ML19312C642
| ML19312C642 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 10/06/1969 |
| From: | Moore V US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Boyd R US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 7912190850 | |
| Download: ML19312C642 (12) | |
Text
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OCT
- 1969 1
Roger S. Boyd, Assistant Director for Reactor Projects, DEL THRU: Saul Levine, Assistant Director for Reactor Technology, DRL THE DUKE POWER COMPAIIT OCONEE UNITS 1, 2, AND 3, QUESTIONS RELATING TOINSTRUMENTAT10tt,CONTROLANDPOWER,DOCIETMmC50-269)-270,-287 N
Please include the enclosed questions among those now in preparation for transmittal to the applicant. The review was performed by Messrs. O. D. Parr and R. D. Follard of the I&PT Branch.
We wish to point out that no provision has been made to detect i
abnormal neutron flux distributions, other than the incore neutron detectors. This portion of the Ocones design is a significant departure from previously approved designs.
Original signed by l
Voss A.Mccre V. A. Moore, Chief Instrumentation & Power RI-744A Technology Branch DRL:I&PTB:RDF Division of Reactor Licensing
Enclosure:
Questions cc w/cac1:
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Form AEC-318 tRev. 9-53) AECM 0240 m s. novameuver peinvene aerica, ises o-aee..it 7912190ffC
OCT 6 1969 OCOIIEE talITS 1, 2, AMD 3 QUESTICIIS RELATING TO INSTRUMENTATICII, CONTROL, AMD POWER 1.
Page 7-1 of the FSA1 states that the protective systems which actuate reactor trip and engineered safety feature action are designed to meet the intent of IEEE 279. Please identify any features of the design which differ from the criteria of IEEE 279 and the Comunission's General Design Criteria.
Explain the reasons for any differences.
2.
What are yout seismic design bases for the reactor protection system (RPS), the emergency electric power system, and the instrumentation and controls for both the engineered safety features (ESP) and the decay heat removal system? Will the systems be designed to be capable of actuating reactor trip or engineered safety feature action during the maxinema peak acceleration? If a seismic disturbance occurred after a major accident, would emergency core cooling be interrupted? Whsc tests and analyses will be performed to assure that the raismic design bases are met? What seismic specifications are eoployed in the instrumentation and control purchase order (s)7 3.
Please describe the quality control procedures which i:pply to the equipment in the 375, the ESF and containment isolation Thia systems, and the associated emergency power systems.
description should include, but not rescessaruy oe lumned w.
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SURNAME >
DAM >
Form AEC-Sis iRev. 9 53) AECM 0240 as,,,,,,,,,,,,,,,,,,,g,,,,
Oconee 1, 2, & 3 2
OCT 6 1969 (a) quality control procedures used during equipment fabrication, shipment, field storage, field installation, and system component checkout; and (b) records pertaining to (a) above.
4.
Pages 7-8, 7-10, 8-9 and 8-10 of the FSAR present some information concerning electrical installation criteria but not in sufficient depth to allow an evaluation of the insta11stien of the reactor protection system.
Please submit your cable installation design criteria for preserving the independence of redundant RPS and ESF circuits (instrumentation, control and pcwer).
For the purpose of cable installation, the protection system circuits should be interpreted in their broadest sense to include sensors, instrument cables, control cables, and power cables (both a.c. and d.c. ), and the actuated devices (e.g., breakers, valves, pumps).
Cable separation should be considered in terms of space and/
a.
or physical barriers between redundant cables. Please address (1) the separation of power cables from those used for control and instrumentation, (2) the intermixing of control and instru-ment cables within a tray (or conduit, ladder, etc.), (3) the intermixing within a tray (or conduit, ladder, etc.) of cables for different protection channels, and (4) the intermixing of non-vital cabling with protection system cabling.
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SURNAME >
DATE>
form AEC-518 i Rev. 9-53) AECM 0240 g,
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i oconee 1, 2, s. 3 3
OCT 6 1969 b.
What are your criteria with respect to (1) tha separation of penetration areas, (2) the grouping of penetrations in each area, and (3) the separation of penetrations which are mutually redundant 7 Please discuss cable tray loading, insulation, derating, c.
and overload protection for the various categories of cables.
d.
Flesse discuss your criteria with respect to fire stops, protection of cables in hostile environments, temperature moni-toring of cables, fire detection, and cable and wireway markings.
e.
Please discuss the administrative responsibility for, and control over, all of the foregoing (a - d) during design and installation.
f.
Flesse discuss your design criteria for locating the process instrumentation inside containment to include (1) separation of redundant sensors and sensing lines, (2) protection provided for detectors and sensing lines, and (3) protection provided to cables between sensors and electrical penetrations.
OFFICE >
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Form AEC-Sto tRev. 9-53) AECM 0240 a s.
,s.. set nienne amca 1 imao-aw.cr
oconee 1, 2, s. 3 4
OCT 6 1969 5.* What are your design criteria for reactor protection system and engineered safety feature related elsetrical and mechanical equipment located in the containment, or elsewhere in the plant, which take into account the potential effects of radiation on these co g nents due to either normal operation or accident conditions (superimposed on long-term normal operation)?
Des-cribe the analysis and testing performed to verify compliance with these design criteria.
6.*
Please identify all equipment and components (e.g., motors, cable, filters, pump seals) located in the primary containment which are required to be operable during and subsequent to a loss-of-coolant or a steam-line break accident. Describe the qualifications tests which have been or will be performed on each of these items to insure their availability in a combined high temperature, pressure and humidity environment.
7.
What criteria have you established relative to assuring that loss of the air conditioning and/or ventilation system will not adversely affect operability of safety related control and electrical equipment located in the control room and other equip-ment roomsi Describe the analysis performed to idencify the
- These questions relate to the engineered safety feature chapter of the FSAR and should be forwarded to the applicant with other questions concerning that chapter.
OFFICE >
SURNAME >
om,
l Form AEC-Sla (Rev. 9-53) AEC.\\t 02+0 as.,,,,,,,,,,,m,,,,,,,,,,,,,,,,,,,,
I v - -
ocone. 1, 2,4 3 5
OCT 6 1969 worst case environment (e.g., temperature, humidity). What is the limiting condition with regard to tempers cure that would require reactor shutdown, and how was this determined 7 Describe any testing (factory and/or onsite) which has been or will be l
performed to confir:a satisfactory operability of control and
)
electrical equipment under extreme environmental conditi ma.
1 8.
Describe how reactor protection system and engineered safety equipment will be physically identified as safety equipment in the plant.
9.
Sactions 1A.11 and 7.4.5 of the FSAR discuss the capability of maintaining the reactor in a safe shutdown condition if access to the control room is lost. The FSAR appears not to discuss obtaining this condition, as required by GDC 11, from outside l
the control room. Please document your intention with respect to this criterion and give your basis for auf departure from it.
10.
Pages 7-41 and 7-44 of the FSAR discuss the Control Rooms.
Please provide the following additional information:
What communication systeem are available to the Centrol a.
Rooms for:
(1) Special purpose (e.g., sound powered phones);
(2) Emergency (e.g., provide a brief discussion of the LM 21m== :ptr);
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Oconee 1, 2, & 3 6
00T 6 1969 b.
What facilities for emergency lighting are provided?
- 11. With regard to the bypass of the reactor coolant pressure actuation signal in the HP and LF Injection Systems, please supply the following additional information:
De number, type, and operation sequence of switches used to a.
initiate the bypass in each system.
b.
The indication available to the operator that each bypass has been and/or is capable of being actuated.
The justification for the selection of the 1900 psi setpoint c.
used in the HP Injection system.
Since the reactor trip occurs at a lower pressure, it appears that reactor operation is possible with the HP Injection system bypassed.
d.
The provisions made to enable the operator to remove each l
bypass below its respective automatic removal setpoint.
f 12.
Describe theinformation available to the operator which will inform him that a test is in progress on either a RFS or ESF channel.
In addition, describe the indication available to identify,down to the channel level, the instruments which caused a particular pro-tactive action to occur. These descriptions should be in sufficient detail to permit a determination of the system's compliance with Sections 4.13 and 4.19 of IEEE 279.
OFTICE >
SURNAME >
DATE >
Form AEC-318 (Rev. 9-33) AECM 0240
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,as eser reinvise or, ca. me e-weir
Oconee 1, 2, 4,3 7
OCT 6 1969
- 13. The typical control circuit for an ESF pump motor (Figure 7-4) appears not to agree with the statement (Page 7-15) that power for operating the C1 relays is taken from the power source of the associated device (pump, valve, etc.).
In addition, the typical control circuit for solenoid operated valves appears not to meet Section 4.14 of IXEE 279 in that daliberate operator action is not required to restore the valves to their normal operating positions.
Please resolve the above.
14.
The following items in the FSAR appear to be inconsistent and need clarification or correction:
Table 7-1, Figure 7-1, and Section 15.2.3 disagree on the s.
setpoints of the Pcwer/ Flow and Power /RC Pump reactor trips, b.
Figure 7-6 and 7-7 disagree as to how the SC1 gating cir-cuits are disabled on a reactor trip.
Figure 7-6 indicates that power from the programmers to the group power supplies is interrupted. Figure 7-7 indicates that 120 vac input power to the programmer, which is shown as part of the group power supply, is interrupted.
- 15. The following questions concern items in the Bod Drive Control System for which the FSAR contains insufficient information.
They are grouped together only because it appears likely that j
l they concern the same circuits or protective actions.
OmCE >
SURNAME >
DATE >
Forra AEC-He (Rev,9-531 AECM 024, as. sowne== sat *eierme aarica ines e-ase.eir
oconee 1, 2, & 3 a
OCT 6 1969 Page 7-22 states that monitors are provided to sense a.
asyuumetric rod patterns.
Please describe these monitors.
The description should include, but not necessarily be limited to, the detection circuitry, alar:ns provided and their setpoints, control or protective actions served, and design bases, b.
Page 7-23 states that faults serious enough to warrant immediate action produce automatic correction cosanands.
Please identify and/or describe these faults, their detec-tion circuits, the automatic correction provided and the method of i:nplementing it, and the design bases of the circuits.
Section 14.1.2.7 discusses the dropped or stuck rod acci-c.
dent and references Sections 7.2.2 and 7.2.3.
Please provide a more complete description, similar to the:s requested in a. and b. above, of the referenced circuitry and the automatic protection it provides.
16.
Section 7.1.3.3.4 describes the periodic testing of the ESF actuation circuitry. As pressure switches are used as detectors, this section is not entirely applicable to channels 7 and 8, perticularly with respeec to logic relay lamps, bistable set-points, and comparative readings of analog indications.
Please l
CmCE >
SURNAME >
DATE >
Form AEC-318 (Rev. 9-ih AECM 0240 e s. novae.sent mutine arrics e sees e-aee-eir
OCT 6 1969 oconee 1, 2, a 3 lo interlocks ? If safety is involved, are the interlocks designed in accordance with the protective system design basis stated in Section 7.1.1 of the FSA17 19.
Please describe, or indicate in Figure 7-6 or 7-7, how the rod withdrawal interlocks (e.g., high SUR) are implemented in the rod drive control system.
20.
The Integrated Control System (ICS) and its design bases are discussed in Section 7.2.3 of the FSAR.
This discussion does not identify which, if any, of the functions provided by this system are required for reactor protection.
Please supply the following additional information:
a.
Identification of the safety reisted functions provided by the ICS; b.
The limitations placed on reactor operation if the ICS or any of its subsystems (unit load demand, integrated master, steam generator control, and reactor control) is not operat-ing properly.
21.
Please provide the following informa tion concerning the instru-mentation which provides protective signals:
a.
For the process instrumentation which provides signals to the RFS and ESF setuation circuitry, please provide a table d ich-lists"the-fat owing-in tion-:
SURNAME >
DATE >
Form AEC.)Ie iRev. 9-53) AECM 0240 m s. eavesesser.einv>=e orrica ; issa e-ase it
Ocon2a 1, 2, & 3 11 007 6 1969 (1) The parameter being sensed; (2) The type of sensor (e.g., Bailey pressure), manufactu-rers specified accuracy, repeatability and expected failure mode (s);
(3) The type of readout (e.g., indicating, blind);
(4) The type of power required (e.g., external, self);
(5) Use of channel (to provide information to RFS or ESF actuation circuitry);
(6) Identification of sensors connected to a cousnon sensing lino.
b.
The Oconee Nuclear Station represents the first nuclest power plant to reach the operating license review for which Babcock and Wilcox supplied nuclear instrumenttttion is used. For each type of nuclear detector being pro ided, v
state its operating history.
Briefly describe the design concepts utilized for the signal c.
conditioning and readout circuitry for the process and nue1 ear instrunentaeion.
- 22. Our preliminary review of the reactor coolant flow sensing scheme (Pados 7-34 and 7-35) has indicated that the design is susceptible to a single failure which could both result in the need for pro-taction (failed control channel) and, at the same time, negate OFFME >
$URNAME >
DATE >...............
Form AEC.He iRev.1-5h AECM 0240 as. sevennesse eeisme omes. i.ee e-m.eir i
Oconee 1, 2, & 3 u
OCT 6 1969 protection (four protection channels).
Plasse provide the results of tha analyses which were used to show that this design consplies with IEEE 279.
23.
In the Ocones design, safety injection initiation is provided by low reactor pressure or high containment pressure.
Initiation by either of these diverse signals, assuming failure of the other signal, should taeet ECCS requirements.
It is not clear whether the analysis of the effectiveness of the Oconee safety injection design takes credit for a reactor trip.
If a reactor trip is j
required, please provide the rationale used to show how the use of high containment pressura signals :neets this requirement.
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Form AEC-lis tRev.9-33) AECM 0240 a s.
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