NL-07-082, Supplemental Submittal Regarding Relief Request RR-02, Proposed Alternative for Regenerative Heat Exchanger Welds, and RR-05, Risk-Informed ISI (Relief from B-F and B-J Examination Requirements)
| ML072080472 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 07/19/2007 |
| From: | Jones T Entergy Nuclear Northeast |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-07-082, TAC MD4696, TAC MD4700 | |
| Download: ML072080472 (9) | |
Text
SEntergy Enteray Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB Buchanan, NY 10511-0249 T.R. Jones Licensing Manager Tel (914) 734 6670 July 19, 2007 Re:
Indian Point Unit 2 Docket No. 50-247 NL-07-082 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-P1-17 Washington, DC 20555-0001
SUBJECT:
Supplemental Submittal Regarding Relief Request RR-02, "Proposed Alternative for Regenerative Heat Exchanger Welds,"
and RR-05, "Risk-Informed ISI (Relief from B-F & B-J examination requirements)." (TAC Nos. MD4696 and MD4700)
- 1. Entergy Letter dated February 28, 2007, P.W. Conroy to Document Control Desk, ' 4 th Ten-Year Interval Inservice Inspection and Containment Inservice Inspection Program Plan at Indian Point Unit 2 (IP2)"
REFERENCES:
Dear Sir or Madam:
letter dated February 28, 2007 (Reference 1) Entergy Nuclear Operations, Inc. submitted the 4'. Ten-Year Interval Inservice Inspection and Containment Inservice Inspection Program Plan for the period March 1, 2007 through April 3, 2016 for IP2. Appendix B of the enclosure contained seven (7) relief requests. The NRC staff requested additional information via Teleconference on June 19, 2007 in order to complete its review of Relief Requests RR-02 and RR-05. The purpose of this letter is to provide the responses to the questions discussed at the Teleconference. Responses to NRC questions on RR-02 and RR-05 are provided in to this letter.
If you have any questions or require additional information, please contact Mr. T.R. Jones, Manager, Licensing at (914) 734-6670.
T. R(Jones Licensing Manager Indian Point Energy Center 4'/047
NL-07-082 Docket 50-247 Page 2 of 2 Attachments:
- 1.
Response for Additional Information Regarding Relief Request RR-02 and RR-05 (TAC Nos. MD4696, MD4700) cc:
Mr. John P. Boska, Senior Project Manager, NRC NRR DORL Mr. Samuel J. Collins, Regional Administrator, NRC Region I NRC Resident Inspector's Office, Indian Point 2 Mr. Paul Eddy, New York State Dept. of Public Service
ATTACHMENT 1 TO NL-07-082 Response for Additional Information Regarding Relief Requests RR-02 and RR-05 (TAC Nos. MD4696 and MD4700)
ENTERGY NUCLEAR OPERATIONS, INC.
INDIAN POINT NUCLEAR GENERATING UNIT NO. 2 DOCKET NO. 50-247
NL-07-082 Docket No. 50-247 Page 1 of 6 Response for Additional Information Regarding Relief Request RR-02, "Proposed Alternative for Regenerative Heat Exchanger Welds."
Responses to the questions discussed at June 19, 2007, Teleconference regarding Relief Request RR-02, "Proposed Alternative for Regenerative Heat Exchanger Welds." (TAC No.
MD4696) are as follows:
- 1.
Pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(a)(3)(i),
Request RR-02 proposes an alternative to the examination requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section Xl, Subarticle IWB-2500 on the basis that the alternative examination will provide an acceptable level of quality and safety. Request RR-02 specifically proposes that the provisions of ASME Code Case (CC) N-706, "Alternative Examination Requirements of Table IWB-2500-1 and Table IWC-2500-1 for PWR Stainless Steel Residual and Regenerative Heat Exchangers,Section XI, Division 1," be used as an alternative to the examination requirements of Table IWB-2500-1, Examination Categories B-B and B-D for the specified regenerative heat exchanger (RHX) 21 vessel welds.
CC N-706 states that the alternative examination requirements specified in Table I of the CC may not be applied to any heat exchanger nor to any heat exchanger design or configuration that has experienced a through wall leak. Note (1) from Table I of the CC states that the application of the alternative examination requirements of Table I of the CC is limited to those welds that are part of the as-received RHX assembly. The RHX assembly may be formed from multiple smaller heat exchanger subcomponents connected by sections of piping. According to Note (1), all of the smaller heat exchanger subcomponents and the connecting piping are considered to be within the boundary of the RHX assembly.
Section E of RR-02 states that a leak occurred in January 2004 at San Onofre Nuclear Generating Station (SONGS), Unit 3 in the letdown line exiting the RHX.
Please indicate whether the design or configuration of the RHX at SONGS 3 corresponds to that of RHX 21 at Indian Point 2. If the design or configuration of the RHX at SONGS 3 does correspond to the design or configuration of RHX 21 at Indian Point 2, please provide additional detail regarding how the leak that occurred in the SONGS 3 RHX letdown line was determined to be outside the scope of CC N-706, taking into consideration the provisions of Note (1) under Table I of the CC and the definition of RHX 21 at Indian Point 2 as an RHX assembly which includes all of the smaller heat exchanger subcomponents and the connecting piping.
NL-07-082 Docket No. 50-247 Page 2 of 6
Response
The event at SONGS Unit 3 identified a small leak that developed in a pipe weld on the letdown outlet side of the Regenerative Heat Exchanger. Vibration resulting from operation of Positive Displacement Charging Pumps with inadequate discharge pressure pulsation dampeners, and a locked spring can hanger was the primary cause. Initial inspections and system monitoring revealed that the charging line into the RHX and the letdown line out of the heat exchanger were vibrating at elevated levels. The leak was from the weld on the Letdown piping outlet at the bottom of the vertically oriented RHX.
The SONGS Unit 3 Regen Heat Exchanger is of different design, manufacturer, and configuration then at Indian Point Unit 2. The SONGS Unit 3 RHX is of Whitlock design with a vertical orientation.
The IP2 RHX is designed and manufactured by Atlas Industrial Mfg Corp., for Westinghouse, and is of horizontal configuration. In addition, IP2 installed suction stabilizers and discharge pulsation dampers on each of the charging pumps to reduce vibration-induced fatigue on the pump and associated piping.
- 2.
CC N-706, Table 1, Note (2) states that all welds under Item No. 1.10 of the table shall have received at least one volumetric examination and that the preservice or construction code volumetric examination may be used to meet this requirement.
The interpretation of this requirement is that, for a particular component, all items that would be subject to the alternative examination requirements of Table I must have received this volumetric examination.
Section E of RR-02 states that the pressure retaining welds in RHX 21 had received at least one volumetric preservice examination. Please verify that all welds in RX 21 that are within the scope of the proposed alternative had previously received this volumetric examination.
NL-07-082 Docket No. 50-247 Page 3 of 6 For all welds identified in RR-02 that have received a volumetric examination, please state whether these examinations included only the preservice examinations required by Subarticle IWB-2200 or additional successive inspections during previous Inservice Inspection (ISI) intervals required by Subarticle IWB-2400. Please discuss any relevant conditions that were found during these examinations.
Response
A pre-service volumetric Inspection was performed on all Regenerative Heat Exchanger Head-to-shell, Tube sheet-to-shell and nozzle-to-vessel welds no recordable indications noted. Additional successive examinations were performed on these welds during the 2 nd and 3 rd interval which ended in 2000 in accordance with IWB-2400 and approved Relief Requests. Eleven (11) were volumetrically examined while the remainders were surfaced examined with no recordable indications identified during the inspections.
- 3.
Please discuss the programs that are currently in place for Reactor Coolant System (RCS) leakage monitoring in the vicinity of RHX 21.
Response
Unidentified leakage is monitored in accordance with IP2 Improved Technical Specifications. Additionally, the containment atmosphere particulate radioactivity is monitored per Technical Specification requirements.
The heat exchanger will also continue to receive a system leakage test (Procedure 2-PT-R75) prior to startup after each refueling outage. During this system test, the components receive a visual (VT-2) examination. The corresponding piping and component supports will also continue to be inspected per the requirements of the Code, as they are not affected by this relief request.
NL-07-082 Docket No. 50-247 Page 4 of 6 Response for Additional Information Regarding Relief Request RR-05, "Risk-Informed ISI (Relief from B-F & B-J exam. requirements)."
Responses to the questions discussed at June 19, 2007, Teleconference regarding Relief Request RR-05, "Risk-Informed ISI (Relief from B-F & B-J exam. requirements)." (TAC No.
MD4700) are as follows:
- 1.
The newer versions of the American Society of Mechanical Engineers (ASME)
Code have reduced the exempted portions of high pressure safety injection piping from nominal pipe size (NPS) 4 to NPS 1.5 for volumetric examination. This reduction in exempted piping has caused other licensees to add ASME Class 1 pressurized water reactor (PWR) high pressure safety injection piping to the scope of their risk-informed inservice inspection (RI-ISI) programs, and to implement their chosen RI-ISI methodology to classify, risk-rank, and to select, as necessary, additional locations for the next ISl interval. Please describe how you treated this issue in your RI-ISI program for the fourth 10-year ISl interval when you updated your code of record from the 1989 edition to the 2001 edition with 2003 addenda.
Response
For the fourth 10-year ISI interval Indian Point Unit 2 included the high pressure safety injection piping with nominal pipe size >1.5 for volumetric examination in the scope of the Risk-Informed In-service inspection (RI-ISI) program. This is described in the original submittal for the Third ISI Interval (SER TAC No. MC0624).
- 2.
The submittal states that the number and locations of inspections have not changed and therefore the original changes in risk estimate remain valid. The change in risk estimate may change because of changes in the probabilistic risk assessment (PRA) and/or degradation mechanism even if the risk ranking does not change enough to require changing of the inspection program. Please explain why the previous changes in risk estimate remain valid considering changes that have been made to the PRA and degradation mechanism.
Response
The inputs from the updated probabilistic risk assessment (PRA) model were reviewed against inputs used in the original RI-ISI consequence assessment. The PRA inputs reviewed include initiating event conditional core damage probability and system unavailabilities (e.g. valve failure rates). The impact on the consequence ranking results is shown below.
NL-07-082 Docket No. 50-247 Page 5 of 6 2ne U*pde Ifomtaonl:
RiI4$Coequence Input" Impact Initiator IE Re*q COF CCDP Description
!CCDP Rank IE-T3 1.87E400 1.53E.06, 82E-D7 TU.RBINE TRIP WTH MAIN FEEDNATERAVAILABLE, 12E.O Medum None IE-T2 3.86E601 8.70E.07 2.26 EO06 LOSS OF MAIN FEEDWATER N/A IE-A
,.00E,06 1.58E.08 1315E-03 LARGE LOCA 6.07E003 High
- None, IE-t
-400E.05 1.16E07 2,91E.03 INTERMEDIATE LOCAK 3,02E'03 NHig None IEt-'
51E00ED4 3ZE.07' 7,80E04 -SMALLLOCA 2J81 E`4 High None IE-T4.
4.27E-.4 7.08E.09 1,6E.05 MAIN STEAM LINE BREAK INSIDE.CONTAINM ENT WNA' FOiure 'Ries HourlRate Yearly R e Description 5.36E-07 Per Hr 4.70E.03 Per Yr CHECKVALVE FAILS TO'REMAIN CLOSED 212E602 None 3.006EM Per Hr 2.63E.02 Per Yr AOV FAILS TO REMAIN'CLOSED.
.86E403 None 3.HOE-09 Per Hr 263E-05 Per Yr MANUAL VALVE FAILS TO' REMAIN CLOSED 8.12E04
- INone, 3A.E.03 Per Demand MOV FAILS TO CHANGE PSITION 1'59E-'03 None*
1.00E.03 Per Demand CHECKVALVE FAiLsT CLOSE*
77EO5 N one'k 1.00E.03 Per Demand AOV DOES NOTCLOSE e8.2E-04 None A (1)=: Scientech Report No. 17184C-2 Consequence Analysis ofClass-1 Piping in Suppot of CocdC*ase N,578, Indian Point 2 Nudear'Power PIntý (2)ý= Representatve of several sy,,emmspecific values (3) Per the above, several valve lure rates from the updateid PRA are higer wita thn that used in the RI.ISI application, Per Reference these values were sed in segments determiiedto be Med*um consequence rank (See 0-.). Assuch; the fo lowing bound any impact of these changes and ishs that the R.IS8
,consequdhce rank doesvnot change:
Upper Bound LOCA CCDP= 3.15E,03 Upr BoundValve Failure =3.00E03:
CdDPL, V elve=
9.45E-06 or a Medium conse nce rank
NL-07-082 Docket No. 50-247 Page 6 of 6 The results of this review concluded, there are no required changes to the RI-ISI consequence ranking due to the updated PRA input. That is, high consequence segments remained high and medium consequence segments remained medium.