ML20216J421
| ML20216J421 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 03/18/1998 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20216J409 | List: |
| References | |
| NUDOCS 9803230438 | |
| Download: ML20216J421 (6) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION
&!.TERNATIVE TO INSPECTION OF REACTOR PRESSURE VESSEL CIRC,UMFERENTIAL WELDS GRAND GULF NUCLEAR STATION EtLTERGY OPERATIONS. INC.
DOCKET NO: 50-416 '
1.0 Introduction By letter dated February 11,1998, Entergy Operations, Inc. (Entergy or the licensee) requested an altemative to performing the reactor pressure vessel (RPV) circumferential shell weld examinations requirements of both the American Society of l@chanical Engineers (ASME)
Boiler and Pressure Vessel Code,Section XI,1992 Edition, with portions of the 1993 Addenda ,
(inservice inspection), and the augmented examination requirements of 10 CFR 50.55a(C)(6)(ii)(A)(2) for the Grand Gulf Nuclear Station (GGNS). The s!temative was proposed pursuant to the provisions of 10 CFR 50.55a(a)(3)(1) and 10 CFR 50.55a(g)(ii)(A)(5),
and is consisient with information contained in Information Notice (IN) 97-63, " Status of NRC Staff Review of BWRVIP-05."
The alternative proposed by Entergy is the performance of inspections of essentially 100
- percent of the GGNS RPV she'llongitudinal seam welds and essentia!!y zero percent of the RPV shell circumferential seam welds during Refueling Outage RF11, which will result in partial examination (2 to 3 percent) of the circumferential welds at or near the intersections of the longitudinal and circumferential welds.
The requirement for inservice inspections, which include RPV circumferential weld inspection, derives from the Technical Specifications (TS) for GGNS which state that the inservice inspection (ISij anc bsting of the American Society of Mechanical Engineers (ASME) Code Class 1,2, and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, and applicable addenda, as required by 10 CFR 50.55a(g).
Pursuant to the requirements of 10 CFR 50.55t(g)(4), ASME Code C'ess 1,2, and 3 components shall meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, " Rules for Inservice inspection of Nuclear Power Plant Components," to the extent "
practical within the limitations of design, geometry, and materiais of construction of the components. The regulations require that inservice examination of components an't system ,
Attachment i
9803230438 980318 PDR ADOCK 05000416 P PDR
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pressure tests conducted durinp the first 10-ysar interval and subsequent intervals comply with the requirements in the latest edition and addenda of the ASfWE Code,Section XI, incorporated by reference in 10 CFR 50.55a(b) on the date 12 months prior to the start of the 120-month interval,' au'oject to the ! imitations and modifications listed thefein. The applicable ASME Code, y ~ Section X!, for GGNS, during the currunt 10-year ISI interval is the 1993 Edition, with portions of the 1993 Addenda.
Section 50.55a(g)(6)(ii)(A) to Title 10 of the Cocle of FederalRegeist/ons (10 CFR 50.55a(g)(6)(ii)(A)) squires that licensees perform an expanded RPV shell weld examination as specified in the 1989 Edition of Section XI of the ASME Code, on an j " expedited" oasis. " Expedited,"in this context, effectively meari during the inspochon mtsrval ;
when the Rule was approved or the first period of the next inspection interval. The final Rule
, wag published in the FederalRegister on Aucus' 6,1992 (57 FR 34666).' By incorporating into
( the regulations the i389 Edition of the ASME Cede, the NRC staff sequired thallicensees perferm volumetrb examinaticn of" essentially 100 percent" of the RPV pressure-retaining shell vmids during all inspection intervals. Section 50.55a(s)(3)(i) (10 CFR 50.55a(a)(3)(!)) indicates that alternatives to the requirements in 10 CFR 50.55a(g) are justified when the proposed eftemative provides an acceptable level of quality and safey.
By letter dated September 28,1995, as supplemented by letters dated June 24, and October 29,1996, and May 16, June 4, Juno 13, and December 18,15?07, the Boiling Water Reactor Vessel and intemals Project (BWRVIP), a technical committee of the BWR Owners Gioup (BWROG), submitted the proprietary report, "BWR Vessel and Intemals Project, BWR Reactor Vessel Shell Weld inspection Recommendations (BWRVIP-05)," which proposed to reduce the scope ofinspection of the BWR RPV welds from essentially 100 percent of all RPV shell welds to 50 percent of the axial welds and 0 percent of the circumferential welds. By letter dated October 29,1996, the BWRVIP modified their proposal to increase the examinatim of the axial welds to 100 percent from 50 percent while still proposing to inspect essentially 0 percent of the circumferential RPV shell welds, except that the intersection of tne axial and circumferential welds would have included approximately 2 to 3 percent of the circumferential welds.
~ On May 12,1997, the NRC staff and members of the BWRVIP met with the Commission to -
discuss the NRC staffs review of the BWRVIP-05 report. In accordance with guidance provided by the Commission in Staff Requirements Memorandum (ORM) M970512B, dated May 30,1997, the staff has initiated a broader, risk-informed review of the BWRVIP-05 proposal.
In IN 97-63, the staff indicated that it would consider technically-justified attematives to the augmented examination in accordance with 10 CFR 50.55a(a)(3)(i),10 CFR 50.55a(a)(3)(ii),
and 50.55a(g)(6)(ii)(A)(5), from BWR licensees who are scheduled to perform inspections of the BWR RPV circumferential welds during the fall 1997 or spring 1998 outage seasons.
Acceptably-Justified attematives would be considered for inspection delays of up to 40-months or two operating cycles (whichever is longer) for BWR RPV circumferential shell welds only.
. o 2.0 Background - Staff Assessment of BWRVIP-05 Report The stafs independent assessment of the BWRVIP-05 proposal is documented in a letter dated August 14,1997, to Mr. Carl Terry, BWRVIP Chairman. The staff concluded that the industry's assessment does not sufficiently address risk, and additional work is necessary to provide a complete risk-informed evaluation.
1 The stafs assessment was performed for BWR RPVs fabricated by Chicago Bridge and Iron (CB&l), Combustion Engineering (CE), and Babcock & Wilcox (B&W). The staff assessment identified cold over-pressure events as the limiting transients that could lead tc failure of BWR RPVs. Using the pressure and temperature resulting from a cold over-pressure event in a foreign reactor and the parameters identified in Table 7-1 of the sta#s independent assessment, the staff determined the conditional probability of failure for axial and circumferential welds fabricated by CB&l, CE, and B&W. Table 7-9 of the stafs assessment identifies the conditional probability of failure for the reference cases and the 95 percent confidence uncertainty bound cases for axial and circumferential welds fabricated by CB&l, CE and B&W. B&W fabricated vessels were determined to have the highest conditional probability of failure. The input material parameters used in the analysis of the reference case for B&W fabricated vessels resulted in a reference temperature (RT ) at the vessel inner surface of 114.5'F. In the uncertainty analysis, the neutron fluence evaluation had the greatest RT, value (145'F) at the inner surface. Vessels with RTc values less than those resulting from the stats assessment will have less embrittlement than the vessels simulated in the stats assessment and should have a conditional probability of vessel failure less than or equal to the
. values in the stafs assessment.
The failure probability for a weld is the product of the critical event frequency and the conditional probability of the weld failure for that event. Using the event frequency for a cold over-pressure event and the conditional probability of vessel failure for CB&l fabricated circumfersntial welds, the best-estimate failure frequency from the stafs assessment is
<6.0 X 10-" m per reactor year and the uncertainty bound failure frequency is <2.8 X 1F*
per reactor year, 3.0 Licensee Technical Justification The licensee indicated in the February 11,1998, letter that the basis for requesting the attemative inspections is the BWRVIP-05 report, which stated that the probability of failure of BWR RPV circumferential shell welds is orders of magnitude lower than that of the axial shell welds. This conclusion was also demonstrated in the sta#s independent assessment of the BWRVIP-05 report. The BWRVIP-05 report indicates that, for a typical BWR RPV, the failure probability for axial welds is 2.7 X if and the failure probability for circumferential welds is 2.2 X 10* for 40 years of plant operation.
The licensee calculated the RT, value for the limiting GGNS circumferential welds at the end of the requested relief period using the methodology in Regulatory Guide (RG) 1.99, Revision 2.
2 Insufficient or no failures to accurately determine reference case failure probability.
4 Since there are no circumferential welds in the beltline region, the limiting circumferential welds are weld seams AB and AC which are approximately 5 inches below and 22 inches above the reactor core, respectively. The RT, values calculated in accordance with RG 1.99, Revision ,
2, depsnd upon the neutron fluence, the amounts of copper and nickel in the circumferential j weld, and its unirradiated RT . The licensee determined the highest neutron fluence at the end of the next two operating cycles at the inner surface of the circumferential welds to be 0.102 x 105 n/cm2. This value resulted from linearly interpolating the peak fluence in the beltline region to the end of the relief request period. The amounts of copper and nickel in weld seam AB are 0.03 percent and 0.81 percent, respectively. The amounts of copper and nickelin weld seam AC are 0.04 percent and 0.95 percent, respectively. The plant -specific unirradiated 1 RT, for weld seam AB is -40*F, and the plant-specific unirradiated RT, for weld seam AC is I
-20*F. Using these parameters and the methodology in Regulatory Guide 1.99, Revision 2, the licensee determined that the RT, values for circumferential weld seams AB and AC at the end of the relief period are -5.4*F and 25.5'F, respectively. Both values are less than the reference ,
case for the CB&l fabricated vessels in the staffs assessment. Since the RT, values of the I GGNS circumferential welds are less than the values in the staffs independent assessment, the licensee concluded that the GGNS vessel circumferential welds are bounded by the staffs independent assessment, thus providing additional assurance that the vessel welds are also bounded by the BWRVIP-05 report. l l
The licensee assessed the systems that could lead to a cold overpressurization of the Grand l Gulf reactor pressure vessel (RPV). These included the high pressure core spray (HPCS),
reactor core isolation cooling (RCIC), standby liquid control (SBLC), reactor feed pumps, condensate system, low pressure core spray (LPCS), low pressure core injection (LPCI),
control rod drive (CRD) and reactor water cleanup systems (RWCU).
The RCIC pumps are steam driven and do not function during cold shutdown. The licensee stated that the RCIC turbine was designed to operate on Auxiliary Steam for testing purposes, ,
however, operation with Auxiliary Steam is not allowed by procedure. The HPCS system is ,
motor operated and could be operated during cold shutdown. Startup of HPCS requires either )
manual initiation or inadvertent initiation. The HPCS system has a high level interlock for the HPCS injection valve to prevent overfilling the RPV. This interlock cannot be overridden and therefore, overpressurization of the RPV should not occur. The licensee stated that there were no automatic starts associated with SBLC. Operator initiation of SBLC should not occur during shutdown, however, the SBLC injection rate is approximately 42 gpm which would allow operators sufficient time to control reactor pressure if manual initiation occurred.
The reactor feed pumps are the high pressure makeup system during normal operations. The reactor feed pumps are steam driven and therefore, cannot be operated during cold shutdown.
The condensate system is the supply source to the reactor feed pumps.- The condensate pumps require manualinitiation and line up for injection during cold shutdo.vn. If the resulting reactor level increase was ignored, the reactor pressure-temperature limit would not be exceeded due to the condensate shutoff head of approximately 150 psig. The LPCS and LPCI systems are low pressure ECCS systems with low shutoff heads. If either one of these systems were manually or inadvertently initiated during cold shutdown, the resulting reactor pressure and temperature would be below the pressure-temperature limits. The CRD and RWCU systems use a feed and bleed process to control RPV level and pressure during normal
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cold shutdown conditions. Per plant procedures, the reactor head vents are open when reactor coolant temperature is less than 190*F. The ORD pumps injection rate is less than 60 gpm; !
this flowrote and the opened reactor head vents allow sufficient time for operators to react to unanticipated level changes.
In all cases, the operators are trained in methods of controlling water level within specified limits in addition to responding to abnormal water level conditions during shutdown. The licensee also stated that procedural controls for reactor temperature, level, and pressure are an integral part of operator training. Plant-specific procedures have been established to provide guidance to the operators regarding compliance with the Technical Specification pressure-temperature limits. On the basis of the pressure limits of the operating systems, operator training, and established plant-specific procedures, the licensee determined that a non-design basis cold over-pressure transient is unlikely to occur during the requested delay. Therefore, the licensee concluded that the probability of a cold over-pressure transient is considered to be less than or equal to that used in the stafs assessment of BWRVIP-05.
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4.0 Staff Review of Licensee Technical Justification The staff confirmed that the RT, values for the circumferential welds at the end of the relief period are less than the values in the reference case and uncertainty analysis for the CB&l fabricated vessels. RT,is a measure of the amount ofirradiation embrittlement. Since the RTa values are less than the values in the reference case and the uncertainty analysis for CB&l fabricated vessels, the GGNS RPV will have less embrittlement than the CB&l fabricated vessels and will have a conditional probability of vessel failure less than or equal to that estimated in the stars assessment.
The staff reviewed the information provided by the licensee regarding the GGNS high pressure injection systems, operator training, and plant-specific procedures to prevent RPV cold over-pressurization. The information provided sufficient basis to support approval of the l
sitemative examination request. The staff concludes that the probability of a non-design basis cold over-pressure transient occurring at GGNS during the requested delay !: 'ow, which is consistent with the stafs assessment.
5.0 Conclusions Based upon its review, the staff reached the following conclusions:
- 1) Based on the licensee's assessment of the materials in the circumferential welds in the ,
Grand Gulf RPV, the conditional probability of vessel failure should be less than or equal ,
to that estiinated from the stafs assessment.
- 2) Based on the licensee's high pressure injection systems analysts, operator training, and ;
plant specific procedures, the probability of a non-design basis cold over-pressure !
transient is low during the requested delay period and is consistent with the stafs l
assessment. -
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- 3) Based on the previous two conclusions, the staff concludes that the Grand Gulf RPV can be operated during the requested delay period with an acceptable level of quality and safety and the inspection of the circumferential welds may be delayed for the .
requested two operating periods.
~ Therefore, the proposed attemative to performing the RPV examination requirements of the ASME Code,Section XI,1992 Edition, with portions of the 1993 Addenda, and the augmented examination requirements of 10 CFR 50.55a(g)(6)(ii)(A)(2) at Grand Gulf for circumferential shell welds for two operating cycles is authorized pursuant to 10 CFR 50.55a(a)(3)(i).
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