ML19220A732
| ML19220A732 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 08/08/1974 |
| From: | Stello V US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Moore V Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7904240188 | |
| Download: ML19220A732 (17) | |
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p-Cocket No. 50-320 V. A. Mcore, Assistant Director for LWRs, Group 2. L METROPOLITAN EDISON COMPANY, JERSEY CEhTRAL POWER & LIGHT CGMPANY AND PENNSYLVANIA ELECTRIC CG4PANY - TiiREE MILE ISLAND NUCLEAR STATION, UNIT 2; FIRST RCUND QUESTIONS Plant Nare: Three Mile Island 2 Licensing Stage: Operating License Oceket Nu::ter: 50-320 Responsible Branch and Project Manager: LVR 2-2, B. Washburn Technical Review Branch Involved:
EISCS Branch Requested Co=pletion Date:
August 2 1974 Applicant's Response Date Necessary f'or Complet.it..: of Next Action Planned on Project: October 18, 1974 Description ofRResponse: First Set of Questions Review Status: Awaiting Inform
- tion Enclosed is a listing of the first set of questions prepared by the L:RS. Electrical, Instrumentation and Control Systems Branch, for transmittal to the applicant. This set of questions reflects the results of our review through Amendment 18.
It should be noted that Three Mile Island Unit 2 differs in scme respects (larger standby diesel generators) from Unit 1.
Also some design changes (Auxiliary Feedwater Systaa) and additional documentation will be required to meet present review standards.
Significant review effort appears necessary in the following areas:
1.
Seis:nic and environmental qualification of safety-related equipment..
2.
Conforsance to applicable Standards and Regulatory Guides.
3.
The Auxiliary Feedwater System design.
4.
Diesel generator controls and qualification testing.
5.
Conformnce of the Class I Electric systrsts to the siggle failure criterion and to Regulatory Guide 1.6.
O n e:nal si::ce bz
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. Victor Stello. dr., Assistant Director
. for Iteactor Safet:4 Directmte cf Licensing ea rs *
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Form AEC.518 I Rev. 9 53) AZCM 7240 W u. s: oovuoseusset seenvisie oawicri s.te. ese.n ee e.
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N EtiCLOSURE THREE MILE ISLATID Ut4IT 2 tiUCLEAR POWER STATIOil FIRST SET OF QUESTI0tlS
22-1 22.0 Electrical, Instrumentation & Control Systems 22.1 Supplement the information contained in FSAR Section 3.10, (3.10)
Seismic Design of Category I Instrumentation and Electrical Equipment, as follows:
(1) Provide a summary listing (tabulation) correlating all safety related electrical equipment, equipment locations, seismic design bases at each location, seismic qualification method used (test and/or analysis) and seismic test and/or analysis results.
This should include the 4.16 kV switchgear, 480 V switchgcar, unit substations, and motor control centers; 120 v a-c system cccconents; 125 v d-c system components including batteries, battery racks, battery chargers, distribution centers, and panelboards; static inverters; process control equipment; protection and safeguards actuation racks; nuclear instrumentation; electrical penetration assemblies; motor operated valves; diesel generators etc.
(2) Provide a more detailed description of the seismic qualification method (test and/or analysis) used for each Class I component and module. This description may be incorporated in the summary listing of (a) above.
(3) Confirm that the seismic qualification testing demonstrated the capability to change state or operate during a SSE for all components and medules which are required to so operate in performance of their design safety functions. Provide the bases for the methods of simulating the net effect of the design basis seismic event which were used in the qualification tests.
(') Verify that the auxiliary equipment (local control panel, lube oil system, etc.) which is required for the operation of the emergency diesel generators has been seismically qualified. Describe the testing and/or analysis performed to seismically qualify this equip-ment, and state the results of this testing and/or analysis.
(5) With reference to the seismic qualification methods described in BAW-10003 Revision 3, confirm the items stated below:
(a) The responses of cabinet assemblies at the var 1ous instrument or device mounting locations due to actual earthquake distur-bances are determined either by testing or analysis.
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22-2 (b) The maximum response detemined in (a) above at the instrument or device mounting locations are less thar 1.
9 (c) The instrument or device responses resulting from a coa.tinuous sine wave test input are shown to be equivalent to the responses due to actual earthquake disturbances in magnitude and frequency content.
(6) Table 5-2 of BAW-10003 provides a list of instruments "
to be used in the Nuclear Instrumentation, Reactor Protection and Safety Features Actuation systems.
Identify the suoplier of the cable assemblies and cable connectors to be used with these instruments and provide a detailed description of the seismic qualification (test and/or analysis) for these assemblies and connectors.
22.2 Supplement and revise the information in Section 3.11 (3.11) regarding environmental qualification of safety related components located in the primary containment to include the folicwing:
(1) Provide a concise statement of the limiting DBA environmental conditions in the containment. This should include temperature, pressure, humidity, radiation, and chemical environment.
(2) State the length of time from occurrence of the DBA that each component is required to operate in order to perfom its design safety function.
(3) Describe the environmental qualification testing performed for each component and identify the applicable test documentation. Provide this test documentation if it has not been previously submitted.
Your resconte should (a) state whether the tests were perfomed on prototype equipment, (b) contain sufficient detail to permit a direct ccmparison between the test anditions and the limiting DBA conditions (superimposed on normal aging) for all parameters, and (c) discuss the adequacy of the environ-mental qualification for each component.
(4)
If environmental qualification is based (in whole or in part) on the analyses or on use of data from tests perfomed on other than prototype equipment, describe and justify each instance of the use of these methods and justify the applicable documentation.
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22-3 22.3 The system design for the ccmbustible gas control systems (6.2.5.2)
(hydrogen recombiner system and as a backup, the hydrogen purge system) as described in Section 6.2.5.2 does not contain sufficient information to ascertain the adequacy of the instrumentaticn, control and electrical equipment for these systemr Therefore, provide this description in sufficient detail to pemit an evaluation. This description should include the folicwing:
(1) Explicitly, identify the source of power for the liydrogen recombine unit or,assumina this unit is inoperable,th source of power for the hydrogen control exhaust fan.
(2) Identification of any instrumentation, control or electrical equipment which may be common to both the hydrogen recombiner system and the hydrogen purge system.
22.4 The design of the motor-operated isolation valves for the (6.3.2.15) core flooding tanks (as described in Section 6.3.2.15)do not provide sufficient assurance that these valves will open when required. An acceptable variation of your design would include the following features:
(1) Valve position visual indication (open or closed) for each valve which is not d.apendent on power being available to the valve controller.
(2) Redundant visual and audible alams for each valve when the valve is not fully open and reactor coolant pressure is above a preset value. These alarms shall be actuated by redundant and independent valve rsition sensing circuitry including at least one position sensor sensing actual valve position, and by redundant and independent pressure signals.
(2) A Technical Specification requirement that the reactor shall not be made critical or shall be shut down unless the motor operated isolation valve in the discharge line of each cara flooding tank is open and the breaker supplying power to the valve operator is locked open and tagged.
Please indicate your plans and schedule to modify the design of the isolation valving to include the preferred features stated above or to conform to other criteria that provide equivalent assurance that these valves will be open when required.
22-4 22.5 (RSP)
Based on the infor ation presented in Section 6.3.2.17, we (6.3.2.17) believe that:
(1) The proposed design of the circuits used to change over to the cross-T/ar nada (using LPI aumps as boosters for the HPI Pumps) and tne recirculation mode of operation following a LOCA does nec confcra to the requirements of IEEE Std 279-1971, and (2) The complexity of the proposed change-over may not provide adequate assurance that the operator will correctly perform the required actions.
The Staff's position is that the instrumentation and controls provided to acccmplish the change-over to the recirculation mode and the cross-over mcde should be designed to meet the requirements of IEEE Std 279-1971 including the require-ments for automatic and manual initiation of protective actions at the systems level.
In addition, the procedures should be of such simplicity as to provide a high degree of assurance that the operator will perform correctly all actions that are necessary to protect the health and safety of the public.
Therefore, either modify your design to shew conformance with the Staff's position, or justify your design, as opposed to a design which provides automatic and system level manual initiation of switch cver as required by the literal interpretation of IEEE Std 279-1971.
Justification of the present design should include the folicwing:
(a) Define the time av'.ilable for the operator to complete the necessary monitoring and switching functions for the cross-wer node (i.e., using LPI pumps as boosters for HPI pumps) and tte recirculation mode prior to enset of conditiona (assut:e and define worst case) that are unacceptable frra the scandpoint of plant safety.
(b) Provide a description of the control panei arrangement of the indicators and switches which the operator must monitor and operate to affect switchover.
(c) Describe the permissive interlocks that are provided (for equipment protection or other reasons)between the various ECCS components that are operated during switchover.
(d) Identify the system conditions that require the use of the cross-over mcde prior to or coincident with the initiaticn of the recirculation mode of operation.
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22-S 22.6 Supplement the information contained in Section 7.0 (7.0 and 15.0) wi' regard to the Reactor Protection System (RPS),
Control Red Drive Control System (CRCCS) or wnere appropriate to the Safety Featur e Actuation System (SFAS) as folicws:
(1 ) Table 7.1-2 indicates that the RPS meets the require-ments of IEEE Std 279-1968.
Icentify those features of the RPS design which do not meet the requirements of the 1971 version of this standard. Your response should specifically address conformance to Sections 4.7, 4.17 and 4.22 of IEEE Std 279-1971.
(2) Section 7.1.1.2 of the FSAR indicates that no supporting systems are necessary for the safe operation of the RPS. Discuss the failure of varicus Feating, Ventilating and Air Conditioning Systems located throughcut the plant and identify thoss (if any) who's failure could result in the RPS not performing its design function. The portions of the RPS considered should include (as a m.inimun) the process-to-sensor coupling, sensors,instfument channels, decision logic, and actuation devices.
(3) The information presented in Sections 7.1.2.8 and 7.2.2.4 is not sufficient to demonstrate conformance of your RPS design with all of the regulatory positions of Regulatory Guide 1.22 (Safety Guide 22), particularly Positions 1(a),1(b) and 3(a). This information should be provided in the detail required for this evaluation. Your response should address conformance of your design with all the positions of Regulation Guide 1.22 and partirularly including a description of the interlocks which preclude the bypassing of more than one RPS channel.
(4) Sections 7.'.
and 7.2.2.2 state that the RPS does not comply wi.a IEEE Std 338-1971, Trial-Use Criteria for the Pericdic Testing of Nuclear Power Generating Station Protection Systems, because this standard was issued subsequent to equipment procurement.
This standard is primarily concerned with periodic testing not with system design. Define the degree of conformance of the test and surveillance program for the RPS and SFAS embodied in the Technical Specifications (Section 16.0) with the provisions of IEEE Std 338-1971.
Identify any system design features that preclude testing in conformance with this standard.
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22-6 (5)
>rovide a nore detailed description of the RPS manual trip switches including tne'r installation.
Your response should address those features of the design which implement the separation and independence requirements for redundant safety circuits.
(6) Table 15.1.1-1 of the FSAR indicates that thc maximum rod withdrawal speed (used for the analysis of the startup accident) is 30 in/ min. Mcwever, it appears from Figure 7.7-1 of the FSAR that if the programmer motors were to overspeed, this maximum withdrawal speed could be exceeded.
Provide a discussion of the design features, other than the use of synchronous programmer motors, that would limit rod withdrasal speed to 30 in/ min.
(7) Describe the methods for periodically testing the RPS response time for the trip parameters. Of concern is the time history changes fer these responses times.
Include a discostien cf the response times in relation to the safety limits and state the worst case margin in terms of time.
(8) Describe the means by which a control rocm operator is appraised of a reduction in engineered safety features redundancy on an overall systems basis.
22.7 With regards to IEEE Std 336-1971, Installation, Inspection (7.1.2.6) and Testing Requirements for Instrumentation and Electric Equipment During the Construction of Nuclear Power Generating Stations, it is not clear frcn the discussion presented in Section 7.1.2.6 whether the requirements of this standard have been or will be met for the installation, inspection, and testing procedures for instrumentation, sensing lines, electrical 'and instrumentatie enetrations, cabling and racewa/s, switchgear and panel.
"ide an additional discussion defining the degree c c
mance to the require-ments of this standard for the abov, items.
22.8 Section 7.3.1.1.4 of the FSAR indicates that certain valves (7.3.1.1) in the containment isolation system are provided with air storage resevoirs.
Identify all valves in the isolation system which are of this type and provide a typical control diagram for these valves. Also, describe the degree of conformance of the instrumentation and controls for these s to the safety criteria.
This should include a sion of their testability, redundancy and power rs t remen ts.
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22-7 22.9 Ti. FSAR (Section 7.3.2.lp) indic,
.s that the SFAS and (7.3.2.1 )
ESF systems are designed so that once a protective action is initiated, it continues until.i t is halted by deliberate operator action.
Secticn 4.16 of IEEE Std 279-1971 states that "The protection system shall be so designed thate cnce initiated, a protective action at the system leve? shall go to ecmpletion. Return to operation shall require subsequent deliberate operator action." Confirm that in ynur design a protective action at the system level will go to comoletion once initiated. Identify and justify any exceptions.
22.10 Incidents have occurred at a nuclear power plant that (None) indicate a deficiency in the control circuit design.
These incidents involved the inadvertent disabling of a component by racking out the dircuit breaker for a different component.
As a result of these occurrences, we request that you perform a review of the control circuits of all safety related equipment at the plant, so as to assure that disabling of one component does not, through incorporation in other interlocking or sequencing controls, render other components incoerable. All modes of test, separation and failure should be considered.
Also, your procedures should be reviewed to ensure they provide that,whenever a part of a redundant system is removed from service, the portion remaining in service is functionally tested innediately after the disabling of the affected portion and,if possible, before disabling of the affected portion.
2.11 State the extent of conformance of the safety systems (None) to Regulatory Guides 1.40, 1.41, 1.47, 1.68, 1.73, 1.75 and 1.80.
With regard to Regulatory Guide 1.47:
(1) The conditions of positions 3(b) and 3(c) are interpreted to include bypasses that result from manipulaticn of permanently installed electrical control devices located in any accessible area of the plant, and (2) The design criteria for the indication system should reflect the imoortance of both providing accurate infomation for the operator and reducing the possibility.for the indicating equipment to affect adversely the monitored safety systems.
In discussing the Three Mile Island Unit 2 design criteria, the following should be considered:
(a) The bypass indicators should be arranged
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to enable the operator to assess readily e
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g 22-8 the operating status of ecch safety system and determine whether continued reactor operation is permissible.
(b) Means by which the operator can cancel erroneous bypass indications, if provided, should be justified by demonstrating that the postulated causes of erroneous indications cannot be eliminated by another practical design.
(c) Unless the indication system is designed in conformance with criteria established for safety systems, it should not be used to perform functions that are essential to the health and safety of the public. Neither should administra'ive procedures require immediate opere 'or actions based solely on the bypass
- indi, ions.
(d) The indication systc... should be designed and installed in a manner which precludes the poss-
'ity of adverse effects on the plant's sa. % sfi c.
Failure or bypass of a protection 5
fu:'i ? .JuM t L.9 a Credible Consequence or fa sur
. curring in the indication equipment and thL, v"
'ndication shoulo not reduce the r e'. ui re in.:gndence between redundant safety
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(e) The indication system should include a capability of ussuring its operable status during normal plant operation to the extent that the indicating and/or annunciating function can be verified.
(f) Means provided in the control room for manually activating system level bypass indications, and the associated p"ocedural or administrative controls, should be described.
22.12 Table 7.5-1 which lists the information readouts available (7.5.1.2) to the operator for monitoring conditions in the plant should be revised or supplemented to include the following:
(1) Number of available channels.
(2) Number of required channels.
(3) Type, number and location of readouts (indicator / recorder).
(4) The purpose or operator usage of each of the monitored parameters during normal and accicent or post-accident condi tions.
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22-9 22.13 Provide the criteria and design bases which established the (None) heat tracing requirenent, tenperature control, monitoring, and pcwer requirements for the boric acid tanks, borated water storage tanks, and spray additive tanks, and related piping of the enemical addition systen. Discuss the consequences of a single failure in the heat tracing temperature control or instrumentation of each of the above mentioned systems.
22.14 Provide the results of a review of your operating,
(:;ane) maintenance and testing procedures to determine the extent of usage of jumpers or other temporary forms of bypassing functions for operating, testing or maintenance of safety related systems.
Identify and justify any cases where the use of the above methods cannot be avoided.
Provide the criteria for any use of jumpers for testing.
22.15 (RSP)
We have concluded f rcm the informaticn presented in (6.0, 7.0 and the FSAR concerning the Auxiliary Feedwater System 15.0)
(AFS) that this. system is. essential to plart safety and must be capable of satisfying its functional requirement after sustaining a break in its piping inside containment and a single electrical failure.
We will require that the instrumentation control, and electrical subsystems associated with the AFS be designed to conform to IEEE Std 279-1971 and IEEE Std 308-1971. Therefore, (1) Modify your design of these subsystems to conform to these standards and. criteria or justify the present design on some other defined basis.
(2) 'If your design is modified, provide a sufficiently detailed description, in;luding design bases and supporting analyses, to enable evaluation of the new design for conformance with the stated criteria and standards.
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O 22-10 22.16 (RSP)
The Staff has recently identified a concern with regard (None) to the application of the single failure critarion to manually-controlled electrically-operated valves.
It has been concluded that where a single failure in an electric system can result in loss of capability to perfom a safety function, the effect on public safety must be evaluated. This is necessary regardless of whether the loss of safety function is caused by cn active ccmponent failing to perform a requisite mechanical motion, or by a passive component perfoming an undesirable mechanical mo ti on. The following Staff position presents an acceptable means for meeting the single failure criterion with regard to this type of single failure.
(1) Single failures of both active and passive ccmoonents in the electric systems of valves and other fluid system ccmponents should be considered in designing against single failures, even thcugn the fluid system component may not be called upon to function in a given safety system operational sequence.
(2) Where it 1s determined that failure of a single active or passive ccmponent in an electric systen can causa mechanical motion of a passive component in a fluid system and this motion results in total ~ loss of the system safety function, it is acceptable, in lieu of design changes that also may be acceptable, to disconnect pcwer to the electric systems of the component.
The plant technical specifications should include a list of all electrically-cperated passive valves, and the required positions of these valves, to which the requirement for removal of electric power is applied in order to satisfy the single failure criterion.
O) Electrically-operated valves which are classified as active valves, but which are manually-controlled should be operated frca the main control rocm. Such valves may not be included among those valves from which power is removed in order to meet the single faiiure criterion unless: ( a) electric power can t:e restored to the valves frem the main control rocm,( b) valve operation is not necessary for at least 10 minutes following cccurrence of tne event requiring such operation, and (c) it is demonstrated that there is reasonable assurance that all necessary operator actions will be perfomed within the time shown to be adequate by the analysis. The plant technical specificacions should include a list of the required positions of manually-contrclied, electrically-operated valves and should identify those valves to wnich the requirement for removal of electric power is applied in order to satisfy the single failure criterion.
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22-11 (4) Wnen the single failure criterion is satisfied by removal of electric power frca passive valves or frca active valves meeting the recuirements of (3),
above, the associated valves should have redundant position indication in the main control rocm and the position indication system should itself meet the single failure criterion.
(5) The phrase " electrically-operated valves" includes both valves operated directly by an electric device (e.g., a motor operated valve and a solenoid-operated valve) and those valves operated indirectly by an electric device (e.g., an air operated valve who:e air supply is controlled by an electric solenoid valve).
Therefore, please provide:
(a) An evaluation of all safety related' fluid systems to identify all valves whose fcilure can result in the total loss of the system function.
(b) A description of the means provided to meet the single failure criterion in safety related fluid systems where it is identified that a single failure will result in total loss of that system function.
22.17 Section 8.3.1.1.5 indicates that a synchronizing signal is (8.3.1.1 )
provided to the inverter frcm the standby regulated power source. Provide a discussion of the consequences of the loss of this synchronizing signal and any indicators which may be available to indicate the signal loss.
22.18 Provide the follcwing additional informaticn with regards to (8.3.1.1) the Diesel Generator (D/G) controls and interlocks.
(1) Section 8.3.1.1.S.3 states that several permissive interlocks are to be satisfied for oceraticn in the fully automatic start mcde. Describe the indications available to the operator to appraise him of the status of these permissive conditions.
(2) With regard to the information presented in Section 8.3.1.1.8.2e, describe the capability for testing the two out of three lube oil pressure (low) or two out of three crankcase pressure (high) coincident circuits used for tripping the D/Gs.
(3) Provide the design basis for the shutdcwn reset push-button permissives (item a5 of Section 8.3.1.1.8.3) for operation in the fully automatic start mode.
(4) Describe che permissive function of the start and stop E.SMlE9
pushbutton (item b of Section 3.1.1.8.4) on the diesel generator control pane ~t in the control rocm.
It appears that this item is inconsistent with the opening sentence of this section.
22.19 The FSAR states that all essential motor circuits nave (8.3.1.1) thermal overload devices and that small motors, below 5 HP are also equipped with a thermal switch built intc the motor.
Discuss the ef fect of low bus voltage (e.g., during diesel operation) on motor torque and the resultant possibility for 0/L trip of motor operated valves prior to completion of their stroke.
Also, discuss any provisions for bypassing these 0/L's especially during emergency conditions.
22.20 (RSP)
The standby diesel generators (3000 kW continuous)
(8.3.1.2) described in Section 8.3.1.2.3b are in a size range that has not been previously qualified for use in nuclear power plants.(It is noted that the diesel generators provided for Three Mile 1sland 'Jnit 1 are rated 2500 kW continuous.)
We will require qualification testing for these un'ts similar to that performed on the Zion 4000 kW diesel generators. An. acceptable test program would include the following requirements.
(1) At least two tests acceptable to the Staff shall be performed.on each diesel to demonstrate the start and load capability of these units with some margin in excess of the design requirements.
(2) Prior to initial criticality, performance of at leatt 300 valid start and load tests, with no more than three failures allcwed. This would include all valid tests performed offsite.
(A valid start and load test shall be defined as a start frca design cold ambient conditions with loading to at least 50% of the continuous rating within the required time intarval, and continued operation until temperature equilibrium is attained.
(3) A failure rate in excess of one per hundred will require further testing as well as a review of the system design adequacy.
State your intent vith regard to meeting these requirements and provide a datriled description of your test program.
22.21 Provide a description of the switchyard batteries (None) installation. This description should include a discussion ci the independence of these power supplies.
GS 230
22-10
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22.22 Verify that the batteries serving the protection systems (8.3.2.2) loads are independent Seismic Class I installations housed in separate rooms.
Describe these installations in more detail including the bettery room ventilation systems.
22.23 (RSP)
The information presented in the FSAR concerning the battery (S.3.2.2) test program is not complete. We require that your design criteria include:
(1) The " Procedure for Battery Capacity Tests," as specified in Section 5 of IEEE Std 450-1972, and (2) The frequency of " Performance Discharge Test," as specified in Section 5.3.6 of IEEE Std 508-1971 unless a demonstratable technic:1 oasis can be established for a greater interval between performance tests.
Revise the FSAR to include the above requirements and justify any exceptions taken. Also, specifically address the capability of the batteries (assuming nc other available sources) to perfarn their intended safety functions for the indicated degraded condition of 203 volts across the battery.
22.24 Regulatory Position ad of Regulatory Guide 1.6 (Safety (None)
Guide 6) requires that at least one. interlock be provided to prevent an operator error that would parallel redundant power sources. Pescribe the interlocks which prevent an operator from paralleling the emergency diesel generators by manually closing the folicwing breakers.
(1) Manually closing breakers TlE-2E-2 and T2E-lE-2 or breakers T3E-4E-2 and T4E-3E-2 to parallel buses 2-lE and 2-2E or buses 2-3E and 2-4E respectively (reference Figure 8.3-3).
(2) Manually closing breakers Tile-21E2 and TELE-llE2 or breakers T12E-22E2 and T22E-12E2 to parallel buses 2-llE and 2-21E or buses 2-12E and 2-22E respectively (reference Figure 8.3-5).
(3) Manually closing breakers T31E-41E-2 and T41E-31E-2 to parallel buses 2-31E and 2-41E (reference Figure 8.3-7).
It is noted that for most of these pairs of tie breakers, one of each pair is shown on the respective Figures as being normally closed (NC). With these tie breakers normally closed, the scheme does not meet single failure.
Confirm that these breakers are all normally open or justify your design on scme other defined basis.
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llh 22-14 22.25 Your design provides alternate feeds from redundant buses (None) for make-up pump 1B (reference Figure 8.3-3), reactor building air coviing fan "C" (reference Figure 8.3-5),
nuclear service closed cooling pump "NS-P-1C" and make-up pumo "13" aux liaries (reference Figure 8.3-13), the i
mechanical trash rack SW-S-lC, the traveling screen SW-S-2C and the screen wash pump discharge strainer SW-5-3 respectively (reference Figure 8.3-7).
Describe the interlocks and administrative contr,ls required for these circuits in order to meet the single failure criterion and position 4d of Regulatory Guide 1.6.
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