ML042740478
| ML042740478 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 11/02/2004 |
| From: | Dan Collins NRC/NRR/DLPM/LPD1 |
| To: | Bakken A Public Service Enterprise Group |
| Wunder G, NRR/DLPM, 415-1494 | |
| References | |
| TAC MC5710, TAC MC5711 | |
| Download: ML042740478 (5) | |
Text
November 2, 2004 Mr. A. Christopher Bakken, III President & Chief Nuclear Officer PSEG Nuclear - X15 P.O. Box 236 Hancocks Bridge, NJ 08038
SUBJECT:
SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2, CORRECTION TO ISSUANCE OF AMENDMENT NOS. 263 AND 245 (TAC NOS. MC5710 AND MC5711)
Dear Mr. Bakken:
On September 16, 2004, the Nuclear Regulatory Commission (NRC) issued Amendment Nos. 263 and 245 to Facility Operating License Nos. DPR-70 and DPR-75 for the Salem Nuclear Generating Station, Unit Nos. 1 and 2, respectively. These amendments consisted of changes to the Technical Specifications (TSs) in response to your application dated July 29, 2002, as supplemented by letters dated March 28, and May 1, 2003, and August 20, 2004.
Based on discussions between your staff and the NRC on September 23, 2004, we are reissuing a corrected copy of pages 11 and 22 of the Safety Evaluation dated September 16, 2004. Omitted text from the Fuel Handling Building Closure administrative requirements is re-inserted on page 11. Additionally, Page 22 is corrected to remove an incorrect reference to a footnote in TS 3.9.4 and correct the required containment closure time from 30 minutes to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
Sincerely,
/RA/
Daniel Collins, Sr. Project Manager, Section 2 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-272 and 50-311
Enclosures:
As stated cc w/encls: See next page
Salem Nuclear Generating Station, Unit Nos. 1 and 2 cc:
Mr. Michael H. Brothers Vice President - Site Operations PSEG Nuclear - X15 P.O. Box 236 Hancocks Bridge, NJ 08038 Mr. John T. Carlin Vice President - Nuclear Assessments PSEG Nuclear - N10 P.O. Box 236 Hancocks Bridge, NJ 08038 Mr. Patrick S. Walsh Vice President - Eng/Tech Support PSEG Nuclear - N28 P.O. Box 236 Hancocks Bridge, NJ 08038 Ms. Christina L. Perino Director - Licensing & Nuclear Safety PSEG Nuclear - N21 P.O. Box 236 Hancocks Bridge, NJ 08038 Jeffrie J. Keenan, Esquire PSEG Nuclear - N21 P.O. Box 236 Hancocks Bridge, NJ 08038 Ms. R. A. Kankus Joint Owner Affairs PECO Energy Company Nuclear Group Headquarters KSA1-E 200 Exelon Way Kennett Square, PA 19348 Lower Alloways Creek Township c/o Mary O. Henderson, Clerk Municipal Building, P.O. Box 157 Hancocks Bridge, NJ 08038 Dr. Jill Lipoti, Asst. Director Radiation Protection Programs NJ Department of Environmental Protection and Energy CN 415 Trenton, NJ 08625-0415 Brian Beam Board of Public Utilities 2 Gateway Center, Tenth Floor Newark, NJ 07102 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Senior Resident Inspector Salem Nuclear Generating Station U.S. Nuclear Regulatory Commission Drawer 0509 Hancocks Bridge, NJ 08038 Mr. Carl J. Fricker Plant Manager PSEG Nuclear - N21 P.O. Box 236 Hancocks Bridge, NJ 08038
November 2, 2004 Mr. A. Christopher Bakken, III President & Chief Nuclear Officer PSEG Nuclear - X15 P.O. Box 236 Hancocks Bridge, NJ 08038
SUBJECT:
SALEM NUCLEAR GENERATING STATION, UNIT NOS. 1 AND 2, CORRECTION TO ISSUANCE OF AMENDMENT NOS. 263 AND 245 (TAC NOS. MC5710 AND MC5711)
Dear Mr. Bakken:
On September 16, 2004, the Nuclear Regulatory Commission (NRC) issued Amendment Nos. 263 and 245 to Facility Operating License Nos. DPR-70 and DPR-75 for the Salem Nuclear Generating Station, Unit Nos. 1 and 2, respectively. These amendments consisted of changes to the Technical Specifications (TSs) in response to your application dated July 29, 2002, as supplemented by letters dated March 28, and May 1, 2003, and August 20, 2004.
Based on discussions between your staff and the NRC on September 23, 2004, we are reissuing a corrected copy of pages 11 and 22 of the Safety Evaluation dated September 16, 2004. Omitted text from the Fuel Handling Building Closure administrative requirements is re-inserted on page 11. Additionally, Page 22 is corrected to remove an incorrect reference to a footnote in TS 3.9.4 and correct the required containment closure time from 30 minutes to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
Sincerely,
/RA/
Daniel Collins, Sr. Project Manager, Section 2 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-272 and 50-311
Enclosures:
As stated cc w/encls: See next page DISTRIBUTION PUBLIC D. Collins G. Matakas, RGN-I S. LaVie ACRS PDI-2 Reading C. Raynor M. Blumberg OGC G. Miller D. Cullison ADAMS Accession Number: ML042740478 OFFICE PDI-2/PE PDI-2/PM PDI-2/LA PDI-2/SC(A)
NAME GMiller GWunder CRaynor DCollins DATE 10/28/04 10/28/04 10/28/04 10/29/04 OFFICIAL RECORD COPY to close the equipment hatch opening (i.e., restrict air flow out of the containment), such as a refueling hatch closure device, is staged at the work area along with the necessary installation tools.
e.
A sufficient number of personnel are designated and available with the responsibility for expeditious closure (within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) of the containment equipment hatch opening following the FHA.
5.
If containment closure would be hampered by an outage activity, compensatory actions will be developed.
6.
Either the Containment Purge system or the Auxiliary Building Ventilation System with suction from the containment atmosphere, with associated radiation monitoring will be available whenever movement of irradiated fuel is in progress in the containment building and the equipment hatch is open. If for any reason, this ventilation requirement can not be met, movement of fuel assemblies within the containment building shall be discontinued until the flow path(s) can be reestablished, or close the equipment hatch (or an equivalent closure device is installed) and personnel airlocks. Periodic verification (once per shift) of this administrative control will ensure that air flow will be directed from containment to the Auxiliary Building or the Plant Vent where continuous monitoring will be in effect thus minimizing the potential for unmonitored releases out the open containment hatch following the FHA.
7.
Personnel responsible for Containment Building Closure shall be trained and knowledgeable in using the procedure for executing containment closure. Walkdowns should be considered to demonstrate the closure capability including compensatory actions in the event of loss of electrical power.
Fuel Handling Building Closure:
The following requirements shall be maintained to ensure defense-in-depth.
Closure Controls are in effect during operations within the Fuel Handling Building involving movement of irradiated fuel assemblies.
1.
The Fuel Handling Building doors shall be maintained closed except for normal entry and exit unless a designated person is available to close the open door(s) should a[n] FHA occur within the Fuel Handling Building.
2.
The FHAVS [fuel handling area ventilation system], with associated radiation release monitoring will be available for the release flow path. If for any reason operation of the fuel handling area ventilation system flow path must be discontinued and the fuel building is open to the outside environment, fuel movement within the fuel handling building shall be l
discontinued until the flow path can be reestablished or until the openings l
to the outside environment are closed.
l The NRC staffs position is that the FHA is the only event during CORE ALTERATIONS that is postulated to result in fuel damage and radiological release. This position is documented in TSTF-51. Under the revised FHA analyses, the potential for a radioactive release only exists during the movement of fuel within the containment or the SFP. The proposed change to the applicability statement leaves the LCO and required actions applicable during activities which could result in an FHA with fuel damage and radiological release. Therefore, the deletion of CORE ALTERATIONS is acceptable.
3.3.6 TS 3/4.9.9 Delete TS 3/4.9.9, Refueling Operations, Containment Purge and Pressure-Vacuum Relief Isolation System, into TS 3.9.4, Containment Building Penetrations. This proposed change will implement consistency with the ITSs. Part of SR 4.9.9 will be relocated to TS 3/4.9.4 such that the verification that each containment purge isolation valve actuates closed on a manual actuation signal at least once-per-18-months would be retained.
TS 3/4.9.9 Containment Purge and Exhaust Isolation System, the entire section, would be deleted and the page is marked intentionally left blank.
Removal of automatic isolation of the purge system is partially compensated by implementation l
of administrative controls to close the containment within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> using designated personnel l
after an FHA. The staff has determined that these administrative controls provide an important element of defense-in-depth, and with these administrative controls in place, it will assure that the licensee will manage the consequences of an FHA in a manner that will afford adequate protection to the public. As such, the staff finds that the removal of automatic isolation is acceptable with the addition of administrative controls to effect closure. The staff reviewed the request to delete TS 3/4.9.9 and agrees, as further explained below, that the TS section may be deleted since the purge isolation system is not credited in the DBA analysis.
The NRC staff notes that the containment purge and exhaust isolation system (CPEIS) is not a form of instrument or a process variable, design feature or operational restriction that is an initial condition of a DBA or transient that either assumes the failure of, or presents a challenge to, the integrity of a fission product barrier. Therefore, Criterion 1, and 2 of 10 CFR 50.36(c)(2)(ii) do not apply.
The licensee has shown, on the basis of their FHA design-basis analysis that operation of the CPEIS is not required to satisfy the dose values of 10 CFR 50.67. Thus, the system is not on the primary success path for a DBA. As such, Criterion 3 of 10 CFR 50.36(c)(2)(ii) does not apply.
Since the purge isolation system is not credited in the DBA analysis, it is not considered to be risk-significant to public health and safety by either operating experience or probabilistic safety assessment; therefore, Criterion 4 of 10 CFR 50.36(c)(2)(ii) does not apply. The staff finds that the purge isolation system does not meet the criteria contained in 10 CFR 50.36(c)(2)(ii) and its removal is, therefore, acceptable.