GO2-04-125, Request for Permanent Relief from Inservice Inspection Requirements of 10 CFR 50.55(g) for the Volumetric Examination of Circumferential Reactor Pressure Vessel Welds

Request for Permanent Relief from Inservice Inspection Requirements of 10 CFR 50.55(g) for the Volumetric Examination of Circumferential Reactor Pressure Vessel Welds

ML042150393
Person / Time
Site:	Columbia
Issue date:	07/15/2004
From:	Atkinson D Energy Northwest
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
GO2-04-125
Download: ML042150393 (11)

Text

ENVERGY NORTHWEST PO. Box 968 . Richland, Washington 99352-0968 July 15, 2004 G02-04-125 10 CFR 50.55a U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555

Subject:

COLUMBIA GENERATING STATION, DOCKET NO. 50-397 REQUEST FOR PERMANENT RELIEF FROM INSERVICE INSPECTION REQUIREMENTS OF 10 CFR 50.55a(g) FOR THE VOLUMETRIC EXAMINATION OF CIRCUMFERENTIAL REACTOR PRESSURE VESSEL WELDS

References:

(1) NRC Generic Letter 98-05, "Boiling Water Reactor Licensees use of the BWRVIP-05 Report to Request Relief from Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Welds."

(2) Letter Dated July 28, 1998, NRC to Carl Terry (BWRVIP Chairman),

"Final Safety Evaluation of the BWR Vessel and Internal Project BWRVIP-05 Report (TAC No. M93925)."

Dear Sir or Madam:

Pursuant to 10 CFR 50.55a(a)(3)(i), Energy Northwest herein requests NRC approval for permanent relief from inservice inspection requirements of 10 CFR 50.55a(g) for the volumetric examination of circumferential reactor pressure vessel welds for Columbia Generating Station (Columbia). This category of welds is specified in the American Society of Mechanical Engineers Code Section XI, Table IWB-2500-1, Examination Category B-A, Item B 1.11, Circumferential Shell Welds. The attached request (2ISI-027) is for the remaining term of Columbia's current operating license and demonstrates, as described in reference 1, that at the expiration of Columbia's license, the welds will continue to satisfy the limiting conditional failure probability for circumferential welds as specified in reference 2.

REQUEST FOR PERMANENT RELIEF FROM INSERVICE INSPECTION REQUIREMENTS OF 10 CFR 50.55a(g) FOR TIE VOLUMETRIC EXAMINATION OF CIRCUMFERENTIAL REACTOR PRESSURE VESSEL WELDS Page 2 of 2 The request also describes operator training and procedures that are established at Columbia to limit the frequency of low temperature over-pressurization events to the amount specified in reference 2. Energy Northwest requests approval of 21SI-027 by March 2005, to support planning for the next refueling outage scheduled for May 2005.

If you have any questions or require additional information regarding this matter, please contact Mr. DW Coleman, Manager, Regulatory Programs, at (509) 377-4342.

Respectfully, DK Atkinson Vice President, Technical Services Mail Drop PE08

Attachment:

10 CFR 50.55a Request cc: BS Mallet - NRC - RIV WA Macon - NRC - NRR NRC Sr. Resident Inspector - 988C RN Sherman - BPA/1399 TC Poindexter - Winston & Strawn

10 CFR § 50.55a Request Attachment Page I of 9 10 CFR 50.55a Request Number 21SI-027 Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)

-Alternative Provides Acceptable Level of Quality and Safety-ASMEE Code Components Affected Weld No.' Description Code Category Item No.

AA Bottom head to #1 shell course B-A B1.11 AB #1 to #2 shell course B-A B1.11 AC #2 to #3 shell course B-A B1.11 AD #3 to #4 shell course B-A B1.11

'See ISI Diagram RPV-101 (pg. 12)

Applicable Code Edition and Addenda

The applicable code is American Society of Mechanical Engineers (ASME)Section XI 1989 Edition with no Addenda. Columbia Generating Station (Columbia) is in its second inservice inspection (ISI) interval, third inspection period.

Applicable Code Requirements Volumetric examination of these welds is required once every inspection interval in accordance with ASME Section XI, 1989 Edition, with no Addenda, Table IWB-2500-1, Code Category B-A, Item Number B 1.I 1.

Reason for Request

Energy Northwest requests this reduction in the number of Reactor Pressure Vessel (RPV) circumferential welds requiring examination in order to avoid unnecessary inspections and to reduce radiological dose. Implementation of the proposed alternative would provide these benefits while maintaining an adequate level of quality and safety.

10 CFR § 50.55a Request Attachment Page 2 of 9 Proposed Alternative and Basis for Use On November 10, 1998, the NRC issued Generic Letter (GL) 98-05 (reference 2) informing Boiling Water Reactor (BWR) licensees that they may request permanent relief from the inservice inspection requirements of 10 CFR 50.55a(g) for the volumetric examination of RPV circumferential welds, Code Category B-A, Item Number B 1.11 by demonstrating the following:

1. At the expiration of the current license, the circumferential welds will continue to satisfy the limiting conditional failure probability as specified in reference 4.

2. Licensees have implemented operator training and established procedures that limit the frequency of Low Temperature Over-Pressure (LTOP) events as specified in reference 4.

Energy Northwest herein requests approval to implement this alternative examination methodology for Columbia as allowed by GL 98-05 and proposes to modify Columbia's ISI schedule to perform inspections of essentially 100 percent of the RPV axial shell welds and essentially zero percent of the RPV circumferential welds (item B 1.11, reference 1).

Approximately two to three percent of circumferential welds will continue to be examined at their points of intersection with the axial welds. These inspections are being proposed as an alternative to the ISI requirements for circumferential welds in the ASME Code,Section XI, 1989 Edition (no Addenda).

The Boiling Water Reactor Vessel and Internals Project (BWRVIP) report BWRVIP-05 (reference 3) provides the technical basis for eliminating inspection of RPV circumferential shell welds. This report was transmitted to the NRC in September 1995 and based on the report, the NRC staff concluded in their final safety evaluation of July 28, 1998 (reference 4),

that licensees may request relief from performing the inspections described therein. In this safety evaluation the staff found that the limiting plant specific failure frequency for RPV circumferential shell welds is 8.2 x 10-8 /year. On March 7, 2000, the NRC issued a supplement (reference 5) to the July 28, 1998, final safety evaluation of the BWRVIP-05 report. In this supplement, the NRC concluded that the resultant RPV failure frequencies due to failure of the limiting BWR axial welds are below the acceptable Regulatory Guide (RG) 1.154 (reference 6) threshold of 5.0 x 10 4 /year, given the assumptions described in the supplemental safety evaluation. Together, these safety evaluations establish a conclusion that the failure frequency of BWR RPV circumferential welds is orders of magnitude less than that of the axial welds.

In GL 98-05, the NRC stated that the estimated failure frequency of the BWR RPV circumferential welds is well below the acceptable Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) criteria discussed in RG 1.174 (reference 7). The NRC also indicated that the estimated frequency of RPV circumferential weld failure bounds the corresponding CDF and LERF that may result from an RPV weld failure. In the July 28,

10 CFR § 50.55a Request Attachment Page 3 of 9 1998, safety evaluation, the staff justified elimination of ISI for RPV circumferential welds because the failure frequencies for circumferential welds in BWR plants are significantly below the criteria specified in RG 1.154 and the CDF of any BWR plant, and that continued future inspections would result in negligible decrease in an already acceptably low value. The NRC further concluded in GL 98-05 that the proposal in the BWRVIP-05 report, as modified by two criteria, was acceptable and that the BWR licensees may request permanent relief from the inservice inspection requirements of 10 CFR 50.55a(g) for the volumetric examination of circumferential RPV welds provided they meet the two criteria discussed below. The GL specifically states that licensees still need to perform the required inspections of "essentially 100 percent" of all axial welds.

Generic Letter 98-05, Criterion 1

'At the expiration of their license, the circumferential welds will continue to satisfy the limiting conditional failure probability for circumferential welds in the NRC's July 28, 1998, safety evaluation."

Energy Northwest Response The NRC evaluation of the BWRVIP-05 report included a Probabilistic Fracture Mechanics (PFM) analysis to estimate RPV failure probabilities. Three key assumptions in the PFM analysis are:

1. The neutron fluence was that estimated to be the end-of-license mean fluence;

2. The chemistry values are mean values based on vessel types; and

3. The potential for beyond design basis events is considered.

For Columbia's RPV, the single circumferential weld joint located between shell course 1 and shell course 2 within the beltline region (identified as weld AB on ISI diagram RPV-101) is the limiting circumferential weld. For plants such as Columbia, with RPVs fabricated by Chicago Bridge & Iron (CB&I), the mean end-of-license neutron fluence used in the NRC PFM analysis was 5.lE+18 n/cm2 at 32 Effective Full Power Years (EFPY). The fluence value for the limiting circumferential weld at the expiration of Columbia's current operating license (assuming 33.1 EFPY) is 3.09E+ 17 n/cm 2 . Thus, the fluence effect on embrittlement is lower for Columbia than that assumed for the corresponding weld in the NRC PFM analysis.

Changes in the nil ductility reference temperature (ARTNDT), may be used as one of the means for monitoring the amount of irradiation embrittlement. The calculated embrittlement shift in RTNDT (ARTNDT) for Columbia at the end-of-license (33.1 EFPY) is 28 degrees F. By comparison, Table 2.6-4 of the NRC's final safety evaluation of the BWRVIP-05 report (reference 5) indicates an embrittlement shift of 109.5 degrees F at 32 EFPY for CB&I fabricated vessels. Table 1 of this Attachment captures this comparison data and illustrates that

10 CFR § 50.55a Request Attachment Page 4 of 9 the calculated ARTNDT value for Columbia's RPV is less than, and thus bounded by, the embrittlement shift assumed in the NRC's safety evaluation for the BWRVIP-05 report. An additional margin of conservatism exists considering Columbia's fluence and ARTNDT values are calculated for a greater value of EFPY than that assumed in the Staff's analysis.

For these reasons, the limiting circumferential weld at Columbia is less brittle than the corresponding weld in the PFM case study and is therefore bounded by the Staff's limiting conditional failure probability for CB&I circumferential welds (P(FIE)) in table 2.6-4 of the NRC's safety evaluation (reference 5). Thus, Criterion 1 of GL 98-05 is met.

Generic Letter 98-05 Criterion 2 "Licensees have implemented operator training and established procedures that limit the frequency of cold over-pressure events to the amount specified in the NRC's July 28, 1998, safety evaluation."

Energy Northwest Response In GL 98-05, the NRC stated that beyond design-basis events occurring during plant shutdown could lead to LTOP events that could challenge RPV integrity. The BWRVIP assessment indicated that the major contribution to LTOP event frequency results from unmitigated injections from condensate or control rod drive systems and a failure to properly realign the reactor water cleanup system following a reactor trip at low temperatures could potentially cause an LTOP event. For a BWR to experience such an event would require several operator errors. Although no LTOP events have occurred at a domestic BWR, the NRC identified several events that could be considered precursors to such an event and cited one actual LTOP event that occurred at a foreign BWR. This event involved a series of operational errors that allowed a Control Rod Drive (CRD) pump to run until the vessel went water-solid with no outflow from the reactor resulting in a maximum RPV pressure of 1150 psi within a temperature range of 79F to 88F. The probability that the operator fails to take action to mitigate coolant injection is a key variable in assessing the frequency of LTOP events.

Procedural Controls to Prevent LTOP Events Operating procedures and Operator training programs at Columbia are barriers that make an LTOP event unlikely during low temperature evolutions such as RPV pressure testing at the conclusion of a refueling outage. These procedures require continuous monitoring and control of reactor water level, pressure, and temperature during cold shutdown and refueling operations.

The Operations procedure governing control room activities requires that operators continuously monitor indications and alarms, to detect abnormalities as early as possible, and immediately notify the control room supervisor of any changes or abnormalities in indications.

This procedure requires that changes, which could affect reactor water level, pressure, or

10 CFR § 50.55a Request Attachment Page 5 of 9 temperature, be performed only under the auspices of a Senior Reactor Operator (SRO). This ensures any deviations in reactor water level or temperature from specified parameters will be promptly identified and corrected. Additionally, at each shift turnover, operators discuss the status of plant conditions, any on-going activities that could affect critical plant parameters, and contingency planning. This ensures that on-coming operators are aware of any activities that could adversely affect reactor water level, pressure, or temperature. These procedures minimize the likelihood of an LTOP event from occurring and are reinforced through periodic operator training.

Work Management Control A review of industry operating experience indicates that inadequate work management is a potential contributor to a cold over-pressure event. At Columbia, an outage management group schedules work performed during outages. All work activities are reviewed against a shutdown safety plan and coordinated through an outage control center, which provides operations oversight. In the control room, the SRO is required to maintain continuous attention to any work activity that could potentially affect reactor level or decay heat removal during refueling outages. Pre-job briefings are conducted for work activities that have the potential of affecting critical reactor parameters. The individuals involved in the work activity attend these briefings and discuss expected plant responses and contingency actions to address unexpected events or conditions that may be encountered.

Operator Training to Prevent RPV LTOP Events Procedural controls for reactor temperature, level, and pressure are an integral part of operator training. Specifically, operators are trained in methods of controlling water level within specified limits, as well as responding to abnormal water level conditions outside the established limits.

Licensed operator training further reduces the possibility of an LTOP. The initial licensed operator training curriculum covers brittle fracture and vessel thermal stress; operational transient procedures, including the operational transient on reactor high water level; technical specifications limiting conditions for operation; and, simulator training of plant heat up and cool down including performance of surveillance tests which ensure pressure/temperature curve adherence. In addition, periodic operator training reinforces management's expectations for strict procedural compliance and conservative decision-making.

Industry Events Review Energy Northwest continuously reviews operating experience to ensure Columbia's procedures and training are revised to benefit from lessons learned from industry events, including LTOP events. This is done with the objective of precluding similar events from occurring at Columbia.

a- X 10 CFR § S0.55a Request Attachment Page 6 of 9 Considering the operational and administrative barriers discussed above, the probability that the operator fails to take action to mitigate coolant injection is low enough to assure the frequency of an LTOP event at Columbia is bounded by the amount specified in the NRC's safety evaluation (reference 4). Thus, Criterion 2 of GL 98-05 is met.

Conclusion The BWRVIP-05 report provides the technical basis for eliminating inspection of 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 of the axial shell welds. Based on an assessment of the materials in the limiting circumferential weld in the beltline of Columbia's RPV, the conditional probability of RPV failure is less than or equal to that estimated in the NRC's analysis through the end of the current operating license.

Based on established operator training, practices and procedural controls, the frequency of an LTOP event at Columbia is less than or equal to the frequency assumed in the NRC's July 28, 1998, safety evaluation.

Duration of Proposed Alternative This proposed alternative will be implemented upon NRC approval and remain in effect until the expiration of Columbia's current operating license on December 20, 2023.

Precedents The following plants have been granted permanent relief from performing examinations of Code Category B-A, item number B1. 11, RPV circumferential welds in accordance with GL 98-05:

• LaSalle Station Units 1 and 2, SER dated January 28, 2004

• Pilgrim Nuclear Power Station, SER dated April 11, 2003

• Brunswick Units 1 and 2, SER dated September 14, 2000

• Grand Gulf Unit 1, SER dated April 11, 2001

• Susquehanna Units 1 and 2, SER dated February 28, 2001

• Peach Bottom, SER dated June 15, 2000

• Hope Creek, SER dated November 1, 1999

• Perry, SER dated February 15, 2001

• Monticello, SER dated July 27, 2001

• Fitzpatrick, SER dated February 22, 2000 References

1. American Society of Mechanical Engineers (ASME)Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," 1989 Edition, No Addenda.

10 CFR § 50.55a Request Attachment Page 7 of 9

2. Generic Letter 98-05, "Boiling Water Reactor Licensees Use of the BWRVIP-05 Report To Request Relief From Augmented Examination Requirements On Reactor Pressure Vessel Circumferential Shell Welds," November 10, 1998

3. EPRI Report Proprietary Report TR-105697, "BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05)," dated September 1995

4. Letter, Gus C. Lainas (NRC) to Carl Terry, BWRVIP Chairman, "Final Safety Evaluation of the BWR Vessel and Internals Project BWRVIP-05 Report (TAC No.

MA3925), dated July 28, 1998

5. Letter, Jack R. Strosider (NRC) to Carl Terry, BWRVIP Chairman, "Supplement to Final Safety Evaluation of the BWR Vessel and Internals Project BWRVIP-05 Report (TAC No. MA3395), dated March 7, 2000

6. RG 1.154 Regulatory Guide 1.154 - Format and Content of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors (ML003740028) (Draft SI 502-4 published 0111986)

7. RG 1.174 Regulatory Guide 1.174 - An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis Revision 1, November 2002

I1, 10 CFR § 50.55a Request Attachment Page 8 of 9 TABLE 1 Columbia Generating Station RPV Circumferential Weld AB Information Columbia's USNRC Limiting Plant Specific Limiting Weld Analyses Parameters at 32 EFPY, Wire (2) Safety Evaluation Table 2.6-4 Neutron fluence at the end-of-license (1)(n/cm2) 3.09E+ 17 5.1E+18 Initial (unirradiated) reference temperature (RTNDT) -50.0F -65.0F Weld chemistry factor (CF) 108 134.9 Weld copper content 0.08% 0.10%

Weld nickel content 0.936% 0.99%

Increase in reference temperature due to irradiation (ARTNDT) 28.0F 109.5 0F Mean adjusted reference temperature (Mean ART = RTNDT + ARTNDT) -22.0F 44.5 0 F Notes:

1. The end-of-license fluence is projected to 33.1 EFPY

2. Weld wire (5P6756) of weld AB is limiting weld material in RPV, GE Nuclear Energy, NEDC-33144P, "Pressure-Temperature Curves for Energy Northwest Columbia," San Jose, CA, April 2004, pg 21

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