ML20215H572
| ML20215H572 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 04/14/1987 |
| From: | Zimmerman S CAROLINA POWER & LIGHT CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| NLS-87-074, NLS-87-74, NUDOCS 8704200418 | |
| Download: ML20215H572 (35) | |
Text
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Cp&L Carolina Power & Light Company APR 141987 SERIAL: NLS-87-074 10CFR50.62 United States Nuclear Regulatory Commission ATTENTION: Document Control Desk Washington, DC 20555 BRUNSWICK STEAM ELECTRIC PLANT UNIT NOS.1 AND 2 DOCKET NOS. 50-325 & 50-324/ LICENSE NOS. DPR-7l & DPR-62 ATWS - SYSTEM DESCRIPTIONS Gentlemen:
On July 1,1986, Carolina Power & Light Company (CP&L) submitted a letter stating that detailed information regarding the Alternate Rod injection (ARI) and Standby Liquid Control (SLC) systems would be submitted af ter issuance of the NRC Safety Evaluation Report (SER) of the BWR Owners' Group licensing 21,1986 and included m, topical report (N a request from the NRC da+ed SER was issued on October January 8,1987 for plant-specific reviews related to the Anticipated Transients Without Scram ( ATWS) Rule (10CFR50.62).
In response to both items referenced above, CP&L hereby submits information to demonstrate the adequacy of the SLC, AR1, and Recirculation Pump Trip (LC system.
RPT) systems. Enclosure 1 contains the information requested concerning the S The details of the BSEP ARI design and its conformance to the ARI design basis requirements and objectives are contained in the checklist included in Enclosure 2. The RPT design is described, and a comparison of the BSEP design with the Monticello design is included in Enclosure 3.
The SLC system operation will follow the two-pump operation requirements specified in the BWROG topical report and approved in the SER. The ARI design also conforms to the design basis requirements provided in the topical report and approved in the SER.
The RPT system design makes use of single-trip coils, an op'. ion provided in the topical report, but not approved in the SER. Carolina Power & Light Comp,any believes the system description provided herein will justify approval of our existing RPT system design. Approval of the BSEP RPT design is required by July 1,1987 to support its installation in the Unit 2 outage currently scheduled to commence January 2,1988.
Please refer questions regarding this submittal to Mr. S. D. Floyd at (919) 836-6901.
Yours very truly, Y
f S. R. Zimn rman Manager Nuclear Licensing Section BAT /lah (5168 BAT) 8704200418 070414 Enclosures PDR ADOCK 05000324 Y
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Dr. J. Nelson Grace (NRC-Ril)
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Mr. W.11. Rutand (NRC-BNP)
Mr. E. Sylvester (NRC) f 411 Fayettmite Street
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v Standby Liquid Control System The Brunswick Plant will modify the existing Standby Liquid Control System (SLCS) and update the Technical Specifications to support the two-pump operation alternative for compliance with the ATWS rule,10CFR50.62.
The existing SLCS at BSEP was reviewed to determine the impact of two-pump operation on system design. The existing design was found to be acceptable with one change; the pump relief valve setpoint must be changed from 1400 psig to 1450 psig to accommodate the higher line losses associated with increased flow while still maintaining an adequate working margin. Technical Specificaticn Surveillance Requirement 4.1.5.3.c will be changed to reflect this.
Both units at the Brunswick Plant are of the GE BWR/4 design, with 218-inch inside diameter vessels. The 86-gallon per minute injection rate requirement and 13-weight percent sodium pentaborate solutica' concentration requirement specified by paragraph (cX4) of 10CFR50.62 are values used in NEDE-24222, " Assessment of BWR Mitigation of ATWS, Volumes I and II," December 1979, for BWR/5 and BV!R/6 plants with a 251-inch inside diameter vessel. NEDE-24222 recognized that different values would be equivalent for smaller plants and states that a 66-gpm control liquid injection rate in a 218-inch inside diameter vessel is equivalent to the 86-gpm injection rate for a 251-inch vessel. This position was accepted by the NRC in a Safety Evaluation Report dated October 21,1986.
Each of the two positive displacement pumps in the existing SLCS is designed to deliver 43 gpm to the reactor vessel. Technical Specification Surveillance Requirement 4.1.5.c.2 states that each pump discharge must be demonstrated to be at least 41.2 gpm at a pressure greater than or equal to 1190 psig. This is well above the requirement for a total flow of 66 gpm from the two pumps for this particular reactor design.
The concentration of sodium pentaborate in the SLC tank is specified on Technical Specification Figure 3.1.5-1. Currently, the concentration, as shown on the figure, is allowed to drop below 10 percent sodium pentaborate at high tank volumes. A TS change will be submitted to make the lower limit on sodium pentaborate thirteen percent by weight.
The changes outlined above will ensure that BSEP will meet the requirements of 10CFR50.62. Appropriate Technical Specification changes for Uni I were submitted on December 2,1986 (NLS-86-416). The modifications on Unit I will be complete and this system operable prior to the end of the current refueling outage. The modifications on Unit 2 are scheited for the next refueling outage, currently scheduled to commence January 2,1983. Technical Specification changes fcr Unit 2 will be submitted prior to the start of that outage.
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Alternate Rod Insertion 1
ARI System Function Times NRC Requirement: Rod injection motion will begin within 15 seconds and be completed within 25 seconds from ARI initiation.
CP&L Response: Sufficient blowdown capability will be installed in the existing scram pilot air header to provide depressurization of the air header which will ensure that rod injection motion will begin within 15 seconds and be completed within 25 seconds of ARI initiation.
2.
Safety-Related Requirements NRC Requirements: (a) Class IE isolators are used to interface with safety-related systems, (b) Class IE isolators are powered from a Class IE source, and (c) Isolator qualification documents are available for staff audit.
CP&L Response: The system will be designed to minimize interface with safety-related systems. Where interfaces exist, Class IE isolators powered from Class IE sources will be used and qualification docur ents will be required.
3.
Redundancy NRC Requirement: The ARI system performs a function redundant to the backup scram system.
CP&L Response: The ARI system will perform a function redundant to the existing backup scram system. The ARI valves will be DC powered, energized-to-function in the same way as the backup scram valves. Actuation signals to the ARI valves will be independent of those actuating the backup scram system.
4.
Diversity from Existing RTS NRC Requirements: (a) ARI system is energize-to-function,(b) ARI system uses DC-powered valves, and (c) Instrument channel comp)onents (excluding sensors but including all signal conditioning and isolation devices are diverse from the existing RTS components.
CP&L Response: The ARI valves will be energized-to-function and DC powered and will not require off-site power to function. The instrument channel components, including the sensors, signal conditioning and isolating devices, will be diverse from the existing RTS components.
5.
ElectricalIndependence from the Existing RTS NRC Requirements: (a) ARI actuation logic is separate from RTS logic and (b) ARI circuits are isolated from safety-related circuits.
CP&L Response: The ARI logic will be separate from the RTS logic and will perform an ATWS/ARI function only. The ARI circuits will be isolated from safety-related circuits thereby ensuring that ARI circuits will not degrade a safety-related circuit or prevent it from performing its design function.
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Physical Separation from Existing RTS NRC Requirement: The ARI system is physically separated from RTS.
CP&L Response: The ARI equipment and circuitry will be located in such a way to ensure that physical separation from RTS components is maintained.
7.
Environmental Qualification NRC Requirement: The ARI equipment is qualified to conditions during an ATWS event up to the time the ARI function is completed.
CP&L Response: The equipment used for the ARI will be qualified for the environmental conditions expected during the anticipated transient during which they are required to function.
8.
Quality Assurance NRC Requirement: Comply with Generic Letter 85-06.
CP&L Response: The ARI system components will comply with Generic Letter 85-06.
9.
Safety-Related Power Supply NRC Requirements: (a) ARI system power independent from RTS and (b) ARI system can perform its function during any loss of off-site power event.
CP&L Response: The ARI will be designed such that the ARI system power supply is independent of any RTS components. DC power will be used to ensure it can perform its function during any loss of off-site power event.
- 10. Testability at Power NRC Requirements: (a) ARI be testable at power and (b) bypass features conform to bypass criteria used in RTS.
CP&L Response: The ARI will be testable at power up to, but not including, the final actuating device (i.e., the ARI valves). A Control Room readout will be provided when a trip channel is placed in the test position. The test switch will be used as a maintenance bypass. The test position annunciates in the Control Room and allows work on one train (two instruments).
- 11. Inadvertent Actuation NRC Requirements: (a) ARI actuation setpoints will not challenge scram setpoints and (b) coincident logic is utilized in ARI design.
CP&L Response: The ARI actuation setpoints will not challenge scram setpoints.
The reactor pressure setpoints will be 1100 psig, which is higher than the normal reactor high-pressure setpoint. The reactor low-level setpoint will be +118 inches decreasing, which is lower than the normal reactor low-level setpoint for a scram.
Coincident logic will be used in the ARI design. BSEP will modify its existing recirculation pump trip logic to enhance its reliability with the ARI design. The (5168 BAT /l ah )
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logic will be a two-out-of-two coincident and will consist of two trip systems. Each trip system will have two level and two pressure channels. By modifying its existing logic, spurious trips from single sensor failure will be eliminated. The logic design will incorporate portions of the Monticello and Modified Hatch designs.
- 12. Manual Initiation NRC Requirement: Manual initiation capability is provided.
CP&L Response: Manual initiation will be provided for ARI through two series-mounted control switches. The series switches will prevent inadvertent manual ARI initiation from operator action.
- 13. Information Readout NRC Requirement: Information read'out is provided in the Main Control Room.
CP&L Response: Information readout of ARI initiation will be provided upon manual or automatic ARI initiation. Control Room annunciators and computer inputs / readouts will be incorporated into the ARI design to provide ARI actuation indication.
- 14. Completion of Protective Action NRC Requirement: The protection action be completed once it is initiated.
CP&L Response: The ARI system will be designed so that once initiated, the protective action will go to completion. Either the automatic or manual actuation signals " seal-in" to assure that all control rods have time to fully insert.
Reset of the ARI function will be automatically prohibited for the duration of the seal-in time. The ARI function can be manually reset by the operator af ter completion of the seal-in time if the automatic signals have cleared. No automatic return to normal operation is provided.
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To Serial No. NLS-87-07't Recirculation Pump Trip i
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Recirculation Pump Trip BSEP will modify its original BWR/4 RPT design to use a redundant two-out-of-two logic arrangement for ARl/RPT initiation. Each individual level and pressure instrument can be tested during plant operation without initiating the ARl/RPT systems since two level or two pressure signals must be present in one channel to complete the initiating logic (see Figure 1). The above logic design has incorporated a combination of designs similar to those presented in the Licensing Topical Report, NEDE-31096-P, specifically the Monticello and Modified Hatch designs.
This combination of designs ensures that plant reliability is not reduced when using the RPT logic for the new ARI system. It will also eliminate recirculation pump trips caused by spurious operation of a single process sensor. The original BSEP RPT logic is represented by Figure 2. BSEP will also replace the existing pressure switches presently used for RPT with transmitters and analog trip units. This change will enhance system reliability, provide ease of testing while at power, and modify logic so that one logic train trip will trip both recirculation pumps. Technical Specifications will be revised to reflect component changes being made.
The BSEP RPT design employs a single trip coil in each recirculation system motor generator set drive motor breaker. Carolina Power & Light Company believes that the requirement for redundant trip coils is inconsistent with the ATWS rule (10CFR50.62).
The Statement of Considerations entitled " Considerations Regarding System and Equipment Specifications" clearly states that redundancy is not required for either the Diverse Reactor Trip System or for Mitigation Systems, which include RPT. In addition, an analysis was performed of the reliability of the Monticello RPT design as compared to the BSEP RPT design. The Monticello design was found to be no more reliable than the proposed BSEP design. The major difference between the Monticello and BSEP designs is that Monticello trips the recirculation motor generator set field breaker utilizing redundant trip coils, whereas the BSEP design trips the main 4kv breaker and does not use redundant trip coils. The analysis was accomplished by drawing fault trees for
" Failure to Trip" and "inadver tent Trip" for both designs by reviewing accumulated data for the RPT circuit breaker under consideration and by discussing the designs with experienced plant personnel.
The following assumptions were made for the fault trees:
- 1. The systems have identical components, except as shown in Figures 3 and 4.
- 2. ATWS causes either a High-Pressure Trip or Low-Level Trip, but not both.
- 3. Level and pressure instruments are solid state analog with low failure rates.
- 4. Common cause failures would affect the " failure to trip" trees equally; therefore, these contributors to the overall unavailability are not included.
The fault trees are shown in Enclosures 5 through 8. Failure rate data from WASH 1400 as presented in the Shoreham PRA was used to calculate the component unavailabilities for this analysis as shown in Tables 2 and 3. Using this data, the circuit breaker failures dominate and there is no difference in unavailability between the designs as shown in Table 1.
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~ In addition, the trip coil of the BSEP RPT circuit breaker has an indicator light connected in series with it to indicate whether the breaker is "open" or " closed".
Therefore, the light is continuously monitoring the status of the trip coil and would effectively annunciate an open circuit within the trip coil.
Also, the Nuclear Plant Reliability Data System (NPRDS) was interrogated for failures of the Model 5HK-350 circuit breaker that is used as the BSEP RPT breaker. Since January 1983,21 failures were found at 9 plants. The NPRDS list did not show any trip coil failures contributing to the 21 listed failures.
Therefore, it was concluded that the Monticello design would be no more reliable than the BSEP design.
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TABLEI FAULT TREE DATA Tree: Monticello - Failure to' Trip (Tree Name: RPT)
Source Breaker X Trip Coil X Instrumentation X Top Event ' X VVash 1400 -
1.2E-3 1.6E-6 1.6E-3
- 2.4E-3 Tree: BSEP - Failure to Trip (Tree Name: RPT2)
Source BreakerX Trip Coil X Instrumentation X Top Event X VVash 1400-1.2E-3 1.6E-6 1.6E-5 2.4E-3 Tree: Monticello - Inadvertent Trip (Trip Name: RPTIT)
Source BreakerX Trip Coil X Instrumentation X Top Event X VVash 1400 6.3E-3 1.3E-4 1.3E-3 1.3E-2 Tree: BSEP - Inadvertent Trip (Tree Name:. RIT)
Source BreakerX Trip Coil X Instrumentation X Top Event X Wash 1400 6.3E-3 1.3E-4 1.3E-3 1.3E-2 (5168 BAT /l ah )
TABLE 2 DATA: ' FAILURE OF BREAKER TO OPEN Breaker Failure Data - Fails to Transfer -
Wash 1400 11.25'X 10-3/d Coil Failure Data. Coil Open (i.e., keeps breaker from tripping)
Wash 1400 2.68X10-7/hr Since a light is in series with the coil, discovery time would equal one shift. Assume 12-hour shift.
X = i AT = i (2.68X10-7)(12) = 1.6X10-6 Instrumentation General Wash I'400 2.68X 10-6/hr Since these are or will be analog instruments, discovery time would equal one shift.
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TABLE 3 DATA - INADVERTENT BREAKER TRIP Breaker Failure Data Premature Trip Wash 1400 1.25X 10-6/hr X = AT, where T = Mission' Time = (Capacity Factor)(24)(365.25) = 5,000 X=
(1.25Xl'0-6)(5,000) = 6.3X10-3 Coil Failure Data Shorts to Power Wash 1400 2.68X 10-8/hr X = AT, where T;= Mission Time = (Capacity Factor)(24)(365.25) 5,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> avg CRITICAL X = (2.68X10-8)(5000) = 1.3X10-4 Instrumentation and Control Solid State Device Fails Shorted Wash 1400 7.68X 10-7 X = AT, where T = Mission Time 5,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> avg CRITICAL X = (2.68X 10-7) (5,000) = 1.3X 10-3 (51688AT/bmc)
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INAlvERTENT TRIP OF EITHER FAULT TRE: MONTICELLO INDUERTDR TRIP Fl!I 2PE: RPTIT DATE:
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I L O LEVEL LON LEW L LON LEVEL LON LEVEL HIGH PRESS.
HIGH PRESS.
Irli. A INST. C INST. B INST. D INST. A INST. C SPURIOUS SPURIOUS SPURIOUS SPURIOUS SPURIOUS SPURIOUS ACTUA110N ACTUATION ETUATION ACTUATION ACTUATION ACTUATION RRS-ACT4A-LLA RRS-ACT4A-LLC RRS-ACT44-LLB RRS-ACT4A-LLD RRS-Ati4A-HPA RRS-ACT4A-HPC 0
0 0
0 0
0 1
s
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+-
w-w w--r e-ae vm v
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TRANSFER TD PAGE 2 HH E
F4 ULT 1 RED MONTICELLO 1mDU0tTENT TRIP Flti mE: RPTIT -
CHAN I FAILS DATC 12-19-86 KV. 3 RPili-69 l
.r I
1 HIGH PRESS.
HIGH PRESS.
INST. I INST. D SPURIOUS SPURIOUS ETUATION ETUATION r
RRS-ET-44-NPB RRS-E HPD 0
^
e 1
5
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l
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I i
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i To Serial No. NLS-87-074 Fault Tree: BSEP Inadvertent Trip (5168 BAT /lah)
J
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PAGE 1 DF 3
INABVERTENT TRIP OF EITHER FAULT TRE: BSEP JMDUERTENT TRIP IREAKER FitI MlO RIT DATE:
12-19-86 REU. 1 Rii-61 1 t
BREAKER A BREAKER 8 TRIPS TRIPS RIT-62 I RIT-63 i PAGE 2 BREAKER I TRIP COIL 01 PtRIOUS TRIP SPURIOUS TRIP SPURIOUS
$10 MAL ACTUATION II RRS-BKR{0-BKRB RRS-TCL-4A-TCBI Ali-64 1 0
0
&2
(
(
.T P A GIE 2 0F 3
FAULT TRE: BSEP l*l)UERTENT TRIP FILE MtC PIT TRANSFER TO PAGE I DATE:
12-18-86 REU. I NEAKER A TRIPS AIT-62 I i
i j
MEAKER A TRIP COIL Al SPWIOUS TRIP W W10US TRIP SPURIOUS SIGNAL ACTUATION RRS-tKR 0-BKRA RRS-TCL TCA1 RIT-64 I i
i i
i LON LEVEL LON LEVEL HIGH PRESS.
HIGH PRESS.
CHAN. A CHAN. 8 CHAN. A CHAN. B SPURIOUS TRIP SPURIOUS SPURIOUS SPJIIOUS
(
ACTUATION ACTUATION ACTUAil0N RIT-65 i RIT-66 i RIT-67 i RIT-68 I I
I i
f I
I LD LEVEL LON LEVEL LON LEVEL LON LEVEL HIGH PRESS.
HIGH PRESS.
IST. A INST. C INST. I INST. 8 INST. A INST. C SPURIOUS SPURIOUS SPURIOUS SPUR 100$
SPURIOUS SPURIOUS ACTUATION ACTUATION ACTUATION ACTUATION ACTUATION ACTUATION RRS-ACT
-LLA RRS-ACT
-LLC RRS-ACT -LLB RRS-ACT -LLD RRS-ACT
-HPA RRS-ACT -HPC l
l
- l. -.T PAGE 3 0F 3
TR MSFER TO PAGE 2 HIGH PRESS.
FnULT TMD SKP IMDUERTDE 1 RIP cum, 3 FILE DWE: RIT i
SPal005 MTC 12-10-M KV.1 i
l ETMilm RIT-68 I l
l HIGH PRESS.
HIGH PRESS.
INST. 3 IllST. O SPURIOUS SPWIOUS ACillATIN ACTUATION RRS-ACT$-WB 0
RRS-ACT-44-WD t
l 4
I l
[
i l
1
. - - - _ - - - - - _ - _