel were in the area during the UGS removal and radiological conditions remained acceptable.

Operability of the containment isolation system during core alterations ensures that the containment is automatically isolated upon detection of high radiation levels within the containment. The system will

. restrict the release of radioactive material from the containment atmosphere to the environment in the event of a fuel handling accident. During this event, core alterations or fuel movement were not in progress, al1d therefore, the potential for a fuel handling accident did not exist. Additionally, the CIAS was anticipated based on previously measured radioactivity levels on the UGS and additional HP protective measures were implemented prior to UGS movement to ensure personnel exposure was minimized.

As a result of this event, incore instrumentation at core locations R-9 and T-13 will not be available during St. Lucie Unit 2, Cycle 10, and simulated, partial length ICI assemblies will be used at these locations. FPL performed an engineering safety evaluation to determine the acceptability of this modification and concluded that the installation of dimensionally equivalent ICI assemblies will not adversely affect reactor coolant system (RCS) integrity. Additionally, more than the minimum number of incore detectors, as defined in the St. Lucie Unit 2 Final Updated Safety Analysis Report (FUSAR),

will be available for St. Lucie Unit 2, Cycle 10 operation. The evaluation also concluded that the remnant ICI segments located in core locations R-9 and T-13 will have no adverse effect on plant operation or safety.

There was no adverse impact on radiological safety or release of radioactive material as a result of this event. All radiation monitoring instrumentation functioned as designed and health physics measures were in place at all times to ensure that personnel exposure was minimized during movement of the UGS. Based on the above, the protection of the health and safety of the public was not affected by the event.

Following the CIAS on'April 23, 1997, operators verified that required. containment isolation components had properly actuated. The CIAS was then reset following the completion of the UGS relocation to the lower reactor cavity area.

'NRC FORM 366A (4.95I

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSIO I4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION YEAR SEQUENTIAL REVISION ST. LUCIE UNIT 2 05000389 5 OF 7 97 001 00 TEXT itfmore speceis required, use edditionel copies of NRC Form 366AI I17I

2. The UGS was moved to its refueling storage location, and the incore detector material found in the ICI guide tubes was either removed for disposal or repositioned to the lower thimble assembly portion of the guide tube. Following this, the UGS was raised to the normal travel height to verify that radiation levels remained normal. The UGS was subsequently moved back to its operating position in the core without incident.

3. Simulated ICI assemblies were installed in core locations R-9 and T-13, and an engineering safety evaluation was performed which verified the acceptability of operation with assemblies in place and remnant ICI segments remaining in the two incore guide the'imulated tubes. FPL will further determine if this modification will remain for one cycle or as a permanent change to the facility.

C 4, The Instrument and Control tIRC) procedure related to the removal of the incore detectors was revised to include a positive verification of proper ICI removal through the use of remote video equipment. Additionally, FPL will assess the type and application of the load cell being used during ICI removal to determine if further improvements can be made.

5. FPL is continuing to assess the failure mechanism associated with the incore instrumentation assemblies located in core positions L-16, R-9 and T-13. Additional corrective actions will be implemented as needed based on the results of the assessment. For example, the use of non-depleting incore detectors, which do not require frequent replacement, is being evaluated for possible u'se at St. Lucie Plant.

6. Radiological surveying practices associated with the upper guide structure will be reviewed for enhancements in identifying potentially radioactive ICI material in the UGS prior to its removal from the reactor core.

None None NAC FOAM 388A I4.95I

NRC FORM 366A U.S. NUCLEAR REGULATORY COMMISSIO (4-95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION YEAR SEQUENTIAL REVISION ST. LUCIE UNIT 2 05000389 6 OF 7 97 001 00 TEXT iifmore spece is required, use addi tionel copies oi AIRC Form 366A/ I17I INCORE INSTRUMENTATION INSTALLATIONARRANGEMENT UGS SUPPORT PLATE

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