ML20209E097
| ML20209E097 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 04/23/1987 |
| From: | Cutter A CAROLINA POWER & LIGHT CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.E.4.2, TASK-TM NLS-87-088, NLS-87-88, NUDOCS 8704290428 | |
| Download: ML20209E097 (2) | |
Text
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--.o P. O. Box 1s51 e Releegh, N. C. 27602 m e m 23i SERIAL: NLS-87-088 APR 2 3 W 10CFR50.90 a
A B. CUTTER Vce President Nucieer Engineermg & Ucensmg United States Nuclear Regulatory Commission ATTENTION: Document Control Desk hashington, DC 20555 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 DOCKET NOS. 50-325 & 50-324/ LICENSE NOS. DPR-71 & DPR-62 i
CONTAINMENT ISOLATION ON HICH RADIATION NUREG-0737, ITEM II.E.4.2.(7) i I
Gentlemen:
On March 5,1987, the Staff issued a Safety Evaluation Reporting relating to Carolina Power & Light Company's design to meet the requirements of NUREG-0737, item II.E.4.2.(7) at the Brunswick Steam Electric Plant. The Company proposed to use a i
high radiation signal from the main stack monitor to isolate the containment vent and f
purge isolation valves. The SER determined that this design complies with NUREG-0737, Item II.E.4.2.(7) but requested that Technical Specifications be submitted for the operability of the radiation isolation signal circuitry and for the stack monitor I
setpoints. Carolina Power & Light Company has evaluated the request and has determined the existing Technical Specifications for the stack radiation monitor provide
- sufficient assurance of operability and that additional specific Technical Specifications for the operability of the radiation isolation signal circuitry and for the stack monitor setpoints are not necessary.
l On February 6,1987, the Commission published its Proposed Policy Statement on Technical Specification Improvements for Nuclear Power Reactors (52FR3788). This policy statement provided criteria for determining which Limiting Conditions for Operation should be included in the Technical Specifications. The Company has reviewed I
the requested Technical Specifications against the Commission's criteria and determined i
that:
1.
The installed radiation isolation signal circuitry is not used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. This function is performed by the existing drywell area radiation monitors, low reactor water level instrumentation, and high containment pressure instrumentation. The stack monitor and the associated logic circuitry are considered nonsafety related. The nonsafety-related circuit is isolated from the existing safety-related circuit and, as such, will not af fect l
the existing safety-related isolation signals.
2.
The modification does not affect any process variable that is an initial condition of a Design Basis Accident or Transient Analyses that either assumes the failure t
of or presents a challenge to the integrity of a fission product barrier. As i
stated above, the isolation signal circuitry is isolated from the existing i
safety-related circuit and, therefore, will not af fect the existing safety related j
isolation signals.
8704290428 870423 i
PDR ADOCK 05000324 n0 P g,D P
. United States Nuclear Regulatory Commission NLS-87-088 / page 2 APR 2 31987 3.
The installed nonsafety-related radiation isolation signal circuitry is not part of the primary success path which functions to mitigate a loss-of-coolant accident inside containment. The safety-related low reactor water level and high containment pressure isolation signals fulfill this function. Technical Specifications 3.3.2, 3.3.5.9, 3.3.5.3, 3.11.2.1, and 3.11.2.8 govern the operability of the water level and drywell pressure instrumentation, as well as the operability of the main stack radiation monitor and the drywell area radiation monitor, venting and purging procedures, and gaseous radioactive effluent activity. The main stack radiation setpoints are listed and controlled in the Brunswick Offsite Dose Calculation Manual, which is submitted to the Staff as part of the Semi-Annual Radioactive Effluent Release Report in accordance with Technical Specification 6.13.2. In addition, CP&L has revised Abnormal Operating Procedure 6.2 to address this isolation capability. These measures assure that 10CFR100 dose limits are not inadvertently exceeded.
Based on the above assessment and the Company's and the NRC's commitment to improve Technical Specifications, Carolina Power & Light Company has determined that the additional requested Technical Specifications are not necessary and are not consistent with NRC policy on this subject.
Please refer any questions regarding this matter to Mr. Sherwood R. Zimmerman at (919) 836-6242.
Your very y,
A. B. Cutter S
MAT /lah (5182 MAT) cc:
Dr. J. Nelson Grace (NRC-RII)
Mr. W. H. Rutand (NRC-BNP)
Mr. E. Sylvester (NRC)
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