ML20041B073
| ML20041B073 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 02/19/1982 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20041B072 | List: |
| References | |
| NUDOCS 8202230238 | |
| Download: ML20041B073 (25) | |
Text
_ _ - _ - _ _ _ - _ _ _
SAFETY EVALUATION OF REACTOR WATER CLEANUP SYSTEM PIPE REPAIRS COMMONWEALTH EDISON COMPANY DOCKET NO. 50-265 QUAD-CITIES STATION, UNIT 2 License No. DPR-30 1.0 Introduction Since start-up on December-26, 1981, following a 15-week refueling outage, the Quad-Cities Station, Unit 2 had been monitoring a slowly increasing rate of leakage from an unidentified source in the drywell.
On January 2, 1982, an increase of 0.5 gallons per minute was observed.
On January 15, 1982, the 2B Raactor Recirculation MG Set was tripped
-due to a problem with the exciter brushes. At that time the licensee decided to identify the existing drywell leakage, even though the leak rate was within the Technical Specification limit of five gpm (for leakage from an unidentified source).
While in the drywell, the Operating Engineer noticed that water was leaking from the insulation around the Reactor Water Cleanup (RWCU)
System suction piping.
It was determined that a pipe to pipe butt weld (Weld S14) downstream of the inboard isolation valve was the source of the leak. The unit was shut down, t'202230238 820219 PDR ADOCK 05000265 P-PDR l
. The RWCU System is a filtration and ion exchange system for maintaining the purity of the water in the reactor vessel and recirculation lines. Water purity is necessary to reduce the 6eposition of water impurities on the fuel surface, thus minimizing any effect on heat transfer, and to reduce secondary sources of beta and gamma radiation by removing the corrosion products, impurities, and possible fission products from the primary system water.
The system is normally operated continuously during all phases of reactor plant operation, startups, shutdown, and refueling. While the RWCU System serves no safety function, automatic isolation of the system is necessary to maintain drywell integrity and to ensure proper operation of the Standby Liquid Control System.
The RWCU System pumps take reactor water from the "A" recirculation pump suction line and discharge it through a series of regenerative and non-regenerative heat exchangers where the temperature is reduced to less than 150*F.
At this point, the reactor water is passed through filter /demineralizer units, and impurities are removed. The flow path continues back through the regenerative heat exchangers where the water is heated before returning to the reactor vessel through one of the feedwater lines. An additional source of water is provided by a line from the vessel bottom head drain. Water is removed from the bottom head to provide mixing and prevent thermal stress caused by the colder water in this area.
It is also used to remove crud from the bottom of the vessel. A temperature element is provided to give bottom head drain temperature indication.
. The RWCU System piping penetrates the drywell as the inlet water is taken from the recirculation loop and the outlet water returns to the feedwater line. The inlet line is equipped with motor operated inboard and outboard drywell isolation valves. The return line is equipped with one motor operated outboard isolation valve because it penetrates the feedwater line.
At the time the facility was shutdown on January 16, 1982, the leakage rate into the drywell from an unidentified source was about 2 gpm.
The Icaking water was collected in the drywell drain system, from whence it was pumped to the plant's radioactive waste treatment system for processing. There was no release of radioactivity to the environment as a result of the leakage.
Ultrasonic examinations were performed to determine the crack extent in the leaking butt weld.
In addition, two adjacent piping butt welds were ultrasonically examined. As a result of discovering linear indications on these two welds, the inspection program was expanded to inspect all butt welds in the RWCU System piping up to the outboard isolation valve. These examinations identified that: (1) eight out of eleven butt welds downstream of the inboard isolation valve had ultrasonic reflectors indicative of cracks, (2) four out of ten butt welds upstream from the inboard isolation valve showed ultrasonic l
_J
indications, and (3) none of the four welds outside the drywell had unacceptable indications.
While the UT inspectors were performing inspections on the piping butt welds, one of the inspectors observed a drop of water on the socket weld which attaches the 2-inch reactor drain piping to the sock-o-let on the RWCU System piping (Weld FIA).
Subsequent visual and high temperature penetrant examination failed to reveal any linear indications indicative of a through-wall crack.
Radiation levels were very high in the vicinity of the RWCU System piping (400-800 mR per hour), and required 17.9 Man Rem to complete the ultrasonic inspection program.
By letter dated January 22, 1982, Region III confirmed an understanding with the licensee that repair of the non-isolatable portion (upstream of the inboard isolation valve) of the RWCU System line would not be initiated without the concurrence of the NRC Region III Regional Administrator's office.
In a telephone discussion on January 23, 1982, and in a meeting on January 26, 1982, Commonwealth Edison described their proposed repair program and assessment of safety considerations during the repair.
2.0 Description of the Proposed Repair Design and Installation The modification program consists of replacing the existing 6" NPS schedule 80S RWCU System piping inside the drywell from the inboard isolation valve M02-1201-2 downstream to the drywell penetration.
Becausa there is not full core offload capacity at present, the licensee proposed a program to repair the crack indications which does not involve pipe replacement. Repairs will be made between the tie-in to the 20" recirculation piping and the inboard isolation valve M02-1201-2, shown in Figure 1.1.
Figure 1.2 shows an isometric drawing of the RWCU System piping, and identifies locations and types of cracks subject to repair. The modifications are required to correct certain conditions in the RWCU System piping and allow the plant to resume normal operation until the next regularly scheduled refueling outage, planned for the spring of 1983.
The existing pipe material is ASTM A312, TP 304. The existing fittings are ASTM A403, Gr. WP304. The valve body is ASTM A351, Gr. CF8M.
The 6" NPS schedule 80S piping from the inboard isolation valve downstream to the drywell penetration will be replaced with ASME SA-312 type 304L and have mechanical properties of type 304. The fittings will be ASME SA-403 Grade WP304 with carbon content limited to 0.035% maximum.
The pipe to pipe weld joint location F6 containing the circumferential crack indication will be repaired using a sleeve shown in Figure 2.1.
The sleeve will provide an independent pressure boundary around and to either side of weld F6.
. The suspected leak in the 2" side of the sock-o-let (location FIA) will be repaired with a sleeve as shown in Figure 2.2.
This sleeve will provide an independent pressure boundary around the entire existing sock-o-let by attaching to the 6" RWCU System piping and to the 2" branch piping above the sock-o-let.
The sleeves for F6 and FIA will be machined from wrought material, split, and welded around the pipe using full penetration welds. The sleeve ends will be welded to the pipe using partial penetration welds with cover fillets.
The material for the sleeves will be type 304L stainless steel with 0.035% maximum carbon content and having mechanical properties of type 304. This material is highly resistant to IGSCC in the welded condition.
The longitudinal crack indications in the elbows at locations S9, S10, and F12 will be repaired by establishing an overlay of weld metal deposited 360* around and to either side of the existing welds, as shown in Figure 2.3.
The weld deposited bands over the longitudinal crack indications are similar to sleeves and provide independent pressure boundaries around'the existing welds. Weld metal will be type 308L, which is resistant to propagation of IGSCC cracks.
. The weld overlay. repairs to the axial flaws will be performed by depositing circumferential stringer beads. The deposition will be accomplished in layers with each layer completed before beginning the-next layer. The overlay will be applied at low heat input and the water present inside the pipe will produce rapid cooling..The degree of sensitization produced by the weld overlay process will be evaluated
'by a representative mock-up and metallurgical testing.
The mock-up study will evaluate:(1) feasibility of using mechanized welding on elbows. (2) evaluation of quality of weld metal, (3) UT inspectibility of the pad weld, (4) Extension of existing crack during pad weld repair, and (5) concern over welding on marginally sensitized material.
In the event that leakage occurs during the repair, back-up measures have been established to stop the leak, and allow repairs to proceed.
Mechanical clamps have been designed for each generic repair zone,and j
a freeze plug will be established and maintained during the first 1/8" thickness of weld overlay on the flaw repairs for weld joint S-9. S-10, and F-12.
The equipment will be maintained in place for the remainder of the weld repairs such that the freeze plug could be re-established.
During the course of the repair, the non-isolable piping is being supported with temporary restraints. The supports wil.1 account for normal, pipe / water weight, repair equipment-weight, and relocation of existing restraints.
7
- 8.
3.0 Evaluation of Proposed Repairs The following evaluation of the proposed repairs includes metallurgical aspects (materials, welding, and related contingency measures); stress and fatigue-analyses and systems safety considerations.
Included in the metallurgical evaluation are our assessments pertaining to the need I
for. additional surveillance (e.g., technical specification changes).
~
3.1 Metallurgical Considerations I
3.1.1 Repair Evaluation t
For.the pipe-to-pipe weld repair by sleeving, all welding will be l
done with water in the pipe acting as a heat sink, and will be
~
performed on uncracked base metal.
The weld buildup over cracked welds is more critical. Many l
undersurface cracks will " chase" the succeeding weld beads through..and terminate at or near to the finished weld surface.
This is particularly true when water is present. Melt through to a water filled crack could cause the weld to " blow out" locally, causing a leak and making continued welding impossible.
For this reason, positive measures should be in place to handle such contingencies.
a 4
4 e
+---w-
=w.-
r
,-.*w--
e
- er e
c c-
-vv -
%ew- - -
.es
,- - + - -
- -------e
_.9 However, if.the weld overlay is properly performed, the reinforced joint is not likely to fail by-intergranular stress corrosion cracking (IGSCC), and there is reasonable assurance that the repair will be safe for operation during the remainder of'the operating cycle.
The proposed repair for the sock-o-let is to reinforce this joint with a tailored sleeve. This involved simple fabrication and welding operations; therefore, no problem is anticipated. The
. joint should be mechanically supported such that even if a leak should develop during operation, the reinforced joint would not fail in a manner to cause a leak of significant size. Welding procedures and welders are to be qualified in accordance with ASME Section IX prior to start of work.
A problem that could arise by ' reinforcing the cracked welds is that the extra stiffness of the repairs and joints may induce higher stresses in the unreinforced welds than were there' previously. This change in stress and strain pattern could cause increased cracking propensity in the unrepaired welds in this piping. However, current leak detection capability is considered adequate to detect leakage from such cracking should it occur.
'3.l.2 Generic Implications The licensee states that the RWCU System welds have not been inspected to any significant extent prior to the recent events.
. A total of five welds have been inspected during the service lifetime. All their welds were found to be free of cracks, and are still free of cracks, according to results of the recent 100% inspection.
BWR piping systems that are used intermittently, or are stagnant during operation, often have the high level of stress and aggressive oxidizing water chemistry necessary to cause IGSCC. The BWR piping systems that have been most prone to this problem include recirculation bypass lines, core spray lines, and reactor water clean up lines. The NRC position is that " service sensitive" lines such as these should be inspected for cracks associated with the welds every refueling outage.
Inspection on a sampling basis is considered acceptable for systems with many welds. This position and related criteria and recommendations are contained in.
NUREG-0313, Revision 1.
3.1.3 Contingency Measures During Repairs i
l Although no major problems are expected, the fact that this piping cannot be isolated from the reactor vessel requires that contingencies be inplace in the event that repair operations cause a'significant leak. The following contingency plans and precautions shall be taken:
l
' - ll -
a.
Methods to temporarily stop leaks,'such as' clamps or formed-rubber sleeves, shall be prepared for each type of joint to
~
- be repaired. The accessibility and clearances available-
.in which to use these methods shall be verified before beginning the repairs. These will be used to temporarily stop a leak to permit effective freeze plugs to be generated if needed.
b.
Freeze plugging equipment will be installed ready'for use before repair, operations begin. The freeze plug shall be established prior to initiating repair operations on weld locations S9, S10 and F12, and shall be maintained until the first 1/8" lof weld metal is deposited.
c.
Personnel trained in establishing both the temporary
~
measures identified above and freeze plugs shall be at the work place. Communications shall be established between the welding operator supervisor and the control room during welding.
d.
Plans shall be made to obtain and install fuel storage capabity expeditiously, if needed.
e.
The finished repair will be nondestructively examined to ensure that the weld deposit is free of cracks, and hydro tested at cold operating pressure and at hot standby conditions.
J f.
- A temporary. support system will be installed to ensure that normal stressLlimits will not be exceeded for unreinforced (unrepaired) welds, i
3.1.4 Conclusions Regarding Material Aspects of Repairs We have concluded that the proposed repair may be accomplished a.
in a safe manner, if precautions listed above are taken, b.
We conclude that, provided the weld reinforcements are-accomplished as described, the repaired joints are safe 4
to operate until the next refueling outage.
Because there is a possibility that the unrepaired and c.
unreinforced weld joints may be subject to higher stresses as a result of these modifications, they may-have a somewhat higher propensity for cracking; however current leak detection capability is considered adequate during this interin period of operation.
d.
No additional Technical Specification changes are needed at this time to assure continued safe operation.
3.2 Stress Considerations 4
3.2.1-Stress and Fatigue Evalutions 4
, - -..~, -.
..-r
~
. Stress and fatigue analyses were performed to determine the acceptability of the repairs for service for the remainder of the operating cycle. The strength of the repairs was evaluated by finite difference analyses performed at the proposed pipe sleeves located at F6 and FIA, and cloced form solution analyses performed for weld overlays at S9, S10 and F12.
The mathematical model used for the F6 location extended from the center of the pipe sleeve to 7" beyond the end of the pipe sleeve. This model was used to evaluate both mechanical and thermal loadings. Mechanical loadings taken into account included gravity loads, OBE loads, seismic anchor movements and internal pressure loads. Thermal loadings were selected for the most severe thermal impact, which included the startup/ shutdown and hot isolation transients. The forces due to mechanical loadings were combined in such a manner as to maximize each of the six force and moment components. The resultant forces and moments were applied to the model such that the maximum shear stress in the pipe was at the same location of maximum normal stresses. Temperature profiles caused by transients were generated by a finite element model. Thermal gradients causing maximum thermal stresses from each transient were selected for stress analyses. Results from the two stress analyses were then combined with those from mechanical loading analysis and then compared with design limits to determine whether they were acceptable.
The existing piping section was accounted for in the most-conservative manner in the boundary conditions used for the analyses.
For the repair at FIA, the analytical procedure was similar to that used for F6.
One conservative assumption was adopted by assuring that pipes were stressed by mechanical loadings to the maximum stress value permitted by ANSI B31.1.
The effect of" internal pressure (1250 psig) was analyzed in two ways; reactor water leaking (and not leaking) through the existing sock-o-let.
The higher of the stress values from these two cases was selected to combine with the thermal stress analyses results.
Closed form solutions were used to determine the adequacy of weld overlays. The meclunical loadings adopted were those used in the original design analysis for the same cross-section, and resisted by the new material. Thermal. stresses were obtained by assuming the temperature gradients existed across a thick cylinder wall.
The limits used for all stress analyses were ASME Boiler and.
Pressure Vessel Code Section III, class 2 design limits.
'In addition to the stress analyses, a fatigue analysis was also performed for each of the five repair locations. The transients used were 10 start up and shutdown cycles and 2 hot isolations.
Usage factors were determined by following the ASME Boiler and Pressure Vessel Code.Section III design fatigue curve. The appropriate fatigue strength reduction factors were selected and used in the analyses.
________j
. 3.2.2 Conclusion Regarding the Stress and Fatigue Analyses We conclude that the results of stress analyses and fatigue analyses demonstrate that from a mechanical strength and fatigue standpoint, the repairs proposed will withstand the anticipated loadings for the remainder of the current operating cycle. The criteria and limits used for these analyses are conservative and therefore, acceptable.
3.3 Systems Considerations During Repairs 3.3.1 Description of Present Plan Consideration According to the licensee, the present conditions of the reactor are as follows:
Reactor Water Level Normal Shutdown Level, i.e, about 19 ft. from the top of the active fuel Temperature / Pressure 160*F/ atmospheric Heatup Rate 10*F/Hr.
3.3.2 Evaluation of Systems During Repairs The shutdown pumps are operated intermittently at 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> intervals.
16 -
This indicates low decay heat in the reactor. The boiloff rate is not significant. Since the reactor is in a stable condition, i.e., in cold shutdown, the consequences of a pipe break are manageable.
If a circumferential break is postulated in the RWCU System piping and if no action is taken by the operator, the worst scenario is the exposure of 1/3 of active fuel in the core. This is based on the' location of reactor recirculation suction nozzle and the jet pumps. Since the reactor coolant temperature is only 160*F and there is no pressure, " flashing" would not occur.
Since the repair is performed with a water level of about 19.ft.
from the top of core, there is reasonable time for the operator to start feedwater pumps or ECCS pumps to maintain the reactor level. The approximate calculations of the-system heatup rate, based on estimates of system heat capacity and the fission-product decay heat at the present time, are in reasonable i
agreement with values measured at the plant. The estimated decay heat values are so low that several hours would be available for corrective operator action even if a large break occurred l
in the nonisolable portion of the RWCU piping, or automatic i
i initation of ECCS would not occur.or vessel water level decreased i
j to the jet pump nozzles.
i l
I L
. The licensee has prepared special administrative procedures for the repair. The licensee is also committed to a detailed training session with a mock-up prior to starting the repair on the nonisolable.section of the piping. The licensee has developed contingency plans to mitigate the consequences of a leak should it occur on the 6" line. The licensee also has developed emergency procedures for a failure that causes significant.
leakage and that cannot be stopped. The procedures require manual starting of feedwater and ECCS pumps if the reactor level drops.
Additionally, plans to obtain core offload capability and initiate core offload are in place.
3.3.3 Conditions to be Observed During Repairs Prior to initiating the repair program and procedures, the following conditions, which are-not exclusive, must be satisfied.
a.
Reactor water level shall be closely monitored at all times during this repair program for all Control Room Reactor Water Level instrumentation.
b.
Reactor water temperatures shall be properly controlled using the intermittent operation of the Shutdown Cooling Mode of RHRS.
_.mm.
i
. c.-
A Condensate-Feedwater System flow patheshall be available
.with pumping capability.
d.
The LPCI Mode of RHRS shall be' operable at all times: including -
-RHR pumps and valves constituting the injection paths being.
in their normal configurations. A pump operability test.will be performed before the repair to show that the pumps are operable.
e.
Both Core Spray subsystems shall.be operable at all.-times, including the Core Spray pumps and valves constituting the injection paths being in their normal configuration. A pump' operability test.will be performed before.therrepair' 4
to show that the pumps-are operable.
f.
Both the 1/2 and 2 Diesel Generators shall be operable.
i p.
Suppression Pool water level shall be within Technical
~
Specification limits.
j h.
Condensate Storage Tanks ~shall be maintained at sufficient levels to provide back-up water to Suppression Pool for Core Spray and LPCI.
1 9
, + - -
7
,p
-,,-E r
~-3--
e,-*
--r--
-w-
_ 19 _
i.
Secondary Containment'shall be in effect with both SBGT Systems operable and the. capability for Reactor Building-Ventilation isolation in effect.
j.
Communications will be established between the Control Room and the Drywell.
k.
Administrative controls are in place for RWCU inboard i
isolation valve M02-1201-2 to prevent inadvertent opening.
a 1.
To ensure that any postulated additional cracking will be detected prior to potential pipe rupture, leakage within the drywell shall'be closely monitored. A change in unidentified leakage of greater than 1 gpm in four hours shall be cause for the licensee to inspect the RWCU system repairs for leaks.
j 3.3.4 Conclusions Regarding Systems During Repairs i
In summary, there is reasonable assurance that-repairs can be performed safely with the core present in the vessel.
4.0 Conclusion
.It-is concluded:
(1) that there is reasonable assurance that the repairs can be safety performed if all contingencies and conditions discussed are met, and acceptable nondestructive examinations and leak tests are conducted subsequent to repairs, and (2) the repaired joints are safe to operate until the next refueling outage.
i M0
[
]
RWCU LINE
/
2-1201-2 l
2-1202-6"A p
r
/
/
/
/
\\
CORE f
7.
/
o DRYWELL
- RHR 5HUTC0WN COOLING LINE 2-1025-20"A l
[
]
i l
RPV II RECIRC SUCTION l
LINE 2-02028-28"A y
1 RECIRC PUMP 2B FIGURE 1.1 CONCEPTUAL DRAWING OF RWCU.LINE AND RELATED LINES
- - - - + ~
X
>-e e Existing Pipe Support a Esisting Cuide
.y
- weld s9 /
C)*p '
9 9
IDC7 TION F12 2 longitudinal cracks on ellw side at 10:00 position,' facing valve),
one -954 thru, one -50%
thru walls one longitudinal crack at 9:30 position,
-50% thru wall s
-q
!* 4 IDCATION FIA IO'*2 I/O
=l; 1
Possible small leak J.
in fillet weld r
r between 2" pipe and g'f
- 1, sockolet ir IDCATION F6 2*-5=
Ciretsuferential J
IDCATION S10 crack, 1.5" long, Longitudinal crack
-954 thru wall in 3/4" long, -954 thru
- t downstream IIAZ wall on elbow side y
IDCATION S9
}
g Longitudinal crack 3/4" long, <50s thru wall, in HAZ on elbow FIGURE 1.2 side of weld ISOMETRIC DRAWING OF RWCU LINE NO. 2-1202-6" - A 9
li S
DA P
g s/
D L
~
E g_
=.,
W u/:I x
ys.ii! e t 1
A 1
s w
D s
s A
s w
s E
E x
'o 1
V sO 2
E
/
'l
/
E
+
w E
L s
R S
A1 U
s i
s'_
G E
I E
P r
A P
F I
I i'
i! I a
P P_
G
- L' E
x N
e V
I E
T E
S
/
L x
l Ii S
E uI:I D
b I
A P
4
\\7 D
/
PY L
E T
W
/
/
/
A l
/
lA T
/
E D
///
/
ts i
l,1.l
b I
., r.
O t
\\
D s
sA. w. 7 s
8 30
\\
SOCKOLET SLEEVE I
t
,s I
\\
m_m [m
'/+ 7 s
umv
's
[
-t
+
f DETAIL D
'm
_s j m-s _ m I
FIGURE 2.2 SOCKOLET SLEEVE
O r
YAL l
R i
F
,^
E N
V m
O
(
i '[
EV L
A V
-WO B
L E
w LA ii t
C v
I S
PY Y
T 3
A 2
LR E
E R
V U
O G
D IF L
o E
t W
s Y
~~
A L
R
'q E
~
9 N~
V 5
O j
?
W
,7 ilk __
_.=
O B
L E
_~
L A
('AU C
IPYT T
Y 3
,.