ML20039F220
| ML20039F220 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 12/21/1981 |
| From: | Adensam E Office of Nuclear Reactor Regulation |
| To: | Parker W DUKE POWER CO. |
| References | |
| NUDOCS 8201120207 | |
| Download: ML20039F220 (27) | |
Text
r DISTRIBUTION Docket File
' Local PDR LB #4 reading NRC PDR dig 2 L IM DEisenhut NSIC/ TIC EAdensam TERA Locket hos.:
50-413 KJabbour ACRS (16) end 60-414 MDuncan A
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RTedesco RVollmer Q/
%C Dr. Williaa 0. Parker, Jr.
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vice President - Steam Production RMattsen RHartfie'.d,kFPA4 CgO/987gj Duke Power Laapeny N7
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A P.O. box 33189 OELD charlotte, Nortn Carolina 28242 OIE (3)
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Dear Hr. Parr.er:
Suoject: tEL Instrucentation and Control Systens trancn (ICSii)
Agenua Items for Discussion with Duke on Catawba Station Attachec (Cnclosure 1) is a list of items which the IL5u woulo like to discuss with you in a series of meetings. The initial orientation r.4eeting is scheouled for Jariary 13,.19ebin betnesda. Tne purpose of this ueeting woula be to confira the schedule for later n.eetings and to outline the objectives of the review process. We would not expect to discuss the tech-nical details of the agenda items, houever we would attempt to ass 1 n 9
priorities end schedules far adoressing each item. Subsequent reetings will be scheduled to review the hSSS and buP cesigns and interfaces.
We enticipate that other questions and conctrns may arise as a result of the meetings and froa our consultant's, Argonne National Laboratory, review.
Thus, the attached list should not be considered a final or coreplete list of iteas to be resolved prior to issuing a Safety-Evaluation Report.
Our review in other areas will be completed in the near future; and we will.
send you separate requests for additional inforuation related to those areas.
If you require any clarification of this letter, please contact the project aanager, Kahtan Jabbour, at (301) 492-7821.
The reporting and/or record keeping requirements contained in this letter af fcct fewer tnan ten respondents; therefore 0110 clearance is not required under P.L.96-511.
Sincerely, Llinor b. isdenson, Chief Licensing uranch !4 Division of Licensing
Enclosure:
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CATAWBA I
l Mr. William O. Parker Vice President Steam Production Duke Power Company P.O. Box 33189 Charlotte, North Carolina 28242
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cc: William L. Porter, Esq.
North Carolina Electric Membership Duke Power Company Corp.
P.O.' Box 33189 t.
3333 North Boulevard' Charlotte, North Carolina 28242 P.O. Box 27306 Raleigh, North Carolina 27611 J. Michael McGarry, III, Esq.
Debevoise & Liberman Saluda River Electric Cooperative, 1200 Seventeenth Street, N.W.
Inc.
Washington, D. C.
20036 207 Sherwood Drive Laurens, South Carolina 29360 North Carolina MPA-1 P.O. Box 95162 James W. Burch, Director Raleigh, North Carolina 27625 Nuclear Advisory Counsel 2600 Bull Street Mr. R. S. Howard Columbia, South Carolina 29201 Power Systems Division Westinghouse Electric Corp.
Mr. Peter K. VanDoorn P.O. Box 355 Route 2, Box 179N Pittsburgh, Pennsylvania 15230 York, South Carolina 29745 Mr. J. C. Plunkett, Jr.
NUS Corporation 2536 Countryside Boulevard Clearwater, Florida 33515 Mr. Jesse L. Riley, President Carolina Environmental Study Group 854 Henley Place Charlotte, North Carolina 28208 Richard P. Wilson, Esq.
Assistant Attorney General S.C. Attorney General's Office P.O. Box 11549 Columbia, South Carolina 29211 Walton J. McLeod, J r., Esq.
General Counsel South Carolina State Board of Health J. Marion Sims Building 2600 Bull Street Columbia, South Carolina 29201 4
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UNITED STATES y
g NUCLEAR REGULATORY COMMISSION 5
- y WASHINGTON, D, C. 20555
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DEC 21 #
Docket Nos.:
50-413 and 50-414 Mr. William O. Parker, Jr.
Vice President - Steam Production Duke Poier Company P.O. Box 33189 Charlotte, North Carolina 28242
Dear Mr. Parker:
Subj ect: NRC Instrumentation and Control Systems Branch (ICSB)
Agenda Itens for Discussion with Duke on Catawba Station Attached (Enclosure 1) is a list of items which the ICSB would like to discuss with you in a series of meetings. The initial orientation meetina is scheduled for January 13, 1981 in Bethesda. The purpose of this meetir.3 would be to confirm the schedule for later meetings and to outline the objectives of the review process. We would not expect to discuss the tech-nical details of the agenda itens, however we would attempt to assign priorities and schedules for addressing each item. Subsequent meetings will be scheduled to review the NSSS and B0P designs and interfaces.
We anticipate that other quest!ons and concerns may arise as a result of the meetings and from aur consultant's, Argonne National La%ratory, review.
Thus, the attached list should not ce considered a final or complete list of itens to be resolved prior to issuing a Safety Evaluation Report.
Our review in other areas wH1 be completed in the near future; and we will send yoc separate requests for additional infonnation related to those areas.
If you require any clarification of this letter, please contact the project manager, Kahtan Jabbour, at (301) 492-7821.
The reporting and/or record keeping requirements contained in this letter affect fewer than ten respondents; therefore, OMB clearance is not required under P.L. %-511.
Sincerely,
.,/
Elinor G. Adensam, Chief Licensing Branch #4 Division of Licensing
Enclosure:
As stated cc:
See next page
6 f
CATAWBA Mr. William 0. Parker Vice President - Steam Production Duke Power Company P.O. Box 33189 Charlotte, North Carolina 28242 cc: William L. Porter, Esq.
North Carolina Electric Membership Duke Power Company Corp.
P.O. Box 33189 t.
3333 North Boulevard Charlotte, North Carolina 28242 P.O. Box 27306 Raleigh, North Carolina 27611 J. Michael McGarry, III, Esq.
Debevoise & Liberman Saluda River Electric Cooperative, 1200 Seventeenth Street, N.W.
Inc.
Washington, D. C.
20036 207 Sherwood Drive Laurens, South Carolina 29360 North Carolina MPA-1 P.O. Box 95162 James W. Burch, Director Raleigh, North Carolina 27625 Nuclear Advisory Counsel 2600 Bull Street Mr. R. S. Howard Columbia, South Carolina 29201 Power Systems Division Westinghouse Electric Corp.
Mr. Peter K. VanDoorn P.O. Box 355 Route 2, Box 179N Pittsburgh, Pennsylvania 15230 York, South Carolina 29745 Mr. J. C. Plunkett, J r.
NUS Corporation 2536 Countryside Boulevard Clearwater, Florida 33515 Mr. Jesse L. Riley, President Carolina Environmental Study Group 854 Henley Place Charlotte, North Carolina 28208 Richard P. Wilson, Esq.
Assistant Attorney General S.C. Attorney General's Office P.O. Box 11549 Columbia, South Carolina 29211 Walton J. McLeod, J r., Esq.
General Counsel South Carolina State Board of Health J. Marion Sims Building 2600 Bull Street Columbia, South Carolina 29201
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ENCLOSURE 1
, Abkg* j AGEt!DA ITEt15 FOR t1EETItiG(S) WITH DUKE POWER C0ftPANY ON CATAWBA STATI0tl'S INSTRUMENTATION AND CONTROL Following is a list of items for discussion at meetings with the applicant to pro-vide the NRC staff with information required to understand the design bases and design implementation for the instrumentation and control systems at Catawba fluclear The applicant should be prepared to use both simplified and detailed Station.
instrument, control and fluid system schematics during the meetings to explain system designs and to provide verification that design bases and regulatory criteria are met.
The staff requested a review of the adequacy of emergency operating 1.
procedures to be used by control room operators to obtain safe shutdown upon loss of any Class lE or non-Class lE bus supplying power to safety or non-safety related instruments and controls. This was addressed in a letter from R. Tedesco (NRC) to W. Parker (DPC) dated April 16,1981, as Item 222.01.
The response to question 222.01 (Volume 13 of the FSAR) in-dicates that a review was made using the guidelines of IE Bulletin 79-27 and concluded that no design modifications are required.
(1) Confirm that all a.c. and d.c. instrument buses that could affect the ability to achieve cold shut condition were reviewed and identify these buses.
(2) Confirm that clear, simple, unambiguous annunciation of loss of power is provided in the control room for each bus addressed in item 1 above.
Identify any exceptions.
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- (3) Confirm that th2 effect of loss of pow:r to each l'oad on each bus identified in item 1 above, including ability-to reach cold shutdown, was considered in the review.
2.
The staff requested the applicant to perform a re.iew to determine what, if any, design changes or operator actions would be necessary to assure that high energy.line break will not cause control system failures to complicate the event beyond the FSAR analysis.
This issue was addressed in the letter from R. Tedesco (NRC) to W. Parker (DPC) dated April 16, 1981, as Item 222.03.
In the response to question 222.03 (Volume 13 of FSAR) the applicant states that the required review is not completed and results will be documented in a later revision. This item will remain _ open until applicant has submitted the results and our review of those results is completed.
3.
The staff requested the applicant to provide information to assure that the design basis event analyses (Chapter 15 of FSAR) adequately bound other more fundamental credible failums.
This issue was addressed in the letter from R. Tedesco (NRC) to W. Parker (DPC) dated April 16, 1981 as Item 222.04.
In the response to question 222.04 (Volume 13 of FSAR) the applicant states that an analysis will be provided in the future. This item will remain open until the applicant has submitted the required information and our review of the information is completed.
4.
Volume 13, Page 220-7, Response to IEB 80-06.
Provide a schedule for conducting the test to demonstrate that all equipment remains in its emergency mode upon removal of the actuating signal and/or manual resetting of the various isolating or actuation signals.
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- 9.
Prov'ida an ovsrview of the plant. electrical distribution system, with
,, s emphasis on vital buses and separation divisions, as background for addressing various Chapter 7 concerns.
6.
Indicate whether Catawba 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.
7.
Using detailed schematics describe the operation of circuits used for the NSW system. Discuss the design criteria for the instrumentation and control (i.e., indicators available, testability, automatic switchover). Also discuss the interface with the bypass and inoperable status panel.
8.
Using detailed schematics describe the operation of circuits used for isolation of NSW to the air compressors.
Discuss periodic testing and indicate which components (including sensors) 'are located in seismic quali-j fied structures.
9.
For the RTS and ESFAS, revise the FSAR Sections 7.2.2.1 and 7.3.2.1 to include the. basis, assumptions, and results of the referenced FMEA.
Confirm that the FitEA, referenced in Section 7.3.2.1, for ESFAS includes 10.
(1) all B0P scope and (2) design changes subsequent to the design analyzed in the WCAP.
- 11. Section 6. 2.1.1. 3. 2. 2 states tnct " Instrumentation provided to monitor and record the containment pressure and temperature and sump temperature during the course of an accident within the containment is discussed in Chapter 7".
In our review of Chapter 7.we find some information on the containment pressure in Section 7.3.1.1.2 and Tables 7.5.1-1 and 1.5.1-2; however, no information on the instrumentation for the containment ar.d step 6
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12.
Identify all plant safety-related systems, or portions thereof, for which the design is incomplete.
13.
Identify where microprocessors, telemetry systems, multiplexers, or computer systems are used in or interface with safety-related systems.
14.
Identify any sensors or circuits used to provide input signals to the protection system or perform a function required for safety which are.
located or routed through non-seismically qualified structures.
This should include sensors or circuits providing input for reactor trip, emer-gency safeguards equipment such as 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.
Identify any testing or analyses performed which insure that failures of non-seismic structures, mountings, etc. will not cause failures which could interfere with the operation of any other portion of the protection system.
15.
Describe how the effects of high temperatures in reference legs of steam i
l generator water level measuring instruments subsequent to high energy breaks I
are evaluated.
Identify and describe any modifications planned or taken' in response to IEB 79-21.
Describe the level measurement errors due to environ-mental temperature effects on level instruments (excluding steam generator l
level) including reference legs.
1 l
l
- 16. Describe features of the Catawba environmental control system which insure that instrumentation sensing and sampling lines for systems important to Discuss safety are protected from freezing during extremely cold weather.
the use of environmental monitoring and alarm systems to prevent loss of, or damage to systems important to safety upon failure of the environmental
b.
Discuss electrical independence of-the environmental control control system.
system circuits.
Identify the lead-lag constants used in the RPS and ESF's.
17.
Provide and describe the following information for flSSS and B0P safety 18.
Dis-related setpoints: (a) Provide a reference for the methodology used.
cuss any differences between the referenced methodology and the methodolog Catawba. (b) Verify that environmental error allowances are based on the for highest value determined in qualification testing. (c) Identify protection 4
channels where the Technical Specification setpoint, with allowance for channel statistical error, falls within 5% of the instrument range limit 1
For those cases, or within 5% of the range between level measurement taps.
specify the remaining margin to the end of the range. (d) Document the environmental error allowance that is used-for each reactor trip and enginee safeguards setpoint. (e) Identify any time limits on environmental qualifica of instruments used for trip, post-accident monitoring or engineered safety Where instruments are qualified for only a limited-features actuation.
time, specify the time and basis for the limited time. (f) Address the effect of test equipment accuracy on setpoint errors. (g) As an example, derive the setpoints for the low-low steam generator level trip.
The information provided in Section 2.2.1 (including Table 2.2-1) and 19.
3.3-1, 3.3-2, and 4.3-1 of the Tech. Specs. include a trip for Tables This Steam /Feedwater Flow Mismatch and Low Steam Generator Wat Clarify this apparent discrepancy.
trip is not discussed in Section 7.2.1.1.2.
Table 15.0.6-1 lists the following limiting trip points assumed in the 20.
accident analysis:
Power range high-neutron flux, high setting --- 118%
a.
)
b.
Low pressurizer pressure -- 1921 psig.
Low-low steam generator level -- 25% of narrow range level span.
c.
The trip setpoints shown in Table 2.2-1 of the Tech. Specs. for items
- b. and c. above are >1865 psig (allowable values >11155 psig) and >10%
(allowable value >9%) respectively.
These values are considerably lower (less conservative) than those used for the accident _ analysis, rather than being more conservative to account for instrument errors. Also, replacing the 109% value used in Table 15.0.7-1 with the allowable value of 110% of Table 2.2-1, would result in a change in maximum overpower trip setting from 118% to 119%.
Discuss how the results of the accident analysis would be affected if the allowable trip settings given in the Tech.- Specs. would be used in lieu of-the values shown in Table 15.0.6-1.
Revise the allowable trip settings, if needed, to prevent the safe operating limits from being exceeded.
- 21. The trip setpoints specified in Section 7.2.1.1.2 for item 4b, Reactor Coolant Pump Undervoltage, is 70% of rated voltage and for item 4c, Reactor Coolant Pump Underfrequency,is 56 Hz.
No values have been provided_for.these setpoints-in Table 2.2-1 of the Tech. Specs.
Specify the trip setpoints and the allowable values for the above parameters Describe also the eN pment used for the monitoring, and i
in Table 2.2-1.
discuss how the periodic tests on these monitors are performed.
Page 7.3-75 provide a di.scussion of accuracy, or a reference to supplement 22.
the 'ttypical" accuracy information given.
Relate the accuracy requirements of the plant, such as for the safety analyses, to demonstrated equipment accuracy.
^
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- 23. Tables 3.3-3, 3.3-4 and 4.3-2 of the Tech. Specs. include the following parameters for ESFAS:
9 Differential Pressure Between Steam Lines 8 Steam Flow in Two Steam Lines -- High, Coincident with T avg Steam Line Pressure -- Low These parameters are not discussed in Section 7.3.
Clarify this apparent discrepancy.
- 24. Tables 3.3-3, and 3.3-4, Engineered Safety Feature Actuation System Instru-mentation,of Technical Specifications do not include the loss of both Main FW Pumps as a parameter to start the auxiliary feedwater pumps. We find, however, the loss of both haiii Ftf Pumps listed as a parareter to start the motor driven pumps in the _ description of the Auxiliary Feedwater System in Section 10.4.9, and also in Section 7.3.2.3.
Clarify the apparent discrepancy.
25.
Tables 3.3-9 and 4.3-6 in the Tech. Specs. specify the following readouts on the Auxiliary Shutdown Panel:
4 Reactor Coolant Temperature - W/R hot leg 4 Reactor Coolant Flow Rate Tables 7.4.7-1 through 7.4.7-2 do not show monitors for the reactor coolant flow rate and show monitors for the reactor coolant temperature N/R to cold leg vice hot leg.
Clarify the apparent discrepancies.
26.
In the Safety Evaluation Report (and supplements) issued for the Catawba con-struction permit the following items required resolution during the operating licensee review:
(1)
Include activation of the RHR and safety injection pumps as an integral part of the ECCS periodic tests.
(2)
Include in the design the capability to test the reactor trip from a safety injection signal without being restricted T
,t.,. '
o or limited by power operation of the reactor.
Discuss the status of the above items.
27.
Identify where instrument sensors or transmitters supplying information to both a protection channel and control channel or to more than one control channel are located in a common instrument line or connected to a coninon instrument tap.
The intent of this item is to verify that a single failure in a common instrument line or tap can neither defeat required separation between control and protection nor cause multiple control system actions not bounded by analyses contained in Chapter 15 of the FSAR.
For control systems, the discussion can be limited to channels used for control of reactivity, reactor-coolant pressure, reactor coolant temperature, reactor coolant -flow, reactor coolant inventory, secondary system pressure, steam generator feedwater flow and steam generator steam flow.
28.
Describe the scram response time testing.
29.
Portions of paragraph 7.3.1.2.6, subparagraph 1, appear not to apply to ESFAS response times.
In particular, the discussion on reactor trip breakers, latching mechanisms, etc. should be replaced by a discussion of ESF pump and valve time responses.
The applicant should provide a revised discussion for ESFAS (a) defining specific beginning and end points for which the quoted times apply and (b) relating these times to the total delay for all equipment and to the accident analysis requirements.
l 30.
The information provided in Section 7.2.2.2.3 on testing of the power range channels of the Nuclear Instrumentation System, covers only the testing of the high neutron flux trips.
Testing of the high neutron flux rate trips is not included.
I
n-t.
Provide a description of how the flux rate circuitry is tested periodically to verify its performance capability.
In the discussion of the auxiliary. shutdown control in Section 7.4.7.2 it is 31.
stated that "The safety evaluation of achieving and maintaining hot shut-down with the controls available at the auxiliary shutdown panels includes consideration of transients whose consequences might jeopardize the safe shutdown conditions".
Provide a detailed discussion on the type and severity of transients con-sidered, and describe the effect these transients have on the auxiliary shutdown control.
Provide a table showing safe shutdown display information and identify 32.
safety grade items.
Discuss the capability of achieving hot and cold shutdown from outside 33.
the control room including:
a) list of qualified displays, location and basis for selection.
b) description and location of auxiliary shutdown panel or equivalent.
c) description of required controls, 1
d) description of isolation, separation, and transfer / override provisions.
Discuss a typical transfer scenario.
e) description of any communications system required to coordinate operator actions, including redundancy, separation, and environ-mental qualification for local environment, I
f) description of control room annunciation of remote control or overridden status of devices under local control.
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4 g) description of_ any auxiliary system. essential to remote e
shutdown capability.
h) description of control of access to the displays and controls located outside the -control room.
i) testing during reactor operation.
j) means for ensuring that cold' shutdown-'can be accomplished before Technical Specification _ limits on hot shutdown are exceeded.
k) address statemert at bottom of page 7.4-20:
" Cold shutdown conditions can be reached from outside the control room with some temporary instrumentation and control ta modi fication. "
Section 7.a.3.1 states that "There are no bypasses capable of preventing 34.
the Component Cooling Water System from performing'its safety function; however,-in the event system control must be. transferred to the ' auxiliary, shutdown panel, some automatic signals are defeated (refer to Section
-7.4.7)".
Provide a list of the automatic signals that are defeated. -Discuss which controls and interlocks, if.any, that~ originate or pass through the control room and/or cable room, remain active.
As discussed in Section 7.2.2.3, isolated output signals from protection 35.
system channels are utilized to generate a control signal to automatic control systems, such as rod drive system, pressurizer pressure and level control,'and others.
The control signal is derived by 'auctioneering the -
redundant protection system channels to select the high or low signal, whichever is chosen based on consideration of safety in case of a failure.
7 u
' Discuss what steps, if any, are taken to prevent unnecessary control action, such as opening of pressurizer relief valves, during the testing of protection system channels with a signal from a test source.
- 36. Will a test be performed to verify the capability of maintaining the plant in a safe shutdown condition from outside the control room?
37.
Page 7.3-16 Testing During Shutdown.
Describe provisions for insuring that the " isolation valves" discussed here are returned to their normal operating positions after test.
38.
Describe compliance with R. G.1.118 and IEEE Std. 338-1975.
Confirm that tech. specs. provide detailed instructions which insure that blocking of a selected protection function actuator circuit is returned to normal operation after testing.
Confirm that tech. specs. include RTS and ESFAS response times for reactor trip functions.
Confirm that tests include all components, from sensor to operation of final actuation device, and describe a typical response time test.
Indicate any area of non-compliance with basis for each.
39.
Page 7.1-8 The statement is made, "The Westinghouse design of protection systems incorporates overcurrent devices to prevent malfunctions in one circuit from causing unacceptable influences on the functioning of the protection system." Provide information on the specific places where this is done and the basis that this design does not compromise protection channel independence.
Discuss conformance with R.G. 1.75 and IEEE 384-1974.
40.
Describe the design criteria and tests performed on the isolation devices in the NSSS and 80P. Address results of analysis or tests performed to demonstrate proper isolation between separation groups.
~.
The discussion in Section 7.1.2.2 states that Westinghouse tests on 41.
the Series 7300 PCS system covered in WCAP-8892 are considered applicable As a result of these tests, Westinghouse has stated that the to Catawba.
isolator output cables will be allowed to be routed with cables carrying The discussion voltages not exceeding 580 volts a.c. or 250 volts d.c.
of isolation devices in Section 7.5.3.3.9 of the FSAR, however, considered the maximum credible fault accidents of 118 volts a.c. or 140 volts d.c.
Also, the statement in Section 7.7.2.1 implies that the isolation only.
In order devices were tested with 118 volts a.c. and 140 volts d.c. only.
to clarify the apparent nonuniformity, provide the following:
(a) Specify the type of isolation devices used for Catawba Process Instrumentation System.
If they are not the same as the Series 7300 PCS tested by Westinghouse, specify the fault voltages for which they are rated and provide the supporting test data.
(b) Provide information requested in (a) above for the isolation devices of the Nuclear Instrumentation System. As implied in WCAP-8892, the tests on Series 7300 PCS did not include the Nuclear Instrumentation System.
(c) Describe what steps are taken to insure that the maximum credible fault voltages which could be postulated in Catawba, as a result of BOP cable routing design, will not exceed those for which the isolation devices are qualified.
The section covering compliance with R.G.1.53 addresses only 42.
Page 7.1-14.
Provide the equivalent Westinghouse equipment and associated topical reports.
information for the B0P portions of plant safety systems and auxiliary systems required for support of safety systems.
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'43. 'Pages 7.1-14.
The section covering compliance with R.G. 1.47 addresses only ESF systems.
State compliance to R.G. 1.47 for other Catawba safety
'E related systems.
44.
Describe the implementation of the bypassed and inoperable status indication provided for ESF and compliance with R.G. 1.47.
Discuss types of status' displays and alarms. Discuss computer utilization and software verification and validation techniques.
~
- 45. The discussion on the operating bypasses of the Reactor Trip System (RTS) s.
in Section 7.2.2.2.3 (Item 13), refers the bypass indication to Section 7.8.
In our review of Section 7.8 we find Section 7.8.3, titled ESF Bypass In-dication, but we can find no discussion on the RTS bypass indication.
Describe the RTS bypass indication system, provide a list of systems for which bypass indication is provided, and discuss how the bypass indication system conforms to the requirements of Regulatory Guide 1.47.
46.
Page 7.1-16.
Provide a schedule for developing IEEE-338 reliability goals and demonstrating the adequacy of test frequencies.
A7.
Discuss the plans and schedule for complying with R.G.1.97, Rev. 2.
Describe the conformance of the present design.
48.
Describe how separation criteria for protection channel circuits, protection logic circuits, and non-safety related circuits complies with R.G. 1.75.
Indicate the separation method between these circuits.
~
Discuss a typical example for each type circuit including an intra-panel wiring circuit.
.s.
49.
Discuss the method of redundantly tripping the turbine following receipt of reactor protection signals requiring turbine trip.
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Discuss the diverseyfeatures of tha undarvoltage and shuntitrips of the
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reactor trip breakeis'.
Indicate if they Lean be tested independeritly.
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- 51. 'Using detailed schem'atics,/ describe the d'esign cf the pressurizer.PORV control and the block valve control.
Does the current design st'ill pro-vide for actuation of the' pressurizer spray or relief valves up,on a single
-i instrument-fai. lure.
Identify;andEdjscribedesion;featureswhich' ensure x
.s m
s that the RCS pressure is sa'fely controlled during' lcw temperature opera-tion to include parameters utilized and trhnitpred(for ' alarm indication.
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Discuss the degree of' redundancy in the logic for the' low temerature m
a interlock for the RCS pressure control.
Describe the design features used to provi'de ' direct indication of 52.
pressurizer safety and relief valve positions in the control room.
53.
Page'7.2-29. The fourth _paragraphsimplies that a turbine trip may.
open the pressurizer code safety valves.
Please discuss.
- 54. IProvide analysis indicating whether the pressurizer PORV will be actuated follcwing a turbine trip below the power Fd ooint of P-9.
The analysis should cover core physics parameters br,
..g those expec et d throughout the core lifetime.
Describe the electrical power supply arrangement, air supply design 55.
features, and any interlocks associated with control and operation of the steam generator PORV.
Section 7.7.1.7 and Figure 7.7.1-6 conflict concerning the basis for
_55.
g-programming steam generator water level. Also there is no input label s_
for the filter in Figure 7.7.1-6.
Correct FSAR as appropriate.
Describe the steam generater level instrumentation.
Identify the 57.
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- channels for protection functions, control functions, and post-accident t
monitoring.
58.
Provide an analysis indicating the time between reaching each high steam generator level alam setpoint and filling the steam generator with water assuming failure hf the level channel used for'contrci in the low direction.
Since only +wo'out of three logic is used for high steam generator. level, the rereining two channels do not meet the single failure criteria. Assune that the isolation function does not occur. The initial plant power level 3?
'# 'resulting in the most rapid steam generator filling should be assumed. The applicant should be prepared to discuss the probable consequences of filling the steam generator and causing water to flow into the stean piping and the r
consequences of a steam generator level control channel failure, l,
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- 59. Using detailed schematics describe the protection for locked rotor l
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or sheared shaft of the reactor coolant pumps.
l 60.
Describe the procedures to borate the primary coolant from outside the
- control room when the main control room is inaccessib1'c.
How much time is there to do this?
t 61." The discussion of the philosophy of protection for the reactor coolant pumps which'is presented in FSAR Section 7.3.2.3 is inadequate. Therefore:
c (a)' identify any situations in which a control room alarm will not be actuated upon loss of component cooling flow to the re-actor cooling pumps.
(b) Quantify the time delay between a loss of component cooling to the reactor coolant pumps and the initiation of an alarm.
(c) Describe how the operator corrects for a loss of component cooling water flow to the reactor coolant pumps during a seismic event or at.any other time flow is lost.
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t (d) Quantify the time it will take an op:;rator (after an alarn is received) to attempt to take the corrective action which is identified in response to part (e) above and the time which is required to evaluate the results of this attempt.
(e) Describe the consequences of a failure to take effective ccrrective action (including reactor trip) within 10 minutes.
Using detailed system schematics, describe the power distribution for the 62.
accumulator valves and associated interlocks and controls including bypass indicator light arrangement.
The discussion of the accumulator motor-operated valves in Section 7.6.4 63.
Provide does not include information on the valve position indication.
this information and discuss how the requirements 2 and 3 of the' Branch Technical Position ICSB 4 regarding the valve position indication and alarms, are followed.
In the event of a boron dilution transient, describe the operation of the 64.
detection system and the intarface arrangement with the protection system Indicate if the neutron detector is qualified both for valve actuation.
environmentally and seismically.
Confirm quality of detectors is Category I.
Confirm that *,he reactor coolant pump breakers are designed and qualified 65.
to meet all criteria applicable to equipment performing a safety function.
If not, provide the basis for determining that tripping the pump breakers on underfrequency is not a safety function.
Using detailed schematics, verify that no single failure will preclude 66.
reactor coolant system letdown capability.
Discuss whether the motor-operated valves in the safety inje: tion pump 67.
lines from RWST receive an automatic signal following SI initiation.
Confirm that tech. specs. will include surveillance requirements for 68.
the RTD bypass loop flow alarms.
Identify all remotely controlled valves in the Engineered Safety Features 69.
Systems which require power lockout during a certain mode of operation, and discuss how the design meets Branch Technical Position EICSB-18.
Table 1.3.1-1 states that there are no significant differences between 70.
Our review the ESFAS for Catawba and those of FcGuire and Hatts Bar.
shows that several parameters associated with the safety injection and/or containment and steam line isolation for McGuire and Hatts Bar are not provided for Catawba.
These parameters are:
(1) High Differential Pressure Between Steam Lines (2) High Steam Flow (3) Pressurizer Water Level (Watts Bar only)
Although the above parameters are not included in Table 7.3.1-1 and are not discussed in 7.3.1.2.6, credit appears to be taken for monitoring pressurizer water level in 7.3.2.4.1.
Clarify the apparent discrepancies and amend Table 1.3.1-1 as necessary.
Using detailed system schematics, discuss the bypass, bypass interlock, 71.
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.
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'. ~721. Describ] the interface betw:en th; radiation monitoring system (RMS) and ESFAS for containnent ventilation and fuel building isolation to include the use of non-safety grade equipment in RMS and ESFAS.
73.
Describe the method of providinp redundancy for equipment in certain ventilation systems such as Cable Spreading Room, Exhaust Isolation, Control Room Exhaust, Isolation and Control Building Outside Air.
Pre-pare a list of Ventilation Isolation Control System actuated equipment and indicate number of actuation channels for each.
74.
Describe automatic and manual design feature permitting switchover from injection to recirculation mode for emergency core cooling including pro-tection logic, conoonent bypasses and overrides, parameters monitored and controlled, and test capabilities.
Discuss design features which insure that a single failure will neither cause premature switchover nor prevent switchover when required.
Discuss the reset of Safety Injection actuation prior to automatic switchover from injection to recirculation and the po-tential for defeat of the automatic switchover function.
Confirm whether the low-low level refueling water storage tank alarms which determine the time at which the containment spray is switched to recirculation mode are safety grade.
75.
Using detailed system schematics, describe the sequence for automatic initiation, operation, reset, and contrcl of the auxiliary feedwater system. The following should be included in the discussion:
a) the effects of all switch positions on system operation.
b) the effects of single power supply failures including the effect of a power supply failure on auxiliary feedwater con-trol after automatic initiation circuits have been reset in a post accident sequence.
o b
c) any bypassss within the system including the means by which it is insured that the bypasses are removed.
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 This should include required by the auxiliary feedwater system.
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.
h) logic circuits used to transfer pump suction from the Condensate Storage Tank to the Nuclear Service Water System including verification that all equipment used for this function is seismic-ally qualified and protected from failure of near-by structures which may not be seismically qualified.
- 1) design features to insure that no single failure can result in an open flow path from the Nuclear Service Water System to the Condensate-Storage Tank.
The information in Section 10.4.9 on the auxiliary feedwater system does 76.
not specify the criteria applied in the design of the control and instru-Describe the mentation systems for the modulating level control valves.
. instrumentation and controls, identify the power sources used for each of the valves, and provide an analysis to show that no single failure can pre-vent supplying auxiliary feedwater when required.
e
'7 7<, Describe the instrumentation provid;d for monitoring tha loss of both main feedwater pumps and how the design meets the requirements of IEEE 279-1971.
78.
Revise Table 7.3.1-3 to include, under P-ll, the automatic resetting of the
" Auto-Start Defeat" logic for auxiliary feedwater pumps as discussed on page 7.4.3.
79.
For main steam and feedwater line valve actuation, describe control circuits for isolation valves and include automatic, nanual and test features.
In-dicate whether any valve can be manually operated and whether each valve actuation level is alarmed in the control room.
Indicate specific inter-faces with the safety systen electrical circuits.
80.
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 instru-mentation is within each CCW train or is accomplished by having one inter-lock per train.
81.
Regarding CCW pumps; discuss (a) circuits which automatically start a pump in a CCW train on low pressure in the pump discharge and (b) circuits which automatically start a pump in the operating train on an SIS.
Include subsequent operator control of the pumps.
82.
Using detailed s:.hm tics, describe the sequence of operation for the RHR isolation valves.
The discussion should include the effects of various single failures in power supplies for the valves and their controls.
Indicate any single instrument bus failure which could cause inadvertent closure of RHR suction valves in both trains when the system is in use for decay heat removal.
Describe how proper operation of the RHR isolation t
e
',I rs, valves will be insurcd in th2 cv::nt of interlock or power failura.
Id::ntify i
testing planned for this area.
83.
Identify and describe features of the RHR system motor operated isolation valve interlocks designed to prevent overpressurization of the RHR system to include separation and independence measures.
Discuss compliance with ICSB BTP No. 3, 84.
As described in Section 10.4.9.2, the Nuclear Service Water System is used in emergencies to supply water to the auxiliary feedwater motor-driven pumps.
This supply is initiated automatically by a two-out-of-three low pressure signal in the condensate suction line.
Describe the type and qualification of the pressure monitoring equipment and circuits.
85.
Table 7.5.1-1 shows that the Post-Accident Monitoring System includes two Wide-Range T channels, and two T channels.
It is stated that "The hot cold T
channels are on separate power supply from the T channels." From hot cold this statement we conclude that both T channels are on one power supply, hot and both T channels are also on one power supply.
This would imply cold that a loss of a power supply would result in a loss of data from either both T or T channels whichever is applicable.
Discuss what consider-hot cold ation was given to the loss of T or T data.
hot col d 86.
Using detailed schematics, describe the operation of the UHI systen.
I From the description of upper head injection (UHI) interlocks in Section 7.6.3, it appears that the requirements of Branch Technical Position ICSB 4 in providing automatic opening of the valves whenever either primary coolant system pressure exceeds the preselected value, or a safety injection signal is present, are not folkwed.
If so, justify the approach taken.
Also confirm that the ac control power supply used for
',as 4 the valve position indicating lights is independent of the power supply used for the annunciators that alarm if the valve is not fully closed above a set pressure, as required by ICSB 4.
87.
Using detailed schematics, describe the operation of the containment spray system.
Discuss the redundancy of the spray additive tank iso-lation valves which are closed on low additive tank level.
Describe how the transmitter used to close these valves are monitored.
Indicate periodic test requirements for the instrumentation and controls used.
88.
Describe the design features used in the rod control system which
- 1) Limit reactivity insertion rates resulting from single failures within the system.
- 2) Limit incorrect sequencing or positioning of control rods.
The discussion should cover the assumptions for determining the maximum control rod withdrawal speed used in the analyses of reactivity insertion transients.
89.
List the basis, assumptions and results from the FMEA (WCAP-8976) for the rod control systems.
90.
Section 7.8.2, Monitor Light Panels, includes a statement that "An energized light on a monitor panel normally indicates that the monitored equipment is in its safety position or mode.
Exceptions to this convention are identified to the operator".
Discuss the reasons why the exceptions were taken, list the equipment involved, and describe how the exceptions are identified to the operator.
[,
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- .. 1,4, 91.
Using detailed schematics describe the design of boric acid addition control and the volune control tank level control. Discuss the recent Westinghouse generic deficiency regarding volume control tank level and its applicability to Catawba.
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