ML20079S132
| ML20079S132 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 07/07/1983 |
| From: | Bayne J POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | Vassallo D Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0619, RTR-NUREG-619 JPN-83-64, NUDOCS 8307110314 | |
| Download: ML20079S132 (11) | |
Text
123 Main Street White Plains, NewYork 10601 914 681.6240 A NewYo.rkPbwer a ""iaa v"-
Executwe Vice President Q[
Nuclear Generation July 7, 1983 JPN-83-64 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 Attention:
Mr. Domenic B. Vassallo, Chief Operating Reactors Branch No. 2 Division of Licensing
Subject:
James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 Control Rod Drive Return Line Modifications (NUREG-0619)
References:
1.
PASNY Letter, J.P. Bayne to H.R. Denton, dated.ianuary 11, 1982 (JPN-83-6).
Enclosures 1.
Description of Defect in the Control Rod Drive Return Line --
June, 1983 2.
Proposed Control Rod Drive Return Line Modification 3.
Control Rod Drive Return Line Operating Historf (1975-1983) 4.
NUREG-0619 Implementation Schedule Sumnary for the Control Rod Drive Return Line Dear Sir Inservice inspection of the Control Rod Drive return line (CRDRL), during the current refueling outage, resalted in the detection of a through-wall weld defect in the three-inch piping upstream of the vessel penetration.
A description of this defect is provided in Enclosure (1).
As a result of this defect in the CRD return line, the Authority is proceeding to permanently cap the vessel nozzle. The Authority is scheduled to perform modifications, as described in NUREG-0619, to the CRD return line during the 1985 refueling outage (Reference 1).
Enclosures (1) through (4) to this letter include a description of the defect found in the CRD return line, the proposed modification for capping the reactor vessel nozzle, an operating historf of the CRD return line and an implementation schedule for all NUREG-0619 modifications.
It should be noted that the Authority's proposed approach is slightly different from that described in NUREG-0619 8307110314 930707 ll PDR ADDCK 05000333 P
The NRC is requested to review the Authority's proposed course of action for resolving the immediate requirement to cap the CRDRL nozzle, including the schedule and technical information provided in this letter. The Authority requests your review by July 11, 1983. This is necessary to permit finalization of planning and completion of the modifications without delaying the refueling outage now in progress.
If you have any further questions, please contact Mr. J. A. Gray, Jr., of my staff.
Very truly yours, J
,B' Executive Vic President Nuclear Generation cc:
J. Linville Resident Inspector U.S. Nuclear Regulatory Commission P.O. Box 136 Lycoming, N.Y.
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ENCLOSURE 1 TO JPN-83-64 DESCRIPPION OF THE CRDRL DEFECf Seepage f rom the reactor side of the joint was discovered during the loosening of a pipe support.in preparation for ultrasonic examination.
The rate of leakage was extremely slow, and could only be observed by the formation of a reflective surface.
i The crack was located at Elevation 321, approximately 6 f t.
from the nozzle, at the 11 o' clock position on the pipe as facing the vessel.
Subsequent ultrasonic examination determined that the crack ran parallel i
to the fusion zone of the weld, in the heat affected zone for a circumferential distance of 3.5 inches.
Review of the original construction radiographs revealed a transverse (axial) indication at approximately the same position as the discovered crack. Metallographic examination of this joint will be performed af ter removal of the affected piping to determine the extent and possible cause of this cracking.
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ENCIDSURE 2 TO JPN-83-64 PROPO3ED CONTROL ROD DRIVE RETURN LINE MODIFICATIO3
Background
i Enclosure (3) to this letter provides a detailed operating history of the Control Rod Drive return line (CRDRL).
In sammary, tne CRDRL was in service for less than two years.
In-vessel non-destructive testing, completed one full operating cycle af ter the line was isolated, detected no indications in the blend radius or vessel wall. Control Rod Drive System perforaance has been normal througn tne period daring whicn the return line has been isolated.
it Based on in-vessel video camera inspections daring the current outage, has been determined that tne CRDRL nozzle theraal sleeve actually extends It is the past the vessel wall into the vessel approximately 1/2".
Authority's conclusion that this f act, in conjunction with the short time period that the CRDRL was in service, explains the absence of identified defects at the FitzPatrick plant wnich have been detected at some other plants.
CRDRL Vessel Penetration Capping Modification the Since the presently identified piping defect must be corrected now, Authority intends to permanently cap the CRDRL vessel nozzle during tha current refueling outage. A contingency modification f or capping the CRDRL nozzle was previously prepared.
Because the NUREG-0619 modifications were not scheduled to be completed until the 1935 refueling outage, complete engineering, procedures, material, and planning are not available to support full implementation of all modifications reqaired by NUREG-0619 during the current refueling outage.
In planning for the capping of the vessel nozzle per this contingency modification, the Authority has identified significant potential problems in the areas of personnel radiation exposare ( ALARA), control of heavy loads (NUREG-0612), and personnel safety resulting f roa a need to perform in-vessel work during the current refueling outage. To eliminate the need to enter the vessel to install the reqpired welding purge plug, a plug will be designed and f abricated to allow installation from the drywell side of the nozzle and remote removal from inside the vessel.
The remaining required in-vessel work involves a final liquid penetration examination of the CRDRL nozzle blend radius and vessel wall.
The significant problems associated with in-vessel work are as follows:
a)
It is necessary to emplof personnel inside the reactor vessel to perform the liquid penetrant inspection. Due to the close proximity of tne nozzle to the top of the core, it will be possible to maintain only 2-3 feet of water above the core for shielding.
As a result of j
radiation levels from the fuel and vessel internals, all i
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work cast b3 parfocand froc a lead-shieldsd ptreonnsi The carrise carpandsd from ths Rsactor Building crans.
total estimated partonnel radiation exposure for this work is 60 man Rem.
b) Presently, JAF has a licensing restriction which precludes the handling of heavy loads above irradiated fuel. The estimated weight of the lead-shielded personnel carrier is in excess of 3 tons.
Once the carrier has been designed, a heavy-load analysis per NUREG-0612 must be performed in order to assure the safet/ of the lift, and the previously mentioned restriction would have to be lif ted.
c) Refueling of tne reactor during the current outage has been completed.
It would be a significant impact on the present outage senedale to perform a full core off-load to support this modification. A full core off-load would decrease the personnel radiation exposare for the in-vessel work, but significant exposure would still exist f rom the vessel internals.
A{.ternative Approach to NUREG-0619 Requirements The Authority proposes an alternate approach which is to defer the final liquid penetrant inspection of the nozzle required bf NUREG-0619, and restrict the depth of the nozzle bore inspection to approximately one-half (7") of the total nozzle (14") length.
The total personnel radiation exposure for this work is estimated to be 16 man Rom.
NUREG-0619 Implementation and Schedule Enclosure (4) provides a detailed listing of NUREG-0619 requirements and the Authority's schedule for implementation. Tne method and schedule for implementation is the same as that previously provided by the Authorit/
(in Reference 1 to this letter) with the following exceptions:
a) The Authority intends to cap the CRDRL vessel nozzle during the present outage in advance of completing the other modifications required by NUREG-0619, which will be installed in accordance with the previously submitted schedule.
Justification:
Since the FitzPatrick plant has demonstrated satisf actorf operation of the CRD Sf ates since 1977 with the CRDRL isolated, the remaining NUREG-0619 modifications need not be completed at this time and will be performed during the 1985 refueling outage..
f b) A final liquid panstrent examination of the outsr 7" of ths CRDRL nozzle bore will be parformsd now.
Tna liquid penetrant examination of tha innar 7" of the bore, the blend radius and vessel wall will be deferred and performed from inside the vessel during the refueling outage for Cfcle 12, or tne next time in-vessel penetrant inspection of the f eedwater nozzle /sparger is regslired of NUREG-0619, Table 2.
Extensive planning to minimize personnel radiation exposure and resolution of NUREG-0612 heavy load concerns will precede these operations.
Not performing a liquid penetrant examination inside the vessel is considered acceptable based on the following:
L. Previous PT, UT and visual examinations of the nozzle blend radius and local vessel wall as described in Enclosure (3) have found no indication of defects.
These examinations have confirmed the effectiveness of the thermal sleeve during the short period of time during wnich the CRDRL was in operation.
- 2. Tne plant operated a relativelf short period (less than 2 years) with the CRDRL in service.
- 3. To compensate for the extended period until final PT, the Authority proposes to perform a television visual inspection (less than 1 mil resolution) of the nozzle blend radius and adjacent vessel wall during each refueling outage until the final PT is performed.
c) The Authoritf proposes not to perform the flow capacitf test required bf NUREG-0619 wnen capping the CRDRL vessel penetration during the current outage, f
Justification:
- 1. The flow requirement for a adR/4 (21d" vessel) is ontf 135 gpm, whicn is well within the capacitf of the CRD System (2 pump operation). Two CRD pump capability was demonstrated during the isolated CRD System testing performed in 1977.
- 2. Performance of this test to properly demonstrate the desired flowrate, requires haat-up of the vessel to normal pressure and temperature followed bf a reactor scram. Intentional scrams of this type are undesirable and the pumping of large quantities of water past the CRD mechanism seals can significantly contribute to seal failure.
- 3. The Authority is aware that this requirement has not been imposed on other plants who have proven an..
altsentte shutdown ctpsbility for plant fires. Tha Authoritf has significantif upgradsd its fira prott:ction. separation, mitigation and alternate shutdown capabilites 'in recent years.. A complete alternate shutdown capabilit/ (Appendix R) will be fullf operational following the 1985 outage.
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ENCLOSURE 3 TO JPN 6B CONTROL ROD DRIVE RETORN LINE OPERATING HISTORY References 1.
General Electric Service Information Letter (SIL) 200, Supplement 2 - Control Rod Drive Hydraulic return line Modification.
2.
G. E. letter 0-EP-7-143 f rom G. E.
(T. Collopf) to PASNY (J.D. Leonard) dated August 4, 1977.
3.
Authority letter, G.T. Berrf to R.W. Reid dated Juif 7, 1977.
4 G.E. letter G-EPL-7-199 from G. E.
(J. Collopf) to PASNY (J.D. Leonard ) dated November 9,1977.
5.
Authority letter P.J. Earlf to J. A.
- Ippolito, dated March 14, 1979 (JPN-79-14).
History 1975-1977 JAF operates for Cycles 1 and 2 with CRDRL in service to the reactor vessel.
June, 1977 Based on the recommendation of General Electric in Reference (1), the Control Rod Drive return line (CRDRL) was permanentlf isolated for reactor power operations.
June, 1977 CRD System performance test with the CRDRL isolated was sucesssfullf performed and test data evaluated bf General Electric (Ref erence 2).
These test results dere forwarded to NRC via Ref erence 3.
June, 1977 Vessel directly below CRDRL nozzle was ultrasonically inspected f rom outside witn no, indications (Reference 4).
Sept. 1977-Control rod scram time testing frequencf was Aug. 1978 increased to ensure proper scram function with CRDRL isolated during Cycle 3.
In November,1978, the CRDRL nozzle blend radius and adjacent vessel wall area was liquid penetrant inspected with no indications. The thermal sleeve was not removed for this inspection. The results of this inspection were forwarded to the NRC via Reference 5..
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6 Dec. 1978-CRD System has been operated with the CRDRL 4
May 1983 isolated. System performance and corrective maintenance requiremants have been normil. Scram times have been normal. Augmented ISI per NORE3-0313 was performed during outage periods on the stainless steel sections of the CRDRL.
1 June, 1983 An in-vessel visual inspection (with less than 1 mil resolution) was performed on the CRDRL nozzle and on adjacent vessel wall with no indications.
This inspection provided a view into the nozzle bore as far as the thermal sleeve mounting pads.
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RSCLOfGRE 4 TO JPN 64 NUREG-0619 IMPLEMENIAfION SCHEDULE FOR CONfROL ROD DRIVE REfGRN LINE NOZZLE The following is a summary of the implementation plans and dates for NUREG-0613 actions relative to the CRDFL, which were not addressed elsewhere in the letter.
NUREG-0619 Section 8 Implementation 8.1.(1)
Tnis item is partially complete.
It will be completed during the Cycle 12 refueling outage.
8.l.(2)
Interim valved-out operation will end with the capping of the vessel penetration during the current outage. CRD performance testing was completed in 1977. Augmented ISI of the line will terminate with capping of the vessel penetration.
8.1.(3)
The Authority is reviewing les previous commitment to 8.1.(4) reroute the CRDRL with continuous flow.
If the advantages of rerouting are not confirmed, the Authority will perf orm the modifications of 4(a' ) -
(c') during the 1985 refueling outage.
If the CRDRL is rerouted, this modification will be perfocned during the 1985 outage.
8.1.(5)
If the line is rerouted with continuous flow during the 1985 outage, the required pressure control station will be installed.
8.1.(6)
Tnis item is not applicable to the FitzPatrick plant.
9.1.(7)
Operating procedures for achieving CRD flow to the reactor vessel, equal to or greater than the required bolloff rate, will be implemented prior to startup for Cycle 6 (August 1983).
8.2.(1)
Final dye penetrant testing will be performed in accordance with the schedule provided in Enclosure (2) to this letter. Augmented ISI of the CRDRL will not be required following the present outage because all stainless steel portions of the CRDRL will be eliminated. )
EsICIOSURE 4
-,s 8.2.(2)
Final d/e penetrant testing will be performed in accordance with the schedule provided in Enclosure 2.
If the CRDRL is rerouted (valved open) during the 1935 outage, tne pressure control station will be installed j
and the return flow capacity demonstrated. Welded branch connections will be inspected during each refueling outage.
8.2.(3)(a,b)
Refer to d.2. (2) and d.2.(1) 8.2.(3)(c) and If carbon steel is maintained in anf part of the CRD 8.2.(4) exhaust water s/ stem after the modifications are installed, plant maintenance procedures will be expanded to require flushing the normal drive-movement I
exhaust water header and cleaning the filters in the insert-exnaust lines.
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