ML20040A671
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UNITEo STATES N
NUCLEAR REGULATORY COMMISSION yT)1 o
WASHINGTON, D. C. 20555
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Jg 2 51979
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MEMORANDUM FOR:
R. Satterfield, Chief, Instrumentation and Control Systems Branch, Division of Safety Systems i
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G. Iainas, Chief, Plant Systems Branch, Division of Operating Reactors Ol FROM:
D. Tondi, Section Leader, Plant Systems Branch, Division of Operating Reactors 2.
SUPJECT:
INSTRUMENT PIPING ON REACTOR COOLANT FLOW TRANSMITTERS (Ij FPS, B&W PLANTS 4
I The purpose of this memorandum is to ensure consistency of reviews perfomed by DSS and DCR on a specific area of concern.
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Condition 2.C (7) of the Crystal River Unit 3 Operating License
'l require's Florida Power Corporation to submit a fix for the single set of instrument piping per reactor coolant loop which serve the l
four differential pressure transmitters used in the RPS. We are requiring the Licensee to backfit the plant with a second set of instrument piping per reactor coolant loop in order that the RPS instrumentation' meet the single failure criterion.
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-h-It is requested that this requirement of, "a minimum of two instrument piping systems per reactor coolant loop," be required'of those
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applicants of B&W plants where the RPS overpower trip based upon reactor coolant flow / axial power imbalance utilize only one set of instrument piping for four differential pressure transmitters per reactor coolant
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loop.
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D. Tondi, Section Leader Plant Systems Branch
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Division of Operating Reactors
Contact:
J. Burdoin, X28077 cc:
J. Burdoin F. Rosa W. Russell F. Asme G. Iainas C. Miller ~
D. Tondi M. Srinivasan i
k 8201210389 810403 PDR FOIA MADDEN 80-515 PDR
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SER BAW-10003 " Qualification Testing of Protection Instru entation," Revision 4 March 1974 The protection system is cocprised of four redundant and independent channels.
- 1 Each channel monitors the following parameters which cause a bistable trip which in s
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turn actuates a trip relay within a reacter trip module: (1) neutron flux. (2) reactor coolant flow. (3) neutron flux power imbalance (4) reactor coolant pump f
status. (5) reactor coolant pressure (high and low). (6) high reactor coolant outlet l
temperature, and (7). high containment vessel pressure.
Each trip module combines the four channel trip outputs in a two-out-of-four logic to y
trip the control rod power supply breakers. Each channel is completely testable during power operation.
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- The present design of all four redundant reactor coolant flow transmitters measuring flow in each loop to the steam generators, which are part of the reactor coolant system flow and axial power imbalance protection function, share comon processs
, sensing lines to the reactor coolant flow tube. We will require that design of the
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reactor coolant flow parameters be modified to meet the single failure criteria in order to preclude loss of the reactor coolant flow / axial power protection function..
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j The applicant is reviewing the staff position and will advise us regarding the details of the design modification and the schedule for the completion of its installation.
l We will report this matter in a supplement to the Safety Evaluation Report.
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We have previously reviewed the design Jefhth gencri and ecific items g
applicable to the Rancho Seco Operating License application.
The Davis Besse reactor protection system design differs from t'at of Rancho Seco 1 in h
two respects. These involve the power supply interrupt interface between the reactor protection system and the control rod drive mechanisms. The Davis Besse facility uses two manual reactor trip switches in series (instead of one) to interrupt power to each I
undervoltage coil of the main alternating current feeder breakers and thereby disconnects the power to the rod drive mechanisms. The facility has a redundant diverse method of power interrupt using silicon control rectifiers in the rod group power supplies instead of dire et current breaker interrupt of the holdir; power supplies.
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I Since the diverse method of power interrupt originally proposed was not seismically qualified and therefore unacceptable, the applicant modified the design by adding two I
qualified Class IE alternating current main feeder breakers in series with the l
existing alternatir.g current breakers. The non-qualified diverse silicon enntrol i
rectifier trip has been retained as a backup. These changes conform to the require-mea
- of the Institute of Electrical and Electronic Engineers Standard 279-1958, the standard applicable to the Davis Besse facility, and are therefore acceptable ionditioned only on the satisfactory documentation of this change in the Final Safety Y
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valuation Report.
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7.0 INSTRUMENTATION AND CONTROL 5 7.1 General In our Safety Evaluation Report we stated that the tiectrical drawings of the reactor protection system, the engineered safety features system and the Class 1E support systems that were submitted in the Final Safety Analysis Report were incom-plete in part or were not presented in sufficient detail to verify that the de' sign had been implemented adequately. We required the applicant to submit a final design package for all safety-ralated equipment in sufficient detail to facilitate our review. Revised final design drawing packages were submitted by the applicant with sufficient detail and pemitted us to, conduct an independent review. We conclude that the drawings presently docketed in the Final Safety Analysis Report are adequate for an operating license review and are acceptable. Therefore, we consider this matter resolved.
7.2 Reactor protection System In our Safety Evaluation Report we stated that we would report the results of our evaluation regarding modification of the reactor coolant system flow sensors y
in regard to the common pressure sensing line to all four differential pressure
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transmitters. We have now detemined that the system should be modified to reduce the susceptibility of the system to false flow indication in the event of single failure (e.g., a break, leak or plugging of either the high pressure or low pressure sensing line.)
We have infomed the applicant of the need to modify the system to reduce the susceptibility to false flow indication. We will review the proposed modifica-tions when the applicant completes its assessment and detemines what modifications can be made, and we will require that approved modifications be implemented during or prior to the first refueling outage.
j We conclude that until this matter is satisfactorily resolved, the surveillance requirements imposed by the plant technical specifications on the reactor pro-l tection system instrumentation (Table 4.3-1) and on the reactor coolant system l
operational leakage (Section 3.4.6.2) provide an acceptable assurance that breaks l
or leaks in the sensing lines will be detected. Based on operating experience, we l
also conclude that for the interim period, plugging of the sensing lines is highly unlikely.
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