ML20046D458
| ML20046D458 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 08/04/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20046D457 | List: |
| References | |
| RTR-REGGD-01.097, RTR-REGGD-1.097 NUDOCS 9308200104 | |
| Download: ML20046D458 (4) | |
Text
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UNITED STATES j
j NUCLEAR REGULATORY COMMISSION f
WASHINGTON, D.C. N0001 ENCLOSURE-1 p'
SUPPLEMENTAL SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION CONFORMANCE TO REGULATORY GUIDE 1.97 IOWA ELECTRIC LIGHT AND POWER COMPANY DUANE ARNOLD ENERGY CENTER DOCKET NO. 50-331
1.0 INTRODUCTION
By letters dated July 3 and October 16, 1985, March 31, 1987, and May 3, 1989, Iowa Electric Light and Power Company, the licensee for Duane Arnold Energy Center (DAEC), submitted information regarding the implementation of Regulatory Guide (RG) 1.97, Revision 2.
The NRC staff's Safety Evaluation Report (SER) dated May 9, 1990, accepted the licensee's proposal for compliance with RG 1.97, Revision 2.
During the period of December 9-13, 1991,' Region III conducted an inspection of the DAEC facility to examine the licensee's implementation of RG 1.97.
No deviations were found during the inspection.
- However, as a result of later discussions held with the licensee, several additional action items were identified and by letter dated June 26, 1992, the licensee submitted ~a report addressing these additional items.
One item regarding Class IE power for Category'1 instruments required further clarification and by letter dated July 9,1993, the licensee provided the staff with the additional information.
2.0 EVALUATION The staff has reviewed the licensee's submittals and has concluded that the licensee either conforms to, or is justified in deviating from,_the guidance of RG 1.97 for each post-accident monitoring variable. The staff also concluded that the licensee's commitment changes described below are acceptable:
a)
By letter dated September 20, 1991, the licensee submitted a license amendment requesting to remove the component lists from-
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l Section 3.7_of the Technical Specifications (TS), based on the guidance of NRC Generic _ Letter (GL) 91-08, " Removal of. Component Lists from Technical Specifications." The licensee evaluated which valves _ would-be subject to administrative controls for containment isolation valves and revised the containment isolation valve position indication list accordingly. Also, the licensee's RG 1.97 Master Equipment List was updated to include the additional valves identified as primary containment isolation valves and position indication for some of these valves was added to the licensee's equipment qualification program.
The licensee 9308200104 930004 PDR-ADOCK 05000331 p.
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has committed to environmentally qualify these instruments.
The staff has reviewed the licensee's accounting of primary containment isolation valve position indication instrumentation and found it to be acceptable.
b)
RG 1.97 recommends a plant-specific implementation for indication-of status of standby power to control room operators. The DAEC design has this indication, along with the associated equipment, located in the essential switchgear rooms. The essential i
switchgear rooms are located within the control room / control building envelope. The licensee performed an engineering evaluation of the adequacy of the location of this indication and found the equipment in the control room adequate; however, certain l
annunciators were added to the licensee's RG 1.97 Master Equipment List. The staff has reviewed the licensee's indication of status of standby power and found it to be acceptable.
c)
Section 1.3.1.b of RG 1.97 states that no single failure within either the accident monitoring instrumentation or its power sources concurrent with the failures that are a condition or result of a specific accident should prevent the operators from being presented the necessary information. The licensee's review of RG 1.97 instrumentation identified a potential accident scenario in which a high energy line break causing an instrument line failure within primary containment coincident with loss of a single instrument power supply could result in the loss of Category I reactor pressure instrumentation in the control room.
The licensee, however, determined that High Pressure Coolant-Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) steam inlet pressures could be used as a backup indication and committed to upgrade tFese instruments to Category 1.
The staff has reviewed the licensee's use of this alternate instrumentation and has found it to be acceptable, d)
In GL 82-33, " Supplement I to NUREG 0737 - Requirements for Emergency Response Capability," Item 8.2, " Technical Support Center (TSC)," and Item 8.4, " Emergency Operations Facility (E0F)," state that variables required for TSC or EOF functions must be available in each facility. Currently, the TSC and E0F have only the information provided by the plant process computer (which does not monitor certain RG 1.97 variables). The licensee has stated that dedicated phone lines exist for the transmission of data from the control room to the TSC and E0F.
In addition,
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hard-copy data can be transmitted to the TSC and EOF via facsimile. The staff has reviewed the licensee's TSC and EOF accident monitoring capabilities and found them to be acceptable, e)
Section 1.3.1.c of RG 1.97 states that Category 1 instrumentation j
should be energized from station standby power sources and should be backed up by batteries where momentary interruption is not j
tolerable. The licensee identified certain primary containment 1
isolation valve position indications which are supplied by power
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from the reactor protection system (RPS). These valves include i
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the recirculation pump mini-purge isolation valves and the drywell i
equipment and drain sump isolation valves. The RPS is powered i
from Class IE, safety-related 480V AC electrical busses via two j
interruptible motor-generator sets and in the event of a complete i
loss of offsite power, these primary containment isolation valves would fail closed and their associated position indication would be lost.
By letter dated July 9, 1993, the licensee provided information concerning the use of alternate indication methods for the operators to use in the event of an accident and stated that, for the recirculation pump mini-purge isolation valves, the operators can observe that local mini-purge flow indication is zero.
Local visual verification of these valve positions is also available. The licensee has also stated that, for drywell equipment and drain sump isolation valves, operators can verify the fail-closed position of these valves by local visual verification from the torus room and if, during an accident, this area is not accessible, the tank levels to which these sumps discharge can be monitored. The staff has reviewed the licensee's use of alternate means for verifying valve position for primary containment isolation valves and found it to be acceptable.
i In addition, an engineering study by the licensee found the power supplies of all RG 1.97 instruments to be acceptable, except the power supplies for standby liquid control flow and cooling water flow to ESF components. The licensee has agreed to upgrade the l
power supplies for these variables to a more reliable power source.
The staff has reviewed the licensee's power supplies to RG 1.97 instrumentation and found them to be acceptable.
f)
The licensee has performed an engineering evaluation of DAEC's compliance with RG 1.97 in accordance with RG 1.75 for the physical independence of electrical systems.
This evaluation covered identification, cable spreading area and main control room, isolation devices, documentation of analyses, associated l
circuits, and internal separation. Although the construction and licensing of DAEC predates the issuance of RG 1.75 and RG 1.97, the licensee has made a good faith attempt to comply with these regulatory guides within the constraints of pre-existing structures, systems, and components. The licensee has also i
committed to comply with the guidance of RG 1.97 wherever j
practicable when installing new or replacement equipment. The staff has reviewed the physical independence of the licensee's electrical systems and found it to be acceptable.
I g)
Certain RG 1.97 instruments at DAEC require seismic qualification a
but did not require seismic qualification at the time of construction and licensing. The licensee has agreed to evaluate this instrumentation within the Seismic Qualification Utility Group methodology. The licensee has also committed to follow the guidance of RG 1.97 wherever practicable when installing new or j
replacement equipment. The staff has reviewed the licensee's seismit qualification of RG 1.97 instrumentation and found it to i
be acceptable.
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3.0 CONCLUSION
Based on our review of the licensee's submittals, we find that the Duane Arnold Energy Center design is either in conformance with, or justified in deviating from, the guidance of RG 1.97 for each post-accident monitoring variable.
Principal Contributor:
H. Rathbun Date: August 4, 1993 L
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