LR-N09-0103, Submittal of 2008 Summary of Revised Regulatory Commitments

Submittal of 2008 Summary of Revised Regulatory Commitments

ML091340091
Person / Time
Site:	Salem
Issue date:	05/05/2009
From:	Braun R Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LR-N09-0103
Download: ML091340091 (7)

Text

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, Ne'" Jersey 08038-0236 MAY -0 Zoos LR-N09-0103 U. S. Nuclear Regulatory Commission o PSEG NuclearLLC ATTN: Document Control Desk Washington, DC 20555-001 Salem Generating Station, Units 1 and 2 Facility Operating License Nos. DRP-70 and DRP-75 NRC Docket Nos. 50-272 and 50-311

Subject:

2008 Summary of Revised Regulatory Commitments In accordance with the Nuclear Energy Institute (NEI) process for managing Nuclear Regulatory Commission (NRC) commitments and associated NRC notifications, PSEG Nuclear LLC (PSEG) submits this correspondence to discuss commitments that were changed and not reported by other means during 2008.

The attached commitments were evaluated in accordance with the requirements of PSEG's Regulatory Commitment Change Process, which is consistent with the guidance in NEI 99-04, "Guideline for Managing NRC Commitments." Additional documentation is available for your review.

There are no new commitments in this letter.

If there are any questions, please contact Howard Berrick at 856-339-1862.

Sincerely, Robert C. Braun Site Vice President - Salem Attachments (1) 95-2168 REV. 7/99

Document Control Desk MAY.o02009 LR-N09-0103 2 C Mr. Samuel Collins, Administrator - Region 1 U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19046 Mr. R. Ennis, Project Manager - Hope Creek and Salem U. S. Nuclear Regulatory Commission Licensing Project Manager - Salem Mail Stop 08B1 Washington DC 20555-001 USNRC Senior Resident Inspector - Salem (X24)

Mr. P. Mulligan, Manager, IV Bureau of Nuclear Engineering PO Box 415 Trenton, NJ 08625

Attachment LR-N09-0103 Page 1 of 5 Revised Commitment Description Justification For Change Original Commitment: Equivalent closure devices for the equipment As approved in Amendments 251/232 the Fuel Handling Accident (FHA)

The design and fabrication of hatch in accordance with Technical analysis was revised using the Alternate Source Term (AST) methodology.

the Outage Equipment Hatch Specification (TS) 3/4.9.4 will be designed The revised AST FHA assumed a 100% release over a 2-hour period (OEH) at Salem I and 2 is in and fabricated in accordance with commercial through the open equipment hatch. During the approval of Amendments accordance with the ASME grade requirements to serve as a ventilation 263/245 to allow the containment equipment hatch to be open during the Boiler and Pressure Vessel barrier and are not required to be designed movement of irradiated fuel within the containment, PSEG committed to (B&PV) Code,Section VIII, and fabricated in accordance with ASME establishing the closure of the containment equipment hatch within 1-hour 1989 edition or later. Section VIII. As stated in Salem TS 3/4.9.4 following an FHA to provide additional defense in depth.

bases, 'any equivalent closure device used to In 1998 when the Nuclear Regulatory Commission (NRC) approved Source Document: satisfy the requirements of TS 3/4.9,4.a will be Amendments 217/199 to allow an equivalent closure device to be used in Salem Unit 1/2 Amendments designed, fabricated, installed, tested, and lieu of the inner equipment hatch, PSEG stated that this equivalent closure 217/199 and 263/245 Safety utilized in accordance with established device would be designed in accordance with the ASME B&PV Code, Evaluation Reports (SER) (as procedures to ensure that the design Section VIII, 1989 edition or later. As stated in the SER for Amendments corrected by letters LR-N04- requirements for the mitigation of a fuel 217/199 and the current Salem TS 3/4.9.4 basis, 'any equivalent closure 0359 dated March 11, 1999, handling accident (FHA) during refueling device used to satisfy the requirements of TS 3/4.9.4.a will be designed, and LR-N07-0288 dated operations are met." The closure of the fabricated, installed, tested, and utilized in accordance with established November 19, 2007), ventilation barrier is to provide additional procedures to ensure that the design requirements for the mitigation of a fuel defense-in-depth by closing the equipment handling accident during refueling operations are met."

hatch release pathway although the FHA

Reference:

analysis performed using the AST, approved Prior to the approval of the AST analysis for the FHA on October 10, 2002, CR 80094905 by Amendment 251/232, assumed a 100% the closure of the containment was performed to mitigate the consequences release over 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> through the open of the FHA by limiting the release pathway.

Date of Change: equipment hatch. Since pressurization events However, with the implementation of the FHA AST analysis, closure of the 03/01/08 are unlikely during MODE 6 (refueling) containment is no longer an action to mitigate the consequences of the FHA.

operation, the closure requirement for the Although it is not credited in the mitigation of the FHA, PSEG committed in equivalent device need only be sufficient to the approval of Amendments 263/245 to have the ability to close the provide an atmospheric ventilation barrier to containment within 1-hour following an FHA to add additional defense in restrict radioactive material released from a depth.

fuel element rupture during refueling operation. As was recognized in the SER for Amendments 217/199, pressurization events are unlikely during Mode 6 (refueling) operations; the closure requirement for the equivalent device need only be sufficient to provide an atmospheric ventilation barrier.

Attachment LR-N09-0103 Page 2 of 5 Revised Commitment Description Justification For Change Calculation S-2-RC-MDC-2151 was performed to evaluate the containment environment during modes 5 and 6 following a loss of RHR event. This calculation concluded that only an atmospheric-pressure closure device (ventilation barrier) is necessary to meet the containment closure commitment following a FHA.

The closure device can be fabricated from metal or flexible materials that will prevent gross ventilation flow through the equipment hatch. It is capable of being closed within one hour (or prior to core boiling) and is designed to accommodate the tracks and other materials that may be installed through the hatch.

In view of the above, an equivalent closure device need only be an atmospheric ventilation barrier to restrict radioactive material release. In the case of the FHA, containment pressure would be atmospheric pressure to the extent that an equivalent barrier would not need to resist internal containment pressure.

Attachment LR-N09-0103 Page 3 of 5 Revised Commitment Description Justification For Change Original Commitment: PSEG is deleting the explicit commitment to WCAP-1 1394 is not a Salem specific report, but a generic Westinghouse PSEG will use the NRC WCAP-1 1394. topical report that is an inherent part of the Westinghouse reload safety approved methodology in analysis process. Per Westinghouse, this WCAP is referenced in their WCAP-1 1394-P-A for each dropped rod protection analysis guidance, which is utilized for Salem.

fuel cycle to ensure the minimum departure from The WCAP is also referenced in the Salem UFSAR Section 15.2.3, "the nucleate boiling ratio (DNBR) transient response analysis, nuclear peaking factor analysis, and is maintained above the DNBR performance of the departure from nucleate boiling (DNB) design basis safety limit. confirmation are performed in accordance with the methodology described in Reference 15 [WCAP-1 1394]. Note that the analysis does not take credit for Source Document: the power-range negative flux rate trip." Salem implemented this WCAP as LR-N06-0035 the dropped rod event design basis in the 1990s and from that time has not taken credit for the negative flux rate trip (NFRT) in the accident analyses.

NRC Amendment SER 278/261 dated March 19, 2007 The current PSEG Nuclear procedures associated with core reload and Salem Updated Final Safety vendor reload analyses are: NF-AA-100 (Reload Control Procedure), NF-Analysis Report (UFSAR), AA-1 00-1001 (Core Reload and Cycle Management Configuration Changes Chapter 15 Using SAP), and NF-AP-100-7000 (Westinghouse NSSS Reload Design Control Implementation).

Reference:

70085988 In addition PSEG maintains a formally signed interface agreement with Westinghouse (NFS-0180, "Reload Design Interface Document for Salem Units 1 and 2"). These documents do not include any specific list of Chapter Date of Change: 15 events that are evaluated on a reload basis, or any specific references to 09/11/08 Westinghouse codes, methods, or topical reports.

However, the procedures and interface documents do cover specific reload milestones and information transmittals affecting the accident analyses.

These include the Reload Safety and Licensing Checklist (RS&LC) and Reload Safety Analysis Checklist (RSAC). The RS&LC is a document the customer prepares and sends to Westinghouse that covers any anticipated plant, operating and licensing changes to be implemented for the upcoming cycle. The RS&LC is included along with other reload documentation in a

Attachment LR-N09-0103 Page 4 of 5 Revised Commitment Description Justification For Change transmittal referred to as the Reload Design Initialization (RDI).

Westinghouse provides the RSAC to the customer; it lists those parameters that most significantly affect the accident analysis for the reload'design, along with the current limits utilized in the analyses.

Implementation of any new Westinghouse method is specifically called out in the above procedures: NF-AA-1 00 Step 4.83, "if required then ensure a 50.59 review is performed for any identified changes such as methodology changes..,"; NF-AP-100-700 Step 4.8.10, "review and comment to Westinghouse on the draft Safety Assessment and RSAC current limits.

Request calculations from Westinghouse, and review for (1) any non-routine analyses, (2) any analyses being performed for the first time, and (3) and analyses that had a methodology change."; NF-AP-1 00-700 Attachment 3 (Reload Design Finalization Checklist), #8, "are all methods and models used in the reload design fully accepted by PSEG and the NRC"; NFS-0180, "methods and computer code for nuclear design are those developed or adapted by Westinghouse. These are described in the methods, codes and design manuals. Any exceptions to this use must be agreed upon and documented."

Although the commitment to use WCAP-1 1394 (for the dropped rod event supporting NFRT removal) is not explicitly included in the Nuclear Fuels reload procedures, it is an inherent part of the Westinghouse accident analysis design basis. Since the above procedures and interface document with Westinghouse explicitly cover any changes to their methods that can affect Salem reload design and licensing requirements, the necessary commitment is already implicitly covered and there is no need for any additional explicit commitment in the noted Nuclear Fuel reloads design procedures.

Attachment LR-N09-0103 Page 5 of 5 Revised Commitment Description Justification For Change Original Commitment: PSEG provided an "initial" complete formal The present training program is covered under the systems approach to PSEG to provide "... a training and a qualification program for training [as defined in 10 CFR 55.4] as outlined in 10 CFR 50.120 (b). The complete formal training Emergency Diesel Generator (EDG) training programs incorporate the instructional requirements necessary to program be implemented for all maintenance for applicable personnel who provide qualified personnel to operate and maintain the facility.

the mechanical and electrical work on and supervise EDG maintenance maintenance, quality control, activities.

and operating personnel, Continuing (future) training will be decided including supervisors who will using the systems approach to training lAW be responsible for the 10CFR 50.120 and the training process maintenance and availability of description.

the diesel generators.

The depth and quality of this training program shall be at least equivalent to that of training programs normally conducted by major diesel engine manufacturers." This commitment was implemented.

Source Document:

Salem Unit 2 SER 8.3.4 (d)

Supplement 5

Reference:

70077701 Date of Change:

03/01/08