ML19209B726
| ML19209B726 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 10/02/1979 |
| From: | Stewart W FLORIDA POWER CORP. |
| To: | Reid R Office of Nuclear Reactor Regulation |
| References | |
| 3--3-A-3, 3-0-3-A-3, NUDOCS 7910100361 | |
| Download: ML19209B726 (26) | |
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3-0-3-a-3 Mr. Robert W. Reid, Chief Operating Reactors Branch No. 4 Division of Operating Reactors U.S. Nuclear Regulatory Commission Wa r>hing ton, D.C. 20555
SUBJECT:
Crystal River Unit 3 Docket No. 50-302 Operating License No. 09R-72
==Dear
>k. Reid:==
On September 19, 1979, Florida Power Corporation received your letter dated Septecher 7,1979 requesting additional information concerning the upgrade of the anticipatory reactor trips on turbine trip or loss of main feedwater at Crystal River Unit 3.
You specifically requested Florida Power Corporation to provide the additional information identified in the enclosure of your letter and evaluate the possibility of improving the installation schedule for the safety grade anticipatory trips previously identified in our response to IE Bulletin 79-05B.
In that regard, enclosed is our response to items 1-9 identified in the enclosure of your letter.
Our previous implementation schedule for this design modification was approximately 12 months following NRC approval of our proposed design.
This schedule was based on the long lead times necessary for the manufacture, delivery and installation of safety grade equipment.
As indicated in our response to item 5, we have been able to locat a some existing qualified equipment from a utility who is ext ariencing delays in construction of its nuclear plant.
These components can be delivered to CR#3 within 22 weeks from the time of NRC appreial of our proposed design.
The installation of this equipment would be accomplished at the first refueling outage or outage of,af ficient duration to permit installation, following receipt of the equipment on site.
We believe that this improved schedule is consistent with your September 7,1979 letter.
h b0 n~q0JU
/d 7910eao % Gl r.
Petershiore n733 813-ee,6-sisi General Office 3201 Thirty-fourin street souin. e o Box 14042. st
Mr. Robert W. Reid Page Two October 2, 1979 Please contact us if you wish to discuss our response further.
Very truly yours, FLORIDA POWER CORPORATION W.2 uuo W. P. Stewart Manager, Nuclear Operations ECShewT01 D88N 1138 039
STATE OF FLORIDA COLNTY OF PINELLAS W.
P. Stewart states that he is the Manager, Nuclear Operations, of Florida Power Corporatioe; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the information attached hereto; and that all such statements made and matters set forth therein are true and correct to the best of his knowledge, information and belief.
I f
di.kAk
. V.
P.
Stewart Subscribed and sworn to before Le, a Notary Public in and for the State and County above named, this 2nd day of October,1979.
!w Notary Public Notary Public, State of Florida at Large, My Commission Expires: August 8, 1983 CameronNotary 3(D12)
QUESTION 1.
For your proposed design, state the degree of conformance with the acceptance criteria listed in Column 7.2 of Table 7-1
(" ACCEPTANCE CRITERIA FOR CONTROLS") of the Standard Review P;an.
Justify any non-conformance.
QUESTION 2.
Provide a discussion of the following:
a.
design basis information required by Section 3 of IEEE-2 79-1971, and b.
conformance with the design requirements of Section 4 of IEEE-279-1971.
RESPONSE TO QUESTIONS 1 AND 2:
The proposed design for safety grace anticipatory trips contains four redundaat and independent channels which monitor the main feedwater pumps and the turbine.
This equipment will initiate an RPS reactor trip on the tripping of both main feedwater pumps or on a turbine trip.
The cabinet mounced equipment will be installed in and become an integral part of the existing four channel RPS-1.
As auch, the additional equipment will meet or exceed the design bases of the RPS at CR #3 and will meet or exceed the acceptance criteria and design requirements of the RPS.
The description of the conformance of the RPS with the acceptance criteria and design require-ments of Crystal River Unit 3 can be found in Section 7.1 of the FSAR.
QUESTION 3.
Provide a description of any changes to and/or interfaces with the existing protection system.
Include diagrams (block, location, functional and/or elementary wiring), as necessary, to clearly depict the changes and/or interfaces.
In addition, provide an analysis which demonstrates that these changes and/or interfaces will not degrade the existing protection system.
RESPONSE TO QUESTION 3:
The anticipatory trip equipment will be added to the RPS cabinets and will interface as new trips in the present bistable trip string.
Figures 1 and 2 of the Attachmet.t I show the functinal interface of the added equioment with the RPS.
Drawings 51079DBG-1 and 51079MLG-1 of the attachment describe the inputs, outputs, and logic of the new trip functions.
The added modules will consist of contact buf fers, bista" 1,
and auxiliary relays, which have been tested and qualified for use 1 safety system.
Existing RPS power supplies, flux signals, interlock circuits, and indica-tors will be used as r2 quired by the added equipment.
The requirements for the RPS, e.g., cooling, power, seismic, environmental, will be the same for the system with anticipatory trips as the requirements prior to addition of the new trips.
A failure analysis of the RPS-1 was performed and is contained in Topical Re po rt BAW-10003, " Qualification Testing of Protection System Instrumen-tation."
This fa:. lure analysis was predicated on the use of qualified ECShewT01 D88 1138 041
4 RESPONSE TO QUESTION 3 (Cont'd) modules and concluded that any single failure in the RPS will not prevent performance of its protection action when required.
The added equipment uses qualified modules and the failure analysis of BAW-10003 is still applicable to the RPS containing anticipatory trips.
The anticipatory trips provide additional protection and conservatism beyond that provided by the rest of the RPS.
No credit is taken for any of these trips in the FSAR accident analyses.
The sensors will be redundant, separated, and designed such that a single failure will not prevent them from performing their intended f unction.
The sensors are anticipatory to other diverse parameters which will cause a reactor trip.
Thus, the protection system will not be degraded br thesu trips since functioning of the anticipatory trips is not required to provide safety action and contact isolation of 500 volts is provided.
The sensor contacts are closed during normal operation and open to cause a reactor trip when either the main feedwater pumps trip or the turbine trips.
The contacts serve to interrupt power to cause a reactor trip.
Loss of power to the trip circuitry will also initiate a reactor trip.
QUESTION 4.
Identify equipment which is identical to equipment utilized in existing safety grade systems.
For the equipment which is not identical, briefly describe the differences.
RESPONSE TO QUESTION 4:
The equipment to be used are bistables, contact buf fers, and auxiliary telays.
These modules are updated versions of modules already in use in B&W safety systems of the operating plants.
Significant changes are:
the bistable output to the RPS trip string has been converted from a relay contact to a solid state output; the contact buffer now uses one transformer with a rectified output to monitor the field contacts instead of two transformers with AC outputs; and transistor buffer amplifiers for driving relay coils from current liaited grounded input signals have been added to the auxiliary relay.
These changes were made to improve the performance of the modules.
QUESTION 5.
For all critical components, provide an expected delivery date.
RESPONSE TO QUESTION 5:
The Reactor Protection System components, contact buf fers, bistables, and auxiliary relays, which are the critical path components in this design, are available from existing systems which have been delayed in construction.
These components can be delivered to CR #3 within 22 weeks from the time of NRC approval of our proposed design.
ECShewT01 D88 1138 042
QUESTION 6.
In general, the equipnent shall be seismically and environmentally qualified.
Therefore, provide the following descriptive informatioa for the qualification test program:
a.
equipment, design specification require" tt s,
b.
test plan, c.
test setup, d.
test procedures, and e.
acceptability goals and requirements.
RESPONSE TO QUESTION 6:
The modules to be used have been qualified for use in B&W safety systems.
Attachment II contains the seismic and environmental summary reports f)r each module which describe the test programs and report the acceptability of the modules.
The detailed test procedures and test data are available for audit.
QUESTION 7.
Identify the portion (s) of the design which are within the scope of supply of B&W and/or other contractors.
RESPONSE TO QUESTION 7:
The B&W scope of supply is limited to the RPS modules, i.e., contact buffers, bistables, and auxiliary relays centained within the RPS system cabinets.
The cable, relays, and pressure swir.ches used in the design will be purchased by Florida Power Corporation directly.
QUESTION 8.
Provide the criteria for the overall reactor protection system installation testing which will demonstrate that the new trip haa been installed properly.
If this information is not available at this time, provide a schedule for its submittal.
RESPONSE TO QUESTION 8:
Detailed installation instructions and test procedurcs uill be providec cc ensure that the anticipatory trip equipment is properly installed and performs the functions described.
In addition, the cabinet mounted equipment to be supplied by B&W will be f ully testable from the RPS cabinets.
The equipment will have provision for simulating input signals and verifying the proper response of the RPS channel.
This testing will be similar to that presently performed on the RPS and will be integrated into the periodic testing of the cabinets.
Cverall reactor protection system installation testing to be performed by the utility could include actuation of the new sensors and verification of proper reponse at the CRDCS cabinets.
ECShewT01 D88
t QUESTION 9.
The safety evaluations for the anticipatory trips are either missing or are incomplete.
Submit supporting analysis to justify the proposed trip signals by addressing the following items:
a.
Provid< an analysis that quantifies the improvement in the time-to-reactor-trip for both the turbine trip and the loss of main feedwater signals; b.
Address the need to bypass these trips at 20% power versus bypass at a lower power (approximately 10%);
c.
Discuss the adequacy of t'ae proposed trip signals for loss of main feedwater 5.or a variety of iailure scenarios (such as feedwater valv.t closures),
i.e., see the Oconee 1 transients of 6/11/79; and d.
Provide an evaluation of why a reactor trip on low steam generator level is not a viable anticipatory trip signal when the other signals are bypassed, i.e., see tha Crystal River 3 transt:nt of 8/2/79.
RESPONSE TO QUESTION 9:
a.
The primary purpose of anticipatary reactors trips (ARTS) is to reduce the proba'aility of lif ting the PORV for turNine trip / loss of main feedwater type events.
For a reactor high pressure trip setpoint of 2300 psig, it was shown in References 1 and 2 *. hat the PORV would not lift with a setpoint of >2400 psig.
The margin to the PORV setpoint can be increased, however, by use of ARTS.
Figure 9a-1 shows the pressure increase from nominal operating pressure as a func' tion of time to trip for the loss of main feedwater event.
From this figure, it can be seen that an ART that detects and trips the plant at 4 seconds results in a peak pessure increase of 60 psi; whereas the high R.C. pressure trip which would occur at 8 seconds resalts in a peak pressure increase of 184 psi.
The anticipatory trip signals which have been selected will initiate a reactor trip in less than one second.
As seen on Figure 9a-1, a one second time to trip results in a 12 psi pressure increase, compared to a 184 psi pressure increase for the high pressure trip at 8 seconds.
The analyses presented above are for a loss of main feedwater event which produces higher peak pressures than turbine trips produce.
The time to reactor trip af ter a turbine trip from full power is, however, approximately the same as that for a loss of main feedwater.
b.
Sensitivity studies on time to reach the PORV setpoint vs. power level for a loss of feedwater event have been performed.
Table 9B-1 displays the results of these analyses.
The results are for a trip on high RC presaure since that gives the shortest tiue to steam generator ECShewT01 D88 1138 044~
RESPONSE TO QUESTION 9 (Cont'd) dryout assuming no auxiliary feedwater.
For power levels <25% FP, it can be seen that sufficient time for operator action exists to initiate feedwater and any bypass setpoint below this value should be a matter of providing sufficient operati nal flexibility.
For the turbine trip event, the system 14 4 sufficient res ponsiveness such that, at lower power levels (<25%), a reactor trip is not anticipated if the turbine trips.
The power level >25% fo; which the turbine trip-reactor trip may be bypassed is plant specific, and is dependent on the condenser bypass and atmospheric dump valve capacities, c.
The Oconee 1 transient of 6/11/79 was a reactor startup situation with 1 MFWP reset and not operating.
When the operating feedwater pump tripped, the reactor did not automatically trip on loss of feedwater because the low discha$ge pressure trip on the reset MFWP was not reached prior to the operator manually tripping the plant.
There are two important points to be made with resoect to the above situation.
First of all, a reactor trip based on feed pump operation, such as the proposed safety grade trips will be, would have detected this loss of feedwater event.
Secondly, at a startup condition such as this transient occurred at the ARTS would have been bypassed.
However, as discussed in res ponse to b. above, there is sufficient operator action time.
It should also be noted that the purpose of ARTS is to decrease the probability of PORV actuation on turbine trip / loss of main feedwater type events.
Since it has been demonstrated (1,2) that with a reactor trip of 2300 psig and PORV setpoint >2400 psig, no lifting of the PORV will occur, the addition of ARTS only increase the margin to PORV setpoint pressure.
d.
The Crystal River transient of 8/2/79 was similar to the Ococee transient briefly described in c. above, only the operating pump lost flow slowly and the reactor trip was by automatic control grade trip on low steam generator level instead of a manual trip.
The RC pressure rise @v2255 psig at time of trip) would have tripped the plant had the level trip not functioned.
As was shown in Response 9b, an ART in this power level is not really needed, although it may indeed trip the plant before the high RC pressure trip.
REFERENCES:
1.
B&W Report to the NRC, May 1, 1979, " Evaluation of Transient Behavio r and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant."
2.
Toledo Edison Report to the NRC, May 15, 1979, " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant, Volume 3."
ECShewT01 D88 r
i138 043
ATTACHMENT I NEW SAFETY-GRADE REACTOR TRIPS FOR RPS-I 1138 046
CABINET MOUNTED EQUIPMENT FOR ADDITION OF RPS TRIPS ON LOSS OF MAIN FEEDWATER AND TURBINE TRIP 3 Contact Buffers 2 Bistables Per Channel 2 Auxiliary Relays Modules will be installe in a pre
' red mounting case and tested as a unit prior to shipment.
The mounting cc.e is to be installed in an empty row of each RPS channel and connections made to the RPS wiring.
Trip response time of the RPS cabinet mounted equipment will be 5150 ms.
Isolation of the contact buffer nodule is 600 volts with the contact input lines not grounded.
Customer contact input requirements:
Continuous 90 ma, P-P Surge 250 ma, P-P Vol tage 118 VAC Closed contact indicates pump running Open contact indicates pump tripped 1133 047
This report describes the implementation of safety-grade reactor trips into the RPS-I for loss of main feedwater and turbine trip.
Loss of Main Feedwater Trio - Control cil pressure switches on both main feedwater pumps will input an open indication to the RPS on feedwater pump trip.
Contact buffers in the RPS will sense the contact inputs and initiate an RPS trip when both pumps have tripped.
This trip will be bypassed below a predetermined flux level, typically 20", FP.
Reference Figure 1.
Turbine Trio - Contact outputs from the main turbine electro-hydraulic control unit wi!1 input an open indication to the RPS on turbine trip.
Contact buffers in the RPS will sense the contact inputs and initiate an RPS tri >
when < turbine trip is indicated.
This trip will be bypassed below a pre-determined flux level, typically 20% FP.
Reference Figure 2.
Pressure switches for both trips will be supplied by the customer.
B&W will supply all RPS cabinet mounted equipment.
Attachment I lists the cabinet mounted equipment and gives the trip response time. also gives the contact buffer isolation voltage and the customer requirements for the contact inputs.
Figure 1 is a simplified drawing of the main feedwater pump trip.
Figure 2 is a simplified drawing of the turbine trip.
Drawing 510790GB-1 shows the generic logic for the new trips.
Drawing $1079MLG-1 is a legend for the generic logic drawing.
1138 048
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ATTACHMENT II SEISMIC AND ENVIRONMENTAL
SUMMARY
REPORTS 1138 055
Equipment:
The Bailey Controls Company Solid State Bistable module P/N 6628492A1 was environmentally tested providing type test data.
The Solid State Bistable is a standard 2-unit-wide mcdule designed for plug-in moun ti ng.
Brief Summary of Test Results:
Tests were performed to verify that the performance characteristics of the Solid State Bistable Module qualify it for use in a Nuclear Power Generating Facility.
The test unit failed to meet the acceptance criteria for output load vol tages during humidity effect tes ting.
Engineering replaced a reference amplifier.
Upon ratest, the module met specified acceptance criteria.
Based on the test data, the Solid State Ristable meeting all the design relge ' requirements.
I138 056
Test title Units Units sequence Set up conditions Environment conditions Acceptance criteria tested acceptable Solid State Bistable
- 1. Repeatability of set
- a. Nonnal input / output Standard test conditions
<0.02% set point span 2
2 point trip conditions configuration Temperature: 75 F i.5%
- b. Powe upplies il5V Humidi ty: 50% i 20% Ril
- c. Load: 3K ohms
- 2. Power supply effect Same as test No.1 Standard test conditions For 1% variation <0.02%
2 2
except power supplies:
set point span 115 dc with 11% varia-For 25% variation <0.1%
tion from references set point span repeated using 15%
variations
- 3. kbient temperature Same as test No.1 Temperature: 40 F to for 40 F to 140 F 2
2 effect except set internal 1400F Trip point accuracy <0.1%
set point span shift-set point to 8.00 V dc tiumi di ty : R}i <50%
and apply external set Response Time: >32 ms-low paint voltage Ivl contact volt. <0.5V dc High Ivl contact vilt.
<2.0V dc
- 4. An' nt relative Same as temperature Temperature: 110 F Trip point accy: <0.1% set 2
2 hu. dity effect tests p int span chge in internal tiumi di ty : 80% R}i for set point <0.1% low lvl con-96 h tact <0.5fdc-hi Ivl con-0 H for f4 tact 32.0V dc u
Response Time: 332 ms m
- 5. Dri f t, long term Same as test No.1 Standard test Change in trip accuracy 2
2 C3 (30 day) except setpoint ad-condi tions
-<0.07% setpoint span Ul justed to 9.00V dc,
N
Equipment:
The Bailey Controls Company Solid State Bistable Module, P/N 6628492A1, was seismically tested providing type test data.
Test Mounting:
The Solid State Bistable Module was mounted in a standard Module Mounting Case with backplate.
A standard 32-blue ribbon connector was utilized to interface the module with the signal source.
The standard Module Mounting Case was then securely attached to the Seismic Test Mounting Box.
The Seismic. Test Mounting Box was attached to the Qualifica-tion Test Lab-45 Biaxial Test Table.
The uae of the 450 Biaxial Table results in equal horizontal and vertical components.
Electrical interface, nardware, and mounting were equivt. lent to field installation.
Seismic Testing:
Exploratory Testing:
The resonant survey consisted of a sinusoidal vibration input of 0.2 9's vertical peak acceleration at frequencies from 1 to 35 Hz and back to 1 Hz.
The resonant survey was conducted at a sweep rate of I octave /mi nute.
The constant input was applied to the 450 Biaxial Table and continuously monitored.
Proof Testinc:
A biaxial multifrequency excitation was applied to the Solid State Bistable for a period of 30 seconds.
Each 30-second event consisted of dependent biaxial pseudorandum excitation.
The random input frequencies were adjusted in 1/3-octave bandwidths until the Test Response Spectrum (TRS) enveloped the Required Response Spectrum (RRS) within the limits of the biaxial table displacement.
A damping of 5 percent (Q of 10) was utilized for the control accelerometer in testing.
The TRS did not envelope the, RRS (below 6.0 Hz worst case) in the low-frequency range.
No resonant frequencies exist in the range not enveloped during test; therefore, this is an acceptable deviation.
Test Monitoring:
Seismi c:
The Solid State Bistable Module was monitored with accelerometers through appropriate signal conditioning to determine its mechanical response.
The location of the monitoring accelerometers is delineated in the seismic report.
The control accelerometer was mounted directly to the biaxial test table for controlled input.
Electrical :
The unit's outputs were monitored and documented on a strip cnart recorder during these events.
Brief Summary of Test Results: The Solid State Bistable Module was within the specifications cited in the module test procedure acceptance criteria section during and af ter the SSE tests.
Consequently, the Solid State Bistable Modules are considered qualified for nuclear applications.
Specified Features Demonstrated by Test:
The purpose of this test was to satisfy seismic level testing requirements before, during, and after test of the Solid State Bistable.
Module functional operability and solid-state relay state were maintained throughout the exploratory and seismic events.
Structural integrity of enclosures was maintained.
nr8 1138s w
Equioment: The Bailey Controls Company Contract Buffer, P/N 6628908A1 was environmentally tested providing type test data.
The Contact Buffer module is a 2-unit-wide module designed for plug-in moun ting.
Electrical connections are made through a standard 32-pin Blue Ribbon connector at the rear of the module.
The vi tal bus uses a separate plug-in connec to r.
Brief Summary of Test Resul ts:
Tests were performed to veri fy that the performance cnaracteristics of the Contact Buffer module qualify it for use in a Nuclear Power Generating Facility.
Based on the test data, the Contact Buffer abdule meets all the design range
~
requi remen ts.
Type Test Justification: Because of the nature of application, this product consists of various types, versions, or ranges.
A worst case representative sampling has been tested by BCCo Qualification Test Laboratory to verify that this product performs the required functions within the required operating and environmental conditions.
Part Number Nature of Difference 6628908A2 Variation of Frontplate Silkscreening 6
I138 059
/
e a
Contact Buffer Test title Units Uni ts sequence Set up conditions Environnent conditions Acceptance criteria Tested accep table l
- 1. Functional test
- a. Normal input / output Standard test conditions No faulty operation 2
2
. configuration Temperature: 75 F 5F j
- b. Power supply: 118 Humidity: 50% i 20% RH V ac f
- c. Load
- 2. Power supply ef fect Sane as test No.1 Standard test conditions Same as test No.1 2
2 except power supplies:
Minimum: 105 V ac Maximum: 130 V ac
- 3. Anbient tempera-Same as test No.1 Temperature: 40 F to No fault operation for 2
2 ture for function test 1400F function test.
Vac = 106 for re.
Ilumi di ty: Rif <50%
<12 ms. for response time sponse time tes t tes t
- 4. Acbient relative Same as tes t No. 3 Temperature: 110 F Sane as test No. 3 2
2 humidi ty ef fect Humidi ty: 80% Mi for 96 h 90% mi for 24 h
- 5. Dri f t, long term Sane as tes t No. 1 Standard test conditions Relays do not change state 2
2
~
(30 day) wi th both relays during drif t period energized during drif t test CO CJ~
CD
Equi pmen t:
The Bai1ey Controls Company Contact Buffer Module, P/N 6628908Al, was seismically tested providing type test data.
Test Mounting:
The Contact Buffer Module was mounted in a standard Rodule Mounting Case with backplate.
A standard 32-pin blue ribbon connector and a separate standard 2-prong connector for the vital bus were utilized to interface the module with the signal source.
The standard Module Mounting Case was then securely attached to the Seismic Test Mounting Box.
The Seismic Test Mounting Box was attached to the Qualification Test Lab 450 Biaxial Test ' Table.
The use of the 450 Biaxial Table results in equal horizontal and vertical components.
Electrical interface, hardware, and mounting were equivalent to field ins tal la tion.
Seismic Testina:
Exoloratory Testing:
The resonant survey consisted of a sinusoidal vibration input of 0.2 g's vertical peak acceleration at frequencics from 1 to 35 Hz and back to 1 Hz.
The resonant survey was conducted at a sweep rate of 1 octave / minute.
The constant input was applied to the 45 Biaxial Table and continuously moni tored.
Proof Testing: A biaxial multifrequency excitation was applied to the Contact Buffer Module for a period of 30 seconds.
Each 30-second event consisted of dependent biaxial pseudorandom excitation.
The random input frequencies were adjusted in 1/3-octave bandwidths until the Test Response Spectrum (TRS) enveloped the Required Response Spectrum (RRS) within~ the limits of the biaxi5l table displacement.
A damping of 5 percent (Q of 10) was utilized for the control accelerometer in testing.
The TRS did not envelope the RRS (below 5.0 Hz worst case) in the low-frequency range.
No resonant frequencies exist in the range not enveloped during test; therefore, this is an acceptable devi a tion.
Test Monitorino:
Seismic:
The Contact Buffer was monitored with accelerometers through appropriate signal conditioning to determine its mechanical response.
The location of the monitoring accelerometers is deljneated in the sei mic report.
The control accelerometer was mounted directly to the biaxial test table for controlled input.
Electrical :
The unit's outputs were monitored by chatter detectors per MIL-STD-202D, Method 310.
Resul ts :
The Contact Buffer was within the specifications cited in the module test procedure acceptance criteria section during and after the SSE tests.
Consequently, the Contact Buffer Modules are considered qualified for nuclear applications.
Specified Features Demonstrated by Test:
The purpose of this test was to satisfy seismic level testing requirements before, during, and af ter test of the Contact Buffer.
Module functional operability and predetermined relay state were maintained throughout the exploratory and seismic events.
Structural integrity of enclosures was maintained.
1138 061
Equipment:
The Bailey Controls Company Auxiliary Relay, P/N 6628807 B1 was environmentally tested providing type test data.
The Auxiliary Relay Module is a 2-unit-wide module designed for plug-in mounting.
Brief Summary of Test Results:
Tests were performed to verify that the performance characteristics of the Auxiliary Relay-qualify it for use in a Nuclear Power Generating Facility.
Based on the test data, the Auxiliary Relay meets al1 the design range requirements.
1138 062
P i
Auxiliary Relay P/N 6628807 B1 Test title uni t Uni ts sequence Set up conditions Environment conditions Acceptance criteria tested acceptable
- 1. Functional veri fi-
- a. Normal input / output Standard test conditions Proper operatidn of relays 1
1 ca tion configuration
~
0 Tenperature: 75 F i 5 F
- b. Power supplies Humi di ty: 50%
20f RH
-15 V de
- c. Load; none
- 2. Power supply Same as test No.1 Standard test conditions Same as test 1 1
1 effect(DC) except power supplies:.
from -13.5 V dc to
-16.5 V dc
- 3. Ambient temperature Same as test No.1 Temperature: 40 F to Same as test 1 1
1 effect 1400F Humi di ty : RH < 50%
- 4. Ambient relative Same as test No.1 Temperature: 110 F Same as test 1 1
1 humidity ef fect Humidi ty: 801 RU for 96 h 9C% Mi for 24 h
- 5. Dri f t, long tenn Same as test No.1 Standard test conditions Sane as test 1 1
1 (30 day)
, - =
- b CX3 C7' W
Equipment: The Bailey Controls Company Auxiliary Relay, P/N 6628807Bl, was seismically tested providing type test data.
Test Mounting: The Auxiliary Relay Module was mounted in a standard Module Mounting Case with backplate.
Two standard 32-pin blue ribbon connectors were utilized to interface the module with the voltage source.
."'The standard Module Mountir)g Case was then securely attached to the Seismic Test Mounting Box.
The geismic Test Mounting Box was attached to the Qualification Test Lab 45 Biaxial Test Table.
The use of the 45 Biaxial Table results in equal horizontal and vertical components.
~
Electrical interface, hardware, and mounting were equivalent to field installtion.
Seismic Test:
Exploratory Testina:
The resonant survey consisted of a sinusoidal vibration input of 0.2 9's vertical peak acceleration at frequencies from 1 to 35 Hz and back to 1 Hz.
The resonant survey was conducted at a sweep rate of g
1 octave /ninute. The con ~stant input was applied to the 45 Biaxial Table and continuously monitored.
Proof Testing:
A biaxial multifrequency excitation was applied to the Auxiliary Relay Module for a period of 30 seconds.
Each 30-se'cond event consisted of dependent biaxial pseudorandom excitation.
The random input frequencies were adjusted in 1/3 octave bandwidths until the Test Response Spectrum (TRS) enveloped the Required Response Spectrum (RRS) within the limits of the biaxial table displacement.
A damping of 5 percent (Q of 10) was utilized for the control accelerometer in testing.
The TRS did not envelope the RRS (below 7.0 Hz worst case) in the loGfrequency range. No resonant frequencies exist in the range not enveloped during test; therefore, this is an acceptable deviation.
Test Monitoring:
Seismic: The Auxiliary Relay Module was monitored with accelerometers through appropriate signal conditioning to determine its mechanical response.
The location of the monitoring accelerometers is delineated in the seismic report.
The control accelercmeter was mounted directly to the biaxial test table for controlled input.
Electrical :
The unit's outputs were monitored with chatter detectors per MIL-STD-202D, Method 310 during these events.
Brief Summary of Test Results: The Auxiliary Relay Module was within the specifications cited in the module test procedure acceptance criteria section during and after the SSE tests.
Consequently, the Auxiliary Relays are con-sidered qualified for nuclear applications.
Specified Features -Demonstrated by Test:
The purpose of this test was to satisfy seismic level testing requirements before, during, and after test of the Auxiliary Relay Module.
Module functional operability and predetermined relay state were maintained throughout the exploratory and seismic events.
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1138 065 e
e t
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TABLE 98-1 POWER LEVEL SENSITIVITY TIME TO REACH TDiE TO FILL POWER LEVEL PORV SETPOINT PRESSURIZER 100%
3 min.
10 min.
75%
6 min.
11 min.
50%
12.3 min.
13 min.
25%
>>15 min.
16.6 min.
NOTE:
RESULTS ARE FOR I"dE CASE OF NO AUXILIARY FEEDWATER INITIATION WHICH RESULTS IN THE SHORTEST ACTUATION TIMES.
REACTOR TRIPS ON HIGH RC PRESSURE TRIP (2300 psig).
ECShewT01 D88 1138 366