ML19263F259
| ML19263F259 | |
| Person / Time | |
|---|---|
| Issue date: | 05/02/1978 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19263F257 | List: |
| References | |
| REF-GTECI-A-10, REF-GTECI-RV, TASK-A-10, TASK-OR NUDOCS 8001110332 | |
| Download: ML19263F259 (7) | |
Text
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Proposed Revision 1 CATEGORY A TECHNICAL ACTIVITY NO. A-10
Title:
BWR Feedwater Nozzle Cracking (Including Non-Destructive Examination Techniques for Inservice Inspection)/BWR Control Rod Drive Return Line Nozzle Cracking Lead Responsibility: Division of Operating Reactors Lead Assistant Director:
Darrell G. Eisenhut, Assistant Director for Systems & Projects, D0R Task Manager: Dick Snaider, DDR 1.
Problems Description A.
BWR Feedwater Nozzle Cracking Of the 22 operating BWR's with feedwater nozzle /sparger systems (normally 4 nozzles /spargers per BWR, nominal nozzle dianeter being 10" - 12"), 20 have been inspected to date (4/1/78), resulting in the discovery of blend radius and/or bore cracking in each vessel.
Al-though most cracks have been in the range of 1/2" to 3/4" total depth (including cladding), one crack penetrated the cladding into the base metal for a total depth of approximately 1.50 inches. The initiation of cracking is due to high cycle fatigue caused by fluctuations in water temperature within the vessel in the sparger-nozzle region during periods of low feedwater temperature when the flow may be unsteady and intermittent.
Once initiated, the cracks are driven deeper by the larger pressure and thermal cycles associated with startup and shutdown.
Fracture analyses indicate that the cracks found to date in the feedaater nozzles constitute a potential safety problem because the observed rate of crack growth with time in service is such that the margin of safety against fracture will be reduced below acceptable values unless the cracks are detected and ground out every few years.
Obviously, repair by grindout can be repeated only a few times before ASME Code limits for nozzle reinforcement are exceeded.
- However, repair by welding buildup of the grindout has not been demonstrated to be acceptable.
In addition, the inspection and renoval of cracks by grinding has caused enough radiation exposure to personnel to be deemed unacceptable as a long-term solution.
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. B.
Control Rod Drive Hydraulic Return Line Nozzle Crackino
^
TOiDRL Nozzle)
Each of the applicable BWRs has one CRDRL nozzle of 3" - 4" diameter, which is normally located approximately four feet below the level of the feedwater nozzles (In the Oyster Creek and Nine Mile Point vessels, the CRDRL nozzle is located at the same level as the feed-water nozzles). Thermal fatigue cracks have been found by dye penetrant (PT) inspection of CRDRL nozzles at 9 of the 18 domestic units inspected to date (4/1/78).
These cracks resemble those found in the BWR feedwater nozzles, and the cause of cracking appears to be thermal fatigue.
All but 5 of the operating domestic BWRs have some sort of thermal sleeve (there are several designs) in the CRDRL nozzle, but because of the number of cracks found during inspections of nozzles with sleeves, the efficacy of the sleeves is in doubt.
To date, the principal activity of licensees has been to re-route or temporarily valve out the CRDRL.
Although both accomplish the in-tended purpose of shutting off cold water flow to the nozzle, General Electric Company (GE) has further reconmended, in GE Service Information Letter (SIL) No. 200, Supplement No. 2 (November 18, 1977),
that the CRD system oe operated in an isolated mode. GE reconmends against retention of the present CRDRL, even valved out, because of the potential for stress corrosion in the stagnant line.
GE also recommends against operation with a rerouted CRDRL open to the reactor vessel. The reconmendation to isolate the rerouted line was made on the basis that return to the vessel is unnecessary for proper CRD system operation and that CRD makeup capability to the vessel will be maintained even when the return line is eliminated entirely.
The staff still considers the matter of CRDRL isolation to be an unresolved issue because of questions regarding the amount of CRD pump flow which will be available to the vessel, the possible effects of isolation upon various drive parameters, and recently-reported potential long-term deleterious effects on certain conponents of the CRD hydraulic system. GE has begun an evaluation of component per-formance of affected portions of the CRD hydraulic system and has commenced investigation of possible system modifications. The staff must assess these proposals prior to completion of its review of this subject.
In the interim, the staff will review control rod test information from each facility which has modified its present CRD 2213 133
t system by valving out or re-routing.
Additionally, to increase assurance of safety for continued operation, the staff is reconmending inspection of the CRDRL nozzle blend radius and bore at each BWR during its next scheduled refueling outage. As in the case of feed-water nozzles, we are especially concerned, particularly in the case of older units, that a potential safety problem could arise from deep cracks which would necessitate weld repair.
2.
Plan for Problem Resolution Briefly stated, the plan for generic resolution of the BWR feedwater nozzle and CRDRL nozzle cracking problems will involve the following:
A.
Issue interim guidance to operating units.
Such guidance includes criteria for inspection based upon present knowledge of crack growth and available techniques and has been issued as NUREG-0312 in July 1977.
B.
DDR and DSS follow advancements in the following areas:
- 1) Development and testing of effective feedaater nozzle thermal sleeves and spargers to protect the nozzle bore and blend radius from thermal cycling and thus minimize or renove the source of crack initiation. GE has completed such development and testing and is in the process of writing a final detailed topical report after having met with the staff to discuss the results of testing.
- 2) DSS will folicw the Brookhaven National Laboratory (BNL) Structural Analysis Group review of the testing involved in the topical reoort referenced above. This BNL review will verify that the new design provides a satisfactory lono-terr solution to the feedwater nozzle cracking problem. The review will include detailed assessment of GE-empicyed analytical techniques, computer codes, and interpretation of test data.
Such information will not appear in the GE topical report but is considered essential for assurance that the design is efficacious as a long-term solution.
This determination is necessary in particular for plants undergoing CP and OL review, but will also assist D0R in assessing the requirements for inservice inspection of nozzles fitted with the new design. Preliminary review of the GE topical report and discussions with cognizant GE personnel have not produced any information which would make the staff believe the new GE design is not a viable solution, especially since cladding removal is an integral part of nozzle preparation for installing the new sparger/ thermal sleeve.
Therefore, the staff has allowed the 2213 134
1 installation of the new design on one operating reactor (Brown's Ferry Unit No. 2) to date, and has determined that operation of this plant is satisfactory during the period of the BNL review.
This also applies to additional facilities for which the staff may approve modification prior to completion of the BNL task.
- 3) DDR and DSS will follow the life-cycle testing of certain CR0 system valves. GE is performing such testing to determine if long-time reverse flow will lead to valve degradation. GE also is pursuing various CRD system modifications on " requisition" (new) facilities. These modifications, which will eliminate valve reverse flow, require no CRD return line to the vessel.
D0R and DSS will review the proposed modifications, which GE will also offer as " suggested" modifications to the owners of operating plants.
- 4) Development of viable ultrasonic test (UT) techniques by the nuclear industry to allow reliable and consistent early deter-mination of cracking (and credible claims for the absence of cracking) from positions exterior to the reactor vessel.
Such development of UT is important to both DDR and DSS final positions, especially since two operating plants and several plants in OL review have a welded thermal sleeve-to-nozzle safe-end design. The development of UT procedures for these plants is important because certain regions of the nozzle inner radius and bore are inaccessible to surface examination. This portion of the program will be coordinated with Task No. A-14, Flaw Detection.
5)
Development of various feedwater system and CRD system modifica-tions as part of th: generic effort toward problem resolution.
6)
Issuance of Branch Technical Position paper (CP and OL plants) and final NUREG document (operating plants) upon satisfactory completion of subtasks 1) through 5) above.
3.
NRR Technical _0rganizations Involved A.
Engineering Branch, Division of Operating Reactors. Has overall lead responsibility for review of all generic inspection, repair, in-service inspection technique development, weld-repair / annealing study, and modification (such as clad removal and new design thermal sleeves /spargers) efforts. Will gather and disseminate critical information (fluid flows and temperatures) on operating plants.
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k Will manage fracture mechanics consultants as listed in section 4 below.
Issue final NUREG documents.
Manpower estimates:
1.7 manyear FY 1978, 1.7 manyear FY 1979 B.
Plant Systems Branch, Division of Operating Reactors.
Has lead responsibility for review and approval of any proposed generic feedwater or CRD system modifications. Will assist in development of NUREG documents.
Manpower estimates:
.2 manyear FY 1978,.2 manyear FY 1979 C.
Mechanical Engineering Branch, Division of Systems Safety. Will work with DDR on developnent of criteria and will issue BTP for CP/0Ls similar to NUREG guidance issued for operating facilities.
Will manage consultant on review of test and analytical information leading to GE topical report. Will review information related to CRD system modifications.
Manpower estimates:
.8 manyear FY 1978,.8 manyear FY 1979 D.
Materials Engineering Branch, Division of Systems Safety. Will assist DSS-MEB as necessary in the development of criteria.
Manpower estimates:
.3 manyear FY 1978,.3 manyear FY 1979 E.
Task Manager, Division of Operatino Reactors.
Has overall responsi-bility for coordination of DDR and DSS technical tasks and for the development and issuance of criteria documents.
Manpower estimates:
.2 manyear FY 1978,.2 manyear FY 1979 4.
Technical Assistance Requirements Amount Contractor FY 76 FY 79 Program Objectives Washington University
$5K Perform fracture analyses Paul Paris of feedwater nozzle cracks (Managed by DOR) detected in operating reactors. This is necessary for critical crack length calculaticns.
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. Amount Contractor FY 78 FY 79 Frogram Objectives Brookhaven National
$25K S20K Perform indepth review of Laboratory GE test and analytical (Managed by DSS) information to assure thermal sleeve /sparger design is viable as a long-term solution.
5.
Interactions With Outside Organizations A.
General Electric Company The NRC staff has followed all GE generic testing and developmental work, especially those tests designed to determine the cause of cracking and those developments related to UT enhancement.
This coordination will continue.
B.
Electric Power Research Institute The NRC staff will follow closely EPRI UT optimization development work for the conplex nozzle geometry.
This work has other generic implications (see Task No. A-14).
C.
Individual Licensees and Applicants of BWR Facilities Each licensee has already been involved in discussions and written correspondence with the NRC concerning inspections to be performed.
This interaction, as well as discussions on a generic basis, will continue until problem resolution, although the NRC position has been spelled out clearly in the interim position paper. Applicants for BWR OLs will also be involved in similar interaction with DSS.
6.
Assistance Requirenents From Other NRR Offices Office of Nuclear Regulatory Research (RES), RES is responsible for the Heavy Section Steel Technology (HSST) program.
Information ob-tained from this program will be useful in the developnent of generic fracture analysis methods for a flaw at a geometric discontinuity.
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. 7.
Schedule for Problem Resolution The major milestones for the generic feedwater and CRDRL nozzle issues are as follows:
A.
Issue interim NRC guidance to licensees - August 1977.
Complete.
B.
Start review of completed GE testing - June 1978.
C.
DSS approve generic feedwater nozzle design for new plants and issue interim guidance - April 1979.
D.
Resolve CRDRL issue regarding return line isolation, reverse flow through CRD system valves, and proposed CRD system modifications -
June 1979.
E.
With input from Task A-14, resolve UT issue, evaluating techniques for use on complex geometry - August 1979.
F.
Issue final guidance to applicants (Branch Technical Position) and licensees (NUREG Document) - October 1979.
8.
Potential Problems The most serious potential problem facing the NRC staff and licensees at this point is the discovery of a crack large enough to exceed the ASME code criteria for required reinforcement area.
This would result in the need for a vessel repair (other than grinding) which would be an undertaking of potentially large proportions and of safety significance.
A generic contingency plan is presently being outlined by DDR. As scoping of such a contingency plan develops, we will document the plan as Appendix A to this report.
The schedule may be lengthened by extension of UT analysis in the performance of related Task A-14.
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