ML20073C939
| ML20073C939 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 04/19/1991 |
| From: | Sieber J DUQUESNE LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9104260055 | |
| Download: ML20073C939 (15) | |
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$Norefquvt. PA 1!477 (XX4 April 19, 1991 IN1s$$
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Nuclear Regulatory Commission Attn:
Document Control Desk Washington, DC 20555
Subject:
Beaver Valley Power Station, Unit No. 1 and No. 2 BV-1 Docket No. 50-334, License No. DPR-66 BV-2 Docket No. 50-412, License No. NPF-73 Potentially Invalid Leak Detection Tests Used as an Alternative for Required ASME,Section XI Hydrostatic Tests (Supplement Report)
In our submittal of March 15, 1991, we provided an operability assessment for Beaver
- Valley, Unit No. 1 to supplement our May 3,
- 1990, submittal concerning potentially invalid leak detection testing.
We also stated that we were evaluating all piping that was excluded from the Instrumented Inspection Technique (IIT) Program to determine if any additional piping should have been included.
We have completed this review and have determined that the following Unit No. 1 piping should have been tested:
Coils}ing Water Subsystems (Containment Air Recirculation Component Coo Containment Air Compressor Aftercooler, Control Cooling Room Redundant Cooling Coils *,
"C" Reactor Coolant Pump Cooling Lines)
Fuel Pool Cooling System
- Portions of the Chemical and Volume Control System (CVCS) (VCT, Non-Regenerative and Excess Letdown Heat Exchangers, Boric Acid Transfer Pump Lines, and Associated Piping)
Portions of the CVCS (Regenerative Heat Exchanger Piping)
Portions of the Auxiliary Feedwater System
+
Steam Generatcr Blowdown Sample Lines
+
Portions of the River Water System Three of these systems were exempted from the pressure testing program during the first period of the second ten year interval at Unit No. 1 because the IIT requirements were still in place and being followed.
These are identified above with an asterisk (*).
9104260055 910419 P,DR.
ADOCK0500g4 g g477 I/ /
Beaver Valley Power Station, Unit No. 1 and No. 2 BV-1 Qocket No. 50-334, License No. DPR-66 BV-2 Docket No. 50-412, License No. NPF-73 Page 2 Attached are operability assessments for che above piping which conclude that the piping is structurally sound and functionally operable.
These operability assessments have been reviewed by the Onsite Safety Committee.
Although the justification for excluding these lines from the first ten year interval pressure testing program (i.e., hydrostatic and leakage testing) is considered valid and was handled in accordance with requirements established during the IIT licensing
- process, we will include these lines in the ASME Section XI second ten year interval pressure resting program or otherwise file for relief in accordance with 10 CFR s0 as warranted.
If you have any questions, please contact Mr. Steve Sovick at (412) 303-5211.
Sincerely, ohd
/r av J.'D.
Sieber Vice President Nuclear Group
Attachment:
Supplemental Report cc:
Mr. J.
Beall, Sr. Resident Inspector Mr.
T.
T. Martin, NRC Region I Administrator Mr. A.
W. DeAgazio, Project Manager Mr.
R. A. McBrearty, NRC Region 1 Inspector M
i
)
SUPPLDIENI'AL REIORI' System Operability Awsent for the EW-1 Comnanent Coolirrt Subsystems
'Ihe Control Room Redurdant Cooling Colls, Containment Air Recirculating Cooling Ccils (wlth exception to the containment penetrations), ard the Containment Air Compresscr After Cooler lines are considered to have maintained structural integrity and functional operability based on the followirg activities and surveillances:
'Ibe oInrating pressures of the piping subsystems were compared with the pressure ratirgs for the pipe sizes ard schedules. 'Ihe smallest mrgin between the values was approximately 1024 psig. Ample margin exists in the design of this line to accomtiodate operating pressures.
Operations Surveillance Test (OST 1.44C.2) Containment Air Recirculation Cooler 'Ibst is performd nonthly.
Imkage through the subject lines would be detected by this OST, Component coolirg subsystems are in continuous operation durire the life of the station. Routine pre-startup inspections are perfonned along with periodic observation ard monitorirs of system parameters during operation.
Camponents and systems served by the component cooling water subsystem are monitored; therefore, any malfunctions causing low flow, low pressure, high temperature, or high radioactivity levels could be detectcd. 'Ihe affected components ard systems causing the increased levels will be isolated, shutdown and repaired.
Any leakage in the Containment Air Recirt:ulating cooling coils or the Containment Air Compressor After Cooler lines would increase the containment sump levels ard containment sump pump cyclire rate.
Icakage examinations at operating pressure were performed on all of the pipiny in question as required by ASME Section XI durirg the first two periods of the first 10 year inspection interval.
Leakage examination at operating pressure was performed on the Control Room Redundant Cooling Coils as required by ASME Section XI during the secord period of the second 10 year inspection interval.
I A Working Group of the ASME XI Ctmmittee has recwucrded that a leak test (normal operatiry pressure) be required in lieu of currunt hydro +W.
'Ihis is based on experience showing minimal elevated pressure tests do not generate adequate stress loads to prxpagate existing defects. 'Ihis ra > - erdation was passed by the Ctde Sub-comittee ard is presently urder review by the Main Code Ctamtittee.
Based on the above, the Control Roam Redundant Coolirg Coils, Ccntainment Air Recirculatiry CooliIg coils, ard the Containment Air Compressor After Cooler lines have been determined to be structurally sound ard functionally oprable.
l System Operability Im m e nt for the IN-1 Canpanent Coolirg Subsystems (Pinirn between TV-CC-105C & 1CG-296. 299, 302)
The portion of the Unit 1 Reactor Plant Cmponent Coolirg System not tested (i.e., pipiry frcxn valve 7V-CC-105C and 1CCR-296, 299 ard 302) is considered to have unintained structural integrity aM its functional operability based on the followirq activities aM surveillances:
The operating pressures of the pipiry subsystems wem emparul with the pressure ratirgs for the pipe sizes ard schcdules. The snallest margin between the values was a; proximately 984 psig.
Anple nargin exists in the design of this line to ahuadate operating pressures.
Ctznponent coolire water to the Reactor coolant Pumps is in continuous operation durirg the life of the station. Routine pre-startup inspections are perfomtd along with periodic observation and monitoring of system parameters during operation.
The "C" Reactor Coolant Punp served by the cxmponent cooling water subsystem is nonitored, therefore, any malfunctions causing low flow, low pressure, high tenperature, or high radioactivity levels, could be detected. The affected ccraponents ard systems causirq the irr~ eased levels will be isolated, shutdown and repaired.
Any leakage in the "C" Reactor Coolant Punp CCR pipirq would increase the containment sunp levels and containment sump pump cycliry rate.
Irakage examinations at cperatirg pressure were perfomed on all of the piping in question as required by ASME Section XI durirg each period of the first 10 year inspection interval. These exams were performed by certified VI-2 pelsonnel.
A Working Gmup of the ASME XI Ctanittee has rey arded that a leak test (normal operatirq pressure) be required in lieu of current hydro-testirg. This is based on experience showing minimal elevated pressure tests do not generate adequate stress loads to propagate existirg defects. This rex eMation was passed by the code Sub-Committee ard is presently uMer review by the Main Code Committee.
Based on the considerations listed above, the portion of the "C" Reactor Coolant Pump CooliIg lines identified have been determined to be' structurally sound ard functionally operable.
i System Operability Assessmnt of the IN-1 Riel Pool Coolim System On 4/11/86, a system hydrostatic test (IT-20-221-1) was performed to fulfill the testim requiremnts of DCP 221 (the addition of valve 1PC-145 to the fuel pool ocolim system). S e piping that was tested was bouMed by valves 1K-145,1PC-110,1PC-111,1M-115 aM 1K-116.. A VP-2 inspection was perfonmd on this pipig by a certified inspector and documented on visual examination reports VER
- V13092 aM VER # V13093. Se piping bouMed by valves 1PC-110, 1PC-111 and 1PC-105 (i.e., the fuel pool suction lines) were not covered by hydrostatic test (IT-20-221-1). m is portion of piping is considered to have mintained its stnictural integrity and functional operability based on the following activities and surveillances:
Se operating pressures of the piping subsystems was ccxupared with the pressure ratings for the pipe sizes aM schedules. Se smallest margin between the values was approximately 587 psig. Ample margin exists in the design of this line to am sdate operating pressures.
his portion of the subject line is pressurized to nonral system operatirg pressure during nomal system arrangemnt.
Normal plant tours by operations persaucl are performed on a shift basis in the plant areas where this line is accessible. Any major pressure boundary leakage could, in conjunction with installed instrumentation, be detected during thee tours.-
A leak in the subject line would result in a decreasing level in the spent fuel pool which would be detected in the control room with a loa level annunciator. An annunciator for high fuel pool temperature also is in the control room.
Station 1 cgs monitor fuel pool level aM temperature on a shift basis.
R e Fuel Pool Cooling System is a radioactive system.
Serefore, any major leakage from this piping would result in an increase in airborne radioactivity which would be detected by various plant radiation monitors.
i
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Any leakage frca this pipirq would erd up in the fuel buildirg sump which would result in ircreased sump levels or stmp pump cyclirg rate.
A weekly Operations Sutveillanoc 'Ibst (OST 1.20.1) is performed to verify the " spent fuel pool icvel 1cu" annurriator is not illuntinated.
I.cakage examinations a'; operatirg pressure were perfonted on these lines in the first two periods of the first inspection interval at Unit 1 per the requiruments of ASKE Section XI.
These exam were perfonncd by VI-2 certified personnel.
A Workirg Group of the ASME XI cuanittee has t-nded that a leak test (nonnal operatirq pressure) be required in lieu of current hydro-testing, his is based on experience shcwing mininal elevated pressure tests do not generate adequate stress loads to propagate existing defects. his r-ndation was passod by the Code Sub-camntittee ard is prescr'tly under review by the Main 0:rb ccmittee.
Based on the above, the subject line has been determined to bo structurally sourd and functionally operable.
System Operability Assess mnt of the IN-1 Cicmical and Volume Control System (CVC3)
(VCT, Non-Regenerative ard D: cess Intdown Heat D:changens, Doric Acid Transfer Pumps, and Associated Pinirn)
W e portion of the CVCS not tested is considered to have mintained structural integrity and is functionally operable based on the following activities ard surveillances:
Se operatirn pressures of the piping subsystems was compared with the pressure ratirgs for the pipe sizes ard schedules. We smilest mrgin between the values was approximtely 318 psig. Ample mrgin exists in the design of this line to arr.umalate operatire pressures.
2e Volum Control Tank (VCT) and Non-Regenerative Heat
+
Dcchanger pipinct are prescurized to normal system operatire pressure during normal system arrargement.
Re control room operators perform Operatirg Surveillance Test (OST) 1.6.2 " Reactor Coolant System Water Inventory Balanoe" every three (3) days when the plant is operating at steady state corditions.
Isakage through the subject lines would be detected by this OST.
me inventory in the liquid waste system is logged daily (log L3-11). Since leakage frun these lines would be collected by the liquid waste system, a through-wall leak would be apparent in this inventory. We inventory is reviewed daily by the Shift Supervisor and weekly by the Site Radwaste Coordinator. Radiation monitors within the liquid waste system would also detect any leakage fram these lines.
Monthly, the Radiological Control Department personnel perfonn routine radiation surveys outside containment.
Seepage from the VCT, Non-Regenerative ard Doric Acid Transfer Pump pipiry that may not be detected by OST 1.6.2 or the liquid waste inventory would be detected durity this survey.
he CVCS is a radioactive system; an:1, therefore, any major leakage from this pipirg would result in an increase of airborne radioactivity which would be detected by various plant radiation mnitors.
1
7.
The VCr ard Non-Regenerative piping aru readily isolable should a leak occur. The piping has double-valve isolation frm the primary system on a pressurizer lw level signal. It could be easily isolated by the control recra operators should the line be discovered to be leakirg.
The Dccess Ictdwn Heat Excharger pipirq is nonnlly double valve isolated frm the ruactor otolant system.
Leakage exams at operating pressure waru perfonned on these linos periodically as required by ASME XI durirg the first two periods of the first 10 ycar Interval. 7tw.se exams were parformod by certified Vr-2 perravw.l.
A Workirg Group of the ASME XI Committee has recmmended that a leak tast (normal operating pressure) be required in lieu of current hydro-testing. This is based on expericnoe shwirq minimal e'evated pressure tests do not generato adequate stress loads to prepagate existing defects. This r-rxiation was passed by the we Sub-Camittee and is presently urder ruview by the Main Code Co d ttee.
Based on the above, the subject lines have been determined to be structurally sound and furctionally operable.
I 1
I System operability Ae w nt for the BV-1 Chemical and Voluno Cbntrol System (CVG)
Portions of the Rectenerative Heat Ihrhainer Ploim)
'Ihe portion of the CVCS (lines 2" 01-1-1502-Q1, 2" Q1-143-1502-Q1, 2" Ql-149-1502-Q1, and 2" OI-144-1502-Q1 between valve ILV-Ol-460B ard valves
'IV-Ol-200A, -200B, ard -200C) not testod is considered to have unintained structural integrity and is functionally operable based on the followirq activities and surveillances:
'Ihe operating prussures of the piping subsystems was compared with the p.m ratirgs for the pipe sizes and schedules. 'Ihe smallest margin between the values was approxinately 781 psig.
Ample margin exists in the design of this line to accumedate operating pressures.
'Ibe portion of the noted lines is pressurized to nomal system a
operatirg pressure durirg normal system arrargement.
'Ihe control room operators perform Operating Surveillance Test (OST) 1.6.2 " Reactor Coolant System Water Inventory Balance" overy three (3) days when the plant is operatire at steady state conditions. Irakage through the subject lines wxtid be detected by this OST.
'Ihe inventory in the liquid waste system is logged daily (log L3-11). Since leakage frm these lines would be collected by the liquid waste system, a through-wall leak would be apparent in this inventory. 'Ihe inventory is reviewed daily by the Shift Supervisor ard weekly by the Site Radwaste (bordinator.
. Radiation monitors within the liquid wasta system would also detect any leakage frm these lines.
'Ihe portion of the noted lines is readily isolatable should a leak occur. 'Ihc portion has double-valve isolation from the primary system on a pressurizer low level signal.
It could be easily isolated by the control rem operators should the line be discovered to be leakirg, leakage exams at operating pressure were perforud on these lines a
periodically as required by ASME XI durirg the first two periods of the first 10 year Interval. Also, these lires were included in the Boric Acid Walkdown performed in September 1989 (7R).
In all these cases, a VI-2 examination, by certified exa h o, was performed.
A Working Group of the ASME XI Otanittee has r-nded that a leak test (nonal operating pressure) be required in lieu of current hydzt>-testing. 'Ihis is based on experience shcuirg minimal elevated pressure tests do n2 generate adequato stress loads to propagate existing defects. '1his ra_>
- -vdation was pasW by the Cbde Sub-comittee and is presently under ruview by the Main Code Ctanittee.
Based on the above, the subject lines have been determird to be structurally sourd ard functionally operable.
System Operability Ae w = nt for the IN-1 Auxiliary Feedwater System
'Ibe cortion of the Unit 1 Auxiliary Feedwater Systcn not tested (i.e., pipirg bounded by valves FCV-W-103A ard 1W-606 ; FU/-N-103B and 1W-608) is cor. sic % rod to have maintainod structural integrity and its functioml operability based on the follwing activities and surveillances:
'Ibe lines in question are schedule 80s ard ratcd at 1662 psig. 'Ibe maxinum operatirg pressure to which these lines will be exposcd under norml ard emergency operatirg conditions is 1155 peig. Anple mrgin exists in the design of these lines to acmeMate operatiry pressures.
Leakage exams at operatirg pressure were performed on these lines periodically as ruquirtd by ASME XI durity the first two periods of the first 10 year interval ard the first period of the secord 10 year interval.
Operational Surveillanx 'Ibsts (OST 1.24.2 ard 1.24.3),
" Motor Driven Auxiliary Fecd Pump Test (1FW-P-3A)" ard
" Motor Driven Auxiliary Feed Pump Test (1FW-P-3B]"
respectively are perfomed gaarterly. leakage through the subject lines would be detected by these OST's.
Operational Surveillance 'Ibst (OST 1.24.8), " Motor Driven Auxiliary Feed Pump Check Valves ard Flow 'Ibst" is performed durirg each refueliry. Icakage thtugh subject lines kulld be detected by this Cer.
Furctional Tests ('IOP 1-88-06 ard 'IOP 1-89-22) were performed on the subject lines durirg the first period of the second interval with no detection of leakage.
Valves 1FW-606 ard 1FW-608 are normally open and provide an uninhibited recirculation flow path to the Primary Plant Demineralized Water Storage Tank. In all practicality, the subject lines could be considered as open-ended pipirg where an unimpared flow verification would fulfill the hydrostatic testirg requirement.
Any leakage from this pipiry would erd up in the safeguarrls area sump which would result 5 increased sump levels or sump pump cyclirg rate.
i
A Workirg Group of the ASME XI Ctannittee has thsded that a leak test (rcrmal operatirq pressure) be required in lieu of current hydro testirg. 'Ihis is based on experience showiry mininal elevated pressure tests do not generate adequate stress loads to propagate existing defects. '1his
-a -xdatlan was passed by the Code Sub-car,unittee aml is In presently under review by the thin Ctde Ctxanittee.
Based on the above, the subject lines have been determired to be structurally sourd and functionally operable.
i.'
System Operability Awwat for the BV-1 Steam Generator Blowdwn Sanole Lines The portion of the Unit 1 Steam Generator Blowdown Sanple System (i.e., tubing bounded by valves 1BD-13 ard TV-SS-117B ; 1BD-14 ard TV-SS-117C ; 1BD-15 and TV-SS-117A) are considered to have maintained structural intcgrity ard its functional operability based on the following activities ard surveillances:
The operatirg pressures of the Steam Generator Blowdown Sample System tubirg were cognrud with the pressure ratings for the tubiry sizes and schodules. The smallest margin between the values was approximately 4246 psig. Ample rnrgin exists in the design of these lines to acconti Wte operatirq pressums.
Isakage examinations at operating pressure were perforned on all of the tubing-in question as regaired by ASME Section XI during the first two periods of the fire.10 year inspection interval.
Isakage examinations at operatirq pressure were performed on the Steam Generator Blcw: lown Sample lines as required by ASME Secticn XI durirq the first period of the secord 10 year inspection interval.
Per Chemistry funual Ompter 3 procedure 1-3.49.
Samplirg ard itsting of the Steam Generator Blowdown is performed on a daily basis. Isakage through the subject lines would be detected by this procedure.
A Workirq Group of the ASME XI Canittee has recctumorded that a leak test (normal operating pressure) be required in lieu of current hydro-testirg. This is Lased on experience showirg minimi elevated pressure tests do not generate adequate stress loads to propagate existirq defects. This reactumendation was
!=r# by the code Sub-camittee and is presently urder review by the Main code comittee.
Based on the activities frttu above, the Unit 1 Steam Generator Blowdown Sample System Tubirq (i.e., piping bourded by valves 1BD-13 ard TV-SS-117B ; 1BD-14 ard TV-SS-117C ; 1BD-15 ard TV-SS-117A) have been determined to be structurally sourd and functionally operable.
System Ooerability Assessment for the BV-1 River Water System The portion of the Unit 1 River Water System not tested (i.e.,
piping bourded by 1W-206 and im-207) is considered to have mintained structural integrity and its functional operability based on the follwing activities and surveillances:
The line in question is schedule 40S ard ratal at 1793 psig. The mvimm operatirg pressure to which this line will be exposed under normal and operating conditions is 100 psig. Anple mrgin exists in the design of this line to acotatodate operating pressures.
Isakage exams at operating pressure were perforaxl on these lines periodiaally as rcquired by ASME XI during the first two periods of the first 10 year interval.
In order to properly perfom HAFA 1bst procedures IIT 24.4 and 30.2, the subject line was filled and vented. At the time of test set-up, any leakage fran this piping wuld have been detected.
A Workirq Group of the ASME XI Committee has rended that a leak test (nomal operatirg pressure) be requirui in lieu of current hydro testing. This is based on experience showing minimal elevated pressure tests do not generate adequate stress loads to propagate existing defects. This r e ndation was passed by the Code Sub-Ocnnittee ard is presently urder review by the Main Code Ctmaittee.
Based on the above, -the subject line has been detennined to be structurally sourn and functionally cperable.
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