ML20118A689
| ML20118A689 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 09/14/1992 |
| From: | Starkey R CAROLINA POWER & LIGHT CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20118A690 | List: |
| References | |
| NLS-92-237, NUDOCS 9209240453 | |
| Download: ML20118A689 (31) | |
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SEP 141992 fi e $1 Mar y, J4 vier Pecentwo N e w f.e % en De p eiem SERIAL: NLS :,,2 237 10 CFR 50.90 TSC 02TSB01 United States Nuclear Regulatory Commission ATTENTION: Docurc'.ent Control 9esk Washington, DC 20555 BRUNSWICK STEAM ELECTRIC PLANT, UNir NOS 1 AND 2 DOCKET NOS. 50 325 & Sf' 124/ LICENSE NOS. DPR 71 & DPR-62 REQUEST FOR LICENSE AMENDMENT STEAM LEAK DETECTION INSTRUMENTATION NUMAC UPGRADE Gentlemen:
In accordance with the Code of Federal Regulations, Title 10, Parts 50.90 and 2.101, Carolina Power & Light Company hereby requests a revision to the Technical Specifications for the 9tunswick Steam Electric Plant (BSEPL Units 1 and 2.
The proposed amendment revises the Technical Specifications to reflect the replacement of existing Riley, GEMAC and Fenwal steam leak detection equipment with General Electric NUMAC leak detec'.on equipment. The proposed amonc ment also revises surveillance requirementa for steam leak detection instrumentation associated with the reactor water cleanup system, the high pressure coolant injection system, and the reactor core isolation cooling system. The specific changes include:
(1) delete the CHANNEL CHECK surveillance test for the reactor water cleanup system isolation high differential flow function, (2) extend and standardize the CHANNEL FUNCTIONAL TEST and CHANNEL CAllBRATION surveillance frequencies for the reactor wi er cleanup system isolation differential flow function and the reactor water cleanup, high pressure coolant injection, and reactor core isolation ::ooling system isolation ambient and differential temperature functions,
- 3) increase the reactor water cleanup system isolation diffuential flow time delay TRIP SETPOINT and ALLOWABLE VALUE from "less than or equal to 45 seconds" to "less than or equal to 30 minu es,*
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e Document Control Desk NLS-92 237 / Page 2 (4) increase the reactor water cleanup system isolation differential flow TRIP SETPOINT and ALLOWABLE VALUE from "less than or equal to 53 g.-1/ min" to "less thsn or equal to 73 gal / min,"
(b) delete the instrument response time requirement for the high pressure coolant injection system isolation steam line tunnel temperature - high function, and (6) delete the instrument response time requirement for the reactor water cleanup system isolation area temperature high and area ventilation differential temperature high fune.tions,
- 17) delete the instrument response time requiremont for the reactor water cleanup system isolation differential flow high function, and (8) revise the description of the reactor water cleanup isolation differeritial flow delay trip.
function to reflect elimination of the time delay relays por the new system configuration,-
(9) add a new reactor water cleanup system isolation area temperature function for piping outside of the reactor water cleanup room,
- provides a detailed description of the proposed changes and the basis for the changes. details, in accordance with 10 CFR 50.91(a), the basis for the Company's determination that the proposed chanke do not involve a significant hazards consideration. provides an environmental evaluation which demonsuates that the poposed amendment meets the eligibility criteria for categorical exclusion set forth in N CFR 51.22(c)(9).
Therefore, pursuant to 10 CFR 51.22(b), no environmental assessment needs to be prepared in connection with issuance of the amendment. provides page change instructions for incorporating the proposed revisions. provides the proposed Technical Specification pages for Unit 1.
t provides the proposed Technical Specification pages for Unit 2.
Carohna Power & Lloht Company is providing,in accordancs with 10 CFR 50.91(b), the State of North Carolina with a copy of the propossd license amendment.
The Brunswick Unit 1 steam leak detection system modification is scheduled for installation during Refueling Outage No.10, which is currently planned to begin March 4,1993. The corresponding Unit 2 mcdification is scheduled for installation during Brunswick Unit 2 Refuelinn Outage No.- 11, which le currently planned to begin Snptember 9,1993. - CP&L requests that the proposed amendment for Unit 1 be issued by March 1,1993 and for Unit 2 by September 1,1993. In order to allow for procedure revisions and orderly incorpora..J into copies of the Technical 1
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Document Contsol Desk NLS 92 237 / Page 3 Specifications, CP&L requests that the proposed amendments, once approved by th NRC, be issued with an affective date to de no later than 90 days from the issuance of the amendment.
Please refer any questions regarding this submittal to Mr. M. R. Oates at (919) 540 6033.
Yours very truly, M
R. B. Starkey, Jr.
WHM/wrm (numactsc.wpf)
Enclosures:
1.
Basis for Change Request 2.
10 CFR 50.92 Evaluation 3.
Environmental Considerations 4.
Page Change Instructions 5.
Technical Specification Pages 'Init 1 6.
Technical Specification Pages Unit 2 R. D. Starkey, Jr., having been first duly sworn, did depose and say that the information contained herein is true and correct to the best of his information, knowledge and belief; and the sources of his information are officers, employees, contractors, and agents of Carolina Power & Light Company.
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l ENCLOSURE 1 DRUNSWICK STEAM ELECTRIC PLAN 1, UNITS 1 AND 2 NRC DOCKET NOS. 50 325 & 50 324 OPERATING LICENSE NOS. DPR 71 & DPR 62 l
REOUEST FOR LICENSE AMENDMENT STEAM LEAK DETECTION INSTRUMENTATION NUMAC UPGRADE DASIS_E0 LLC 1WiGMQlRS.I DACKG110WD:
The leak detection systems covered by this licent,e amendment request include ambient and differential temperature monitoring for the reactor water cleanup, high pressure coolant injection and cactor core isolation cooling systems and differential flow monitoeing for the reactor water cleanup system.
The temperature monitoring portions of the reactor water cleanup, high pressure coolant injection, and reactor core isolation cooling system leak detection instrumentation 2 1volved in this request utilire ambient and differential air temperature measurements. Two types of temperature instrument channels are presently used. The first type of channel consists of local thermocouples that provide inputs to temperature switch modules in the control room, which in turn perform indication, alarm, and isolation f mctions. The temperature switch modules utilized in theso channels havo historically experienced a high drift rate, have been prone to spurious alarms and trips, and have been difficult to maintain.
The second type of instrument channel consists of local temperature switches whose contacts provide input to control room relay logic. These local switches have the disadvantage of not providing indication for the monitored areas and have been the cause of signihcant personnel radiation exposure due to the time spent inside the reactor building performing monthly testing of these switches.
The reactor water cleanup system differential flow leak detection involves comparison of the reactor water cleanup system inlet flow to the sum of the two output flows (one is return to the feedwater system; the other is reject flow to the train condenser or the radwarto system). The existing reactor water cleanup system differential flow leak detection instrumentation has several technical limitations. The function consists of a complex three element flow loop with analog computing modules and density compensation is not provided to account for varying process water temperatus. As a result, this instrument channelit subject to significant signal drift and inaccuracy. The loop inaccuracy is particularly evident during reactor water cleanup system transient conditions, such as system fill and start up. These limitations have led to unwarranted process isolation signals that unnecessarily challenge the reactor water c!eanup system containment isolation function, degrade operator confidence in the control system, creato additional work load for operations personnet in dealing with the control system, and result in an excessive numler of licensee event reports documenting those unnecessary system isolations.
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e Carolina Power & Light Company plans to upgrade tne leak detection instrumentation described above by instsilation of a General Electric microprocessor based NUMAC system. The NUMAC system will process signals from the ambient and differential temperature sensors of the reactor wster cleansp, high pressure coolant injection, and reactor core isolation cooling systems and from the flow sensors in the reactor water cleanup system. The NUMAC system will reduce instrument drift, improve Instrument accuracy, simplify maintenance activities, enhance the i
operator / equipment interface, and prowde a means of density compensation for the reactor water cleanup system differential flow measurement. In conjunction with NUMAC installation, the local ambient temperature switches will be replaced with thermocouples connected to NUMAC channels.
As a result of the significant system perfurmance and reliab!Iity improvements that will result from this installation, the following changes to the BSEP Technical Specifications are proposed:
111M.1:
C11BBfRL REQUIREMENI:
1 Technical Specification Table 4.3.21 (Isolation Actuation Instrumentation Surveillance Requirements), item 3.a (Reactor Water Cleanup System isolation, Differential Flow High) currently specifies the performance of a CHANNEL CHECK on a D (Daily) frequency.
)
fjlOfQSED CHANGE:
Revise the surveillance frequency for the reactor water cleanup system, differential flow high containment isolation actuation instrument CHANNEL CHECK from D (Daily) to NA (Not Applicable)
DASJS:
-1 The NUMAC system performs a continuous self test and alerts the operator via annunciation when a problem is detected. The following diagnostic and self test features are provided:
1.
Continuous monitoring of each flow and each density compensation input signal for out-of-bounds values.
2.
Continuous monitoring of the two internal powe: supplies (NUMAC remains functional with only one internal power supply).
i 3.
Continuous monitoring of the external power input.
4.
A self check of each channel to confirm functionality at least once per 30 minutes.
I 5.
Continuous monitoring to assure that the system is not left in an inoperable cuadition (card out of file, status switch left in the inop mode).
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i Based on installation of the improved NUMAC system and the NUMAC sistom's self diagnostic features, the D tDaily) surveillance frequency for the reactor water cleanup system differential flow -
high containmont isolation actuation Instrument CHANNEL CHECK is no longer noodod.
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CWiliULLilLQUlBEMuiD Technical Specificatlon Table 4.3.21 (Isolation Actuation instrumentation Surveillance Requiroments) currently specifies the following CHANNEL FUNCTIONAL TEST and CHANNEL CAllBRATION frequencies for trip functions affected by the NUMAC installation modification:
Item 3.a (Reactor Weter Cleanup System isolation, a Flow High) specifies a CHANNEL FUNCTIONAL TEST on a M (Monthly) frequency.
Item 3.b (Reactor Water Cleanup System isolation, Area Temperature High) specifies a CHANNEL FUNCTIONAL TEST on a M frequency, item 3.c (Reactor Water Cleanup System Isolation, Area Ventilation a Tempuature High) specifies a CHANNEL FUNCTIONAL TEST on a M frequency, item 3.f (Reactor Water Cleanup System isolation, Differential Flow High Time Delay Relay) specifies a CHANNEL FUNCTIONAL TEST on a M frequency, item 4.a.4 (thgh Pressure Coolant injection System isolation, HpCl Steam Line Tunnel Temperature High) specifies a CHANNEL FUNCTIONAL TES1 on a M frequency and a CHANNEL CAllBRATION on a 0 (Quarterly) frequency, item 4.a.7 (High Pressure Coolant injection System isolation, HPCI Steam Line Ambient Temperature - High) specifies a CHANNEL FUNCTIONAL TEST on a M frequency.
Item 4.a.8 (High Pressure Coolant injection Systnm trolation, HPCI Steam Line' Area a Temport.ture High) specifies a CHANNEL FUNCTIONAL TEST on a M frequency.
Jtem 4.a.9 (High Pressure Coolant injection System isolation, HPCI Equipment Area Temperature High) specifies a CHANNEL FUNCTIONAL 1EST on a M frequency and a CHANNP. CAllDRATION on a O frequency, item 4.b.4 (Reactor Core Isolation Cooling System Isolation, RCIC Steam Line Tunnel Temperature Hight speciCes a CHANNEL FUNCTIONAL TEST on a M frequency.
Ito.n 4.b.7 (Reactor Core Isolation Cooling System Isolation, RCIC Steam Line Ambient Temperature - High) specifies a CHANNEL FUNCTIONAL TEST on a M frequency, Item 4.b.8 (Reactor Core isolation Cooling System isolation, RCIC Steam Line Area a Temperature High) specifies a CHANNEL FUNCT'ONAL TEST on a M frequency.
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1 It vn 4.b.9 (Reactor Core Isolation Cooling System isolation, RCIC Equipment Roorn Atabient Temperature Hight specifies a CHANNEL FUNCTIONAL TEST on a M frequency and a CHANNEL CALIBRATION on a O frequency.
Item 4.b.10 (Reactor Core isolation Cooling System isolation, RCIC Equipment Room a Temperature High) specifies a CHANNEL FL.NCTIONAL TEST on a M frequency and a CHANNEL CAllDRAllON on a G frequency.
Item 4.b.11 (Reactor Core laulation Cooling System isolation, RCIC Steam Line Tunnel Temperature High Time Delay Relay) specifies a CHANNEL FUNCTIONAL TEST on a M frequency.
PROPOSED CHANGE:
Revise Technical Specification Table 4.3.21 (Isolation Actuation Instrumentation Surveillance Requiremeists) such that the required CHANNEL FUNCTIONAL TEST and CHANNEL CAllBRATION frequencies for the trip functions affected by the NUMAC installation modification are extended as follows:
ltem 3.a (Reactor Water Cleanup System Isolation, 6 Flow High): revise the CHANNEL FUNCTIONAL TEST from M (Monthly) to SA (Semi Annual) frequency.
Item 3.b (Reactor Water Cleanup System Isotation, Area Ternperature High): revise the CHANNEL FUNCTIONAL TEST from M to SA frequency, item 3.c (Reactor Water Cleanup System Isolation, Area Ventilation a Temperature High):
revise the CHs NNEL FUNCTIONAL TEST from M to SA frequency.
Item 3.f (Reactor Water Cleanup System Isolation, Differential Flow High Time Delay Relay): revise the CHANNEL FUNCTIONAL TEST from M to SA frequency.
Item 4.a.4 (High Pressure Cnolant injection System Isolation, HPCI Steam Line Tunnel Temperature - High): revise the CHANNEL FUNCTIONAL TEST from M to SA frequency and the CHANNEL CAllBRATION from O (Quarterly) to R (Refuel) frequency.
Item 4.a.7 (H gh Pressure Coolant injecthn System Isolatiori, HPCI Steam Line Ambient.
Temperature High): revise the CHANNEL FUNCTIONAL '.'EST from M to SA frequency.
Item 4.a.8 (High Pressure Coolant injection System isolation, HPCI Stuam Line Area a Temperature - High): revise the CHANNEL FUNCT!ONAL TEST from M to SA frequency, item 4.a.9 (High Pressure Coolant injection System isolation, HPCI Equipmont Area Temperature - High): revise the CHANNEL FUNCTIONAL TEST from M to SA frequency and thei CHANNEL CAllBRATION from Q to R frequency.
item 4.b.4 (Reactor Core isolation Cooling System isolation, RCIC Steam Line Tunnel Temperature - High): revise the CHANNEL FUNCTIONAL TELT from M to SA frequency.
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Item 4.b.7 (Reactor Core isolation Cooling System isolation, RCIC Steam Line Ambient Temperature Hich): revise the CHANNEL FUNCTIONAL TEST from M to SA frequency.
Item 4 b.8 (Reactor Core Isolation Cooling System isolation, RCIC Steam Line Area a Temperature High): revise the CHANNEL FUNCTIONAL TEST from M to SA frequency.
Item 4.b.9 theactor Core isolation Cooling System Isolation, RCIC Equipment Room Ambient Temperature. High): revise the CHANNEL FUNCTIONAL TEST from M to SA frequency and the CHANNEL CAllBRA rl0N from 0 to R frequency, item 4.b.10 (Reactor Core Isolation Cooling System isolation, RCIC Equipment Room a Temperature High): revise the CHANNEL FUNCTIONAL TEST frc.m M to SA frequency and the CHANNEL CAllBRATION from 0 to R frequency.
Item 4.b.11 (Reactor Core Isolation Looling System isolation, RCIC Steam Line Tunnet Temperature High Time Delay R lay): revise the CHANNEL FUNCTIONAL TEST frorn M to SA frequency.
BAS $:
CHANNEL FUNCTIONAL TE.SIA:
The preposed change would increase the CHANNEL FUNCTIONAL TEST surveillance interval from M (Monthly) to SA (Sem! Annual) for the identified temperature and differential flow monitoring Technical Specification Trip Functions. The NUMAC system provides a comprehensive self test feature that is capable of detecting most of the potential failures that the CHANNEL FUNCTIONAL TEST is intended to identify.
As described in item 1 above, the NUMAC system performs a continuous self test and alerts the' operator via annunciation when a problem is detected. The following diagnostic and self test features are provided:
1.
Continuous monitoring of each flow and each density compensation input signal for out of bounds values.
2.
Continuous monitoring of the two internal power supplies (NUMAC remains functional with only one internal power supply).
3.
Continuous monitoring of the external power input, 4.
A self check of each channel to confirm functionality at least once per 30 minutes.
5.
Continuous moritoring to assure that the system is not left ir an inoperable condition (card out-of file, status switch left in the inop model.
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CHANNEL CAllBRATIONs:
The proposed change wouU i~ crease the CHANNEL CAllBRATION surveillance interval from O I
i (Quarterly) to R (Refuelinyl for four of the temperature monitoring Technical Specificat on Trip functions. The NUMAC system features a high degree of stability, thereby permitting longer j
intervals between calibrations without significant instrument drift.
A primary reason for performing a CHANNEL CAllBRATION is to detect and manage instrument drift. For the temperature channels (both ambient and differential), three primary potential sources of instrumcat drift exist: (1) the NUMAC processing unit, (2) the analog to-digital (A to D) converter I
located in the NUMAC system, and (3) the temperature sensors (thermocouples). The NUMAC processing units utilize digital circuitry and are essentially drift free. The NUMAC input analog to-digital converter is subject to nominal drift due to the analog signal processing involved. The external temperature sensing devices are subject to virtually no signal drift. The thermocouples are inherently stable devices with little potential for significant drif t. They can be expected to either operate within manufacturer's specifications or else fait completely. Carolina Power & l.ight calculations utilizing NUMAC specified accuracy and drift values demonstrate that adequate margin exists between the actual field calibration setpoint and the existing, or proposed revised, Teu nical Specification ALLOWABLE VALUES to justify a 22.5 month cahbration interval (18 months plus the 25 percent surveillance !nterval extension allowed by Technical Specification 4.0.2).
ITEM. 3:
CURRENT REQUIREMENT:
Technical Specification Table 3.3.2 2 (Isolation Actuation Instrumentation Setpoints), item 3.f t
(Reactor Water Cleanup System isolation, Differential Flow High - Time Delay Relay) currently specifies a TRIP SETPOINT and ALLOWABLE VALUE of less than or equal to 45 seconds.
PROPOSED CHANGE:
Revise the Technical Specification Table 3.3.2 2 (Isolation Actuation Instrumentation Setpoints),
item 3.f ' Reactor Water Cleanup System isolation, Differential Flow - High Time Delay Relay) to increase the TRIP SETPOINT and ALLOWABLE VALUE from "less than or equal to 45 seconds" to l
"less than or equal to 30 minutes."
DASIS:
~
The existing setpoint of less than or equal to 45 seconds has led to many unwarranted system isolations of the reactor water cleanup *ystem and has been a chronic problem for Operations personnel during initial system fill, pressurization and startup. LERs 192-01t. b91-010i 2 90-011, 2 88-010 and 2 88 003 provide documentation of specific cases where incorrect leakage indications resulted in unnecessary system isolations, i
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4 The reactor water cleanup high differential flow afarm/ trip is based on the difference between the input flow from thu reactor coolant system and the sum of the two output flows (one is return to the feedwater system; the other is reject flow to the mein condenser or the radwaste system). The difference between the input and output flows is assumed to be leaksge. The proposed NUMAC system willincorporate the enhanced feature of process flow density compensation based on process temperature. This will provide the operator more accurate reictor water cleanup system leak rate information. See Figures 1 and 2 for a basic diagram of the RWCU Differential Flow Instrument / Control System.
The sole design basis for the reactor water cleanup system high differential flow isolation function is to assure compliance with 10CFR100 and iOCFR20. The high differentita flow isolation function is not intended for protection of reactor pressure vessel water level or for limiting the reactor building environment for equipment qualification purposes. Current reactor water cleanup system high energy line break scenarios rely on ambient temperature detection as the isolation initiation signal.
General Electric Company's proprietary report GE NE 77014 0592 evaluates the consequences of a 300 gal / min reactor water cleanup system cold leak remaining unisolated for 30 minutes. That analysis utilized conservative source terms and assumptions. The resultant control mom, site boundary and low population zone doses are within the limits proscribed in Standard Review Plan 6.4 and 10CFR100 and are less than dose consequences previously ev,luated for other BNP events. These documents provide the basis for establishment of 300 gal / min as CP&L's engineered ANALYTICAL LIMIT for use in setpoint calculations related to this differential flow function and for the establishment of 30 minutes as the At.LOWABLE VALUE/ TRIP SETPOINT LIMIT for this differendal flow lsolation time delay function.
Plant sump monitoring Instrumentation, room flood alarms and plant leakage response procedures preclude the possibility that adverse room flooding conditions could result from the increased time delay on th<a automatic RWCU system isolation.
As a comparkon, Plant Hatch has a two hour override on the RWCU fiferential flow high trip signal to assist the operators during system start-up and during system transients. The FitzPatrick Plant has no RWCU differential flow high trip.
jlE(4:
CUHRENT REQUIREMENT:
Technical Specification Table 3.3.2 2 (Isolation Actuation Instrumentation Setpoints), item 3.a (Reactor Water Cleanup System isolation, Differential Flow High) currently specifies a TRIP SETPOINT and ALLOWABLE VALUE of less than or equal to 53 gallons per minute, E17
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PROPOSED CH ANGE:
Revise the TRIP SETPOINT and ALLOWABLE VALUE for the reactor water cleanup system iso:stion, differential flow high containment isolation function from "less than or equal to E3 gal / min' to "less than or equal to 73 gal / min.'
BASIS:
l i
CP&L'n calculation ORWCU 0010, prepared in accordance with the methodology defined la ISA.
S67.04, "Setpoints for Nuclear Safety-Related Instrumentation," defines the magnitude of the.
uncertainty acsociated with the reactor water cleanup system isolation diff,erential flow trip function setpoint. The uncertainties ara characterized as either " measurable" or "unmeasurable."
The measurable uncertainties are those attributable to effects that may be present during surveillance testing, The unmeasurable uncertainties are those related to effects that will not be present during survcillance testing (e.g., flow orifice effects, seismic events, post accident environmental conditions).
The requested increase in the Technical Specification ALLOWABLE VALUE is intended to establish a difference between the actual field calibration setpoint and the new ALLOWABLE VALUE that is large enough to bound the sum of the measurable uncertainties present during surveillance testing-onditions and a nominal additional 'LER avoidance" margin. The CP&L calculation demonstrates
..iat satisfaction of the proposed ALLOWABLE VALUE during surveillance testing will assure that the 300 gal / min ANALYT! CAL LIMIT will not be exceeded during any postulated plant events.
ITEM O' s
CURRENT REQUIREMENT:
Technical Specification Table 3.3.2 3 (Isolation System Instrumentatien Response Times)
Item 4,a.4 (High Pressure Coolant injection System isolation, HPCI Steam Line lunnel Temperature High) currently specify a response time of s 13 seconds.
PROPOSJ;D CHANQs:
Revise the instrument response time requirement for the high pressure coolant injection system isolation steam line tunnel temperature high trip function from s 13 seconds to NA (Not applicable).
BASIS:
The response time requirements for containment isolation instrumentation are based on the need to -
limit damage to safety-related equipment or to mitigate the consequences of an accident or.
equipment failure. The sensing devices for the high pressure coolant injection system isolation steam line tunnel temperature - high trip function channel we thermocouples, Thermocouples ara -
inherently stable devices with little potential for shift in response time. They can be expected to either respond within manufacturer's time constant specifications or else fail completely, A given l
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break in the high pressure coolant injection system steam supply hne will always re, ult in a predictable respr.nse from the sensing thermocouples, regardless of how long the thermocouple has l
been installed (i.e., the response time of a thermocouple is constanth Therefore, the response time of a thermocouple need not be routinely measured.
Isolation of the high pressure coolant injection system is required in the event of a process high l
energy line bre8S (HELB)'and has been analyzed in the CP&L Paport No. 9527-058-S MS 001,
" Reactor Building Environmental Report". The isolation trip signalinitiators considered in that analysis included both temperature channels and high flow sensors. The high flow trips ware used in the analysis to mitigate the large (a 300% flow) break and the temperature channels were used to mitigate sma:ler breaks. The high temperature isolation trip function is capable of providing a timely response to either size break.
The worst case 10 inch double ended guillotine break was analyzed. The rapid ambient ternporature increase fiorn this break creates an immediate response from the temperature sensor thermocouples, in addition, smaller break.s which cause a slower increase in ambient temperatures, also provide timely isolation signals. Smaller breaks yield a lower mass energy release and thus reduced environmental consequences. The temperature scnsors do not react as quickly to the more slowly increasing temperatures from these smaller breaks, however, this is not of concern since the environmental effects of a small break are not as severe as for large breaks. Thus, the response times of those sensors need not be as quick. Therefore, the thermocouple response time is not critical in assuring high energy line break mitigation (i.e., valve isolation) and response time testing is not necessary, i
it should be noted that other boiling water reactors similar to the Brunswick Plant do not have response time testing requirements in their Technical Specifications for these same Trip Functions.
ITEM 0-
.CUBRENT REQUIREMENT:
- t Technical Specification Table 3.3.2-3 (Isolation System Instrumentation Response Times) Item 3.b (Reactor Water Cleanup System Isolation, Area Ternperature - High) and item 3.c (Reactor Water Cleanup System isolation, Area Ventilation Differential Temperature High) currently specify a response time of "s 13 seconds."
n EflOPOSED CHANGE:
Revise the instrument response time requirement for the reactor water cleanup system area temperature high and area tentilation differential temperature high isolation functions from s 13 seconds to N/A (Not Applicable).
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The response time requirements for conta!nment isolation instrumentation are based on the need to limit damage to safety-rrlated equipment or to mitigate the consequences of an accident or equipment failure. The sensing devices for the reactor water cleanup system isolation area temperature high and ventilation area differential temperature high instruments are thermocouples.
Thermocouples are inherently stabla devices with little potential for shift in response time. They can be expected to either respond within manufacturers time constant specifications or else ail r
completely. A given break in the reactor water cleanup system piping wi!l always result in a predictable ruspor.se from the sensing thermocouples, regardless of how long the inermocouple has been installed (i.e., the response time of a thermocouple is constant). Therefore, the respons6 time-of a thermocouple need not be routinely measured.
Isolation of the reactor water cleanup system is required in the event of a process nigh energy line break (HELB) and has been analyzed in the CP&L Report No. 9527 058 S MS-001, fReactor '
Building Environmental Report," The isolation trip signallnitiators considered in that anansis for the RWCU line break were the temperature channn!s The high temperature isolation trip function is capable of providing a timely response to either a double ended guillotine break or smaller breaks.
s it should be noted that several other boil;ng water reactors similar to the Bru isw'ck Plant do not have response time testing requirements in their Technical Specifications for these same Trip Functions.
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CUhBENTEDUIPEMENT:
Technical Specification Table 3.3.2 3 (Isolation System Instrumentation Response Times) Item 3.e (Peactor Water Cleanup System isolation, 4 Flow High) currently specifies a response time of s 45 seconds.
PROPOSED CHANGE:
P.evise the instrument response time ree.uirement for the reactor water cleanup system isolation, a flow high trip function from s 45 seconds to NA (Not applicable).
DAS13:
The intent of response time requirements for containment isolation instrumentation is to monitor for) subtle performance changes in trip channels whose prompt retpora,e is taken credit for in event l
detection and mitigation. The analysis performed to justify the pmposed increases in the -
differential flow time delay TRIP SETPOINT and ALL sWABLE VALUES (item 3 above) and the differential flow TPIP SETPOINT and ALLOWABLE VALUFS (item 4 above) demonstrates that rapid response time from the differential flow trip channelis not critical.
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The involved differentiel pressure transmitters, thermocouples and NUMAC equipment are subject to very little changa in response time relative to the 30 minute setting proposed for the isolation time delay associated with that funct..M. Any subtle response time changes that might occur.
f would be insignificant relativo to margins present in the model and result of General Electric Report GE-NE 77014-0592.
The current Technical Specification identifies the exist;ng 45 second differential flow delay time as both a response time requirement (Table 3.3.2 3, item 3.a) and as a setpoint (Table 3.3.2 2, item 3.fl. It is CP&L's position that this duplication is unnecessary and that the "30 minute
- differential flow delay time parameter would best be treated as a setpoint, rather than as a response time. The surveillance testing of this parameter when treated as a setpoint wi!! continue i
to ensure that the trip channel perfvenance is adequate to satisfy the performance assumptioiss i
utilized in the GE leakage analysis report lh f,MRB.ENT REQUIREMENI:
Technical Specification Tables 3.3.2-1 (Isolation Actuation Instrumentation),3.3.2 2 (Isolation Actuation Instrumentation Setpoints),3.3.2 3 (Isolation System instrumentation Response Times) and Tabic 4.3.2-1 (Isolauon Actuation inctrumentation Surveillance Requirements) currently identify item 3.f in each table as " Reactor Water Cleanup System isolation, a Flow High Time Delay Relay."
PROPOSED CHANGE:
In each of those tables, drop the word
- relay
- so that the revised description of item 3.f will read
" Reactor Water Cleanup System isolation, 4 Flow High Time Delay."
DANS in the existing configuration of the differential flow trip function, the time delay function is performed by actual time delay relays. Within the replacement NUMAC system, the time delay function is performed via software within the NUMAC microprocessor.
l This item description change is necessary to avoid potential confusion in future application and interpretation of this Technical Specification line item.
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l C.URRENT REQUIREMENI:
None.
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' Add to Technical Specification Tables 3.3.21,0.3.2 2,3.3.2 3, and 4.3.21 a new reactor water cleanup system isolation actuation instrumentatlors functiof 6 covuing piping outside of the reactor water cle6nup system room.
DASIS:
Two area high tamperature instrumentetion channels for initiating reactor water cleanup system
- Isolation were installed and declared operablo on Febtuary 13,1991 (Unit 1) and August 10,1990 (Unit 2). The design sad function of these leak detection instruments is similar to the existing leak detection ternperature monitoring instruments currently identified in the Technical Specifications; therefore, responsa times and surveillance frequencies that are the same as those establiohed for -
existing leak detection to:nperature monitoring instrumentation are being proposed; CONCLUSION:
The p.oposed changes as described above are (ntended to reflect the physical configuration changes that will result from the NUMAC Installation, to optimize and standardire surveillance requirements consistent with the impeoved reliability and technology that the NUMAC system provides relative to the existing Riley, GnMAC and fenwal instrwnentation, to accommodete the results of upgraded setpoint uncertainty and line break conse urtnce calculations, and to recognize s
an additional RWCU isolatior trip function installed sesoral years ago.
it is concluded that these changes are all fully justifiablo. Continued safe operation of the =
Drunswick Steam Electric Plant is assured.
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ENCLOSURE 2 BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 1.
NRC DOCKET NOS. 50 325 & 50 324 OPERATING LICENS6 NOS. DPR 71 & DPR 02 REQUEST FOR LICENSE AMENDMENT STEAM LEAK CETECTION INSTRUMENTATION NUMAC UPGRADE 10 CFR 50.92 EVALUATION The Commission has provided standards in 10 CFR 50.92(c) for determining whe".her a significant.
hazards consideration exists. A proposed amendment to an operating ' license it a facility involves no 64,nificant hazards consideration if ooeration of the facility in accordance wit i the proposed amendment would not: (1) involve a significant increase in tne probability or consequences of an accident previously evaluated, (2) create the possibility of a new or different kind of accident from ar>y accident previously evaluated, er (3) involve a significant reduction in a margin of safety, Putsuant to 10 CFR 50.91(a)(1), Carolina Power & Light Company has reviewed this proposed license amendment request and determined that its adoption would not involve a significant harudt, consideration. The bases for this determination are as follows:
Standard 1:
The proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated because:
llCm_1.:
The proposed change will el'tninate the requirement to perform a daily CHANNEL CHECK for the reactor water cleant p system high differential flow isolation function. No accident initiators or precursors are changed by the proposed elimination of the daily CHANNEL CHECK for the affected steam leak detection instrument, Based on the significantly reduced drift rate of the NUMAC based system rind the NUMAC system's self test and self-diagnostic features, the affected leak detection instrumentation will continue to perform its des!gn isolation function, Therefore, the proposed change will not significantly increase the probability of an accident previously evaluated.
The proposed change will not alter the assumptions used in analyses of reactor water cleanup system leaks. The NUMAC estem orift characteristics are superior to those of the existing UEMAC and Fenwalleak detection equipment. The microprocessor based NUMAC system being installed has diagnostic and self test features that will simplify maintenance activities and improve the operator / equipment interface. The proposed deletion of the daily CHANNEL CHECK for the affected leak detection instrumentation will not degrade the ability of the equipment to perform its isolation function in addition, other leak detection instrumentation specified by the Technical Specifications will continus to provide alterr. ate E21
~_
t means of detecting and mitigating the consequences of reactor water cleanup system leaks. Therefore, the proposed changes to the surveillance intervals will not significantly affect the consequences of an accident previously evaluated.
hidIL2:
The proposed changes wilh (U revise the requirement to perform a CHANNEL FUNCTIONAL TEST from a monthly frequency to a semi annual frequency for the reactor water cleanup system high differential flow, high area temperaturn. high area ventilation differential temperature, and h!gh differential flow time delay relay isolation functions, for the high pressure coolant injection system high steam line tunnel temperature, high steam line ambient temperature, high steam line area differential temperature and high equipment area temperature and for the reactor ccre isolation cooling system high steam line tunnel temperature, high steam line ambient temperature, high steam line area differential tempeiature, high equipment room ambient temperature, high equipment room differential temperature and high steam line tunnel temperature tirne delay relay isolation trip functions, and (2) revise the requnement to perinrm a CHANNEL CAllBRATION from a quatterly frequency to a refueling frequency for the high pressure coolant injection system high steam line tunnel temperature and high equipment area temperature, and for tha reactor core isolation cooling system high equipment room ambient temperature, and high equipment room differential temperature isolation trip functions. No accident initiators or precursors are char.ged by the proposed changes to the surveillance intervals for the affected steam leak detection instruments. Based on the significantly reduced drift rate of the NUMAC based system and the NUMAC system's self test and self diagnostic features, the affected leak detection instrumentation will continue to perform its design isolation function. Therefore, the proposed changes will not significantly increase the probability cf an accident prnviously evaluated.
The proposed changes to the surveillance intervals for the effected steam leak detection instruments will not change or alter the assumptions useo in analyses of either reactor water cleanup, high pressure cools ! injection, or reactor cose isolation cooling system leaks. The NUM/ C system possesses drift characteristics superior to those of the existing GEMAC leak detection equipment. The microprocessor based NUMAC system being installed has diagnostic and self test features that will simplify maintenance activities and improve the operator / equipment inte-f ace. The preposed changes to the surveillance frequencies for the affected leak detection instrumentation will not degrade the abiliiy of the equipment to perform its isolation function. In addition, other leak detection instrumentation specified by the Technical Specifications will continue to provide alternate means of detecting and mitigating the consequences of reactor water cleanup system and reactor core isolation cooling system leaks. Therefore, the proposed changes to the surveillance frequencies will not significantly affect the consequences of ar. accident previously evaluated, htLO The proposed change revising tha TRIP SETPOINT and ALLOWABLE VALUE for the reactor watet cleanup system high differential flow time delay relay from " s 45 seconds" to E2-2 1
1
l f
- s 30 minutes" will not significantly increase the probability of an accident previously evaluated. The sole design basis func'.ico for the reactor water cleanur system high differential flow isolation function is to assure compliance with the offsite and control room dose limitations imposed by 10 CFR 100 and 10 CFR 20. The high differential flow isolation function is not intended for protection of reactor pressure vessel water levels or for limiting the reactor building environment for environmental qualification purposes. The proposed changa to the reactor water cleanup system high differuntlat flow isolation function will not affect any hitiating mechanism for a previously evaluated accident.
Therefore, the proposed change wi!! not significantly increase the probability of an accident previously evaluated.
The proposed change revising the TRIP SETPOINT and ALLOWADLE VALUE for the reactor water cleanup system high differential flow time delar relay will not significantly affect the consequences of an accident previously evaluated. General Electric Company's proprietary report GE NE 770-14 0592 evaluates the consequences of a 300 gal / min reactor water cleanup system cold leak romaining unlr.olated for 30 minutes. That analysis utilized conservative source terms and assumptions. The resultant control room, site boundary and low population zone doses are within the limits prescribed in Standard Review Plan 6.4 and 10CfR100 and are less than dose consequences previously evaluated for other BNP events.
Plant sump monito= ring instrumentation, room flood alarms and plant leakage responte procedures preclude the possibility that adverse room flooding conditions could result from the increased time delay on the autcmatic RWCU system isolation. Therefore, the proposed change will not significantly affect the consequences of an accident previously evaluated.
hMn.d The proposed change revising the TRIP SETPOINT and ALLOWABLE VALUE for the reactor water cleanup system high differential flow time delay relay from 's 53 gal /m!n" to 's 73 gal / min' will not significantly increase the probability of an accident previously evaluated.
The sole design basis function for the reactor water cleanup system high differential ficw isolation function is to assure complianca with the offsite and control room dose limitations imposed by 10 CFR 100 and 10 CFR 20. The high differential flow isolation function is not intended for protection of reactor pressure vessel water levels or for limiting the reactor building environment for environmental qualification purposes. The proposed change to the reactor water cleanup system high differential flow isolation function will not affect any initiating mechanism for a previously evaluated accident. Therefore, the proposed change will not significantly increase the probability of an accident previously evaluated.
The proposed change revising the TRIP SETPOINT and iLLOWABLC VALUE for the reactor water cleanup system high differential flow trip function Ai'l not significantly affect the consequences of an accident previously evaluated. General Electric Company's proprietary report GE NE 7 70-14-0592 evaluates the consequences of a 300 gal / min reactor water cleanup system cold leak remaining unisolated for 30 minutes. That analysis utilized cohorvative source terms ad assumptions. The resultant control room, site boundary and low population zone doses are within the limits prescribed in Standard Review Plan 0.4 and 10CTR100 and are loss than dose conseo' ences previously c"aluated for other BNP events, and documerta that 300 gal / min v'.s
- the engineered ANALYTICAL LIMIT. The E2-3 l
proposed TRIP SETPOINT and ALLOWABLE VALUE increase establishes a surveillance test acceptance criteria that will assure that the 300 gal / min ANALYTICAL LIMIT cannot be exceeded during any postulated plant condition. Therefore, the proposed change will not significantly affect the consequences of an accident previoutly evaluated.
homi The proposed change to delete the Instrument response time requirement for the high pressure coolant injection system steam line tunnel temperature high isolation function will not significantly increar,e the probability of en accident previously evaluated. No occident initla' ors or precursor 6 are changed by the proposed chtnge to the HPCI steam line tunnel temperature high response time requirement. Therefore, the proposed change will not significantly increase the pobability of an accident previously evaluated.
The proposed change to delete the instrument response time requirement for the affected steam leak detection instrument will not change or after the assumptions used in analyses of high pressure coolant injection system steam line breaks. The current Reactor Duilding _
Environmental Report does not take credit for the response times of the high pressure coolant injection system temperature trip functions since leak detection and Isolation initiation for large L:.aks (a 300% flow) is provided by high flow instrumentation, The high flow instrumentation provides a faster, more direct indication of large breaks. The worst case is the 10 inch high pressure coolant injection double ended guillotine break. The temperature based leak detection instrumentation is intended to detect small breaks
(<300% flow) that are below the threshold of the high flow instrumentation. The response time of the temperatute based leak detection instrunsentation is fut a critical parameter since the reduced flow rates associated with smaller breaks permits a lor ger response time for detection and isolation initietion of the leak. For the high pressure coolant injection smallline break (< 300% flow), the temperature profile extrapo!ation of the postulated high energy line break conditions in the reactor building at the 20 foot and 17 foot elevations, until valve closure is achieved, is within the previously established envlronmental qualification profiles and does not exceed the peak temperatures. Thus,0 response time of the instrumentation is not critical in assuring high energy line break mitigation / valve isolau n and response time testing is not warranted. The proposed changes to delete the response time requirement for the affected leak detection instrumentation will not degrade the ability of the equipment to perform its isolation function. The NUMAC system possesses drift characteristics superior to those of the existing GEMAC leak detection equipment. The microprocessor based NUMAC system.
being installed has diagnostic cnd self test features that will simplify maintenance activities and improve the operator / equipment interface. Based on the significantly reduced drift rate of the NUMAC based system and the NUMAC system's self test and self diagnostic features, the affectod leak detection instrumentation will continue to perform its design isolation function. in addition, other leak detection instrumentation specified by the Technical Specifications will continue to provide alternate means of detecting and mitigating the consequences of high pressure coolant injection system steam line leaks.
Therefore, the prooosed change will not significantly affect the consequences of an accident previously evaluated.
E2-4 l
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The proposed change to delete the instrument response time requirement for the reactor water cleanup system area temperature high and area ventilation differential temperature
- high isolation functions will not significantly increase the probability of an accident previously evaluated. No accident inltiators or precursors are changed by the proposed change to the reactor wi' cleanup system area temperature response time requirements.
Therefore, the pr n ised change will not significantly increase the probability of an accident previously evaluated.
The reactur water cleanup system area temperature trip function is provided to detect system steam leaks. The system utilizes thermorouples as the temperature sensing devices. Thermocouples are inherently stable devices with little potential for s!gnificant changs in response time. The CHANNEL FUNCTIONAL TESTS and CHANNEL Call 0 RATIONS required by Technical Specifications provide adequate assurance of the instruments' abllity t'., sense a reactor water cleanup system steam leak and to isolato the system. The NUMAC system possesses drift characteristics superior to those of the existing GEMAC leak detection equipment. Also, the NUMAC system has diagnostic and self test features that will simplify maintenance activities and improve the operator / equipment Interface. Based on the significantly reduced drift rate of the NUMAC system, as well as the system's self test and self diagnostic features, the proposed change.
to delete the instrument response timo requirement for the affected steam leak detection instruments will not charv s or alter the assumptions used in analyses of rear: tor water cleanup system steam line breaks, nor will the deletion degrade the ability of the equipment to perform its isolation function. Therefore, the proposed change will not significantly affect the consequences of an accident previously evaluated, httIL72 The proposed change to delete the instrument response time requirement for the reactor water cleanup system differential flow - high isolation function will not significantly increase the probabit;ty of an accident previously evaluated. No accident initiators or precursors are changed by the proposed chango to the reactor water cleanup system differential flow response time requirements. Therefore, the proposed change will not significantly increase the probability of an accident previously evaluated.
The reactor water cleanup system d;fferential flow trip function is provided to detect system leaks at temperatures below the sensitivity of the system's temperature trip channels. CHANNEL FUNCTIONAL TESTS and CHANNEL CAllBRATIONs required by Technical Specifications provide adequate assurance of the lostrumer s' ability to sense a reactor water cleanup system steam leak and to isolate the system. Based on the significantly reduced drift rate of the NUMAC system, as well as the system's self test and self-diagne Q features, the propesci change to delete the instrument response time requirement for the affected steam leak detection instruments will not change or alter the assumptions used in 'alyses of reactor water cleanup system steam line breaks, nor will the deletion degrade ability of the equipment to perform its isolation function.
E2-5
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Therefore, the proposed changs will not significantly affect the consequences of an s.ccident previously evaluated.
htt(ILD:
The proposed chsnoe to update the descriptive title of this Technical Specification is required only to reflect the configuration change associated with the NUMAC upgrade. The time delay function will still exist but will be oorformed within the NUMAC software rather that by a discrete tima delay relay equipment item. This change will not significantly increase the probacility of an accident previously evaluated. Installation of this instrumentation will have no impact on accident precursors or initiators. Therefore, the proposed change will not significantly increase the probability of an accident previously evaluated.
The proposed change to update the descriptivt title will not significantly affect the consequences of an accident previously evaluated. This description change will not adversely affect radiological releases from a postulated reactor water cleanup system leak.
Thus, the proposed change will not significantly affect the consequences of an accident previously evaluated.
001I1 E:
The proposed change to add new requirements to the Techrical Specifications for a new reactor water cleanup system isolation actuation instrumentation function covering pipir.g outside of the reactor water cleanup system room will not significantly increase the
- probability of an accident previously evaluated, installation of this instrumentation will have no impact on accident precursors or initiators. Therefore, the proposed change will not significantly increase th: probabili y of an accident previously evaluated.
t The proposed change to add new requirements to the Technical Specifications for a new reactor water cleanup system isolation actuation instrumentation function covering piping outside of the reactor water cleanup system room will not significantly affect the consequences of an accident previously evaluated. Installation of the new instrumentation will not adversely effect radiological releases from a postulated reactor water cleanup system leak. Indeed, addition of the leak detection instruments for piping outsido of the ruactor water cleanup system room will enhance the abilty to detect and isolation reactor water cleanup system leaks. Thus, the proposed change will not significantly affect the consequences of an accident previously evaluated.
$1MldN.d_2:
The proposrtd amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated because:
E2 6
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i laati:
The proposed change will eliminate the regulrernent to perform a daily CHANNEL CHECK for the reactor water cleanup high differential flow isolation function, No accident initiators or precursors are changed by the proposed elimination of the daily CHANNEL CHECK for the affected steam leak detection instrument The proposed change will not adversely offect the availability of the steam leak detection instrument to detect and mitigate reactor water cleanup system leaks. Tbs instrument involved will continut, to function as currently designed. No now modes of plant operation will be created as a result of the proposed changes. In addition, the proposed change to climinate the daily CHANNEL CHECK surveillance for the steam leak detection instrument will not cause the initiation of any accidents nor create any new credible single failure. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
P lMUL2:
The proposed change will optimize and standardito the CHANNEL FUNCTIONAL TEST and CHANNEL CAllBRATION surveltlance frequency requirements for the high pressure coolant injuction, reactor core isolation cooling and reactor water cleanup systems ambient and differential temperature and the reactor water cleanup system differential flow isolation trip functions. The proposed change will not *dversely affect the availability of the steam leek detection instrument to detect and mitigate high pressure coolant injection, reactor core isolation cooling or reactor water cleanup system leaks. The instruments involved will continue to funulon exactly as currently designed. No new modes of plant operation will be created as a result of the proposed changes. In addition, the proposed change to the surveillance testing requirements for the steam leak detection instrumentation will not cause the initiation of any accidents nor create any new credible single failure. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated, item 3:
The proposed change revising the TRIP SETPOINT and ALLOWADLE VALUE for the reactor water cleanup system high differential flow time delay relay from "s 45 seconds
- to
- s 30 minutes" will not create the possibility of a new or different kind of accident from any accident previously evaluated. The proposed change to the TRIP SETPOINT and ALLOWABLE VALUE for these instruments will not adversely affect the availability of the instrument to detect and mitigate reactor water efeanup system leaks. The instrument will otherwise continue to functioa exactly as currently design 9d. No new modes of plant operation will be created as a result of the proposed change. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
i E2-7 o
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fte m 4 The proposed change revising the TRIP SETPOINT and ALLOWABLE VALUE for the reactor watet cleanup system high differential flow time delay relay from "s 53 gal / min' to
- s 73 gal / min
- will not create the possibility of a new or different kind of accident from any accident previoutli evaluated. The proposed change to the TRIP SETPOINT and ALLOWABLE VALUE for these ir'struments will not adversely affect the availability of the instrument to detect and mitigate reactor water cleanup system leaks. The instrument will otherwise continue to perform the some function as does the cucrent design. No new modes of plant operation will be created as a result of the proposed change. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accideitt previcusly evaluated.
Item 5:
The ; toposed change to delete the instrument response time requi'ement for thn High Pressure Coolant injection System Steam Line Tunnel Temperature High isolation function will not create the possibility of a new or different kind of accident from any accident previously evaluated. The replacement NUMAC instrumentation involved will continue to perform the name function as does the existing hardware configuration. The proposed changs deleting the instrument response time requirement for these leak detection instruments will not cause the initiation of any accidents nor create any new credible 4, ingle failure. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
Item _0:
The proposed change to delete the instrument response time requirement for the Reactor Water Cleanup System Area Temperature and Area Ventilation Differential Temperature isolation functions will oci create the possibility of a new or different kind of accident from any accident previously evaluated. The replacement NUMAC instrumentation involved will continue to perform the same fun:: tion as does tiie existing hardware configuration. The proposed change deieting the instrument response time requirement for these leak detection instruments will not cause the initiation of any accidents nor create any now credible single failure. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
Item 7:
The proposed char:ge to delete the instrument response time requirement for the Reactor Water Cleanup System Differential Flow Time Delay Relay isolation functions will act create the possibility of a new or different kind of accident from any accHent previously evaluated. The replacement NUMAC instrumentation involved will continue to perform the same function as does the existing hardware configuradon. The proposed change deleting the instrument response time requirement for these leak detection instruments will not cause the initiation of any accidents nor creato a ny new credible single failure. Therefore, E2-8 I
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i i
the pre;,cted change does not create the possibility of a now or different kind of accident from any accident previously evaluated.
ama:
The proposed change to update the descriptive title of this Technical Specification is required on.f V.eflect the configuration change associated with the NUMAC upgrade. The time delay function will still exist but will be performed within the NUMAC software rather that by a discrete time delay relay equipment item. This change will not cause the initiation of any accidents, Will not create any new credible single failure, and will not after any accident initiator or precursor, furthermore, no new modes of plant operation will be
)
created as a result changing this line item's description. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
r hem. 9:
The proposed change to add new requirements to the Technical Specifications for a new reactor water cleanup system isolation actuation instrumentation function covering piping outside M the reactor water cleanup system room will not create the possibility of a now or different kind of accident from any accident previously evaluated. Addition of the new instrumentation will not cause the initiation of any accidents, will not create any new credible single failure, and will not alter any accident initiator or precursor. Furthermore, no new modes of plant operatloa will be created as a result of adding the new reactor water cleanup system leak detection instrumentation, lherefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
- S.tandEd 3:
The pioposed amendment does not involve a significant reduction in the margin of safety because:
Itsn.1:
The proposed deletion of the daily CHANNEL CHECK for the reactor water cleanup differential flow trip function w%l rd Micantly reduce a margin of safety, Due to the NUMAC system's enhanced self tm e.W sif diagnostic features and the NUMAC system's enhanced operator interface, elimination of this daily CHANNEL CHECK will not reducu the reliability of the instrumentation. Therefore, the proposed change does not involve a significant reduction in a margin of safety.
E2-9 s-
o hCIIL2:
t The proposed changes to the surveillance intervals for the affected steam leak detection instruments will not significantly redure a margin of safety, Due to the improved drift characteristics of the NUMAC based leak defection instrumentation, the NUMAC system's enhanced self test and self diagnostic features, and the NUMAC system's enhanced operator interface, decroasing the surveillance frequencies of the leak detection instrument functions involved wi!! not reduce the reliability of the instrumentation. Therefore, the proposed change does not involve a significant reduction in a margin of safety, item.3:
The proposed change revising the TRIP SETPOINT and ALLOWABLE VALUE for the reactor water cleanup system high differential flow time delay from
- s 45 seconds" to
's 30 minutes" wili not significantly reduce a margin of safety. The sole design basis function for the reactor water cleanup system high differential flow isolation function is to assure compfbnce with 10 CFR 100 and 10 CFR 20. The high differential flow isolation function is not intended for protection of reactor pressure vessel water levels or for limiting the reactor building environment for environmental qualification purposes. General Electric Company's proprietary report GE NE 77014 0592 evaluates the consequences of a 300 gal / min reactor water cleanup system cold leak remaining unisolated for 30 minutes. That analysis utillied conservative source terms and assumptions. The resultant control room, site boundary and low pooulation zone dosas are within the limits prescribed in Standard Review Plan 6.4 and 10CFR100 and are less than dose consequences previously evaluated for other BNP events. Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Item 4:
The proposed change revising the TRIP SETPOINT and ALLOWABLE VALUE for the reactor water cleanup system high differential flow from
- s 53 gat / min' to 's 73 gal / min' will not significantly reduce a margin of safety. The sole design basis function for the reactor water cleanup system high differential flow isolation function is to assure compliance with -
10 CFR 100 and 10 CFR 20. The high differential flow isolation function is not intended for protection of refetor pressure vessel water levels or for limiting the reactor building environment for environmental qualification purposes.' General Electric Company's proprietary report GE-NE 770-14-0592 evaluates the c( tequences of a 300 gal / min reactor water cleanup system cold leak remaining unisolated for 30 minutes. That a >
is utilized conservative source terms and assumptions. The resultant control room, site boundary and low population zone doses are within the limits prescribed in Standard Review Plan 0.4 and 10CFR100 and are less than dose consequences previously evaluated for other BNP events. Therefore, the proposee change does not involve a significant reduction in a margin of safety.
E210 a
e htLf!LE:
Tne prnposed change to delete the instrument response time requirement for the High Pressure Coolant injection System Steam Line Tunnel Temperature Hloh isolation function will not significantly reduce a margin of safaty. Leak detection system thermocouples are inherently stable devices which exhibit little drift or changes in responto time characteristics over long periods of time. Thus, eliminating response time testing of the affected instrument channels will not result in a degradation of the leak detection system's ability to respond. The periodic performance of required CHANNEL FUNCTIONAL TESTS and CHANNEL CAllBRATIONS for the High Pressure Coolant Injection System Sterm Line Tunnel Temperature. High isolation functiors provides adequate assurance of the ability of ~
the instrument to respond. Therefore, the proposed change does rmt involve a significant reduction in a margin of **fety.
h0m.01 The proposed change to delete the instrument response time requirement for the Reactor Water Cleanup System area temperature and area ventilation differential temperature trip functions will not significantly reduce a margin of safety. Leak detection system thermocouples are inherently stable devices which exhibit little drift or changes in response time characteristics over long periods of time. Thus, eliminating response time testing of the affected Instrument channels will not result in a degradation of the leak detection system's ability to respond. The periodic performance of required CHANNEL FUNCTIONAL TESTS and CHANNEL CAllBRATIONs for these Reactor Water Cleanup temperature channel trip functions will provide adequate assurance of the ability of the Instrument to respond.
Therefore, the proposed change does not involve a significant reduction in a margin of.
safety.
html:
The proposed change to delete the instrument response time requirement for the Reactor Water Cleanup System differential flow trip function will not significantly reduce a margin of safety. The involved differential pressure transmitters, thermocouples and NUMAC equipment are subject to very little change in response time relative to the 30 minute setting proposed for the isolation time delay associated with that function. Any subtle response time changes that might ocsur would be insignificant relative to margins present in the model and the result of GE report GE-NE 770-14-0592. Thus, eliminating response time testing of the affected Instrument channels will not result in a degradation of the leak detection system's ability to respond. The periodic performance of required CHANNEL FUNCTIONAL TESTS and CHANNEL CAllBRATIONs for these Reactor Water Cleanup differential flow channel trip function will provide adequate assurance of the ability of the instrument to respond. Therefore, the proposed change does not involve a significant reduction in a margin of safety.
E2-11
e htLritJ:
The proposed change to update the descriptive title of this Technical Specification is required only to reflect the configuration change associated with the NUMAC upgrade The time delay function will still exist but will be pe formed within the NUMAC software rather that by a discrete time delay relay equinmane !.em. This change will not will not significantly reduce a margin of safety.
Jtem. 9:
The proposed change to add new requirements to the Technical Specifications for a new reactor water cleanup system isolation actuation instrumentation function covering piping outside of the reactor <ater cleanup system room will not significantly reduce a marglu of safety. Installation of new leak detection instrumentation covering piping outside of the reactor water cleanup system room provides additlo.1al assurance of the timely detection and mitigation of reactor water cleanup system leaks. Therefore, the proposed change actually increases the margin of safety.
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.e ENCLOSURE 3 BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 NRC DOCKET NOS, 50-325 & 50 324 OPERATING LICENSE NOS. DPR 71 & DPR 62 REQUEST FOR LICENSE AMENDMENT STEAM LEAK DETECTION INSMUMENTATION NUMAC UPGR ?E ENVinONMENTAL CON! ' RATIONS 10 CFR 51.22(c)(9) provides criterion for and identification of licensing and regulatory actions oligible for categorical exclusion from performing an environmental assessment. A proposed amendment to an operating license for a facility requires no environmental assessment if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant hazards considerailon; (2) ren!! n a significant change in the types or significant increase in the i
amounts of any effluents that may be released offsite; (3) result in an increase in individual or; cumulative occupational radiation exposure. Carolina Power & Light Company has reviewed this request and determined that the proposed amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9), Pursuant to 10 CFR 51.22(b), no environmentalimpact statement or environmental assessment needs to be prepared in connection with the issuance of' the amendment. The basis for this determination follows:
Pronosed Chanags.
The proposed changes include reflection.t the physical r enf:guration changes that will result from the NUMAC int allaton, optimization and standardizatiot, of surveillanN Nuirements consistent '
i with tha improved reliaYiity and technology that the NUMAC system provides relative to the -
existing Riley, GEMAC and Fenwal instrumentation, accommodati'.n of the results of upgraded setpoint uncertr'nty ars line break consequence calculations, and recognition of an additional -
RWCU isolation trip function installed several years ago.
fMig:
The change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) for the fol lowing reasons:
1.
As demonstrated in Enclosure 2, the proposed amendment does not involve a significant hazards consideration.
2.
The proposed amendment does not result in a significant change in the types or significant increase in the amounts of any effluents that may be released offsite.
The only line break for which the detection and/or mitigation action will be any different after the propused modification ed Techn! cal Specification change is the reactor water cleanup cold leak as mitigated by the differential flow instrumentation. The isolation time t: 1
delay on this break will be increased from 45 seconds to 30 minutes. The types of potential effluents remain unchanged. The increased isolation time delay will result in a proportionalincrease in the amount of potential effluent; however, the results of GE report GE NE 77014 0592 demonstrate that the magnitude of those relea:es will be significantly less than the applicable regulatory limits and significantly less than those of BNPs bounding high energy line break as described in UFSAR Section 15.6.3 Main Steam Line Break Accident. As such, the change has no impact on the types and will not increase the amounts of ar7y effluents that may be released beyond the values previously evaluate,d and approved for BNP, 3.
The p aposed amr '.dment does not ren.lt in an increase in individual or cumulative occupational radiation exposure.
The only line break for which the detection and/or mitigation actie. will be any different after the proposed modification and Technical Specification change is the reactor water cleanup cold leak as mitigated by the differential flow instrumentation. The isolation time delay on this break will be increased from 45 seconds to 30 minutes. The increased isolation time delay will result in a rroportional increase in the potential resultant radiation exposures; however, the results of GE report GE NE-77014-0592 demonstrate that the magnitude of those exposures will be significantly less than the applicable regulatory limits and significantly less than those of DNPs bounding high energy line break as described in UFSAR Section 15.6.3 Main Steam Line Break Accident. Therefore, the amendment has no affect on either individual or cumulative occupational radiation exposure beyond the values previously evaluated and approved for BNP.
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