ML14254A401

Response to Request for Additional Information for Relief Request 3ISI-14

ML14254A401
Person / Time
Site:	Columbia
Issue date:	08/27/2014
From:	Javorik A Energy Northwest
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
G02-14-118, TAC MF3435
Download: ML14254A401 (11)

Text

ENERGY O ~Alex ColumbiaP.O.

Ae IL Javorik Generating Box 968,Staftio PE04 NORTHWESTRic WA 99=

Ph. 509.377.85551 F. 509.377.4150 AUG 2 7 201%

G02-14-118 10 CFR 50.55a U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

COLUMBIA GENERATING STATION, DOCKET NO. 50-397 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION FOR REUEF REQUEST 3101-14

References:

1) Letter, dated February 6, 2014, AL Javork (Energy Northwest) to NRC, Inservice Inspection (ISI) Program Request 31S1-14"

2) Letter, dated July 31, 2014, Cad F. Lyon (NRC) to Mark E. Reddemann, "Columbia Generating Station - Request for Additional Information Related to Inservice Inspection Program Request 31S1-14 (TAC No.

MF3435)"

Dear Sir or Madam:

By Reference 1, Energy Northwest submitted for approval relief request 31S1-14.

Via Reference 2, the Nuclear Regulatory Commission (NRC) submitted a Request for Additional Information (RAI) to Energy Northwest. Enclosure 1 provides the requested information. Enclosure 2 provides a revision to the proposed relief request 31S1-14.

This letter and its enclosures contain no regulatory commitments. If there are any questions or if additional Information Is needed, please contact Ms. L. L. Williams, Licensing Supervisor, at 509-377-8148.

I declare under penalty of perjury that the foregoing is true and correct. Executed this 2-.L day of , 20.I*.

Respectfully A. L. Javonk Vice President, Engineering 64V,-7

. RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION FOR RELIEF REQUEST 31S1-14 Page 2 of 2

Enclosures:

1) Response to Request for Additional Information (RAI)

2) 31S1-14 Revision 1 cc: NRC RIV Regional Administrator NRC NRR Project Manager NRC Senior Resident Inspector/988C M Jones - BPA/1 399 (email)

JO Luce - ESFEC (email)

RR Cowley - WDOH (email)

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION FOR RELIEF REQUEST 31S1-14 Page 1 of 1 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)

NRC Reauest:

EVIB-RAI-1: Relief Request 31 SI-1 4 applying the ASME Code Case N-702, "Aftemative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds,, for the CGS third 10-year inservice inspection (ISI) interval does not mention the examination method for the nozzle radii. The NRC has approved use of ASME Code Case N-702, except for the VT-1 visual examination specified in the code case, for all past similar relief requests.

Therefore, please modify your request specifying your examination methodology to be either ultrasonic (UT) examination or VT-1 in accordance with the conditions placed upon the use of Code Case N-648-1, "Alternative Requirements for Inner Radius Examinations of Class 1 Reactor Vessel Nozzles,* as specified in the most recent edition of Regulatory Guide (RG) 1.147, "lnservice Inspection Code Case Acceptability, ASME Section XI, Division 1,* or provide additional information as to why your request should not be modified.

Please note that the condition placed by RG 1.147, Revision 16 (ADAMS Accession No. ML 101800S36), on the use of Code Case N-648-1 is to be revised, and a different condition is proposed in RG 1.147, Revision 17 for Code Case N-648-1 , which is expected to be issued near the end of this year.

Enemy Northwet Reomue:

Energy Northwest (ENW) used UT examination for the inner radius exams performed earlier in the interval. ENW will use UT examination for the inner radius exams for the remaining population identified in the relief request. Enclosure 2 contains Revision 1 to 31S1-14.

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION FOR RELIEF REQUEST 31SI-14 Page 1 of 8 10 CFR 50.55a Request Number 31S1-i 4 Revision 1 Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)

--Alternative Provides Acceptable Level of Quality and Safety--

1. ASME Code Component(s) Affected The components in the following table are affected by this request.

Table I Identification Code Itm Number Description Catgory Number Nl-0 RRC Nozzle to Vessel Weld @ 0 Deg B-D B3.90 N1-0-IR RRC Nozzle Inner Radius @ 0 Deg B-D B3.100 NI-180 RRC Nozzle to Vessel Weld 0 180 Deg B-D B3.90 NI-180-IR RRc Nozzle Inner Radius @ 180 Deg B-D B3.100 N2-30 RRC NOZZLE TO-VESSEL WELD 0 30 B-D B3.90 Deg N2-30-IR RRC NOZZLE INNER RADIUS @ 30 Deg B-D B3.100 N2-60 RRC NOZZLE TO VESSEL WELD @ 60 B-D B3.90 Deg N2-60-IR RRC NOZZLE INNER RADIUS 0 60 Deg B-D B3.100 N2-90 RRC NOZZLE TO VESSEL WELD 0 90 B-D B3.90 Deg N2-90-IR RRC NOZZLE INNER RADIUS 0 90 Deg B-D B3.100 N2-120 RRC NOZZLE TO VESSEL WELD @ 120 B-D B3.90

_ _,_.Degog N2-120-IR RRC NOZZLE INNER RADIUS @ 120 B-D B3.100 tDg N2-150 RRC NOZZLE TO VESSEL WELD 0 150 B-D 63.90 Deg N2-150-IR RRC NOZZLE INNER RADIUS 0 150 B-D B3.100 Deog N2-210 RRC NOZZLE TO VESSEL WELD 0 210 B-D B3.90 Deg N2-210-IR RRC NOZZLE INNER RADIUS 0 210 B-D 63.100 Deg N2-240 RRC NOZZLE TO VESSEL WELD @ 240 B-D B3.90

, DIg N2-240-IR RRC NOZZLE INNER RADIUS 0 240 1 B-D B3.100

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION FOR RELIEF REQUEST 31S1-14 Page 2 of 8 Table 1 IdentIfication Code Item Number Description Category Number Deg N2-270 RRC NOZZLE TO VESSEL WELD @ 270 B-D B3.90 Deg N2-270-lR RRC NOZZLE INNER RADIUS 0 270 B-D B3.100 Deg .,

N2-300 RRC NOZZLE TO VESSEL WELD @ 300 B-D B3.90 Dog N2-300-IR RRC NOZZLE INNER RADIUS 0 300 B-D B3.100 Deg N2-330 RRC NOZZLE TO VESSEL WELD @ 330 B-D B3.90 Deg N2-330-1R RRC NOZZLE INNER RADIUS 0 330 B-D 83.100 Deg RRC Reactor Recirculation

2. Aoolicable Code Edition and Addenda Columbia Generating Station ASME Section Xl Code is the 2001 Edition through the 2003 Addenda (Reference 3). Additionally, for ultrasonic examinations,Section XI, Appendix VIII, 'Performance Demonstration for Ultrasonic Examination Systems," is implemented as required (and modified) by 10 CFR 50.55a(b)(2)(xv).

3. Applicable Code Reouiremernt Class 1 nozzle-to-vessel weld and nozzle inner radii examination requirements are given in Subsection IWB, Table IWB-2500-1, 'Examination Category B-D Full Penetration Welds of Nozzles in Vessels - Inspection Program B," Item Numbers B3.90 and 83.100, respectively. The method of examination is volumetric. For the extent of examination, all nozzles with full penetration weld to the vessel shell (or head) and integrally cast nozzles must be examined each interval.

4. Reason for Reauest The identified nozzles (see Table 1) are scheduled for examination prior to the end of the current Inspection interval for Columbia. The interval began in December 2005 and will end in December 2015. The proposed alternative provides an acceptable level of quality and safety, and the reduction In scope could provide a dose savings of as much as 2500 mRem for the next refuel outage.

/ RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION FOR RELIEF REQUEST 31S1-14 Enclosure 2 Page 3 of 8

5. Progosed Alternative and Ba1sis for Use Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested from performing the required examinations on 100% of the identified nozzle assemblies in Table 1 above. As an alternative, for all welds and inner radii identified In Table 1, Columbia proposes to examine a minimum of 25% of the nozzle-to-vessel welds and inner radius sections, including at least one nozzle from each type and nominal pipe size, in accordance with Code Case N-702 (Reference 4). The proposed schedule to complete the minimum of 25% from each group is presented in Table 2 below. Columbia has completed three assembly (nozzle and inner radii) examinations already this interval using automated ultrasonic testing techniques. Greater than 97% coverage was: achieved on each location and no relevant indications were reported. With the acceptance of the relief request only one nozzle-to-vessel weld and one inner radius would be required to meet the 25% for the current interval. These examinations will be performed using ultrasonic testing techniques.

__Table 2 Number to Total be Group Number Examined Comments RRC Outlet 2 1 1 Completed in R19 (N1)

RRC Inlet (N2) 10 3 2 Completed in RI 9, 1 scheduled in I I1 R22 R19 Refuel Outage 19 (2009)

R22 Refuel Outage 22 (2015)

Basis for Use Boiling Water Reactor (BWR) Vessel Internals Project (BWRVIP) topical report BWRVIP-241, IBWR Vessel and Internals Project Probabilistic Fracture Mechanics Evaluation for the Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii" (hereafter referred to as BWRVIP-241) (Reference 2), documents supplemental analyses for BWR reactor pressure vessel (RPV) recirculation inlet and outlet nozzle-to-shell welds and nozzle inner radii. BWRVIP-241 was submitted to address the limitations and conditions specified in the December 19, 2007, safety evaluation (SE) for the BWRVIP-108NP report, "BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Nozzle-to-Vessel Shell Welds and Nozzle Inner Radii."

(Reference 1) The BWRVIP-108NP report contains the technical basis supporting American Society of Mechanical Engineers Boilerand PresSure Vessel Code (ASME Code) Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds," for reducing the inspection of RPV nozzle-to-vessel shell welds and nozzle Inner radius areas from 100 percent to 25 percent of the nozzles for each nozzle type during each 10-year interval. Based on the two

/ RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION FOR RELIEF REQUEST 31S1-14 Enclosure 2 Page 4 of 8 evaluations (BWRVIP-241 and BWRVIP-108NP), the failure probabilities due to a low temperature over pressure (LTOP) event at the nozzle blend radius region and the nozzle-to-vessel shell weld for Columbia recirculation nozzles are very low and meet the NRC safety goal.

Based on the results of this evaluation, the report concluded that the inspection of 25%

of each nozzle type is technically justified as per Code Case N-702.

EPRI report BWRVIP-241 received a final NRC Safety Evaluation (SE) on April 19, 2013 (ML13071A240) (Reference 6). In the SE, Section 5.0 OConditions and Umitations" indicates that each licensee who plans to request relief from the ASME Code, Section Xl requirements for RPV nozzle-to-vessel shell welds and nozzle inner radius sections may reference BWRVIP-241 report as the technical basis for the use of ASME Code Case N-702 as an alternative. However, each licensee should demonstrate the plant-specific applicability of the BWRVIP-241 report to their units in the relief request by demonstrating all of the following:

(1) The maximum Reactor Pressure Vessel (RPV) heatup/cooldown rate Is limited to less than 11 50 F per hour.

The Columbia Technical Specification limits the heatup/cooldown rate to less than or equal to 100 degrees F in any one hour period.

For the Recirculation Inlet Nozzles (N2) the following criteria must be met:

(2) (pr/t)/CRPCv<1.15.

(3) [p(ro 2 2 2 0 +r, )/(ro -rg )]/CNozzL<1 .47.

For the Recirculation Outlet Nozzles (Ni) the following criteria must be met:

(4) (pr/t)/CRpv< 1.15.

(5) [p (r02+r*0)(r?-

2 )/CNOozzE<1 C .59.

The terms to be used in the NRC SE Section 5 applicability evaluations criteria 2-4 are:

CRPV = recirculation Inlet nozzles N2 (from BWRVIP.241 model) = 19332 psi CNOZZLE = recirculation inlet nozzles N2 (from BWRVIP-241 model) = 1637 psi CRPV = recirculation outlet nozzles Ni (from BWRVIP-241 model) = 16171 psi CNOZZLE = recirculation outlet nozzles NI (from BWRVIP-241 model) = 1977 psi p = RPV normal operating pressure (psi) r = RPV inner radius (inch) . .. .

/ RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION FOR RELIEF REQUEST 31S1-14 Enclosure 2 Page 5 of 8 t = RPV wall thickness (inch) r, = Nozzle inner radius (inch) ro= Nozzle outer radius (inch)

Table 3 below summarizes these results.

......... . .. T. able 3 Outlet Nozzles (NI)

Cno09 CRPV

( p r t(n* r r0 Criteria (4)

<1.15 Criteria (5)

<1.59 1977 16171 102 12 9.5 10.8 15.4 0.84 1.51

___ 1~0 7__ _ _ _ _ _

Came o.,V oi Inlet Nozzles (N2) t r,.o<1.16 <1.47 1637 19332 102 12 .5 5.8 10.0 0.71 1.25 0 7 ____ d I_ __ _ __ _ _ _ __ _

Based upon the above information, the RRC Inlet and Outlet nozzle-to-vessel shell welds and nozzle inner radii sections identified in Table I meet the SE criteria and therefore Code Case N-702 is applicable. In addition, the reactor vessel is low alloy steel plate specification SA-533 grade B class I. The nozzle is low alloy steel forging specification SA-508 class 2. The weld metal used in the welds specified in Table I is carbon/low alloy steel and the welds are outside the beltline fluence region. Therefore, use of Code Case N-702 provides an acceptable level of quality and safety pursuant to 10 CFR 50.55a(a)(3)(i) for all RRC Inlet and Outlet nozzle-to-vessel shell welds and nozzle inner radii sections identified in Table 1.

6. Duration of ProDosed Alternative The duration of this request is for the third inservice inspection interval ending December 12, 2015.

7. preedrits None.

8. References

1. BWRVIP-108NP: BWR Vessel and InternalsProject, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water ReactorNozzle-to-

/ RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION FOR RELIEF REQUEST 31S1-14 Enclosure 2 Page 6 of 8 Vessel Shell Welds and Nozzle Blend Radii. EPRI, Palo Alto, CA: 2007.

1016123.

2. BWRVIP-241: BWR Vessel and Internals Project,ProbabilisticFracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to- Vessel Shell Welds and Nozzle Blend Radii. EPRI, Palo Alto, CA: 2010. 1021005.

3. ASME Boiler and Pressure Vessel Code,Section XI, "Rules Jor Inservice Inspection of Nuclear Power Plants," 2001 Edition through 2003 Addenda.

4. ASME Boiler and Pressure Vessel Code, Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds,Section XI, Division 1," February 20, 2004.

5. Matthew A. Mitchell, Office of Nuclear Reactor Regulation, to Rick Libra, BWRVIP Chairman, 'Safety Evaluation of Proprietary EPRI Report, 'BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-totVessel Shell Welds and Nozzle Inner Radius (BWRVIP-1 08)', December 19, 2007.

6. Sher Bahadur, Office of Nuclear Reactor Regulation, to Dennis Madison, BWRVIP Chairman, 'Safety Evaluation of the Boiling Water Reactor Vessel Internals Project (BWRVIP-241) Report, 'Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-To-Vessel Shell Welds and Nozzle Blend Radii (TAC NO. ME6328)* April 19, 2013.

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