05000255/LER-1993-009, :on 930916,identified Pressurizer Penetration Safe End Crack Which Resulted in PCS Leakage.Crack Initiated Due to PWSCC in HAZ of PORV Inconnel 600 Safe End. Engineering Evaluation of Failure Conducted
| ML18059A520 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 11/18/1993 |
| From: | Roberts W CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| Shared Package | |
| ML18059A518 | List:
|
| References | |
| LER-93-009, LER-93-9, NUDOCS 9312010345 | |
| Download: ML18059A520 (5) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2) |
| 2551993009R00 - NRC Website | |
text
.-...,--*-"'-----*--~--~----...._._._.._,_~--...---.-....~-~** -*---**-
- *I NRC FOl'm 388 U.S. NUCLEAR REOl.JlATORY COMMISSION lt-l!31 APPROVED OMB NO. 31~104 EXPIRES: 8131186 LICENSEE EVENT REPORT (LERI FACILITY NAME 111 DOCKET NUMBER 121 PAGE 131 Palisades Plant olslololol2lsls 1 I OF 0 Is TITLE 141 PRRC:::SURTZFR I'
N~:T.,A.TTfftil SAFE END CRACK RESULTS IN PCS LRAKAGR EVENT DATE 161 LER NUMBER 181 REPORT DATE 181 OTHER FACILITIES INVOLVED 181 SEQUENTIAL I
REVISION FACILITY NAMES MONTH DAY YEAR YEAR NUMBER NUMBER MONTH DAY YEAR N/A 0161010101 I
.01 9
~ 6 91 3 -
01 q 9 -
01 111 118 91 3 9 3 1
N/A o I 6 I o I o I o I I THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR I: fC/tecJr -
°',,_. ol dw fallowing) 1111 OPERATING N
MODE 191 20.4021bl 20.4061cl 60.731all211lvl 73.711bl 20.4061a111 IUI POWER 60.381cll11
- 60. 7 31all21M 73.711cl LEVB.
I i--.-
. 1101 0
20.4061all1 lliil 60.381*1121
- 60. 7 31all211vill
'OTHER (Specify In --
i--
1 ****************)**1~ **** :
- .* j
. 20.4061all1 lliiil 60.731all2101 60.731all211viii1W
- - endlnText,
~
i--
20.4061a1(110vl 60.731ill211iil
- 60. 7 31*112llviii11BI NRC Form 3HAI
?.---
20.4061*11.UM 60.731all211iiil
. 60.731*11211xl LICENSEE CONTACT FOR THIS LER 1121 NAME TELEPHONE NUMBER William L. Roberts Staff Licensin2 En2ineer AREA CODE I
- Sl.1ls 11sl41-lalsl1l3 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT 1131 MAHUFAC*
REPORTABLE *>
MANUFAC,:
REPORTABLE
CAUSE
SYSTEM COMPONENT TUR ER
- To NPRDS
CAUSE
SYSTEM COMPONENT TURER '
TO NPRoS I
I I
I I
I
- 1
'[%
I I
I I
I. I I
I I I I I I I I
I I
I I
I I.
SUPPLEMENTAL REPORT EXPECTED 1141 MONTH DAY YEAR
~NO
- EXPECTED
--, YES Uf yu.,_,,,.,_ EXPECTED SUBMISSION llo4 TEI SUBMISSION I
I I
DATE 11SI ABSTRACT 1Um1t ID 1400 _.
.. I.e., _.,,,;,,,.r.1y flftHn.;,,g1e-_. typewrlttwl -*I 1181 On September 16, 1993, at approximately l900 hours, 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 leak in the power operated relief valve (PORV) line near the nozzle connection to the pressurizer. The plant was returned to cold shutdown.
The crack initiated due primary water stress corrosion cracking (PWSCC) in the heat affected zone (HAZ) of the power operated relief valve (PORV) Inconel 600 safe end.
Cotrective 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 s1fe end to replace the piping that was removed, and examining and evaluating the remaining pressurizer nozzles and other primary cooJ ant system nozzles to provide assurance of operability prior to returning the plant to service.
9312010345 931i1B 127 PDR ADOCK 05000255 ~:
S PDR
~
NRC Form :39e;,,
18*8.31 FACILITY NAME C1I Palisades Plant
EVENT DESCRIPTION
_LICENSEE EVENT REPORT (LERI TEXT CONTINUATION DOCKET NUMBER 121 YEAR LER NUMBER 131 SEQUENTIAL NUMBER
*-*,1 _,,I U.S. NUCLEAR REGULATORY COMMISSION APPROVED OMB NO. 31~1CM EXPIRES: 8131186 REVISION NUMBER PAGE 141.
0 5
0 0
0 2
5 5
9 3
Q Q
9 -
Q 1
Q 2
OF Q
5 On September 16, 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 psi a) when plant operations personnel identified an increasing trend in cgntainment sump level indication, possibly indicating
- a leak in the PCS.
A few mi nut es later, an aux Hi ary operator conducting rounds in the.
containment reported a steam leak near the pressurizer.. Closer inspection found an un-iso~atable leak in the power operated relief valve (PORV) line nea~ the pr!ssurizer 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 nozzle.
On September 17, 1993 the plant achieved cold shutdown and direct vi~ual and NOE examination of the crack are*a was performed.
The leak ~rea 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 ev~nt 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 t.he final 5 to 10% of crack growth being transgranualar.
- During the inservice inspection of this 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 plant 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 plant was returned to a cold shutdown condition. Jransition to cold shutdown required no abnormal operations or any safeguards equipment performing a design basis function.
NRC Fonr. 381A (9-8~1 FACILITY NAME 111 LICENSEE EVENT REPORT (LERI TEXT CONTINUATION DOCKET NUMBER l2l YEAR LER NUMBER 131 SEQUENTIAL NUMBER U.S. NUCLEAR REOUlATORY COMMISSION APPROVED OMB NO. 3160-<>1°'
EXPIRES: 1/31/86 REVISION NUMBER PAGE 141 Palisades Plant 0
5 0
0 0
2 5. 5 9
3 0
Q 9 -. 0 l
0 3 OF Q
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 significance of 1 thi*
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 b~en 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 l~ak increased in size monitoring the increasing leakage fnto 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 le~k 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
Th~ sp~cifi~ 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 i~vestigations were reported to the NRC in a letter dated October 7, 1993. Specific plan actions 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.
.NRC Forin 3HA 111-8.11 FACILITY NAME (1 I LICENSEE EVENT REPORT (LERI TEXT CONTINUATION OOCK£T NUMBER 121.
YEAR LER NUMBER 131 SEQUENTIAL NUMBER U.S. NUCLEAR REGULATORY COMMISSION APflROVED OMBNO. 31~104 EXPIRES: 8/31 /86 REVISION NUMBER PAGE 141 Palisades Plant 0
5 0
0 0
- 2.
5 5
9 3
Q '.
Q 9 -
Q 1
Q 4
OF 0 5 Short Term
.1.
- An engineering evaluation bf the failure has been conducted.*
The specific elements of this engineering analysis included the following:
a.
Meta 11 urgi ca 1 ana 1 ys is of the fa i 1 ed 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.
Pipin~ stresses and weld residual stresses are being evaluated and will be reviewe.d 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 ar~ 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.
simi.lar 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 end, based on the root cause analysis of the failur~, has shown that the lifetime of the repaired saf~
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 been identified.
Other safe ends that may be susceptible to.the same failure.~ave 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.
NRC Fotm 3HA
- 18*&31 U.S. NUCl.EAR REGULATORY COMMISSION APPROVED OMB NO. 31&0-<>1CM EXPIRES: Bl31 /86 FACILITY NAME 111 Palisades Plant Long Term LICENSEE EVENT REPORT (LERI TEXT CONTINUATION DOCKET NUMBER 121 YEAR
. LER NUMBER 131 SEQUENTIAL
.NUMBER REVISION NUMBER PAOE l'I 0
5 0
0 0
2 5
5 9
3. -
Q 0
9 -
0 1 0 5 OF s.
5.: Corrective actions necessary to ensure long term safe operation of Palisades will be identified.
a.
b.
Engineering evaluation of additional repairs which may be necessary for long term operation of the pressurizer safe ends will be performed and.additional
- repairs and corrective actions, if required, will be completed by the end of the next refuel*ing shutdown.
Further evaluation of non-desttuctive examination techniques in light of the pressurizer safe end crack will be conducted.
Enhanced ultrasonic technique~
will be employed in an augmented inspection program for safe ends beginning in the next refueling shutdown.
The NRC has requested additional information on the subject in two letters dated October 7, 1993. Additional information concernfog this occurrence.and our follow-up a6tions will be contained in follow-up correspondence to the NRC.
ADDITIONAL INFORMATION
Consumers Power Company submitted to,the NRC information concerning this event in letters dated September 29, 1993, October 1, 1993~ October 4, 19~3, October 7, 1993, and J October 15, 1993, two letters dated October 20, 1993, October 27, 1993, October 29, 1993, November 1, 1993, Novembet 2, 1993, and.November 15, 1993.
NRC correspondence on this issue included a Confirmatory Action Letter dated October 1, 1993, two requests for additional information concerning the safe end crack.dated J
October 8, 1993, and a letter dated October 29, 1993~