ML19312C629
| ML19312C629 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 09/21/1977 |
| From: | DUKE POWER CO. |
| To: | |
| References | |
| PROC-770921, NUDOCS 7912190834 | |
| Download: ML19312C629 (44) | |
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ATTACEIENT A OCONEE NUCLEAR STATION UNIT 2 INSERVICE I::S?ECTION AND TESTING PROG?J.'4 4
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ATTAC'd MENT A I OCONEE NUCLEAR STATION UNIT 2 INSERVICE INSPECTION AND TESTING FROGR.Di The following information is provided to describe the Inservice Inspection
, and testing program required by 10CFR50.53a for oconee Unit 2. This in-formation'is in the format requested by Mr. A. Schwencer's letter of November 30, 1977. -
I Inservice Inspection Program A. Applicable ASME Boiler and Pressure Vessel Code:
Section XI 1974 Edition through and including Summer 1975 Addenda (hereaf ter referred to as "Section XI")
B. Period of which the program is applicable:
Forty month period beginnin3 January 10, 1978' C. Components to be examined:
Vessels, heat exchangers, pumps, valves, and piping, will be classified in accordance with NRC Regulatory Guide 1.26
, Revision 3. Ta,ble 1 lists the systems for which Regulatory Guide 1.26 applies; however, it will be'necessary to consult the appropriate Duke drawings for the exact system, or partial system, boundaries.
. i-D. Examinations and Repairs:
Appropriate examination categories and methods are contained in Table IWB-2600, Table IWC-2600, or Subsection IWD of Section XI.
These tables are referenced, as required, in Table 1 of this Attachment.
In general, volumetric examination will be performed by ultrasonic techniques. Main steam and feedwater, however, will be radiographed where possible. Surface examination will be
,- performed by dye penetrant or magnetic particle.
Repair procedures will be prepared as necessary by the Duke Power Company Steam Production Department. The Quality Assurance Department will review these procedures for compliance with Section XI. Re-examination-to Section XI vill be included in the' repair process.
1 E. Request for relief:
Specific relief is requested from the following ASME Code Section XI~
l requirements which have been determined to be impractical.
- 1. Reactor Pressure Vessel (50-270-1)
- a. Component for which relief is requested:
1)' Name and Number: Reactor Pressure Vessel; 4.
NRC Docket No. 50-270
, 2) Function: Reactor Core Support, Reactor Coolant Pressure
- Boundary j
- 3) ASME Section 111 Code Class: Equivalent Class 1~per NRC Regulatory Guide 1.26, Revision 3 i
- o. ASME Section XI requirement that has been determined to be impractical:
- ASME Boiler and Pressure Vessel Code Section XI, 1974 Edition through Summer 1975 Addenda. Paragraph IWB-2411; Subarticle IWB-2500; Table IWB-2500 Category 3-D; Table IW3-2600 Item
- No. Bl.4 t c. Basis for Requesting Relief
l
(~ ,.
The net effect 5f the above Code requirements is that four nozzles, of a total of eight, must be examined by the end of 30 monthsaof com=ercial operation. Due to core support structures design of Oconee 2r only the two reactor coolant cutle nozzles are accessible without removing the core barrel, which in turn requires-
- complete defueling. This requirement is, therefore, considered
- to.b'e impractical.
- d. and e. Alternate Examination and Implementation Schedule i The following examination program is proposed in lieu of the Code 7
.equirements.
Examination Schedule
- ,' Ccaponents to be Examined '(Elapsed Time Since Ccamercial' i Service Date) ,
. _1 Reactor Coolant Outlet Nozzle Approximately40 months {
1-Reactor Coolant Outlet Nozzle Approximately 80 months <
's 4 Reactor Coolant Inlet Nozzles Approximately'120 months 2 Core Flooding Nozzles Approximately~120 months i
I f .-
L L
- a. . e, 1
Dif ferent nozzles will be examined each inspection.
. ;^
This program reflects the Reactor Vessel Nozzle examination previously contained in the Oconee Technical' Specifications.
- 2. Core Flood and Decay Heat Removal Syste= (50-270-2) i
- a. Component for which relief is requested.
. 1. Name and Number: Core Flood and Decay Heat Removal System (Duke System No. 53A)
Attach =ent Welds 53A and 10ZA
- 2. Function: The Core Flood and Decay Heat Removal System perfor=s two major functions, j (1) Source of low pressure :coling water during l loss of coolant accident.via the low pressure injection'pu=ps and core flood tanks. .
(2) Source of low pressure cooling water for de-cay heat removal during reactor shutdcwn.
Weld 53A is an attach =ent veld for a rwin variable spring hanger. Weld 10ZA is the attach-
'=ent weld for a rigid restraint.
! 3.
~
ASME Section 111 Code Class:
.i Equivalent Class 1 per'NRC Regulatory Guide 1.26, Revision 2
- b. ASME Section X1 require =ent that has been determined to be impractical:
( ASME Boiler and Pressure Vessel Code Section X1 1970 Edition J
including Winter 1970 Addenda. Table 1S-261, Iten No. 4.5, Category K-1 Volumetric Examination.
j c. 3 asis for requesting relief:
i The weld geometry of the attachment welds-on the core flood lines prevent a meaningful volumetric e:: amination of these attachments.
- d. Alternate exasinations:
A surface examination using the liquid dye penetrant. technique !
was perforced on weld 53A and 10ZA. It is requested that this i
examination replace the volumetric examination required in Table 1S-261 of the applicable Section X1.
J
- e. Implementation:
These examinations were performed during the 1977 refueling cutage.
i
- 3. High Pressure Injection System (50-270-3)
- a. Component for which relief is requested:
- 1. Name and Number: High Pressure Injection System (Duke System No. 51A) Attachment Welds 93Z and 89C
. 2. Function: The High Pressure Injection System provides normal make-up water to the Reactor Ccolant System and high pressure cooling water during loss of coolant accident. Welds 93Z and 39L are attachment welds for rigid restraints.
- 3. ASME Section 111 Code Class:
Equivalent Class 2 per NRC Regulatory Guide 1.26, Rev 2.
Currently carried as Class 1 pending review of updated Inservice Inspection Plan per 10CFR50.55a.
- b. ASME Section X1 requirement that has been determined to be 12-practical:
ASME 3ciler and Pressure Vessel Code Section X1 1970 Edition including Winter 1970 Addenda. Table 15-261, Item No. 4.5, Category K-1 Volsmetric Examination.
- c. Basis for requesting relief:
The weld geometry of the attachment welds on the core flood lines prevent a. meaningful volumetric examinatior. of these attachments.
Under the provisions of the updated Inservice Inspection Code, these' welds will not require a volumetric examination.
- d. Alternate examinations:
A surface examinatica using the liquid dye penetrant technique was performed on welds 93Z and 89C. It is requested that this examination replace the volumetric examination required in Table 15-261 of the applicable Section X1.
- e. Implementation:
These examinations were performed during the 1977 refueling outage.
Page 2 of 5 Table 1 Description of Inservice Inspection Program ;
l Quality Group Description Category and Method A Reactor Coolant Pressure Boundary Table IUB-2600 of Referenced (Ref: 10CFR50.2) Code A liigh Pressure injection System Table IWB-2600 (1) Letdown line from RCS to OClV (2) Injection lines from RCS to 0CIV B liigh Pressure injection Table IWC-2600 (3) Letdown line from OCIV to LDST and portions of Chemical Addi tion, Puri fication, and Coolan t Treatment Systems (4) Injection lines from LDST to OCIV'-
C Iligh Pressure injection ,
Subsection IWD (5) LDST branch lines to Nitrogen, Ilydrogen, Sampiing, Radwaste, and RV seal return (6) RCP seal supply; seal bypass; standpipe fill; seal leak off A Low Pressure Injection, Core Flood, and DilR IWB-2600 (1) LPl f rom branch on core flood 1 ine to outside containment isolation valve (2) Core flood lines from RV to second check valve
.- on line to CFT (3) DilR from RCS to OCIV
'See page 'S for lis t of abbreviations.
Part 1 Page 3 of 5 Table i Description of inservice inspection Program Quality 2 Group Description Category and Method 3
-B Low Pressure injection, Core Flood, and DilR Table IWC-2600 (1) Core flood 1ine from 2nd check valve of f RV to 6 l CFT (5) CFT and branch connections to CFT (e.g. , llPI, !!2, and CA)
(6) Remainder of LPI sys tem as indica ted on P0-102A-1, PC-102A-2, PO-102A-3 B Reactor Dullding Spray (Entire System) Table IWC-2600 C Spent Fuel Cooling: Portion as indicated on Subsection IWD drawings P0-101 A-1, P0-101sA-2, P0-10iA-3 f
8 Main Steam Table IWC-2600 (1) its line from OTSG to turbine stop valves C Main Steam Subsection IWD (2) Branch linc to emergency feed pump turbine -
B Feedwater Table IWC-2600 (1) . From OlSG to fi rs t check valve outside containment; includes normal and emergency feedwater lines Q
i Part't Page la of 5 Table 1
- Description of inservice inspection Program i
l Quality 2 ,
Group Description Category and itethod 3 1C Feedwater Subsection IWD .
(2) Emergency feedwater lines from UST to emergency FW pump to connection on normal and emergency feedwater 1Ines (3) Emergency feedwater lines from CCW to correction line to upper feed ring C service Water Systems Subsection IWD Only the portion of the appropriate service wa ter sys tems required to perform' the safety function per Regulatory Guide 1.26 will be subject to inservice i ns per. t i on . These portions of systems will be indicateel on the appropriate Duke diagraninatic layout draw;ngs. ._
0 e
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Part 1 Page 5 of 5 Table 1 List of Abbreviations CA Sampling System CCW Condenser Circulating Water System CFT Core Flood Tank DHR ' Decay Heat Removal FW Feedwater HPl High Pressure injection LDST Letdown S torage Tank MS Main S team Sys tem N2 Ni trogen Sys tem OCIV Outside Containment isolation Valve OTSG Once Through S team Generator RCP Reactor Coolant Pump RCS Reactor Coolant Sys tem RV Reactor Vessel UST Upper Surge Tank e
e O
4
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II Pump Testing A. Applicable ~ASME Boiler and Pressure Vessel Code i
Section X1 1974 Edition through and including Summer 1975 Addenda B. Period for which the program is applicable twenty month period 4
beginning January 10, 1978.
C. Pumps to be tested 4
Pumps which'are considered to be ASME Class 1, 2 or 3 when classified in accordance with NRC Regulatory Guide 1.26 and i
i which are provided on emergency power source will be tested. ;
A listing of these pumps, the parameters to be measured and I
the test internal is provided in Table 2.
1 i D. Request for Relief
)
The following specific relief is requested for the provisions of the.ASME Section X1 requirements which have been determined j to be impractical to meet.
!- 1. (a) Requirement: IWP-3300, IWP-3400 (a) Monthly. testing during normal operation for LPI Pump "2A".
( (b) 7.eason: During normal plant operation, LPI pumps can be run only in recirculation made to the BWST. The j
"2A" pump can caly be tested using a line-up which contains a 3" inch section of pipe. This restrices i
flosi ot a range _frca 1150 to 1550 Gpm. At this icv flow, the installed Flow and DP instrumentation lacks the required accuracy and, due to pump head curve i
' .,' characteristics, repeatability is not readily assured.
(c) Proposed Testing: During cold shutdowns (or monthly j
in the event of frequent shutdowns) the "A" pumps can be fully tested in Decay Heat Removal mode. During normal plant operation, the pumps could be operated in recirculation mode for 15. minutes or until -vibration 1- readings are taken, whichever is longer.
i j
- 2. (a) Requirement: IUP-3300 (Table IWP-3100-1) Flcw for Concentrated. Boric Acid Low Pressure Boric Acid Pumps, l and auxiliary service water pump. '
(b) Reason: Flow measurement devices do not exist in these i{
lines. A station modification would be required to in-stall instrumentation.
- (c) Proposed Testing
- ,None possible for this parameter.-
1 i
l
- s. ..,
- 3. (a) Requirement: IWP-3300-(Table IWP-3100-1), Inlet Pressure Pi, for all pumps which are in operation on a routine basis at the time the test is started.
(b) Reason: Several Systems are normally in operation with one or more pumps running. Taking inlet pressure prior to pump startup would require an additional swap-over to another pump. This (1) increases time required for the test, (2) causes additional wear and tear on the pumps, (3) on some systems could require additional Radiation dose during valve line up prior to swap-over and (4) presents additiona?
opportunity for human error during swap-over which might damage system ccaponents.
(c) Proposed Testing: Inlet pressure will be taken prior to startup of any standby pumps. Since in most systems stand-by and operating pumps are alternated periodically, all pumps will be checked at one time or another. Also, on systems where the inlet piping is ccamon, the operational pump will affect the inlet pressure of the standby pump so that operating pressure on one pump would be the same as pre-star pressure en the standby pump.
- 4. (a) Requirements: IWP-3300 (Table IWP-3100-1), Lube .011 Level for CSAP,LP 3oric Acid Pumps, and auxiliary service water-pump.
(b) Reason: No indication exists to verify lube oil level with-out partial disassembly of the pump.
(c) Proposed Testing: None en this parameter, t .:
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f TA3LE 2 CCONEE NUCLEAR STATION UNIT 2 PUMPS TO BE TESTED IN ACCORDANCE WITH AS:fE SECTION XI o
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x = - = w > a e m ITEMS High Pressure Injection Pumps Mo HS X X X X X X X (2A, 2B, 2C) Primary Make-Up 1 Low Pressure Injection Pumps Mo NA X X X X X X X (2A, 23, 2C) Decay Heat Removal, 2 4 Reactor Building Spray Pumps. . Mo HS X X X X X X X (2A, 23) -
1 AP Emergency Feedwater Pump Mo NA X X X X X X X Concent. 3oric Acid Pump Mo NA 2 2 2 X 2 X X Boric Acid Addition Low Pressure Boric' Acid Pumps Mo LA 2 2 2 X 2 X X (2A, 23) Boric Acid Addition Notes:
i
- 1. HPI and 23 Spray pumps cannot be operated at cold shutdown.
Therefore, per IWP-3400 (a), they will be tested within 7 days after any cold shutdcwn which coincides with the due date of the test.
- 2. See attached list of requested e:cemptions for e:cceptions.
III Valve Testing
( A. Applicable ASME Boiler and Pressure Vessel Code Section XI 1974 Edition through and including Summer 1975 Addenda
- 3. Period for which the Program is applicable.
Twenty month period beginning January 10, 1978 C. Valves to be Tested The inservice testing program for valves at Ocenee Nuclear Station will be conducted in accordance with the methods described in Subsection IWV of Section XI of the ASMF Boiler and Pressure Vessel Code where practicable, consistent with axisting station design.
The valve inservice testing program has been examined to determine which valves should be inspected and the extent of testing required for each valve. The following discussion describes the selection of valves to be tested:
- 1. Categorv A Valves These valves are those for which seal leakage is limited to a specific maximum.a=ount in the closed position for fulfillment of their function. In accordance with IWV-3410, exercise tests of Category A. valves will be performed only on those valves which may be required to change position for the fulfillment of their function in an accident or for safe shutdown of the unit.
- 2. Category 3 Valves Thes~e are Class 1, 2 or 3 valves which may be required to change position for the fulfillment of their function in an accident or for safe shutdown of the unit but for which seat leakage is inconsequential.
- 3. Categorv C Valves Class 1, 2 and 3 check valves which must change position for fulfillment of their safety function will be exercise tested.
The only relief valves for which testing is considered necessary are the pressurizer and main steam relief valves as all other relief valves serve no safety-related function.
- 4. Category D Valves There are no Category D valves at Oconee.
- 5. Category E Valves These valves shall be verified to be locked or sealed.
y Table 3 is a listing of valves which are encompassed by the criteria 7
listed above. Specific instances in which the code requirements are considered impractical are indicated.
D. Specific Relief The valve listing in Table 3 identifies those valves for which certain requirements of the code are impractical. Valves with a comment code "1" are to be tested at time other than power operation as is permitted by Subsections IWV-3410 and IWV-3520. Those valves with a comment code "2" are impractical to leak test due to the lack of appropriate test connections or isolation valves. Those valves with a comment code "3" are impractical to exercise test. No testing will be performed in lieu of the ASME Code Section XI requirements that are impractical.
The intent of ICCFR30.55a is to require that throughout the service life of a nuclear facility the inservice inspection program shall meet the requirements of Section XI of editions of the ASME Boiler and Pressure Vessel Code and Addenda which become effective to the extent practical. It is our understanding that the code does not require up3rading of the design of the facility, but rather where practical,_to improve the inspection or testing criteria or methods. In the case of Subsection IWV for valve testing, certain provisions of the code have been identified which are not practical to meet. These primarily are the results of insufficient test connections or isolation valves to enable leak tests to be performed; inappropriate piping configurations to permit exercise testing of check valves; or unaccessibility of ccaponents to verify operation of the valve to be tested. Continued operation without inservice inspection of these com-s ponents is not considered to:be detrimental to the public health and safety over the "as licensed plant" for the following reasons: Many of the valves are subjected to high system pressures during normal operation and unacceptable leakage would be readily detected, e.g. , core flood tank valves CF-3, CF-19, CF-33, etc. Some valves which cannot be tested perform containment isolation functions, however, the downstream piping is adequately designed for accident conditions, e.g., NP-20 reactor coolant pump seal return; many valves which cannot be specifically tested are but one of two redundant isolation valves.
Additionally, all systems are functionally tested during the periodic contain-cent integrated leak rate test to provide assurance of operability. In con-sideration of the burden which would be imposed to enable testing of these com-ponents in accordance with the code, it is not felt that the health and safety of the public would be significantly improved.
Additionally, the following specific relief from the code is requested:
- 1. a) Requirement: IWV-3410(c) Power operated valves.
b) Reason: Power operated valves which operate in very short time periods (in the order of one second) are difficult to accurately time. In these instances, the specified limiting valve of the full stroke time will-generally be considerably greater than the actual full stroke time. In accuracies in timing contribute to not being able to meet the acceptance criteria of IW7-3410(c) (3).
c) Proposed testing: If any valve with a previously measured stroke
, time less than or equal to one second is observed to increase in stroke time to slower than 1.5 seconds, test frequency shall be increased to once each month until corrective action is taken, at which time the original test frequency shall be resumed. In any case, any abnormality or erratic action shall be reported.
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2LP-15 LPI A Ildr. to llPI P X B 2LP-16 LPI il Ilcir. to llP1 P X 11 2BS-5 A RilS Check C X C 2BS-6 11 RBS Check C X C 211 S - 7 A LPI'Ildr. to KilS C X C 2PS-9 B LPI Ild r. to RBS C X C
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X C 2BS-14 A RilS Disch Check C X C 1 211S- 1 6' B RilS Disch Check C .- X C 2BS-19 !! RilS D. isch Check C - X C 1 Will be tested every 5 yrs.
PO-104A-1 2S F-60 Fuel Transfer Canal Fill fl X '
A 2 2SF-61 Fuci Transfer Canal Fill ' ?! 'X A 2 PO-106-A-1 2CS-64 CliAST Outlet P X B PO-l06E-2 2FW-64 Filtered Water to RS fl X A 2 2 FW--65 Filtered Water to RB fl X, A 2-
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ATTACH' MENT 3 I '
OCONEE NUCLEAR STATION UNIT 3 INSERVICE INSPECTION AND TESTING PROGRAM The following information is provided to describe the Inservice Inspection and testing program required by 10CFR50.55a for oconee Unit 2. This in-formation is in the format requested by Mr. A. Schwencer's letter of' November 30, 1977. '
I Inservice Inspection Program 1
A. Applicable ASME Boiler and Pressure Vessel Code:
i 4
Section XI 1974 Edition through and including Summer 1975 Addenda (hereaf ter referred to as "Section XI")
- 3. Period of which the program is applicable:
Although the forty =onths period is required to begin April 17, 1978
- it will be voluntarily implemented on January 10, 1978 to conform
, with Oconee 2 schedule.
C. Components to be examined:
- Vessels, heat exchan,gers, pumps, valves, and piping, will be classified in accordance with NRC Regulatory Guide 1.26 j Revision 3. Table 1 lists the system's for which Regulatory 4
Guide 1.26 applies; however, it will be necessary to consult the appropriate Duke drawings for the exact system, or partial system, boundaries.
D. Examinations and Repairs: ,
Appropriate examinction categories and. methods are contained in
. Table IW3-2600, Table IWC-2600, or Subsection IWD of Section XI.
These tables are referenced, as required, in Table 1 of this Attachment.
In general, volumetric examination will be performed by ultrasonic techniques. Main steam and feedwater, however, will be radiographed where possible. Surface examination will.be
, performed by dye penetrant or magnetic particle.
Repair procedures will'be prepared as necessary by the Duke-Power Company Steam Production Department. The Quality Assurance Department will review these procedures for compliance-with Section XI. Re-examination to Section'XI will be included in the repair process.
4 l'
- 1. Reactor Pressure Vessel (50- 287-1)
I a. Component for which relief is requested:
- 1) Name and Number: Reactor Pressure Vessel; NRC Docket No. 50-270
- 2) Function: Reactor Core Support, Reactor Coolant Pressure Boundary
- 3) ASME Section 111 Code Class: Equivaient Class 1 per NRC Regulatory Guide 1.26, Revision 3
- b. ASME Section XI requirement that has been determined to be i= practical:
ASME Boiler and Pressure Vessel Code Section XI, 1974 Edition through Summer 1975 Addenda. Paragraph IW3-2411; Subarticle IWB-2500; Table IWB-2500 Category 3-D; Table IW3-2600 Ite No. 31.4
- c. Basis for Requesting Relief:
The net effect of the above Code require =ents is that four nozzles, of a total of eight, must be examined by the end of 80 months of commercial operation. Due to core support structures design of Oconee 2, only the two reactor coolant outlet no=zles are accessible with, cut removing the core barrel, which in turn requires complete defueling. This requirement is, therefore, considered to be impractical.
- d. and e. Alt'ernate Examination and Implementation Schedule The',following examination program is proposed in lieu of the Code requirements.
Examination Schedule Components to be Examined (Elapsed Time Since Ccm=ercial Service Date) ,
1 Reactor Coolant Outlet No: le Approximately40 months {
1 Reactor Coolant Outlet. Nozzle Approximately 50 conths 4 Reactor Coolant Inlet Nozzles Approximately 120 nonths 2 Core Flooding No::les Approximately 120 conths Different nozzles will be examined each inspection.
This program reflects the Reactor Vessel Nozzle examination previously contained in the Oconee Technical Specifications.
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Part 2 Page 5 of 5 Table 1 List of Abbreviations CA Sampling System CCW Condenser Circulating Water S/ stem CFT Core Flood Tank DHR Decay Heat Removal FW Feedwater HPl High Pressure injection LDST Letdown S torage Tank MS Main S team Sys tem N2 Nitrogen System OCIV Outside Containment Isolation Valve OTSG Once Through Steam Generator RCP Reactor Ccolant ? cmp RCS Reactor Coolant Sys tem RV Reactor Vessel UST Upper Surge Tank O
t O
II Pump Testing A. Applicable ASME 3 oiler and Pressure Vessel Code Section X1 1974 Edition through and including Su=mer 1975 Addenda 3 Period for which the program is applicable twenty conth period beginning January 10, 1978.
C Pumps to be tested Pumps which are considered to be ASME Class 1, 2 or 3 when classified in accordance with NRC Regulatory Guide 1.26 and which are provided on energency power source will be tested.
A listing of these pumps, the paraceters to be measured and the test internal is provided in Table 2 .
t c
f
(
D. Request for Relief The follcwing specific relief is requested for the provisions l of the ASME Section X1 requirements which have been determined to be impractical to meet.
l l 'l . (a) Requirement: IWP-3300, INP-3400 (a) Monthly testing during normal operation for L?I Pump "3A".
(b) Reason: During normal plant operation, LPl pumps can be run only in recirculation made to the BWST. The "3A" pump can only be tested using a line-up which contains a 3" inch section of pipe. This J:stricts flow to a range from 1150 to 1550 Gpa. At this low flow, the installed Flow and DP instrumentation lacks the required accuracy and, due to pump head curve characteristics, repeatability is not readily assured.
(c) Proposed Testing: During cold shutdowns (or monthly in the event of frequent shutdowns) the "A" pumps can be fully tested in Decay Heat Removal mode. During normal plant operation, the pumps could be operated in recirculation = ode for 15 minutes or until vibration readings are taken, whichever is longer.
- 2. (a) Requirement: IUP-3200 (Table IWP-3100-1) Flow for Concentrated Boric Acid Low Pressure Boric Acid Pumps, and auxilia'ry service water pump.
(b) Reason: Flow measurement devices do not exist in these lines. A station modification would be required to in-stall instrumentation.
(c) Proposed Testing: None possible for this parameter.
- 3. (a) Requir ement: IWP-3300 (Table IWP-3100-1), Inlet Pressure Pi, for all pumps which are in operation on a routine basis
. at the time the test is started.
(b) Reason: Several Systems are normally in operation with one or more pumps running. Taking inlet pressure prior to pump startup would require an additional swap-over to another pump. This (1) increases time required for the test, (2) causes additional wear and tear on the pumps, (3) on some systems could require additional Radiation dose during valve line up prior to swap-over and (4) presents additional opportunity for human error during swap-over which might damage syste= components.
(c) Proposed Testing: Inlet pressure will be taken prior to startup of any standby pumps. Since in most systems stand-by and operating pumps are alternated periodically, all pumps will be checked at one time or another. Also, on systems where the inlet piping is co==on, the operational I pump will affect the inlet pressure of the standby pump !
so that operating pressure on one pump would be the same (
as pre-start pressure on the standby pumo. l
- i
- 4. (a) Requirements: IWP-3300 (Table IWP-3100-1), Lube 011 Level l l
for C3AP,LP Boric Acid Pumps, and auxiliary service water pump. I
i t
TA3LE 2
/
OCONEE NUCLEAR STATION UNIT 3
- PUMPS TO BE TESTED I
- ; ACCORDANCE WITH ASME SECTION XI o
b a
e u o - u u u o o
- : > =.
m m m o a m
= m a a o o s u o o = H P y c u u o * -
C u 4 L *** *** M C
$ u . n U e u o u 2 u v u u y - e u o A A C 3 u e : - - ~ c o =
w w ~ = w > a = m ITE'i High Pressure Injection Pumps Mo HS X X X X X X X (3A, 33, 3C) Primary Make-Up 1 AP Lev Pressure Injection Pumps Mo NA X X X X X X X (3A, 33, 3C) Decay Heat Re= oval 2 Reactor Buildin;; Spray Pumps Mo HS X X X X X X X (3A, 33) 1 AP Low Pressure Service Water Pumps Mo NA X X X X X X X (2A, 33)
Spent Fuel Fool Cooling, Pumps Mo NA 2 2 X X X X X (3A, 33)
Emergency Feedwater Pump Mo NA X X X X X X X Concent. Boric icid Pump Mo NA 2 2 2 X 2 X X (Boric Acid Addition)
Low Pressure Loric Acid Pump Mo NA 2 2 2 X 2 X X (3A, 33) Boric Acid Addition Auxiliary Service Water Pucp Mo NA 2 2 2 X 2 X X Notes:
- 1. HPI and R3 Spray pump cannot be operated at cold shutdown.
Therefore, per IWP-3400 (a), they will be tested within 7 days after any cold shutdown which coincides with the due date of the test.
- 2. See attached list of exemptions for exceptions.
- - ~. . . - .
4 l *.
(b) Reason: No indication exists to verify lube oil level with-
- < out partial disassembly of the pump.
l (c) Proposed Testing: None on this parameter.
l
- 5. (a) Requirement: IWP-3300 (Table IWP-3100-1) Flow Measurement for Low Pressure Service Water Pump 33.
(b) Reason: 3-LPSW pumps supply two headers, LPA and LP3. A l header can be isolated for testing flow through A pump.
i However B pump flow cannot be measured since 3 header supplies i all essential loads which can't be isolated. Neither can I
~
B pump be lined up to A header.
(c) Proposed Testing: All other parameters will be tested ca 3 pump. The ability of 3 pump to supply the normal require-ments of 3 header (which is approxi=ately-the same as .73 flow) will verify the general performance of the pump.
- 6. (a) Requirement: IWP-3300 (Table IWP-3100-1) Suction pressure measurement for Spent Fuel Pool Cooling (Unit 3) Concen-trated Boric Acid. Low Pressure Soric Acid pumps, and auxiliary service water pump.
(b) Suction pressure instrumentation does not exist for these pumps and station modifications would be required for installation of gauges.
(c) Proposed Testing: Level indications exist for the pool /
i tanks which supply these pumps. These levels, along with ,
j known station head deferences from "0" level to pump suction, '
} will provide a rough indication of pump suction pressure.
Velocity losses should be relatively constant from test _to test, assuming repeatability of flow rates and valve positions.
III Valve Testing A. Applicable.ASkEBoilerandPressureVesselCode i
! Section.31 1974 Edition through and including Summer 1
1975 Addenda.
1 3. Period for which the Program is applicable although the twenty month period is required beginning April 17, 1977 the program will be im-i plemented on January 10, 1977 to conform to the Oconee 2 schedule.
j C. Valves to be Tested The inservice testing program.for valves at Oconee Nuclear _ Station vill be conducted in accordance with the methods described in Subsection IWV of Section XI of the ASME.3 oiler and Pressure Vessel Code where practicable, consistent with existing station design.
The valve inservice testing program has been_ examined to determine which valves should be inspected and the extent of testing required i
for'each valve. The following discussion describes the. selection of valves to be tested:
g ~x- -eo , y y ,,, a rte - p q p e
i
~1. Category A Valves These valves are those for which seal leakage is limited to a specific maximum amount in the closed position for fulfillment
- of their function. In accordance with IWV-3410, exercise tests of Category A valves will be performed only on those valves which may be required to change cosition for the fulfillment
, of their function in an accident or for safe shutdown of the unit.
, 2. _Cate2orv 3 Valves These are Class 1, 2 or 3 valves which may be required to change position for the fulfillment of their function in an i accident or for safe shutdown of the unit but for which seat leakage is inconsequential.
i 3. Categorv C valves Class 1, 2 and 3 check valves which cust change position for j fulfillment of their safety fanction will be exercise tested.
The only relief valves for which testing is considered necessary are the pressurizer and main steam relief valves as all other relief valves serve no safety-related function.
! 4. Category D valves
, There are no Ca,tegory D valves at Oconee.
- 5. Catecorv E valves 4
These valves shall be verified to be locked or sealed.
i Table 3 1,'s a'11 sting of valves which are encompassed by the criteria listed above. Specific instances in'which the' code requirements are considered i= practical are indicated.
D. Specific Relief The valve listing in Table 3 identifies those valves for which certain requirements of'the code are impractical. Valves with a comment code j "1" are to be-tested at time other than power operation as is permia.ted
! by Subsections INV-3410 and IWV-3520. 'Those valves with a co= ment. code
, "2" are impractical to leak test due to the lack of appropriate test l
connections or isolation' valves. Those valves with a comment code -
j "3" are impractical to exercise test. No testing will be performed in lieu-of 'the AS}E Code Section XI require =ents that are impractical.
The intent of 10CFR50.55a is to require that throughout the service. life of a nuclear facility:the inservice inspection program shall meet the requirements of Section XI of editions of the ASME Boiler and Pressure Vessel Code and Addenda-c- :
, . - - , - . - - - - - ---m a p , ,,4,,.- ,- -e ~ -r- --g- , -
i.
which become effective to the extent practical. It is our understanding that the code does not require upgrading of the design of the facility, but rather where practical, to improve the inspection or testing criteria or methods. In the case of Subsection IWV for valve testing, certain provisions of the codi have been identified which are not practical to meet. These primarily are t.e results of insufficient test connections or isolation valves to enable leak tests to be performed; inappropriate piping configurations to permit exercise testing of check valves; or unaccessibility of components to verify operation of the valve to be tested. Continued operation without inservice inspection of these com-ponents is not considered to be detrimental to the public health and safety over the "as licensed plant" for the following reasons: M.2ny of M.e valves are subjected to high system pressures during normal operation nd unacceptable leakage would be readily detected, e.g., core flood tank va7.ves C7-3, CF-19, CF-33, etc. Some valves which cannot be tested perform cor.tainment isolation functions, hcwever, the downstream piping is adequately designed for accident conditions, e.g., HP-20 reactor coolant pump seal return; many valves which cannot be specifically tested are but one of two redundant isolation valves.
Additionally, all systems are functionally tested during the periodic contain-ment integrated leak rate test to provide assurance of operability. In con-sideration of the burden which would be imposed to enable testing of these com-ponents in accordance with the code, it is not felt that the health and safety of the public would be significantly improved.
Additionally, the following specific relief frca the code is requested:
- 1. a) Requirement: IWV-3410(c) Power operated valves.
b) Reason: Power operated valves which operate in very short time
- periods (in the orderJaf one second) are difficult or accurately time. In these instances, the specified limiting valve of the full stroke time will generally be considerably greater than the actual full stroke time. In accuracies in timing contribute to not being ableito meet the acceptance criteria of IWV-3410(c) (3).
c) Proposed tasting: If any valve with a previously measured stroke time less than or equal to one second is observed to inpresse in stroke time to slower than 1.5 seconds, test frequency shall be increased to once each month until corrective action is taken, at which time the original test frequency shall be resumed. In any case, any abnormality or erratic action shall be reported.
, - . - - -- . - =_. . _ . - . . -~ _ - - . - - . - . - _
TAI!LE 3
, OCONi:fi IJf!1T 3 COh'POPJ1ANCl: 1:ITil ASiiF. SECT 1011 XI. SUBSECTION IW t
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' Valve No. Valvo Name o n 7' oo nom o o ~: %; c' 3 Conmenta i
PO-100A-3 3RC-67 Pressurizer Relief R X C 3RC-68 Pressurlzer Relief R X C PO-101A-3
' 311P-24 A IIPI Pump Suct. From IBiST P X II
-311P-25 C llPI Pump Suct. From IlWST P X B 311P-101 A IIP 1 Suct. Check Viv. C X C 1 311P-102 C IIPI Suct. Check Viv. C X C 1 311P-105 A IIP 1 Disch. Check Viv. C X C 311P-109 B llPL Disch. Check Viv. C X C 311P-113 C llPI Disch. Check VLv. C X C
.3CA-85 BAMT to LDST C X C 3CA-73 CilAST to LDST C X C 311P-16 Makeup to LDST P X B 3LP-57 LPI to llPI C Train C X C 1 3LP-55 LPI to 11P1 A Train C 'f X C 1 1
i PO-101B-3 311P-3 A LD Cooler Outlet P X X .. A/Il 2 311P-4 . B LD Cooler Outlet P X X . A/B 2 311P-5 LD Cooler Iso.lation P X X A/B l' J 311P-20 RC Pump Seal Return P- ^*
X X A/B 1,2 311P-21 RC Pump Seal Return P X X , A/B 1 311P-26 A Loop Injection P X '
B L 311P-27 B Loop Injection P X B 3tIP-188 h Loop Check Valve C X C 1 3lIP-153 B Loop Check Valve C X C 1 311P-152 B Loop Check Valve C X C 1 3tlP-126 A Loop Check Valve C X 1 4 311P-127 A Loop Check Valve C X 1 311P 194 A Loop Check Vatve C X 1 PO-iO2A-3 3CF-3 A CFT Sampic/ Drain P Y A 2 l3CF-4 B CI'l' Samp'ic/l) rain P :s , A i
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wI - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
a i.
rV FFFFFFFFFFFFFPPLLLL1P CCCCCCCCCCCCC1 P P P P. P. P,P P P P.
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OCO;il:E UNIT 3 CCIOR:IANCE Kill! AfdE SECTIO'! XI, M ..;c.Cif 0:. ;-N .
t-*
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(j er s3 sD$ s?9 S 0$
Drawing No. ,j g g ;;- ?yo gy* 3 g]
V.ilyn Mo. Valv N;m , m. n r.- nm n ra ' n o ;-* =< r. u Cor.cen ta 3LP A LPI lidr. Check VLv. C X C 1 3hP-48 B LP L lhlr. Check Viv. C X C 1 3LP-103 Doron Dilution Viv. P X B 1 3LP-104 Baron Dilution Viv. P X B 1 3LP-15 LPI A lidr. to llPI P X 11 3LP-l6 LPI C lidr. Lo llP1 P X B 3115-5 A RBS Check C X C 3 11 S - 6 . Il Riis check C X C 3llS-7 A LP1 lidr. to It!!S C X C 3 11 S - 9 B LP1 lidr. Lo RBS C X C 311S-3 A RBS Suct. P X B 3 11 S - 4 B RIIS'Suct. P X B
'PO .103A-3 3 11 S - 1 A RBS RB Isol. Valve P X B 3BS-2 B RBS RB Isol. Valve P . X B
, 3flS-ll A RBS Disch Check C X C 311S - 1 4 A RBS Disch Check C X C 1
. Will be tested every 5 yrs.
3BS-16 B RilS D. isch Check C -
X C 3BS-19 B RBS Disch Check C -
'- X C ]
, ,, Will be tested every 5 yrs.
PO-104A-3 -
3SF-60 Fuel Transfer Canal F111 11 X A 2 3SF-61 Fuel Transfer Canal Fill 11 X A 2 PO-106-A-3 3CS-64 CllAST Outlet P X B
'PO-106E-3 3FW-34 Filtered Water to RB M X A 2 3FW-65 Filtered Water to RB M X A 2 3DW-59 DW to RB M X A 2
'a
OCOIIPE UNIT 3 CONPorJ:A;,CI. WP13 A5iE SECTION XI, SUB$ECTION IWV t..
tit n ,o o
- a
- 1 <j n 4 A td H () H o e4 H& :r tr: oo
. Draw ing t,,o.
' <: mt m - oho c. P rr, o r, o
't s en, ~ te v"- in .: rv <n <: n s et, m 7.11ve No. Valve F: nae c.- r w no r- c .- now < sa Comments 3DW-60 DW'to Ril 11 X A 2 PO-107A-3 3CS-5 QT RB Isol. P X X A/t 2
.3CS-6 QT Ril Isol. P X X A/B 3CS-12 QT Recirc. Check C X X A/C 1,3 3CS-11 QT Recirc. Check C X X A/C 3 3GWD-12 QT Vent P X X A/B 2 3GWD-13 QT Vent P X X A/B PO-107B-3
- 3 LWD-1 Normal Sump Suct. P X 'X A/B 2 3 LWD-2 Normal Sump Suct. P X X A/B PO-107D-3 3 LWD-97. Ril Sump to LAWT M, X A 2 PO-110A-3 3CA-17 IIAMT to Makeup Filters C X C 3CA-18 BAMT to flakeup Filters M X; B 3CA-39 Caustic to LP Soction M X , B 3RC-5 Press. Steam Sample P X X A/B
' Press. Water Sampic P X X '- A/B 3RC-6 3RC-7 -Press. Sample P X X A/B ,
3FDW-105 OTSG 11 Sample P X X s A/B 3FDW-106 OTSG A Sample P X X A/il 3FDW-107 OTSG A Sample l' X X A/li 3FDW- LO8 OTSG h Sample P X X A/B PO-116A3 3PR-1 Ril Purge Outlet P X X A/B 3PR-2 RB Purge Outlet P X X A/B 3 Pit-7. RB Radiation Ifonitor P X X A/II i 3PR-8 RB Radiation Monitor P X X A/il
(_ . .
OCONEC UNIT 3 CONrGU!!/siiCF. W1'ill AS:17. SECTION XI, SU3SECTION ItN U 9 ,a R . : !? .: n 6' O*
-Ditwinau No- a N' m ci
<.: .: n '. 3 ci ei in u: -: o n Wilve No. Valve Nat.n 5 o "r no nc< n o N 43 Comnents 3PR-9 RB Radiation Monitor X X P A/B 3PR-10 RB Radiation ?!onitor P X X A/II 13PR-6 Ril Purge Inlet P X X A/B 3PR-5 RB Purge Inlet P X X A/B PO-121 A--3 3FDW-93 EFDW OTSG A C X C 1 13FDW-95 EFDW OTSG 11 C X C 1 PO-121B-3 #
3FDW-101 EFDW to OTSG B C X C 1 3FDW-99 EFDW to OTSG A C X C 1 3FDW-33 EFDW to OTSG A P X B- 1 3FDW-36 EFDW to OTSG A P X 11 1 3FDW-38 EFDW to OTSG A P X 11 1 3FIM-42 El0W to OTSC B P X B 1 3FDW-45 EFDW to OTSG B P X 11 1 3FDW-47' EFDW to OTSG B P X X A/B 2 2G-23 OTSG B Dra1.n F1 X A 2 FDW-103 OTSG A Drain P X X :- A/B 2
.2G-23 OTSG A Drain il X -
A 2 PO-122A-1 '...
3MSI-16 Flain Steam Relief R X -
C PO-124C 3LPSW-6 IISW to RCP Oil Coolers P X X A/B 1 3LPSW-15 LPSW from RCP 011 Coolers P X X A/B 1 31PSW-18 LPSW from RBCU A P X d 3LPSW-21 LPSW from RilCU B P X B 3LPSW-24 LPSW from RBCU C P X 11 3tPsW-4 LPSW DII Cooler Outlet P X B 3LP.SW-5 LPSW Dil Cooler outlet P X B 3LPSW-75 LPSW D11. Cooler Outlet C X C
(_
OCONEE UNIT 3 CONPORM.WCE WITil ASME SECTION XI, SUBSECTION IW M
!? 9 ~8 -
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3tPSW-76 LPSW Dil Cooler Outlet C X C 31,PSW-404 LPSW Dil Cooler Outlet P X 11 3LPSW-405 LPSW D!I Cooler Outlet P X B 3!.PSW-108 RI!CU Outlet ?! E X
[0,-12711 J K- t % N Is lation 11 X A 2 2
3N-107 !! X A 2 3N-116 ?! X A 2 3N-119. M X A 2 30A-27 M X A 2 3CA-29 M X A 2 3N-131 C X X A 2,3 3N-129 C X X A 2,3 PO-137- ,,
3 fi A-5 IIA 1 solation Valve M Xs A 2 3hA-33 IIA 1 solation Valve M X A 2 PO-144A '
3CC-20 CC to RCP C XR X A/C 1,3 3CC-24 CC to RCP C, XR ,
X A/C 1,3 3CC-76 CC to CitD Service Structure C XR X A/C 1,3 3CC-77 CC to CRI) Service Structure C XR
,X A/C 1,3 3CC-7 CC from RCP P X X A/B 1,2 3CC-8 CC from RCP P XR X A/B 1 0-472 ,
31A-90 Inst. Air to Ril 11 X A 2 3[A-91 Inst. Air to Ril M X A 2 0-672 3LPT-24 Leak Rate Test M X A 2 3L'R"r-25 ' Leak' Rate Test M X A 2 3LRT-17 ., Leak Rate Test M A ,,
\
l Valve Type R - Relief valve '
j P - Power-operated valve, electric or pneumatic C - Check valve M - Manual valve
( Leak Test X - Required Safety Valve Test X - Required Check Valve Test. X - Required Locked Open or Closed X - Required to be locked open or closed during power o,neration Comments 1 - Valve cannot be exercised during power operation 2 - Provisions for Icak testing valve do not exist due to piping configuration 3 - Provisions for exercising valve do not exist Attachment 3, Comment 2 R in Leak Test Column indicates test will be done at Refueling Shutdown (all lehk tests on valves will be done annually, at refueling)
Attachment 3, Comment 3 SD Deside Exercise Testi. Indicates test will be done at cold shutdown per 1WV-3410(b)(1),otherwise all exercise tests are done quarterly.
Check tests are done as indicated on thin list and attach'ed pages eb e
a
'q. .
g
-