ML17261A596
| ML17261A596 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 08/31/1987 |
| From: | Kober R ROCHESTER GAS & ELECTRIC CORP. |
| To: | Stahle C NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM), Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8709100239 | |
| Download: ML17261A596 (14) | |
Text
REGULATORY iFORMATION DISTRIBUTION SY
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'A(CESSION NOR: 870'7100239 DOC. DATE: 87/08/31 NOTARIZED:
NO DOCKET FACIL: 50-244 Robert Emmet Qinna Nuclear Plant'nit ii Rochester Q
05000244 AUTH. NAME AUTHOR AFFILIATION KOBERt R. W.
Rochester Qas 5 Electric Corp.
RECIP. NAME RECIPIENT AFFILIATION BTAHLEz C.
Office of Nuclear Reactor Regulationi Director (Post 870411 BTAHLE> C.
Document Control Branch (Document Control Desk)
SUBJECT:
Forwards preliminary design info for ATWB sys scheduled for installation dur ing 1'V8'P refueling outagei per commitment.
Final design phase in progress. Detailed answers to NRC 860922 questions will be provided in Apr 1988.
DISTRIBUTION CODE:
AQSSD COPIES RECEIVED: LTR i
ENCL j SIZE:
TITLE: OR/Licensing Submittal:
Salem
- TWB Events QL-83-28 NOTES: License Exp date in accordance with 10CFR2i 2. 109(9/19/72).
05000244 RECIPIENT ID CODE/NAME PD1-3 LA STAHLEI C INTERNAL: ARM/DAF/LFMB NRR/DEBT/ICBB NRR/DEBT/RSB NRR/DOEA/QCH OQC/HDS2 REB/DE/EIH EXTERNAL:
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1 RECIPIENT ID CODE/NAME PD1-3 PD NRR LASHERI D NRR/DEBT/PBB NRR/DLPG/GAB NRR MAS/ILRB 01 NRC PDR COPIES LTTR ENCL 3
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ROCHESTER GAS AND ELECTRIC CORPORATION o 89 EAST AVENUE, ROCHESTER, N.Y. 14649-0001 ROGER W, KOBER vie pResiDCNT eLeCTRIC PRODUCT)ON August 31, 1987 TCLEPHONe ARfAcoDe Tlo 546-2700 U.S. Nuclear Regulatory Commission Document Control Desk Attn:
Mr. Carl Stahle PWR Project Directorate No.
1 Washington, D.C.
20555
Subject:
Anticipated Transients Without Scram R.E.
Ginna Nuclear Power Plant Docket No. 50-244
References:
- 1. September 22, 1986 letter from D. DiIanni (NRC) to R.
Kober (RGGE)
- 2. October 27, 1986 letter from R. Kober (RGaE) to G. Lear (NRC)
Dear Mr. Stahle:
The attached material is presented to fulfillour commitment to provide preliminary design information for the ATWS system which RG&E will install during the 1989 refueling outage.
The final design phase of this project is now in progress.
The detailed answers to the questions in Reference 1 will be provided in April 1988 instead of March 1988 as previously committed because of the extensive engineering support necessary for the 1988 refueling outage beginning in February.
Ver truly yours, Roger W. Kober Attachment S7O9iOOiW 87OaSl PDR ADOCK'5000244 P
I
ATTACHMENT KEY ELEMENTS R.E.
GINNA ATWS SYSTEM The plant specific submittal should indicate the degree of diversity that exists between the AMSAC equipment and the existing Reactor Protection System.
Equipment diversity to the extent reasonable and practicable to minimize the potential for common cause failures is required from the sensors output to, but not including, the final actuation device, e.g., existing circuit breakers may be used for the auxiliary feedwater initiation.
The sensors need not be of a diverse design or manufacture.
Existing protection system instrument-sensing lines,
- sensors, and sensor power supplies may be used.
Sensor and instrument sensing lines should be selected such that adverse interactions with existing control systems are avoided.
~Res onse The tentative RGSE design utilizes Feedwater Low Flow to sense and respond to an ATWS event.
Figure 1 illustrates how the ATWS system inputs are to be obtained from the existing feedwater flow instrumentation.
For diversity, the new ATWS hardware will be other than Foxboro H Line (Consotrol) and Westinghouse BF relays.
Lo ic ower su lies The plant, specific submittal should discuss the logic power supply design.
According to the rule, the AMSAC logic power supply is not required to be safety-related (Class 1E).
However, logic power should be from an instrument. power supply that is independent from the reactor protection system (RPS) power supplies.
Our,review of additional information submitted by WOG indicated that power to the logic circuits will utilize RPS batteries and inverters.
The staff finds this portion of the design unacceptable, therefore, independent power supplies should be provided.
~Res onse The power to the logic circuitry will be provided by the existing Technical Support Center (TSC) battery which is independent of the RPS and will not fail upon loss of offsite power.
Safet -related interface The plant specific submittal should show that the implementa-tion is such that the existing protection system continues to meet all applicable safety criteria.
1 M
R~es onse The. existing RPS is isolated from the ATWS logic by means of Foxboro M/66BR-OH current repeaters.
Therefore, the existing RPS will continue to meet all applicable safety criteria.
ualit Assurance The plant specific submittal should provide information regarding compliance with Generic Letter 85-06, "Quality Assurance Guidance for ATWS Equipment that, is not Safety-Related".
R~es onse Currently RG&E plans to implement an existing Quality Assurance Program developed for "Non-Seismic Category I Safety-Related Items" for the design, fabrication, installation, inspection,
'test and operation of ATWS equipment.
This program meets or exceeds the guidance provided in Generic Letter 85-06.
The program is defined in Section 2.2 of Appendix A to the Ginna Station Quality Assurance Manual.
A copy of Appendix A has been previously provided to the NRC.
Maintenance B
asses The plant specific submittal should discuss how maintenance at, power is accomplished and how good human factors engineering practice is incorporated into the continuous indication of bypass status in the control room.
R~es onse Maintenance at power will require placing the ATWS system in a bypass mode which will isolate the ATWS output relays from the turbine trip and auxiliary feedwater pump start logic.
The continuous indication of bypass status in the control
- room, as well as all control room modifications, will require a human factors engineering review.
The conceptual design is reviewed by the RG&E Nuclear Engineering group to ensure that the modification meets the principles of human factors engineering as established by the guidelines of NUREG-0700.
Modification design cannot be finalized until approval of the human factors engineering considerations is obtained.
0 eratin b
asses The plant specific submittal should state that operating bypasses are continuously indicated in the control room; provide the basis for the 70% or plant specific operating bypass level; discuss the human factors design aspects of the continuous indication; and discuss the diversity and independence of the C-20 permissive signal (Defeats the block of AMSAC).
K
~Res onse The ATWS operating bypass will be continuously indicated in the control room.
The ATWS syst: em actuation signals are blocked below a level of 40 percent. power, as determined by one of two turbine load signals being below predetermined setpoints.
Both of the turbine load signals exceeding their setpoint (corresponding to 40 percent power) will arm the ATWS logic, and permit actuation of the turbine trip and auxiliary feedwater start circuits.
This interlock is provided consistent with WCAP 11436, since it has been demonstrated that the reactor coolant system pressure does not approach the ASME stress level C limit of 3200 psig when an ATWS event, occurs below 40 percent power.
This will be provided to insure that spurious ANSAC actuations do not occur at low power operations and during start-up.
The block will automatically be removed as power increases above the 40% level and reinstated as power decreases below the 40% level.
A human factors review of the bypass continuous indication will be performed as discussed in Item 5.
Diversity and independence of the C-20 permissive will be achieved in the same manner as for the feedwater flow analog inputs described in Items 1 and 3.
7.
Means for b assin The plant specific submittal should state that the means for bypassing is accomplished with 'a permanently installed, human factored, bypass switch or similar device, and verify that disallowed methods mentioned in the guidance are not utilized.
8.
~Res onse The means for bypassing the ATWS actuation logic will be a
permanently installed, human factored, manual bypass switch located on the main control board.
The manual bypass will not involve lifting leads, pulling fuses, manually tripping
- relays, or physically blocking relays.
Manual initiation The plant specific submittal should discuss how a manual turbine trip and auxiliary feedwater actuation are accomplished by the operator.
R~es onse A manual turbine trip is accomplished by actuating the single Manual Turbine Trip pushbutton on the main control board. Auxiliary feedwater flow is manually initiated by moving both motor driven auxiliary feedwater pump breaker control switches on the main control board from auto to the close position.
When each pump is started, its respective discharge valve opens to supply its corresponding steam generator.
t i
II
9.
Electrical inde endence from existin reactor rotection s stem The plant specific submittal should show that electrical independence is achieved.
This is required from the sensor output to the final actuation device at which point non-safety-related circuits must, be isolated from safety related circuits by qualified Class 1E isolators.
Use of existing isolators is acceptable.
- However, each plant specific submittal should provide an analysis and tests which demon-strates that the existing isolator will function under the maximum worst case fault conditions.
The required method for qualifying either the existing or diverse isolators is presented in Appendix A.
~
R~es onse Electrical independence from existing reactor protection system hardware is achieved with existing Foxboro M/66BR-OH current repeaters (isolators).
The qualification and effec-tiveness of these isolators'bility to provide electrical isolation of maximum credible faults has been evaluated under SEP Topic VII-1.A and, on July 30, 1981, the NRC formally concluded that the Foxboro M/66BR-OH isolators in use at Ginna Station are suitably qualified for providing Class IE isolation and satisfy all current licensing criteria.
10.
Ph sical se aration from existin reactor rotection s stem Physical separation from existing reactor protection system is not required, unless redundant divisions and channels in the existing reactor trip system are not physically separated.
The implementation must be such that separation criteria applied to the existing protection system are not violated.
The plant specific submittal should respond to this concern.
~Res onse To ensure physical separation from the existing RPS, ATWS system field cables will not be run in RPS cable trays or
- conduits, and ATWS system instrumentation and relays will be installed in a cabinet. physically separate from RPS cabinets.
11.
Environmental uglification The plant specific submittal should address the environ-mental qualification of ATWS equipment for anticipated operational occurrences only, not for accidents.
R~es onse The ATWS system hardware will be suitable for the environ-ment in which it is intended to operate.
ATWS equipment is
.not subject to 10CFR 50.49.
12.
Testabilit at Power Measures are to be established to test, as appropriate, non safety related ATWS equipment prior to installation and periodically.
Testing of AMSAC may be performed with AMSAC in bypass.
Testing of AMSAC outputs through the final actuation devices will be performed with the plant shutdown.
The plant specific submittals should present the test program and state that the output signal is indicated in the control room in a manner consistent with plant practices including human factors.
R~es onse At power, operability of the ATWS system will be testable from each analog input to the output actuation relay.
During shutdown, operability of the system will be tested from the analog inputs to verification of turbine trip latch mechanism and auxiliary feedwater pump start.
The output signal will be indicated in the control room in a manner consistent with plant practices including human factors.
13.
Com letion of Miti ative Action AMSAC shall be designed so that, once actuated, the comple-tion of mitigating action shall be consistent with the plant turbine trip and auxiliary feedwater circuitry.
Plant specific submittals should verify that the protective action, once initiated, goes to completion, and that. the subsequent return to operation requires deliberate operator action.
R~es onse The ATWS system output actuation relay will be of the latching type to ensure that the protective action, once initiated, goes to completion.
A manual reset of the ATWS system will be required following each actuation.
A time delay will be included in the circuit to insure that, a manual reset cannot be accomplished prior to completion of the protective action.
Technical S ecifications i
Technical specification requirements 'related to AMSAC will have to be addressed by plant specific submittals.
R~es onse The WOG has taken a position that Technical Specifications for AMSAC are unnecessary and do not enhance the overall safety of nuclear power plants.
Normal nuclear plant admini-strative controls are sufficient to control AMSAC.
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