ML20212N659

Forwards Test Procedure Showing Test Configuration & Acceptance Criteria & Test Results Verifying Max Credible Fault Applied to Output of Rochester Instrument Model SC-1302,per SPDS SER Open Item

ML20212N659
Person / Time
Site:	Three Mile Island
Issue date:	03/09/1987
From:	Hukill H GENERAL PUBLIC UTILITIES CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20212N663	List: ML20212N659 ML20212N671
References
5211-87-2051, NUDOCS 8703130107
Download: ML20212N659 (4)

Text

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GPU Nuclear Corporation b

NggIg7 Post Office Box 480 Route 441 South Middletown, Pennsylvania 17057-0191 717 944-7621 TELEX 84 2386 Writer's Direct Dial Nurnber:

March 9,1987 5211-87-2051 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Gentlemen:

Three Mile Island Nuclear Generating Station Unit 1 (TMI-1)

Safety Parameter Display System Isolation Device Test Results In the NRC Safety Evaluation Report on the TMI-l Safety Parameter Display System (Reference 1), the NRC identified an open item with respect to a test program and test results for the Rochester Instrument Model SC-1302, which is

< used to buffer reactor coalant pump power from the computer. The test report was to demonstrate that the device is qualified to protect safety systems (i.e., RPS related pump power monitors) from maximum credible faults. The SER requested that GPUN provide the following:

1. Test procedure showing test configuration and acceptance criteria

2. Test results to verify that the maximum credible fault was applied to the output of the Model SC-1302 in the transverse mode (between signal and return) and no perturbations were seen on the Class lE input.

A simplified sketch of the reactor coolant pump power monitor is provided as Attachnent 1. Rochester Instrumnt Systems was contracted to perform the testing. The test procedure is provided as Attachment 2. Test results are provided as Attachment 3.

TEST RESULTS Test 2Cc is the specific transverse fault requested. Tests 2Ca and 2Cb were also performed for completeness. (Terminals 5 & 6 are the output terminals)

Test 2Ca: 120 volts with a 20 amp capability was applied to terminal 5 and chassis ground.

A 20 millivolt .2 millisecond spike was observed on the input.

Test 2Cb: 120 volts was applied to terminal 6 and chassis ground.

A 65 millivolt .67 millisecond spike was observed on the input.

8703130107 870309 PDR h0 ADOCK 05000289 PDR g g GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation ,

5211-87-2051 March'9, 1987 Test 2Cc: 120 volts was applied to terminals 5 & 6.

After approximately 2.5 seconds the input voltage increased from 1 volt to 4.5 volts DC~ (zero volts AC)

DISCUSSION OF TEST RESULTS Yoltage spikes during tests 2Ca'and 2Cb are not of significance.

Test 2Cc was a destructive test, as expected. The output circuit failed open. Inspection revealed evidence of high temperature associated with the destruction of the output filter capacitor, output transistor and output transistor drive circuits.

, The SC-1302 input resistor is located in proximity to output circuitry components. When these output components were destroyed, they exposed the input' resistor to high temperature, thus increasing its resistance as .noted by the input. voltage increase from 1 to 4.5 volts D.C. with a constant current

' input. The fact that the output circuit failed open is significant in that further- degradation was prevented.

It is important to note the absence of A.C. voltage on the input, which demonstrated that isolation was maintained.

Thus, it is necessary only to discuss the impact of the increased input impedance on the isolator's safety related circuit side (see Figure _l, attached). The reactor coolant pump watt transducer output is 0-1 milliamp D.C. into an output impedance of- 0-10,000 ohms. = The transducer's output load is two SC-1302 isolation devices in series. This increase in the SPDS 1302 voltage drop represents an inpedance change (2000 to 5500 ohm) which is within

, the watt- transducer's capability. The other SC-1302, associated with the Reactor Protection System, would be unaffected.

Rochester-has provided the test procedures and test results, a copy of which are attached. A GPUN representative observed the testing, which was performed L on January 22, 1987.

i Sincerely, t

l .

s

. D Hukill Vice President & Director, TMI-1

- HDH/SK/pa(1968g) l l Attachments: (1) Simplified Sketch of RC Pump Power Monitor l (2) Test Procedure (3) Test Results

Reference:

(1) NRC Letter, "TMI-1 Safety Parameter Display System",

John F. Stolz to Henry D. Hukill, Decenter 23, 1985.

cc: T. E. Murley l R. Conte

ATTACPRFET 1 RPS n

SC-1302 CTS y RIS WATT g

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m 0-1 MAMP

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PTS } (INTO 0-10,000 OHM) SC-Eb 1302 COMPUTER

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l SIMPLIFIED SKETCF. OF RC PUMP POWER MONITOR FIGURE 1 l

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C 4 4 ATTACHMENT 2 TEST PROCEDURE T

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