View source for LR-N14-0147, Response to Request for Additional Information Request for Relief SC-I4R-140 from American Society of Mechanical Engineers Boiler & Pressure Vessel Code, Section XI Requirement Re the Reactor Vessel Head Flange Seal Leak..

{{Adams
| number = ML14161A444
| issue date = 06/10/2014
| title = Response to Request for Additional Information Request for Relief SC-I4R-140 from American Society of Mechanical Engineers Boiler & Pressure Vessel Code, Section XI Requirement Re the Reactor Vessel Head Flange Seal Leak..
| author name = Duke P
| author affiliation = PSEG Nuclear, LLC
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000272, 05000311
| license number = DPR-070, DPR-075
| contact person = 
| case reference number = LR-N14-0147, TAC MF3898, TAC MF3899
| document type = Letter type:LR
| page count = 12
| project = TAC:MF3898, TAC:MF3899
| stage = Response to RAI
}}

=Text=
{{#Wiki_filter:PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, NJ 08038-0236 PSEG Nuclear LLC 10 CFR 50.55a JUN 10 2014 LR-N14-0147 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Salem Generating Station Units 1 and 2 Renewed Facility Operating License Nos. DPR-70 and DPR-75 NRC Docket Nos. 50-272 and 50-311

==Subject:==
Response to Salem Nuclear Generating Station, Units 1 and 2 - Request for Additional Information Regarding Request for Relief SC-14R-140 From American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section X I Requirement Regarding the Reactor Vessel Head Flange Seal Leak Detection Piping (TAC Nos. MF3898 and MF3899)

==References:==
: 1) Letter from Paul Duke (PSEG Nuclear LLC) to USNRC, "Request for Relief from System Pressure Test Pressurization Requirements on Class 1 and Class 2 Reactor Vessel Flange 0-ring Leak Detection System, " dated April 8, 2014 (ADAMS Accessioq No. ML14098A308)
: 2) Letter from USNRC to Thomas Joyce (PSEG Nuclear LLC), "Salem Nuclear Generating Station, Units 1 and 2 - Request for Additional
--- -- ---- - - - ---- -lnformation -Regarding- Request-for -RelietS       - e-14R"'1-4G- From- America
                                                                                                      - n Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI Requirement Regarding the Reactor Vessel Head Flange Seal Leak Detection Piping (TAC Nos. MF3898 and MF3899), " dated June 2, 2014 (ADAMS Accession No. ML14135A291)
In Reference 1, PSEG Nuclear LLC (PSEG) submitted relief request SC-14R-140 for Salem Generating Station Units 1 and 2, requesting relief from specific requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, Rules for lnservice Inspection of Nuclear Power Plant Components.
In Reference 2, the NRC requested additional information. Attachment 1 to this letter provides PSEG's response.
There are no regulatory commitments contained in this letter.

JUN 1 ()'2014 LR-N14-01 47                                                         10 CFR 50. 55a Page 2 Should you have any questions concerning this matter, please contact Mr. Brian Thomas at 856-339-2022.
Sincerely, /1)
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Paul R. Duke, Jr.
Manager - Licensing PSEG Nuclear LLC  :       Response to Request for Additional Information for Relief Request SC-14R-1 40 :        Excerpt from Drawings 205201 (Unit 1) and 205301 (Unit 2) cc:    W. Dean, Administrator, Region I, USNRC NRC Senior Resident Inspector, Salem J. Lamb, Project Manager, Salem, USNRC P. Mulligan, Manager IV, NJBNE L. Marabella, Corporate Commitment Tracking Coordinator T. Cachaza, Salem Commitment Tracking Coordinator LR-N14-0147 Attachment 1 Response to Request for Additional Information for Relief Request SC-14R-140 Page  1 of 7 LR-N14-0147 RESPONSE TO REQUEST FOR ADDITIONAL INFORMAT ION RELATED TO REQUEST FOR RELIEF SC-14R-140 FROM AMERICAN SOCIETY OF MECHANICAL ENGINEERS BOILER AND PRESSURE VESSEL CODE, SECTION XI REQUIREMENTS REGARDING THE REACTOR VESSEL HEAD FLANGE SEAL LEAK DETECTION P IPING SALEM NUCLEAR GENERATING STATION, UNITS 1 AND 2 DOCKET NOS. 50-272 AND 50-311 By letter dated April 8, 201 4, Public Service Enterprise Group Nuclear LLC (PSEG or the licensee) requested relief from the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code), Section X I, for pressure testing the reactor vessel head flange seal leak detection piping at Salem Nuclear Generating Station, Units 1 and 2 (Salem 1 and 2), for the duration of the fourth 1 0-year inservice inspection (lSI) interval.
Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(a)(3)(ii), the licensee requested to use an alternative on the basis that complying with the system leakage test that is required by the ASME Code, Section XI, Table IWB-25001-1, Examination Category B-P; and Table IWC-2500-1 , Examination Category C-H would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the submittal by PSEG and has determined that additional information is needed as set forth below.
REQUEST FOR ADDITIONAL INFORMATION
: 1. Provide a detailed diagram of the reactor pressure vessel (RPV) head flange seal leak-off lines. _ClearlY' identifY' and discuss the boundar)' between t.._._.he ASME Code Class 1 and Class 2 piping.

===Response===
The boundary between Class 1 (specification 44A) 3/4" piping and Class 2 (specification 33B) 3/8" tubing is isolation valves 1 (2)R-C2 and 1 (2)RC3.
The Class 2 tubing extends to air operated valve 1(2)RC4, the tubing at the outlet of 1(2)RC4 becomes non-safety (specification 53E). See Attachment 2 for excerpts of Salem Unit 1 drawing 205201 Sheet 1 and Unit 2 drawing 205301 Sheet 1 showing the RPV head flange leak-off lines.
Page 2 of 7

Attachment 1 LR-N14-0147
: 2. Is any part(s) of the Class 1 and Class 2 RPV flange leak-off piping insulated and inaccessible for the ASME Code required VT-2 visual examination? If yes, discuss how the licensee will perform the VT-2 visual examination of the insulated and inaccessible part(s) of the subject piping.

===Response===
For Salem Unit 1 , none of the Class 1 piping is insulated and for Salem Unit 2 only approximately two feet of piping, just upstream of isolation valves 2RC2 and 2RC3, is insulated for personnel burn protection. The Class 1 non-insulated piping is accessible for direct VT-2 examination except where it penetrates the bio-shield wall and the portion of piping closest to the reactor vessel flange as it passes through cold leg reactor pressure vessel nozzle inspection plugs (also known as sandboxes). The VT-2 examination of inaccessible areas will include inspection of the surrounding areas below the lines, at ends of penetration sleeves and at the bottom of the reactor vessel annular gap for evidence of leakage as permitted by ASME Section X I Code, 2004 Edition, no Addenda IWA-5241 and IWA-5242. Additionally, the portion of leak-off lines that is inaccessible inside the sandboxes will receive a VT-2 examination any time the sandboxes are opened for piping weld examination or gasket replacement, typically at least once during each lSI 10-year interval.
All of the Class 2 tubing is encased in insulated tubing trays but is readily accessible for VT-2 examination. Any leakage would be obseNed by evidence of leakage emanating from the tubing tray joints. All class 2 areas will be examined as permitted by ASME Section XI Code, 2004 Edition, no Addenda IWA-5242(b).
: 3. The NRC staff notes that in Section 5, Attachment 1 of RFA SC-14R-1 40, the
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                 "In lieu of the requirements of IWB-5222{b) and IWC-5220, a VT-2 visual examination of the Class 1 and Class 2 accessible portions of the leak detection system piping will be performed while the system is subjected to the static pressure from the head of water when the reactor cavity is filled to its -normal refueling water level for at least four hours. . . . "
The ASME Code requirement for Class 1 piping in IWB-2500, Table IWB-2500-1 , Examination Category B-P, states that the system leakage test shall be conducted in accordance with IWB-5220 and the associated VT-2 visual examination according to IWA-5240, each refueling outage. The system leakage testing requirement in IWB-5220 consists of two subparagraphs, IWB-5221 pressure requirement and IWB-5222 boundary requirement. The pressure requirement in IWB-5221 consists of two subparagraphs, IWB-5221 (a) and IWB-5221 (b). The boundary Page 3 of 7 LR-N14-0147 requirement in IWB-5222 consists of two subparagraphs, IWB-5222(a) and IWB-5222(b).
It appears that the licensee is only requesting relief from the boundary requirement of IWB-5222(b) during system leakage testing performed at or near the end of each inspection interval for Class 1 segment of the subject piping.
: a. Discuss whether the licensee is also asking relief from IWB-5221(a), IWB-5221(b), and IWB-5222(a) in addition to IWB-5222(b).

===Response===
In addition to IWB-5222(b), relief is also being requested from IWB-5221(a), IWB-5221(b) and IWB-5222(a) for the Class 1 portion of the leak-off line piping. In lieu of the pressure not less than the pressure corresponding to 100% rated reactor power, PSEG requests an alternate pressure test to be conducted at ambient conditions after the refueling cavity has been filled to its minimum water level for refueling operations of 23 feet above the top of the RPV flange for at least four hours. A static pressure of approximately 1 0 psig is expected to be experienced at the top of the RPV flange with a minimum of 23 feet of borated water above the flange. The alternative test will be performed each refueling outage with the isolation valves 1(2)RC2 and 1(2)RC3 in the open position to allow concurrent Class 2 portions of the line to be tested.
: b. Has the licensee performed the system leakage test of Class 1
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refueling outage? If no, discuss whether the licensee is asking relief from the system leakage testing requirements of IWB-5221(a), IWB-5221(b), and IWB-5222(a) for each refueling outage.

===Response===
All the accessible Class 1 portions of the leak-off line piping have been examined as required by IWB-5222(a) every outage in the past refueling outages, although the pressure requirement of IWB-5221(a) has not been achieved.
The ASME Code requirement for Class 2 piping in IWC-2500, Table IWC-2500-1, Examination Category C-H, states that the system leakage testing shall be conducted in accordance with IWC-5220 and the associated VT-2 visual examination according to IWA-5240, each inspection period. The Page 4 of 7 LR-N14-0147 system leakage testing requirement in IWC-5220 consists of two requirements, the IWC-5221 pressure requirement and the IWC-5222 boundary requirement. The boundary requirement in IWC-5222 consists of two subparagraphs, IWC-5222(a) and IWC-5222(b).
It is not clear to the NRC staff from which subparagraph(s) the licensee is asking relief for Class 2 segment of the subject piping.
: c. In light of IWC-5221, IWC-5222(a), and IWC-5222(b), discuss specific subparagraph(s) of IWC-5220 for which the relief was requested.

===Response===
Relief is being requested only from IWC-5221 pressurization requirements for the Class 2 portion of the leak-off line tubing. In lieu of testing at Reactor Coolant System (RCS) pressure, PSEG requests an alternate test to be conducted at ambient conditions after the refueling cavity has been filled to its minimum water level for refueling operations of 23 feet above the top of the RPV flange for at least four hours. A static pressure of approximately 1 0 psig is expected to be experienced at the top of the RPV flange with a minimum of 23 feet of borated water above the flange. This test will be performed each refueling outage in conjunction with the Class 1 portion of the line. The normal operating pressure of this system is approximately 5 psig but in the event of inner RPV flange 0-ring leakage the pressure could reach RCS pressure. The pressure retaining boundary requirements in subparagraph IWC-5222(a) will be achieved during alternative pressure test and sub paragraph IWC-5222(b) does not apply.
: 4. For justifications that the structural integrity and leak tightness of the Class 1 and Class 2 RPV flange leak-off lines will be reasonably ensured without compliance with the ASME Code system leakage testing requirements, discuss whether there has been any plant specific, fleet, and industry operating experience regarding potential degradation of the subject lines due to known degradation mechanisms that would lead to leakage.

===Response===
There is no site related operating experience (OE) regarding degradation of these RPV flange leak-off lines. During Salem Unit 1 Cold Leg Alloy 600 weld mitigation using Mechanical Stress Improvement, a portion of these lines was removed and replaced due to tooling interferences.
Seabrook Station has experienced a through-wall leak on a leak-off line in the normally inaccessible area under the reactor cavity seal ring as Page 5 of 7 LRN14-0147 documented in their March 24, 2014 relief request (ADAMS Accession No. ML14086A017). This leak was detected during performance of Alloy 600 bare metal visual examinations and the affected pipe section was replaced. Based on an OE search, no additional degradation of a leak-off line was identified.
: 5. Discuss leakage detection capabilities at the plant, or any measure(s) that would be taken, to monitor and identify leakage in an unlikely event of a through wall leak in the RPV flange seal leak-off lines concurrent with leak or failure of the RPV flange 0-ring seal during normal operation.

===Response===
The RPV flange 0-ring leakage detection system at Salem consist of RPV flange leak-off line temperature indication (TI401) which is monitored in the control room and is logged on a shiftly basis in accordance with procedure S1(2). 0P-DL.ZZ-0003 "CONTROL ROOM LOG - MODES 1 -
4". In the event a temperature increase is detected in the RPV flange leak-off line, isolation valve 1(2)RC2 would be closed to isolate the inner 0-ring from the reactor coolant drain tank (RCDT) and 1(2)RC3 would be opened to line up the outer 0-ring leak-off line to the RCDT. Since the RCDT collects both reactor coolant system (RCS) and non-RCS in leakage, all leakage collected into RCDT is considered unidentified leakage subject to the limits of Unit 1 Technical Specification (TS) 3.4.6. 2.b and Unit 2 TS 3.4.7.2.b. In the event reactor coolant system unidentified leakage exceeds 1 gallon per minute, TS require the leakage rate to be restored to within limits within 4 hours. Otherwise the plant is required to be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
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through wall leak to the containment atmosphere concurrent with a leak or failure of the RPV flange 0-ring seal, this leakage would be detected by the RCS leakage detection systems described in Salem Unit 1 Technical Specification (TS) 3.4. 6.1 and Salem Unit 2 TS 3.4.7.1. The RCS leakage detection systems consist of the following:
* the containment atmosphere particulate radioactivity monitoring system
* the containment pocket sump level monitoring system, and
* either the containment fan cooler condensate flow rate or the containment atmosphere gaseous radioactivity monitoring system.
The above RCS leakage detection systems have overhead alarms in the control room. In the event of increasing radiation levels associated with the radiation monitors, procedures direct the operators to investigate the Page 6 of 7

Attachment 1 LR-N14-01 47 source of the leakage. For the containment sump and containment fan cooler condensate flow, the alarm response procedures direct the performance of an RCS leak rate calculation. The RCS leak rate procedure provides guidance for identifying the source of the RCS leakage.
: 6. The licensee in the last paragraph of Section 4, Attachment 1 of RFA SC-14R-1 40 stated, in part, "The configuration of the leak-off fine also precludes pressurizing the line externally with the head installed. . . . The time and radiation exposure associated with removing and reinstalling the closure head, replacing the outer 0-ring and re-cleaning of the vessel flange mating surface prior to head installation would be an undue hardship. . . . "
For the above mentioned hardship on the basis of radiation exposure, provide an estimate for roentgen equivalent man exposure with consideration of an as low as reasonably achievable, if the licensee performs the ASME Code required system leakage test with the vessel head installed.

===Response===
The radiation exposure associated with performing the RPV flange Leak off line pressure test with head installed using an external pressure source is estimated as follows:
* lifting and moving RPV head from head storage stand to vessel is estimated to be 1 .483 roentgen equivalent man (REM);
* assembling RPV head for pressure testing is estimated to be 2.220
__ ____________        HEM;
* disassembling the RPV head to replace 0-rings is estimated to be 1.1 45 REM; and
* lifting and moving RPV head back to head storage stand is estimated to be 1 .483, The total estimated dose would be 6. 331 REM per unit during each refueling outage. Over a ten-year lSI interval with six outages for two units, this would be 75.972 REM. Installation and removal of the RPV Head for the purposes of pressure testing the flange leak-leak off lines would result in significant dose which would be contrary to keeping dose as low as reasonably achievable (ALARA) in accordance with the requirements of 1 0 CFR 20.1101 (b).
Page 7 of 7 LR-N14-0147 Attachment 2 Excerpt from Drawings 205201 (Unit 1 ) and 205301 (Unit 2)
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