ML20216B808
| ML20216B808 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 04/02/1998 |
| From: | Geoffrey Edwards PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9804140114 | |
| Download: ML20216B808 (10) | |
Text
A v
(
PECO NUCLEAR nm e w l
Nuclear Group Heaquarters A Unir or ITCO EMucy 965 kward sj L
April 2,1998 Docket No. 50-277 License No. DPR-44 i
l U.S. Nuclear Regule.' '/ Commission Attn: Document Control Center Washington, DC 20555 Subject-Peach Bottom Atomic Power Station, Un;t 2 1
Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)
References:
1.
Letter from J. F. Stolz (U.S. Nuclear Regulatory Commission (USNRC)) to G. A. Hunger, Jr. (PECO Energy Company), dated i
j July 2,1997 2.
Letter from J. F. Stolz (USNRC) to G. A. Hunger, Jr. (PECO l
l Energy Company), dated October 7,1997 l
l
Dear Sir / Madam:
1 Attached for your review is an alternative plan for the examination of the Peach Bottom Atomic Power Station (PBAPS), Unit 2 Reactor Pressure Vessel (RPV) shell welds. This plan is similar to previously approved alternative plans that were the subject of the Reference 1 and 2 safety evaluations for PBAPS, Unit 3. We request your approval by September 1, l
1998 in order to support the upcoming PBAPS, Unit 2 refueling outage scheduled to begin in early October,1998.
f l
If you have any questions, please contact us.
I l
Very truly yours,
=b t
Garrett D. Edwards jCjC Director-Licensing f(OM
)
Attachment cc:
H. J. Miller, Administrator, Region I, USNRC A. C. McMurtray, USNRC Senior Resident inspector, PBAPS 9004140114 980402 PDR ADOCK 05000277 P
Dockst No. 50-277 License No. DPR-44 s
PROPOSED ALTERNATIVE REACTOR VESSEL SHELL WELDS Peach Bottom Atomic Power Station, Unit 2 Proposed Alternative PECO Energy Company proposes an alternative in accordance with 10 CFR 50.55a(a)(3)(i), pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5). The following is a discussion of l
the proposed alternative. This proposed alternative is composed of two parts, which are discussed below in Sections 1 and 2.
j A proposed alternative is requested for the examination of the RPV circumferential shell welds (Section XI Exam Cat. B-A, item No. B1.11) as required by 10 CFR l
i 50.55a(g)(6)(li)(A)(2), and the inservice inspection requirements for circumferential welds in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI,1980 Edition through Winter 1981 Addenda (Table IWB-2500-1, Examination Category B-A, Item No. 81.11) for two (2) operating cycles. The basis for this alternative is provided in Section 1.
i Additionally, PECO Energy Company (PECO Energy) is unable to meet the 90% volume coverage requirement for each longitudinal weld of the PBAPS, Unit 2 reactor vessel as required by 10 CFR 50.55a(g)(6)(ii)(A)(2). Therefore, PECO Energy proposes that an I
alternative plan be accepted in lieu of the 90% volume coverage < The basis for this l
alternative is provided in Section 2.
l Section 1: Basis for Proposed Alternative to Circumferential Shell Weld insoections PECO Energy requests a proposed alternative from the examination of the RPV circumferential shell welds (Section XI Exam Cat. B-A, item No. B1.11) as required by 10 CFR 50.55a(g)(6)(ii)(A)(2), and the inservice inspection requirements for circumferential welds in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI,1980 Edition through Winter 1981 Addenda (Table IWB-2500-1, j
Examination Category B-A, item No. B1.11) for two (2) operating cycles. PECO Energy I
will be performing an examination of the reactor vessel longitudinal shell welds to the maximum extent praatical from the inner diameter, within the constraints of vessel internal restrictions. The, extent of weld examination coverage anticipated for the longitudinal shell welds is identified on Table 1, and is further discussed in Section 2. It should be noted that our current examination plan is designed to provide longitudinal weld coverage; however, incidental coverage will result in ari estimated portion of 2-3 percent of the intersecting circumferential weld.
The basis for this request is documented in the report "BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05)",
1 j
Docket No. 50-277 i
License No. DPR-44 that was transmitted to the NRC in September 1995. The BWRVIP-05 report provides the technical basis for eliminating inspection of Boiling Water Reactor (BWR) RPV circumferential shell welds. The BWRVIP-05 report concludes that the probability of failure of the BWR RPV circumferential shell welds is orders of magnitude lower than that I
of the axial shell welds. The NRC staff has conducted an independent risk-informed
{
assessment of the analysis contained in BWRVIP-05. This assessment also concluded that the probability of failure of the BWR RPV circumferential welds is orders of magnitude l
lower than that of the axial shell welds. Additionally, the NRC assessment demonstrated that inspection of BWR RPV circumferential welds does not measurably affect the l
probability of failure. Therefore, the NRC evaluation appears to support the conclusions of J
i l
This independent NRC assessment utilized tho FAVOR code to perform a probabilistic l
fracture mechanics (PFM) analysis to estimate RPV failure probabilities. Three key assumptions in the PFM analysis are: the neutron fluence was that estimated to be end-l of-license mean fluence, the chemistry values are mean values based on vessel types, and the potential for beyond design basis events is considered. Although BWRVIP-05 l
provides the technical basis supporting an alternative, the following information is provided to show the conservatisms of the NRC analysis relative to the PBAPS, Unit 2 vessel.
l For plants with RPVs designed by Babcock & Wilcox, such as those at Peach Bottom, the mean end-of-license neutron fluence used in the NRC PFM enalysis was 5.3x10" n/cm.
2 However, at PDAPS, Unit 2, the highest fluence anticipated at the end of the requested l
period (October,2002 (2R14)) is 5.2 x 10"n/cm'. Thus, embrittlement due to fluence l
effects is lower, and the NRC analysis as described at an August 8,1997 meeting with industry, is conservative for PBAPS, Unit 2 in this regard. At this August 8,1997 meeting, l
the NRC presented the results of their independent risk assessment of BWR reactor vessel shell welds. Therefore, there is conservatism in the already low circumferential l
weld failure probabilities as related to PBAPS, Unit 2.
l The following table illustrates that the PBAPS, Unit 2 plant has additional conservatism in comparison to the NRC's Independent Assessment Fracture Analysis limiting case (i.e.
l B&W SN 2 in Table 7-7). The chemistry factor, ARTum, margin term, mean ART, and upper bound ART are calculated consistent with the guidelines of Regulatory Guide 1.99, Rev. 2.
2
o Dock:t No. 50-277 License No. DPR-44 i
Parameter PBAPS, Unit 2 NRC Independent I
Description Comparative Assessment Limiting Parameters at 19 EFPY Fracture Analysis (Bounding Cire. Weld)
Parameters Fluence, n/cm' 5.2 x 10" 1.25 x 10" Initial RTwor, 'F
-32
-5
)
Chemistry Factor 82 190 Cu %
0.06 0.287 Ni %
0.97 0.60 A RTuor 20.1 87.9 Margin Term 20.1 62.2 Mean ART
-11.9 82.9 Upper Bound ART 8.1 145.1 As shown above, each parameter used in the limiting NRC Independent Assessment report (excluding Ni%) bounds the circumferential shell weld information for PBAPS, Unit 2 at 19 EFPY, the EFPY at the end of the requested deferral period. (The combination of the Ni% and Cu% determine the Chemistry Factor, which is itself bounded by the NRC Independent Assessment.)
At the August 8,1997 meeting, the NRC staff indicated that the potential for, and consequences of, non-design basis events not addressed in the BWRVIP-05 report should be considered. In particular, the NRC staff stated that non-design basis, cold, over-pressure transients should be considered. It is highly unlikely that a BWR would experience a cold, over-pressure trans!9nt. At the August 8,1997 meeting, the NRC staff described several types of events that could be precursors to BWR RPV cold, over-pressure transients. These were identified as precursors because no cold, over-pressure event has occurred at a U. S. BWR. Also at the August 8 meeting, the NRC staff identified one actual cold, over-pressure event that occurred during shutdown at a non-U. S. BWR.
This event apparently included several operational errors that resulted in a maximum RPV j
l pressure of 1150 psi with a temperature range of 79 F to 88 F.
As provided in the following discussion, PECO Energy has in place procedures which monitor and control reactor pressure, temperature, and water inventory during all aspects i
l of cold shutdown which would minir Qe ce likelihood of a Low Temperature Over-Pressurization (LTOP) event from ocr mng. Additionally, these procedures are reinforced through operator training.
3 i
l
Docket No. 50-277 License No. DPR-44 The code Leakage Pressure Test and the code Hydrostatic Pressure Test procedures l
which have been used at PBAPS, have sufficient procedural guidance to prevent a cold, l
over-pressurization event. The Leakage Pressure Test is performed at the conclusion of l
each refueling outage, while the Hydrostatic Pressure Test is performed once every ten years. Other pressurizations required for informational leakage inspections are performed 1
in accordance with a procedure similar to the ASME Code test procedures. These l
pressurizations are infrequently-performed, complex tasks, and the test procedures r.ce considered Plant Evolution / Special Tests. As such, a requirement is included in them for l
Operation's Section management to perform a " pre-job briefing" with all essential personnel. This briefing details the anticipated testing evolution with special emphasis on:
conservative decision making, plant safety awareness, lessons learned from similar in-house or industry operating experiences, the importance of open communications, and, i
finally, the process in which the test would be aborted if plant systems responded in an adverse manner. Vessel temperature and pressure are required to be monitored throughout these tests to ensure compliance with the Technical Specification pressure-l temperature curve. Also, the procedures require the designation of a Test Coordinator for the duration of the test who is a single point of accountability, responsible for the coordination of testing from initiation to closure, and maintaining Shift Management and line management cognizant of the status of the test, i
Additionally, to ensure a controlled, deliberate pressure increase, the rate of pressure increase is administratively limited throughout the performance of the test. If the pressurization rate exceeds this limit, direction is provided to remove the CRD pumps, which are used for pressurization, from service.
l With regard to inadvertent system injection resulting in an LTOP condition, the high i
pressure make-up systems (High Pressure Coolant injection (HPCI) and Reactor Core Isolation Cooling (RCIC) systems, as well as the normal feedwater supply (via the Reactor Feedwater Pumps)) at PBAPS are all steam driven. During reactor cold shutdown conditions, no reactor steam is available for the operation of these systems. Therefore, it is not possible for these systems to contribute to an over-pressure event while the unit is in cold shutdown.
In the case of low pressure system initiation, the PBAPS, Unit 2 pressure-temperature limit curves for hydrostatic testing (PBAPS, Unit 2 Technical Specifications Figure 3.4.9-1),
permit pressures up to 312 psig at temperatures from 70 F up to 100 F. Abovo 100 F, the permissible pressure increases immediately to above 560 psig and increases rapidly with increasing temperature. The maximum discharge pressure for the PBAPS Core Spray and Residual Heat Removal Pumps, taking a suction from the Torus at atmospheric pressure, are approximately 373 psig and 360 psig, respectively. Therefore, the potential for an over-pressurization event which would significantly exceed the pressure-temperature limits, due to an inadvertent actuation of these systems, is very low.
Procedural control is also in place to respond to an unexpected or unexplained rise in reactor water level which could result from a spurious actuation of an injection system.
Actions specified in this procedure include preventing condensate pump injection, l
A
\\
f I
t __
b
l Docket No. 50-277 License No. DPR-44 securing ECCS system injection, tripping CRD pumps, terminating all other injection sources, and lowering RPV level via the RWCU system.
In addition to procedural barriers, License'd Operator Training has been held which further reduces the possibility of the occurrence of LTOP events. During Initial Licensed Operator Training the following topics are covered: Brittle fracture and vessel thermal stress; Operational Transient (OT) procedures, including the OT on reactor high level; Technical Specification training, including Section 3A.9, "RCS Pressure and Temperature (P/T)
Limits"; and, Simulator Training of olant heatup and cooldown including performance of surveillance tests which ensure pressure-temperature curve compliance. In addition, operator training has been provided on the expectations for procedural compliance, as provided for in the Stations' Operations Manual.
In addition to the above, ongoing review of industry operating plant experiences is conducted to ensure that the PECO Energy procedures consider the impact of actual events, including LTOP events. Appropriate adjustments to the procedures and associated training are then implemented, to preclude similar situations from occurring at PBAPS.
Based upon the above, the probability of a cold over-pressure transient is considered to be less than or equal to that used in the NRC analysis described at the August 8,1997 meeting and is conservative for PBAPS, Unit 2, Section 1: Summary Based on the documentation in BWRVIP-05, the risk-informed independent assessment performed by the NRC staff and the discussion above, PECO Energy believes a delay for two (2) cycles, which corresponds to October, 2002 (2R14)hfied.
in completing the inspection of the RPV circumferential shell welds at PBAPS, Unit 2 is jus 5
t Docket No. 50-277 License No. DPR-44 Sectior.. Basis for Proposed Alternative to 90% Lonaitudinal Volume Coveraae 10 CFR 50.55a(g)(6)(ii)(A)(2) states that all licensees shall augment their reactor vessel examinations by implementing the examination requirements for Reactor Pressure Vessel (RPV) shell welds specified in item B1.10 of Examination Category B-A, " Pressure Retaining Welds in Reactor Vessel," in Table IWB-2500-1 of Subsection IWB of the 1989 Edition of Section XI, Division I, of the ASME Boiler and Pressure Vessel Code, subject to l
the conditions specified in 50.55a(g)(6)(ii)(A)(3) and (4). For the purpose of this augmented examination, essentially 100% as used in Table IWB-2500-1 means more than 90% of the examination volume for each weld.
PECO Energy Company (PECO Energy)is unable to meet the 90% volume coverage requirement for each longitudinal weld of the PBAPS, Unit 2 reactor vessel as required by 10 CFR 50.55a(g)(6)(ii)(A)(2). Therefore, PECO Energy is proposing an alternative to the 90% volume coverage requirement for each longitudinal weld, in accordance with 10 CFR 50.55a(a)(3)(i), pursuant to 10 CFR 50.55a(g)(6)(ii)(A)(5). As discussed previously in Section 1, PECO Energy is also requesting a proposed alternative from the examination of the RPV circumferential shell welds (Section XI Exam Cat. B-A, item No.
B1.11).
PECO Energy intends to inspect the Peach Bottom Atomic Power Station (PBAPS), Unit 2 RPV during the upcoming PBAPS, Unit 2 refueling outage (2R12), currently scheduled to begin in early October,1998. The proposed alternative is to perform an examination of the RPV longitudinal shen welds to the maximum extent practical from the Inner Diameter (ID), within the constraints of vessel internal restrictions. This examination would be performed for longitudinal shell welds specified in item B1.10 of Examination Category B-A, " Pressure Retaining Welds in Reactor Vessel," in Table IWB-2500-1 of Subsection IWB of the 1989 Edition of Section XI, Division I, of the ASME Boiler and Pressure Vessel Code. Further examination from the ID is not practical without disassembly of the vessel internal components. Table 1 provides the longitudinal welds, the estimated volumetric examination coverage from the ID, and the ID restrictions that will potentially obstruct scanning. The volumetric examination coverage and restrictions from the ID are based on a detailed weld coverage scaa plan completed in February,1998. As shown in Table 1,8 i
of 15 welds achieve greater than 90% volumetric coverage, crediting ID inspections only.
j There are no vessel internals which pose a restriction to the ID oxamination which are i
easily removable. The only removable components (i.e., not welded to the vessel) which limit scan coverage are the feedwater spargers. However, it is impractical to remove the feedwater spargers due to the potential for damage to the sparger seals and nozzles.
Therefore, there are no components which can be removed to increase coverage from the ID.
For those longitudinal welds where greater than 90% volumetric examination may not be achieved from the ID, the estimated supplemental coverage and physical constraints on the vessel Outer Diameter (OD) are identified in Table 1. As noted in Table 1, further review has determined that two (2) welds (RPV-V1 A and RPV-VSA) would exceed 90 percent volume coverage with a supplemental OD examination. In some locations, the 6
Docket No. 50-277 l
l License No. DPR-44 additional weld volume that can be accessed from the OD is a subset of the ID examinations.
The restrictions which prohibit unrestrained access to 100% of the longitudinal weld volume from the OD are the vessel insulation and the vessel nozzles.
l The percentage of longitudinal weld volume coverage estimated from the ID examination l
reflects a significant portion of the total reactor pressure vessel weld length. Attempting to perform supplemental OD examinations would result in a minimal increase to longitudinal weld volume coverage. Obtaining this additional coverage would only permit two I
additional welds to exceed the 90% volumetric coverage requirement. Additional 1
disassembly and reassembly of portions of the reactor vessel biological shield and insulation would result in further increases in personnel radiation exposure of laborers and technicians assigned to these tasks as well as an increase in the general area dose rates in the Drywell for the entire population of workers. Additional doses to the entire population to perform the supplemental (OD) examinations contained in Table 1 are estimated to be 24 man-Rem, in order to perform the OD examinations, an estimated cost I
of approximately $460,000, beyond the cost of performing the ID examinations, would be l
incurred. Therefore, based on the incremental cost and radiation dose, in conjunction with the limited additional volumetric coverage, PECO Energy has concluded that performing
{
OD examinations to increase coverage beyond that achieved from the ID would result in undue hardship without a compensdng increase in safety.
During the fabrication process of the PBAPS, Unit 2 RPV, all of the shell welds were thoroughly examined using several examination methods as required by the original construction code. Additionally, all of the shell welds received volumetric examinations prior to initial plant operations, as prescribed by the ASME Section XI Preservice inspection requirements. Selected shell welds have received volumetric examinations during the first inservice Inspection interval in accordance with ASME Section XI Inservice Inspection requirements. No reportable indications were identified during any of these examinations. The PSI and ISI examination summaries were reviewed to identify planar
)
and/or linear indications and are listed in Table 1. All indications in Table 1 were code i
acceptable.
I The General Electric (GE) GERIS-2000 System will be used to perform the remote controlled, automated UT examinations of the RPV. This tool has been used previously at PBAPS, Unit 3 and other Boiling Water Reactors for the purpose of RPV examinations.
j GE demonstrated this system at the Performance Demonstration Initiative (PDl),
qualification Session No. 61-02, in accordance with the 1992 Edition,1993 Addenda of ASME Boiler and Pressure Vessel Code,Section XI, Appendix Vill requirements.
i Appendix Vill was developed to ensure the effectiveness of UT examinations within the nuclear industry by means of a rigorous, item specific, performance demonstration. The performance demonstration was conducted on an RPV mockup containing flaws of i
various sizes and locations. The demonstration established the capability of equipment,
}
procedures, and personnel, which are similar or the same as those that wit: be used at PBAP6, Unit 2, to find flaws that could be detrimental to the integrity of the RPV. Although 7
Dockst No. 50-277 License No. DPR-44 Appendix Vill is not currently required by regulation, the qualification of equipment, procedures, and personnel to Appendix Vlli criteria demonstrates examination and evaluation techniques that surpass the requirements of the ASME Boiler and Pressure Vessel Code,Section XI referenced by the rule.
PECO Energy will provide a summary report of the results of the RPV examinations to the NRC no later than 90 days from the completion of the outage, in the " Summary Report Submittal," required by Article IWA-6000 of the ASME Boiler and Pressure Vessel Code,Section XI. This submittal will include the individual weld coverage obtained with the GERIS-2000 System, and the examination results.
l
(
Section 2: Summarv As described in BWRVIP-05; 1) the inherent flaw tolerance of the Boiling Water Reactor (BWR) vessel due to lower radiation embrittlement and less challenging design and l
operational loadings,2) the quality of the original vessel fabrication, 3) the lack of I
significant degradation mechanisms, and 4) the results of previous vessel examinations at other operating Boiling Water Reactors, which was confirmed during the inspections at PBAPS, Unit 3 (October,1997), provides the basis to conclude that the proposed i
alternative plan to perform extensive and distributed, high-quality, vessel-shell-weld examinations from the ID will provide an acceptable level of quality and safety, l
l l
l 1
l I
l 1
1 8
1 Dock t No. 50-277 License No. DPR-44 TABLE 1 RESPONSE TO REQUEST FOR ADDTIONALINFORMATION REACTOR PRESSURE VESSEL ALTERNATIVE PLAN PBAPS, UNIT 2 IWiliflONEishiV01'IDWRiittlitl6dsidMEissiV61'00TRiislitl6Ws00~M P81'l 18t iddliistl6hil RPV-V1A 81.2 %
SSBP, JPRE, 10%
1,NOZ NONE NOZ RPV V1B 81.2 %
SSBP, JPRE 0%
1,NOZ NONE J
RPV Vic 81.2 %
SSBP, JPRE 0%
1,NOZ NONE RPV V2A 97.7 %
JPRB N/A N'A NONE RPV V2B 99.6 %
JPRB N/A N/A LAMINAR SPOT RPV V2C 99.6%
JPRB N/A N/A NONE RPV V3A 72.9 %
CSP, FWS 10%
l CLAD INTERFACE RPV V3B 72.9%
CSP, FWS 10%
i NONE RPV V3C S3.6%
FWS, CSDC 10%
i NONE RPV V4A 99.9 %
NOZ N/A N/A NONE RPV V4B 100.0 %
None N/A N/A NONE RPV V4C 100.0 %
None N/A N/A NONE RPV V5A 88.0%
SDB 12%
None N/A N/A CLAD INTERFACE RPV VSC 100.0 %
None N/A N/A NONE Abbreviations:
SSBP Shroud Support Baffle Plate CSDC - Core Spray Downcomer NOZ VesselNozzle JPRE - Jet Pump Riser Elbow JPRB - Jet Pump Riser Brace CSP Core Spray Pipe FWS Feedwater Sparger SDB - Steam Dryer Bracket I-Insulation N/A - Not Applicable 9