ML20059C673
| ML20059C673 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 08/19/1993 |
| From: | AFFILIATION NOT ASSIGNED |
| To: | |
| References | |
| OLA-2-I-MFP-084, OLA-2-I-MFP-84, NUDOCS 9401060025 | |
| Download: ML20059C673 (31) | |
Text
_. _.
3' Q - 2 'f 5 3 2 3 - OL R YN O
T - IV FP~ WY g Qy p;
/
DC2-91-MM-N069 D14 February 2, 1993 93 m IB *
'O MANAGEMENT
SUMMARY
At 2155 PDT, August 13, 1991, while Unit 2 was operating at approximately 100% power, an Unusual Event was declared due to the results of a routine test on Unit 2 that indicated unidentified leakage from the Reactor Coolant System of approximately 1.4 gallon; per minute (gpm),
which is in excess of the NRC license limit of 1.0 gpm.
In accordance with plant procedures, actions were immediately taken to identify and reduce the leakage to within the limit.
A review of the previous leak check surveillance determined a math error caused the leakage to be under-estimated.
this under-estimate resulted in violation of Technical Specification 3.4.6.2.
The leakage was identified as originating in the Charging Subsystem of the Chemical Volume Control System.
The Charging Subsystem returns coolant to the Reactor Coolant System during normal plant operations following chemical treatment by the Chemical Volume Control System.
At 0019 PDT, August 14 1991, with the leakage path identified and with the leakage below the NRC license limit, the Notification of an Unusual Event was terminated, and all response organizations were notified.
All NRC license requirements were met, and no discharge of I
radioactive materials were made to the environment.
T.m event 11. no way affected the health and satety of the public.
During 2R4, some valve bolting torque was checked and verified and/or reestablished, valve bolting was replaced and valves were inspected.
A program has been developed, to be facilitated through a
" HIT Team," for replacing bolting on valves that now have B7 bolting that should be replaced with 630 SS.
A list of all bolted connections in boric acid service has been developed for an inspection plan for 1R5.
l The remainder of the bolted connections not addressed in 2R4 will be taken care of during 2R5.
> 14 0 2 j) 9 scawesmus*9n Page 1
of 29 9401060025 930819 PDR ADOCK 05000275 g
i
?
s DC2-91-MM-N069 D14 February 2, 1993 CVCS UNIDENTIFIED LEAKAGE I.
Plant Conditions Unit 2 was in Mode 1 (Power Operation) at 100% power.
II.
Description of Event A.
Event:
CVCS Leak On August 13, 1991, at 0700 PDT, Surveillance Test Procedure (STP) R-10C, " Reactor Coolant System Water Inventory Balance," was performed.
The data collected had a calculated unidentified leakage rate of 0.8 gallons per minute (gpm).
)
On August 13, 1991, at 1700 PDT, a limiting condition for operation (LCO) was exceeded when the 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> action b.
of Technical Specification (TS) 3.4.6.2 was inadvertently exceeded with an unidentified leakage of > 1.0 gpm.
On August 13, 1991, at 1755 PDT, STP R-10A/B,
" Reactor Coolant System Leakage Evaluation," was performed.
STP R-10A/B determined containment sump unidentified in-leakage to be 1.2 gpm.
On August 13, 1991, at 2010 PDT, STP R-10C was performed and determined that Reactor Coolant System (RCS) unidentified leakage wac 1.9 gpm.
Excess letdown was isolated to determine if it was the source of the unidentified leakage.
There was no change in the unidentified leakage rate.
Excess letdown was returned to service, d
On August 13, 1991, at 2155 PDT, an Unusual Event (UE) was declared when the completion of the allowed four hour period for mitigation of unidentified leakage in the Unit 2 containment from the RCS in excess of the TS 3.4.6.2 limit of 1.0 gpm was completed.
Operations declared the UE to inform management and the appropriate agencies of the Unit 2 condition.
On August 13, 1991, at 2200 PDT, Abnormal Procedure AP-17, " Loss of Charging," was entered.
mnwwmiuww acN Page 2
of 29 i
,~
t f --
. I...#
- ' f s.-
L.
aw l'
71; Fi;
(.'
D::
. -./
{l b CH :w r5 ?-
s.
?E l O' $ l ',
p'-
._/_._...
~ _ _..
l f?
-a-l6 d
l t
S 4
k'n ksl}ty l'(;Q< ;
_ f I iI ?
)
_. j_.. _ _, _... _
l
.l < I,'., l
-] : l', ',
l 4
i l
i 4
DC2-91-MM-N069 D14 February 2, 1993 j
i The charging header was isolated to determine if it was the source of the unidentified leakage.
Visual inspection observed that the leakage stopped.
An STP R-10C was performed.
On August 14, 1991, at 0019 PDT, the results of STP R-10C showed RCS unidentified leakage to be less than 1.0 gpm.
Thus, it was confirmed that the unidentified leakage was coming from the charging header.
With the leakage path identified j
as originating in the charging subsystem of the Chemical Volume Control System (CVCS) (CB) and with the leakage below the TS limit, conformance with TS 3.4.6.2 was achieved and the UE was terminated.
j On August 21, 1991, Justification for Continued Operation-(JCO) 91-05 was approved to authorize Uni + 2 operation with the unidentified RCS leakage less than (but close to) 1.0 gpm.
On August 31, 1991, Management directed Unit 2 shutdown for a one week early start to the fourth refueling outage due to an increase in the unidentified RCS leakage from 0.78 gpm to > 0.9 gpm.
On September 2, 1991, insulation was removed from the normal charging line and leakage was i
identified at the body-to-bannet joint of valve CVCS-2-8378B, the first-off check valve.
In-place inspection showed that bonnet' stud degradation had occurred and two of the twelve studs had faile' with one more close to failure (tLa third stud failed during valve bonnet disassembly).
Current Engineering judgement is that a small gasket leak developed, possibly aggravated by thermal cycling of the valve.
The small leak initiated stud degradation through steam erosion or boric acid wastage or a combination of the two.
The weakening and eventual failure of the studs relieved the gasket compression and resulted in a progressively increasing leak rate.
Inspection of the alternate charging line check valves showed a similar leak on the bonnet gasket of the first off check valve, CVCS-2-8378A.
The evidence of leakage was similar but not as advanced.
One stud had failed and three were 91NCR%P91MMN069)CN Page 3
of 29 r
e
-u.~
s v-
,----n----
9 DC2-91-MM-N069 D14 February 2, 1993 degraded severely.
The studs showed material degradation at the base of the stud where it enters the valve body.
Engineering judgement considers that degradation characteristic resulted from the presence of borated water leakage on the body joint surface which puddled around the base of the studs.
The effects of steam erosion was not evident on these studs.
The valve history for CVCS-2-8378A is the same as CVCS-2-8378B (the normal charging line first-off valve) and has the same original B7 material fasteners torqued to 83 ft-lb.
On September 9, 1991, JCO 91-06 was approved, authorizing continued Unit 1 operation with enhanced attention to unidentified RCS leakage.
The design of the charging line and the alternate charging line on Unit 1 also includes identical 3 inch diameter Velan model 3C58 check valves.
The valves cannot be inspected at power for the same reasons of insulation coverage and accessibility as was the case for Unit 2.
The maintenance history for the Unit i valves is the same as for Unit 2 with the exception of the Unit 2 second-off normal charaing line valve.
The maintenance history for the Unit 1 charging line check valves shows that the valves have not been retorqued or have not had studs or gaskets changed since initial installation.
The bonnet studs would be expected to be B7 material torqued to the 83 ft-lb value which was in effect at that time and with original flex 1tallic asbestos gaskets.
l The bonnet studs on Unit 1 are the same material as those on the Unit 2 charging lines check valves which experienced severe degradation.
The potential agents to cause degradation are boric acid corrosion and steam erosion.
Boric acid corrosion is most aggressive where there is a concentrating mechanism, where the metal remains wetted, and where the metal temperature is less than 400-500
'F.
The temperature of the check valve for normal conditions (with normal charging and with normal letdown) would be approximately 450 *F (approximately the same temperature as the charging water).
At this temperature the rate of attack would be expected to be low if a borated l
emenwr9t uxsora>cw Page 4
of 29 I
i
i DC2-91-MM-N069 D14 February 2, 1993 water leak were to occur.
The Unit 2 charging line check valve, however, has also operated with normal letdown isolated.
This configuration introduces a thermal transient and an operating temperature in the charging line of about 150
- F.
In the prasence of a leak at this lower temperature the studs may remain wetted and may be subject to aggressive boric acid corrosion rates.
l The presence of steam erosion is also indicated by the preliminary visual inspection of the studs.
The stubby appearance of the failed studs at the i
point of failure can be associated with erosion from a steam jet exiting the gasket and impacting the stud.
For additional information see JCO 91-06.
On October 7, 1991, JCO 91-05 was closed following inspection and repair of Unit 2 charging line check valves.
For more information see JCO 91-05.
STP R-10C Cal _qulation Error On August 16, 1991, engineers reviewing STP R-10C test results determined that an error had been made in the August 13, 1991, 0700 PDT unidentified leakage calculation.
The corrected calculation showed unidentified leakage as 1.4 gpm.
On August 20, 1991, a Technical Review Group (TRG) met and determined that, on August 13, 1991, at 1700 PDT, a Limiting Condition for Operation (LCO) of Technical Specification (TS) 3.4.6.2, Activo b.
was inadvertently exceeded when no action was taken to mitigate the RCS unidentified leakage rate that was greater than 1.0 gpm. The TRG also determined that the event met the criteria of 10 CFR 50.73 (a) (2) (1) (B) as a violation of TS.
B.
Inoperable Stractures, Components, or Systems that Contributed to the Event:
None.
C.
Dates and Approximate Times for Major Occurrences:
1.
August 13, 1991; 0700 PDT:
STP R-10C performed with error and usenwmWWKMKN Page 5
of 29
r
~
DC2-91-MM-N069 D14 l
February 2, 1993 l
determined RCS unidentified leakage of 0.8 gpm.
2.
August 13, 1991; 1700 PDT:
Event date.
LCO inadvertently exceeded when undiscovered leak rate > 1.0 gpm not mitigated and Unit 2 not placed in Hot Standby.
3.
August 13, 1991; 1755 PDT:
STP R-10A/B was performed and showed an unidentified in-1eakage rate of 2
1.2 gpm to the containment sump.
4.
August 13, 1991; 2010 PDT:
STP R-100 performed and showed actual leakage 1.9 gpm.
5.
August 13, 1991; 2155 PDT:
A UE was declared when RCS leakage was not controlled
< 1.0 gpm.
6.
August 14, 1991; 0019 PDT:
Leak controlled below TS limit.
UE terminated.
7.
August 16, 1991:
STP R-10C from August 13, 1991, 0700 PDT found to have a calculation error.
Unidentified leakage was 1.4 gpm not 0.8 gpm.
8.
August 20, 1991; 1100 PDT:
Discovery date.
A Technical Review Group (TRG) met viNcawestmee9xw Page 6
of 29
l s
DC2-91-MM-N069 D14 February 2, 1993 and determined that TS 3.4.6.2, Action b.
had been exceeded.
9.
August 21, 1991:
JCO 91-05 approved.
- 10. September 2, 1991:
Insulation removed from charging line, leak located.
I
- 11. September 9, 1991:
JCO 91-06 approved.
- 12. October 7, 1991:
JCO 91-05 closed.
D.
Other Systems or Secondary Functions Affected:
Seventeen additional Unit 2 RCS and SI valves were inspected and had maintenance done as necessary during 2R4 to ensure their body-to-bonnet joint was leak tight, the studs were sound and the 4
torque on the bonnet nuts was acceptable.
E.
Method of Discovery:
STP R-10C Calculation Error On August 16, 1991, engineerc reviewing STP R-10C test results determined that an error had been made in an wiil'untified leakage calculation.
The calculation done for the August 13, 1991, 0700 PDT 4
test recorded a leak rate of 0.8 gpm.
This calculation was found to have an error.
A TRG convened on August 20, 1991 to review this event for reportability.
It was determined that records show that unidentified RCS leakage i
exceeded 1.0 gpm on or before August 13, 1991 at 0700 PDT and action was not taken to reduce the unidentified leakage rate within four hours or be in Hot Standby within the next six hours.
- Thus, the LCO in TS 3.4.6.2, Action b. was exceeded.
l l
F.
Operators Actions:
91NCRWP91 MMN069 JCN Page 7
of 29 j
DC2-91-MM-N069 D14 February 2, 1993 I
operations isolated charging by closing valve I
CVCS-2-8146 in an attempt to mitigate the leak.
G.
Safety System Responses:
None required.
III.
Cause of the Event A.
Immediate Cause:
Inspection showed a body-to-bonnet gasket leak on valve CVCS-2-8378B as well as a smaller body-to-bonnet gasket leak on valve CVCS-2-8378A.
B.
Determination of Cause:
1.
Human Factors:
a.
Communications: N/A.
b.
Procedures: N/A.
c.
Training:
N/A.
l d.
Human Factors:
The R-10C calculation was simple so a calculator was not used.
This contributed to the factor of ten error, c.
Management System:
N/A.
2.
Equipment / Material:
a.
Material Degradation:
Body-to-bonnet joint relaxation and stud wastage
- occurred, b.
Design: Current design recommends a "Flexicarb" gasket with an increased torque value to better allow for thermal cycling as experienced by letdown isolation.
c.
Installation:
fi/ A,
simtwr9t memxw Page 8
of 29
DC2-91-MM-N069 D14 February 2, 1993 d.
Manufacturing:
N/A.
e.
Preventive Maintenance: The PM program had not yet provided for a periodic re-torquing of I
threaded fasteners.
f.
Testing:
Testing detected the leak.
g.
End-of-life failure:
N/A.
C.
Root Cause:
1.
CVCS Leakane The root cause of the leakage was due to a loss of joint preload due to a combination of low (compared to current vendor recommendations) bolting torque, a low resiliency gasket and thermal cyclic fatigue.
Joint relaxation lead to leakage which, in the presence of relatively low temperatures and B7 bolting material, promoted boric acid attack and stud wastage.
Effect Cause Unit 2 experiences Body-to-bonnet leakage unidentified leakage.
on 8378A and 8378B.
Body-to-bonnet leakage. Joint relaxation permitted leakage.
Joint relaxation.
a.) Gasket creep of low resilience gasket.
b.) Time dependent joint compliance causes a drop in threaded fastener torque.
c.) Boric acid wastage permits degradation of body-to-bonnet joint and increased leakage.
91NCRWP 91MMN069)CN Page 9
of 29 a
I DC2-91-MM-N069 D14 I
February 2, 1993 i
This root cause cannot be eliminated with existing technology, however, the programmatic j
corrective actions developed will minimize the l
probability of future unidentified reactor coolant system leakage due to degraded carbon steel bolted joints.
2.
STP R-10C Calgulation Error l
The root cause of this violation of the LCO l
for TS 3.4.6.2, Action b. was personnel error l
due to inattention to detail and a lack of l
self verification by the operator performing I
STP R-10C calculations.
a.
The ACO made a mental calculation.
This calculation is not a difficult calculation to make and the use of a calculator did not seem warranted at the time.
b.
The calculation was made at the end of a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> graveyard shift.
The ACO was tired and his relief was there waiting for a turnover.
The ACO felt hurried by the circumstances.
c.
The calculated results of the R-10C were' in line with previous R-10Cs, therefore the result was considered reasonable by the ACO performing the test.
D.
Contributory Cause(s):
1.
CVCS Leakaae a.
The service environment of these valves provided for thermal cycling (STPs and letdown isolation) which accelerated joint relaxation.
b.
There is no Mechanical Maintenance program I
for periodically monitoring the tightness of bolted connections and retorquing those joints which have relaxed with time.
I 2.
STP R-10C Calculation Error a.
Shift personnel who reviewed and approved emem9tuusooars Page 10 of 29
__m..
DC2-91-MM-N069 D14 February 2, 1993 STP R-10C results did not perform an adequate verification in accordance with Administrative Procedure A-56, " Signatures and Signature Responsibility."
b.
STP R-10C results as recorded were reasonable and consistent with value previously collected.
IV.
?nalysis of the Event A.
Safety Analysis:
The inadvertent violation of TS 3.4.6.2 resulted from a calculation error during the perl umance of STP R-10C on August 13, 1991 at 0700 PDT.
When the observed leak rate did exceed the TS limit at 1755 PDT on the same day, appropriate investigative actions were performed to identify the source and mitigate the leakage.
On September 2, 1991, following orderly shutdown of Unit 2 on August 31, 1991, the lagging was removed from the normal charging line in the vicinity of the check valves and the leak site was located.
The leak location site was observed to be in the body-to-bonnet gasket on the charging line check valve closest (first-off) to the RCS, CVCS-2-8378B.
A body-to-bonnet gasket leak is typically small and once initiated increases g.adually with time.
This charac cristic was apparent with this leak.
PG&E considers the body-to-bonnet gasket leakage would not have produced an abrupt release of reactor coolant.
This evaluation is supported by EPRI Report No. NP-5769, April 1988.
This technical program analyzed various primary pressure boundary closures including check valves.
Pressure boundary bolting integrity had the highest priority in that industry program.
Based on the program results, reactor coolant pressure boundary joint degradation was determined not to be a safety issue.
Therefore, the body-to-bonnet gasket leak does not generate significant safety concerns due to rapid RCS depressurization and impact on core cooling.
J 9tscawretuusanoxs Page 11 of 29
DC2-91-MM-N069 D14 February 2, 1993 Ample time is available to detect leakage, evaluate it, and take the proper and prudent compensating action.
The leakage rate was carefully monitored and the rate of increase was evaluated by the trend of the results of STP R-10C.
When the detected leak rate did exceed TS limits, appropriate actions were taken to mitigate the leak rate.
All leakage was contained in the Unit 2 Containment.
No radioactive materials were released to the environment.
JCO 91-05 was issued to address Unit 2 continued operation with valve CVCS-2-RV-8117 leaking.
On August 12, 1991, while restoring the letdown system to service following the performance of STP-V-3K7A, relief valve CVCS-2-RV-8117 was inadvertently lifted, and did not rescat.
Repeated attempts to induce RV-8117 to reseat properly were unsuccessful, consequently the normal letdown system was not restored to service, and Unit 2 continued to operate on excess letdown.
As a compensatory measure PG&E performed surveillance of the RCS leakage every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, as opposed to the TS requirement of every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
JCO 91-05 has a dedicated safety evaluation.
For more information see JCO 91-05.
JCO 91-06 was issued to address continued Unit 1 operation without detailed, sp'ecific inspection of first-off RCS check valves.
There are two non-return valves installed in 'he normal and ir 'he j
alternate charging lines close to the point where the lines connect to the Reactor Coolant System j
(RCS) cold leg piping.
Both charging lines use the same design, 3 inch diameter swing disc non-return check type valves, Velan model 3C58.
Unit 2 had experienced RCS leakage in this area of the piping of the normal charging line.
Inspection during the refueling outage found the leak site to be the body to bonnet gasket joint on the valve closest to the RCS.
A similar leak was found in the bonnet gasket of the corresponding check valve in the alternate charging line.
The design of the charging line and the alternate charging line on i
Unit 1 also includes identical 3 inch diameter Velan model 3C58 check valves.
The maintenance 9l NCRWP.91MMNOM.JCN Page 12 of 29 i
i
.-v..,.
---e
,.-s,
~.,
~.
DC2-91-MM-N069 D14 February 2, 1993 history for the Unit i valves is the same as for Unit 2 with the exception of the Unit 2 second-off normal charging line valve.
The Unit 1 valves cannot be inspected for the same reasons of insulation coveraga and accessibility as was the case for Unit 2.
Although there is no knowledge of current body to bonnet leakage of the Unit 1 valves, this JCO was prepared to identify prudent compensatory actions in response to the potential occurrence of a similar leek in Unit 1.
JCO 91-06 I
has a dedicated safety evaluation.
For more information see JCO 91-06.
Thus, the health and safety of the public were not adversely affected by this event.
B.
Reportability:
1 1.
Reviewed under QAP-15.B and determined to be non-conforming in accordance with Section 2.1.3.
2.
Reviewed under 10 CFR 50.72 and 10 CFR 50.73 per NUREG 1022 and determined to be reportable as a violation of TS in accordance with 10 CFR 50.73 (a) (2) (1) (B).
See LER 2-91-004 for more information.
3.
This problem does not re3 aire a 10 CFR 21 report.
4.
This problem does not require reporting via un INPO Nuclear Network entry at chis time.
5.
Reviewed under 10 CFR 50.9 and determined to be not reportable since this event does not have a significant implication for public health and safety or common defense and security.
6.
Reviewed under the criteria of AP C-22 requiring the issue and approval of a JCO and determined that a JCO is required.
JCOs 91-05 and 91-06 were generated to address the conditions that led to this NCR.
V.
. Corrective Actions 91scnwestuusce9xw Page 13 of 29
l I
DC2-91-MM-N069 D14 i
February 2, 1993 i
Although the root cause of the condition identified in this NCR was determined to be a combination of several i
factors leading to leakage of the body-to-bonnet j
conncction of a first-off (nonisolable) RCS check valve, it was decided by the TRG to conservatively
]
apply this root cause to all first-off valves with bolted body-to-bonnet connections.
This conservative action was decided on since any leakage from a first-1 off RCS valve is nonisolable and could potentially require a plant shutdown to pcrmit repair.
The effects of similar conditions (i.e.,
boric acid leakage) on second-off valves are not potentially as severe since the first-off valve can be utilized to j
isolate leakage from the valve (i.e., RCS) and thereby minimize the potential for bolting corrosion and nonisolable RCS leakage.
I A.
Immediate Corrective Actions:
}
Charging was isolated by closing valve CVCS-2-8146 j
in an attempt to mitigate the leak (ref. 1).
i l
B.
Investigative Actions:
I j
1.
A containment entry was made to determine the j
location of the leak (ref. 2) 2.
Plant Engineering will determine the magnitude of the thermal cycles charging line 2-S6-246-3 has experienced in service for comparison with a fatigue calculation for the limiting point in that line.
RESPONSIBILITY:
S.
Chesnut COMPLETE Plant Engineering (PTEP)
AR A0238965. AE # 01 3.
Radiographs for charging line 2-S6-246-3 will be retrieved for review to ensure there are no unidentified flaws.
RESPONSIBILITY: K.
Palmer COMPLETE Mechanical Maintenance (PGMA)
AR A0238965. AE # 02 4.
Radiographs for charging line 2-S6-246-3 will be reviewed between the RCS loop four back to the second-off check valve to ensure there Simm9tuumxw Page 14 of 29
J DC2-91-MM-N069 D14 February 2, 1993 were no unidentified flaws.
RESPONSIBILITY:
D. Gonzalez COMPLETE ISI (PTPI)
AR A0238965. AE # 03 5.
Mechanical Maintenance will coordinate the development of an interdisciplinary action plan to investigate charging line 2-S6-246-3 and the vicinity about the line to determine the source of the leakage in the event of a forced outage prior to 2R4.
RESPONSIBILITY:
R.
Powers COMPLETE Mechanical Maintenance (PGMT)
AR A0238965. AE / 04 i
6.
Mechanical Maintenance will coordinate the development of an interdisciplinary action plan to investigate charging line 2-S6-246-3 l
and the vicinity about the line to determine the source of the leakage during Mode 3 and repair it during 2R4.
RESPONSIBILITY:
R.
Powers COMPLETE Mechanical Maintenance (PGMT)
AR A0238965. AE # 05 7.
Plant Engineering will develop a statement providing the reason (s) why thermal cycling was not the leak causative factor.
RESPONSI;:7 TTY:
S.
Chesnut COMPLETE Plant Engineering (PTEP)
AR A0238965. AE # 06 8.
NECS - Engineering will develop statements providing the reason (s) why high cycle fatigue and erosion / corrosion were not leak causative factors.
RESPONSIBILITY:
R.
Klimczak COMPLETE NECS - Engineering (NCFD)
AR A0238965. AE # 07 9.
NECS - Engineering will develop a statement providing the reason (s) why low cycle fatigue was not the leak causative factor.
91NCRWP\\91MMfM9 JCN Page 15 of 29
l
\\
DC2-91-MM-N069 D14 February 2, 1993 RESPONSIBILITY:
R.
Waltos COMPLETE NECS - Engineering (NCFM) i AR A0238965. AE / 08
- 10. Mechanical Maintenance will develop statements providing the reason (s) why material defects and workmanship (welding) were not leak causative factors.
I RESPONSIBILITY: K.
Palmer COMPLETE Mechanical Maintenance (PGMA)
AR A0238965, AE / 09
- 11. Verify the material the studs are made of and determine if the wastage was caused by j
corrosion or erosion.
This study to be done by Failure Prevention, Inc.
RESPONSIBILITY:
R.
Nanninga COMPLETE Mechanical Maintenance (PGMA)
AR A0238965, AE # 12
- 12. Investigate CMTRs or other procurement i
paperwork to determine what material was specified and certified for the bolting material of the valves which have exhibited leakage.
RESPONSIBILITY: K.
Palmer COMPLETE Mechanical Maintenance (PGMA)
AR A0238965, AE # 13
- 13. Ensure the work orders ('40s ) for the val" n which fail the leakage criteria have instructions for collecting data during disassembly such as breakaway torque, gap measurements and cross section measurements similar to the actions in the WOs for valves CVCS-2-8378A & B.
RESPONSIBILITY:
R.
Powers COMPLETE Mechanical Maintenance (PGMT)
AR A0238965, AE # 14
- 14. NECS - Engineering will perform a safety margin calculation to verify that the valves that had stud wastage had adequate safety margin.
AE completed for preliminary data, eiscR WP.91MMN(rr9.JCN Page 16 of 29
.-1
-7
DC2-91-MM-N069 D14 2
Februar'f 2, 1993 then reassigned at November 15, 1991 TRG for recalculation based on revised stud measurements.
RESPONSIBILITY:
C.
Nichols COMPLETE NECS - Engineering (NCED)
AR A0238965, AE # 15
- 15. NECS - Engineering to identify which first-off check valves are subject to thermal cycling due to normal system operation.
j RESPONSIBILITY: R. Klimczak RETURN NECS - Engineering (NCFD)
AR A0238965, AE # 26
- 16. Provide a complete list of first-off ASME piping class I valves with flange body-to-bonnet connections that will have bolting replacement and retorquing performed during 2R4, 1R5 and 2RS.
l RESPONSIBILITY:
R. Waltos ECD: 2/1/93 Mechanical Maintenance (PGMA)
AR A0238965, AE # 27
- 17. Mechanical Maintenance to provide a list of code class II components in borated service in containment.
This is required to support prudent action D.1.d.
RESPONSIBILITY:
R. Nanninga RETURN Mechanical Maintenance (PGMA)
AR A0238965, AE # 19
- 18. Mechanical Maintenance will verify which Unit 2 first-off valves have not had B7 studs replaced with 630.
Replacement of the studs with 630 material will be performed during 2RS and tracked under corrective action C.1.d.
RESPONSIBILITY:
R. Nanninga RETURN Mechanical Maintenance (PGMA)
AR A0238965, AE # 20
- 19. Prepare a visual inspection plan for inspection of the four Velan charging line check valves in Unit 1, to be conducted in the 9nGWPM MMNMlCN Page 17 of 29
--. _ - _ _ - ~
I DC2-91-MM-N069 D14 i
February 2, 1993 event of an unscheduled Mode 3, 4 or 5 outage.
RESPONSIBILITY:
D.
Gonzalez RETURN
)
Inservice Inspection (PTPI)
AR A0238965, AE # 22 1
C.
Corrective Actions to Prevent Recurrence:
1.
CVCS Leakaoe All nuts and studs that exhibit wastage a.
will be replaced as well as the body-to-bonnet gasket.
Valves that have low torque on the body-to-bonnet nuts and no observably leakage will have one nut at a time removed, cleaned, lubricated and torqued to the current vendor recommended value.
See AR A0238965, AE # 17 (ref. 3) for the selection criteria for this
- activity, b.
Per the September 18, 1992 TRG, the action recommended by NECS - Engineering in AE #
17 to develop a PM program for retorquing bolted connections is being considered a prudent action (reference prudent action D.1.c.).
c.
Confirm completion of inspection and remedial actions identified in ARs A0248563, A0251722 and A0258088 during IRS.
RESP:
R. Waltos (PGMA)
COMPLETE Mechanical Maintenance AR A0238965, AE / 23 Outage related; 1RS.
No JCO required.
Not an NRC commitment.
Not a CMD commitment.
d.
Replace B7 bolting in Unit 2 valves in class I piping systems that were not replaced during 2R4.
RESP:
R. Waltos (PGMA)
ECD: 5/7/93 Mechanical Maintenance AR A0238965, AE # 24 91NCRWI*.91MMNte9 xw Page 18 of 29
DC2-91-MM-N069 D14 February 2, 1993 Outage related; 2R5.
No JCO required.
Not an NRC commitment.
Not a CMD commitment.
2.
STP R-10C The personnel involved were counseled on the need for attention to detail and self-verification.
D.
Prudent Actions:
1.
CVCS Leakace Although there is no written instruction a.
or tracking mechanism, B7 bolting in boric acid service valves is being replaced on in-service valves as well as newly procured valves before they are installed.
At this time (October 24, 1991) there is 1
no commitment to time-table to complete replacing B7 bolting.
b.
A valve bonnet bolting retorque initialization program will be developed jointly by Mechanical Maintenance and NECS which will establish that a minimum
)
acceptable preload exists on each bolted joint as a baseline for a future program of periodic retorquing.
This program will include valves in ASME class I piping syste s.
RESPONSIBILITY:
R. Nanninga RETURN Mechanical Maintenance (PGMA)
AR A0238965, AE # 16 f
c.
A future program for inspection-retorquing-replacement of bolted connections in boric acid service will be developed for B7 bolting in containment.
RESPONSIBILITY:
R. Nanninga COMPLETE Mechanical Maintenance AR A0289139 d.
A comprehensive program will be developed 9mcR%T*.9 t MMNCm9)CN Page 19 of 29
...=.
. _ ~
~
DC2-91-MM-N069 D14 February 2, 1993 to identify which bolted connections in boric acid service in containment will be part of the inspection / replacement
- program, i.e. those bolted connections which are most sensitive to boric acid wastage.
j RESPONSIBILITY:
V.
Vanderzyl RETURN l
NECS Engineering (NCED) 1 AR A0238965, AE # 18 j
l e.
Mechanical Maintenance will update the list of valves attached to STP R-8C,
" Containment Walkdown For Evidence Of Boric Acid Leakage," to delete valves which have had B7 studs replaced with 630.
RESPONSIBILITY:
R. Nanninga PETURN Mechanical Maintenance (PGMA)
AR A0238965, AE # 21 VI.
Additional Information l
A.
Failed Components:
The B7 bolting of the body-to-bonnet connection of j
primary systems valves with insufficient installation torque.
i B.
Previous Similar Events:
None documented.
C.
Operating Experience Review:
1.
NPRDS:
None.
2.
NRC Information Notices, Bulletins, Generic Letters:
a.
IEIN 86-108, Supplement 2,
" Degradation of Reactor Coolant System Pressure Boundary Resulting from Boric Acid Corrosion."
b.
Generic Letter 88-05, " Boric Acid Corrosion of Carbon Steel Reactor Pressure HNCRWrMON9 JCN Page 20 of 29 1
,o r
-,-,-._---4 4
w,--
ni
- -.m.
m
DC2-91-MM-N069 D14 February 2, 1993 Boundary Components in PWR Plants."
3.
a.
SOER 84-5, " Bolt Degradation or Failure in Nuclear Power Plants."
b.
SER 35-87, "Non-isolable Reactor Coolant System Leak."
D.
Trend Code:
MM (Mechanical Maintenance) - B2 (procedure incomplete).
E.
Corrective Action Tracking:
1.
T~4e tracking action request is A0238965.
2.
The corrective actions are outage related.
F.
Footnotes and Special Comments:
None.
G.
References:
i 1.
Initiating ARs A0238817 and A0238818.
2.
Work order C0090298.
3.
AR A0238965, AE / 17.
4.
Failure Prevention Inc. Report, dated December 19, 1991.
H.
TRG Meeting Minutes:
1.
On August 20, 1991, at 10:00 a.m.
PDT, in Room 424 of the Administration Building the initial TRG for NCR DC2-91-MM-N069 met.
The scope of the NCR will include actions necessary to repair whatever is leaking in the RCP 2-4 space.
The highest probability is the charging line 2-S6-246-3, downstream of the second-off check valve, is leaking at one of the two check valves or one of the four vent i
valves.
9 tNCRWl%91MhCM9Jm Page 21 of 29 t
--..,.w a-
j DC2-91-MM-N069 D14 February 2, 1993 The first-off check valve (CVCS-2-8378B) has not been touched since 1982.
Inspection and refurbishment of this valve is scheduled for 2R4.
The second-off check valve (CVCS-2-8379B) was refurbished during 2R3.
The B-7 bonnet studs were replaced with A564 TY630 studs and torqued to 170 ft-lbs (vs original torque of 83 ft-lbs).
This event is reportable, not for the leak but for the evidence that the leak was detected to have occurred some time prior to 0700 PDT, August 13, 1991.
Steps to mitigate the leak were not taken until 2230 on August 13, 1991.
The applicable TS (3.4.6.2 b. Action b.)
requires Hot Standby within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.
The root cause of the TS violation was personnel error in that a math error was made in the manual R-10C calculation on the morning of August 13, 1991.
The incorrect result of the calculation was consistent with the previous unidentified leakage which contributed to both the operator who made the calculation and the STA who reviewed the calculation accepting the result without question.
The proposed initial corrective actions were to: (1) Perform a general review of STP R-10C with Operations personnel' emphasizing the need for attention to the calculation and (2)
Counseling of the individuals involved in the i
calculation and its review.
j The initial leak was determined to be only liquid water that was relatively cool and I
uncontaminated, implying it was from the CVCS and not the RCS.
Charging was isolated and the leak essentially stopped.
It was decided the valve CVCS-2-8146 would be closed and gagged.
This controlled the unidentified leakage to less than 1.0 gpm.
Additional details of this event can be obtained from JCO 91-05.
The area of the leak currently has a two phase nature (liquid and steam) and is contaminated.
This implies that 91senw9 uume;cn Page 22 of 29 9
0 w,
..n.
-.,n.---.---.. - - - -.,.,
---.,o
,--.,.rn--
~
_ - - ~.
i DC2-91-MM-N069 D14 February 2, 1993 I
the CVCS pressure is isolated and there is back-leakage from the RCS.
With letdown line failures, excess letdown has been put into service in the past.
This results in a thermal transient of approximately 350 F.
It is possible that the accumulation of stress cycles could have fatigued a weld or fitting in the charging line resulting in a crack, This will be investigated.
As noted in the investigative actions, statements will be developed and included in this NCR why the various possible failure
]
mechanisms are unlikely for both Units 1 & 2.
Due to the difficult access and the mirror insulation on the lines in the area of the leak, the exact location of the leak cannot be determined with the Unit at power.
The TRG will reconvene when information becomes available on the leak source or if any significant changes to the leak occur.
2.
On September 16, 1991, at 1:00 pm PDT in Room 604 of the Administration Building, the TRG reconvened to evaluate the information
- ollected to date.
i Field inspection has found a number of body-to-bonnet etuds severed due, apparently, to wastage caused by boric acid.
These studs are made of B7 (carbon steel).
All primary system valves that have the i
potantial to have a leak of boric acid bearing fluid through a bolted flange connection have been considered.
An inspection and retorquing plan has been developed.
If the vendor's i
recommended torque value is greater than the torque value maintenance records show for the as-left condition and there is evidence of leakage at the valve, the bolting material and the gasket will be replaced and the studs will be torqued to the vendor's recommended valtle.
91NCRWM91MMNG59.JCN Page 23 of 29
'-e-rv s-mm--+
r-e
-%am.w.wam er-
-"m F-'-'-*m
- +mw-N
-c' 1P---*M==
DC2-91-MM-N069 D14 February 2, 1993 If a valve on the developed "of interest" list does not show evidence of leakage, the nuts will be removed one at a time, the studs lubricated and the nut replaced with the vendor recommended torque value.
AR A242040242040lists the valves of interest.
The as-found torque seems to be a function of the operating temperature of the valve.
High temperature normal operation seems to correlate to a relaxation of the joint clamping force and produces low as-found breakaway torque.
To date, five valves have been found with evidence of leakage.
1 A discussion was held on the advi.sabilit'f of
{
replacing the B7 carbon steel studs with 630 stainless steel (SS).
The B7 studs tend tc l
increase clamping force with increasing temperature due to a difference in thermal expansion coefficient between carbon steel and SS.
This increased clamping force at temperature would tend to reduce leakage probability for B7 stud use.
The only drawback of the B7 material is when there is leakage of boric acid bearing fluid that will ctuse wastage.
If the installation totyue value is sufficient, the B7 studs have sufficient clampf ~ force and no leakage will occur.
There are three considt.. nations that influence whether a valve will leak at the body-to-bonnet joint:
1.
gasket relaxation, ii.
operational thermal cycling and lii.
installation bolt torc.uing at a value lower tnan the current vendor recommendation.
The TRG considered potential corrective action such as revising STP R-10C for Unit 2 and the procurement of class I valves sith B7 bolting.
91NCR%791MMN069 JCN Page 24 of
,9
DC2-91-MM-N069 D14 February 2, 1993 The TRG will reconvene on Thursday, September 26, 1991 at 1:00 pm PDT to discuss additional i
as-found information for the valves of interest and identify a root cause for the leakage.
Reservation should be made for a large meeting room and Operations and W.
Barkhuff need to attend the meeting.
3.
On October 1, 1991, at 1:00 pm PDT in the Canyon Room of the Administration Building the TRG reconvened to discuss the status of the valve inspections.
Attachment One shows the status of the Unit 2 i
valve inspections as of October 1, 1991.
The TRG will reconvene on Tuesday, October 15, to discuss the valve inspections and 19 A results prior to the completion of 2R4.
4.
On October 22, 1991, at 1:30 pm PDT, in Room 533 of the Administration Building, the TRG reconvened to discuss bolted connection preload relaxation with time.
Information and event chronology of the valve inspections and concerns have been incorporated into the text of this NCR writeup.
The focus of this NCR has chifted significantly since it was originated.
Initially this NCR addresses a missed LCO due to a calculation error and unidentified RCS l
leakage in Unit 2.
Currently, this NCR addresses valve body-to-bonnet bolting and a program to periodically retorque valve bolting in both Unit 1 and Unit 2.
Historical information will be retained in this writeup.
Mechanical Maintenance and NECS Engineering will develop a program to baseline and maintain the torque of valve bolting as a leak prevention measure.
i This TRG will reconvene on November 15, 1991, l
at 10:00 am PST to discuss the torque initialization program.
i S.
On November 15, 1991, at 10:00 PST in Room 604 j
MNCRWP91MMN009KN Page 25 of 29
--._-.-,-. __. = _ _ -. _
'l DC2-91-MM-N069 D14 February 2, 1993 of the administration building the TRG reconvened to discuss the development of a program to baseline.the soundness of existing 4
bolted connections with B7 bolting in boric acid service in containment.
Additionally a periodic reinspection program similar to the c
five year program recommended by Velan valves will be evaluated.
f Mechanical Maintenance will provide a list of Code class II components in boric acid service in containment.
NECS Engineering will provide a program for inspection /retorque/ replace for B7 bolted connections of Code class I j
components in boric acid service in i
containment and the criteria for determining the sensitivity of the joints to wastage or deterioration.
l l
As a minimum, the first off check valves from the reactor coolant system will be inspected and B7 bolting will be replaced since they are j
the most sensitive to wastage in containment.
1 The TRG will reconvene on Thursday, January 16, 1992 at 10:00 PST to discuss progress of the actions underway.
6.
On January 16, 1992, in room 533 of the administration building at 10:00 PST, the TRG reconvened to discuss the results of the Failure Prevention Inc. (FPI) analysis, j
Attachment 3., of the valve stud sent to them.
FPI concluded that the wastage was due to general corrosion.
FPI did not have the drawing prepared by Will Barkhuff however.
From the drawing, it appears that erosion certainly was a factor.
l NECS will provide analysis of the acceptability of leaving particular stuck B7 studs in place.
Mechanical Maintenanr9 will be retorquing the bolting of identified valves and flanges during the upcoming outages and replacing the B7 studs of any valve opened for inspection with 630 studs.
Large scale B7 stud replacement will be delayed until the RTD manifolds are removed to support ALARA.
9tscawn91Mumxs9xw Page 26 of 29 4
T 4
F DC2-91-MM-N069 D14 February 2, 1993 i
Mechanical Maintenance will verify which first-off valves have not had B7 studs replaced with 630 studs.
Mechanical Maintenance will also update the list of valves from STP R-8C to delete those that have had B7 studs replaces with 630 studs.
AEs were added to AR A0238965 to track these two activities.
The pressurizer manway studs are B7.
It was recommended that these studs be retorqued during 1R5 and 2R5.
AEs # 17 and # 18 were revised to state the scope of AE # 17 was only 1R5 and 2RS and AE #
18 was 1R6 and 2R6 and beyond.
The TRG will reconvene on February 19, 1992 at 10:00 PST to discuss the lists of identified i
valves for stud changeout.
The estimated closure date of this NCR is December 1,
1992.
7.
On February 19, 1992, the TRG reconvened in room 533 of the administration building at 10:00 am PST to discuss the status of the lists of valves to be inspected / reworked.
As i
noted in the AEs to the tracking AR, these lists have been updated.
A new action item for preparation of an unscheduled outage action plan for Unit 1 was determined following the TRG meeting and has been includad above as AE / 22 to AE A0238965.
The TRG will reconvene on April 15, 1992 to agree on and finalize the inspection / rework scope for 2RS, the long-term program plan and the class II valve list.
8.
On May 13, 1992, at 1:00 pm PDT in room 533 of the administration building the TRG reconvened to discuss the progress of the long term corrective actions.
Revisions were incorporated into the correctivc actions as noted above to update their status.
Additional corrective actions were identified and assigned, as noted above.
91NCR%791MMN069)CN Page 27 of 29
\\
. - _. ~. -
4 t
DC2-91-MM-N069 D14 February 2, 1993 The TRG plans to reconvene on July 8, 1992 to discuss the progress in developing a long term program for maintaining the integrity of components in boric acid service.
9.
On July 16, 1992, at 10:00 am PDT in room 424 of the administration building, the TRG reconvened to review the status of corrective actions.
During 2R4, some valve bolting torque was checked and verified and/or reestablished, valve bolting was replaced and valves were inspected.
A program has been developed, to be tacilitated through a " HIT Team," for replacing bolting on valves that now have B7 bolting that should be replaced with 630 SS.
A list of all bolted connections in boric acid service has been developed; by system, component and by bolt material.
This list will be forwarded to NECS Engineering for the development of a " program," that will address all bolted connections inside containment for the further development of an inspection plan for 1RS.
The remainder of the bolted connectiona not address in 2R4 will be taken care of during 2RS.
This TRG will reconvene following 1R5 to discuss the results of inspections and actions performed during that outage.
The ECD for this NCR is April 1, 1994.
- 10. On September 18, 1992, the TRG reconvened in room 527 of the administration building at 1:00 pm PDT to discuss reorganizing investigative, corrective and prudent actions as shown above.
Revisions and additions were made as shown above.
Overall ECD for NCR closure 91NCRwP.91MM?KM JCN Page 28 of 29
F 4
l DC2-91-MM-N069 D14 February 2, 1993 j
is 7/1/93.
The NCR will be sent to PSRC and this TRG is not exp':cted to reconvene.
I.
Remarks:
None.
J.
Attachments:
1.
2R4 B-7 BOLTING RETORQUE/ INSPECTION LIST 2.
QCR-91-003, CHECK VALVES 8378 A & B (UNIT 2)
STUD DAMAGE i
91NCRwB91t(MNt29)CN Page 29 of
'29 I
,. _ -,.