ML20059G697
| ML20059G697 | |
| Person / Time | |
|---|---|
| Site: | Catawba, McGuire, Mcguire  |
| Issue date: | 01/10/1994 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20059G696 | List: |
| References | |
| NUDOCS 9401250144 | |
| Download: ML20059G697 (23) | |
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Marked-un Technical Specification Paces l McGuire , r I s 7 r Y f
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v_ ~ = r TABLE 2.2-1 ' S ~ 5 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS E . i FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES c - 5 1. Manual Reactor Trip N.A. N.A. g 2. Power Range, Neutron Flux Low Setpoint 5 25% of RATED Low Setpoint 1 26% of RATED e., THERMAL POWER THERMAL POWER E m High Setpoint 1 109% of RATED High Setpoint 1 110% of RATED THERMAL POWER THERMAL POWER
- 3. Power Range, Neutron Flux, 5 5% of RATED THERMAL POWER with i 5.5% of RATED THERMAL POWER High Positive Rate a time constant 1 2 seconds with a time constant 1 2 seconds
- 4. Intermediate Range, Neutron 5 25% of RATED THERMAL POWER 1 30% of RATED THERMAL POWER Flux y 5. Source Range, Neutron Flux $ 10s counts per second 5 1.3 x 10s counts per second
- 6. Overtemperature AT See Note 1 See Note 3
- 7. Overpower AT See Note 2 See Note 4 8.~ Pressurizer Pressure--Low 1 1945 psig 1 1935 psig k k 9. Pressurizer Pressure--High 1 2385 psig 5 2395 psig EE gg 10. Pressurizer Water Level--High < 92% of instrument span
~ ~< 93% of instrument span ?? 91 % 90 % - . - . 11. Low Reactor Coolant Flow - D L of minimum measured > M of minimum measured .E .E how per loop
- T109 per loop
- _
- Minimum measured flow is 96,250 gpm per loop. "
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LIMITING SAFETY SYSTEM SETTINGS ' I BASES l Pressurizer Pressure In each of the pressure channels, there are two independent bistables, each with its own trip setting to provide for a High and Low Pressure trip
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thus limiting the p assure range in which reactor operation is permitted. The Low Setpoint trip protects against low pressure which could lead to DNB by. tripping the reactor in the event of a loss of reactor coolant pressure. On decreasing power the Low Setpoint trip is automatically blocked by P-7 , (a power level of approximately 10% of RATED THERMAL POWER with turbine i impulse chamber pressure.at approximately 10% of full power equivalent); and on increasing power, automatically reinstated by P-7. _ The High Setpoint trip functions in conjunction with the pressurizer relief and safety valves to protect the Reactor Coolant System against system overpressure. l' Pressurizer Water Level --- The Pressurizer High Water Level trip is provided to prevent water relief through the pressurizer safety valves. On decreasing power the Sressurizer 1' High Water Level trip is automatically blocked by P-7 (a power livel of I approximately 10% of RATED THERMAL POWER with a turbine impulse enamber pressure at approximately 10% of full equivalent); and on increasing power, automatically reinstated by P-7. Low Reactor Coolant Flow The Low Reactor Coolant Flow trips provide core protection to prevent DNB by mitigating the consequences of a loss of flow resulting from -the loss of one or more reactor coolant pumps. On increasing power above P-7 (a power level of approximately 10% of RATED THERMAL POWER or a turbine impulse chamber pressure at approximately 10% of full power equivalent), an automatic Reactor trip will occur if the flow in more than one loop drops below'89%,of nominal full loop flow. Above P-8 (a
-power level of approximately 48% f RATED THERMAL POWER) an automatic Reactor trip .will occur if the flow in a single loop drops below '89%, of nominal full loop flow. Conversely on decreas ng power between P-8 and he P-7 an automatic Reactor trip will occur n loss of flow in more han one loop and' below P-7 the trip function is auto atically blocked.
9f '{' R I '[, McGUIRE . UNITS 1 and 2 B 2-6 p -
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'i POWER DISTRIBUTION LIMITS .
l 7 i i 3/4.2.5 DNB PARAMETERS SURVEILLANCE REQUIREMENTS
-4.2.5.1 Each of the parameters of Table 3.2-1 shall be measured by averaging the. indications (meter or computer) of the operable channels and verified to be ,
3 within their limits at least once per 12 hours.
- 4. ;2. 5. 2 The RCS total flow rate indicators shall be subjected to a CHANNEL t CALIBRATION at least once per 18 months.
4.2.5.3 The RCS total flow rate shall be determined by peesis4or heat balance-measurement at least once per 18 months. I k l l( i i i
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I l i l i l I1 l 4: - 1 McGUIRE - UNITS 1 AND 2 3/4 2-22a Amendment No. (Unit 1) Amendment No. / ? (Unit 2) }-
i i POWER DISTRJBUTI N LIMITS ( Figure 3.2 - 1. Reactor Coolant System Total Flow Rate Versus Rated Thermal Power - Four Loops in Operation l 388850 A penairy of 0.1% for undetectec fe cwater , Permissible ventun fouting and a measurement uncertainty Operanon cf 4-7% for flow are inctuced in tnis figure. . Regen g (98.385000). 385000 ---------------------------------------- p (96.381150) E , ! a. S 381150 - Restncted 2 Oooratton
'5 Region .
(94.377300) C 377300 - E 2a
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' (92.373450) Coormoon ~;
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i 365750 D 361900i . . . . . . . 86 88 90 92 94 96 98- 100 102,
- 1, Fraction of Rated Thermal Power McGUIRE - UNITS 1 AND 2 3/4 2-24 Amendment No. d (Unit 1) l Amendment-No. t (Unit:2)'
POWER' DISTRIBUTION LIMITS . BASES , 3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distri-bution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during STARTUP testing and periodically during power operation. The limit of 1.02, at which corrective action is required provides DNB and linear heat generation rate protection with the x y plane power tilts. The peaking increase that corresponds to a QUADRANT POWER TILT RATIO of 1.02 is included in the generation of the AFD limits. The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correc-tion of a dropped or misaligned rod. In the event such action does not cor-rect the tilt, the margin for uncertainty on F q(X,Y,Z) is reinstated by reducing the power by 3% from RATED THERMAL POWER for each percent of tilt in excess of 2.0%. For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles. 3/4.2.5 DNB PARAMETERS The limits on the DNB-related parameters assure that each of the para- l meters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a design limit DNBR throughout each analyzed transient. As noted on Figure 3.2-1, RCS flow rate and THERMAL POWER may be " traded off"- ; against one another (i.e., a low measured RCS flow rate is acceptable if the ! power level is decreased) to ensure that the calculated DNBR will not be below the design DNBR value. The relationship defined on Figure 3.2-1 remains valid as long as the limits placed on the nuclear enthalpy rise hot channel factor, Fg (X,Y), in Specification 3.2.3 are maintained. The indicated T,yg values and the indicated pressurizer pressure values correspond to analytical limits of 592.6 F and 2220 psia respectively, with allowance for indication instrumen-tation measurement uncertainty. When RCS flow rate is measured, no additional l allowances are necessary prior to comparison with the limits of Figure 3.2-1 since a measure ent error of 1.7% for. RCS total flow rate has been allowpd for in determination of the design DNBR value. t heAsa w WemmpW, GREMeg Du ck E9aM ~To *%c VpM The measurement errar fer RCS total-flow-rate is based upon performing-s .5 mfg p)
) precision heat-baiance and using-the-restrit to celibrste the RCS flow rate indi hg 3.2-cater:;. Potent 4al-fculing of the feedwater venturi which might not be detected could bies the-result-frem the precision heat bslence in a non conversative McGUIRE - UNITS 1 AND 2 B 3/4 2-5 Amendment No. (Unit 1)
Amendment No. R(Unit 2)
POWER DISTRIBUTION LIMITS . BASES , 3/4.2.5 DNB PARAMETERS (Continued) eenner. Therefore, a penalty of 0.1% for undetected fouling of the feedwater N venturi is included in Figure 3.2-1. Any fouling which might bias the RCS flow rate measurement greater than 0.1% can be detected by monitoring and trending various plant performance parameters. Ifdetected,actionshallbe leithertheeffectofthefoulingshallbequantifiedandcompen taken before performing subsequent precision heat balance me*surements, i.e., ; the RCS flow rate measurement or the venturi shall be cleaned to eliminate the fouling. The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. Indication instrumentation measurement uncertainties are accounted for in the limits provided in Table 3.2-1. The measurement error for RCS total flow rate is based upon the performance of past precision heat ! balances. Sets of elbow tap coefficients, as determined during these heat balances, were averaged for
- each elbow tap to provide a single set of elbow tap coefficients for use in calculating RCS flow. This set of coefficients establishes the calibration of the RCS flow rate indicators and becomes the set of elbow tap coefficients used for RCS flow measurement. Potential fouling of the feedwater venturi, which might I i not be detected, could bias the result from these heat balances in a non-conservative manner.
4
'McGUIRE - UNITS 1 AND 2 B 3/4 2-Sa -
Amendment No.130 (Unit 1)'
. Amendment No.112 (Unit-2)
9 4 Attachment Ib Marked-un Technical Specification Paggs Catawba 1 i
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v v v UNIT 1 DNLY TABLE 2,2-1 20 . REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPolNT ALLOWABLE VALUE f
- 1. Manual Reactor Trip N.A. N.A.
- 2. Power Range, Neutron Flux
- a. High Setpoint 5109% of RTP* s110.9% of RTP*
- b. Low Setpoint 525% of RTP* s27.1% of RTP*
- 3. Power Range, Neutron Flux, 55% of RTP* with a 56.3% of RTP* with High Positive Rate time constant a time constant a 2 seconds a 2 seconds
- 4. Intermediate Range, Neutron Flux s25% of RTP* s31% of RTP*
~ 5 g 5. Source Range, Neutron Flux s105 cps s1.4 x 10 cps
- 6. Overtemperature AT See Note 1 See Note 2
- 7. Overpower AT See Note 3 See Note 4
- 8. Pressurizer Pressure-Low 21945 psig 21938 psig***
- 9. Pressurizer Pressure-High 52385 psig s2399 psig
?/ 10. Pressurizer Water Level-High 592% of instrument span s93.8% of instrument span E . 9/ V9.7 @
& 11. Reactor Coolant Flow-Low 2%% of loop minimum a8tr9% of loop minimum
@ measured flow ** measured flow **
c. E 2 *RTP - RATED THERMAL POWER C j
** Loop minimum measured flow = 95,500 gpm
*** lime constants utilized in the lead-lag controller for Pressurizer Pressure-Loe are 2 seconds for lead and I second for. lag. Channel calibration shall ensure that these time.ccustants are adjusted to these
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*d-UNIT 2 ONLY ,
TABLE 2.2-1 E2 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPoINTS gj . SE TRIP SETPOINT ALLOWABLE'VALUE FUNCTIONAL UNIT
- 1. Manual Reactor Trip N.A. N.A.
EE n, 2. Power Range, Neutron Flux
- a. High Setpoint s109% of RTP* s110.9% of RTP*
- b. Low Setpoint 525% of RTP* s27.1% of RTP*
- 3. Power Range, Neutron Flux, s5% of RTP* with a 56.3% of RTP* with High Positive Rate time constant a time constant
= 2 seconds a 2 seconds Intermediate Range, Neutron Flux 525% o'f RTP* 531% of RTP*
4. 5 3*, 5. Source Range, Neutron Flux s105 cps s1.4 x 10 cps Overtemperature AT See Note 1 See Note 2 6. See Note 3 See Note 4
- 7. Overpower AT a1945 psig alc38 psig***
- 8. Pressurizer Pressure-Low Pressurizer Pressure-liigh 52385 psig 52399 psig 9.
s92% of instrument span s93.8% of instrument span (( 10. Pressurizer Water Level-High
- 87.7 he/%
% of loop minimum a88:9% of loop minimum Et 11. Reactor Coolant Flow-Low measured flow **
a measured flow ** a E. _. *RTP - RATED THERMAL POWER S
** Loop _ minimum measured flow - 96,250 gpm
*** Time constants utilized in the lead-lag controller for PiNssurizer Pressure-Low are 2 seconds for lead and I second for lag. Channel calibration shall ensure that these time constants are adjusted to these values.. ~ ._______.-
4 LIMITING SiFETY SYSTEM SETTINGS BASES Pressurizer Pcessure In each of the pressurizer pressure channels, there are two independent bistables, each with its own trip setting to provide for a High and Low Pressure trip tius limiting the pressure range in which reactor. operation is permitted. The '.ow Setpoint trip protects against low pressure which could lead to DN8 by tiipping the reactor in the event of a loss of reactor coolant pressure. On decreasing power the Low Setpoint trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, it automatically reinstated by P-7. The High Setpoint trip functions in conjunction with the pressurizer relief and safety valves to protect the Reactor Coolant System against system overpressure. Pressurizer Water Lev 11 The Pressurizer Figh Water Level trip is provided to prevent water relief through the pressurize
- safety valves. On decreasing power, the Pressurizer High Water Level trip "s automatically blocked by P-7 (a level of ap'proxi -
i mately 10% of RATED TH1 RMAL POWER with a turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing pe;ser, is automatically reinstateI by P-7. Reactor Coolant Flow The Low Reactor Coolant Flow trips provide core protection and prevents ' DNB by mitigating the consequences of a loss of flow resulting from the loss of one or more reactor coolant pumps. On increasing power abtve,P-7 (a power level of approximately 10% of RATED THERMAL POWER or a turoine impulse chamber pressure at approximately 10% of full power equivalent), an automatic Reactor trip will occur if the flow in more than one loop drops below'90%,of nominal full loop flow. Above P-8 (a power level of approximately 48%4of RATED THERMAL POWER) an automatic Reactor , trip will occur if the flow in any single loop drops below DS%gof nominal 'l full loop flow. Conversely, on ecreasing power between P-8)and P-7 an- I automatic Reactor trip will occu on low reactor coolant flow in more than ! one loop and below P-7 the trip f nction is automatically blo ked. ] 9/7l
$70 t .- l J
CATAWBA - UNITS 1 & 2 8 2-6 i
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l E0WER DISTRIBUT10t1 LIMITS 3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION
- 2. Within 24 hours of initially being within the region of prohibited operation specified on Figure 3.2-1, verify that the combination of TilERMAL POWER and Reactor Coolant System total flow rate are ,
restored to within the regions of restricted or permissible operation, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next-2 hours. SURVEIt. LANCE REQUIREMENTS
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4.2.5.1 Each of the parameters of Table 3.2-1 shall be verified to be within their limits at least once per 12 hours. 4.2.5.2 The Reactor Coolant System total flow rate indicators shall be sub-jected to a CHANNEL CALIBRATION at least once per 18 months. The measurement instrumentation shall be calibrated within 7 days prior to the performance of the eclee4 metric- flow measurement. 4.2.5.3 The Reactor Coolant System total flow rate shall be determined by ,
-prec4sion-heat bahce-measurement at least once per 18 months. (. ,
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i UNIT 1 ONLY l
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l 385,820 I 4 3 e tA t a t 0."s f or .noe i eci n
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Restricted E328,180 Operation (96.378,1AO) i S Regten
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E ' c. (n . E 370,5401 (92.370,MO) I i S o o : O ' 3 Prohibited g . Operation 5366,720. (90.366,720) Region c:: 1 362,900 4 I : 359;080 > 86 88 90 92 94 96 98 10 0 10 2 Fraction of Rated Thermal Power (
) Figure 3.2-1 Reactor Coolant Systen Total Flow Rate Versus Rated Thermal Power Four Loops in Operation CATAWBA UNIT 1 3/4 A2 16 heen&sent No.113 i
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UNIT 2 ONLY: 38881o
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365750 1 361900 a6 6a- so s2 e4 se sa m io2 Fraction of Rated Thermal Power Figure 3.2 1 Reactor Coolant System Total Flow Rate Versus Rated Thermal Power Four Loops in Operation CATAldM UNIT 2 3/4 B2 16 Amendment No. 107
POWER DISTRI.BUTION LIMITS ( BASES 3/4.2.5 DNB PARAMETERS (Continued) l to maintain a design limit DNBR throughout each analyzed transient. As noted on Figure 3.2-1, Reactor Coolant System flow rate and THERMAL POWER may be
" traded off" against one another (i.e., a low measured Reactor Coolant System flow rate is acceptable if the THERMAL POWER is also low) to ensure that the calculated DNBR will not be below the design DNBR value. The relationship defined on Figure 3.2-1 remains valid as long as the limits placed on the nuclear enthalpy rise hot channel factor, FAH(X,Y) in Specification 3.2.3 are maintained. The indicated T value and the indicated pressurizer pressure value correspond to analyticU limits of 594.8'F and 2205.3 psig respectively, with allowance for measurement uncertainty. When Reactor Coolant System flow rate is measured, no additional allowances are necessary prior to comparison with the limits of Figure 3.2-1 since a as:surement error of 2.1% for Reactor Coolant System total flow rate 4h as been allowed for in determination of thg design DNBR value. MAW 6HEAK V60EKQiq GREnrcKT1/MJ aK CUAL The measurement error for 1b n s vaul6 cTMT6b od f/buttC Receter4eelant-Sy'Aem-tot:1 ficw rate 3.uk is mased upon-parforMng--a-prac4s4ca-heat-balance-and-us4ng-the-result tc calibr:tc the
-Reaeter-Goolant-System Flew rate ind4c+ters. Potenth! fou-ling of the fccd= ,
7 water-venteri whieh-might-not-de-detected could-was the result < rem the precision haat balance ia 2-nanconservathe manner. Therefore, a penalty of 0.1% for undetected fouling of the feedwater venturi is included in Figure 3.2-
- 1. Any fouling which might bias the Reactor Coolant System flow rate measure-ment greater than 0.1% can be detected by monitoring and trending various plant performance parameters. If detected, action shall be taken before performing subsequent precision heat balance measurements, i.e., either the effect of the fouling shall be quantified and compensated for in the Reactor Coolant System flow rate measurement or the venturi shall be cleaned to eliminate the fouling.
The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their l limits following load changes and other expected transient operation. Indica-tion instrumentation measurement uncertainties are accounted for in the limits provided in Table 3.2-1. The measurement error for RCS total flow rate is based upon the performance I balances. Sets of elbow tap coefficients, as determined during these heat ba
! each elbow tap to provide a singie set of elbow tap coefficients for use in ca set of coefficients establishes the calibration of the RCSihflow r tap coefficients used for RCS flow measurement. Potential fouling of the not be detected, could bias the result from these heat balances in a non-conservative m l
CATAWBA - UNITS 1 & 2 B 3/4 2-4 Amendment No.) (Unit 1) Amendment No. 1 (Unit 2)
Attachment II 4
.1 Justification and Safety Analysis 1
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Technical Justification For Performing The RCS Flow Surveillance Ily Using Cold Leg Elbow Tap Flow Indications llackground The calculated Reactor Coolant System flowrates as detennined by the current Technical , Specification surveillance method have changed significantly over the past several fuel cycles at the McGuire and Catawba Nuclear Stations. These changes are not substantiated by the changes that have occurred in the system hydraulics, and are not confinned by other indications ofloop flow. These changes have on occasion resulted in closely approaching the Technical Specification minimum measured flow limit, with a minimum flow margin of as low as 0.1% having occurred at McGuiru Unit 2. This situation has resulted in Technical Specification changes to reduce the ; minimum measured flow and has impacted the themial margin and operating space in the reload .; designs. The current surveillance method calculates RCS flow based on steam generator thermal output from a calorimetric measurement, divided by the enthalpy difference acmss the reactor vessel as indicated by the hot and cold leg RTDs. Uncertainty in the hot leg temperature as indicated by the RTD has been identified as the main contributor to calculated decreases in RCS flow. Changes in core reload designs have resulted in core exit temperature distributions that, when combined with incomplete flow mixing and asymmetric flow pattems in the upper plenum, ' pmduce varying hot leg temperature indications. The net effect of these phenomena has resulted in what has been refermd to as hot leg streaming. The three hot leg RTDs are oriented approximately at 120 angles on the cmss section of the hot leg pipe. The RTDs can indicate different > temperatures in each loop, between loops, and can also change during the fuel cycle as the core power distribution changes. Due to the observed error in this method of flow surveillance and the ; consequences related to core thennal margin and operating space, an attemate method of j performing the flow surveillance using the cold leg elbow tap indication of flow is pmposed. The ; cibow taps are used for the Reactor Protection System monitoring ofloop flow and are a tme measurement of flow. The elbow taps were originally used for the Technical Specification flow surveillance at McGuire and other Westinghouse plants, but a change to the calorimetric based flow method was adopted with the intent of benefiting fmm supposed better accuracy. The l unanticipated impact of hot leg streaming has climinated the benefit of the calorimetric method. In j the proposed method, the existing historical calorimetric data is used to establish a calibration of _; the elbow taps, and then the future flow surveillance is performed by using the elbow tap flow indications. i l Justification and Safety Analysis Technical Specification 3/4.2.5.3 Irquires the detennination of the RCS flow rate by precision heat balance measurement at least once per 18 months. The precision heat balance measurement is used to calculate RCS flow based on gross steam generator thennal output and measurements of RCS ; hot and cold leg temperatures. Once the RCS flow is determined by this method, the cold leg , elbow tap flow coefficients are adjusted to repmduce the RCS flow indication determined by the calorimetric. Recently, the precision heat balance has been adversely affected by the hot leg _ 1 temperature streaming effect and has been the cause of unsubstantiated RCS flow measurement 1 decreases. Changes in the calorimetric result fmm changes in core exit temperature and/or flow distributions which can sigr.ificantly affect the average T-hot as measured by the hot leg RTDs. 1 1 I
u . Elbow tap APs and analytical predictions of flow do not confirm the RCS flow rates determined by the periodic calorimetrics. . H is proposed that TS 4.23.5 be changed to mmove the requirement for detennining the RCS flow rate by precision heat balance at least once per 18 months and replaced with the requirement to determine the RCS flow rate by measumment at least once per 18 months. His will allow RCS flow determination by elbow tap AP measurement. We elbow tap AP method of flow detennination was initially used at McGuim Nuclear Station and was dropped in favor of a precision calorimetric each cycle. At the time, November 1982, McGuire was having difficulty meeting the Technical Specification flow limit and the precision calorimetric was considered more accurate and allowed substantial margin gain. With the advent of more severe temperature and flow gradients in the hot legs, the precision calorimetrics were affected and resulted in nonphysical indicated flow decreases in the RCS, he elbow tap AP indications will provide an indication of. flow which is not dependent on future changes in hot leg temperature measurement and therefore will provide an improvement in flow surveillance accuracy. Le elbow taps, installed in a crossover leg cibow in each h>op and used to measure RCS flow, are. not calibrated prior to the installation of the elbow in the RCS. De RCS consists oflarge diameter piping which would make the calibration of the elbow taps in a laboratory environment expensive and difficult. Bis means the constant of proponionality (elbow tap coefficient or K value) for each tap must be detemlined by some other method. De flow indications from the AP transmitters are adjusted by setting the elbow tap coefficients to the values detennined by the flow calorimetric cach cycle. This nomialization effectively changes the elbow tap coefficients each cycle. It is proposed that the previously perfomied precision RCS flow calorimetrics be the basis for establishing final values for the elbow tap coefficients. Once these final values am established, the elbow tap coefficients will no longer be adjusted to a calorimetric, ne RCS flow surveillance will then be perfonned by using future elbow tap AP indications. Elbow Tao Flow Measurement Rencatability Elbow flow meters are a fomi of centrifugal meter, using the momentum forces developed by the change in flow direction. De principal parameters that detennine the AP for a specified flow are the radius of curvature of the elbow and the diameter of the flow channel through the elbow. Experiments on elbow meters have detennined that the flow measurements are not affected by differences in surface roughness and have a high degree of repeatability. Specific phenomena that have affected other types of flow meters, or that might affect the elbow meters in the RCS application, have been evaluated to detennine whether any of these phenomena would affect cibow meter repeatability. In addition, data from an operating plant equipped with a highly accurate flow meter has been compared with the elbow meter measurements at that plant to demonstrate elbow meter repeatability, l .. Venturi Fouling Venturi meters am affected by deposits, called fouling, that affect surface mughness and throat area. He fouling is apparently caused by an electrochemical ionization plating of copper and magnetite particles in the feedwater, a process associated with the large vehwity increase as the flow appmaches the venturi throat. This condition is not present in
- - ~ . . . - - - . . - _-
4 an RCS cibow; there is no large change in cmss section to produce a velocity increase and , ionization, and changes in surface roughness do riot affect the elbow flow measurement. .
- 2. Meter Dimensional Changes The elbow meter will not be susceptible to dimensional changes due to pressure and temperature since these temperatures and pressures would be approximately the same (full-power conditions) each time flow measurements are made. Erusion of the elbow surface is a unlikely since stainless steel is used, and the velocities are not large (42 fps) relative to vek>citics known to cause emsion. The effects of any dimensional change or of erosion !
could only affect flow by changing elbow radius or pipe diameter, both large relative to i any possible dimensional change. Therefore, the elbow meter is considered to be a highly stable flow measurement element.
- 3. Upstream Velocity Distribution Effects l
The vek) city distribution entering the elbow meter will be skewed by the upstream 40 cibow on the steam generator outlet nozzle, and the velocity distribution entering the steam generator outlet nozzle may be skewed due to i's off-centerlocation relative to the tube ; sheet. These geometric effects will remain constant through a fuel cycle, so the elbow ! meter AP would not change. r Steam genemfor tube plugging is usually randomly distributed acmss the tube sheet, so the vek) city distribution approaching the outlet nonle would not change. The velocity distribution could change if extensive tube plugging occurred in one k> cation on the tube sheet, but the change would not be transmitted thmugh the outlet nozzle to the elbow meter. The veh> city within the steam generator plenum is small (6 fps) compared to the i downstream cold leg pipe velocity. This large change in flow area would significantly ~ _ g' decrease or flatten any upstream velocity gradient. Therefore, any tube plugging, even if asyrmnetrically distributed, would not affect the elbow flow measurement repeatability.
- 4. Flow Measurement Comparisons Flow measurement comparisons have been conducted at Prairie Island Unit 2 which has e other highly n rurate means of RCS flow measurement. The Leading Edge Flow Meter (LEFM) isalled at this plant pmvided a means of confinning repeatability of cibow 1 meters. These comparisons were performed over a period which saw many significant ;
changes in system hydraulics and the cibow meter measurement changes were found to be ; in agreement with those of the LEFMs. , Ebpw Tap Flow Coefficient Determination Selected sets of cibow tap coefficients, as detennined during the calorimetric procedures, are averaged together for each elbow tap to pmvide a new cibow tap coefficient for each tap for use in calculating RCS flow. The set of coefficients detennined for each unit will then become the set of elbow tap coefficients used. The elbow tap coefficients chosen to be used for cach unit are given below. 'I
. Elbow Tap Coefficients .;
Taps McGuire Unit I McGuire Unit 2 Catawha Unit 1 Cataw ba Unit 2 i L lamp A, Tap 1 0.30695 03 0174 0.29773 03 0365 ^ l>>op A Tap II 0.29821 0.29183 0.29348 0.29183 , 12)op A, Tap III 0.30203 0.29781 0.29515 0.30020
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Loop II, Tap I 0.28441 0.29909 0.3N10 0.30021 ! Loop II, Tap 11 0.28409 0.29163 0.30803 0.28332 Loop 11, Tap III 0.28722 0.29173 0.30444 0.30258 : 1 Loop C, Tap I 0.28624 0.29155 0.28915 0.31370 ' l>>op C, Tap 11 0.31312 0.29399 0.28489 0.29362-12)op C. Tap III 0.29923 0.29250 0.29097 0.30 ISO , Loop D, Tap I 0.30704 0.30037 0.30331 0.29698 ; 12)op D, Tap II 0.29401 0.29755 0.29932 0.29685 ; IA>op D, Tap III 0.30174 0.29844 0.31051 0.29886 RCS Flow and Loss of Flow Setnoint Uncertainties f The effects of not perfi)rming a nomialization of the cold leg cibow taps for future cycles were evaluated. The evaluation is based on the premise that a final detemiination of the cold leg clbow tap coefficients is made based on previously performed precision RCS flow calorimetrics. This effectively establishes the flow coefficient for the elbow. For each future cycle, the Technical j Specification RCS flow is confirmed by the flow indicated by the cold leg cibow taps. { The RCS flow uncertainly was detennined based on the following intbrmation: ;
- 1) Installation specific instmmentation (Barton 764 transmitters at McGuire and Tobar/Veritrak transmitters at Catawba). j
- 2) Pn>cedural sensor calibration accuracy and vendor estimates of sensor and rack - I temperature and pressure effects.
- 3) 'The cold leg cibow tap AP transmitters am calibated, either on a bench or in place, such that it responds within the calibration tolemnce for a given AP input once per 18 months. 3
'I Typically, this would occur during a reload shutdown.
Based on the above, calculations were perfonned for the loss of flow reactor trip instrument uncenainty and for the RCS flow indication uncenainty using the plant process computer. The calculations, using the approved setpoint methodology which was submitted to the NRC on October 8,1981 lbr McGuire and July 30,1984 for Catawba, resulted in the fbliowing .i conclusions. I
, - - m . .
Duke Power Company has evaluated the RCS flow measurement uncenainty noted in Mc9uire Technical Specification 3/4.2.5 Figure 3.2-1, and has found that the value will ,; , increase fmm the present value to 1.89 flow (calculated value is 1.74% flow) plus 0.1% for feedwater venturi fouling. The present value is 1.7% flow plus 0.1% for feedwater venturi fouling penalty. An evaluation of the RCS flow measurement uncenainty noted in Catawba Technica' Specification 3/4.2.5, Figure 3.2-1, has found that the new value is well within the previously assumed 2.1% flow plus 0.1% penalty for feedwater venturi fouling. The previous calculated measurement uncenainty value was 1.83% flow. De new calculated measurement uticenainty value is 1.87% flow. The proposed new value to include in Catawba Technical Specification 3/4.2.5, Figure 3.2-1,is 1.9% flow plus 0.1% feedwater venturi fouling penalty. These calculations are based on indication via the plant process computer. Any indication process electmnically upstream of the process computer using a reasonably accurate DVM would result in a smaller indication error. Changes to the McGuire Technical SpeciDeations, Table 2.2-1, low reactor coolant flow setpoint and allowable value am necessary since the new channel statistical allowance (CSA) of 3.72% flow is larger than the current total allowance (TA) of 3.50% flow. A new TA of 4.50% flow is assumed which results in a new low reactor coolant flow setpoint of 91% of minimum measured flow per loop and an allowable value of 90% of minimum measumd flow perloop. Changes to the Catawba Technical Specifications, Table 2.2-1, low reactor coolant flow setpoint and allowable value are necessary since the new CSA of 3.37% flow is close to the current TA of 3.50% flow. %is results in a small difference between die setpoint and the allowable value. His may adversely impact future channel calibrations since nomial instmment drift may cause excessive instrument calibrations when the allowable value is j frequently exceeded. To provide more room between the selpoint mxt the allowable value l a new TA of 4.50% llow is assumed which results in a new low reactor coolant flow - l setpoint of 91% of minimum measured flow per loop and an allowable value of 89.7% of minimum measumd flow perloop. j Summary 1
-1 he pmposed Technical Specification changes will allow more flexibility and pmvide more q stability to the detennination of reactor coolant flow. %c cibow taps are currently used for the ~ .;
Reactor Protection System monitoring ofloop flow and are a true measurement of flow. The ; elbow tap AP indications will provide an indication of flow which is not dependent on future changes in hot leg temperature measurement and therefore will pmvide an improvement in flow - ;l surveillance accuracy. In addition, the reactor coolant llow measurements will be more predictable and stable since the flow is measured directly by the elbow meters and is not significantly affected by phenomena such as fouling, cibow dimensional changes, and vekicity distributions. De elbow taps were originally used for the Technical Specification flow surveillance at McGuire and other ) Westinghouse plants, but a change to the calorimetric based flow method was adopted with the :j intent of benefiting fmm supposed better accuracy. The unanticipated impact of hot Icg streammg - J
has climinated the benefit of the calorimetric method, in the pmposed method, the existing historical calorimetric data was used to establish a calibration of the elbow taps, and then the . future flow surveillance will be perfonned by using the elbow tap flow indications. ,
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, ATTACHMENT III ;
Analysis to Sunnort the Conclusion of No Sienificant Hazard l The following analysis, performed pursuant to 10 CFR 50.91, shows that the proposed amendment l will not create a significant hazards consideration as defined by the criteria of 10 CFR 50.92. I
- 1. This amendment will not significantly increase the probability or consequence of any accident previously evaluated.
- No component modification, system realignment, or change in operating procedure will f occur which could affect the probability of any accident or transient. The change in -l method of flow measurement will not change the probability of actuation of any Engineered Safeguard Feature or other device. The actual flow rate will not change. The consequences of previously-analyzed accidents will not change as a result of the new method of flow measurement. ;
I
- 2. This amendment will not create the possibility of any new or different accidents not previously '
evaluated. No component modification or system realignment will occur which could create the l' possibility of a new event not previously considered. The elbow taps are already in place, and are used to monitor flow for the Reactor Protection System. They will not initiate any new events. t
- 3. This amendment will not involve a significant reduction in a margin of safety.
As described in Attachment II, the change in method of RCS flow measurement will provide a ~i more accurate indication of the flow. The actual flow rate will not be affected. The a revised setpoints for low reactor coolant flow are driven by changes to statistical allowances and do not represent substantive, or less conservative, changes. There is no significant reduction in a margin of safety. ; i
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