ML18059A446

LER 93-009-00:on 930919,identified Leak in PORV Line.Caused by PWSCC in HAZ of PORV Line to Pressurizer Nozzle Safe End Weld.Engineering Evaluation Will Be Performed & Addl Repairs & C/As Will Be Completed by Next shutdown.W/931015 Ltr

ML18059A446
Person / Time
Site:	Palisades
Issue date:	10/15/1993
From:	ROBERTS W L, SLADE G B CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-93-009, LER-93-9, NUDOCS 9310220185
Download: ML18059A446 (6)

Text

• consumers Power POWERIN& MICHl&AN'S PRD&RESS Palisades Nuclear Plant: 27780 Blue Star Memorial Highway, Coven, Ml 49043 October 15, 1993 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555

• GB Slade General Manager DOCKET 50-255 -LICENSE DPR-20 -PALISADES PLANT -LICENSEE EVENT REPORT 93-009, PRESSURIZER PENETRATION SAFE END CRACK RESULTS IN PCS LEAKAGE Litensee Event Report (LER) is attached.

This is reportable in accordance with 10 CFR 50.73(a)(2)(ii).

Gerald B Slade General Manager CC Administrator, Region Ill, USNRC NRC Resident Inspector

-Palisades Attachment 9310220185 931015. -r:*** PDR ADOCK 05000255 f.. S . PDR * !!,. r A CMS ENERGY COMPAN)'°

• NRC Form 366 U.S. NUCLEAR REGULATORY COMMISSION (9*83)

• APPROVED OMB NO. 3160-0104 EXPIRES: 9/31196 LICENSEE EVENT REPORT (LERI FACILITY NAME 111 DOCKET NUMBER 121 PAGE 131 Palisades Plant 01s101010121s1s 1 I OF 01 5 TITLE 141 PRESSURIZER PENETRATION SAFE END CRACK RESULTS IN PCS LEAKAGE I EVENT DATE 161 LER NUMBER 161 REPORT DATE 161 OTHER FACILITIES INVOLVED !Bl SEQUENTIAL REVISION FACILITY NAMES MONTH DAY YEAR YEAR NUMBER NUMBER MONTH* DAY YEAR N/A 0161010101 I --01 9 i I 9 9 3 3 01 o I 9 o I o 11 0 i I s 3 N/A 0161010101 I THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR I: !Ch.ck OM or more of tM following/

1111 OPERATING N MOOE 191 20.4021bl 20.4061cl 60.731a)(211iv) 73.71 (bl *---POWER I 20.406(*111

)(ii 60.36icll1 I 60. 7 3(a)(21M 73.71 lei LEVEL I I o --I---1101 20.4061al11 l(iil 60.36(c)(21

60. 73(al12)(vii)

OTHER (Specify in Abatract ---20.406(a)(1

)(iiil 60.73(all211il

60. 7 3 (a)(2) (viii) IA) below end in Text, ---* 20 .406la)(1

)(iv) x. 60.731all211iil

60. 7 31all211viiillBI NRC Form 366Ai --20.4061all1 IM 60.731all211iiil

60. 7 31all211xl LICENSEE CONTACT FOR THIS LER 112) NAME TELEPHONE NUMBER William L. Roberts, Staff Licensing Engin_eer 1 7 1 6 1 4 1 -I a I s I , I 3 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT 1131 MAN.UFA.C*

REPORTABLE MANUFAC* REPORTABLE CAUSE SYSTEM COMPONENT TUR ER TO NPROS . CAUSE SYSTEM COMPONENT TUR ER TO NPROS I I I I I I I I I I I I I I I I I I I I I I I I I I I i SUPPLEMENTAL REP.ORT EXPECTED 1141 MONTH DAY YEAR I EXPECTED I l YES Vf yes, complere EXPECTED_

SUBMISSION DATE! lxl NO SUBMISSION I I I DATE 1161 ABSTRACT {Limit to 1400 spaces. i.e .* *pproximat*iv fiftHl'I ling/trsp<<*

typewrimm lin!al 11 si On September 19, 1993, at approximately 1900 hours0.022 days <br />0.528 hours <br />0.00314 weeks <br />7.2295e-4 months <br />, the plant was in the process of heating up following a refueling outage. The plant's primary coolant system (PCS) was in a hot shutdown condition (532°F and 2060 psia) when plant operations personnel identified a ieak in the power operated relief valve (PORV) line near the nozzle connection to the pressurizer.

The plant was returned to cold shutdown.

The crack tnitiated due primary water stress corrosion cracking (PWSCC) in the heat affected zone (HAZ) of the power operated relief valve (PORV) Inconel 600 safe end. Corrective actions include removing a portion of the safe end containing the crack for evaluation, examining the remaining safe end to.establish its condition for future use, rewelding the PORV pipe to the examined safe end to replace the piping that was removed, and examining and evaluating the remaining pressurizer nozzles and other primary coolant system nozzles to provide assurance of operability prior to returning the plant to service. I \

><RC F"'m 388A U.S. NUCLEAA llEGUL-" TORY COM ... ISSICI< AP9'10VEO OMB NO: 3 160-01 C* EXPIRES: 8131.'86 19-831 FACILITY NAME 11 I LICENSEE EVENT REPORT !LERI TEXT CONTINl)ATION OOCl(ET NUMBER 121 LEll NUMBER 131 SEQUENTIAi.

YEAA NUMBER REVISION NUMBER Palisades Plant Q 5 . Q Q Q 2 5 5 9 3 -

Q

• Q 9 -Q Q Q 2 OF Q 5 EVENT DESCRIPTION On September 19, 1993, at approximately 1900 hours0.022 days <br />0.528 hours <br />0.00314 weeks <br />7.2295e-4 months <br />, the plant was in the process of heating up following a refueling outage. The plant's primary coolant system (PCS) was in a hot shutdown condition (532°F and 2060 psia) when plant operations personnel . identified an increasing trend in containment sump level indication, indicating a leak in the PCS. A few minutes later, an auxiliary operator conducting rounds in the containment reported a steam leak near the pressurizer.

Closer inspection found an isolatable leak in the power operated relief valve (PORV) line near the pressurizer relief valve nozzle. The plant was returned to cold shutdown.

While cooling down, a second visual examination of the leak was performed with primary system pressure about 200 psig. This visual inspection characterized the leak to be a partial circumferencial crack, in or very near to/the Inconel 600 safe-end on the pressurizer On September 17, 1993 the plant achieved cold shutdown and direct visual and NOE examination of the crack area was performed.

The leak area found the circumferencial crack to be approximately 3-inches in length (about 30 percent of the circumference) in the Inconel safe-end to pipe weld. Review of containment sump level information during the event indicates the steam leak was on the order of 0.2 gpm equivalent water. -This event is reportable in accordance with 10 CFR 50.73(a)(2)(ii) as an event that resulted in one of the nuclear power plants principal safety barriers being seriously degraded.

CAUSE OF THE EVENT The crack initiated due to primary water stress corrosion cracking (PWSCC) in the heat affected zone (HAZ) of the PORV line to pressurizer nozzle safe end weld. The cracking mode was intergranular from the inside diameter pipe surface with the final 5 to 10% of crack growth being transgranualar.

• During the inservice inspection of th1s weld earlier this refueling outage, the cap on this weld was ground down to facilitate resolution during volumetric NOE of a recordable indication.

The earlier presence of the weld cap may be why the crack did not open up sooner. ANALYSIS OF THE EVENT At the time of the event the was in the process of heating up from a refueling outage. Plant conditions were such that the PCS was nearly at temperature and pressure and all associated safeguards equipment supporting this stage of plant operation was operable.

The was returned to a cold shutdown condition.

Transition to cold shutdown required no abnormal operations or any safeguards equipment performing a design basis function.

• 'l l<RC FOfm 388A 19*831 Palisades Plant LICENSEE EVENT REPORT (LERI TEXT CONTINUATION DOCKET NUMBER 121 YE.AA LER NUMBEI'\ Ill

• SEQUENTIAL NUMBER U.S. NUCLE.AA qEGULl<TORY APP'IOVEO OMB 1<0. 3'60-0'C4 EXP!RES: 1131/86 RE'\llSlON NUMBER 0 5 0 0 0 2 5 5 9 3 -

0 0 9 -0 0 0 3 o* 0 5 Analysis is ongoing to determine the margin-to-failure for the cracked PORV nozzle safe-end.

This analysis will provide the conclusion for the severity of the crack growth. SAFETY SIGNIFICANCE These three scenarios are hypothesized to judge the possible safety 6f

• event had it occurred while the plant was at power operation.

1. A leak of the same size as was experienced during this event could have resulted.

The Technical Specifications leakage limit of 1 gpm unidentified leakage may have been approached or exceeded and a normal shutdown would occur. 2. A leak could have started small and gradually progressed in size over time. As the leak increased in size monitoring the increasing leakage into the containment sump would have allowed us to take actions to shut the plant down. Exceeding 1 gpm leakage unidentified would require plant shutdotwn.

3. The cracking could have been initially more severe such that a much larger leak developed.

In the case of the PORV piping, the limiting size of the leak is the cross sectional area of the pressurizer relief valve nozzle. Analysis has been completed for small break LOCAs for up to a one square foot break size * (CENPD-137 Supplement 1-P, "Small Break Model Calculative Methods for the CE Small Break LOCA Evaluation Model," January 1977). Since the PORV nozzle is a three inch nozzle, the area available to leak would be 0.05 square feet. Analysis indicates that the maximum fuel cladding temperature predicted for a small break LOCA of this size is 660°F. The maximum thickness of cladding oxidation is predicted to be 0.002% of the cladding wall thickness.

These values are much less than the 10 CFR 50.46 acceptance criteria of a peak cladding temperature of 2200°F and maximum cladding oxidation of 17%. These analysis are also predicted based on assuming the breaks are located on the bottom of the PCS cold legs. Breaks on the top of the pressurizer are much less limiting due to a predicted lower mass ejection rate. CORRECTIVE ACTION The specific actions that have been or are being taken to address the pressurizer relief valve nozzle safe end crack are summarized below. Initial results of the action plan investigations were reported to the NRC in a dated October .7, 1993. Specific plan actiqns are listed below and have been divided into short term and long term actions. Short term actions will be completed prior to start up from the present refueling outage.

F"'m J99A U.S. NUCLEAA 'IEGUL.UORY APl'llOVED OMS 1<0. J' 60-C' C* . E)(P,RES:

Bil 1 .116 19*8JI LICENSEE EVENT REPORT (LERI TEXT CONTINUATION FACILITY NAME 111 OOCltEi NUMBER 121 LER NUMBER 131 PAGE 141 SEQUENTIAi.

YEAA NUMBER REVISION NUMBER Palisades Plant O 5 0 O 0 2 5 5 9 3 -

0 0 9 -0 0 0 4 o* 0 5 or erm 1. An engineering evaluation of the failure has been conducted.

The specific elements of this engineering analysis included the following:

a. Metallurgical analysis of the failed safe end to identify the cause of the crack. b. Analysis of factors contributing to PWSCC in the pressurizer nozzle safe end. This includes evaluation of material properties and stresses that may have contributed to the failure. Piping stresses and weld residual stresses are being evaluated and will be reviewed by an independent third party. . c. Evaluation of other nozzles safe ends in the primary coolant system, based on the engineering evaluation of the failure, to identify other locations which may be susceptible to the same failure cause. This evaluation includes both nozzles with Inconel 600 safe ends and other safe end materials.

Contributing factors to PWSCC will also be evaluated for the safe ends that are identified as being susceptible to the same failure cause .. This will include evaluation of material properties and stresses.

d. Evaluation of appropriate non-destructive examination techniques to identify similar flaws in other susceptible safe ends. 2. Corrective actions for the specific safe end that failed have been identified.

An engineering evaluation of the repair to the pressurizer safe based* on the root cause analysis of the failure, has shown that the lifetime of the repaired safe end well exceeds the length of the next operating cycle.

• 3. Corrective actions for other safe ends that may be susceptible to the same failure have peen identified.

Other safe ends that may be susceptible to the same failure have been inspected for flaws using appropriate non-destructive examination techniques.

4. Necessary corrective actions to ensure safe operation of Palisades during the next . operating cycle prior to returning the plant to service will be completed.

a. Repair of the failed pressurizer safe end. b. Non-destructive examinations of other safe ends potentially susceptible to the same failure cause. Results of the short term corrective actions were transmitted to the NRC in a report dated October 7, 1993.

• RC FOfm JeeA f9-8JI . FACILITY NAME 11 I LICENSEE EVENT REPORT !LERI TEXT CONTINUATION OOCIC:E"T NUMBER 121 LER NUMBEl'I 131 SEQUENTIAL.

YEAR NUMBER u:s. NUClEAA REGUUITORY C0"'"5SlC><

APl"IOVEO OMB '<0. J'60-01C*

EXPIRES: 8/31 i86 RE'\11SION NUMBER PAGE 141 Pal.isades Plant 0500025593 -oo 9 -0 0 0 5 o* O 5 Long Term 5. Corrective actions necessary to ensure long term safe operation of Palisades will be identified.

a. Engineering of additional repairs which may be necessary for long term operation of the pressurizer safe ends will be performed and additional . repairs and

• actions, if required, will be by the end of the next refueling shutdown.

-* -* b. Further evaluation of non-destructive examination techniques in light of the pressurizer safe end crack will be coriducted.

Enhanced ultrasonic techniques will be. emplbyed in an augmented ,inspection program* for safe ends beginning in next refueling shutdown.

• The NRC has requested additional information on the subject in two letters dated October 7, 1993. Additional information concerning this occurrence and our follow-up actions will be contained in follow-up correspondence to the NRC. ADDITIONAL INFORMATION

• Consumers Company submitted to the NRC information concerning this event in. letters dated September 29, 1993, October 1, 1993, October 4, 1993, and October 7, 1993. NRC correspondence on this issue included a Confirmatory Action Letter dated October l,* 1993 and two requests for additional information concerning the safe .end crack dated. October 8, 1993. '