ML20235R194
| ML20235R194 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 09/30/1987 |
| From: | Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| 85-316G, NUDOCS 8710070823 | |
| Download: ML20235R194 (52) | |
Text
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i VIHOINIA 13LECTHIC AND l'OWEli COMPANY l Hicn >5onn,Vinoisir un e61 W. L. Str:wAar N.N2[I)$u"S$m. September 30, 1987 I
l
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United States Nuclear Regulatory Commission Serial No. 85-316G l Attention: Document Control Desk Docket Nos. 50-338 Washington, DC 20555 50-339 License Nos. NPF-4 I NPF-7 1
Gentlemen: j Virginia Electric and Power Company North Anna Power Station Units 1 and 2 Anticipated Transient Without Scram-AMSAC Design Virginia Electric and Power Company stated in our letter of August 28, 1987, )
(Serial No. 85-316F), that we would submit our AMSAC plant specific design i features by September 30, 1987. In accordance with that letter, we are submitting our plant specific design as an attachment to this letter. The attachment addresses each of the fourteen itetus identified by the NRC in the SER issued for Westinghouse plants. l l
Engineering is in progress on this system but is not complete. If changes are l made which alter the design presented in the attachment to this letter, we will l provide you information on the changes.
Our schedule for installation of the AWS mitigation system is based on the time required to complete the detailed design and the lead time necessary for equipment delivery. Our current plans are to have the system installed during i the refueling outagea at the end of cycle 7 and 6, respectively for Units 1 and i
- 2. 1resently, the Unit 1 outage is scheduled to be completed in May 1989 and the Unit 2 outage to be completed in December 1988.
Very truly yours, i
p !
h.L. Stewart Attachment m*IB8RBMjk l fob I 't )
-- J
cc: U. S. Nuclear Regulatory Commission Marietta Street N.W.
Suite 2900 Atlanta, Georgia 30322
,Mr. J. L. Caldwell NRC Senior Resident Inspector North Anna Power Station 1
l l
1
e Attachment LICENSING POSITION ATWS MITIGATION SYSTEM ACTUATION CIRCUITRY (AMSAC)
NORTH ANNA POWER STATION - UNITS 1 AND 2 l
1.0 INTRODUCTION
In order to comply with 10CFR50.62 " Requirements for Reduction of Risk .
From Anticipated Transients Without Scram (ATWS) Events fo' ;,ight-Water- l Cooled Nuclear Power Plants," the Westinghouse Owners Group '(WOG) l prepared and submitted for Nuclear Reguiatory Commission (NRC) review and approval topical report WCAP-10858 "AMSAC Generic Design Package." The NRC's acceptance position of the generic topical report and WCAP-10858A "AMSAC Generic Design Package" formed the basis for preparing North Anna's licensing submittal.
2.0 BACKGROUND
The Anticipated Transients Without Scram (ATWS) Final Rule, 10CFR50.62, allowed the NRC to amend its regulations to require improvements in design and operation of pressurized water reactors to reduce the likeli-hood of a failure to scram and to mitigate the consequences of an ATWS.
The NRC does not believe that the current reactor trip system achieves adequate reliability. They believe this- is due to two reasons: (1) reliability standards are not sufficiently developed or qualitatively i documented; and (2) the dominant role played by common mode failures.
Consequently, the ATWS Final Rule requires diversity from sensor output to the final actuation device to automatically initiate auxiliary feed-water flow and trip the turbine under conditions indicative of an ATWS.
3.0 CRITERIA North Anna must implement paragraph (C)(1) of 10CFRSO.62, "Each f pressurized water reactor must have equipment from sensor output to f.inal
)
1 134-BSD-5281NA-4
l actuation device, that is diverse from the reactor trip system, to ,
automatically initiate the auxiliary feedwater system and initiate a turbine trip under conditions indicative of an ATWS. This equipment must be designed to perform its function- in a reliable manner and be independent (from sensor output.to the final actuation device) from the~ 1 l
existing reactor trip system."
Although the required ATWS mitigation system does not have to be safety related, it is part of the class of systems and components defined in General Design Criteria (GDC) 1, which requires that " structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed." Generic Letter 85-06 " Quality Guidance for ATWS Equipment That Is Not Safety Related" provides direction for the Quality Assurance program that must be applied to the ATWS mitigation system. l l
l 4.0 DESIGN DESCRIPTION The Westinghouse Owners Group (WOG) in concert with Westinghouse Electric ,
l Company prepared WCAP-10858, "AMSAC Generic Design Package." This-I document was submitted to obtain NRC approval of the design prior to implementation of the changes required by 10CFR50.62. The application for NRC review was submitted in 1985 with the Draft SER issued in June 1986, and the Final Safety Evaluation published July 7,1986. The Final Safety Evaluation (Final SER) approved WCAP-10858 and accepted the principal of using only one of three proposed functional designs to detect the onset of ATWS. An accepted version of WCAP-10858, WCAP-10858-Revision A, was issued in October, 1986.
By definition, an ATWS is an expected operational transient (i.e., loss of feedwater, loss of condenser vacuum, or loss of offsite power) which is accompanied by a failure of the reactor trip system to shutdown the reactor. The three functional designs proposed to provide detection of a failure of the Reactor Protection System (RPS) to initiate a reactor trip and a loss of feedwater or loss of load are based on the power level 134-BSD-5281NA-5
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l (reactor or turbine power) and the monitoring of (1) steam generator inventory level, (2) feedwater flow, or (3) feedwater pump breaker and-valve position status.
For each functional design, provision for the existing reactor protection system to operate is provided by time delaying the ATWS mitigation l signal. Likewise, automatic arming of each functional design is provided !
by two turbine load signals above a predetermined value and a time delay ;
on de-energizing is used to keep the ATWS mitigation system armed for a preset period even if the existing reactor protection system trips the turbine successfully.
Functional design 1, using steam generator narrow range level as the detection variable, will be used at North Anna. Each of the thrse narrow range level transmitters in each of the three steam generators will be used in conjunction with the first stage turbine pressure channels to derive the ATWS mitigation system. If any two of the three level transmitters in any two of three steam generators are less than or equal to 13 percent of narrow range level span and the turbine is greater than or equal to 40 percent load ATWS mitigation, AMSAC, will be initiated automatically. A time delay of approximately 27 seconds is provided to allow the existing reactor protection system to respond first. In the ,
event of an ATWS event and the expiration of the time delay, the main turbine will be tripped, all three auxiliary feedwater pumps will start, i the steam generator blowdown isolation and sample isolation valves will receive an automatic close signal, and the breakers which supply power for each rod control motor generator will be tripped.
ATWS mitigation by AMSAC is automatically blocked below 40 percent power by a newly installed permissive (C-20) that is derived from the First Stage Pressure (FSP) transmitters. This automatic block will be defeated for approximately 120 seconds following a decrease of FSP below 40 percent. This time delay will be required for the instance wherein an ATWS event occurs and the turbine load reduces causing FSP to drop below 40 percent. The ATWS mitigating actions, AMSAC, will still be 134-BSD-5281NA-6 l ,
initiated automatically if a loss of heat sink (steam generator inventory loss) occurs within'the 120 second time delay.
5.0 SPECIFIC REQUIREMENTS The NRC Staff accepted WCAP-10858 as a generic concept. Consequently, the Staff approved: (1) implementation of any one of the three functional designs for the detection of an ATWS event; (2) use of existing transmitter, impulse lines, transmitter power supplies, and isolators; (3) testing of the ATWS mitigation system in bypass; and (4) the use of an operating bypass, the C-20 permissive, to prevent spurious ,
actuation in either start-up or shutdown.
The Staff also identified 14 key elements which will be reviewed on a case-by-case basis. These 14 key elements, the NRC guidance for each element, and North Anna's positions are discussed below. Engineering is in progress on this system but is not complete, therefore certain changes may be made between details contained in the plans outlined in this letter and the system installed in the plant.
The 14 key elements are:
A. Diversity B. Logic Power Supplies C. Safety Related Interface D. Quality Assurance E. Maintenance Bypasses F. Operating Bypasses j G. Means for Bypassing H. Manual Initiation I. Electrical Independence from Existing Reactor Protection System J. Physical Separation from Existing Reactor Protection System j K. Environmental Qualification !
L. Testability at Power M. Completion of Mitigative Action ]
N. Technical Specifications 134-BSD-5281NA-7 ,,
l A. DIVERSITY-
-)
NRC Guidance )
1 The plant specific submittal should indicate the degree of diversity that exists between the AMSAC equi,, ment 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.
Position Diversity between the existing Reactor Protection System (RPS) and the ATWS mitigation system (AMSAC) to minimize the potential for j common cause failures is required from sensor output, but not including the final actuation device. The instrument sensors do not have to be diverse. Therefore, existing protection system instruments, impulse lines, and transmitter power supplies may be used. Instruments and their related'. impulse lines must be selected to prevent adverse interaction with existing control systems.
North Anna plans to implement Functional Design 1 of WCAP-10858, Steam Generator Level less than or equal to 13% of narrow range level span, to initiate the AMSAC. Attached are preliminary Logic Diagrams which show how Steam Generator Level will be implemented.
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l Each of the three narrow range level instrument loops (Channels I, II, and III) on each of the three steam generators (A, B, and C) ..
will be used to provide isolated non-safety-related signal inputs l to the AMSAC logic system.
An isolated signal from' Channel I, II, and III of steam generator A narrow range level will input to AMSAC Programmable Logic Controller A(PLCA). An isolated signal from Channel I, II, and 1 III of steam generator B narrow range level will input to AMSAC l PLC B, and an isolated signal from Channel I, II, and III of steam generator C narrow range level will input to AMSAC PLC C.
Channel I and II narrow range steam generator A. B, and C level signals provide protection, indication, and computer input functions. The Channel III narrow range steam generator A, B, and C level signals provide protection, indication, computer input, ,
'I and control functions. Channel III narrow range steam generator levels are the only signals that could be compromised by adverse a feedwater control system interaction.
1 If a channel III narrow range steam generator level channel were to be declared inoperable due to either: failing off-scale high, failing off-scale low, or based on a channel check (detecting that the channel had drifted high or low relative'to the other two channels), it would be placed in trip. As a result, the logic for the affected Programmable Controller would conservatively default to a one out of two coincidence.
By using signals from three different channels, i.e., I, II, and III, the level transmitters, their associated impulse lines, and the level transmitter power supplies are electrically and physically independent and, therefore, non-interacting.
Channel III and IV turbine impulse chamber pressure provide protection and control signals. Isolated signals will be used to develop the C-20 permissive. The isolated signals are also used 134-BSD-5281NA-10
to provide control inputs. The implementation of AMSAC does not adversely affect or degrade the turbine impulse chamber pressure signals. !
The independence of the control system signals derived from the protection system is provided by Class 1E qualified isolators.
The signals obtained from the isolators are never returned to the )
protection system. This is consistent with the requirements of I General Design Criterion 24 Separation of Protection and Controi System. Consequently, the use of isolated steam generator level and turbine impulse chamber pressure signals neither compromises the protection system nor introduces an adverse control ' system interaction.
1 North Anna's RPS utilizes a Westinghouse 7300 Process !
Instrumentation and Control System (7300 System). The 7300 System is a voltage based instrumentation system except for the required 1 current interface with the process transmitters. All 11 analog signals will be isolated and provided to AMSAC for the solving of q coincidence logic. Independence of AMSAC from the existing RPS is achieved through the use of e_xisting qualified isolators which j buffer the voltage signals originating in the 7300 System. ]
Connecting AMSAC downstream of the existing isolators will ensure that the non-safety-related AMSAC will not degrade the existing RPS.
Actuation logic diversity will be provided between the RPS and AMSAC. The existing RPS uses a Westinghouse Solid State Protection System (SSPS) to perform this function. The SSPS uses input relays by C. P. Claire, Type GP1, to provide ground inputs to Motorola negative logic gates which in turn solve- the ,
coincidence logic. The negative logic gates drive master output relays which also are C. P. Claire Type GP1, and the master output relays in turn drive slave output relays. The slave output relays are Westinghouse Type AR and provide equipment actuation. AMSAC i will use a programmable logic controller (PLC) to solve the 134-BSD-5281NA-11
1
.g-coincidence logic. The PLC will use a microprocessor manufactured by Intel (8086) which represents an entirely different technology j with respect to Motorola's negative logic gates as used in the SSPS. Input relays will not be required by the PLC. However, output relays are required in order to provide isolated safety-related mitigation permissives to existing final actuation devices. The output relays will b'e Electro Switch Type CSR rotary relays. The SSPS slave output relays are conventional hinged ,
i armature machine tool type relays. AMSAC output relays will use different principles of operation and will be made by a different manufacturer. i i
Therefore, a sufficient degree of diversity is provided through the consistent application of different manufacturers and different operating principles. I B. LOGIC POWER SUPPLIES i
i NRC Guidance The plant specific submittal should discuss the logic power supply design. According to the rule, the AMSAC logic power supply is !
not requi red to be sa fety-related (Class 1E). However, logic I 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 W0G 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.
Position Virginia Electric and Power Company plans to assure independence from the RPS power supplies by utilizing a non-RPS related UninterruptGle Power Supply (UPS) as the AMSAC power 134-8SD-5281NA-12
supply. The UPS will consist of a battery, battery charger, i and inverter, ,
i C. SAFETY-RELATED INTERFACE NRC Guidance The plant specific submittal should show that the implementation is such that the existing protection system continues to meet all applicable safety criteria.
I 1'
Position l
Isolators are the devices which buffer AMSAC from the safety-related equipment and systems. Existing qualified isolators in the 7300 System will provide non-safety-related analog signals to AMSAC. The company will use Electro Switch I Control Switch Relay Series 24 CSR rotary relays, which will be mounted in the top of the AMSAC panel with a steel shelf interposed between the coil section and the contact section, to q provide isolated safety-related AMSAC outputs to actuate safety-related equipment. This approach does r.ot violate any safety criteria applicable to the SSPS, i.e., IEEE Standard.
279-1971, General Design Criteria 20 through 25, and North Anna's UFSAR Section 7.2.
D. QUALITY ASSURANCE NRC Guidance 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."
134-BSD-5281NA-13
Position North Anna's AMSAC complies with the requirements of Generic.
Letter 85-06, " Quality Assurance Guidance for AMSAC Equipment that is not Safety Related."
A detailed response to each of the eighteen specific guidelines follows:
- 1. Organization NRC Guidance i
The normal line organization is expected to verify compliance-with this guidance. . A separate organization is not required.
If desired, the existing Appendix B QA organization may be involved but this is not required.
Position 1
Virginia Electric and Power Company will purchase only portions of the AMSAC equipment as non-safety-related. These purchases will be made in accordance with the Nuclear Operations Department Standards by personnel who are involved i in both safety-related and non-safety-related purchases. The l Appendix B QA organization may be involved as deemed l
appropriate.
II. Program NRC Guidance It is expected that the existing body of plant procedures or practices will describe the quality controls applied to the ;
subject equipment. A new or separate QA program is not required, i
134-BSD-5281NA-14
n
. Position.
Virginia Electric and Power Company will use the existing j program of Nuclear Operations Department Standards and Station- l Administrative Procedures which apply to non-safety-related equipment-for the non-safety-related AMSAC equipment. A new ;
and dedicated QA program will not be implemented. l III. Design Control NRC Guidance MeasuresS/ are to be established to ensure design specifications are included or correctly translated into design documents S/ and to ensure that all design control activities are consistent with the requirements of 10 CFR 50.59. Normal supervisory review of the designer's work is an adequate control measure.
Position Virginia Electric and Power Company has determined- that a portion of the work involved in the installation of AMSAC is safety-related as it involves the interface of safety-related and non-safety-related equipment. As such design work will be
)
controlled in accordance with the standards for safety-related work as identified in the Virginia Power Nuclear Design Control Program.
IV. Procurement Document Control NRC Guidance Measures are to be established to ensure system specifications and quality requirements, where applicable, are included in procurement docunients.S/
134-BSD-5281NA-15
Position Virginia Electric and Power Company through the use of Nuclear Design Control Program, the Nuclear Operations Department Standards, and the Station Administrative Procedures will l ensure system specifications and quality requirements.'are included as applicable in non-safety-related AMSAC' procurement documents.
V. Instructions, Procedures and Drawings !
NRC Guidance Measures are to be established which ensure that quality controls will be applied to activities that affect quality.
These measures may include such things as written instructions, plant precedures, cautionary notes on drawings and special instructions on work orders. Any methodology which provides the appropriate degree of guidance to utility personnel performing quality-related activities will satisfy 1/ Except for design control measures, where the utility is responsible for ensuring that design control measures are applied at contractor or subcontractor organizations, the term " measures" applies only to activities within the licensee's or applicant's organization. However, the design control measures to be applied at contractor or subcontractor organizations need be no more stringent than those required of the utility.
2/ Except for the record keeping requirements of 10 CFR 50.59 and requirements XVII of this guidance document, any records that are generated as a result of implementing these QA controls are not required to be maintained.
134-1SD-5281NA-16 1
this' requirement. Maintenance on the equipment s'all h be based on the appropriate use of vendor information. Any departure.from such vendor guidance shall be based on an y adequate engineering rationale.S/
l Position j Virginia Electric and Power Company will impl ement- this modification through' a separate Design Change Package (DCP) l for each unit. The DCPs are being prepared safety related.
The Nuclear Operations Department Standards provide for this l means of implementation via the Nuclear Design Control Program. Each DCP will be issued by engineering and approved by the Station Nuclear Safety and Operating Committee prior to implementation. Each DCP will also provide procedures, .
instructions and drawings sufficient 'to provide for proper installation and testing. Maintenance information supplied by vendors will be included.
I VI. Document Control i NRC Guidance i
4 Measures are to be established to control the issuance of and changes to documents affecting quality.S/
Position l Virginia Electric and Power Company will control and retain implementation documents in accordance with the Virginia Power Nuclear Design Control Program and will control procurement documentation in accordance with the Nuclear Operations Department Standards.
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I 4
I VII. Control of Purchased Items and Services NRC Guidance Measures are to be established to ensure that all purchases conform to appropriate procurement documents.2/ Such measures may include the performance of receipt inspections by stores or warehouse personnel or plant engineering personnel.
1 Position l
The Company will assure the control of purchased items and/or services for AMSAC which are non-safety-related in accordance with the Nuclear Operations Department Standards which include ;
provisions for inspections as required. I Vill. Identification and Control of Purchased Items NRC Guidance l
Measures are to be established, where necessary, to identify ]
and control purchased items. Examples of- circumstances ]
requiring such control include the storage of environmentally l sensitive equipment or material and the storage of equipment or material that has a limited shelf-life. '
Position The company will assure the identification and control of non-safety-related material purchased for AMSAC in accordance with the Nuclear Operations Department Standards and the Virginia Power purchase order requirements determined in accordance with the Nuclear Design Control Program. No limited shelf ~ife items are included in the present design. ,
134-BSD-5281NA-18 1
IX. Control of Special Processes NRC Guidance Measures are to be established to control special processes, including welding, heat treating, and non-destructive testing. i Applicable codes, standards, specifications, criteria, and i 1
other special requirements may serve as the basis of these controls..
Position '
' j The Company at this time is not planning to use any special processes in the purchase, fabrication or installation of the non-safety-related AMSAC materials. However, work performed- .
at a vendor would be perfctmed as normally done on standard l products or would be in accordance with the purchase order or specification or approved procedure as required for non-safety-related AMSAC.
X. Inspection NRC Guidance l
Measures are to be established to inspect activities affecting quality. Inspections are to be accomplished in order to verify that these activities are in conformance with the available documentation, or, if no documentation is available, to verify that these activities are being satisfactorily accomplished. In general, the line organization is responsible for determining the inspection requirements and for ensuring that sufficient inspections are ;
performed. Inspections need not be performed by personnel who.
are independent of the line organization. Inspections shall be performed by knowledgeable personnel.
134-BSD-5281NA-19
I 1
Position The Company will have inspections performed on non-safety-related ATWS equipment as deemed necessaryLbased on l compliance with the Nuclear Operations Department Standards )
and the Nuclear Design Control Program.
XI. Testing NRC Guidance Measures are to be established to test, as appropriate, non- )
safety-related AMSAC equipment prior to installation and l operaticn and periodically. Results of the tests should be evaluat>d to ensure that the test requirements have been satisfied.
Position The Company will, in accordance with the Final Design Testing Section of the Desiga Change Package as required by the Nuclear Design Control Program, assure that the system performs properly prior to operation. The periodic testing is ;
discussed in Section "L" later in this' response, i XII. Control of Measuring and Test Equipment 4 1
NRC Guidance Measures are to be established to control, calibrate, and adjust' measuring and test equipment at specific intervals.
Position Measuring and Test Equipment will be maintained and calibrated in accordance with Station Administrative Procedures.
l l
l 134-BSD-5281NA-20 1'
XIII. Handling, Storage and Shipping N_RC Guidance Measures are to be established to control handling, storage, shipping, cleaning, end preservation of purchases' .in I accordance with utility practices and manufacturer's recommendations.
Position i
This will be performed in accordance with the Nuclear ]
Operations Department Standards.
XIV. Inspection, Test, and Operatine Status I
NRC Guidance Measures are to be established to indicate status of !
1 inspection, test, and operability of installed non-safety- '
related ATWS equipment. )
Position i The inspection and testing of installed non-safety-related AMSAC equipment will be as discussed in Section "L" of this response. The operating status of AMSAC will be indicated by.
annunciators in the control room and status lamps on the AMSAC panel.
XV. Nonconformances NRC Guidance l
Measures are to be established to identify nonconformances.
134-BSD-5281NA-21
)
l i
Position ;
1 The Company will identify and disposition' nonconformances in accordance with' the Nuclear Operations Department Standards I and Station Administrative Procedures.
)
XVI. Corrective Action System j
NRC Guidance ,
i
! Measures are to be established for prompt correction of )
! a to quality conditions which are adverse (i.e., l nonconformances), and to preclude repetition of conditions )
adverse to quality. -
Position !
The Company will identify and disposition nonconformances in accordance with the Nuclear Operations Department Standards' and Station Administrative Procedures.
i XVII. Records NRC Guidance Measures are to be established to maintain and control records of activities in accordance with the requirements of 10 CFR 50.59. In addition, measures are' to be established to maintain and control appropriate records to ensure that .the requirements specified in the table accompanying the ATWS rule (49 FR 26036, pp. 26042-26043) have been met.
134-BSD-5281NA-22 1
Position The Company will maintain records in accordance with the Nuclear Operations Department Standards and the Nuclear Design Control Program.
XVIII. Audits NRC Guidance Audits which are independent of line management .are not required, if line management periodically reviews the adequacy 1 of the quality controls and takes any necessary corrective action. Line management is responsible for determining I whether reviews conducted by line management or audits conducted by an organization independent of line management are appropriate. i I
1 Position Independent audits are not planned at this time but may be l performed as required. '
E .- MAINTENANCE BYPASSES ;
I NRC Guidance i l
Tne 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.
Position AMSAC maintenance during unit power operation will be accomplished through operation. of either of two bypass switches. One is located in the Main Control Room on l Benchboard Section 2 and the other is located within the AMSAC 1
I 134-BSD-5281NA-23
L
, s. I I
i i
panel. In.neither case will the lifting of. leads, tripping of. 1 breakers, use of physically blocking relays, nor the pulling of - 'l fuses be required to bypass AMSAC. Bypass status. will be :)
annunciated in the Main Control Room above Vertical Section 1.s l The alarm will be located to provide bypass status to the !
reactor operator. The new alarm will meet accepted human I factors guidelines'as delineated in Virginia Electric and Power Compa ny 's Human factors Standard STD-GN-0005. In accordance ;
with the Virginia Electric and Power Company Nuclear Design Control Program, a review of the human factors. acceptability of j this modification will be performed and' the results will .be l noted in the implementation document. I For maintenance bypass the following human factors principles will be implemented:
1
- 1. The information provided by displays and control equipment added to the Main Control Room as a result of implementing the AMSAC Final Rule will not increase the potential for operator error under both normal and abnormal plant conditions. Bypass for maintenance will be clearly displayed to the operator.
- 2. AMSAC will be integrated into the applicable Emergency Operating Procedures and the- applicable . Maintenance Frocedures.
- 3. AMSAC will be integrated into the operator training program and the North Anna simulator will also be modified l to incorporate the implementation of AMSAC.
l
- 4. AMSAC will be time delayed to allow the existing reactor protection system to respond first. Consequently, the alarm "AMSAC ACTUATED" should always be received after the l
existing reactor protection system commences mitigation. <
Since AMSAC will be installed to mitigate a failure of the ' '
RPS, the AMSAC ARMED and AMSAC ACTUATED alarms may be prioritized. During normal operation the operator will be .
i 134-BSD-5281NA-24 ;
\ .
> 1 1
.i trained to expect AMSAC BYPASSED, AMSAC TROUBLE, and AMSAC ARMED alarms. As.AMSAC BYPASSED and AMSAC TROUBLE wili be 1
status alarms, prioritization may not be required. The j alann AMSAC ARMED will be a pre-trip annunciation which l I could prompt operator responses and prioritization may be '
required. This will be reviewed; however, due to the 3 brief time delay, 27 seconds, operator action based on this alarm is not expected.
F. OPERATING BYPASSES NRC Guidance l I
i The plant specific submi ttal should state that operating bypasses are continuously indicated in the control room; provide the basis for the 70 percent or plant specific operating bypass level; discuss the human factors design l aspects of the continuous indication; and discuss the diversity and independence of the C-20 permissive signal (defeats the ]
block of AMSAC).
Position The design bases for the new AMSAC unique C-20 permissive, which defeats the operating' bypass, two out of two turbine i
first stage pressures increasing, are:
l l 1. " Westinghouse Anticipated Transients Without Trip Analysis," WCAP-8330 August 1974. )
- 2. " Anticipated Transients Without Scram for tight Water Reactors," NUREG-0460, December 1978.
- 3. Anderson, T. M. , "AMSAC Submittal," Westinghouse Letter i NS-TMA-2182 to S. H. Hanauer of the NRC, December 1979.
1 134-BSD-5281NA-25
l These three documents demonstrated that ATWS mitigatior need not be initiated below 70 percent turbine load because reactor .
coolant system pressure does not approacn the ASME Boiler and I Pressure Vessel Code Level C Service Limit of 3200 psig (NRC criteria for successful ATi- 4>-
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