ML18033B700
| ML18033B700 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 05/10/1991 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML18033B699 | List: |
| References | |
| RTR-REGGD-01.097, RTR-REGGD-1.097 NUDOCS 9105150179 | |
| Download: ML18033B700 (6) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 ENCLOSURE SUPPLEMENTAL SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION CONFORMANCE TO REGULATORY GUIDE 1.97 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT UNIT NOS.
1 2
AND 3 DOCKET NOS. 50-259 50-260 AND 50-296
1.0 INTRODUCTION
The staff completed its review of the licensee's conformance to Regulatory Guide (RG) 1.97, Revision 3, by providing the staff's supplemental safety evaluation to the licensee, on February 8, 1990.
The staff found that the licensee's plant design was acceptable with respect to conformance to RG 1.97 with the exception of the neutron flux variable.
By letters dated October 15, 1990 and December 21, 1990, the licensee requested that the staff reevaluate the issues concerning the variables of reactor coolant system (RCS) pressure, residual heat removal (RHR) heat exchanger outlet temperature, primary contain-ment isolation valve position, emergency ventilation damper position, and neutron flux.
The NRC staff has performed a detailed review of these issues.
We conclude that the licensee either conforms to, or has adequately justified deviations from, the guidance of RG 1.97 with the exception of the neutron flux variable.
2.0 EVALUATION We have reviewed the licensee's submittals and conclude that the licensee either conforms to, or has provided an acceptable justification for deviations from the guidance of RG 1.97, for the following variables:
(a)
RCS pressure; (b)
RHR heat exchanger outlet temperature; (c) primary containment isolation valve position; and (d) emergency ventilation damper position.
We also conclude that the licensee does not conform to, or has not provided an acceptable justi-fication for not conforming to the guidance of RG 1.97 for the neutron flux variable.
(a)
RG 1.97 recommends Category 1
RCS pressure instrumentation with a range of 0 to 1500 psig.
The instrumentation provided by the licensee has a
range of 0 to 1200 psig.
The licensee states that the 0 to 1200 psig range covers the full range of pressures for which operator actions would be initiated during accident conditions.
The only postulated need for monitoring pressure greater than 1200 psig would be to document peak pressure during a reactor transient.
RCS pressure is recorded over a
range of 0 to 1500 psig by the Safety Parameter Display System.
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Based on the fact that the 0 to 1200 psig range includes the full range of pressures for which operator actions would be initiated during acci-dent conNtions and reactor transient peak pressure would be recorded, the staff finds the licensee's RCS pressure instrumentation acceptable.
(b)
RG 1.97 recommends Category 2
RHR heat exchanger outlet temperature instrumentation to monitor the operation of the RHR system.
The licensee has provided Category 3
RHR heat exchanger outlet temperature instrumenta-tion.
The licensee has also provided suppression pool water temperature, suppression pool water level, drywell atmosphere temperature, drywell atmosphere
- pressure, reactor coolant level, RCS pressure instrumentation that as a minimum meets the Category 2 criteria of the regulatory guide.
The emergency operating instructions (EOIs) rely on this alternate instrumentation as primary indicators for determining the heat energy remaining in the containment.
The staff finds the licensee's alternate instrumentation, associated
- EOIs, and Category 3
RHR heat exchanger outlet temperature instrumenta-tion acceptable for monitoring the operation of the RHR heat removal system.
c)
RG 1.97 recommends Category I primary containment isolation valve posi-tion indication instrumentation to monitor the accomplishment of contain-ment isolation.
The instrumentation provided by the licensee meets the Category I criteria with the exception of the instrumentation associated with eighteen valves which receive power from the reactor protection system (RPS) power supply, RHR shutdown cooling valves, and Traversing Incore Probe'(TIP) ball valves.
The 18 valves which receive power from the RPS power supply, fail closed on the loss of power.
When the RPS power supply is functional, the position indication for these valves will be available in the control room.
The licensee has stated that the EOIs specifically address other parameters (temperatures, pressures, and radiation) to confirm mainte-nance of primary containment integrity in the event of the loss of the RPS power supply.
Based on the above information, the staff finds the licensee's position indication instrumentation acceptable for these 18 containment isolation valves.
Valves FCV-74-47 and FCV-74-48 are the inboard and outboard containment isolation valves for the RHR shutdown cooling supply line to the RHR pumps.
These valves are normally closed, administratively controlled, and are not required to operate for accident mitigation.
Valve FCV-74-47 is maintained with the power removed for Appendix R compliance and therefore, the position indication would not function with the power removed.
Valve FCV-74-48 is fed power through cables that have not been environmentally qualified.
Valve FCV-74-48 provides positive indication of position indication unless power is removed.
Based on the described use of these valves, and the fact that they are normally closed and administratively controlled, the staff finds the described deviations in position indication acceptable.
(d) e)
3.0 There are five TIP ball valves that perform containment isolation valve functions.
The licensee has provided dual barrier protection through the use of a solenoid operated ball valve and an explosive actuated cable shearing valve.
The position indication circuits are powered from non-Class 1E power.
The TIP system is non-safety related.
The ball valves are designed to prevent automatic reopening of the valves upon reset of the isolation signal.
The ball valves are normally closed and administratively controlled.
Based on the described use of the
- valves, and the fact that they are normally closed and administratively controlled, the staff finds the described deviation acceptable.
RG 1.97 recommends Category 2 emergency ventilation damper position instrumentation to monitor the operation of the ventilation system.
For six dampers, the licensee has provided status lights that provide indica-tion of power applied to move the dampers to their emergency positions.
In addition, the licensee has provided ventilation system radiation and abnormal condition alarms.
The licensee has also stated that the dampers are accessible during and following an accident and can be inspected and manually operated.
Based on the information provided, the staff finds the alternate instru-mentation acceptable to monitor the operation of the ventilation system.
RG 1.97 recommends Category 1 neutron flux monitoring instrumentation to monitor reactivity control.
The licensee has provided neutron flux monitoring instrumentation which complies with the Category 1 criteria except for source and intermediate range monitor drive mechanisms and controls and'the neutron monitoring system power sources.
The staff's position, provided by the February 8, 1990 supplemental safety evalua-tion, is that the existing neutron flux instrumentation is acceptable for interim operation and that the licensee shall install neutron flux monitoring instrumentation which complies with the Category 1 criteria, of RG 1.97, Revision 3.
The licensee has endorsed the Boiling Water Reactor Owners Group (BWROG) appeal of the NRC staff position that directed the installation of upgraded neutron monitoring instrumentation.
The licensee has deferred plant-specific implementation unti 1 the appeal is resolved by the Director of the Office of Nuclear Reactor Regulation (NRR).
The Director of NRR is currently reviewing the BWROG appeal.
It is the staff's position that plant-specific implementation be deferred until the Director resolves the appeal.
CONCLUSION Based on the staff's review of the licensee's submittals, we find that the Browns Ferry Nuclear Plant Units 1, 2 and 3 overall design is acceptable with respect to conformance to RG 1.97, Revision 3, with the exception of the neutron flux variable.
The staff finds acceptable the existing neutron flux instrumentation for interim operation.
It is the staff's position that plant-specific implementa-tion be deferred until the Director of NRR resolves the appeal filed by the BWROG.
Dated:
May 10, 1991 Principal Contributor:
B. Marcus, SICB
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