ML17276A994
| ML17276A994 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 01/13/1982 |
| From: | Bouchey G WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0619, RTR-NUREG-619 GO2-82-36, NUDOCS 8201290100 | |
| Download: ML17276A994 (36) | |
Text
"REGULATORY "
ORMATION DISTRIBUTION SYM
('RIBS)
'ACICESSION NBR:8?01290100 DOC ~ DATE: 82/01/13NOTARIZED:
NO FACIL:.50 397 KPPSS Nuclear'.Projeatr Unit-2r,.washington )Public Powe
'AUTH~ NAME, AUTHOR AFFILIATION BOUCHEYiG,D.
Nashin'gton
<<Publ ic iPower 'Supply 'System RE'CIP ~ NAME RECIPIENT AFFILIATION SCH'4'ENCERrA ~
Licensing Branch '2
SUBJECT:
Forwards util response to NUREG 0619',"-BIIItR Reedwater
- Nozzle, 8 Control Rod Drive Return l.ine Nozzle iCracking,,"
Response
<<should close out open i:tern in facility draft SER ~
DISTRIBUTION CODEi IB001S ACOPIES )RECEIVED:LTR.g
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SIZEt~ p TITLEt PSAR/FSAR AMDTS and Related
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'30 NRR/DE/MTEB l7 NRR/DE/SAB 2'RR/DHFS/HFEBrl0 NRR/DHFS/OLB 3Q NRR/DSI/AEB 26 NRR/DSI/OPB 10 NRR/DSI/ETSB 12 NRR/DSI/PSB 19 NRR ~/RSB 23 tS
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PDR Washington Public Power Supply System P.O. Box 968 3000George Washington Way Richland, Washington 99352 (509) 372-5000 January 13, 1982 G02-82"36 SS-L-02-CDT-82-016 Docket No. 50-397 Mr. A. Schwencer, Director Licensing Branch No.
2 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C.
20555 BRCpg>rpg, J~,4,'jaggy)gyp~
3 RCC
Dear Mr. Schwencer:
Subject:
NUCLEAR PROJECT NO.
2 WNP-2
RESPONSE
TO NUREG-0619 Enclosed are sixty (60) copies of the WNP-2 response to NUREG-0619, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking".
The submittal of this response should close out this open item in the WNP-2 draft SER.
Very truly yours, G.
D. Bouchey Deputy Director, Safety and Security CDT/jca Enclosures cc:
R Auluck -
NRC WS Chin BPA R
Feil NRC Site
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WNP-2
RESPONSE
TO NUREG 0619 "BWR FEEDWATER NOZZLE AND CONTROL ROD DRIVE RETURN LINE NOZZLE CRACKING" M. P.
Reis'anuary 8,
1982
'1 1
FEEDWATER NOZZLE In Safet Evaluation for the General Electric To ical Re ort "BWR
)
-02 the Feedwater Nozzle/S ar er Final Re ort Su lement NRC staff concurs with GE's conclusion that "a particular solution must be devised for each BWR.
.." and stated that "plant specific review will be necessary in order to determine what combination of modifications is acceptable and necessary."
WNP-2's solution to the feedwater nozzle cracking problem is described below in its responses to specific NRC technical and administrative positions as set forth in NUREG-0619.
The Washington Public Power Supply system Nuclear Project No.
2 (WNP-2) is a General Electric BWR/5.
WNP-2 has six feedwater nozzles with forged tee spargers welded to the nozzle thermal sleeves.
WNP-2 sparger has radiused flow holes.(>)
The operating feedwater temperature is approx-imately.4200 at 1005 rated thermal power.
1 'he top mounted elbow option was evaluated for WNP-2.
However, it was concluded that the radiused flow holes (rather than square-edged holes) would adequately resolve the flow hole cracking problem
(
Reference:
WPPSS Position Page t3, Reactor Vessel Feedwater
~5' i,
24, 1978.
TECHNICAL POS ITIONS Position Interference-fit spargers are not acceptable because their
, 4 1(2)(2) eFficacy is expected to decline with time as the interference 4.2, is lost through wear and plastic deformation.
Response
WNP-2 features the welded sparger design.
This design is specifically stated as acceptable in NUREG-0619, Section 4.1, item (3).
The welded sparger/thermal sleeve in conjunction with the forged tee design joint also eliminates cracking at the junction of the sparger arms and thermal sleeve.
The cracks were attributed to vibrations caused by water flowing through the gap between the thermal sleeve and the nozzle safe end.
The welded joint provides a positive seal against any such leakage.
Position
, Clad nozzles are not acceptable because they are more prone to 4.1(4) cracking and more difficult to inspect.
4.4.2
Response
The WNP-2 feedwater nozzles are not, and never have been, clad.
Position 4.2
Response
The NRC staff believes that licensees and certain applicants must
- have, at a minimum, a low-flow controller having the characteristics described in, Section 3.4.4.3 of NEDE-21821-A( 3).
P WNP-2 design currently features a low flow feedwater controller (RFW-FCV-10) al.so called the "startup valve".
This valve may be used to control feedwater flow in the 0-20K power range.
Startup valve position is determined by reactor vessel water level.
While this low flow controller does not meet all the requirements specified by GE, the Supply System believes it is adequate for its intended purpose.
Appendix I evaluates the WNP-2 design with respect to NEDE-21821-A crack propagation assumptions and conclusions.
2)
NUREG 0619 section (and item) 3)
NEDE-21821-A, BWR Feedwater Nozzle/Sparger Final Report, February 1980
Posit'ion The staff believes that 'licensees and certain applicants must 4.2 reroute the Reactor Water Cleanup System to all feedwater nozzles.
Response
The Reactor Water Cleanup System at WNP-2 has been rerouted such that it discharges into all six feedwater nozzles.
This modi-fication was shown in Topical Report NEOE-21821-02 to contribute the largest improvement in the crack initiation usage factor.
Position We require performance of PT in each nozzle prior to installation 4.3.2.5(1) of sparger.
Response
At the time this request was issued,,WNP-2's welded sparger had already been installed.
A PT of all accessible areas was performed.
No reportable indications were found.
Position We require performance of baseline UT of each nozzle after 4.3.2.5(2) installation of the sparger.
The results are to be made part of the plant's permanent records for future reference.
Response
A UT examination will be performed prior to fuel load on each feedwater nozzle using a full'sized mockup of the nozzles as a
reference standard.
Position Routinely inspect feedwater nozzles and spargers as indicated in 4.3.2.2 Table 2 of NUREG-0619.
Response
WNP-2 with unclad nozzles and welded sparger-thermal sleeve design, will perform routine non-destructive examinations at the following intervals:
Exami.nati on UT(4)
Inspection Interval Refuelin C cles Visual Inspection of Sparger(<)
Routine PT(7) 6(8)
A complete description of the preservice and inservice examination programs for these nozzles is included in the Supply System's response to FSAR question 121.8 which is provided as Appendix 2 of this document.
(4)
UT examination to consist of external examination of feedwater nozzle safe
- end, bore and inside blend radius.
(5)
One (1) nozzle will be inspected each refueling outage.
(6) Visual inspection of flow holes and welds in sparger arms and sparger tees and accessibl e porti ons of the nozzle inner r adius.
(7) Accessible areas only.
(8)
A surface examination will be performed on the nozzle inner radii only if an indication has been discovered by ultrasonic testing and if the indication is suspected to have resulted from service induced cracks.
ADMINISTRATIVE RE UIREMENTS Position 4.I(I) 4.4.2 Modifications to the nozzle and sparger/thermal sleeve must be complete prior to receiving an operating license.
Response
WNP-2's nozzle and sparger/thermal sleeve modifications are complete.
Position All plants that will not have received an operating license by 4.2 June 30, 1983 must complete the modifications before issuance of the license.
Response
MNP-2 system modifications and coranitments described previously wi 11 be completed prior to receipt of an operating license.
Position Operating Procedures should include applicable GE 4.4.2 recomnendations.
Response
(1) Per MNP-2 operating procedures, the RMCU system is normally aligned to the feedwater lines during heatup, operations and cooldown.
(2 ) 'HNP-2 procedures provide for startup valve control of feedwater flow between 0 and 10Ã rated thermal power.
(3) It is anticipated that the extraction steam to the feedwater heaters will be cut in during normal power generation, operations (approximately 15% power.'and above).
(4) Operating procedures for turbine startup are not yet developed.
GE's recoranendation will be considered along with the turbine vendor's recormendations when the procedures are developed.
(5) Plant operating procedures will caution against extended periods of large subcooling (large reactor water-feedwater differential temperature),
especially at high feed water flow rates.
As far as is consistent with power generation requirements, operating procedures will minimize total time spent at large subcooling.
~ Position 4.4.3.2 Upon completion of sparger installation and system
- changes, the applicant must submit to NRC the information described in Sections 4.3.2.5 and 4.4.2 and provide information regarding a
leak detection system (if installed).
The report must include detailed informati on regarding systems modifications and procedures which serve -to prevent crack initiation or cr ack
. growth.
These data will be used in determining any possible changes to the inspection intervals of Table 2.
Response
Section 4..3.2.5 requires preservice PT and UT inspections of each feedwater nozzle.
The results of these inspections will be included in the WNP-2 Preservice Inspection final report.
A suranary of this report is scheduled for submittal to the Coranission by September 30, 1982.
Section 4.4.2 states that the Final Safety Analysis Report for each plant should be amended at the earliest date practicable to include all component and system modifications and operating procedures for NRC staff review and approval.
FSAR Figures 3.2-f and 3.2-11 currently reflect the re-routing of the RWCU piping.
Figure 10.4-5 includes the startup valve, RFW-FCY-10.
Additional FSAR changes will be made, as necessary, as system changes and operating procedures are finalized.
Sumnary:
The Supply System believes that WNP-2's design features of high feedwater operating temperature, welded sparger/thermal sleeve design and unclad feedwater nozzles are sufficient to ensure no feedwater nozzle crack initiation within the design operating lif'ctime of the unit.
Frequent inspections and additional system modifications such as RWCU re-routing and installation of the startup valve further reduce the likelihood of feedwater nozzle cracking becoming a significant problem at WNP-2.
CONTROL ROD DRIVE NOZZLES Position Licensees with 251-inch vessel diameter BWR/5's will be permitted 8.1(4)(g) to cut and cap the Control Rod return line (CRDRL) nozzle without re-routing the CRDRL.
Response
WNP-2 is a 251-inch BWR/5 and has selected this option.
Position:
The licensee will be required to install the following 8.1(4)(a') modifications:
Equalizing valves between the cooling water header and the normal drive movement exhaust water header.
Response
This modification has been incorporated into the design of WNP-2's Control Rod Drive System and will be implemented prior to CRD system preoperational testing.
Position Flush ports must be installed at high and low points of the 8.1(4)(b')
normal drive movement exhaust water header piping run if carbon steel i in is retained.
Response
WNP-2's exhaust water headers are stainless steel.
Position Licensee must replace carbon steel pipe in flow stabilizer loop 8.1(4)(c') with stainless steel and reroute stabilizes'loop piping directly to the cooling-water header.
Response
This modification has been incorporated into the design of WNP-2's Control Rod Drive System and will be implemented prior to CRD system preoperational testing.
Position 8.1, 8.2 Each of the applicable licensees will be required to demonstrate, by testing, concurrent two-CRD-pump operation (if necessary to fulfillrequired flow capacity), satisfactory CRD system opera-ti,on, and required return-flow capacity to the vessel.
Flow must be equal to or in excess of the requirements of the base case.
Response
Satisfactory operation of normal CRD System functions (drive, withdraw, settling, scram) will be demonstrated by preopera-tional testing.
However, with regard to return flow, the NRC requested by letter of January 28, 1980, that GE recalculate the makeup flow capacity for the 251-inch BMR/5 without the CRD return line.
This generic information has been provided by letter of May 2, 1980 from Mr. R. L. Gridly, GE, to Mr. D.
G. Eisenhut, NRC.
The results indicate that the 251-inch BNR/5 CRD Syst'm without a return line (capped Nozzle kl0) can achieve a vessel makeup flow in excess of its calculated boiloff rate of 180 gpm.
In view of the above generic informa'tion, and considering that the CRD system is not designed to typical ECCS standards, the Supply System does not believe that additional testing to demonstrate return-flow capacity is warranted and does not comnit to performing such tests.
Position 8.1
Response
All licensees and applicants, regardless of the particular type of modification selected, must establish operating procedures for achieving CRD flow to the reactor vessel equal to or greater than the boiloff rate of the base case discussed in Section 7.3.
The Supply System does not consider the CRD System as necessary to perform an emergency core cooling function in any accident condition credible at WNP-2.
- Hence, the Supply System does not comnit to establishing operating procedures for aligning the CRD System in a core cooling mode.
~ ~
- APPENDIX 1
EVALUATION OF WNP-2 DESIGN WITH RESPECT TO NEDE-21821A ASSUMPTIONS AND CONCLUSIONS
O.
W ~
~
~
Crack Initiati on GE's report(9) indicates that a plant with a welded sparger 420 F feed-0 water temperature rating, unclad nozzles,,RWCU routed to al) nozzles and 1000 psi turbine roll will have a fatigue usage factor of approxi-mately 0.6.
(A usage factor of 1.0 is equivalent to crack initiation.)
WNP-2 has all these features in its design and operating procedures.
Therefore, the Supply System concludes that crack initiation at WNP-2 is highly unlike1y.
Note:
Since WNP-2 is preoperational, the Supply System must assume that the startup/
shutdown and scram cycles used in the GE analysis will be typical of WNP-2.
Crack Pro a ation GE's best estimate for a base case crack growth yields a crack depth of one inch in 35 years, characterized by GE as "marginally acceptable".
(Best estimate results are consistent with the worst crack depths found in any operating reactors and conservatively envelop the cracking observed in the balance of the plants.)>o The base case analysis assumes o
an improved sparger design 425 F feedwater rating o
unclad feedwater nozzles o
an initial crack depth of 0.25 inch o
on/off feedwater cycling at 6 cycles/hour (CPH) o an "average" operating history 'of startup/shutdown and scram cycles.
Since WNP-2 is preoperational, the average history must be assumed as typical for that plant... With regard to the other assumptions.
(1)
WNP-2 has an improved (welded) sparger with a correspondingly low overall heat transfer coefficient, unclad nozzles and 424.80 feedwater temperature rating.
(2)
WNP-2 feedwater nozzles bores and blend radii have been dye penetr ant examined for cracks with no reportable indications
'found.
Since crack initiation is highly unlikely at WNP-2, the 0.25 inch deep crack is, in itself, a conservative assumption made for analytical purposes.
(9)
NEDE-21821-A, Section 4.7.2.6 (10) NEDE-21821-A, Pg.
4-297
(3)
The 6 cph assumption is conservative since it allows for no type of low flow controller.
WNP-2's startup valve will be used to control flow during low ( < 10K) power operations.
With the exception of a rapidly (60 cph) cycled low flow controller, GE's analysis demonstrates that the use of a low flow controller decreases the crack growth rate.
- Hence, use of the startup valve will tend to extend the time required for a 1 inch deep crack to develop.
(4)
NEDE-21821-A does not assume any efforts to detect and arrest cracks during the plant lifetime.
The Supply System has cormitted to periodic non-destructive examination of the feedwater nozzles and wi 11 take corrective measures if any unacceptable cracks are observed.
The Supply System concludes that the current design and operating phi los-ophy at WNP-2, the conservative nature.-of NEDE-21821-A calculations and the Supply System's commitment to non-destructive examination preclude unacceptable feedwater nozzle crack growth.
1-2
~ ~ ~
APPENDIX 2 SUPPLY SYSTEM RESPONSE TO FSAR QUESTION 121.008
WNP-2 AMENDMENT NO-5 August 1979 g.
121.8 Sheet 1 of 10 (In response to this item, refer to the responses to Items 121.15 and 121.18 on tlute Hatch-2 docket.)
Additional infor-.
mation is required to demonstrate that:
(1) the thermal sleeve/sparger design of the feedwater inlet nozzle has been evaluated with respect to potential nozzle cracking resulting from thermal cycling; and (2)- a program of scheduled augmented inservice inspection has been developed.
Thes inservice inspections should be conducted with a method sufficiently sensitive to provide assurance that small cracks can be detected.
Accordingly, we require you to supply the following information:
a.
That technical basis to assure the structural integrity of both the feedwater inlet nozzle and the sparger.
b.
An evaluation of the feasibility of installing automated ultrasonic testing (UT) fixtures on all feedwater inlet nozzles with particular attention focused on the examination of the nozzle bore region.
c.
An evaluation of the feasibility of performing the internal surface examination by magnetic particle methods.
Your response should contain:
(1')
a description of the nozzle and sparger design including the significant dimensions, the materials of construction and the weld locations; (2) a description of the analyses and test data, referencing appro-priate data previously submitted to the NRC staff if. it is applicable for the WNP-2 facility; (3) the detailed projected crack growth rates, stress levels and usage factors for both the nozzle and the sparger; (4) any plant modifications that.
are planned to reduce the temperature differential between the feedwater and the water in the reactor pressure vessel d r'ng low power operation; and (5) a description of any in-strumentation that will be installed in the re'actor pressu u
3.
re vessel to verify the conclusions of your design analysis.
Several ultrasonic testing 'concepts and procedures have bee'n used to examine the feedwater inlet nozzle regions in operating plants.
Identify which of these ultrasonic testing procedures W3.w'll be used in the WNP-2 facility.
Discuss the influence on crack detection, using your ultrasonic testing method in the WNP-2 facility, of local grindouts.
L21. 8-1
AMENDMENT NO-5 August 1979 Sheet 2 of 10 In addition, provide a description of the augmented inservxce inspection (ISX) program to be implemented including scheduled surface examination, ultrasonic testing and verification of the leak tight integrity of the joint between the therma sleeve and the safe end on all nozzles.
The essential elements of an acceptable program are given in the Appendix attached to this set of questions.
Res onse:
A description of the WNP-2 feedwater nozzle and sparger is presented on Figures 121.8-1 and 121.8-2.
The mechanisms which have caused a cracking in operating BWRs are un e understood.
A summary discussion of problems and the solutions incorporated in the WNP-2 design is presented
'nin the following.
A detailed evaluation of the problems of the feedwater nozzle and sparger is presented, in NEDE-21821,."BWR Feedwater Nozzle/
Sparter Final Report" March 1978.
The solution of the feed-water nozzle and sparger cracking problems involves several element..:.,
including material selection and processing, nozzle c a 1 d elimination and thermal sleeve and sparger redesign.
The following summarizes the problems and solutions that have been implemented in the WNP-2 design.
PROBLEM Sparger Arm Cracks CAUSE FIX Uibration Eliminate clearance between thermal sleeve and safe end.
RPV Feedwater Nozzle Thermal Fatigue Eliminate clad, eliminate leakage with a welded joint, between the sparger and safe end.
The sparger vibration has been attributed to a self-excitation caused by instability ofeakage flow through the annular clearance between the thermal sleeve and safe end.
Tests have shown that the vibration is eliminated if the clearance is reduced sufficiently or sealed.
The solution which has been se ec e
1 t d for WNP-2 uses a welded joint, to assure no leakage.the This feature is also an essential part of the solution o
nozz e cr le cracking problem.
Freedom from vibration over a range of conditions has been demonstrated by the tests reporte d I n NEDE-23604.
121.8-2
~ ~
WNF "2
~
DMENT NO.
5 August 1979'heet 3 of 10 The cracking of the feedwater nozzles is a two-part process.
The crack initiation mechanism as discussed above is the result of self-initiated thermal cycling.
Xf this were the only mechanism present, the cracks would initiate, grow to a depth of appr'oximately 0.25 inch, and arrest.'his degree of cracking could be tolerated, but unfortunately there is another mechanism which supports crack growth.
This mechanism is the system induced transients, primarily the startup/
shutdown transients.,
The welded thermal sleeve arrangement also assists in this area because without leakage, the, heat transfer coefficient between the feedwater and the nozzle are reduced to the point where the thermal stresses in the nozzle are not high enough to cause a significant crack growth.
Analyses presented in NEDE-21821, Section 4.7, demonstrates the benefits of the welded thermal sleeve and of using unclad nozzles.
With these demonstrated
- benefits, WNP-2 does not believe it necessary to install instrumentation for design verification.
WNP-2 has installed an automatic feedwater low flow control
- valve, RFW-FCV-10.
This valve has the capabilitiy to control flow down to 362 gpm, or about 1.25% of total flow.
This valve will substantially reduce the temperature differential between the feedwater and the water in the RPV during low power operation.
The following paragraphs address RPV feedwater nozzle examin-ation questions other than Appendix A to Section 121.
Feasibilit of Xnstallin Mechanized Ultrasonic Scanners I
All feedwater nozzle inner radii, safe-end, and bore re.<ions are capable of automated ultrasonic examination.
Both pre-
~ service and inservice inspections of the inner radii and safe 'ee Note 1
/ end welds will be performed usinq such equipment.
~.Tooling has been contracted from the baseline examination agency that will allow nozzle inner radius scanning by contacting an angle beam transducer to the vessel plate surface adjacent to the nozzle to vessel weld.
The scanner mechanism is removable and would be compatible with any of the six (6) feedwater nozzles.
'The
~ Suppl'y 'Sy'tem is'urrently evaluating the'enefi't of'erform-See Note 2
ing an automated examination of the bore region versus a
manual examination in terms of radiation exposure (examination/
- i. setup time) and,examination coverage.
The te'chniqud 'pr'ovM'ing: "
the be'st balan'ce of "those two factors will be chosen.
Ade-quate access exists for either technique.
121. 8-3
NNP-2 AMENDMHNT NO.
5 August 1979 Sheet.
4 of 10 Scanning of the nozzle bore region can be accomplished from the cylindrical section of the nozzle forging.
A manual examination of this region is possible to accomplish in less than ten minutes of scanning time per nozzle by one operator supported inside the biological shield cut-out.
Data record-ing, should it be necessary, can be accomplished by a second examiner positioned outside the shield using redundant elec-tronic instrumentation and a'nalog recorders.'ssuming ten minutes examination time per nozzle and a radiation field of 150 mr/hr, the examiner would receive 25 mr per nozzle.
The automated examination devices would be mounted on tempor-ary tracks, would be installed just prior to the examination and removed following the examination.
It is not considered feasible to leave the equipment installed during plant oper-ation as installation and removal time is minimal and would be quickly offset by equipment. recalibration and maintenance costs considering the adverse environment such equipment would be subjected to during plant operation.
Feasibilit of Ma netic Particle Examination Handheld magnetic yokes will not rehdily fit in the envelope between the sparger body and the nozzle radius, and yet make good contact with the low alloy steel surface.
Poor contact could result in arc-strikes below the electrodes, these surface defects are localized heat affected zones of higher hardness than the surrounding metal.
If the arc-strike was accompanied by localized cracking, then surface grinding would be necessary to restore the nozzle to its original surface condition.
Con-sidering the above, magnetic particle examination methods are not considered feasible inside the reactor vessel with the present sparger configuration.
I Ultrasonic Examination Methods The nozzle inner radius examination will be made by pulse-echo ultrasonic techniques from the exterior of the reactor pressure vessel by contacting the vessel plate surface.
This technique is similar to that used by the General Electric Company and the firm of Lambert MacGill and Thomas.
Proce-dures for the examination will be in a format consistent with others used by the Supply System, but the technical con-tent will be comparable to procedures previously qualified by the above referenced testing organizations.
121. 8-4
0
AM c DHENT NO.
5 August 1979-Sheet 5 of 10 Examination of the nozzle bore region will be performed by pulse-echo ultrasonic techniques from the cyclindrical section of the nozzle forging using sound beam geometry similar to that used by the General Electric Company.
,The Supply System plans to extend the coverage of this technique toward the inner radius by added sound beam refraction.
Prior to use on the WNP-2 feedwater nozzles, a qualification check is intended to be made on a mock-up to demonstrate the techniques validity.
Should local rind outs be made in the examination surface creating a depression with definable sides,
- depth, and length, the ultrasonic techniques being used would obtain reflections from these cavities.
Such reflections can be minimized by-blending the grind cavity into the surrounding base metal.
This would result in improved detection sensitivitiy to postulated thermal fatigue cracks propagating from the grind cavity.
The Supply System will implement the reactor feedwater (RFW)
RPV nozzle inspection program described below, which addresses Appendix A to this question on an item-by-item basis.
Justification for any deviations from the Appendix A require-ments is presented following the response.
I.
AUGMENTED INSERVICE INSPECTION PROGRAM A.
Preservice Examination The Supply System will perform a PSI ultrasonic examination of RPW nozzle inner radii, bore and safe end regions as described in the WNP-2 PSI Program Plan.
The personnel and UT procedures used will be qualified as described in II.C below.
In addition, a preservice liquid penetrant examina-tion will be performed on the accessible areas of all RPW nozzle inner radius surfaces.
B.
Inservice Examination B.l The Supply Syst: em will perform an ultrasonic examin-ation of 1 of 6 reactor feedwater nozzle inner radii, bore and safe end regions each refueling outage using procedures and personnel subject to the same qualifications used during the PSI examinations.
A different nozzle will be examined each outage.
No surface examinations will be performed on the nozzle inner radii unless such a test is required to verify 121. 8-5
Wtl -2 AMENDMENT NO.
10 July 1980 Sheet 6 of 10 the nature of an indication discovered using the ultrasonic technique
.when the indication is suspected to result from service induced cracks on the nozzle inner surfaces.
In the event an indication is discov-ered and found to result from service induced cracks propagating from the nozzle.inner surfaces, the following actions will be taken:
a.
All remaining feedwater nozzles will be examined using both ultrasonic (from the OD) and penetrant techiques during the refueling outage in which the cracking is verified.
I b.
All surface indications determined to be service induced cracks will be removed by local grinding.
c.
An inspection
- method, such as a leak test, will be used to determine the integrity of each of the RFW thermal sleeve to safe end joints.
d.
Appropriate corrective action will be taken as required and as practical to prevent recurrence of crack initiation.
A program and schedule for implementing such corrective action will be pre-pared and submitted to the Commission prior to its implementation.
e.
A RFN nozzle examination program for su@sequent refueling outages will be modified to include an external ultrasonic examination of all feedwater nozzle inner radii, bore and safe end regions for each scheduled refueling outage for 3 consecutive outages.
If no new indications are discovered, or if new indications are determined to not result from s'ervice induced cracks at the nozzle inner surfaces, the original Supply System program will be resumed.
If after 3 additional outages no new indications resulting from surface induced cracks are detected, subsequent examina-tions will be performed in accordance witn normal ASME Section XI requirements.
The conduct of surface examinations of accessible nozzle inner radius surfaces will continue to be used throughout plant life only to confirm or characterize new ultrasonic indications which are suspected to result from service induced cracks at the nozzle inner surfaces.
121. 8-6
WNP-2 Ai<iE, AENT NO.
5 August 1979 B.2 Sheet 7 of 10 As stated in B.l above, the Supply System will per-form a surface (penetrant) examination of accessible inner surfaces on all RFW nozzles during the pre-service examination program.
Subsequent surface examinations of those surfaces will be performed only to verify the nature of an indication discovered using the ultrasonic technique when the ultrasonic indication provides evidence of previously unidenti-fied service induced cracks.
B.3 See response to B.2 above.
If after the sixth planned refueling outage following commercial operation no indications resulting from service induced cracks are found, the subsequent inservice examinations will be performed in accordance with the normal ASME Section XI requirements.
Any indications resulting from service induce cracks which are subsequently found will result in the corrective action described above.
C.
Thermal Sleeve to Safe End Joint As stated in B.l above, the Supply Syst: em will per-form an inspection of the thermal-sleeve-to-safe-end weld joint, such as "a leak test, only if service induced cracks or some other anomoly is discovered which would bring the integrity of the joint into question.
In that case, the feedwater piping will be filled with water and the area of the thermal-sleeve-to-safe-end joint will be inspected for indications of leakage.
II.
ACCEPTANCE CRITERIA A.
The Supply System will comply with this criteria as stated in B.l above.
B.
The supply System will comply with this criteria as stated in B.l above.
C.
The Supply Syst: em.will comply with option (b), in that both the examination personnel and the proce-dures to be used on the nozzles will be qualified on a full size nozzle mock-up.
Supply System examiners will be trained by individual NDE special-ists having previous experience with the General Electric Company procedures and their nozzle test program.
These examiners will undergo further train-ing, practice, and qualifications on a full size 121.8-7
P% V-2 NDMENT NO.
5 August 1979 Sheet 8 of 10 nozzle mock-uq.," The, mock-up will be unclad if negotiaEi.ons can be reached with a utility owning C such a mock-up.
As an alternative, the examiners will qualiiy on a clad mock-up owned Qy ~t e General;
"-.,glectric Fompan~,
Following the qualiticati'on
- process, the eeeaminations will be conducted under the-direct supervision of the experienced NDE specialists responsible for ultrasonic technique and procedure development--for the Supply System.
III. RECORDING AND REPORTING STANDARDS The Supply System will record crack indications and report inspection results in compliance with the require-ments stated in NUREG-0312.
121.8 JUSTIFICATION OF DEVIATION FROM APPENDIX A I.B.1 Ultrasonic Examinations Fre uenc The Supply Sys-tern will examine only one RFW nozzle per
'efueling outage rather than all nozzles using an ultra-sonic technique from the outside of the vessel.
This is justified for the following reasons, which reflects a
significant advance in the WNP-2 design and operating procedures towards the long term solution of the BWR nozzle cracking problems per NUREG-0312, Section 8.0, Part.
1.
b.
Improved Design:
The WNP-2 RFW welded thermal-sleeve-to-safe-end joint provides a "zero leakage" design.
This design essentially eliminates the primary histor'ical initiating source of nozzle cracking in BWRs.
No Nozzle Cladding:
The WNP-2 RFW nozzle surfaces are not clad.
The likelihood of crack initiation in unclad nozzles is more than a factor of a 5
less than for clad nozzles.
All cracks in BWR feedwater nozzles have initiated in the clad metal.
c.
Proven Examination Technique:
The ultrasonic ex-amination equipment and personnel to be used in performi.ng both baseline and inservice ultra-sonic examinations will be qualified on a full scale mock-up of the nozzle, simulating the nozzle geometry and anticipated fatigue crack defects.
Since the WNP-2 reactor feedwater nozzles are unclad as stated in b) above, a more sensitive examination is possible due to lack of clad/basemetal interface.
a 121. 8-8
WNP AYiE 1ENT NO.
10 July 1980 Sheet 9 of 10 d.
Augmented Examination Frequency:
The above stated program provides RFW nozzle examination coverage at nearly twice the frequency of the ASME Section XX requirements, i.e., all RFW nozzles will be examined within 6 years (approximately) rather than within 10 years.
e.
Feedwater Temperature Controls:
As previously
- stated, WNP-2 has incorporated a feedwater, low flow control valve.
The advantages gained from low 'flow control are identified in Section 4.7 of NEDE-21821.
f.
Projected Crack Growth Rates:
As presented in Section 4.7 of NEDE-'21821, the WNP-2 design should have greater than 35 years of operation, considering our low flow control, prior to an initiated crack reaching 1 inch in depth.
This provides for a minimum of 4 examinations per nozzle before reaching a point requiring repair.
Even if the extremely conservative (factor of 5) upper bound crack growth curve is applied, each RFW nozzle would be examined by the Supply System program prior to a crack becoming 1 inch in depth.
Xt is clear that ample conservatism exists in the Supply System examination frequency of the RFW nozzles.
The above factors, when combined, provide a great deal of assurance that the factors which have led historically to BWR RFW nozzle cracking have been vir-tually eliminated.
Furthermore, any cracking which might occur from unanticipated sources will be dis-covered before propogating to a'ignificant depth due to low flow controls and an augmented examination schedule with state-of-the-art qualified ultrasonic examination techniques.
I.B.2&3 Surface Examinations The Supply System will perform surface (penetrant) examinations of the accessible internal surfaces of RFW nozzles during the preservice inspection program.
Xnservice surface examinations will be performed only when indications of service induced cracking are found using the ultrasonic examination technique.
This is. justified as follows:
a.
Reduced probability of crack initiation and growth as stated in the justification under I.B.1a) through f) above.
WN>'-2 AH NDHENT NO.
10 July 1980 Sheet 10 of 10 b.
Access:
In order to obtain access to perform a
pr:net.rant; u> face.
exa nination of the RPW nozzle surfac.e; duri ng a refueling outage, the vessel water leve3 would i>ave t:o be lowered below th' level of th'parcjers and hydrolaser decon-tami.nation perCormed.
A special shielded work plat form would have to be devised to minimize racliation exposure.
This technique was performed at Vermont Yankee. result.ing in about 15 man rem.
Leak TesL r
Therma1-Sleeve-t:o-Safe-Encl Joi.nt:
The Supply System will perform inset.'vic:e inspection to determine the integrity of t:he thermal-sleeve-to-safe-end joint only when indi.cations of: servi.ce induced cracking are detected u" ing the ultrasoni.c examination technique.
The just:iricat:ion Cor this exception is similar to the justification for. not performing inservice surface exam-inations ci,ted
- above, with the following additional justificati.on:
Test Eli;ecLi>>et'te:;s:
The maximum pressure which could practical,ly be placed on the subject weld joint would be that available from the static head of a f lied spargeror.
approximately 6" of water.
The ef fecti.veness of this test to reveal throughwall cracks in the weld joint is question-
- able, since the weld experiences significan'tly higher difCerential pressure and temperature during operati.on.
L"urthermore, this test would not provide evidence oC other than gross throughwall craclcs whic:h, if and when detected via such a test, will in all likelihood have been resulting in some degree of leaking for a signi-fi.cant peri:od of ti.me.
Is was previously demonstrateda>>y cracks developing as a result of such lealcage will.be detected prior to the crack propagat:ing to a depth which would jeopar-di ze the nozzle integri.ty.
There is, therefore, no apprec:iable benefit from performing the leaicage L... t of the sL'arger.'ther than to deter-mine the sLatus of. Lheir int:egrity in the event service inc3uced cr ac]cs are con C irmed.
Since there i.s iso apCreciable
- benefit, and the cost in dollars and man k e)>1 expo u ce Cor such a test is quit:e high, Lhe perforn>anc>> of this test on a
rout:ine basis is not just:ified.,
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NOTES 1.
Preservice inspection of the safe end welds will be done manually.
The mechanized equipment will be demonstrated capable of examining the safe end welds inservice.
The inner radii weld examinations will be done using mechanized equipment.
2.
Based on the experience of performing the baseline exam of nozzle to vessel welds, the Supply System has concluded that the radiation exposure examining the bore region manually would be less than if the exam was done mechanized.
3.
The Supply System has secured an unclad feedwater nozzle from the scrapped Douglas Point Unit 1 reactor..
The Supply System will use this nozzle to qualify the procedures and personnel for feedwater inner radii examinations.
V' l