ML18018A361
| ML18018A361 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 09/30/1982 |
| From: | Mcduffie M CAROLINA POWER & LIGHT CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.3, TASK-2.E.1.2, TASK-2.E.4.2, TASK-2.F.1, TASK-2.K.3.01, TASK-2.K.3.09, TASK-2.K.3.12, TASK-TM NUDOCS 8210050125 | |
| Download: ML18018A361 (69) | |
Text
REGUI.ATOR'NFORMATION DISTRIBUTION SgEM (RIBS)
ACCESSION NBR:8210050125 DOC ~ DATE: 82/09/30 NOTARIZED:
NO FACIL:50-400 Shearon Harris Nuclear Power Planti Unit iE Carolina 50-401 Shearon Harris Nuclear Power Plantr,Unit 2R Carolina AUTH BYNAME AUTHOR AFFILIATION MCDUFFIEiM>>A>>
Carolina Power 8 Light Co ~
RECIP ~ NAME RECIPIENT AFFILIATION DENTONgH>>R>>
Office of Nuclear Reactor Regulationi Director
SUBJECT:
Forwards responses to Instrumentation 8 Control Sys Branch 820730 questions re FSARrconfirming info discussed during 820816-19 8, 0914-16 meetings.
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- CSQE, Carolina Power & Light Company SEP. 3.0 l982 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, D.C.
20555 SHEARON HARRIS NUCLEAR POWER PLANT UNIT NOS ~
1 AND 2 DOCKET NOS.
50-400 AND 50-401 SAFETY REVIEW QUESTIONS RESPONSES-I&C SYSTEMS BRANCH 1
Dear Mr. Denton:
Carolina Power
& Light Company (CP&L) hereby transmits one original and forty copies of the Shearon Harris Nuclear Power Plant (SHNPP) responses to the Final Safety Analysis Report (FSAR) Safety Review Questions from the NRC Instrumentation and Control (I&C) Systems Branch.
This information is being submitted in response to NRC questions transmitted by letter dated July 30, 1982.
These fiftymix (56) questions were tranmitted as part of an agenda for a two-part NRC I&C Systems Branch meeting.
The first meeting, held on August 16-19, 1982 at the SHNPP site, concentrated on those questions identified as Nuclear Steam Supply System (NSSS) scope.
The second
- meeting, held on September 14-16, 1982 at the Ebasco New York office, covered the remaining questions.
These were mainly Balance of Plant (BOP) scope questions.
During these two meetings each of the fiftymix (56)
I&C questions was discussed.
The attached CP&L responses simply formalize the commitments given orally at these meetings.
At the end of each attached
- response, the status of each question is noted as defined by one of the following three categories:
Closed No further CP&L input or action is needed to resolve the NRC concern.
Confirmatory Carolina Power
& Light Company must provide the requested information on the NRC docket, either by letter or FSAR amendment.
Open No resolution possible at this time, CP&L to address this issue at a later date.
8210050125 820930 PDR ADOCK 05000400 PDR 411 Fayettevllle Street
~ P. O. Box 1551 e Raleigh, N. C. 27602
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Harold R. Denton Carolina Power information as committed becomes available.
This a future FSAR amendment.
contact our staff.
& Light Company will provide all additional to in confirmatory item responses as that information information will be submitted by formal letter or by If you have any questions on these responses, please Yours very truly, M. A. McDuffie Senior Vice President Engineering
& Construction JHE/cr (4560C5T5) cc:
Mr. E. A. Licitra (NRC)
Mr. G. F. Maxwell (NRC-SHNPP)
Mr. J.
P. O'Reilly (NRC-RII)
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I&C estion 1
Identify any plant safety related sys'em or portion thereof for which the design is incomplete at this time.
Response
The following is a list of plant safety related systems or portions thereof, for'hich the design is presently incomplete or,undergoing major safety related changes.
Auxiliary Feedwater System Reactor Vessel Level Indication Reactor Head Vent System Subcooling Monitor Hydrogen Sampling System Control Room Ventilation System (hlorine Detection System This item is closed.
I&C estion 2
As called for in Section 7.1 of the Standard Review Plan, provide information as to how your design conforms with the following TMI Action Plan Items as described in NUREG-0737:
(a)
II.D.3 (b)
II.E.1.2 Relief and safety valve position indication Auxiliary feedwater system automatic initiation and flow.
indication (c)
II.E.4' Containment isolation dependability (positions 4, 5, and 7)
(d)
II.F.1 Accident monitoring instrumentation (positions 4, 5, and 6)
(e)
XI.K.3.1 (f)
II.K.3.9 (g)
II.K.3.12 Auto PORV isolation PID controller Anticipatory reactor trip
~Res onse:
A discussion of SHNPP compliance with NUREG-0737 Items IX.D.3, ZI.E.1.2, IX.E.4.2, II.F.l, II.K.3.1, IX.K.3.9, and XI.K.3.12 was provided in the August 16-19, 1982 I&C meeting.
As a result of this discussion, CP&L will revise the FSAR TMI Appendix Item II.D.3 to address the fact that the valve position indicator is environmentally qualified.
'Also, when completed, CP&L will submit its position on TMI Items II.E.1.1 and II.E.4.2.(6).
CP&L will also revise the FSAR TMI Appendix Item II.F.1 to include a cross-reference to applicable portions of FSAR (hapter 6.
CP&L'S position on Item II.K.3.12 is dependent upon the resolution of I&C Question 19.
This item is confirmatory.
IGC estion 3
Provide a brief overview of the plant electrical distribution system, with emphasis on vital buses and separation divisions, as background for addressing various Chapter 7 concerns.
Response
In the August 16-19, 1982 IGC meeting, a brief overview of the plant electrical distribution system, with emphasis on vital buses and separation divisions was provided using electrical one line diagrams.
This item is closed.
IGC Question 4
Provide a discussion of the applicability of the RGs and GDCs identified in Table 7-1 of NUREG-0800 for systems addressed in Sections 7.4 through 7.6 of the FSAR.
Response
FSAR Table 7.1.0-1, Listing of Applicable Criteria for Instrumentation and Control System" provides a cross-reference for applicability of Regulatory Guides and General Design Criteria to specific FSAR Sections.
The information in this table is presently incomplete but will be revised in the FSAR and submitted to the NRC staff as it becomes available.
This item is confirmatory.
Z&C estion 5
We request that the setpoint methodology for each Reactor Protection System (RPS) and Engineered Safeguards Features (ESF) trip setpoint values be provided for both NSSS and BOP scope of supply.
Response
The setpoint methodology for each Reactor Protection System and Engineered Safety Features trip setpoint value is presently under review.
When the setpoint methodology is finalized it will be submitted to the NRC for review.
This information is expected to be ready for submittal by April, 1983.
This item is confirmatory.
I&C Question 6
Describe features of the Shearon Harris environmental control system which insure that instrumentation sensing and sampling lines for systems important to safety are protected from freezing during extremely cold weather.
Discuss the use of environmental monitoring and alarm systems to prevent loss of, or damage to, systems important to safety upon failure of the environmental control system.
Discuss electrical independence of the environmental control and monitoring system circuits.
Response
In the September 14-16, 1982 I&C meeting, the SHNPP Freeze Protection System was discussed.
The discussion included system redundance and the use of environmental monitoring and alarm systems to prevent loss of or damage to systems important to safety upon failure of the Freeze Protection System.
CP&L will submit to the NRC a description of the SHNPP Freeze Protection System and revise FSAR Section 7.7 to include this description.
This item is confirmatory.
Describe design criteria and tests performed on the isolation devices in the Balance of Plant systems.
Address results of analysis or tests performed to demonstrate proper isolation between separation groups and between safety and non-safety systems.
Response
In the September 14-16, 1982 IGC meeting, a discussion of the design criteria and tests performed on the isolation devices was provided.
The discussion included the results of analysis and tests performed to demonstrate proper isolation between separation groups and between safety and non-safety systems.
The NRC staff indicated that the adequacy of the test program would require further review.
This item is open.
I6C estion 8
Provide a list of non-class 1E control signals that provide input to class 1E control circuits.
Response
Non-class lE control circuits that provide input to class 1E control circuits were discussed with the staff during the September 14-16, 1982 IGC meeting.
An informational list detailing Control Wiring Diagram (CWD) references was provided to the staff.
This item is closed.
I&C estion 9
Identify where microprocessors, multiplexers, or computer systems are used in or interface with safety-related systems.
Also identify any "first-of-a-kind" instruments used for safety-related systems.
Response
The Emergency Response'acilities Information System (ERFIS) computer will interface with safety-related systems.
The Safety Radiation Monitoring System is a computer based system which is specified and purchased as a totally safety-related system.
Additionally, the SHNPP Solid State Protection System utilizes a
multiplexer to interface with the non-safety ERFIS computer system.
No other microprocessers, multiplexers, or computer systems are used in or interface with safety-related systems.
There are no "first-of-a-kind" instruments used for safety-related systems.
This item is closed.
I&C estion 10 The FSAR information which discusses conformance to Regulatory Guide 1.118 and IEEE 338 is insufficient.
Further discussion is required.
As a minimum, provide the following information:
1.
Discuss response time testing of BOP and NSSS protection systems using the design criteria described in position C.12 of R.G.
1.118 and Section 6.3.4 of IEEE*338.
2.
The FSAR Page 1.8-149 states that, "Temporary jumper wires, temporary test instrumentation, the removal of fuses and other equipment not hard-wired into the protection system will be used where applicable."
Identify where procedures require such operation.
Provide further discussion to describe how the test procedures for the protection systems conform to Regulatory Guide 1.118 (Rev.
1) position C.14 guidelines.
Identify and justify any exceptions.
3.
Describe typical response time test methods for pressure, differential pressure and temperature sensors.
Response
Response time testing of protection systems was discussed in the August 16-19, 1982 I&C meeting, using the design criteria described in position C.12 of Regulatory Guide 1.118 and Section 6'3.4 of IEEE 338.
The discussion referenced CP&L's response to FSAR Safety Review Question 640.28 concerning response time testing and test summaries in revised FSAR Sections 14.2.12.1.11 and 14.2.12 '.19 submitted in Amendment 4.
CP&L stated that test procedures are being developed, but are not complete.
Specific test procedures which require the use of temporary
- jumpers, temporary test instrumentation, the removal of fuses or other equipment not hard-wired into the protection system have not been identified.
The use of this type of equipment will be minimized but not precluded.
Typical reponse time test methods were also discussed.
CP&L will submit SHNPP test procedures concerning response time testing for NRC review when completed.
This item is confirmatory.
D C estion 11 Provide a commitment that the safety-related instrumentation and controls (I&C) described in Section 7.1 through 7.6 of the FSAR plus safety-related ISC for safety-related fluid systems will be subject to the pertinent requirements of the FSAR QA Program.
This can be done by another footnote to Table
- 3. 2. 1-1.
~Res onse:
In response to the question, the following footnote will be added to FSAR Table 3.2.1-1 in a future amendment.
This table is an overview of all the structures,
'components, and systems.
Refer to the Master Equipment List, process and instrument diagrams, and control wiring diagrams for further detail.
This item is confirmatory.
Z&C estion 12 Figure 7.3.1-1 sheets 1 through 7 are called the solid state protection system functional diagrams, but the logic is a repeat of Figure 7.2.1-1 functional diagrams.
These drawings should be revised to accurately reflect the design and remove errors.
Response
CP&L will revise FSAR Figures 7.2.1-1 and 7.3.1-1 to accurately reflect the design and remove errors.
These figures will be revised in a future amendment.
This item is confirmatory.
I&C estion 13 Indicate whether Shearon Harris has the Westinghouse "General Warning Alarm System".
If so, provide a list of conditions resulting in a general warning alarm and update the FSAR to include the input to the RTS from the general warning alarm.
~lb 8 onse:
Each of the two trains of the Solid State Protection System is continuously monitored by the General Warning Alarm subsystem.
The warning circuits are actuated if undersireable train conditions are set up by improper alignment of train testing systems, circuit malfunction or failure, etc.
as listed below.
A trouble condition in a train is indicated in the control room.
However, if any one of the conditions exists in Train A at the same time any one of the conditions exists in Train B, the reactor will be automatically tripped by the General Warning Alarm Reactor Trip system.
Potential trouble conditions monitored are:
a)
Loss of either of the two 48 volt DC or either of the two 15 volt DC power supplies.
b)
Printed circuit card improperly inserted.
c)
Input Error Inhibit switch in the INHIBIT position.
d)
Slave relay tester Mode Selector in TEST position.
h e)
Multiplexing selector switch in INHIBIT position.
f)
Opposite train bypass breaker racked in and closed.
g)
Permissive or Memory test switch not in OFF position.
h)
Logic Function test switch not in OFF position.
i)
Loss of AC power in the relay cabinets.
The above information will be incorporated in FSAR Section 7.2 in a future amendment.
This item is confirmatory.
Z&C estion 14 Figure 7.2.1-1 Sheet 2 does not show the shunt trip signals.
This is not consistent with the system description.
The functional diagrams should be revised to reflect the design.
Response
FSAR Figure 7.2.1-1 Sheet 2 will be revised to reflect the shunt trip signals design in a future amendment.
This item is confirmatory.
Z&C Question 15 Figure 7.2.1-1 Sheet 8 on steam line isolation logic does not show a seal-in and manual reset capability or manual system level initiation capability.
Explain the difference between the Shearon Harris design vs. other Westinghouse plants.
~Res onse:
FSAR Figure 7.2.1-1 Sheet 8 will be revised to show the seal-in and manual reset capability in a future amendment.
This item is confirmatory.
I&C estion 16 Figure 7.2.1-1 Sheet 8 on High Radiation signal to containment ventilation isolation indicates by note 5 that the circuit is not part of the safeguards system and is not redundant.
Explain the interface arrangement from the non-safety component to the safety system.
Response
FSAR Section 7 will be revised to include a discussion of the redundancy of the containment radiation monitors and the control room isolation monitors.
This item is confirmatory.
Z6C estion 17 Figure 7.2.1-1 Sheet 8 on containment ventilation isolation logic does not satisfy the IE Bulletin'0-06 concern i.e.,
the logic to initiate containment ventilation. isolation uses a radiation high-level signal and ESFAS T signal input to a retentive memory with actuation block gate.
Thus a reset of a high radiation signal would block the ESFAS input signal.
The design should be revised to eliminate this concern.
Response
The logic to inititate containment ventilation isolation will be corrected to ensure that a reset of a high radiation signal will not block the ESFAS input signal.
FSAR Figures 7.2.1-1 and 7'.1-1 will be revised accordingly.
This item is confirmatory.
ZG C Ques tion 18 Figure 7.2.1-1 Sheet ll on pressurizer power operated relief valve control does not show the control switch from the remote shutdown panel as you listed in Table 7.4.1-2.
Also, explain the arrangement between the main control room control switch and the remote shutdown panel control switch which you stated as non-safety train in Table 7.4.1-2.
Response
A discussion of the Auxiliary Control Panel and the transfer function from the Main Control Board was provided in the August 16-19, 1982 I&C meeting.
As a result of this discussion, FSAR Figure 7.4.1-8 will be revised to show the front view of the Auxiliary Control Panel.
FSAR Section 7.4.1 will be revised to include a statement that the physical separation within the Auxiliary Control Panel is the same as the physical separation within the Main Control Board.
This item is confirmatory.
I&C Question 19 The information in Section 7.2.1.1.2.f, "Reactor Trip on Turbine Trip," is insufficient.
Please provide further design bases discussion on this subject, per BTP ICSB 26 requirements.
As a minimum you should:
1)
Using detailed drawings, decribe the routing and separation for this trip circuitry from the sensor in the turbine building to the final actuation in the reactor trip system (RTS).
2.
Discuss how the routing within the non-seismic Category I turbine building is such that the effects of credible faults or failures in this area on these circuits will not challenge the reactor trip system and thus degrade the RTS performance.
This should include a
discussion of isolation devices.
3.
Describe the power supply arrangement for the reactor trip on turbine trip circuitry.
4.
Discuss the testing planned for the reactor trip on turbine trip circuitry.
Identify other sensors or circuits used to provide input signals to the other protection system which are located or routed through non-seismically qualified structures.
This should include sensors or circuits providing input for reactor trip, emergency safeguards equipment such as the auxiliary feedwater
- system, and safety grade interlocks.
Verification should be provided that the sensors and circuits meet IEEE-279 and are seismically and environmentally qualified.
Testing or analyses performed to insure that failures of nonseismic structures, mountings, etc. will not cause failures which could interfere with the operation of any other portion of the protection system should be discussed.
Response
A discussion of the "Reactor Trip on Turbine Trip" was provided at the August 16-19, 1982 IGC meeting.
Included in the discussion was a description of the routing and separation for this trip circuitry including the routing within the non-seismic category 1 Turbine Building.
Also discussed were the power supply arrangement for the reactor trip on turbine trip circuitry and the testing of the circuit.
As a result of this discussion FSAR Section 7.2.1.1
~ 2.f will be revised to include the following information:
Circuit analyses have shown that the functional performance of the protection system would not be degraded by credible faults in circuits associated with reactor trip from turbine trip.
Contacts of sensors on steam stop valves and trip fluid pressure system are closed during operation to complete the AC supply through the Solid State Protection System input relays.
Open circuit or short circuit faults would trip the channel to produce a partial reactor trip.
Z&C estion 19 (cont'd)
FSAR Figure 7.3.1-1 Sheets 5 and 6 will be revised to indicate P-14 logic.
FSAR Table 7.3.1-4 will be revised to add P-14 logic.
A statement will be added to FSAR Section 7 which states that the Reactor Trip on Turbine Trip wiring is in separate conduit.
This item is confirmatory.
I&C estion 20 Identify where instrument sensors or transmitters supplying information to more than one protection channel are located in a common instrument line or connected to a common instrument tap.
The intent of this item is to verify that a single failure in a common instrument line or tap (such as break or blockage) cannot defeat required protection system redundancy.
Response
A discussion of where instrument sensors or transmitters supplying information to more than one protection channel are located in a common instrument line or connected to a common instrument tap was provided in the August 16-19, 1982 I&C meeting.
As a result of this discussion FSAR Section 7.3.2.3 will be revised to include the information presented below which provides justification that a single failure in a common tap will not defeat the protection system redundancy.
The following two cases have been identified as incidences where instrument transmitters supplying information to more than one protection channel share a
common instrument line or tap: (1) the, Reactor Coolant System flow transmitters, three to each loop, (2) each steam generator where at each of two upper taps a steam flow transmitter is connected along with a narrow range steam generator water level transmitter.
In the first case, the high pressure elbow tap for the three redundant Protection System Reactor Coolant System (RCS) flow transmitters is where the shared tap is located.
Redundancy is not compromised by having a shared tap.
If this tap were to break, the instrumentation would fail low which is in the direction of initiation of protective action.
If the shared tap is plugged, which is really not credible, the affected channels will remain static and the condition is easily detectable.
In the second
- case, a failure of the common tap either due to breaking or plugging does not compromise redundancy nor prevent automatic protective action if an initiating condition continued long enough to require protective action.
The protective action is provided either by 2/3 steam-generator low-low water level or by two channels of steam-feedwater flow mismatch due to low feedwater flow indication.
For the discussion on this diverse protection, refer to Section 7.2.2.3.5 which gives recognition to the fact that the steam generator level signal used in the feedwater control originates separately from that used.on a low S.
G. water level signal which, in coincidence with the low feedwater flow signal, will trip the reactor.
Therefore, it can be concluded that a failure of the common tap on the steam generator would not adversely affect the steam generator water level control; and a second random failure does not have to be assumed.
Protective action is therefore available, should an initiating event require it, from the remaining intact diverse instrumentation channels.
This item is confirmatory
Z&C estion 21 Discuss the method of redundantly tripping the turbine following receipt of reactor protection signals requiring turbine trip.
~Res onse:
A discussion of redundantly tripping the turbine following receipt of reactor protection signals requiring turbine trip was provided in both the August 16-19, and September 14-16, 1982 I&C meetings.
The discussions included the use of isolation devices and circuit testability.
As a result of these discussions the turbine trip circuitry design is being reevaluated.
FSAR Figure 7.3.1-1 Sheet 7 will be revised to show the design and a discussion will be added to FSAR Section 7.2 to describe the turbine trip circuitry and logic. Carolina Power
& Light Company will also provide the NRC with a set of process block drawings.
This item is confirmatory.
I6C Question 22 Discuss the diverse features of the undervoltage and shunt trips of the reactor trip breakers.
Indicate if they can be tested independently.
Response
A discussion of the diverse features of the undervoltage and shunt trips of the reactor trip breakers and,the testability of both was provided in the August 16-19, 1982 IGC meeting.
The present SHNPP design does not allow the shunt trip to be independently tested.
CPGL will await generic resolution of the need to independently test shunt trips designed by Westinghouse.
This item is open.
ZGC estion 23 Using detailed system schematics, describe the sequence for automatic initiation, operation,
- reset, and control of the auxiliary feedwater system.
The following should. be included in the discussion:
a) the effects of all switch positions on system oper'ation.
b) the effects of single power supply failures including the effect of a power supply failure on auxiliary feedwater control after automatic initiation circuits have been reset in a post accident sequence.
c) any bypasses within the system including the means by which it is insured that the bypasses are remov'ed.
d) initiation and annunciation of any interlocks or automatic isolations that could degrade system capability.
e) the safety classification and design criteria for any air systems required by the auxiliary feedwater system.
This should include the design bases for the capacity of air reservoirs required for system operation.
f) design features provided to terminate auxiliary feedwater flow to a steam generator affected by either a steam line or feed line break.
g) system features associated with shutdown from outside the control room.
Response
Discussion of the SHNPP Auxiliary Feedwater (AFW) System was presented at both the August 16-19, and the September 14-16, 1982 XGC meetings.
As a result of this discussion CP&L will revise FSAR descriptive text, and logic and flow diagrams associated with the AFW System to accurately reflect the present design of the system.
This item is confirmatory.
ZGC estion 24 FSAR section 10.4.9.2.2 stated that the auxiliary feedwater pumps can also be remote manually aligned to take suction from the emergency service water system in the event of a loss of the condensate storage tank.
Describe the instrumentation available to the operator for his decision to switch over the water source, and the procedure used to transfer pump suction from the Condensate Storage Tank to the Service Water System including verification that all equipment used for this function is seismically qualified.
Response
A discussion of the switchover of the Auxiliary Feedwater Pump Suction from the Condensate Storage Tank to the Emergency Service Water System was provided in the September 14-16, 1982 IGC meeting.
As a result of this discussion
-FSAR Section 7.4.1 will be revised to show the Condensate Storage Tank level indication on the Auxiliary Control Panel.
FSAR Tables 7.4.1-1 and 7.4.1-2 will be revised accordingly.
This item is confirmatory.
ZGC estion 25 Using the detailed system schematics, describe the sequence for periodic testing of the a) main steam line isolation valves, b) main feedwater isolation valves, c) main feedwater control valves (Safety features) d)
auxiliary feedwater system.
The discussion should include features used to insure the availability of the safety function during test and measures taken to insure that equipment cannot be left in a bypassed condition after test completion.
Response
Periodic testing of these valves was discussed in the August 16-19, 1982 ZSC meeting.
The discussion included the features used to insure the availability of the safety function during test and measures taken to insure that equipment cannot be left in a bypassed condition after test completion.
This item is closed.
Z&C estion 26 Using detailed schematics and other drawings as necessary, describe the main steam and feedwater isolation valve hydraulic operators including the interfaces with the safety system electrical circuits.
Response
At the August 16-19, 1982 Z&C meeting, the main steam and feedwater isolation valves'ydraulic operators were described.
This discussion included interfaces with the safety system electrical circuits.
This item is closed.
X&C estion 27 Using detailed system schematics, discuss the bypass, bypass interlock, and test provisions for containment ventilation isolation and control room ventilation isolation.
The discussion should indicate those design features which insure that the safety function is not defeated during system test and that portions of the system are not inadvertently left in a bypassed condition after test.
Response
A discussion of the bypass, bypass interlock, and test provisions for containment ventilation isolation and control room ventilation isolation was provided at the August 16-19, 1982 IGC meeting.
The discussion included those design features which insure that the safety function is not defeated during system test and that portions of the system are not inadvertently left in a bypass condition after test.
As a result of this discussion, FSAR Figure 7.3.1-17 will be revised as necessary.
The corresponding control wiring diagrams (CWD) will also be revised.
This item is confirmatory.
I&C estion 28 Identify any safety systems shared between Units 1 and 2.
Response
At the September 14-16, 1982 IGC meeting, the safety systems shared between Units 1 and 2 were discussed.
As a result of this discussion, FSAR Table 1;2.3-1 will be revised.
This item is confirmatory.
I&C estion 29 Describe the steam generator level instrumentation.
Identify the instrument channel used for protection functions and the control functions.
Address the control and protection interaction concerns.
Response
A discussion of the steam generator level instrumentation using logic diagrams was presented at the August 16-19, 1982 I&C meeting.
The NRC was concerned with the interaction of control and protection channels.
CP&L is reevaluating its steam generator level instrumentation logic.
This item is open.
Z6C estion 30 Using detailed schematics, describe the operation of the containment spray system initiating circuits, bypasses, interlocks and functional testing.
Discuss the redundancy of the spray additive tank isolation valves which are closed on low additive tank level.
Describe how the transmitter used to close these valves is monitored.
Indicate periodic test requirements for the instrumentation and controls used.
Response
A discussion of the operation of the containment spray system with focus on the spray additive tank isolation was presented at the September 14-16, 1982 ZGC meeting.
As a result of this discussion, the adequacy of the instrumentation for terminating NaOH addition in the containment spray system is being reevaluated.
This item is open.
t
IGC estion 31 Using logic and schematic
- diagrams, describe the safety injection syst'm initiating circuits, bypasses, interlocks and functional testing.
Response
The safety injection system initiating circuits, bypasses, interlocks, and functional testing were described in the August 16-19, 1982 I6C meeting.
This item is closed.
~ I
~
Z6C estion 32 Using logic and schematic diagrams, describe the AC emergency power system (diesel generators and sequencer),
initiating circuits, bypasses, interlocks and functional testing.
Response
A discussion of the AC emergency power system initiating circuits, bypasses, interlocks, and functional testing was presented at the September 14-16, 1982 l&C meeting.
This item is closed
IGC estion 33 In FSAR section 7.4.1.2, you stated that systems required for safe shutdown can be periodically tested during normal operation whenever possible to do so
- without adverse affect on plant safety or availability.
Please identify each component which can not be periodically tested during normal operation.
Provide a cross reference to the Technical Specification sections for those components that will be tested during normal operation.
Response
During the August 16-19, 1982 IGC meeting, the components required for safe shutdown which cannot be periodically tested during normal operation were identified.
As a result of this discussion, CPGL is reviewing the proposed Technical Specifications for surveillance requirements of components required for safe shutdown and will submit revised Technical Specifications for NRC review.
CP&L will also revise FSAR Section 7.1 ~ 2.5 accordingly.
This item is confirmatory.
I&C estion 34 The information supplied for remote shutdown from outside the control room is insufficient.
Therefore, provide further discussion to describe the capability of achieving hot and cold shutdown from outside the control room.
As a minimum, provide the following information:
a)
Location of transfer switch'es and remote control station (include layout drawings, etc).
b)
Design criteria for the remote control station equipment including transfer switches.
c)
Description of distinct control features to both restrict and to assure
- access, when necessary, to the displays and controls located outside the control room.
d)
Discuss the testing to be performed during plant operation to verify the capability of maintaining the plant in a safe shutdown condition from outside the control room.
e)
Description of isolation, separation and transfer/override provisions.
This should include the design basis for preventing electrical interaction between the control room and remote shutdown equipment.
f)
Description of any communication systems required to coordinate operator actions, including redundancy and separation.
g)
Description of control room annunciation of remote control or overridden status of devices under local control.
h)
Means for ensuring that cold shutdown can be accomplished.
i)
Discuss the separation arrangement between safety related and non-safety related instrumentation on the auxiliary shutdown panel.
Response
A complete discussion of remote shutdown from outside the control room was provided at both the August 16-19 and September 14-16, 1982 I&C meetings.
As a result of this discussion, CP&L willmodify the door logic to the room containing the Auxiliary Control Panel to both restrict and assure access.
CP&L will also submit for staff review the SHNPP Appendix R Fire Protection Evaluation when it is completed.
This item is confirmatory.
IGC Question 35 Describe the procedures to borate the primary coolant from outside the control room when the main control room in inaccessible.
How much time is there to do this?
~Res onse:
A description of procedures to borate the primary coolant from outside the control room was provided at the August 16-19, 1982 Z&C meeting.
This item is closed.
I6C estion 36 In FSAR section 7.4.1.5, you stated that the automatic charging pump control is not required for safety.
Use detail schematic diagram to explain the interface of protective function and control function of the charging pump circuits.
Response
A discussion of the interface of protective function and control function of the charging pump circuits was presented at the August 16-19, 1982 IGC meeting.
This item is closed.
IGC estion 37 Using detailed drawings (schematics, PGIDs), describe the automatic and manual operation and control of the atmospheric relief valves.
~Res onse:
A discussion of the automatic and manual operation and control of atmospheric relief valves was provided at the September 14-16, 1982 Z&C meeting.
The discussion focused on the environmental qualification of the steam generator power operated relief valves.
This item is closed.
ZGC estion 38 Use detailed schematic to describe the control circuits of the pressurizer heater
- groups, including the interlock and bypass provision from the auxiliary control panel.
Response
A description of the control circuits of the pressurizer heater groups on the Auxiliary Control.Panel was presented at the September 14-16, 1982 ZGC meeting.
As a result of the discussion, CP6L will provide an updated version of the Control Wiring Diagram 2166-B-401, Sheets 152 and 153.
This item is confirmatory.
ISC estion 39 Use detailed schematic to describe the following valves operation on loss of instrument air system:
1)
Boric acid flow control valve 2)
Reactor coolant letdown isolation valve 3)
RHR flow control valve 4)
Pressurizer spray control valve
Response
A discussion of the operation of these valves on loss of instrument air system was presented at the August 16-19, 1982 X&C meeting.
This item is closed.
I&C Question 40 Identify other pneumatically operated valves in the ESF system, besides listed on item 39, and use detailed schematic to describe their operation on loss of instrument air system.
~Res onse:
A discussion of other pneumatically operated valves in the ESF
- system, not listed in Question 39, was provided in the August 16-19, 1982 I&C meeting.
This item is closed.
ZGC estion 41 Using detailed drawings, describe the ventilation systems used to support engineered safety features areas including areas containing systems required for safe shutdown.
Discuss the design bases for these systems including redundancy, testability, etc.
Response
A description of the ventilation systems used to support ESF areas including areas containing systems required for safe shutdown was provided at the September 14-16, 1982 XGC meeting.
This item is closed.
I&C estion 42 Using detailed electrical schematics and piping diagrams, discuss the automatic and manual operation and control of the station service water system and the component cooling water system.
Discuss the interlocks, automatic switchover, testability, single failure, and channel independence, indication of operabil~ity, isolation functions, etc.
Response
A discussion of the automatic and manual operation and control of the station service water system and the component cooling water system was provided in the September 14-16, 1982 IGC meeting.
As a result of this discussion, CP6L will revise the Technical Specifications for component cooling water surge tank level testing.
CP6L will also modify Technical Specifications to ensure that the breaker for "C" component cooling water pump cannot be racked in unless the other associated component cooling water pump breaker is racked out.
The revised Technical Specifications will be submitted for NRC review.
This item is confirmatory.
IGC estion 43 Discuss the plant tests to verify the capability of maintaining the plant in a safe shutdown condition from outside the control room.
Describe design compliance with Regulatory Guide 1.68.2.
Response
A discussion of plant tests to verify the capability of maintaining the plant in. a safe shutdown condition from outside the control room was provided in the August 16-19, 1982 I&C meeting.
The discussion included a reference to CPGL response to PSAR Safety Review Question 640.6 and the test summary in PSAR Section 14.2.12.2.20.
This item is closed.
ZGC estion 44 Discuss. the plans and schedule for complying with R.G.
1.97, Rev.
2.
Describe the conformance of the present design.
Response
CPGL is presently evaluating its compliance with Regulatory Guide 1.97, Revision 2; CPGL will submit. its position on this Regulatory Guide when the evaluation is complete.
This item is open.
I6C estion 45 In a recent letter to NRC dated June 1,
1982, you stated that as a result of detailed control room design review, the main control board was redesigned and certain equipment relocated.
Please discuss the major changes in the control room design and the impact on the accuracy of the information currently provided in section 7.5 of the FSAR.
Response
A discussion of the changes in the control room design as a
result of a human factors review was provided at'he September 14-16, 1982 ISC meeting.
This item is closed.
Z&C estion 46 Using detailed plant design drawings (schematics),
discuss the Shearon Harris design pertaining to.bypassed and inoperable status indication.
As a minimum, provide the information to describe:
1)
Compliance with the recommendations of R.G.
1.47.
2)
The design philosophy used in the selection of equipment/systems to be monitored, 3)
How the design of the bypass and inoperable status indication systems comply with positions Bl through B6 of ICSB Branch Technical Position No.
21, and 4)
The list of system automatic and manual bypasses within the BOP and NSSS scope of supply as it pertains to the recommendations of R G 1 47.
The design philosophy should describe as a minimum the cirteria to be employed in the display of inter-relationships and dependencies on equipment/systems and should insure that bypassing or deliberately induced inoperability of any auxiliary or support system will automatically indicate all safety systems affected.
Response
t A complete discussion of the bypass and inoperable status indication design was presented in the September 14-16, 1982 ZGC meeting.
As a result of this discussion, FSAR Section 7.5.1 will be revised to further clarify the design criteria used for the inoperable status indication.
This item is confirmatory.
Z&C estion 47 Use schematic and layout drawings to discuss the physical separation and wiring for redundant safety related instruments on the main control board.
~Res onse:
A description of the physical separation and wiring for redundant safety related instruments on the main control board was presented at the August 16-19, 1982 Z&C meeting.
As a result of this discussion, FSAR Section 8.3.1.2.30 will be revised to provide a description of the physical separation in the control board.
FSAR Section 8.3.1.2.30 will be modified to include the following information:
Main Control Boards have maintained the separation criteria by use of metal enclosures for all Class 1E circuits within the board.
Class 1E devices are encapsulated in metal enclosures (module cans),
and wirings from such devices are enclosed in flexible metallic conduits running to enclosed sheet metal wireways which are dedicated to trains to point of exit from the board.
Non-Class 1E circuits are separated from Class 1E circuits by provision of wiring runs and exit points physically separated from those for Class 1E circuits.
This item is confirmatory.
I&C estion 48 Provide a discussion (using detailed drawings) on the residual heat removal (RHR) system as it pertains to Branch Technical Positions ICSB 3 and RSB 5-1 requirements.
Specifically, address the following as a minimum:
1)
Testing of the RHR isolation valves as required by branch position E.
of BTP RSB 5-1 ~
2)
Capability of operating the RHR from the control room with either onsite or only offsite power available as required by Position A.3 of BTP RSB 5-1.
This should include a discussion of how the RHR system can perform its function assuming a single failure.
3)
Describe any operator action required outside the control room after a single failure has occurred and justify.
In addition, identify all other points of interface between the Reactor Coolant System (RCS) and other systems whose design pressure is less than that of the RCS.
For each such interface, discuss the degree of conformance to the requirements of Branch Technical Position ICSB No. 3.
Also discuss how the associated interlock circuitry conforms to the requirements of IEEE Standard 279.
The discussion should include illustrations from applicable drawings.
Response
A discussion of the Residual Heat Removal (RHR) System including isolation valve testing,'nd operator actions following a single failure was provided at the August 16-19, 1982 I&C meeting.
As a result of this discussion, the following statement will be added to FSAR Section 5.4.7:
Only one train of the RHR system need be in operation at any one time.
The train which is not in service can be tested.
- Also, CP&L will revise its responses to FSAR Safety Review Questions 430.106, 440.16, and 440.21 to reflect operator actions following a single failure in the power supply to the RHR isolation valves as discussed in the meeting.
This item is confirmatory.
IGC Question 49 Using detailed system schematics, describe the power distribution for the accumulator valves and associated interlocks and controls including bypass indicator light arrangement.
Describe how power is removed and how the system complies to Positions B.2, B.3, and B.4 of BTP ICSB 18 (PSB).
Also, identify any other such areas of design and state your conformance to the positions of BTP ICSB 18.
Response
At the August 16-19, 1982 IGC meeting, a description of the power distribution for the accumulator valves and associated interlocks and controls including the bypass indicator light arrangement was provided.
As a result of the discussion, CP&L will provide the NRC with a listing of all valves where power is removed during operation.
This item is confirmatory.
Z&C Question 50 Using detailed schematics, describe the design of pressurizer PORV control and the block valves control, and verify that no single failure will preclude the automatic actuation logic during the low temperature mode of operation.
Response
The design of the pressurizer PORV control and block valves control was discussed at the August 16-19, 1982 Z&C meeting.
The discussion focused on the possibility that a single failure could preclude the automatic actuation logic during the low temperature mode of operation.
Carolina Power
& Light Company is reviewing the adequacy of the existing low temperature overpressure protection system against single-failure criteria.
This item is open.
IGC estion 51 Using detailed schematics, describe the design of boric acid addition control and the volume control tank level control.
Discuss the Westinghouse generic deficiency regarding volume control tank level (Westinghouse letter to NRC dated May 21, 1981 under 10 CFR 21 requirement) and the CP&L's evaluation to this concern.
Response
A discussion of the design of boric acid addition control and the Volume Control Tank level control was provided in the August 16-19, 1982 ZGC meeting.
The discussion focused on the Westinghouse generic deficiency regarding volume control tank level.
(hrolina Power 6 Light Company's position on this generic deficiency is that the problem can be corrected with timely operator action.
(hrolina Power 6 Light Company will submit the Operating Procedures concerning Volume Control Tank level control for staff review when available.
This item is confirmatory.
ZGC estion 52 Describe the operation of the interlocks used for isolation of the seismic qualified portion of the CCW system.
This discussion should include reference to the fluid system schematics indicating which specific valves are used for the isolation function.
Discuss whether redundancy of instrumentation is within each CCW train or is accomplished by having one interlock per train.
Response
The operation of the interlocks used for isolation of the seismic qualified portion of the Component Cooling Water System was discussed in the August 16-19, 1982 Z&C meeting.
The discussion focused on the Gross Failed Fuel Detector System and its interconnection to the Component Cooling Water System.
This item is closed.
IGC estion 53 Regarding CCW pumps, discuss (a) circuits which automatically start a pump in CCW train on low pressure in the pump discharge and (b) circuits which automatically start a
pump in the operating trains on an SIS.
Include subsequent operator control of the pumps.
Response
A discussion of Component Cooling water pump circuits which automatically start a
pump on low pressure and circuits which automatically start a
pump on a Safety Injection Signal was presented in the August 16-19, 1982 l&C meeting.
This item is closed.
I6C estion 54 On November 7, 1979, Westinghouse notified the Commission of a potential undetectable failure which could exist in the engineered safeguards P-4 interlocks.
Test procedures were developed to detect failures which might occur.
The procedures require the use of voltage measurements at the terminal blocks of the reactor trip breaker cabinets.
In order to minimize the possibility of accidental shorting or grounding of safety system circuits during testing, the staff believes that suitable test jacks should be provided to facilitate testing of the P-4 interlocks.
Provide a discussion on how the above issue will be resolved for Shearon Harris.
Response
Carolina Power
& Light Company has reviewed the NRC position on this issue and agrees that a design change is needed to minimize the possibility of accidental shorting or grounding of safety system circuits during testing.
When the specific design change is finalized, it will be submitted to the NRC for review.
This item is confirmatory.
Z&C estion 55 Discuss your plant computer system which is used for the backup function of the safety related display instrumentation.
Response
A discussion of the plant computer system was presented at the September 14-16, 1982 DC meeting.
This item is closed.
Z&C estion 56 Discuss your steam dump control system which is used for the backup function of the safety system.
Response
A discussion of the steam dump control system was presented in the August 16-19, 1982 I&C meeting.
No credit for steam dump other than code safety relief valves is taken in any safety analysis.
This item is closed.
V cl I&C estion 57 Provide justification for not having manual initiation on the system level for Item 3-8 on Table 7.3.1-3 (This is a new question as a result of the September 14-16, 1982 I&C meeting).
Response
Carolina Power
& Light'ompany will provide the necessary justification when it is completed.
This item is confirmatory.
4 ly