ML19282B616
| ML19282B616 | |
| Person / Time | |
|---|---|
| Issue date: | 02/28/1979 |
| From: | Stello V Office of Nuclear Reactor Regulation |
| To: | Case E Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19282B613 | List: |
| References | |
| REF-GTECI-A-10, REF-GTECI-RV, TASK-A-10, TASK-OR NUDOCS 7903160004 | |
| Download: ML19282B616 (10) | |
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Generic Technical Activity A-10 MEMORANDU" FOR: Edson G. Case, Chairman Task Action Steering Comittee FRO.":
Victor Stello, Jr., Director Division of Operating Reactors SUNECT:
REVISED TASK ACTION PLAN FOR GENEP.IC TECHNICAL ACTIVITY A-10. "BWR N0ZZLE CRACKING
Attached for your use is the approved revised Task Action Plan (TAP) for Generic Technical Activity A-10.
Activity A-10 is an " Unresolved Safety Issue" and revision of the TAP was required by Mr. Denton's directive dated January 23, 1979.
W8fnal Signed b' k Nictor Ste;;,
Victor Stello,s-
.ctor Division of Opcc.
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19 Task A-10 BWR N0ZZLE CRACKING Lead NRR Organization:
Division of Operating Reactors (DOR)
Lead Supervisor:
Darrell G, Eisenhut, Deputy Director (DOR)
Task Manager:
Dick Snaider, 00R Applicability:
Boiling Water Reactors Projected Completion Date:
October 31, 1979
s.
Task A-10 Rev. No, 2 January 1979 1.
DESCRIPTION OF PROBLEM A.
BWR Feedwater Nozzle Cracking Of the 23 operating BWRs with feedwater nozzle /sparger systems (normally 4 nozzles /spargers per BWR, nominal nozzle diameter being10"-12"),21 have been inspected to date (1/25/79) result-2 ing in the discovery of blend radius or bore cracking in all but three vessels. Although most cracks have been in the range of 1/2" to 3/4" total depth (including cladding), one crack pene-trated the cladding into the base metal for a total depth of approximately 1.50 inches. The initiation of cracking is due to high cycle fatigure caused by fluctuations in water temper-ature within the vessel in the sparger-nozzle region during periods of low feedwater temperature when the flow may be unsteady and intennittent. 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 feedwater 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 removal of cracks by grinding has caused enough radiation exposure to personnel to be deemed unacceptable as a long-term solution.
B.
Control Rod Drive Hydraulic Return Line Nozzle Cracking (CRDRL Nozzle)
Each of the 22 applicable BWRs has one CRDRL nozzle of 3"-4" diameter, which is normally located approximately four feet 2
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 feedwater nozzles). Thermal fatigue cracks have been found by dye penetrant (PT) inspection of the CRDRL nozzle and the area irrediately beneath the nozzle at 12 units inspected to date (1/25/79). These cracks resemble those found 2
in the BWR feedwater nozzles, and the cause of cracking appears to be thermal fatigue. All but 2 of the operating domestic BWRs A-10/1
L Task A-10 Rev No. 2 January 1979 have some sort of thermal sleeve (there are several designs) in the CRDRL nozzle, but because of the limited number of inspections of nozzles with sleeves, the efficacy of the sleeves is not known.
To date, the principal activity of licensees has been to re vute or temporarily valve out the CRDRL.
Although both accomplish the intended purpose of shutting off cold water flow to the nozzle, General Electric Company (GE) has further recoseended that the CRD system be operated in an isolated mode. GE recomends against retention of the present CRDRL, even valved out, because of the potential for stress corrosion in the stagnant line. GE also recomends against operation with a rerouted CRDRL open to the reactor vessel. The recomendation to isolate the rerouted line was made on the basis that return to the vessel is unneces-sary for proper CRD system operation and thatCRD 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 questons 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 components of the CRD hydraulic system. GE has begun an evalua-tion of component performance of affected portions of the CRD hydraulic system and has comenced investigation of possible system modifications. The staff aust 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 system by valving out or re-routing. Additionally, to increase assurance of safety for continued operation, the staff is recommending inspection of the CRDRL nozzle blend radius and bore at eacn BWR during its next scheduled refueling outage. As in the case of feedwater 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 2
of crack growth and available techniques and has been issued as NUREG-0312 in July 1977 A-10/2
4 la Task A-10 Rev. No. 2 January 1979 B.
DOR'and DSS Follow Advancements in the Following Areas (1) Development and testing of effective feedwater nozzle thermal sleeves and spargers to protect the nozzle bore and blend radius from thermal cycling and thus minimize or remove the source of crack initiation.
GE has com-pleted such development and tGting and has written a final detailed topical report after having met with the 2
staff to discuss the results of testing. The supplement to this report, addressing additional NRC concerns, is being written now.
(2) DSS will follow the Brookhaven National Laboratory (BNL)
Structural Analysis Group review of the testing involved in the topical report referenced above. This BNL review has been completed and comments have been presented to 2
GE for resolution and inclusion in the report supplement also mentioned above. However, 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 installation of the new GE design on two operating reactors and has approved the use of a similar modification on three l
additonal plants, and has deterTnined that operation of
'2 these plants 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) 00R and DSS will follow the life-cycle testing of certain CRD system valves. GE has performed such testing to deter-mine if long-time reverse flow will lead to valve degrada.
2 tion. A report is being prepared. GE also is CRD system modifications on " requisition" (new) pursuing various facilities.
These modifications, which will eliminate valve reverse flow, require no CRD return line to the vessel. 00R and DSS will review the proposed modifications, which GE may also offer as 2
" suggested" modifications to the owners to operating plants.
(4) Development of viable ultrasonic test (UT) techniques by the nuclear industry to allow reliable and consistent early determination 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 D0R A-10/3
i Task A-10 Rev. No. 2 January 1979 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 cer-tain regions of the nozzle inner radius and bore are inaccessible to surface examination. The staff now recog-nizes that completion of this UT development may be extended oeyond the length of this generic program.
However, this will not hirder resolution of the major issue (crack initiation and growth) and will result in 2
at least a temporarily more conservative stance on inservice inspections by UT until the issue is resolved satisfactorily.
(5) Development of various feedwater system and CR0 system modifications as part of the 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 (4) above.
3.
BASIS FOR CONTINUED PLANT OPERATION AND LICENSING PENDING COMPLETION OF TASK As indicated in Section 2.0 the staff anticipates that this task will result in long term solutions that will provide:
(1) assurance that a conservative margin of safety against vessel failure due to nozzle cracks is maintained at operating facilities, (2) acre stringent licensing requirements concerning selection of saterials and design for nozzles, thermal sleeves, and spargers; (3) more stringent inser-vice inspection and repair criteria; (4) modification of physical systems and/or operating procedures to minimize the occurrence of crack initiation and propagation; and, (5) reliable inservice inspection techniques for detection of nozzle flaws from positions exterior to the reactor vessel.
With respect to feedwater nozzle cracking, specific long term correc-tive seasures will include system and operational changes to reduce the feedwater to reactor water temperature differential during low power operation, an improved thermal sleeve-sparger design to reduce bypass flow which exposes the nozzle surface to fluctuating water temperatures, and removal of clad from the nozzle surface which is believed to provide a surface more resistant to fatigue cracking.
j Implementing some combination of these measures after plants are already under construction or are operating is feasible, e.g., several utilities with operating reactors have already implemented clad removal and the first new thermal sleeve-sparger design has been installed in an operating plant.
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j, Task A-10 Rev. No. 2 January 1979 With respect to control rod drive return line nozzle cracking, specific long term corrective measures will include system modifications that assure proper control rod drive system performance with the return line isolated (if one is installed by design) or eliminated by design.
Control rod drive return line isolation has been implemented at several operating facilities as an interim corrective measure.
Studies are currently underway to determine the acceptability of long tenn operation in this manner.
If these studies (which are scheduled for completion in early 1979) demonstrate no degradation of affected components, no further action in this regard will be necessary for plants so modified.
During the time period required to develop the long term solutions under this task, interim seasures have been taken. Specifically, the staff is requiring inservice inspection using liquid penetrant examina-tions at operating reactors in accordance with the procedures and acceptance criteria set forth in detail in NUREG-0312, Interim Technical Report on BWR Feedwater and Control Rod Drive Re u rn Line Nozzle Cracking, July 1977.
Licensees are also utilizing ultrasonic inspection techniquesin an effort to develop effective techniques that will allow early detection of subsurface flaws.
Enhancement of ultrasonic testing techniques will substantially reduce personnel exposures.
The scheduling and extent of inspection is based upon conservative estimates of crack growth from fracture mechanics analyses assuming undetected flaws. Scheduling is thus dependent upon the reactor's record of past repair (grindouts, clad removal, etc.), operating history (number of startup/ shutdown cycles since last dye penetrant inspection), and licensee actions to minimize crack initiation by procedural or mechanical change.
Preservice inspections and an inservice inspection program are also required of applicants prior to the issuance of an operating license.
The staff has been actively involved in reviewing and approving the results of nozzle inspections and remedial actions proposed by licen-sees to assure continued safe operation. To date the extent of nozzle cracking at operating plants has been limited to depths which can be removed by grinding without exceeding ASME code limits for nozzle reinforcement.
In addition the staff has suggested that measures be taken at affected operating plants and by applicants for plants in the operating license review stage prior to operation, to minimize the occurrence of condi-tions conducive to crack initiation and growth. These measures include monitoring feedwater temperatures and flow, minimizing rapid changes in feedwater flow and temperature, minimizing the duration of cold feedwater injection, avoiding inadvertent or unnecessary HPCI injection, avoiding the unnecessary introduction of cold water from A-10/5
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Task A-10 Rev. No. 2 January 1979 the reactor water cleanup system, and eliminating flow through the control rod drive return line (after assuring proper system operation in an isolated nede).
Although cracking of the pressure vessel nozzles is important to safety, NRC staff analyses indicate that cracking that has penetrated the vessel cladding will grow at a slow enough rate such that the cracking does not pose a critical safety concern today that warrants isrnediate action.
Rather, the staff believes that sufficient time is available, due to the conservative design of the reactor pressure vessel, to permit continued operation of the affected facilities while studies on these events continued on schedule.
Based on the interim measures being taken at operating facilities and being required of applicants for an operating license prior to the issuance of the operating license and the design sargins available in the reactor pressure vessel, we have concluded that operation of such facilities does not preset an undue risk to the health and safety of the public.
For construction permit applications there is reasonable assurance that a variety of long term solutions will be available from this task and from the generic efforts being conducted by the General Electric Company, long before these plants are ready to begin opera-tion..Even if this were not the case additional time would be avail-able since operation could be permitted for a number of years based on inservice inspection and repair procedures using criteria similar to those currently being required.
4.
NRR TECHNICAL ORGANIZATIONS 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. Will manage UT and fracture mechanics consultants as listed in Section 5 below. Issue final MUREG documents.
Manpower Estimates:
8 man-years FY 79 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. Will assist Reactor Systems Branch, DSS, in the 2
development of CRDRL retention / removal criteria.
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Task A-10 Rev. No. 2 January 1979 Manpower Estimates: 0.2 man-year FY 1979, 2
C.
Mechanical Engineering Branch, Division of Systems Safety.
Will work with D0R on development 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 infor-mation leading to GE topical report.
Will review information related to CRD system modifications.
Manpower Estimates:
0.3 man-year FY 1979.
l2 D.
Materials Engineering Branch, Division of Systems Safety.
Will assist DSS-MEB as necessary, in the development of criteria.
Coordinate with D0R on resolution of UT issue.
2 Manpower Estimates: 0.2 man-year FY 1979.
E.
Task Manager, Division of Operating Reactors. Has overall responsibility for coordination of D0R and DSS technical tasks and for the development and issuance of criteria documents.
2 Manpower Estimates: 0.3 man-year FY 1979.
F.
Reactor Safety Branch, Division of Operating Reactors. Will assist Task Manager and Plant Systems Branch in review of 2
CRDL removal issues, especially with regard to vessel makeup flow capability.
Manpower Estimates:
.1 man-year FY 1979 G.
Reactor Systems Branch, Division of Systems Safety. Will develop criteria concerning the removal of the CRDRL of applicable CP/0L facilities.
Hanpower estimate:
.1 man-year FY 1979.
2 5.
TECHNICAL ASSISTANCE Contractor Amount Procram Objectives FY 1978 FY 1979 A. Washington
$5K
$20K Perform fracture analyses University -
of feedwater nozzle cracks Paul Paris detected in operating (Managed by DOR) reactors. This is necessary for generic crack growth calculations.
A-10/7
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Task A-10 Rey, No, 2 January 1979 Contractor Amount Program Objectives FY 1978 FY 1979 D. Brookhaven National
$25K
$20K Perform indepth review of Laboratory GE test and analytical 2
(Managed by DSS) information to assure thermal sleeve /sparger design is viable as a long term solution.
6 INTERACTIONS WITH OUTSIDE ORGANIZATIONS A.
General Electric Company The NRC staff has followed all GE generic testing and develop-mental work, especially those tests designed to determine the cause of cracking and those developments related to UT enhance-ment. This coordination will continue.
B.
Electric Power Research Institue The NRC staff will follow closely EPRI UT optimization development work for the complex nozzle geometry. This work has other generic implications (see Task No. A-14).
C.
Indivicual 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 2
the interim position paper. Applicants for BWR OLs will also be involved in similar interaction with DSS.
7 ASSISTANCE REQUIREMENTS FROM OTHER NRR OFFICES Office of Nuclear Regulatory Research (RES). RES is responsible for the Heavy Section Steel Technology (HSST) program.
Information obtained from this program will be useful in the development of generic fracture analysis methods for a flaw at a geometric discontinuity.
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-10/8