ML20154N598
| ML20154N598 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood, 05000000 |
| Issue date: | 09/20/1988 |
| From: | Morgan W COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| Shared Package | |
| ML20154N603 | List: |
| References | |
| 5140K, IEB-88-008, IEB-88-8, NUDOCS 8809290353 | |
| Download: ML20154N598 (11) | |
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Commonwealth Edloon
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One Fest Nabonal Plaza. CNea0o. Innois Address Reply t* Poet Omco Box 767 CNeago, Infos 60690 0767 September 20, 1988 U.S. Nuclear Regulatory Comission Altne Document Control Desk j
Washington, D.C.
20555
Subject:
Byron Station Units 1 and 2 Response to NRC Bulletin 88-08 &
88-08: Supplement 1 & 2 Dorket Nos.
50-454/455 References (a) NRC Bulletin No. 88-08 Dated June 22, 1988 (b) NRC Bulletin No. 88-08, Supplement 1 Dated June 24, 1988 (c) NRC Bulletin No. 88-08, Supplement 2 Dated August 4, 1988
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Dear Sir The above referenced bulletin and supplements requested that i
licensees review the reactor coolant system (RCS) to identify any connected, i
unisolable piping that could be subjected to temperature distributions which would result in unacceptable thermal stresses and (2) take action, where such piping is identified, to ensure that the piping will not be subjected to unacceptable thermal stresses.
Commonwealth Edison has completed its review pursuant to the request outlined in Bulletin 88-08 and its supplements for Byron Station Units 1 & 2.
1 Additionally Edison has discussed with the staff the methodologies used by Westinghouse Electric Corporation to support Byron Station's proposed a
response. These discussions were held per teleconferences September 2 and 9, 1988.
Per the staff's request Comonwe Ith Edison is submitting Byron Station's proposed response to Bulletin 88-08 along with the Westinghouse Electric Corporation engineering evaluation used to pr5 pare this response.
This information is attached in enclosures 1 & 2.
To the best of my knosledge and belief, the statements cedeained above are true and correct.
In some respect these statements are not based on my personal knowledge, but obtained information furnished by other Comonwealth Edison employees, contractor employees, and consultants. Such information has been reviewed in accordance with company practice, and I believe it to be reliable.
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Please address any questions that you or your staff may have concerninq this response to this office.
Respectfully, WxE W. f. Ho an Nuclear Licensing Administrator Attachments:
cc A.B. Davis Resident Inspector / Byron rf 5140F Subscribed and gwo,rp to be foIe me, thi W_IE day
?ivJE 1988 of
.'i ts & N Hotary Publi
AIIACIDiZliT_1 Byron Station Response To NRC Bulletin 88-08:
"Thermal Stresses in Piping Connected to Reactor Coolant Systen" URC_ Requested _ Action _il Review systems connected to the RCS to determine whether unisolable sections of piping connected to the RCS can be subjected to stresses from temperature str7tification or temperature oscillations that could be induced by leaking valve: and that were not evaluated in the design analysis of the piping.
For those addressees who determine that there are no unisolable sections of piping that can be subjected to such stresses, no additional actions ate requested except for the report required below.
Byron Station'_A_Responsn A review of Byron piping systems connected to the RCS was conducted l
by Westinghouse to determine whether unisolable sections of connected piping could be subjected to stresses from temperature stratification or oscillations induced by leaning valves.
Susceptible sections of piping we.e identified in the $1ngle Aux'11ary Spray line and the four Charging Pump to Cold Leg Injection linei of each Byron Unit. Temperature oscillations may be induced in these lines by leakage of the isolation valves located between the RCS and the charging system.
HRC_ Requested _ Action _l2 For any unisolable sections of piping connected to the RCS that may have been subjected to excessive thermal stresses, examine nondestructively the welds, heat-affected zones and high stress locations, including geometric discontinultles, in that piplng to provide assurance that there are no existing flaws.
Byr_on_ Station's.Responst i
Current nondestructive examination technology does not permit reliable volumetric (radiographic /ultrasonte) examination of small diameter (less than 4-inch) stainless steel piping with high contact radiation I
readings. Therefore, the 1.5-inch diameter Charging Pump to Cold Leg Injection lines cannot be nondestructively examined at Dyron Station.
However, reasonable assutance can be provided that the thermal stress phenomenon potentially caused by the leakage of the Charging Pump to Cold Leg Injection Isolation Valves (1/2SISB01A,B) has not occurred at Byron Station.
Performance of "Reactor Coolant System Isolation Valve Leakage Surveillance Procedure" (1/2 BVS 4.6.2.2-1) determines back leakage from the Reactor Coolant System through the Charging / Safety Injection Check Valves (1/2518815, 1/2SI8900A,B,C,D) to a test tap lo ated between 1/2SI8815 and 1/2 SIB 801A,B valves.
By virtue of the test tap's location, the detection of forward leakage f rom the Charging Pump through the 1/2 SIB 801A,B lsolation valves is an unintended result of 1/2 BVS 4.6.2.2-1 performance.
Procedures 1/2 BVS 4.6.2.2-1 are routinely, rformed as follows for the 1/2SI8815 check valves
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a)
At least once per 18 monchs; b)
Prior to entering Mode 2 (Startup) whenever the plant has been
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i in Cold Shutdown for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or more and if leakage testing has not been performed in the previous 9 months; c)
Prior to returning valves 1/2SI8815 to service following 1
maintenance, repair or replacement work on the valves; d)
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following 1/2SI8815 valve actuation due to automatic or normal action or flow through the valve.
j Data from past performances of 1/2 BVS 4.6.2.2-1 were reviewed and indicate sero leakage in eleven performances and one instance in which leak rate was recorded as 0.000317 gallons per minute (negligible and well within i
the acceptance criteria of 1.0 gpm).
Since all twelve leak tests of the 1/2S18815 valves conducted from January 1985 to April 1988 Indicate that sero 1
or negligible leakage existed, 1/2SI8801A,B valves have not leased; therefore, l
the Charging Pump to Cnid Leg Injection lines have not been subjected to excessive thermal stresses.
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j Regarding the Auxiliary Spray line, Westinghouse recommended j
inspections of a 2-inch sockolet weld at the RCS piping connection and a i
portion of the 6-inch main spray piping.
Due to nondestructive examination j
technique limitations, the 2-inch sockolet weld cannot be volumetrically
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examined, however, it will receive e surface examination and the susceptible t
portion of the 6-inch main spray line will be volumetrically examined prior to l
the end of the Unit 1 second refueling outage (scheduled to begin September 2, l
1988). The necessity to nondestructively examine the Unit 2 main spray line and 2-inch sockolet weld will be determined by a pending stress analysis of l
the pipe.
In the event that nondertructive: examinations of the Unit 2 main spray line and 2-inch sockolet weld are requitad, the examinations will be
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completed prior to the end of the Unit 2 first refueling outage (scheduled to
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begin January 1989).
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tfRC_RequesteLActLon_U I
l Plan and implement a program to provide continuing assurance that
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l unisolable sections of all piping connected to the RCS will not be subjected
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to combined cyclic and static thermal and other stresses that could cause t
i fatigue failure during the remaining life of the unit. This assurance may be l
provided by (1) redesigning and modifying these sections of piping to l
withstand combined stresses caused by various loads including temporal and
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J spatial distributions of temperature resultirg from leakage across valve l
I seats, (2) instrumenting this piping to detect adverse temperature
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distributions and establishing appropriate limits on temperature r
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distributions, or (3) providing means for ensuring that prest.ure upstream from
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block valves which might leak is monitored and does not exce'td RCS pressure.
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Byron _ Station _ Response t I
In order to asrure that the four Charging Pump to Cold Leg Injection l
lines on each Byron Unit will not be subjected to cyclic thermal stresses that i
could cause fatigue failure during the remaining lives of Byron Units 1 and 2, a
surveillance procedures will be developed or revised as netessary to require l
pe iodic tests specifically for leakage past the 1/2 SIB 801A,B isolation valves.
If the 1/2 SIB 801A,B valves leak, the leakage will be discovered L
d"ring the conduct of We surveillance, and action ww.1d ensue to determine i
the leakage source and correct the cause.
Leak testing for the 1/2SI8801A,B l
valves will be routinely performed as fol'ows:
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i a)
At least once per 18 months; b)
Prior to entering Mode 2 whenover the plant has been in Cold Shutdown for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or more and if leakage testing has not been performed in the previous 9 months; c)
Prior to returning valves 1/2SI8801A,B to service following maintenance, repair or replacement work on the valves d)
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following 1/2SI8801A,B valve actuation due to automatic or manual action or flow through the valve.
The periodic performance of the leak test minimizes the potential for long term thermal cycling of the four unisolable Charging Pump to Cold Leg Injection lines by detecting and resolving isolation valve leakage. The 1/2SI8801A,B leak test procedures will be approved for use prior to the end of the Unit 1 second refueling outage (scheduled to begin September 2, 1988).
In order to assure that the single Auxiliary Spray line on Byron Unit I will not be subjected to cyclic thermal stresses that could cause fatigue failure, appropriate sections of the piping will be instrumented with external temperature monitoring devices. The outputs of these devices will be evaluated to determine if leakage past the single isolation valve from the charging system is occurring. The temperature monitoring devices will be j
installed on Unit 1 prior to the end of the second refueling outage (scheduled to begin September 2, 1988).
If a pending stress analysis of the line concludes that the Auxiliary Spray line is not susceptible to fatigue failure, then the temperature monitoring devices may be removed. The necessity to instrument the Unit 2 Aux 111ary Spray line will be determined by a pending stress analysis of the pipe.
In the event that temperature monitoring of Unit 2 is required, the instrumentation will be installed prior to the end of the Unit 2 first refueling outage (scheduled to begin January 1989).
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All Class 1 piping receives a VT-2 Visual Leak Test before and after refueling outages in accordance with the ASME Code. The examinations are intended to detect leakage from Class 1 piping. The Auxiliary Spray and Charging Pump to Cold Leg Injection lines are examined as part of the test, and any leakage due to thermal stress induced pipe cracking would be noted and resolved.
Additionally, the high stress welds identified in these lines are routinely dye penetrant inspected as part of the ASME Inservice Inspection Program.
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5140K L
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ATIAQ91EHL2 Westinghouse Electric Corporation Engineering Evaluation for Commonwealth Edison Byron and Braidwood Nuclear Stations To Address NRC Bulletin 88-08 l
5140K l
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y Westinghouse Energy Systems Qm,,, g, Electric Corporation CAE 88 308 CCE-88-427 FSSE/SS CAE-5530 Mr. D. Elias, Engineering Superintendent S.O. CAE 280 Byron and Braidwood Stations Commonwealth Edison Company August 15, 1988 P. O. Box 767 Chicago, IL 60690 comonwealth Edison Company Byron and Braidwood Nuclear Stations Revised Information Regarding Potential for Temoerature Oscillations in the Reactor Coolant Pioina
Dear Mr. Elias:
The reference letter CAE 88 301 provided Westinghouse's initial input to Comonwealth Edison to address NRC Bulletin 88 08. This letter supersedes CAE 88 301 and provides more information about the screening criteria used to select potential locations of cyclic fatigue.
The potential locations are listed on the attached table.
In the table, only those lines connected to reactor coolant system where temperature induced cyclic fatigue could occur, are included for both units 1 J
and 2 of Byron and Braidwood Stations. The screening criteria for determining these lines are as follows:
1 Adequate driving force must be available. Only lines connecting the charging pumps to the RCS are potential candidates for continuous or cyclic leakage.
Only lines with single, normally closed isolation valves are included.
10CFR50.55a defines the reactor coolant pressure boundary as extending out to the second isolation boundary. Conservatively this would extend the piping line segments reportable under NRCB 88 08 Action 1 back to the second isolation valve; however, tne location of the highest fatigue is the part of the system interconnecting piping between the last check valve and the reactor coolant main loop. The affected table contains only the portion of the piping between the last check valve and the main coolant loop.
Piping peak-to-peak temperature cycles of 50'F were observed in the region of the cracked pipe at the J. M. Farley slant with the overall top-to bottom stratificattor. temperature being a)out 250'F.
Westinghouse's recommendation to Portland Gas and Electric was that fatigue could occur at cycles as low as 20'F.
All temperatures refer to piping outer diameter measurements.
CAE-88 308 CCE 88 427 Mr. D. Elias August 15, 1988 Piping layouts have been reviewed to locate cold traps. When piping runs vertically downward from heat sources (e.g., reactor coolant pipes or regenerative heat exchanger outlet piping), the water (and piping) at the bottom of the cold trap is cooler than the water and piping at the heat sources. This leads to the potential for temperature oscillations.
The charging lines and auxiliary spray line are potentially subject to temperature iduced cyclic fatigue due to the piping layouts that have cold traps. Warm water (470*F) from the regenerativo heat exchanger can propagate to the cold trapped piping where it w'11 be further cooled.
Further propagations of the now cooled leakage to the pressurizer spray or cold leg (s)
(530*F or 556*F, respectively) can cause cold to hot temperature cycles.
One of the charging isolation valves (CV8146 or CV8147) is closed at all times. This reduces the number of design transients that affect the charging nozzle because the operating duty is shared by each charging nozzle. At any given time, the charging line with the closed isolation valve could potentially have cyclic leakage and )iping temperature oscillations. Valves CV8383A and B are spring loaded chec( valves with a cracking pressure of 200 psid intended to be thermal ex)ansion relief valves for the regenerative heat exchanger tube side piping.
Tae high cracking pressure prevents these valves from opening during normal operation, and the flow paths through valves CV8383A and B will not eliminate the potential for cyclic fatigue.
1 The charging /SI branch line potential leakage will be at the temperature of the charging pump discharge.
Leakage could cause cold to hot temperature cycles to occur in the branch line piping adjacent to the reactor coolant cold i
legs.
The attached tables also define whether the potential leakage source temperature applies to normal plant operation or whether it is a transient condition.
If it is a transient condition, a duration for the transient was defined. This information is provided to assist in the evaluation of the severity of any thermal stresses or to makt the judgment that short-term i
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transients will not cause high cyclic fatigue.
A separate document will be provided that documents the components and locations for non destructive examination.
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CAE-88 308 CCE-88-427 Mr. D. Elias August 15, 1988 If we can be of any additional assistance, please call me.
Yours very truly,
$' A 70 7 4T
- 3 (e r H. C. Walls, Manager Commonwealth Edison Projects U.S. Nuclear Projects l
Attachment cc:
C. A. Moerke R. A. Gestor R. Plentewicz R. E. Quario P. Thc as, W R. W. Buchholz, W I
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ATTACHMENT Line Number Subject Normally Closed Leakage Source to Temperature Isolation Valves Temperature Notes Stratification (a) normal operation (b)designbasis (c) design transient 0
RC37A3 CV8146 470 F(a) 1 0
400 F(b) 1 0
375 F(c) 2 0
35 F(c) 3 0
RC28A3 CV8147 470 F(a) 1 0
400 F(b) 1 0
375F(c) 2 35 F(c) 3 i
0 0
RC30M1-1/2 SI8801A or B 100-130 F(a) 1 i
0 (valves in parallel) 60 F(b) 1 0
35 F(c) 2 i
l RC30AB1-1/2 S!8801A or B 100 130 F(a) 1 0
0 (valves in parallel) 60 F(b) 1 0
35 F(c) 2 0
RC30Atl-1/2 SIS 801A or B 100 130 F(a) 1 0
(valvesarein 60 F(b)
I 0
parallel) 35 F(c) 2 i
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ATTACHMENT Line Number Subject Normally Closed Leakage Source to Temperature Isolation Valves Temperature Notes Stratification (a) nomal operation (b) design basis (c) design transient 0
RC30AD1-1/2 S18801A or B 100-130 F(a) 1 0
(valves are in 60F(b) 1 0
parallel) 35 F(c) 2 0
RYl8A2 CV8145 470 F(a) 1 0
400 F(b) 1 0
375 F(c) 2 0
35 F(c) 3 Notes 1)
Potentially, leakage source would occur continuously.
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Leakage source temperature would occur very infrequently--likely less than one hour per year per reactor.
3)
Leakage source temperature based on low pressurizer level signal without coincident safety injection or on phase A containment isolation without coincident safety injection. These scenarios are unlikely to continue for more than 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> per year (for one instance per reactor) until either operator intervention / rectification occurs or else the scenario degrades to a safety injection which isolates normal charging.
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