ML20198P386
| ML20198P386 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 10/31/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20198P384 | List: |
| References | |
| NUDOCS 9711070135 | |
| Download: ML20198P386 (5) | |
Text
_ -.
oa asouq g
y UNITED STATES j
g j
NUCLEAR REGULATORY COMMISSION o
e WASHINGTON, D.C. acN64001
%*...+/
[
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELIEF FROM CERTAIN REACTOR PRESSURE VESSEL INSFlCTION REQUIRQjDWi PUBLIC SERVICE ELECTRIC AND GAS COMP 681 ATLANTIC CITY ELECTRIC COMPANY HOPE CREEK GENERATING STATION DOCKET NO. 50-351 1.0. INTRODUCTION By letter dated August 29,199/, as supplemented by letters dated September 16 and October 1, 1997 Public Service Electric and Gas Company (PSE&G or the licensee) requested an alternative to performing the reactor
)ressure vessel (RPV) circumferential shell weld examination requirements of
)oth the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, 1983 Edition, through the Summer 1983 Addenda (Inservice Inspection), and the augmented examination requirements of 10 CFR 50.55a(g)(6 ii The alternative was)(pro)p(A)(2) for the Hope Creek Generating Station (HCGS).
osed pursuant to the provisions of 10 CFR 50.55a(a)(3)(1) and 10 CFR 50.b5afg)(ii)[A)(5), and is consistent with information contained in Informatton Not9ce (IN) 97-63, " Status of NRC Staff Review of BWRVIP-05." The September 16, 1997, letter contained supplemental information requested by the staff during telephone conference calls with the licensee on September 11 and 15, 1997, related to plant procedures, operator training, and the results of the Inservice Inspection ISI) of the RPV.
The October 1,1997, letter provided clarification regardin(g the regulatory basis for the request and the proposed alternativt.
The alternative proposed by PSE&G is the >erformance of inspections of essentially 100 percent of the HCGS RPV siell longitudinal seam welds and essentially 0 percent of the RPV shell circumferential seam welds during Refueling Outage 7, which will result in )artial examination (2 - 3 percent) of the c' reumferential welds at or near tie intersections of the longitudinal and circumferential welds.
The requirement for ISI, which includes RPV circumferential weld inspection, derives from the Technical Specifications (TSs) for HCGS, which state that the ISI and testing of the 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 10 CFR 50.55a(g)(4), ASME Code Class 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
@ ]Og 37gg ENCLOSURE G
1 practical within the limitations of design sometry and materials of construction of the components. Theregulat$onsrequ,irethat. inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addendi of the ASME 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 l cable ASME Code,Section XI, for HCGS during the curre subject to the limitations and modifications listed therein. The app 1 10-year ISI interval is the 1983 Edition through the Sumer 1983 Addenda.
Section 50.55a(g)(6 (ii)(A) to Title 10 of the Code of federal #egulations (10 CFR 50.55a(g)(6 (ii)(A)) requires that licensees perform an expanded RPV shell weld exam < nat on as specified in the 1989 Edition of Section XI of the ASME Code, on an " expedited" basis.
" Expedited," in this context, effectively-meant during the insp xtion interval when the Rule was approved or the first period of the next P.spection interval. The final Rule was published'in the federal Register '.,o August By incorporating into the i
regulations the 1989 Edition of the ASME Code, t6, 1992 (57 FR 34666)he NRC s licensees perform volumetric examination of " essentially 100 percent" of the RPV pressure-retaining shell welds during all inspection intervals.
Section 50.55a(a)(3)(1) (10 CFR 50.55a(a)(3)(1)) indicates that alternatives to the requirements tn 10 CFR 50.55a(g) are justified when the proposed alternative provides an acceptable level of quality and safety.
l By letter dated September 28, 1995, as supplemented by letters dated June 24 and October 29, 1996, May 16, June 4, and June 13, 1997, the Boiling Water Reactor Vessel and Internals Project (BWRVIP), a technical committee of l
the BWR Owners Group (BWROG), submitted the proprietary report, "BWR Vessel and Internals Project, BWR Reactor Vessel Shell Weld Ins >ection Recommendations (BWRVIP-05)," which proposed to reduce tie scope of inspection 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 examination 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 the axial and circumferential welds would have included approximately 2-3 percent of the circumferential welds.
On May 12, 1997, the NRC' staff and members of the BWRVIP met with the Commission to discus; the NRC staff's review of the BWRVIP-05 report.
In accordance with guidance provided by the Commission in Staff Requirements Memorandum (SRM) M9705128, 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 alternatives to the augmented examination in accordance with 10 CFR 50.55a(a)(3)(1), 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 L
l
o 3-i circumferentini welds during the fall 1997 or spring 1998 outage seasons.
Acceptably justified alternatives 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.
2.0 BACKGROUND
Staff Assessment of BWRVIP-05 Report The staff's independent assessment of the BWPutP-05 proposal is documented in a letter dated August 14, 1997, to Mr. Carl Y ry, 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.
The staff's assessment was performed for BWR RPVs fabricated by Chicago Bridge and Iron (CB&I), Combustion Engineering (CE), and Babcock & Wilcox (B&W). The staff assessment identified cold overpressure events as the limiting transients that could lead to failure of BWR RPVs. Using the pressure and temperature resulting from a cold overpressure event in a foreign reactor and the parameters identified in Table 7-1 of the staff's independent assessment, the staff determined the conditional probability of failure for axial and circumferential welds fabricated by CB&I, CE, and B&W. Table 7-9 of the staff's 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&I, 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 the.,) at the vessel inner surface of Il4.5'F.
In the uncertainty analysis, (RT neutron fluence evaluation had the greatest RT value 145'F) at the staff's assessment will have le., values less than9 hose res(ulting fro inner surface. Vessels with RT ss embrittlement than the vessels simulated in the staff's assessment and should have a conditional probability of vessel failure less than or equs1 to the values in the staff's 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 overpressure event and the conditional
)robability of vessel failure for B&W fabricated circumferential welds, reactor year and the uncertainty bound failure frequency is 3.9 X aest-estimate failure frequency from the staff's assessment is 6.0 X per per reactor year.
G 3.0 DISCUSSION The licensee indicated in the August 29, 1997, letter that the basis for requesting the alternative 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 l
r
(
l 4
was also demonstrated in the staff's independent assessment of the BWRVIP-05 report. The BWRVIP-05 report indicates that, fo a typical BWR RPV, the failure probability for axial welds is,i.7 X 10 [and the failure probability 2
for circumferential welds-is 2.2 X 10' for 40 years of plant operation.
The-licenscecalculated-theRT,ftherequestedreliefperiodusingthe-value for the limitin Hope Creek circumferential weld at -the enlo methodology in Regulatory Guide (RG) 1.99, Revision 2.
The RT values calculated in accordance with RG 1.99 Revision 2, depend upon"t'he neutron fluence, the amounts of copper and nickel in the circumferential weld, and its unirradiated RT.,.
The licensee determined the maximum neutron fluence at the end of the next two operating cycles at the circumferential beltline weld to be 0.0166 X 10, inner surface of the limiting n/cm3(
at the end of the requested relief period is 0.0268 X 10,t,he highest fluence n/c'n ). The amounts of copper and nickel in the limiting circumferential beltline weld is 0.08 percent and 0.63 percent, respectively.
The plant-specific unirradiated RT., for the limiting circumferential beltline weld is -30'F.
Using these parameters and the methodology in RG 1.99, Revision 2, the licensee determined that the RT
'6*F. which is less than the reference case for the B&W fabricated period is II. value for the circumferential weld at the end of the relief vessels in the staff's assessment..Since the RT of the Hope Creek beltline circumferential weld is less than the values in Die staff's assessment, the licensee concluded that the conclusions of the BWRVIP-05 report are bounded for the Hope Creek RPV.
The licensee assessed the systems that could lead to a cold overpressurization of the HCGS RPV. These included the high-pressure coolant injection (HPCI),
reactor core isolation cooling (RCIC), standby-liquid control (SLCS), control rod drive (CRD) and reactor water cleanup (RWCU) systems.
Both the HPCI and RCIC pumps are steam driven and do not function during cold shutdown.
SLCS automatically-initiates if reactor power is not downscale after a low reactor water level or high reactor pressure condition. Automatic initiation of SLCS should not occur during shutdown; however.-the SLCS injection rate is
--approximately 46 gpm, which would allow operators sufficient time to control reactor pressure if manual initiation occurred.- The CRD and RWCU systems use a feed and bleed process to control RPV level and pressure during shutdown.
The CRD pumps injection rate is less than 60 gpa, which allows 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.
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 )ressure limits of the operating systems, operator training, and establisted plant-specific procedures, the licensee determined that a non-design basis cold
4
- overpressure transient is unlikely to occur during the next two operating cycles. Therefore the licensee concluded that the probability of a cold overpressure translent is considered to be less than or equal to that used in the staff's assessment.
4.0 EVALUATION The staff confirmed that the RT., than the values in the reference case and values for the circumferential welds at the end of the relief period are less 1
uncertainty analysis for the BW-fabricated vessels.
RT.,is a measure of the amount of irradiation embrittlement. Since the RT the values in the reference case and the values in the., values are less than uncertainty analysis for BW fabricated vessels,.- the Hope Creuk RPV will have less embrittlement than the BW fabricated vessels and will have a conditional probability of vessel failure less than or equal-to that estimated in the staff's assessment.
Based on pressure limits on the operating systems, and the licensee's operator training and established procedures, the probability of a cold overpressure transient should be minimized during the next two operating periods.
5.0 CONCLUSION
S Based upon its review, the staff reached the following conclusions:
- 1) Based on the licensee's assessment of the materials in the circumferential weld in the beltline of the HCGS RPV, the conditional probability of vessel failure should be less than or equal to that estimated from the staff's assessment.
- 2) Based on the licensee's operator training and established proc 2dures, the probability of cold overpressure transients should be minimized during the next two operating periods.
- 3) Based on the previous two conclusions, the staff concludes that the Hops Creek RPV can be operated during the next two operating periods with an acceptable level-of quality and safety and the inspection of the circumferential welds can be delayed for two operating periods.
'Therefore, the proposed, alternative to perfoming the RPV examination requirements of the ASME Code,Section XI,1983 Edition, through Summer 1983 Addenda, and the augmented examination requirements of 10 CFR 50.55a(g)(6)(ii)(A>(2) at Hope Creek for circumferential shell welds for two operating cycles is authorized pursuant to 10 CFR 50.55a(a)(3)(1).
' Principal Contributors:
K. Kamoski
.K. Kavanagh Date:- October 31, 1997