LR-N04-0510, Change to Commitment Inservice Leak and Hydrostatic Testing
| ML043430160 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 11/23/2004 |
| From: | John Carlin Public Service Enterprise Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LR-N04-0510, TAC 88941, TAC MA1750 | |
| Download: ML043430160 (4) | |
Text
PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 NOV 2 3 2004 Nu clearLLC LR-N04-051 0 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001 CHANGE TO COMMITMENT INSERVICE LEAK AND HYDROSTATIC TESTING HOPE CREEK GENERATING STATION FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354
References:
- 1. NLR-N94050, "License Amendment Application: Inservice Leak and Hydrostatic Testing Exception," March 04, 1994
- 2. Amendment No. 69, Facility Operating License No. NPF-57, April 18, 1994 (TAC NO. 88941)
- 3. LR-N98219, "Request for Change to Technical Specifications:
Inservice Leak and Hydrostatic Testing Requirements," May 13,1998
In accordance with the Nuclear Energy Institute (NEI) process for managing Nuclear Regulatory Commission (NRC) commitments, PSEG Nuclear LLC (PSEG) hereby provides notification of a change to a commitment made in the Reference 1 and 3 letters regarding the conduct of reactor pressure vessel (RPV) inservice leak and hydrostatic testing.
The Reference 1 letter proposed changes to the Hope Creek Technical Specifications (TS) to permit the plant to remain in Operational Condition 4 with average reactor coolant temperature above 200 degrees F when conducting inservice leak or hydrostatic testing, provided certain Operational Condition 3 Limiting Conditions for Operation (LCOs) are met. In Operational Condition 4, average reactor coolant temperature is normally limited to less than or equal to 200 degrees F. In the 95-2168 REV. 7199
Document Control Desk NOV 2 3 2004 LR-N04-051 0 Reference I letter, PSEG stated that leak and hydrostatic tests are performed with all control rods fully inserted. The NRC approved the proposed TS change in Reference 2.
The Reference 3 letter proposed to delete the high drywell pressure secondary containment isolation trip function from the list of Operational Condition 3 LCOs required to be met while remaining in Operational Condition 4 with average reactor coolant temperature above 200 degrees F for reactor pressure vessel inservice leak and hydrostatic testing. In the Reference 3 letter, PSEG again stated that leakage and hydrostatic tests are performed with all control rods fully inserted. The NRC approved the proposed TS change in Reference 4.
During the current refueling outage, PSEG plans to replace 69 control rod drive mechanisms (CRDMs). Scram time testing is required to demonstrate the operability of the replacement CRDMs before plant restart. Surveillance Requirement (SR) 4.1.3.2 requires control rod scram time tests to be performed with reactor coolant pressure greater than or equal to 950 psig. In Operational Condition 4, recirculation pump operation and a water solid or nearly water solid condition in the reactor pressure vessel (RPV) are used to obtain the required temperature and pressure for single control rod scram time testing. These conditions make temperature control difficult since the RCS is isolated from the normal heat sink, and heat input to the RCS is caused by both decay heat and mechanical heat from the recirculation pumps.
The pressure and temperature conditions established for the RPV inservice leak test also satisfy the TS requirements for single rod scram time testing. Suspension of scram time testing during the RPV inservice leak test when reactor coolant temperature is greater than 200 degrees F could unnecessarily extend the time during which the RCS is isolated from the normal heat sink in a water solid or nearly water solid condition. In both Operational Conditions 3 and 4, TS Table 1.2 permits a single control rod to be withdrawn, provided that the one-rod-out interlock is operable. The one-rod-out interlock restricts the movement of control rods to reinforce operational procedures that prevent the reactor from becoming critical. With one control rod fully withdrawn, the shutdown margin requirements of LCO 3.1.1 will continue to be met; thereby ensuring the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.
PSEG has evaluated the change to the commitment regarding the conduct of reactor pressure vessel (RPV) inservice leak and hydrostatic testing and has concluded single control rod withdrawals when reactor coolant temperature is greater than 200 degrees F may be performed in accordance with TS Table 1.2 while the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.
A Document Control Desk LR-N04-051 0 NOV 2 3 2004 If you have any questions or require additional information, please contact Mr. Paul Duke at (856) 339-1466.
ePresident - Nuclear Assessment
Document Control Desk LR-N04-0510 NOV 2 3 2004 C:
Mr. S. Collins, Administrator-Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr. D. Collins, Project Manager - Salem & Hope Creek U. S. Nuclear Regulatory Commission Mail Stop 08C2 Washington, DC 20555 USNRC Senior Resident Inspector-Hope Creek (X24)
Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering PO Box 415 Trenton, New Jersey 08625