ML20198S236
| ML20198S236 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 06/05/1986 |
| From: | Leblond P COMMONWEALTH EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| 1714K, NUDOCS 8606100368 | |
| Download: ML20198S236 (10) | |
Text
r-Z }. 72 West Adams Street, Chictgo, Elinois Commonwealth Edison V
Address Reply to: Post Office Box 767 Chicago, Illinois 60690 0767 June 5, 1986 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Zion Nuclear Power Station Units 1 and 2 ATWS Protection - 10 CPR 50.62 NRC Docket Nos. 50-295 and 50-304 References (a): NRC Generic Letter No. 85-06, dated April 16, 1985.
(b):
G. A. Alexander letter to H. R. Denton dated October 10, 1985.
(c): WCAP-10858, "AMSAC Generic Design Package", dated June, 1985.
Dear Mr. Denton:
10 CPR 50.62 was issued on June 26, 1984 to address the requirements for the reduction of risk from anticipated transients without scram (ATWS).
Reference (a) provided guidance regarding the Quality Assurance required for non-safety related ATWS equipment. Reference (b) documented Commonwealth Edison's plans for implementing ATWS modifications. The purpose of this letter is to transmit a description of the proposed modifications for Zion
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Station. provides a detailed description of the proposed modification. Commonwealth Edison Company believes that this Zion-specific design is consistent with the design philosophy contained in reference (c).
It is our understanding that the NRC is preparing to approve reference (c).
Thus, Commonwealth Edison is providing this submittal in order to expedite the overall implementation of these modifications.
l provides a description of the testing capabilities that have been incorporated into this modification. These capabilities include provisions for both at-power and zero-power testing.
8606100368 860605 00 1
PDR ADOCK 05000295 P
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Mr. H. R. Denton June 5, 1986 Assuming timely NRC review and approval of both reference (c) and this specific design, the final implementation dates discussed in reference (b) remain valid. Those implementation dates are; Unit 1 - Fall 1987 refueling outage Unit 2 - Fall 1988 refueling outage Please direct any further questions regarding this matter to this office.
Very truly yours, P. C. LeBlond Nuclear Licensing Administrator 1m Attachments cc: Resident Inspector - Zion J. A. Norris - NRR f
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ATTACHMENT 1 DESCRIPTION OF PROPOSED ATWS MODIFICATION I
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i FIGURE I ANTICIPATED ATWS MOOlFICATION USING EXISTING STEAM GENERATOR N.R. LEVEL TRANSMITTERS 6#I$
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The general features of the ATWS modification are identified as follows:
1.
The existing reactor protection system, including electrical separation and independence, will not be compromised by this new modification.
2.
The ATWS mitigation system will be capable of performing its function on loss of offsite power. However, this assumes the AC and DC power is ultimately supplied by the existing Safety Related diesel generators and batteries which are presently used for reactor protection and safeguards. There are no plans to add a new diesel generator / battery system for ATWS. Within the existing diesel generator / battery system, power selections for the ATWS mitigation system will minimize common mode failure concerns.
3.
The electronics added for this modification will be located in a non-harsh area and, therefore, Environmental Qualification (10 CFR 50.49) will not be applicable.
4.
per 10 CPR 50.62, the equipment is not required to be safety I
related. However, Commonwealth Edison has elected to procure and install this modification as safety-related. This will ensure the utilization of Commonwealth Edison's pre-existing quality procedures. The ATWS mitigation system will not be specifically i
designed and installed as a seismically qualified system.
l l
5.
ATWS mitigation system initiation will occur when 3 of the 4 steam l
generator levels are below the setpoint. There will be one train of actuation circuitry.
6.
The existing features for manually starting Auxiliary Feedwater Pumps (Aux FW pps) and manually tripping the Turbine from the
-Control Room are satisfactory. Thus, a control room manual ATWS trip is not necessary and will not be installed.
The specific design and features are presented in Figure 1.
The eight notes that follow provide a detailed description and discussion of the proposed system.
Notes 1.
The ATWS modification will use the existing reactor protection steam generator narrow-range level transmitters as the sensors.
l The ATWS signal will be taken from the primary current loop of each transmitter. The existing transmitter has its own safety related power supply module, Instrument Maintenance (IM) Testing points, control Room computer input ( @ ) and control room status light indication (($))asshowninFigure1. Not shown are the other control room annunciators and level indicators and most of the related reactor protection logic.
The use of existing transmitters has reduced the cost of the possi-ble work by a significant amount. Using these existing transmitters does not reduce the integrity of the reactor protection system or the integrity of the proposed system. Steam generator inventory is monitored with annunciation, not only for levels, but independently
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for steam flow versus feedwater flow discrepancies. It is extremely unlikely that a problem with a transmitter could occur without an annunciator alarm. Thus, this diverse combination of indication provides a high degree of assurance that a reliable input to both the RPS and the ATWS mitigation system will be maintained. The use of existing transmitters has eliminated the I
need for mechanics to acquire radiation dose because new transmit-ters would be adjacent to the steam generators. Also, personnel performing maintenance or surveillance workers will not acquire additional radiation doses because new equipment will not be installed in a radiation area. Thus, As-Low-As-Reasonable-Achievable (ALARA) Radiation doses will be achieved by not adding new transmitters. Also, using existing transmitters will minimize training requirements.
2.
The existing reactor protection at Zion consists of the Westinghouse Electric Corp. (W) 7100 series equipment. To emphasize diversity the ATWS equipment immediately downstream of i
the transmitters will be W 7300 series reactor protection electronics, which is a newer generation equipment. A majority of the W 7300 equipment has already been purchased safety related and consists of (a) isolation circuits to prevent adverse feedback to the existing reactor protection, (b) cabinet power supply modules, capable of functioning in a loss of offsite power that are independent from the existing transmitter power supply modules and (c) setpoint and comparator circuitry.
Consistent with Reference (c), the setpoint to initiate the system is 6% of the steam generator narrow range level. The existing j
l reactor protection low-low narrow-range steam generator level reactor trip and Aux. FW pp. initiation is 10%.
3.
The Relay / Matrices / Testing (RMT) Cabinet contains a level relay for each loop that will energize when Steam Generator narrow range level drops to 6% or when the manual trip is activated for testing.
The manual trip can also be used to put the system in a tripped j
state for that loop in the case of failed loop equipment. Each level relay will send a signal to a control room status light, a t
local test light (T.L.) and the matrices as shown in Figure 1.
l l
4.
An existing nuclear power permissive, P-8, will be used to arm the system at 60% power which is consistent with the 70% required by Reference (c). Although the power range detector signals (for the P-8 permissive) would remain at elevated levels after the related accident transients, the time period may not be sufficient to ensure the continued presence of the arming signal. Therefore, a timer is added to the P-8 circuit to ensure that P-8 is armed to 2 minutes beyond the time that the reactor power drops below 60%.
The p-8 nuclear power permissive was selected because it is extremely reliable, has control room alarms when problems arise, has methods established for dealing with a failed channel, and will simplify training and maintenance since station personnel are familiar with it.
The timer and P-8 Test Switch will be in the RMT Cabinet. The Test Switch can be used to manually arm ATWS if the automatic feature was impaired.
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. 5.
The Matrices with the P-8 arming signal feed annunciators and the computer as well as providing an actuating signal. One control room annunciator alarms when the S.G. narrow range level drops to 6% in one loop if the power level is above the P-8 level. A second control room annunciator (which also has first out indication capability) alarms when the S.G. level drops to 6% in three loops if the power level is above the P-8 level. The process computer records the event for each loop when the S.G. level drops to 6% if the power level is above the P-8 level. An event recording also occurs for the sequence of events program when the S.G.
level drops to the ATWS setpoint in three loops if the power level is above the P-8 level.
6.
The control room operator would be alerted by a control room annunciator, if the P-8 Test Switch or the Block test switch were in the non-normal position at any power level.
i 7.
The Block Test Switch would be used with all testing where an actual Turbine Trip was not desired. When the Block Test Switch is in the Block position, varieus matrices combination, the P-8 signal, and the Actuating Relay coils can be tested with a push button (not shown in Figure 1) for a continuity check. The actuating coil does not energize and lock-in during this continuity test because of the low electrical current in this case. Also, the turbine trir solenoid coil would be checked for continuity in a similar manner. Thus, all features upstream of the Actuating Relays can be tested without tripping the turbine. Testing could involve going as far upstream as the IM Test by the transmitters.
8.
With the Block Test Switch in the non-blocking position, an ATWS signal will energize the Actuating Relays and they will stay energized (lock-in) until the control room reset pushbutton is pushed. This lock-in feature eliminates the need for an additional timer. The actuating relay sends a signal to start all the auxiliary feedwater pumps, and to trip the turbine. Sampling and blowdown valve position will not be altered upon system actuation.
This is justified by the small flow rates involved.
1714K
r ATTACIOGDIT 2 TESTING CAPABILITIES OF TifE PROPOSED ATWS MODIFICATION i
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f The anticipated testing for the ATWS modification would involve two basic pieces: at-power testing and zero-power testing.
A typical example of at-power testing is explained here with the use of Figure 1 of Attachment 1 as follows:
At-Power Test
- 1) Start at steady state condition at any power level.
- 2) At the Relay, Matrices, and Testing (RMT) Cabinet verify the P8 Test Light (P8T.L.) is lit if power is above 60%.
3)
If the P-8 Test Light is off, then put the P-8 Test Switch to test and then verify that the P-8 Test Light energizes and verify that the control room annunciator indicates that the switch is in a non-normal position.
- 4) If the P-8 Test Switch is in a non-normal position, then return it to normal and verify the annunciator returns to normal.
- 5) Change the Block Test Switch from normal to block and verify the Block Test Light (B T.L.) energizes and the control room annunciator alarms.
6)
If the power level is below 60%, then put the P-8 Test Switch to the test position. This arms ATWS by simulating a high enough power level.
- 7) At the RMT cabinet for one loop put the manual trip switch to trip, verify (see Figure 1) that the level relay test light (T.L.), the control room status light q$D,thecontrolroomannunciatorandthe control room computer acknowledge a problem with one of four loops. At the RMT Cabinet press the continuity check button (not shown in Figure 1) for each Actuating Relay and verify that the continuity light (not shown in Figure 1) does not go on when pressed. Thus, there is not continuity when there is only 1/4 loops in the simulated ATWS condition, as expected.
- 8) Repeat 7) for other loops.
- 9) At the RMT cabinet put the manual trip switch to trip for three of the four loops. Verify the associated 1/4 loops annunciator and computer indications. Verify the 3/4 loop control room annunciator and computer indication. At the RMT Cabinet, press the continuity check button for each Actuating Relay and verify that the continuity light goes on, thus, there is continuity through the 3/4 loop matrices, the p8 interlock and the Actuating Relay.
- 10) Repeat 9) for other combinations of 3/4 loops tripped.
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. Zero-power testing is expected to be done infrequently, maybe once during each refueling outage. Typical examples of zero-power testing are explained here with the use of Figure 1 of Attachment 1 as follows:
Zero-Power Test (1) At a time during the outage when steam generator levels are below the ATWS setpoint, verify the expected indications as far downstream of the sensors as possible, at least to the level relays in the RMT cabinet.
(2) As an alternate to 1), test signals can be put in at the IM test point. Also, setpoint adjustment would consist of verifying the operation of the level relay at the proper level of signals to the setpoint and comparator circuitry.
(3) At a convenient place in the outage, the P8 Test Switch could be used with the manual trips for the loops to actually demonstrate a start the Aux. FW pps. The turbine trip demonstration could be coordinated at the same time. The Control Room Reset Button would be used afterwards to demonstrate its proper function.
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