ML17158A584
| ML17158A584 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 03/23/1995 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17158A583 | List: |
| References | |
| GL-94-03, GL-94-3, NUDOCS 9503270189 | |
| Download: ML17158A584 (6) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055&4001 EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION
RESPONSE
TO GENERIC LETTER GL 94-03 SUS UEHANNA STEAM ELECTRIC STATION PENNSYLVANIA POWER AND LIGHT COMPANY PP&L DOCKET NOS.
50-387 AND 50-'388
- 1. 0 INTRODUCTION The core shroud in a Boiling Water Reactor (BWR) is a stainless steel cylinder component within the reactor pressure vessel (RPV) that surrounds the reactor core.
The core shroud serves as a partition between feedwater in the reactor vessel's downcomer annulus region and the cooling water flowing up through the reactor core.
In addition, the core shroud provides a refloodable volume for safe shutdown cooling and laterally supports the fuel assemblies to maintain control rod insertion geometry during operational transients and accidents.
In 1990, crack indications were observed at core shroud welds located in the beltline region of an overseas BWR.
This reactor had completed approximately 190 months of power operation before discovery of the cracks.
As a result of this discovery, General Electric Company (GE), the reactor vendor, issued Rapid Information Communication Services Information Letter (RICSIL) 054, "Core Support Shroud Cracking Indications,"
on October 3, 1990, to all owners of GE BWRs.
The RICSIL summarized the cracking found in the overseas reactor and recommended that at the next refueling outage, plants with high-carbon-type 304 stainless steel shrouds perform a visual examination of the accessible areas of the seam welds and associated heat-affected zone (HAZ) on the inside and outside surfaces of the shroud.
Subsequently, a number of domestic BWR licensees performed visual examinations of their core shrouds in accordance with the recommendations in GE RICSIL 054 or in GE Services Information Letter (SIL) 572, which was issued in late 1993 to incorporate domestic inspection experience.
Of the inspections performed to date, significant cracking was reported at several plants.
The combined industry experience from these plants indicates that both axial and circumferential cracking can occur in the core shrouds of GE designed BWRs'.
On July 25, 1994, the NRC issued Generic Letter (GL) 94-03 to all BWR licensees (with the exception of Big Rock Point) to address the potential for cracking in their core shrouds.
GL 94-03 requested BWR licensees to take the following actions with respect to their core shrouds:
inspect their core shrouds no later than the next scheduled refueling outage; perform a safety analysis supporting continued operation of the facility until the inspections are conducted; l
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develop an inspection plan which addresses inspections of all shroud
- welds, and which delineates the examination methods to be used for the inspections of the shroud, taking into consideration the best industry technology and inspection experience to date on the subject;
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develop plans for evaluation and/or repair of the core shroud; and
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work closely with the BWROG on coordination of inspection, evaluations, and repair options for all BWR internals susceptible to intergranular stress corrosion cracking.
Pennsylvania Power
& Light Company, the licensee for Susquehanna Units 1 and 2
responded to GL 94-03 on August 24, 1994.
Part of the licensee's response included a schedule for inspection of the core shroud for the units and safety assessment supporting continued operation of each facility.
2.0 Justificatio fo Continued 0 er tion d Schedu e for Ins ection Re ai The licensee plans to conduct an inspection or repair as appropriate of the core shroud at Susquehanna Unit 1 during the spring 1995 outage and at Unit 2:..
during the fall 1995 outage.
- 2. 1 Susce tibilit of Sus uehanna Units 1
and 2 Core Shrouds to GSCC The core shroud cracks which are the subject of GL 94-03, result from intergranular stress corrosion cracking (IGSCC) which is most often associated with sensitized material near the component welds.
IGSCC is a time-dependent phenomenon requiring a susceptible
- material, a corrosive environment, and a
tensile stress within the material.
Industry experience has shown that austenitic stainless steels with low carbon content are less susceptible to IGSCC than stainless steels with higher carbon content.
The formation of carbides at the grain boundaries upon moderate heating (sensitization) is hindered for type 304 stainless steels with carbon contents below 0.03X.
BWR core shrouds are constructed from either type 304 or 304L stainless steel.
The slightly lower carbon content of type 304L
(<0.035X) makes it less prone to develop IGSCC.
Currently available inspection data indicate that shrouds fabricated with forged ring segments are more resistant to IGSCC than rings constructed from welded plate sections.
The current understanding for this difference is related to the surface condition resulting from the two shroud fabrication processes.
Welded shroud rings are constructed by welding together arcs machined from rolled plate.
This process exposes the short transverse direction in the material to the reactor coolant.
Elongated grains and stringers in the material exposed to the reactor coolant environment are believed to accelerate the initiation of IGSCC.
Water chemistry also plays an important role in regard to IGSCC susceptibility.
Industry experience has shown that plants which have operated with a history of hi'gher reactor coolant conductivity have been more susceptible to IGSCC than plants which have operated with lower conductivities'.
Furthermore, industry experience has shown that reactor coolant systems (RCS) which have operated at highly positive, electro-chemical potentials (ECPs) have been mope susceptible to IGSCC than RCSs that have operated at more negative ECPs The industry has made a considerable effort to improve water chemistry at nuclear facilities over the past 10 years.
Industry initiatives have included the introduction of hydrogen water chemistry as a means of lowering ECPs (i.e., making the ECPs more negative) in the RCS.
The effectiveness of hydrogen water chemistry in reducing the susceptibility of core shrouds to IGSCC initiation has not been fully evaluated; however, its effectiveness in reducing IGSCC in the recirculation system piping has been demonstrated.
Welding process can introduce high residual stresses in the material at the weld joint.
The high stresses result from thermal contraction of the weld metal during cooling.
A higher residual tensile weld stress will increase the material's susceptibility to IGSCC.
Although weld stresses are not easily quantified, previous investigation into weld stresses indicated that the tensile stress on the weld surface may be as high as the yield stress of the material.
The stress decreases to compressive levels in the center of welded section.
The licensee has submitted the following information on the fabrication and operational histories of the Susquehanna core shrouds.
The following is a
discussion of this information.
The materials used to fabricate the core shrouds are as follows:
Material specification American Society for Testing and Materials (ASTM) A-240 Type 304L stainless steel plate.
Carbon content varies from 0.014 to 0.026, depending on the heat
- number, and
'Conductivity is a measure of the anionic and cationic content of liquids.
As a reference, the conductivity of pure water is -0.05 ys/cm.
Reactor coolants with conductivities below 0.20 ys/cm are considered to be relatively ion free; reactor coolants with conductivities above 0.30 ys/cm are considered to have a relatively high ion content.
The electrochemical potential (ECO) is a measure of material's susceptibility to corrosion.
In the absence of an externally applied current, and therefore, for reactor internals in the RCS, the electrochemical potential is equal to the open circuit potential of the material.
Industry experience has shown that crack growth rates in reactor internals are low when the ECP ~
0.230 volts.
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Filler material specification - ASTH Type E-308L and ER-308L.
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The core shroud is fabricated from rolled steel plates which are welded together.
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The weld residual stress levels are considered to be high, based on weld shrinkage stress estimates.
The licensee's core shrouds are constructed from type 304L stainless steel with carbon content less than 0.026X.
The reduced carbon content decreases the potential for initiation of IGSCC in the core shroud in comparison with shrouds fabricated from type 304 stainless steel.
2.2 0 eration 1 Histor A longer time of operation increases the potential for initiation and growth of IGSCC.
Susquehanna Units I and 2 have operated for 8.2 years and 6.5 years respectively.
Other BWRs with shrouds more susceptible to IGSCC which have accumulated considerably longer operational times have inspected their core shrouds.
In none of these cases has a licensee identified cracks that would; diminish the structural margin of the core shroud to an unacceptable level.
The operating times for the Susquehanna units are significantly less than those of other BWRs which have been inspected and found to have significant cracking.
Nevertheless, the possibility exists that the Susquehanna core shrouds contain some degree of cracking.
If any cracking exists, it is not likely to be significant.
The initial 5-year average reactor coolant conductivity for Susquehanna Unit I was 0.205 ps/cm and that of Unit 2 was 0. 199 ys/cm.
The conductivities of both units are well below the average for the entire population of U.S.'WRs (where the conductivities range from -0. 123 ps/cm to -0.717 ps/cm and average
-0.340 ps/cm).
Thus, the low initial reactor coolant conductivity makes it unlikely that the shrouds will contain significant cracks that would affect the structural integrity to any great extent.
2.3 8 s's for Cont'nued 0 erat'o The Unit I shroud was 'inspected visually for weld cracking during the fall of 1993.
Three;vertical and three horizontal welds were selected in the high flux.region-of the core and from previous experience of cracking found in other shrouds.
There were no indications found on any of the inspected welds.
The Unit 2 shroud,
- however, has not been inspected since it has only 6.5 years of operation.
In response to GL 94-03, the licensee submitted a justification for continued operation (JCO) based on the following rationale.
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For both units of Susquehanna, the core shrouds were constructed with low carbon stainless steel material which has a higher resistance to initiation of IGSCC.
Both units of Susquehanna have operated at a low reactor coolant conductivity over the life of the units.
Both units have accumulated low operating time.
The Unit I has operated for 8.2 years which is slightly beyond the minimum number of years of operation to begin inspection of core shroud in accordance with BWROG recommendati.on and the Unit 2 has operated for 6.5 years which does not warrant an inspection of the core shroud.
Taking into account the above factors and qualitatively comparing these factors with those of other BWRs with shroud inspection data, the staff concludes that the licensee's core shrouds have a moderate susceptibility to'GSCC.
This is consistent with the recommendations of the BWRVIP which classify the Susquehanna units as Category B in terms of susceptibility to IGSCC.
The staff has reviewed the inspection results for other BWRs with core shrouds more susceptible to IGSCC and notes that there has been no instance where a
360'hrough-wall crack existed in any plant that was inspected.
- Further, no.
BWR has exhibited any symptoms (power to flow mismatch) caused by leakage through a 360'hrough-wall crack.
'All analyses performed by licensees for higher susceptibility plants show that even if cracking did exist, ligaments would exist to assure structural integrity.
In addition, there is a low probability for an initiating event which could potentially challenge the integrity of the core shroud.
Also, for the uninspected plants with the highest susceptibility, there is only a short duration of operation until the licensee implements necessary inspections or repairs.
3.0 CONCLUSION
Based on th'e evaluation provided in Section 2, the staff finds that the schedule for the inspection or a preemptive repair of the core.shrouds at Susquehanna is acceptable.
The staff concludes that the units can continue to be safely operated until their next refueling outages.
4.0 OUTS AND G IS S
UTURE ACTIONS In accordance with the reporting requirements of GL 94-03, the licensee shall submit to the
- NRC, no later than 3 months prior to performing the core shroud inspection, both the inspection plan and the licensee's plan for evaluating and/or repairing of the shroud based on the inspection results.
In addition, results should be provided to the NRC within 30 days from the completion of the inspection.
In accordance with this specification in the GL, PP&L has already submitted to the Commission its scope and schedule for the core shroud inspection, for Unit I dated December 19, 1994, which is currently under staff review.
The inspection plan and schedule for Unit 2 is yet to be provided to the Commission.
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~ q During these inspections, if the licensee identifies any core shroud cracking requiring an analysis per the ASIDE Code, details of such evaluations must also be submitted to the NRC for review It should be noted that the industry is currently encountering difficulty in performing comprehensive inspections of lower shroud weld due to NDE equipment accessibility prob]ems.
The staff urges licensees to work with various vendors and the EPRI NDE Center in order to develop improved reliable tooling for inspections of shroud welds which are highly obstructed.
Should improved inspection techniques become available, the staff recommendation is for licensees to reinspect the lower shroud welds at the earliest opportunity The licensee indicated in their response that they may adjust their core shroud inspection schedule and scope per guidance from the BWRVIP ~
At
- present, the NRC has not approved the inspection guidelines proposed by the BWRVIP.
Considerable differences remain with regard to the recommended scope of core shroud inspections'he staff cautions the licensee against modifying their plans according to BWRVIP recommendations which have not undergone review and approval by the NRC.
The staff's current position with regard to the scope of inspection is a recommendation for the inspection of lOOX of the accessible core shroud welds.
Should the licensee opt to install a preemptive repair in lieu of performing a comprehensive core shroud inspection, the only required inspection is that mandated in the staff appr oval of the repair option.
Principal Contributor:
Pat Patnaik Date:
I'~arch 23, 1995