ML20117G740
| ML20117G740 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 08/29/1996 |
| From: | ENTERGY OPERATIONS, INC. |
| To: | |
| Shared Package | |
| ML20117G739 | List: |
| References | |
| NUDOCS 9609050307 | |
| Download: ML20117G740 (21) | |
Text
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Reactivity Anomalies
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3.1.2 i
1 3.1 REACTIVITY CONTROL SYSTEMS l 3.1.2 Reactivity Anomalies i
( LC0 3.1.2 'S = ti.ity di'Y;r:nc; bat ::n th ;; niter:d r;d den;ity and the p.edicted red deniit, shall b; ithir _ P.' f k/k. $(
l I APPLICABILITY: MODES 1 and 2.
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ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Core reactivity A.1 Restore core 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> l
difference not within reactivity difference limit. to within limit.
I B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time not met.
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RIVER BEND 3.1-4 Amendment No. 81
Reactivity Anomalies 3.1.2 SURVEILLANCE REQUIREMENTS i SURVEILLANCE FREQUENCY Ac,W^L SR 3.1.2.1 Verify [corereactivitydifferencebetween Once within the M iter:d red der.;ity and the predicted 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after 1% ok/k. reaching redder.;i'h"iswithin2
$64cfNi equilibrium conditions i following startup after fuel movement within the reactor pressure vessel or control rod replacement AND 1000 MWD /T thereafter during operation in MODE 1 i
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i RIVER BEND 3.1-5 Amendment No. 81 4
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I Reactivity Anomalies B 3.1.2
-B-3.1 REACTIVITY CONTROL SYSTEMS B'3.1.2 .Peactivity' Anomalies l
l BASES BACKGROUND In accordance with GDC 26, GDC 28, and GDC 29 (Ref. 11, reactivity shall be controllable such that suberiticality is maintained under cold conditions and acceptable fuel design limits are not exceeded during normal operation and anticipated operational occurrences. Reactivity anomaly is used as a measure of the predicted versus monitoredmeasured l core reactivity during power. operation. The continual confirmation of core reactivity is necessary to ensure that the Design Basis Accident (DBA) and transient safety analyses. remain valid. A large reactivity anomaly could be the result of unanticipated changes in fuel reactivity, control rod worth, or operation at conditions not consistent with those assumed in the predictions of core reactivity,.
and could potentially result in a loss of SDM or violation of acceptable fuel design limits. Comparing predicted versus monitoredmeasured core reactivity validates the l l nuclear methods used in the safety analysis and supports the SDM demonstrations (LCO 3.1.1, " SHUTDOWN MARGIN (SDM)") in ensuring the reactor can be brought safely to cold, suberitical conditions.
When the reactor core is critical or in normal power operation, a reactivity balance exists and the net reactivity is zero. A comparison of predicted and i
monitored::::ured reactivity is convenient under such a j balance, since parameters are being maintained relatively stable under steady state power conditions. The positive reactivity inherent in the core design is balanced by the negative reactivity of the control components, thermal i feedback, neutron, leakage, and materials in the core that absorb neutrons, such as burnable absorbers, producing zero net reactivity.
In order to achieve the required fuel cycle. energy output, l the uranium enrichment in the new fuel loading and the fuel l loaded in the previous cycles provide excess positive i reactivity beyond that required to sustain steady' state j operation at the beginning of cycle (BOC). When the reactor l (continued)
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RIVER BEND B 3.1-14 Revision No. 0l 1 1
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Reactivity Anomalies B 3.1.2 l
! BASES l
BACKGROUND is critical at RTP, the excess positive reactivity is (continued) compensated by burnable absorbers (if any), control rods, and whatever neutron poisons (mainly xenon and samarium) are l present in the fuel.
The predicted core reactivity, :: ::p ::ented by ;;nt cl cd
' den sit y-4the-numbee-o f-cont rol-red--not ches-i-nser ted-e s-a fracti:n of the total numb : cf ;;nt::1 ::d n:tch::; :Et rods-fully-inserted-is-equivalent--to-4004-rod-densit-y+ r- is calculated by a 3D core simulator code as a function of cycle exposure. This calculation is performed for projected operating states and-conditions throughout the cycle. The primary parameter used to measure reactivity anomaly is k-eff (as calculated by the core monitoring system). Other parameters including, but not limited to, power level during coastdown and delta control rod notches can also be used to measure the reactivity anomaly. -The core reactivity is determined by comparison of the reactivity f :: ;;nt:cl ::d densitsee-for actual plant conditione and ke-then ;;mpseed de-the predicted reactivityveive for the cycle exposure.
APPLICABLE Accurate prediction of core reactivity is either an explicit SAFETY ANALYSES or implicit assumption in the accident analysis evaluations (Ref. 2). In particular, SDM and reactivity transients, such as control rod withdrawal accidents or rod drop accidents,' are very sensitive to accurate prediction of core reactivity. These accident analysis evaluations rely on computer codes that have been qualified against available
- test data, operating plant data, and analytical benchmarks.
i Monitoring reactivity anomaly provides additional assurar.ce that the nuclear methods provide an accurate representation {
t of the core reactivity. j l
The comparison between monitoredme::ured and predicted l )
l initial core reactivity provides a normalization for the '
! calculational models used to predict core reactivity. If ,
l the monitoredmeasured and predicted reactivityrod-density l l for identical core conditions at BOC do not reasonably agree, then the assumptions used in the reload cycle design i j analysis or the calculation models used to predict reactivityrod-density may not be accurate. If reasonable agreement-between monitoredmeccured and predicted core reactivity exists at BOC, then the prediction may be normalized to the monitoredme: ur:d value. Thereafter, any significant deviations in the monitored reactivitymeasueed
j Reactivity anomalies satisfy Criterion 2 of the NRC Policy Statement.
- (continued)
RIVER BEND B 3.1-2a Revision No. 1 l
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Reactivity Anomalies (
B 3.1.2 !
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l BASES (continued)
! LCO The reactivity anomaly limit is established to ensure plant operation is maintained within the assumptions l of the safety analyses. Large differences between monitored and predicted core reactivity may indicate I that the assumptions of the DBA and transient analyses are no longer valid, or that the uncertainties in the ,
l Nuclear Design Methodology are larger than expected. 1 A limit on the difference between the monitored reactivitymonitored-rod-density and the predicted l reactivitypredicted-rod-density of 1% Ak/k has been l established based on engineering judgment. A > 1%
deviation in reactivity from that predicted is larger )
than expected for normal operation and should '
therefore be evaluated.
i APPLICABILITY In MODE 1, most of the control rods are withdrawn and l steady state operation is typically achieved. Under I these conditions, the comparison between predicted and monitored core reactivity provides an effective measure of the reactivity anomaly. In MODE 2, control rods are typically being withdrawn during a startup.
In MODES 3 and 4, all control rods are fully inserted, and, therefore, the reactor is in the least reactive state, where monitoring core reactivity is not necessary. In MODE 5, fuel loading results in a 1
continually changing core reactivity. SDM I requirements (LCO 3.1.1) ensure that fuel movements are performed within the bounds of the safety analysis, and an SDM demonstration is required during the first startup following operations that could have altered core reactivity (e.g., fuel movement, control rod replacement, control rod shuffling). The SDM test, required by LCO 3.1.1, provides a direct comparison of the predicted and monitored core reactivity at cold conditions; therefore, reactivity anomaly is not required during these conditions.
ACTIONS A.1 Should an anomaly develop between monitoredmec ured l and predicted core reactivity, the core reactivity difference must be restored to within the limit to ensure continued operation is within the core design assumptions. Restoration to within the limit could be performed by an evaluation of the core design and safety analysis to determine the reason for the anomaly. This evaluation normally reviews the core l
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RIVER BEND B 3.1-33 Revision.jo. 0 l
1
Reactivity Anomalies
, . B 3.1.2 BASES ACTIONS A.1 (continued) conditions to determine their consistency with input to design calculations. Measured core and process parameters are also normally evaluated to determine that they are within the bounds of the safety analysis, and safety analysis calculational models may be reviewed to verify that they are adequate for representation of the core conditions. The required completion Time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is based on the low probability of a DBA during this period, and allows sufficient time to assess the physical condition of ,
the reactor and complete the evaluation of the core design and safety analysis.
B.1 If the core reactivity cannot be restored to within the 1% Ak/k limit, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.1.2.1 REQUIREMENTS Verifying the reactivity difference between the monitored and predicted reactivityeed-density is l within the limits of the LCO provides further assurance that plant operation is maintained within :
the assumptions of the DBA and transient analyses. l The Core Monitoring System calculates the reactivityr:d den:ity for the reactor conditions l obtained from plant instrumentation. A comparison of the monitored reactivity::d den:ity to the predicted reactivityrod-density at the same cycle exposure is used to calculate the reactivity difference. The comparison is required when the core reactivity has potentially changed by a significant amount. This may occur following a refueling in which new fuel assemblies are loaded, fuel assemblies are shuffled l within the core, or control rods are replaced or ,
shuffled. Control rod replacement refers to the !
decoupling and removal of a control rod from a core location, and subsequent replacement with a new l
, control rod or a control rod from another. core l L location. Also, core (continued) l l
I RIVER BEND B 3.1-44 Revision $o. 0 1
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Reactivity Anomalies B 3.1.2 BASES i
SURVEILLANCE SR 3.1.2.1 (continued) l REQUI REMENTS reactivity changes during the cycle. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> interval after reaching equilibrium conditions following a startup is based on the need for equilibrium xenon concentrations in the core, such that an accurate comparison between the monitored and ;
i predicted reactivityrod-density values can be made. l l For the purposes of this SR, the reactor is assumed to be at equilibrium conditions when steady state operations (no control rod movement or core flow changes) are required to maintain power constant (above 80% RTP) for 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.at : Bo 'RTP-have-been obtained. The 1000 MWD /T Frequency was develeped, considering the relatively slow change in core reactivity with exposure and operating experience related to variations in core reactivity. This comparison requires the core to be operating at power levels which minindze the uncertainties and measurement errors, in order to obtain meaningful ,
results. Therefore, the comparison is only done when i in MODE 1.
REFERENCES 1. 10 CFR 50, Appendix A, GDC 26, GDC 28, and i GDC 29. '
- 2. USAR, Chapter 15.
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l RIVER BEND B 3.1-54 Revision $o. O
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APRM Gain and Setpoints l
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( 3.2 POWER DISTRIBUTION LIMITS 3.2.: 'veng- %=r nange "er. iter ( AP'M)
A G in .nd se ipv , c,t 3 LCO 3.2.4 a. T shall be a 1.0; or j
- b. Each required APRM setpoint specified in the R shall !
be made applicable; or ;
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- c. Each required APRM gain shall be adju ed such that the !
adjusted APRM readings result in a culated T a 1.0 ;
when the APRM reading is substitu d for FRTP. ,
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APPLICABILITY: THERMAL POWER = 25% RTP.
ACTIONS CONDITION QUIRED ACTION COMPLETION TIME A. Requirements of the .1 Satisfy the 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> LCO not met. requirements of the LCO.
B. Required Act n and B.1 Reduce THERMAL POWER 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> associated ompletion to < 2S% RTP.
Time not t.
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1 RIVER BEND 3.2-4 Amendment No. 81 <
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APRM Gain and Setpoints 3.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENC 1
i l SR 3.2.4.1 ----------
NOTE--------- --.-------
Not required to be met if SR 3.2.4.2 is satisfied for LC0 3.2.4, Item b or c requirements.
l l Verify T is a 1.0. Once within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after 1 2 25% RTP !
AND
! 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> i thereafter l
l SR 3.2.4.2 ---------------
l fNOTE--------------------
l Not required tg be met if SR 3.2.4.1 is l satisfied f# LCO 3.2.4, Item a
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VerJfy APRM setpoints or gains are adjusted 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for the calculated T.
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l RIVER BEND 3.2-5 Amendment No. 81
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Avorago Power Pengo Monitor
'APRM) Red bloct Gain and SGtpotnts t
TR 3.2.4 ;
.TR 3.2 PCWER DISTRIBUTICN LIMITS !
in ;.I.4--Au=*=;= #-u r ?:n;e '4:n _'
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and ::p.tage - i TLC 3 3.0.4 a.
T shall be 2 1.0; er / '
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l Each required APRM setpoint {
be made applicables or specified in the CCLR shaki
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EachrequiredAPRMgainshallbeadjustedsuchpKIttne adjusted APRM readings result when the APRM reading is substituted for ER*p.in a calculated,t 1
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APPLICABILITY:
1 THERMAL PCWER 2 25% RTP.
ACTICNS j
CCNDITICN REQUIRED AQIION CCMPLETICN TIME A. Requirements of the A.1 Satis f' the requirements TLCO not met. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of the TLCO. 1 l.
S. Required Action and i
associated Completic
/I1ReduceTHERMALPOWERto 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> l
Time not met. < 25% RTP.
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I RIVER. BEND TR 3.2-1 Revision 5 i (51)
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Avorago Pow 3r Pengo Mcntter fAPRM)
Rod block ain and 30tpoints SURVEILLANCE PEQUIREMDITS TR 3.2.4 SURVEILLANCE _
FREOUENCYy/
TSR 3.2.4.1
NCTE--------------------
Not required to be met if TSR 3.2.4.2 is satisfied for TLCO 3.2.4, Item b or e requirements.
i Verify T is 2 1.0. l Once within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after 2 25% RTP AND 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter TSR 3.2.4.2 -----------------NOTE---
Not ---------------
required to be met '. TSR 3.2.4.1 is satisfied for TLCO 3.' 4, Item a requirements.
Verify APRM set for the calcu cedints T.*
or gains are adjusted 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
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- With T < 1.0, /
rather/than adjusting the APRM setpoints, when the APRM reading is substituted for FRTP, provided that the adjusted APRM reading does not/ exceed 100% of RATED THERMAL PCWER, ad3ustment is 76sted on the reactor control panel. and a notice of the
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RIVER BEND TR 3.2-2 Revision 5
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APRM Gain and Setpoints B 3.2.4 B 3.2 POWER DISTRIBUTION LIMITS B-3.2.UAver sp Pwer hnge Meritor (APRM) Go g, and ntgin;3 7
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BASES
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BACKGROUND The OPERABILITY of the APRMs and their setpoints is an /
initial condition of all safety analyses that assume rod insertion upon reactor scram. Applicable GDCs are GOC 10
" Reactor Design"; GDC 13, " Instrumentation and Control";
GDC 20, " Protection System Functions"; and GDC E9,
" Protection against Anticipated Operation Occ #rences" (Ref. 1). This LC0 is provided to require ths APRM gain or APRM flow biased scram setpoints to be adjusted when operating under conditions of excessive power peaking to maintain acceptable margin to the fuel A'ladding integrity Safety Limit (SL) and the fuel cladd Tg 1% plastic strain limit.
The condition of high power peak ng is determined by the ratio of the actual power pea 'ng to the limiting power peaking at Rated Thermal Po (RTP). This ratio is presented as a APRM setpo s T-factor (T). At any off-rated power conditions T s equal to the Fraction of Rated Thermal Power (FRTP) d ided by the Core Maximum Fraction of Limiting Power Densit (CMFLPD). FRTP is the measured thermal power divid by the RTP. CMFLPD is the limiting linear heat gener ion rate (LHGR) divided by the rated LHGR limit. High pow ' peaking exists when:
FRTP CMFLPD As pow is reduced with the design power distribution main ined, CMFLPD is reduced in proportion to the reduction in wer. However, if power peaking increases above the d ign value, the CHFLPD is not reduced in proportion to the eduction in power. Under these conditions, the APRM gains are adjusted upward or the APRM flow biased scraa setpoints are reduced accordingly. When the reactor is optrating with peaking less than the design value, it is not necessary to modify the APRM gains or flow biased scram trip setpoints.
Adjusting the APRM gains or setpoints is equivalent to maintaining CHFLPD less than or equal to FRTP.
DR ETE (continuedl RIVER BEND B 3.2-12 'evision No. O
APRM Gain and Setpoints
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B 3.2.4 BASES l
1 BACKGROUND The normally selected APRM setpoints position the scram (continued) above the upper bound of the normal power / flow operatin region that has been considered in the design of the f el rods. The setpoints are flow biased with a slope th approximates the upper flow control line, such tha an approximately constant margin is maintained betw n the flow biased trip level and the upper operating bound y for core flows in excess of about 45% of rated core fl . In the range of infrequent operations below 45% of ated core flow, the margin to scram or rod blocks is reduc because of the nonlinear core flow versus drive flow re tionship. The normally selected APRM setpoints are s ported by the analyses presented in References 1, 2 and 3 that concentrate on events initiated fro rated conditions.
Design experience has shown that imum deviations occur within expected margins to opera ng limits (APLHGR and MCPR), at rated conditions for ormal power distributions.
However, at other than rated nditions, control rod patterns can be establishe& hat significantly reduce the margin to thermal limits. herefore, the flow biased APRM scram setpoints may be r uced during operation wher, the combination of THERMAL OWER and MFLPD indicates an excessive power peak g distribution.
The APRM neutron ux signal is also adjusted to more closely follow e fuel cladding heat flux during power transients. T APRM neutron flux signal is a measure of
- the core the. al power during steady state operation. ,
During pow transients, the APRM signal leads the actual l core the 1 power response because of the fuel thermal time constan . Therefore, on power increase transients, the APRM signa provides a conservatively high measure of core the 1 power. By passing the APRM signal through an el tronic filter with a time constant less than, but proximately equal to, that of the fuel thermal time constant, an APRM transient response that more closely follows actual fuel cladding heat flux is obtained, while a conservative margin is maintained. The delayed response of the filtered APRM signal allows the flow biased APRM scram
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/ levels to be positioned closer to the upper bound of the normal power and flow range, without unnecessarily causing reactor scrams during short duration neutron flux spikes.
These spikes can be caused by insignificant transients such as performance of main steam line valve surveillances or ,
p momentary flow increases of only several percent. j (coatiaued)
PE 4 7 e-1 RIVER BEND B 3.2-13 Revision No. 0 l
4 APRM Gain and Setpoints B 3.2.4 BASES (continued)
J APPLICABLE The acceptance criteria for the APRM gain or setpoint SAFETY ANALYSES adjustments are that acceptable margins (to APLHGR and PR) be maintained to the fuel claddir.g integrity SL and t fuel cladding 1% plastic strain limit .
USAR safety analyses (Refs. 2 and 3) concentrate n the rated power condition for which the minimum ex cted margin to the operating limits (APLHGR and MCPR) occ s.
LC0 3.2.1, " AVERAGE PLANAR LINEAR HEAT GENE ION RATE (APLHGR)," and LCO 3.2.2, " MINIMUM CRITIC POWER RATIO (MCPR)," limit the initial margins to th e operating limits at rated conditions so that specified ceptable fuel design limits are met during transients init' ted from rated conditions. At initial power level less than rated levels, the margin degradation of either t APLHGR or the MCPR during a transient can be greate than at the rated condition event. This greater. argin degradation during the transient is primarily offset y the larger initial margin to limits at the lower than ated power levels. However, power distributions can b ypothesized that would result in reduced margins to the transient operating limit. When combined with the iner sed severity of certain transients at other than rated c ditions, the SLs could be approached.
At substantially r 'ced power levels, highly peaked power distributions cou be obtained that could reduce thermal margins to the imum levels required for transient events.
To prevent or tigate such situations, either the APRM gain is adjusted ard or the flow biased APRM scram level is required to e reduced. Either of these adjustments effective" counters the increased severity of some events at other ;ha'n rated conditions by proportionally increasing the AP gain or proportionally lowering the flow biased APRM cram setpoints dependent on the incre.ased peaking that may e encountered.
e APRM gain and setpoints satisfy Criteria 2 and 3 of the NRC Policy Statement.
LCO Meeting any one of the following conditions ensures acceptable operating margins for events described above:
- a. Limiting excess power peaking;
,, (continued) lMU7~E RIVER BEND B 3.2-14 Revision No. 0
APRM Gain and Setpoints B 3.2.4 BASES LC0 b. Reducing the APRM flow biased neutron flux upscale (continued) scram setpoints by multiplying the APRM setpoints b the ratio of FRTP and the core limiting value of MFLPD; or
- c. With T < 1.0, rather than adjusting the APRM setpoints, the APRM gains may be adjusted s that the adjusted APRM readings result in a cal .ated T a 1.0 when the APRM reading is substit ed for FRTP. This Condition is to account for e reduction in margin to the fuel cladding integri SL and the fuel cladding 1% plastic strain limi .
In compliance with NRC Generic Letter 02 (Ref. 4), RBS has implemented the BWROG recommendat' s concerning the improved BWR stability Interim Corr ve Actions (ICAs).
The BWROG recommendations include nges to the required administrative operating domain r s rictions defined by the BWROG ICA stability regions.
The BWROG guidelines for the i lementation of the core-average boiling boundarf c t ol require a boiling boundary of 4.0 feet be maintained provide a high degree of stability margin. The r -average boiling boundary controls are implement as an administrative operating limit, FCBB (fractic core boiling boundary), such that FCBB is 1 1 for a c -average boiling boundary of 4.0 feet or higher. Core- rage boiling boundary is the axial elevation of the . 'ransition from sub-cooled to saturated fluid conditio on a core-average basis, and is a function of the core-a erage axial power shape, the core power and flow, and t core inlet subcooling.
To meet e FCBB controls during reactor startup, core-averag axial power distributions which deposit a relatively high action of power into the lower third of the core must be oided. Such power distributions are associated with hi er peaking at higher elevations in the core. However, imizing APRM setpoints T-factor so that APRM trip setdown
's not necessary requires relatively flat core-average axial
/ These flat axial power distributions
/ power result distributions.
in a low core-average boiling boundary, which is
/ associated with increased susceptibility to instability.
Therefore, the requirements of APRM setpoints T-factor
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APRM Goin and Setpoints B 3.2.4 BASES LC0
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result in reduced stability margin. Under these conditi (continued) FCBB stability control is not possible. Furthermore, p 'er distributions necessary for FCe3 control during reactor startup before recirculation pump upshift lead to APR trip setdowns that result in a control rod block. Under ese conditions reactor startup cannot be continued. '
The resolution to the conflict between complyin th the requirements of APRM setpoints T-factor and t FCBS stability control is to use two definitions APRM setpoints T-factor.
The use of two APRM se oints T-factor definitions provides additional operating rgin to achieve power distributions which allow FCBS st i ity control, without compromising LHGR protection a f-rated operatiors. l 3 x FRTP + 1 T p e x
. + 0.4, 4 x CMFLPD otherwise T=
FRTP
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CMFLPD f T is applied only i '< 1.0.
MFLPD is the rat of the limiting LHGR to the LHGR limit for the specif- bundle type. As power is reduced, if the design power, stribution is maintained, MFLPD is reduced in proportion the reduction in power. However, if power peaking in ' eases above the design value .the MFLPD is not reduced proportion to the reduction in power. Under these e ditions, the APRM gain is adjusted upward or the APRM f ow biased scram setpoints are reduced accordingly.
When he reactor is operating with peaking less than the de) gn value, it is not necessary to modify the APRM flow b sed scram setpoints. Adjusting the APRM gain or etpoints is equivalent to maintaining T = 1.0, as stated ;
in the LCO. i For compliance with LCO Item b (APPM setpoint adjustment) or Item c (APRM gain adjustment), oaly APRMs required to be (continued)
DEUTE RIVER BEND B 3.2-16 Revision No. O
4 APRM Goin and Setpoi "s
, , B 3. 4 BASES LC0 OPERABLE per LC0 3.3.1.1, " Reactor Protection Syst m (RPS)
(continued) Instrumentation," are required to be adjusted. addition, each APRM may be allowed to have its gain or se oints adjusted independently of other APRMs that are aving their gain or setpoints adjusted.
APPLICABILITY The T-factor limit, APRM gain adjustment, r APRM' flow biased scram and associated setdowns are rovided to ensure that the fuel cladding integrity SL and he fuel cladding 1%
plastic strain limit are not violated ring design basis transients. As discussed in the Base for LCO 3.2.1 and LCO 3.2.2 sufficient margin to these imits exists below 25% RTP and, therefore, these requi ements are only necessary when the plant is operat ng at a 25% RTP.
ACTIONS A.1 If the APRM gain or setpoin are not within limits while T < 1.0, the margin to the uel cladding integrity SL and the fuel cladding 1% plas c strain limit may be reduced.
Therefore, prompt action should be taken to restore T to within its required lim t or make acceptable APRM adjustments such that he plant is operating within the assumed margin of th safety analyses.
The 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Complet n Time is normally sufficient to restore either the MFLPD o within limits or the APRM gain or setpoints to wi in limits and is acceptabic cased on the low probabilit of a transient or Design Basis Accident occurring si ltaneously with the LCO not met.
E :.1 If t APRM gain or setpoints cannot be restored to within the r required limits within the associated Completion Time, h'e plant must be brought to a MODE or other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to < 25% RTP within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The allowed Completion Time is reasonable, based on operating experience, to reduce THERMAL POWER to
< 25% RTP in an orderly manner and without challenging plant systems.
(continued)
RIVER BEND B 3.2-17 Revision No. O
APRM Gain and Setpoints B 3.2.4 BASES (continued)
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SURVEILLANCE SR 3.2.4.1 and SR 3.2.4.2 REQUIREMENTS T is required to be calculated and compared to APRM dnor setpoints to ensure that the. reactor is operating w' thin the assumptions of the safety analysis. These SRs ar required only to determine T and the appropriate gain or tpoint, and is not intended to be a CHANNEL FUNCTIONAL ST for the APRM gain or flow biased neutron flux scram ci cuitry. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency of SR 3.2.4.1 is chosen to cincide with the determination of other thermal limits, ecifically those for the APLHGR (LC0 3.2.1). The 24 our Frequency is based on both engineering judgment and re ognition of the slowness of changes in power distributio during normal operation. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowance after THERMAL POWER a 25% RTP is achieved is acceptable g' en. the large inherent margin to operating limits low power l'evels.
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency of SR 3.2.4. requires a more frequent verification than if T a 1.0. W n T < 1.0, more rapid changes in power distribution ar typically expected.
REFERENCES 1. 10 CFR 50, Appendix A, G C 10, GDC 13, GDC 20, and GDC 29.
- 2. USAR, Chapter 4, App dix 4B.
- 3. USAR, Chapter 15, ppendix 158.
- 4. NRC Generic Let r 94-02, "Long-Term Solutions and Upgrade of Int rim Operating Recommendations for Thermal-Hydr lic Instabilities in Boiling Water Reactors."
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RIVER BEND B 3.2-18 Revision No. 0
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- RPS Instrumentation
, 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.3.1.1.5 Perform CHANNEL FUNCTIONAL TEST. 7 days SR 3.3.1.1.6 Verify the source range monitor (SRM) and Prior to intermediate range monitor (IRM) channels withdrawing overlap. SRMs from the fully inserted position SR 3.3.1.1.7 ------------------NOTE-------------------
Only required to be met during entry into MODE 2 from MODE 1.
Verify the IRM and APRM channels overlap. 7 days 30C4 I SR 3.3.1.1.8 Calibrate the local power range monitors. 140tr MWD /T average core exposure SR 3.3.1.1.9 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.1.1.10 Calibrate the trip units. 92 days (continued) l l
RIVER BEND 3.3-4 Amendment No. 81 i
'e RPS Instrumentation
,6 . B 3.3.1.1 BASES I
I either APRM downscale rod block, or IRM upscale rod block. {
Overlap between SRMs and IRMs similarly exists when, prior to. withdrawing the SRMs from the fully inserted position, IRMs are above 2/40 on Range 1 before SRMs have reached the upscale rod block.
As noted, SR 3.3.1.1.7 is only required to be met during entry into MODE 2 frem MODE 1. That is, after the overlap requirement has been met and indication has transitioned to the IRMs, maintaining overlap is not required (APRMs may be reading downscale once in MODE 2).
If overlap for a group of channels is not demonstrated (e.g., IRM/APRM overlap), the reason for the failure of the Surveillance should be determined and the appropriate channel (s) declared inoperable. Only those appropriate channel (s) that are required in the current MODE or condition should be declared inoperable.
A Frequency of 7 days is reasonable based on engineering judgment and the reliability of the IRMs and APRMs. ;
l SR 3.3.1.1.8 LPRM gain settings are determined from the local flux profiles measured by the Traversing Incore Probe (TIP)
System. This establishes the relative local flux profile )
I for appropriate representati,ve input to the APRM Syst-m.
The 21000 MWD /T Frequency is based on NA200 and NA300 LrRM ope' rating experience and that resulting nodal power uncertaintyf combined with the other identified uncertainties, remains less than the uncertainty allowed by the GETAB Safety Limit ( 8. 7 % ) .with I DR.". : ncitivity chang : .
(continued) {'
SURVEILLANCE SR 3.3.1.1.9 and SR 3.3.1.1.12 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed on each required i
~
channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The 92 day Frequency of SR 3.3.1.1.9 is based on the reliability analysis of Reference 9.
For Functions 9 and 10 the CHANNEL FUNCTIONAL TEST shall include the turbine first stage pressure instruments.
The 18 month Frequency is based on the need to perfonn this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.
Operating experience has shown that these components usually RIVER BEND B 3.3-262828 Revision No. 1 l
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RPS Instrumsntation B 3.3.1.1
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BASES .
pass the Surveillance when performed at the 18 month Frequency.
SR 3.3.1.1.10 The calibration of trip units provides a check of the actual trip setpoints.
the The channel must be declared inoperable if l trip setting is discovered to be less conservative than the Allowable value specified in Table 3.3.1.1-1. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be readjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology. I For Functions 9 and 10 all applicable trip unit setpoints i i
must be calibrated including the turbine first stage pressure instrument trip unit setpoints.
The Frequency of 92 days for SR 3.3.1.1.10 is based on the reliability analysis of Reference 9. I l
(continued) l SURVEILLANCE SR 3.3.1.1.11, SR 3.3.1.1.13, and SR )
REQUIREMENTS 3.3.1.1.17 (continued) A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor.
This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel ;
adjusted to account for instrument drifts between successive calibrations cons stent with the plant specific setpoint methodology.
For Functions 9 and 10 the CHANNEL CALIBRATION shall include the turbine first stage pressure instruments.
Note 1 states that neutron detectors and flow reference transmitters are excluded from CHANNEL CALIBRATION because of the difficulty of simulating a meaningful signal.
Changes in neutron detector sensitivity are compensated for by performing the 7 day calorimetric calibration l
(SR 3.3.1.1.2) and the 21000 MWD /T LPRM calibration against the TIPS (SR 3.3.1.1.8). Calibration of the flow reference f
l RIVER BEND B 3.3-272&28 Revision No. 1 j