ML20217A701
| ML20217A701 | |
| Person / Time | |
|---|---|
| Issue date: | 10/20/1997 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-3070, NUDOCS 9803250179 | |
| Download: ML20217A701 (17) | |
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ADVISORY COMMITTEE ON REACTOR SAFEGUARDS i
MINUTES OF THE JOINT MEETING OF THE ACRS SUBCOMMITTEES ON GTERIALS AND METALLURGY AND ON SEVERE ACCIDENTS AUGUST 26-27, 1997 ROCKVILLE, MARYL7LND INTRODUCTION The Advisory Committee on Reactor Safeguards (ACRS) Subcommittees i
I on Materials and Metallurgy and on Severe Accidents held a joint meeting on August 26-27, 1997, in Room T-2 B3, 11545 Rockville Pike, Rockville, Maryland, with representatives of the U.S. Nuclear Regulatory Commission (NRC), the Nuclear Energy Institute (NEI),
and the nuclear industry.
The purpose of this meeting was to review the proposed draf t generic letter (GL), and associated draf t regulatory guide (RG) regarding steam generator (SG) tube integrity, and an interim staff Safety Evaluation Report (SER) on the, "BWR Vessel and Internals Project, BWR Reactor Pressure Shell Weld Inspection Recommendations (BWRVIP-05)," prepared by the BWR Owners Group (BWROG).
The entire meeting was open to the public.
Mr. Amarjit Singh was the cognizant ACRS staff engineer for this meeting.
The meeting was convened at 8:30 a.m. on August 26, 1997, and recessed at 6:00 P.M.
on the same day.
It was reconvened at 8 : 3 0 A.M on August 27, 1997 and adjourned on 10:00 a.m. on the same day.
ATTENDEES ACRS Members /ACRS Consultant R.
Seale, Chairman D.
Powers, Member M.
Fontana, Member W.
Snack, Member T. Kress, Member R.
Cheverton, Consultant Nuclear _En2rgy Institute Boilina Water Reactor Owners Groun C.
Callaway C. Terry I
R. Dyle P. Riccardella Principal NRC Sceakers J.
Strosnider, Office of Nuclear Reactor Regulation (NRR)
C.E. Carpenter, NRR A. Lee, NRR T. Reed, NRR E. Murphy, NRR P. Rush, NRR S. Lee, NRR, J. Donoghue, NRR h/ ['l ~ /
c/5 n
9003250179 971020 '
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- 4. -
. Joint Mat'l & Metallurgy August 26-27, 1997 and Sev. Acc.' Minutes J. Hayes, NRR, S. Malik, Office of Nuclear Regulatory Research (RES),..
No written comments or requests for time to make oral statements were received from members of the public.
A complete list-of meeting attendees is kept in the ACRS Office File and will be made available upon request.
The presentation slides and handouts used during the meeting are attached to the office copy of these minutes.
Chairman's Ooenina Remarks Dr. Robert L.
Seale, Acting Chairman of the Subcommittees on Materials and Metallurgy and on Severe Accidents, convened the meeting at 8:30 a.m.
He stated that the purpose of the meeting was to review the proposed draft GL, and associated draft RG regarding SG tube integrity, and an interim SER on the inspection of welds in pressure vessels for BWRs.
NRC Staff Presentation Introduction and Backaround - Mr. C.E. Carpenter, NRR Mr. Carpenter, Lead Project Manager of NRR for the Boiling Water Reactor Vessels and Internals Project, (BWRVIP-05) presented the background for the NRC staff review of the BWR reactor pressure vessel (RPV) shell' weld inspection recommendations discussed in the BWRVIP-05 proprietary report, "BWR Pressure Vessel Weld Inspection Recommendations."
In August
- 1992, the NRC amended Section 50.55a (g) (6) (ii) (A) of Title 10 of the Code of Federal Reculations (10 CFR 50,55a), which required licensees to perform RPV shell weld examination as specified in the 1989 Edition of the American Society of Mechanical Engineers Boiler and Pressure Vessel (ASME) Code Section XI on an " Expedited" basis and revoked all previously granted reliefs for RPV weld examinations.
By incorporating the 1989 Edition of the ASME Code into the regulations, the staff required that licensees perform volumetric examinations of " essentially 100 percent" of the RPV pressure-retaining shell welds during all inspection intervals.
In September 1995, a technical committee of the BWROG, submitted the report BWRVIP-05, which proposed to reduce the scope of inspection of the BWR pressure vessel welds from essentially 100 percent of all RPV shell welds to 50 percent of axial welds and zero percent of circumferential welds.
Later, the BWROG modified its-proposal to increase the inspection of axial welds to 100 percent - (from 50 percent) and zero percent of circumferential welds.
o-
4 Joint Mat'l & Metallurgy August 26-27, 1997 and Sev. Acc. Minutes On May 30,
- 1997, the Commission issued a Staff Requirements Memorandum (SRM) directing the staff to assess the validity of a sampling or tiered approach to the inspection of the reactor vessels.
The Commission also instructed the staff to perform a comprehensive assessment of the probabilistic analysis of the BWRVIP-05 report.
The staff has completed its preliminary assess.nent of the BWRVIP-05 report and concluded that additional information is required to complete its review.
Discussion of the NRR Indeoendent Assessment of the BWRVIP-05 Egoort - Ms. Andrea Lee, NRR Ms. Lee, NRR, discussed the NRR Independent Assessment of BWRVIP-05 report in the areas of f abrication of BWRs and the statistical assessment of a sampling approach to RPV weld inspections.
Fabrication of BWRs Ms. Lee stated that RPVs were f abricated to very high standards, as evidenced by preservice and inservice inspections to date.
However, dif ferent processes were used to fabricate RPV welds under a number of varying conditions.
Additionally, the BWR RPVs were constructed by several vendors.
The staff reviewed the various fabrication methods associated with RPV fabrication in order to support a statistical assessment of a sampling approach to RPV weld inspections.
In particular, the staff focused on whether it could be assumed, considering the various fabrication methods, that all RPV welds represent one statistical population, or if the population of RPV welds could be divided into well-defined subpopulations.
Ms. Lee also stated that the staff reviewed the following three welding techniques and the differences in the types of welds.
1.
Shielded Metal Arc Welding (SMAW) 2.
Submerged Arc Welding (SAW) 3.
Electroslag Welding (ESW)
A number of variables had to be considered when determining which weld process to use for fabrication of an RPV.
One variable was economic impact. The Shielded Metal Arch Welding technique was the most expensive of the three welding processes.
The Submerged Arc Welding technique was less expensive than the SMAW process, and the Electroslag Welding technique was the least expensive of the three; however, ESW could only be used to fabricate axial welds.
The staff attempted to divide the BWR welds into subpopulations for statistical analysis, but each reactor vessel can have unique characteristics that distinguish it from other reactor vessels even if the vessels were f abricated by the same vendor.
It is difficult to categorize the subpopulations of the reactor vessel welds
r Joint Mat'l & Metallurgy August 26-27, 1997 and Sev. Acc. Minutes because of.many variables involved in the welding process.
The staff performed the comparison of the BWR RPVs which were grouped by vessel fabricator weld process.
In terms of cladding for B&W fabricated vessels, the SMAW, the SAW and the semi-automatic gas metal arc welding (GMAW) processes were used.
For CE and CB&I fabricated vessels, the SMAW process or a combination of the SMAW and the GMAW processes was used for back cladding.
With regard to stress relief processes, the standard temperature and time used for stress relief were 1150' F for a cumulative period of approximately 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br />. The guidelines for the amount of stress relief time are stated in Section III of the ASME Code. The ASME Code also allows an alternate requirement that is lower than the minimum of 1100* F.
Statistical Samolina Ms. Lee presented the statistical analysis of a sampling approach to RPV weld examinations.
Of the total length of axial and circumferential welds in a RPV, approximately 60 percent are circumferential welds, and axial welds make up the remaining 40 percent.
The staff concluded that no meaningful statement about circumferential welds can be made on the basis of only a statistical analysis of the axial welds.
Any such statement must
{
be based on the assumption that both the axial and circumferential j
welds are random samples from some parent population of welds.
If
)
this assumption is made, an inference can be made about the expected number of defective circumferential welds on the basis of j
an inspection of the axial welds.
For the best possible situation
)
in which all axial welds are inspected and no defective welds are j
found, an inference can be made about the expected number of j
defective circumferential' welds on the basis of an inspection of the axial welds.
The staff concluded that no useful inference regarding the condition of the circumferential welds can be made on j
the basis of a 100 percent inspection of the axial welds only.
The j
statistical analysis was based on the assumption that both the j
axial and circumferential welds were random samples from the same parent population of welds.
But in reality, the circumferential welds.were not from the same population because of the different weld processes and other sources of variances.
Therefore, a
statistical sampling approach is not valid.
Limitino Transients - Mr. Sam S. Lee, NRR Mr. Le e,. NRR, discussed the staff's independent assessment in identifying reactor transients and how these limiting transients are accounted for in estimating the challenge or the frequency of challenges to the vessel. The most limiting operational transients
Joint' Mat'l & Metallurgy August 26-27,~1997 and Sev. Acc. Minutes i
i with respect to the vessel are loss of feedwater or single safety relief valve (SRV) blowdown events for normal-and degraded conditions.
Normal operating temperature and pressure for a BWR RPV is 500* F and 1000 : psig.
During the hydrostatic test, RPV temperature and pressure are approximately 150' F to 200* F and_1000 psig and are maintained on the pressure-temperature curve-for that 3
particular vessel.
The most. limiting transients for the cmergency and faulted operating conditions are any transient that enuses.or results in a rapid cooldown, and repid depressurization1of the l
vessel shell welds. These transients are limiting for pre-existing cracks in the vessel shell welds.
In comparison, the BWRVIP-05 report indicated that the water-solid leak test condition, or the I
' hydrostatic test, is limiting for small flaws in RPV shell inner diameter.
The BWRVIP-05 report was limited to design-basis
- accidents.
To provide a broader risk-informed assessment, the l staff performed a sampling review cf 17 years of licensee event reports and event notifications to determine if other events (i.e.,
shutdown events) could be potentially more limiting to the vessel.
1 For the hydrostatic test and the leak test, the staff estimated the vessel challenge frequency to be on the order of 7x10-' per reactor
)
year.
For the reactor water cleanup system the staff estimated the vessel challenge frequency to be 1x10-' per reactor year.
For the actual event that occurred, from years of experience the staff i
estimated the frequency to be 6x10-' per reactor year.
Mr. Lee stated that the staff has generated questions for the BWROG to H
refine these estimates and to verify the assumptions associated with its analyses.
Conditional Failure Probabilities - Mr. Shah N. Malik, RES Mr. Malik, presented the conditional failure probabilities for axial and circumferential welds.
The staff performed the probabilistic fracture mechanics analysis for RPVs fabricated by CE, B&W, and CB&I.
The general analysis framework for these analyses is similar to pressurized thermal shock (PTS) probabilistic analysis that is performed for BWR vessels.
In general, material properties, fluence, and flaws were considered as random variables.
For ~ the CB&I vessels with circumferential
- cracks, no failure was predicted for 10 million simulations performed for the various cases and, as such, the conditional failure probability is less than 1.0x10-7 per reactor year.
The major difference in the treatment of fluence between the BWRVIP-05 report and in staff's analyses is that BWRVIP-05 report considered moderate levels of fluence because the staff used a more comprehensive set of fluence. data in its analyses.
In the flaw
- depth and density distributions, the BWRVIP-05 report used Marshall depth distribution with a density of 30 flaws / meter' and the staff
'used SAFT-UT inspection data on a pressurized-water reactor (PWR) vessel' weld to determine flaw depth distribution and flaw density.
i I-
1 Joint Mat'l & Metallurgy August 26-27, 1997 and Sev. Acc. Minutes Since insufficient failures were reported for the references for circumferential welds, their sensitivity to flaw size and density could not be determined.
j The staff performed sensitivity studies to determine the effect that inservice inspection (ISI) would have on the probability of failure (POF). Two probabilities of detection (PODS) of flaws were evaluated.
They were the POD of flaws resulting from the Program for the Inspection of Steel Components (PISC) studies and the POD of flaws defined by Method C in the VIPER code.
The staff believes that the PISC-based POD should be considered a lower bound for ultrasonic inspection methods that have been qualified as meeting Appendix VIII of ASME Code,Section XI.
Mr. Malik concluded that the fluence has the highest effect on the conditional POF related to other variables, such as fracture toughness and flaw distribution, and the 95 percent confidence-bound values are about a factor of five higher than the best-estimate values.
Industry Presentation Mr. Carl Terry, Vice President, Niagara Mohawk, and Chairman, BWR Vessel Internals Project; Mr. Robin Dyle, Southern Nuclear; and Dr.
Pete Riccardella from Structural Integrity presented the introduction, review of BWRVIP-05 methodology, and comparison of the NRR Independent Assessment against the BWRVIP-05 analysis respectively.
Mr. Terry made the following introductory remarks:
The BWRVIP-05 proposal for BWR RPV shell weM inspection is based on sound and thorough technical deterministic risk evaluation.
The BWRVIP-05 proposed 100-percent baseline inspections for axial welds and zero percent inspections for circumferential welds.
The proposed inspections would result in significant savings to the BWR licensees, with no measurable impact on safety.
Different approaches are used in the BWRVIP-05 report and by the NRC r,taff, but the conclusions are the same.
Circumferential shell welds are substantially less risk-I significant than axial welds.
Inupection of circumferential shell welds has no measurable safety benefit.
Joint Mat'l & Metallurgy August 26-27, 1997 and Sev. Acc. Minutes A low-temperature, overpressure event is extremely unlikely for U.S.
BWRs because of BWR operating conditions and operational controls and procedures.
A very low probability of circumferential weld failure exists even for a low-temperature, overpressure event.
Differences in assumptions in the BWRVIP-05 report and the NRR Independent Assessment of the BWRVIP-05 report included a limiting operational transient and flaw density and flaw size distribution.
Mr. Terry stated that overall there is substantial agreement between the staff and the industry regarding the analysis.
He further stated that the BWRVIP-05 team is finalizing responses to the NRR Independent Assessment of the BWRVIP-05 report and the associated request for additional information.
Mr. Dyle discussed the methodology used in reviewing the BWRVIP-05 report and stated that the methodology was developed by NRC for analysis of.PWR PTS (VISA code), including probabilistic treatment of vessel fracture toughness and radiation embrittlement.
He made the following comments regarding the methodology:
Fabrication defects in the vessel were assumed to exist.
A multiple random variable was used to compute vessel failure probabilities.
Several new features specific to BWR vessel inservice inspection were added to the methodology that is stress corrosion crack initiation in cladding and crack growth in low-alloy steel.
Several specific vessel configurations with wide range of fluence and copper content were considered.
i He stated that it is reasonable to conclude that thase vessels did not enter into service with large surface-breaking flaws.
Dr. Riccardella compared the NRR Independent Assessment with the BWRVIP-05 report.
He stated that NRR used significantly different methods and assumptions that resulted in differences in detailed numerical results, but overall conclusions are the same. The major differences between the NRR and the BWRVIP-05 report follow:
a
Joint Mat'l & Metallurgy August 26-27, 1997 and Sev. Acc. Minutes Limitina Ooerational Transient NRR Assumption:
Cold over pressurization event (1150 psi; 88 *F) with' an event frequency of 6x10-* per reactor year BWRVIP-05 Assumption:
Routine cold pressure test (1050 psi; 150-200 *F) with an event frequency of one per reactor Flaw Density and Size Distribution NRR Assumption:
Pressure Vessel Research Users Facility (PVRUF) distribution 3
and flaw density of 994 flaws /m BWRVIP-05 Assumption:
3 Marshall distribution and flaw density of 30 flaws /m Material Fracture Touchness NRR Assumption:
Mean and standard deviations of properties grouped by vessel manufacturers and fracture toughness sampled as a random variable BWRVIP-05 Assumption:
a Mean standard deviations of properties selected for three real vessels that span the BWR fleet and fracture toughness selected as a deterministic variable from one of four curves Crack Procacation NRR Assumption:
a No crack propagation assumed BWRVIP-05 Assumption:
Stress-corrosion crack growth computed on the basis of random sampling from conservative crack growth test data Dr. Riccardella concluded that with common assumptions, the NRR and the BWRVIP-05 analyses yielded similar results, and both analyses showed that for any set of common input assumptions, i
circumferential welds impose orders of magnitude less risk than i
axial' welds..Therefore, the value of inspecting circumferential welds is minimal and certainly not worth the cost and man-rem 1
F exposure for. the licensees.
Probabilistic fracture mechanics j
analyses by the BWRVIP-05 and NRC demonstrated a very low probability of circumferential weld failure.
Dr. Riccardella stated that by adopting the proposed inspection program, BWR J
1 l
Joint Mat'l & Metallurgy August 26-27, 1997 and Sev. Acc. Minutes licensees will continue to perform a substantial amount of inspections on the risk-significant RPV shell welds.
He also stated that. rapid adoption of this proposed inspection program by the industry is critical.
Draft Generic Letter on Steam Generator Tube Intecrity - Mr. Jack Strosnider, NRR Mr. Strosnider, provided the background and the status of the proposed draft generic letter (GL) and the associated draft regulatory guide.
He stated that during the June 1997 ACRS
- meeting, the staff presented the results of the rule risk assessment and regulatory analysis performed in support of the steam generator rulemaking, and concluded that rulemaking was unnecessary and the objectives of the rule could be accomplished through a GL.
In SECY-97-013, " Steam Generator Rulemaking, " dated May 23, 1997, the staff proposed to pursue the following approach in lieu of SG rulemaking:
Request licensees by GL to propose performance-based Technical Specifications (TS) for inspecting, monitoring, and assessing SG tubes.
Develop a regulatory guide to provide guidance for an i
acceptable performance-based approach for inspecting, i
monitoring, and assessing SG tubes.
Provide an option to implement degradation-specific management i
a consistent with guidance in DG-1061, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Current Licensing Basis."
i Evaluate plant-specific risks as part of individual plant examination (IPE) followup.
The Commission approved the staff's approach in the Staff Requirement Memorandum (SRM) dated June 30, 1997, and directed the staff to amend TSs.
i Operatina Experigagg - Mr. Phil Rush, NRR Mr. Rush, summarized the regulatory framework of inspections in the area of current TS regarding SG tube integrity.
He stated that operating experience indicates that compliance with TS requirements does not ensure operation within the licensing basis and, in some cases, the potential existed for a main steamline break with SG tube rupture.
In general, the licensees focus on satisfying TS surveillance requirements during SG tube examinations, and the existing TSs do not ensure adequate tube integrity throughout the next cycle of operation.
_ ___ _j
i o
-6 Joint Mat'l & Metallurgy August 26-27, 1997 and Sev. Acc. Minutes Reculatorv Bases for the Generic Letter - Mr. Tim Reed, NRR Mr. Reed, presented the regulatory bases for the proposed GL.
He stated that the intent of the regulatory framework was to ensure
.that SG tube integrity is maintained consistent with the General Design Criterion (GDC) 14, " Reactor Coolant Pressure Boundary," of Appendix A to*10 CFR Part 50, which states that the reactor coolant pressure boundary shall be designed, fabricated,
- erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture. The SG surveillance requirements were intended to ensure that SG tube structural integrity is maintained consistent with GDC 14 by
. ensuring that the ASME Code margins are maintained and that SG tube
-leakage integrity is maintained consistent with the plant licensing basis.
In general, the existing TS is prescriptive, out of date, and fails to focus on what is important to maintain the SG tube integrity for the entire operating cycle.
As4 stated in GDC 14, SG tube integrity must be maintained during operation, and " Corrective Action," of Section XVI of Appendix B to 10 CFR Part 50, requires that licensees identify significant conditions adverse to quality and to take action to preclude repetition.
Potential loss of SG tube integ: city during the operating cycle clearly meets the definition of a condition adverse to quality.
On the basis of operating experience in which ASME Code margins
. were not maintained, the staff concluded that evidence exists that at least some licensees were not in compliance with their licensing basis.
l Procosed Draft Generic Letter and Guidelines for Imolementation
- Mr. Emmett Murphy, NRR Mr. Murphy, presented the purpose of the proposed GL.
The purpose i
of this proposed GL is to: (1) inform licensees that actions beyond l
current TS requirements may be necessary to ensure SG tube integrity; (2) request that licensees implement the actions described herein to ensure that SG tube integrity is monitored and maintained consistent with regulatory requirements and the plant's licensing bases; and (3) require that licensees submit a written response to the NRC regarding the hplementation of the requested actions.
In addition, this propo_ad GL is intended to provide guidance to licensees that may wish to amend their TSs to permit the use of " defect-specific management (DSM) " methodologies for monitoring and maintaining SG tube integrity.
Draft'Reculatorv Guide DG 1074 - Mr. Emmett Murphy, NRR Mr. Murphy discussed the guidance provided in DG 1074 regarding the SG tube'. integrity.
The draft guide provides detail guidance concerning the implementation of the mocel TSs.
Draft DG 1074 also
O Joint Mat'l & Metallurgy August 26-27, 1997 and Sev. Acc. Minutes provides guidelines concerning the monitoring of operational leakage and the acceptable basis for determining appropriate TSs for operational leakage limits.
The model TSs accompanying the regulatory guide are basically intended to be performance-based, that is based on TSs for performances tied to ensuring SG tube integrity.
Mr. Murphy made the following key points regarding the draft regulatory guide:
Performance criteria commensurate with tube integrity. These criteria are benchmarks against which the effectiveness of licensee actions for ensuring tube integrity is evaluated.
Licensees must adjust their programs as necessary to ensure that these criteria are met.
The performance criteria are consistent with the current design and licensing basis for PWRs.
Inservice inspection (ISI) of the tubing relative to the tube repair criteria. Non-destructive examination (NDE) techniques and personnel must undergo a performance-based qualification to ensure that the tubing can be reliably inspected relative to the tube criteria and that the condition of the tubing can be reliably monitored relative to the performance criteria.
The frequency and scope of inspection must be such as to ensure that the performance criteria will be met.
Operational primary-to-secondary leakage and implementing leakage limits such that the performance criteria are not exceeded.
The proposed GL requests that the licensees submit a proposed change to the TS limiting condition for operation (LCO) limits on operational leakage in any one SG if and as necessary to make these limits consistent with the performance criteria.
The condition monitoring, "as-found" condition of the tubing during each ISI with respect to the performance criteria.
This is a " backward-looking" assessment to confirm that the overall program has been successful in ensuring that the performance criteria are, in fact, being met.
Nuclear Enerav Institute (NEI) Presentation - Mr. Clive Callaway, NEI I
{
Mr. Callaway, presented the industry's views and made the following 1
comments:
The NEI working group will bring generic SG issues to closure by advocating broader and more consistent implementation of
)
industry guidance by the utilities.
I l
l k
Joint Mac'l & Metallurgy August 26-27, 1997 and Sev. Acc. Minutes The industry initiatives offer the best approach.
The TS will j
not provide a satisfactory solution.
Technical Specification amendments are resource intensive for the licensees and the NRC staff.
I Status of Differina Professional Ooinion (DPO) - Mr. Tim Reed, NRR Mr. Reed, presented the background and the status of the DPO evaluations.
He stated that the staff has documented its consideration of the DPO issues, but the document is not complete.
The staff will provide the document to ACRS in near future.
The concerns raised by the author of the DPO and the staff's responses are summarized below:
1.
NDE Issue:
The concern was identified that nondestructive examination (NDE) techniques are not capable of adequately detecting and sizing intergranular stress corrosion cracking (IGSCC) and that correlations of leakage versus an NDE parameter such as bobbin coil voltage cannot be reliably used to calculate leakage for design basis events such as a main steam line break (MSLB). Additionally, it was stated that the complex morphology of IGSCC cracks and the limitations of current NDE technology make it impossible to construct such correlations.
Staff's Response:
The subject of limitations in NDE was specifically evaluated in NUREG-1477, Voltage-Based Interim Plugging Criteria for Steam Generator Tubes," and in the methodology developed in GL fs5-05, Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Af fected by Outside Diameter Stress Corrosion Cracking. "
Further, a major portion of the proposed new regulatory framework is focused on providing the necessary guidance on how to assure that NDE methods are properly qualified and NDE uncertainties accounted for in assessing tube integrity.
2.
Iodine Soikina Issue:
The concern is that the iodine spike (i.e.,
the release of radioactive fission products into the RCS from the fuel through cladding perforations) following a large depressurization transient such as a MSLB, may be greater than the value of 500 assumed in the standard review plan (SRP) dose assessment methodology.
Additionally, the concern is raised that the iodine spike (i.e.,
r Joint Mat'l & Metallurgy August 26-27, 1997 and Sev. Acc. Minutes 500) could increase to higher multipliers as initial RCS coolant activity is decreased so that simply reducing initial RCS activity (via a technical specification change) may not result in a one-for-one reduction in calculated accident doses.
Staff's ResoonggJ, To assess the issue of iodine spiking, the staff has reviewed the industry models and performed independent analyses of available data and models.
Based on these assessments, the staff has concluded that although there are no data corresponding directly to the rapid depressurization conditions associated with a postulated MSLB, the spiking factors assumed in the existing dose assessment methodologies provide a reasonable level of conservatism for calculation of doses against the 10 CFR Part 100 guideline limits.
3.
Main Steam Line Break Issue:
The concern is that elevated tube differential presaure caused by design basis secondary depressurization transients (including MSLB) can cause primary-to-secondary leakage that could be greater than the leakage from a steam generator tube rupture (SGTR).
The leakage may be sufficient to deplete the refueling water storage tank (RWST) inventory via ECCS injection lost to the secondary side of the SGs (and therefore not available for recirculation from the containment sump) thereby leading to core damage.
Staff's Resoonse:
" Voltage-Based Interim Plugging Criteria for Steam Generator Tubes" documented the staff's consideration of proposed voltage-based repair criteria (which subsequently developed into GL 95-05).
In the NUREG-1477, the staff explicitly analyzed steam generator tube leakage during secondary side depressurization I
events, including main steam line breaks (MSLB). Thermal-hydraulic analyses were performed using the REIAP code to assess plant response to MSLB events involving a range of primary-to-secondary leakage.
Calculations were performed over a range of leak rates from a few hundred gpm to over one thousand gpm.
For these calculations, tube leakage was not assumed to begin until the differential pressure increased to above normal operating levels.
The results showed that the primary-to-secondary pressure j
differential decreases immediately following the break, since primary as well as secondary pressures are decreasing.
This is due to the cooling effects on the RCS caused by the high steam flow rates resulting from the secondary side break.
Following depletion of the secondary inventory in the affected generator, the RCS pressure begins.to increase as the cooling effect ceases and the emergency core cooling system (ECCS) injection increases the 4
a
Joint Mat'l & Metallurgy August 26-27, 1997 and Sev. Acc. Minutes reactor coolant inventory.
This causes the pressure differential across the tubes to exceed the normal operating differential pressure.
The pressure differential would increase to approximately 2230 psid assuming no tube leakage and normal operator actions.
The staff concluded that if primary-to-secondary leakage occurred during a secondary depressurization (leak rates as great as that of a tube rupture), that suf ficient margin in RWST capacity existed to mitigate the event before core damage occurred.
4.
Risk Increase Issue:
The concern is that the frequency of core damage with containment bypass may be approximately 3.4 X 10-4 per reactor year. This risk value was initially attributed to the increased risk resulting from the increased potential for RWST depletion as a result of large l
postulated primary-to-secondary tube leakages as discussed in issue 2 above.
More recently, this concern has included references to
]
station blackout sequences, and it has been implied that severe i
accident leakage and failures of degraded tubes under such conditions could lead to higher risk.
Staff's Resoonse:
The risk assessment performed by the staff in support of the development of the proposed new regulatory framework considered the issues raised in the DPO.
A major conclusion of the risk assessment and associated regulatory impact analysis was that a generic backfit requiring licensees to take action to reduce risk associated with SG tube degradation could not be supported per the criteria of 10 CFR 50.109.
However, the risk evaluation also concluded that (1) the conclusion regarding inability to support a generic backfit was based on licensees taking action beyond chose required by the current technical specifications, (2) further assessments are necessary to determine if plant-specific vulnerabilities require additional action, and (3) certain forms of alternate repair criteria could potentially have an adverse impact on risk.
It is the staff's intent to include assessment of plant-specific vulnerabilities as part of the individual plant examination (IPE) followup program.
Also, since some forms of alternate repair criteria could introduce new vulnerabilities and contributions to risk, under the proposed regulatory framework the staff would review supporting risk assessments as part of their determination of the acceptability of new ARC.
14 l
w j
4 Joint Mat'l & Metallurgy August 26-27, 1997 and Sev. Acc. Minutes 5.
Severe Accident Issue:
The concern is that the SG tubes may fail prior to other portions of the reactor coolant pressure boundary (RCPB) due to (1) inadequate NDE characterization (leaving in service potentially large numbers of through-wall flaws or flaws that grow through-wall during the operating cycle and which subsequently fail under severe accident conditions prior to other portions of the.RCPB),
(2) increased flow through the tube cracks (as small as pin-hole leaks) resulting in increased heat transfer to the tubes and a change in the thermal-hydraulic regime analyzed during this portion of the severe accident, (3) the cracks in tubes opening and unplugging due to increased pressure, and (4) the potential for jets from the cracks to erode and fail adjacent tubes leading to a large release.
Staff's Resoonse:
Regarding severe accidents, the staff has concluded that core damage conditions, particularly those associated with high primary pressure, dry steam generator secondary side events, can introduce vulnerabilities that have not been previously considered.
These vulnerabilities are the result of challenges to tube integrity from the high reactor coolant system (RCS) temperatures predicted during these events. The staff considered high temperature ef fects in its risk assessment and also assessed the potential for plant-specific vulnerabilities due to particular forms of degradation.
As a result of these assessments, the staff has concluded that certain vulnerabilities will be considered along with results of the previously mentioned IPE followup program.
Subcommittee Recommendations The subcommittee recommended that representatives of the staff brief the full committee on these issues.
Followun Actions The staff plans to provide the ACRS copy of the final Safety Evaluation Report on BWRVIP-05 by end of Calendar year 1997 or early 1998.
And a copy of the resolution document of the DPO regarding the steam generator tube integrity by the mid September 1997.
Backaround Material Provided to the Subcommittee:
e "BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05), " EPRI TR-105697, September 1995
Joint Mat'l & Metallurgy August 26-27, 1997 and Sev. Acc. Minutes l
e Draft Safety Assessment by the Office of Nuclear Reactor j
Regulation related to the review of the Topical Report by the j
Boiling Water Reactor Vessel and Internal Project:
"BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations" (BWRVIP-05) e SECY-97-152, dated July 18, 1997,
Subject:
Status of Saft v Evaluation Report on Proposed Reduction in Augments Examination Requirements for Boiling Water Reactor Pressure Vessels Pursuant to 10 CFR 50.55a (g) (6) (ii) (A) (SRM M970152B) e NRC Information Notice 97-63: Status of NRC Staff's Review of BWRVIP-05, dated August 7, 1997 e
Letter dated November 20, 1996, from T.
S.
Kress, Chairman, ACRS, to James M. Taylor, Executive Director for Operations,
Subject:
Proposed Rule on Steam Generator Tube Integrity Memorandum dated June 16, 1997, from Joram Hopenfeld, Task Manager, Generic Safety Issues Branch, Division of Engineering Technology, Office of Nuclear Regulatory Research, to L.
Joseph Callan, Executive Director for Operations,
Subject:
Differing Professional Opinion Regarding Voltage-Based Repair Criteria for Steam Generator Tubes, Steam Generator Rulemaking e
Memorandum dated June 27,
- 1997, from L.
Joseph Callan, Executive Director for Operations, to Commissioners,
Subject:
J.
Hopenfeld's Differing Professional Opinion Concerning Voltage-Based Repair Criteria for Steam Generator Rulemaking e
E-Mail Message to the Office of EDO from the Author of the DPO Compilation of Concerns Identified by ACRS Members e
e Summary of Differing Professional Opinion Issues Presented to the ACRS on November 7, 1996 e
Memorandum dated July 22, 1997, from Jocelyn Mitchell, Senior Level Technical Advisor, Office of the Executive Director for Operations, to DPO File,
Subject:
Discussion with J. Hopenfeld e
ACRS Letter dated June 20, 1997, " Proposed Regulatory Approach Associated with Steam Generator Integrity" EDO Letter dated July 15, 1997, responding to the ACRS Letter of June 20, 1997 8
U.
S.
Nuclear Regulatory Commission, NUREG-1570,
" Risk Assessment of Severe Accident-Induced Steam Generator Tube Rupture, Draft Report for Comment," revised May 1997
Joint Mat'l & Metallurgy August 26-27, 1997 and Sev. Acc. Minutes U.
S.
Nuclear Regulatory Commission Regulatory Analysis,
" Regulatory Approach for Steam Generator Tube Integrity," May 1997 Draft Proposed raneric Letter regarding Steam Generator Tube Integrity and b sociated Draft Regulatory Guide DG-1074, dated August 12, 1997 eeeeeeeee..........
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NOTE:
Additional details of this meeting can be obtained from a transcript of this meeting available in the NRC Public Document Room, 2120 L Street, N.W.,
Washington, D.C.
20006, (202) 634-3274, or can be purchased from Neal R.
Gross & Co., Inc., Court reporters and Transcribers,1323 Rhode Island Avenue, N.W., Washington, D.C.
20005, (202) 234-4433.