ML20198E843
| ML20198E843 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 12/31/1996 |
| From: | WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20070L114 | List: |
| References | |
| NSA-SSO-96-525, NSA-SSO-96-525-R01, NSA-SSO-96-525-R1, NUDOCS 9801090233 | |
| Download: ML20198E843 (12) | |
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NSA SSO 96 525, Rev. I Westinghouse Proprietary Class 1 Ocember,1996 JOSEPH M. FARLEY NUCLEAR PLANT SAFETY ANALYSIS INTERMEDIATE RANGE NEUTF.ON FLUX REAC1'OR TRIP SETPOINT CHANGE Introduction The Intennediate Range Nuclear Instmmentation System (NIS) channels are designed to provide infonnation to the plant operators about the neutron flux and power in the reactor core when operating in the low to mid power range. The intermediate Range (IR) chalmels also provide a reactor trip signal on increasing power if the power level rires above the trip setpoint (typically 25% Rated Thermal Power). During controlled plant startup, the IR reactor trip is manually blocked above the P-10 permissive (typically 10% Rated Thennal Power), which is provided by the Power Range (PR)
NIS detectors. Although the safety analyses do not explicitly assume a trip from the IR NIS, it serves as a backup for the PR NIS trip at low power level, thus providing diversity in the reactor trip system (RTS), In addition, the Ik provides a control interlock function (C-1) and a permissive (P-6), The C-1 control interlock blocks control md withdrawal at powe. levels greater than 20% Rated Thennal Power (RTP),
unless manually bypassed above P-10. The function of P-6 is to allow a manual block of the Source Range reacto trip at 10* amps on increasing power and to automatically enable the Source Range reactor trip on decreasing power.
During routine plant operations, problems with the IR trip can arise because the outputs of the IR and PR detectors are subject te diffeient [
1+"'*. The IR channels are typically calibrated one time at the beginning of the cyc!. .vhile the PR channels are nonnalized to a seccadary side power calorimetric on daily basis, Over time, the large magnitudes of the
[ ]+"'* can pot;ntially lead to overlap of the l'-10 permissive (as measured by the PR detectors) and the IR reactor trip setpoint.
When such overlap occurs, routine plant evolution such as plant / reactor shutuwn must be deisyed until the NIS IR channels can be recalibrated. Should the operators
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fail to ooserve an overlap condition, then an inndvertant reactor trip would result. To alleviate these types of problems with this trip function it has been proposed that the Technical Specifications IR reactor trip setpoint be increased from 25% to 35% RTP to increase the operating margin beween PR P-10 and the IR reactor trip setpoint.
The tiip uncertainties (including the [ ]+"'*) were evaluated to support this Technical Specifications change. As part of this effort, Farley-specific data was gehered ard evaluated to better define the [ ]+"' . In Page 1 9001090233 971231 PDR ADOCIA 05000348 i p PDR '
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NSA.SsO.96 525. Rev. I December,1996 addition, all other terms in the IR setpoint calculation were revie,ved with respect to actual plant practices and equipment capabilities. This evaluation provides the basis for updating uncenainty terms in the IR setpoint uncertainty calculation. The results of those evaluations are described in this document and support raising the trip setpoint to 35% RTP.
Current IR Basis When computing the instrument uncertainties for the Farley 1 & 2 IR channels, the Westinghouse setpoint methodology (described in WCAP-13751, Rev. 0) initially
] total uneenainty for the [
assumed ] + "' a [ term. This value was broken down funher into [
]+"'*. The values assumed for IR trip setpoint allowances in the Parley Setpoint Study appear below.
__ _ + a.c The above effects were treated as statistically [ ] + "'*
parameters, and thus were combined using [
]+"'*. Therefore, the total combination is [ ]+"'*. As mentioned before, the assumption in WCAP-13751, Rev. O is [
Test and Data Analyris The IR detectors are approximately centered about the mid-plane of the core. Because of t'e relatively short length of the IR detectors (compared to the longer length PR detectors) the IR detectors are more sensitive to [
) + a.e ,
As a result, the IR channels can see a significant shift in indicated power over the cycle. In addition, since each of the IR detectors is located just a few assemblies away [
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NSA SSO-96 525. Rev.1 December 1996 i
]+ ***. That is, [
] + a,e .
\
In order to better our 'ify the [ ]+*'*, Farle(perfonned a test on Unit I to measure tw effect of changes [ ] ** on the NIS IR detector currents. Reactor power was [
) + a.cas test data was taken. The [
]+*'*. This test provided two sets of data; one for N35 and one for N36 [ ]+"'*. Since the
[
]+*** in the analysis. A test performed at the beginning of the previous cycle in Farley Unit I also provided data for the N35 and N36 detectors over [
]+"'*. Although the rods were not [+***
] , the results are very similar [
]+*** to the data analyzed and confirm the repeatibility of the effects described below.
Several different [
]+"'*. The resulting plots approximate [ ]+*** and demonstrate dramatically the impact of [ ]+*'*. The data shows that the N35 detector output can change from [
]+***. Based on Farley data, the sitivity of the N35 detector output to power at a [
]+"'*. A similar analysis was performed for the N36 data; however, this detector exhibits different characteristics with a variation with respect to [
]+"'*. It should be noted that differences between detector characteristics are not unusual and have been observed at other plants. In order to minimize the errors, and provide for a single calibration point for both detectors, it is recommended that the calibration be perfonned [
]+"'*, This results in a maximum nonconservative error of [
]+"'* for the N36 detector md [ ]+"'* for the N35 detector.
Therefore, the maximum [
]+*'*. The maximum [
) + a.c Since the N36 effect is more limiting, it will be used in the calculation of the setpoint uncertainty. In addition. based on a review of Farley data at 100% RTP [
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l
.4*
NSA SSO-96-525, Rev. I December,1996
]+"'* The Parley results are similar to those seen at other plants, and the
[ ]+"'* accounted for in the setpoint uncertainty is expected to be bounding for both Unit I and Unit 2, power uprating conditions, differences in '
individual detector characteristics, and future core design changes which do not
[ ] "'*.
In addition to [ ] + "'* , t he [ ]+"'" terms were also revised to better reflect plant operating conditions at the Farley Units. The [
]+"'*. The previous [
]+*'*. In addition, the data that was used to detennine the [
) + a,c, Use of the updated allowances (rather than an estimate based on engineering judgment) results it. a more accurrie PMA uncer'ainty calculation as noted below.
.- + n.c The equation for CSA for the Intermediate Range channels used in the Far.ey l Setpoint Study (WCAP-13751, Rev. 0) has the following form.
CSA = ((PMA)* + (PEA)* + (SCA + SMTE + SD)* + (SPE) + (STE)' +
(RCA + RMTE + RCSA + RD)* +(RTE)*}'" + EA Page 4
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l NSA.SSO-96 525. Rev. I December.1996 where:
CSA = Channel Statistical Allowance PMA = Process Measurement Accuracy PBA = Primary Element Accuracy SCA = Sensor Calibration Accuracy SMTE = Sensor Measurement and Test Equipment Accuracy SD = Sensor Drift SPE = Sensor Pressure Effects STE = Sensor Temperature Effects RCA = Rack Calibration Accuracy RhhE = Rack Measurement and Test Equipment Accuracy RCSA = Rack Comparator Setting Accuracy RD = Rack Drift RTE = Rack Temperature Effects EA = Environmental Allowance The value of CSA for Farley Units 1 & 2 in the current WCAP-13751, Rev. O is '
[ ]+*** for a span defined to be 0 to 120 %RTP.
For the detennination of a new uncertainty for the IR reactor trip setpoint, the latest Westinghouse CSA algorithm, which more accurately reflects plant operating pmetices has been used. This enhanced algorithm has the form shown below and is in concert with ISA 67.04 guidelines.
CSA = [
, + a ,c In addition, the IR setpoint calculation iu WCA'. -13751, Rev. O made certain assumptions about the IR channel racks which have been revised to better reflect the actual calibration and perfonnance capabilities of the equipment. Due to the fact that the IR channel is designed to provide information over a span of 8 decades and that -
readings are taken from a log meter, the channel tolerances and allowances must be relatively large in order to be achievable. This revised calculation has, therefore, included increased values for the rack calibration accuracy, the rack temperature effect, and the rack drift. A reference accuracy (RRA) has also been included to account for the [ ]+",
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4 0 -
(
NSA-sso-96-525. Rev. I December,1996 Using the new values determined above for [
]+"'* in conjunction with the enhanced uncenainty algorithm, the new CSA for the IR trip setpoint is [ ]+"'* for the 120% RTP span. Table I lists the various compoaents of the CSA calculation.
As discussed earlier, this trip is not explicit!y credited in any safety analysis; ,
therefore, there is no denned Safety Analysis Limit (SAL). With the nominal trip setpoint at 35% RTP and a CSA of [ ]+"'*, the trip is assured of actuating before reaching 60% RTP. This bounding value is sufGeient to provide a diverse backup (consistent with WCAP-12733, Protection System Diversity Design Bases) for the Power Range Low Setpoint which has a nominal trip setpoint of 25% ,
The Allowable Value (as speciGed in the Technical SpeciGcations) is associated with ancenainties in rack electronics. The intent of the A!!owable Value is to provide the plant with a way to assess the operability ot' the process racks. With an upper calibration limit of 35% RTP and a lower calibration limit of 29.5% RTP, an Allowable Value of 40% RTP will accommodate the [
]+"'*. The present Allowable Value in the Farley Unit 1 & 2 Technical Specifications is 30% RTP with a trip setpoint of 25% RTP. The new recommended Allowable Value is 40% RTP with a Nominal Trip Setpoint of 35 % RTP.
This analysis has no impact on the C-1 control interlock or the P-6 permissive, which are to remain at 20% RTP and 10' amps, respectively. Explicit uncertainty calculations are not traditionally performed for these functions since there are no SALs associated with irnerlocks and permissives which do not provide trips or ESF.
Conclurian The purpose of this analysis is to suppon the revision of the NIS IR nominal trip setpoint from a value of 25% RTP to 35% RTP. la additit.a this analysis suppons an Allowable Value of 40% RTP, k
The revision to the IR trip setpoint required alterations to the [
]+"'* in the CSA calculation. The value of the [ ]+"'* term is based on the assumption that the IR detectors are calibratedj [+a.e, Page 6
o e
o' 3 ,.
't NSA SSO-96-525. Rev. I December,1996 ,
The revised setpoint uncertainty calculation results, using Westinghouse methodology, justifies the proposed Technical Specincation changes to the NIS IR nominal trip setpoint and Allowable Value. Increasing the trip setpoint results in increased operating margins between NIS excore PR and IR : rip functions, which reduces the likelihood of inadveitant plant trips during plant shutdown. ,
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p T ~
NSA-SSO-96-525. Rew 1 December,1996 TABLE 1 INTERMEDIATE RANGE, NEUTRON FLUX Parameter- Allowance *
- Process Measurement Ar. Curacy _,,,,
m "
Primary Element Accuracy sensor calibration
[ ] ,,,
se sor Measurement & Test EquipmC3t ACCurac ,,,
sensor Pressure 'Iffects ;
sensor Temperature Effects
[. ) ....
sensor Drift t 3 .. .
Environmental Allowance Rack Calibration Rack Accuracy [ ]** '
Measurement & Test Equipment Accuracy Rack Reference Accuracy [ ]**
comparator [ ] **
- Rack Temperature Effects [ ]**
Rack Drift c 3 ....
In % span (defined to be 120% Rated Thermal Power)
Channel statistical Allowance =
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ATTACllMENT 111 Joseph M. Farley Nuclear Plant Technical Sp ci6 cations Change Request Intermediate Range Neutron Flux Reactor Trip Setpoint Signi6 cant flazards Evaluation f
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10 CFR 50.92 SIGNIFICANT IIAZARDS EVALUATION ,
Joseph M. Farley Nuclear Plant lat: mediate Range Neutron Flux Reactor Trio Setpoint Chaneg I
As required by 10 CFR 50.91 (a)(1), an analysis is provided to demonstrate that the proposed license amendment to increase the Intermediate Range neutron ilux reactor trip setpoint from 25%
RTP to 35% RTP involves no significant hazards masideration.
Proposed Change The proposed change to the Farley Units 1 and 2 Techni:al Specification Taole 2.2-1 involves a change in the Intermediate Range neut on flux trip setpoint from 's 25% RTP" to "S 35% RTP" and a change in the corresponding allowable value from "s 30% RTP" to "s 40% RTP." The Bases to Technical Specification 2.2.1 has been char.ged to reflect the setpoint ci,av (trip setpoint changed fron 25% to 35%). It also deletes the reference to the reactor trip setpoints m T. S.
3.10.3, "Special Test Exceptions-Physics Tests," and T. S. 3.10.4, "Special Test Exceptions-Reactor Coolant Loops," tn be consistent with the Westinghouse Standard Technical Specifications (NUREG 1431, Revision 1) fomut, since this redundant reference merely restates the analyzed setpoints centained in Specification 2.2.1, " Reactor Trip System Instrumentation Setpoints."
Backgrard Plant operational problems associated with the Intermediate Range (IR) reactor trip can arise (and have occurred at Farley) because the outputs of the IR and Power Range (PR) detectors are subject to different and relatively large uncertainty terms. The IR channels are typcally calibrated one time at the beginning of the cycle, while the PR channels are normalized to a secondary side power calorimetric on a daily basis. The large magnitudes of the uncertainties can potentially lead to overlap of the P-10 permissive (as measured by the PR detectors) and the IR reactor trip setpoint, which can cause an inadvertent trip or result in operational delays during a controlled shutdown.
To alleviate some of the operational problems with this diverse trip function it is proposed that the IR neutron fi 'x reactor trip setpoint be increased to 35%.
Analysis
'Ihe Intermediate Range Nuclest Instrumentation System (NIS) channels are designed to provide information about the neutron flux and power in the reactor core when operating in the to,v to mid power range, The Intermediate Range channels also provide a reactor trip signal on increasing power if the power level rises above the trip setpoint before it can be manua'ly blocked above the P-10 permissive setpoint (which is provided by the Power Range NIS detectors). This trip is a backup for the PR neutron flux low setpoint (25% RTP) reactor trip. The IR trip does not piaide primary protective action for any accident scenario. Although the safety analyses do nct explicitly assume a trip from the Intermediate Range NIS, its function as a backup for the Power Range NIS trip at low power level provides diversity in the protection system.
q-
.W
- Significant liazards Evaluation Page 2 An evaluation of the uncertainties associated with the IR channels was performed using Farley data, ne Process Measurement Accuracy (PMA) uncertainty allowances were adjusted based on the Farley-specific data, and the rack drift uncertainty a!!owance was adjusted to reflect a more realistic value based on equipment performance. Using the Westinghouse setpoint methodology, it was concluded that with a nommal IR reactor trip setpoint of 35% RTP the trip will ocur at an actual power no higher than 60%
RTP This bounding value is sufficient to assure that the IR trip will perform its intended function as a 3 diverse backup for the Power Range Neutron Flux Low reactor trip setpoint.
The analysis performed to support the change of the Intermediate Range reactor trip setpoint from 25%
RTP to 35% RTP is also applicable to the special test exceptions of Specifications 3.10.3 and 3.10.4.
He value of 35% RTP for the Intermediate Range trip setpoint is applicable to these specifications as well since they merely restate the analyzed setpoints cont:.ined in Specification 2.2.1. No separate safety an:. lysis or evaluatica for these special test exceptions is reqmred. Since the IR high flux and PR high flux low trip setpoints referenced in the special test exceptions specifications are redundant to the information contr.ined in Specification 2.2.1, the references to the trip setpoints may be deleted to be consistent with the improved Standard Technical Specifications (NUREG 1431, Resision 1) format.
The proposed change will increase the operating margin between the P 10 setpoint (10% RTP) and the IR reactor trip setpoint. This increased operating margin will minimize delays during plant shutdown and reduce the likelihoW ofinadvertent plarc. trips when operrting at low power.
- 50.92 Evaluation Based on the information presented above, the following conclusions can be reached with respect to 10 CFR 50.92.
- 1) He proposed change in Intermediate Range reactor trip setpoint from 25% RTP to 35% RTP, the associated allov.able value cha,ge, and the deletion of the redundant references to the IR high flux and PR high flux low setpoints do nu mvolve a significant increase in the probability or consequences of an accident previously evaluated in the Farley FSAR. He IR reactor trip neither causes any accident nor provides primary protection for any accident in the Farley FSAR. No new accident initiators have been identified because of this proposed revision. No new performance requirements for any system that is used to mitigate dose consequences have been ;
imposed by this proposed change. No input assumption to any dose consequence calculation is affected by this proposed change. - All previously repor ui dose consequences remain bounding.
Herefore, the radiological consequences to the public resulting from any accident presiously evaluated in the F3AR have not significantly increased.
- 2) The proposed Technical Specifications change to the IR reactor trip setpoint, associated allowable value change, and the deletion of the redundant references to the IR high flux and PR high flux low setpoints do not create the possibility of a new or different kind of accident from any previously evaluated in the FSAR.1Jo new accident scenarios, failure mechanisms or limiting single failures are introduced as a result of the increase in IR setpoint from 25% RTP to 35%
l RTP. No new challenges to the safety-related Reacter Trip System have been identified. The NIS hardware has not been modified, and Farley will continue to perform periodic IR channel j calibration and surveillance in accordance with I
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d' Significant flazards Evaluation Page 3 Technical Specifications. - All previous!';identified accident sec.urios remain bounding since the IR trip setpoint provides no primary accident protection. Therciore, the possibility of a new or -
different kind of accident is not created.
- 3) De proposed increase in the IR reactor trip setpoint from 25% RTP to 35% RTP, the associated allowable value change, and the deletion of the redundant references to the IR high flux and PR high flux low setpoir$ts do not involve a significant reduction in the margin of safety. All previously established acceptance limits continue to be met for all events, since the IR trip does not provide any prunary protective action for any act:ident scenario. Changing the IR setpoint and allownb!c value will not invalidate its beckup function. Here are no physical modifications required for the protection system. This change will not affect the operation of any other safetv-related equipment. Farley-specific setpoint uncertainty calculations support the setpoint change.
Since all acceptance limits continue to be met, there is no significant reduction in the margin of saf:ty.
Conclusion Based upon the pieceding evaluation, it has been deterndned that the proposed changes to the Farley Technical Specifications for increasing the NIS Intermediate Range neutron flux reactor trip etpoint and allowable value and the deletion of the redundant references to the IR high flux and PR high flux low setpoints in the Special Test Exceptions does not involve a significant hazards consideration as defined in 10 CFR 50.92 (c).
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