ML20237A748
| ML20237A748 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 08/07/1998 |
| From: | Sorensen J NORTHERN STATES POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 50-282-98-10, 50-306-98-10, NUDOCS 9808170065 | |
| Download: ML20237A748 (3) | |
Text
-
Northern States Power Company Prairie Island Nuclear Generating Plant 1717 Wakonade Dr. East Welch. Minnesota 55089 August 7,1998 10 CFR Part 2 1
U S Nuclear Regulatory Commission Attn: Document Control Desk l
Washington, DC 20555
]
]
PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket Nos. 50-282 License Nos. DPR-42 l
50-306 DPR-60 i
Reply to Notice of Violation (inspection Report 98010)
Your letter of July 10,1998, which transmitted Inspection Report No. 50-282/98010 (DRS), required a response to a Notice of Violation. Our response to the specific violation is contained in the attachment to this letter.
l The inspection report also noted other weaknesses which contributed to the inability to maintain adequate decay heat removal. Weaknesses pertaining to lack of three part communications and lack of maintaining consistent oversight of plant status are being addressed via operator training. Also, the simulator steam generator PORV fidelity weakness has been corrected and a review of the simulator auxiliary feedwater flow f
and steam generator level model is ongoing.
NSP has determined that the narrow operating range imposed on steam generator level was another factor that contributed to the inability to maintain adequate decay heat removal. This narrow operating range for steam generator level is needed to satisfy the assumptions in the Prairie Island main steam line break (MSLB) accident analysis. A new MSLB accident analysis methodology is currently ur. der NRC review and should provide a less restrictive steam generator level operating range.
In this response we have made two new Nuclear Regulatory Commission l
commitments, as indicated by the corrective actions in italics.
ll.
b l
i
{
r l
9808170065 980807 PDR ADOCK 05000282 G
USNRC NORTHERN STATES POWER COMPANY l
August 7,1998 Page 2 Please contact Jeff Kivi(612-388-1121)if you have any questions related to this letter.
J el P. Sorensen Plant Manager Prairie Island Nuclear Generating Plant l
l c:
Regional Administrator-Region lil, NRC l
Senior Resident inspector, NRC NRR Project Manager, NRC l
J E Silberg
Attachment:
RESPONSE TO NOTICE OF VIOLATION l
l l
L_________________________-.----_____.--
Attachment j
August 7,1998 Page1 RESPONSE TO NOTICE OF VIOLATION VIOLATION -
Criterion V of 10 CFR 50, Appendix B, " instructions, Procedures and Drawings,"
requires, in part, that activities affecting quality shall be prescribed by documented procedures and shall be accomplished in accordance with those procedures.
Contrary to the above, the inspectors identified that on June 5,1998, emergency l
operating procedure 1ES-0.1, " Reactor Trip Recovery," Revision 13, lacked q
adequate guidance for dumping steam using the steam generator power
~
operated relief valves with the main steam isolation valves closed. Specifically, the procedure did not provide instructions for controlling reactor coolant system temperature as evidenced by the inadvertent lift of one steam generator main steam safety valve.
This is a Severity Level IV violation (Supplement I).
RESPONSE TO VIOLATION Reason for the Violation The Prairie Island Nuclear Generating Plant (PINGP) Emergency Operating Procedures (EOPs) are based on the Generic Guidance provided by the Westinghouse Owner's Group (WOG) Emergency Response Guidelines (ERGS),
)
Revision 18. The level of detail of procedural guidance for operat;on of the steam generator power operated relief valves (PORVs) if the reactor coolant temperature is greater than no load temperature and increasing is " Dump steam using SG PORVs." Similarly, the PINGP EOP step was written with the same level of detail. Since it is demonstrated that this level of detail did not provide adequate guidance to the control room operator, additional detail is necessary.
L Corrective Steps Taken and Results Achieved l
A Temporary Memo has been issued adding detail to the step for placing the
)
l steam dump in the " Steam Pressure" mode if the main steam isolation valves l
(MSIVs) are closed, the operator is directed to reduce steam generator PORV controller setpoints to 71.5%, which corresponds to the desired reactor coolant l
l L
l
)
Attachment August 7,1998 Page 2 no load temperature of 547 degrees F. With this action taken, automatic operation of the steam generator PORVs will maintain the desired temperature and prevent r+eam pressure from rising to a steam generator main steam safety valve lift setpoint.
Corrective Steps To Avoid Further Violations 1.
The next revision to 1[2] ES-0.1 willIncorporate the detall of the temporary memo into the standing EOP.
2.
A review of all EOP steps involving the use of the steam generator PORVs will be conducted to identify any other EOP steps where it is appropriate to include additional detailed guidance.
The Date When Full Compliance Will be Achieved Comp: lance has been re-established.
l L- -_ ------_-- - _ - _- -
-