ML20082D485
| ML20082D485 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 11/18/1983 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20082D431 | List: |
| References | |
| TAC-53214, TAC-53215, NUDOCS 8311220443 | |
| Download: ML20082D485 (23) | |
Text
{{#Wiki_filter:. Attachment 1 Proposed Amendments to McGuire Units 1 and 2 Technical Specifications t 1 8311220443 831118 PDR ADOCK 05000369 P PDR
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0.2 0.4 0.6 0.8 1.0 1.2 O FRACTIOfl OF RATED THERMAL POWER FIGURE 2.114 l REACTOR CORE SAFETY LIMIT . FOUR LOOPS IN OPERATION UNIT 1. l 2 '., McGUIRE - UNITS I and 2
665, Flow Per Loop = 95,500 gpm 660f 655I 249C 650 ##d 645 226 0osy, Operation 6404 655 650 2 con Ds1* 625 u
- 1900
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[615-610 605 GEO- . 5 3' Acceptable \ Operation 50 555< l l 550-575< 570 ,
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McGUIRE UNITS 1 and 2 2-2a
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2 TABLE 2.2-1 E S REACTOR TRIP SYSTEM INSTRilHENIATION TRIP SETPOINTS M ALLOW'ABLE VALUES M CTIONAL UNIT TRIP SETPOINT k 1. Manual Reactor Trip H.A. N.A.
$ Low Setpoint - 1 26% of RATED
~ 2. Power Range, Neutron Flux Low Setpoint - 1 25% of RATED THERMAL POWER THERMAL POWER 5 High Setpoint - 1110% of RATED n High Setpoint - 5 109% of RATED THERMAL POWER THERMAL POWER
- 3. Power Range, Neutron Flux, 5 5% of RATED THERHAL POWER with 5 5.5% of RATED THERMAL POWER liigh Positive Rate a time constant 1 2 seconds with a time constant 1 2 seconds
- 4. Power Range, Neutron Flux, 5 5% of RATED TilERMAL POWER with 5 5.5% of RATED THERMAL POWER liigh Negative Rate a time constant 1 2 seconds with a time constant 1 2 seconds
$ 5. Intermediate Range, Neutron 5 25% of RATED TilERMAL POWER $ 30% of RATED THERMAL POWER Flux 5 5
- 6. Source Range, Neutron Flux $ 10 counts per second i 1.3 x 10 counts per second
- 7. Overtemperature aT See Note 1 See Note 3
- 8. Overpower AT See Note 2 See Note 3
- 9. Pressurizer Pressure--Low 1 1945 psig 1 1935 psig
- 10. Pressurizer Pressure--High 1 2385 psig 1 2395 psig
- 11. Pressurizer Water Level--High 1 92% of instrument span 1 93% of instrument span
- 12. Low Reactor Coolant Flow 1 90% of design flow per loop
- 1 89% of design flow per loop *
- Design flow is 97,500 gpm per loop, for Unit 1.
and 95,5oo pm per I.c.p fo r U n;t '2 . 4
i e 3 TABLE 2.2-1 (Continued) REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS i NOTATION (Continued) E NOTE 1: (Continued) P = Pressurizer pressure, psig, P' = 2235 psig (Nominal RCS operating pressure), S = Laplace transform operator, sec-1, and f 1(AI) is a function of the indicated difference between top and_ bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that: (i) for q - qb between -36% and +9.5% for Unit 1 or between -36% and +8% for Unit 2, l f 1(AI = 0, where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the. core respectively, and qt + qb is total THERMAL POWER in percent of RATED THERMAL POWER; (ii) for each percent that the magnitude of qt - Ab exceeds -36%, the AT Trip Setpoint shall be automatically reduced for Unit 1 by 0.863% or for Unit 2 by 1.173% of l its value at RATED THERMAL POWER; and (iii) for each percent that the magnitude of qt - Ab exceeds +9.5% for Unit 1 or +8% for Unit 2, the AT Trip Setpoint shall be automatically reduced for Unit 1 by 0.983% or for Unit 2 by 0.901% of its val'se at RATED THERMAL POWER.
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FIGURE 3.2-4b UNIT 2 R0D B0W PENALTY AS A FUNCTION OF BURilUP ,
POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT CHANNEL FACTOR. and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continueo)
- c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and
- d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.
N FhwillbemaintainedwithinitslimitsprovidedConditionsa.through d..above are maintained. As noted on Figures 3.2-3 and 3.2-4, RCS flow rate and F may be " traded off" against one another (i.e. , a low measured RCS flow rate is acceptable if the measured F is also low) to ensure that the calcu-lated DNBR will not be below the design DNBR value. The relaxation of F N as a function of THERMAL POWER allows changes in the racial power shace for a'l permissible rod insertion limits. R, as calculated in Specification 3.2.3 and used in Figure 3.2-3, accounts for Fg N^ less than or equal to 1.49. This value is used in the various accident analyses where F influences parameters other than DNBR, e.g., peak clad tem-perature, and thus is the maximum "as measured" value allowed. R , as defined, 2 allows for the inclusion of a penalty for Rod Bow on DNBR only. Thus, knowing the "as measured" values of FN and RCS flow allows for " tradeoffs" in excess of R equal to 1.0 for the purpose of offsetting the Rod Bow DNBR penalty. INSERT
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When an qF measurement is taken, an allowance for both experimental error and manufacturing tolerance must be made. An allowance of 5% is accropriate for a full-core mao taken with the Incore Detector Flux Macoing System, and a 3% allowance is appropriate for manufacturing tolerance. Acencment No. 22 ' Unit ' ; McGUIRE - UNITS 1 and 2 8 3/a 2-4 Amentment No. 3 (Unit 2;
;/70#07
d I I KJ S E Ellr >< Fuel rod bowing reduces the value of DNB ratio. Credit is available to partially offset this reduction. This credit comes from generic or plant specific design margin. For McGuire Unit 1, the margin used to partially offset' rod bow penalties is 9.1 percent. For McGuire Unit 2, the margin used to partially offset rod bow penalties is 5.9 percent with the remaining 3.2 percent used to trade off against measured flow being as much as 2 percent lower than thermal design flow plus uncertainties. The penalties applied to F H to account for rod bow (Figures 3.2-4 Unit I and Unit 2) as a function of burnup are consistent with those described in Mr. John F. Stolz's (NRC) letter to T. M. Anderson (Westinghouse) dated April 5,1979 with the difference being due to the amount of margin each unit uses to partially offset rod bow penalties. o 4
. . _ . - _ ._, . , _ _ _ _ . - ., m, ,_r._ . ,
9 Attachment 2 Justification and Safety Analysis The proposed amendments shown in Attachment I would accomplish two related
- purposes:
(a) Reduce by 2% the Reactor Coolant System (RCS) flow rate required for operation of McGuire Unit 2 at 100% power and revise the limits-for safety systems setting to accomodate the RCS flow reduction (b) Provide for a 1% reduction in power for each 1% reduction in , the measured RCS flow below the flow requirement for 100% power for McGuire Unit 2. Attachment 2A contains a safety evaluation which concludes that the 2% reduction in Unit 2 RCS flow rate is acceptable. Appropriate limits for safety system settings are revised, consistent with the safety evaluation, to accomodate the RCS flow reduction. This reduction in the Unit 2 RCS flow requirement is necessary because the most recent measurement of RCS flow by precision heat balance indicated total RCS flow to be 392,500 gpm -- less than the current
' requirement. Thus. power level is limited to 90%.
Attachment 2B evaluates and justifies the proposed change to require a 1% power
~
reduction for each 1% reduction in RCS flow rate below the 100% power flow requirement. The concept of trading power for flow (10% power for 5% flow) is already included in the McGuire Technical Specifications. The proposed amendment would change the ratio for the power / flow' tradeoff from 2/1 to 1/1 and provide limits for each 1% increment. Although the change is proposed only for Unit 2, a similar change for Unit 1 will be proposed along with other changes required for Cycle 2 operation. , h
Attachment 2A McGUIRE UNIT 2 CYCLE 1 RIDUCED RCS FLOW SAFETY EVALUATION The following safety evaluation confirms the acceptability of operation a,t.100 percent of rated thermal power and 98 percent of the current full power RCS flow requirement as specified in the Technical Specifications for McGuire Unit 2, Cycle 1. All of the affected FSAR Chapter 15 accidents and protection system setpoints have been reviewed to determine the impact of the proposed reduction in flow requirement. In addition, Technical Specification changes required to support the reduced flow are included. This safety evaluation applies to Cycle 1 operation of McGuire Unit 2 only. DNB Considerations Thermal-hydraulic sensitivity studies indicate that a 2 percent flow reduction results in a 3.2 percent DNBR penalty. The generic DNBR margin will be reduced from 9.1 percent to 5.9 percent in order to maintain the validity of all previous DNB evaluaticas. Thus, the core DNB limits are unchanged, and the conclusion that the DNB design basis is met for the following FSAR transients remains valid: Excessive Heat Removal Due to Feedwater System Malfunction Excessive Load Increase
- Main Steamline Depressurization Main Steamline Rupture _ -
- Loss of Load / Turbine Trip Partial Loss of Forced Reactor Coolant Flow
- Complete Loss of Forced Reactor Coolant Flow Uncontrolled RCCA Bank Withdrawal from a Subcritical Condition Uncontrolled RCCA Bank Withdrawal at Power
- Startup of an Inactive Reactor Coolant Loop
- Inadvertent ECCS Operation at Power
- Reactor Coolant System Depressurization 1
Non-DNB Considerations In addition to the DNB concern, the following eva'uations are presented for those accidents which are not DNB related or for which DN8R is not the only safety crf terion to be met. Control Rod Withdrawal From a Subcritical Condition A control rod assembly withdrawal incident when the reactor is subcritical results in an uncontrolled addition of reactivity leading to a power excursion (Section 15.2.1 of the FSAR). The nuclear power response is characterized by a very fast rise terminated by the reactivity feedback of the negative fuel temperature coefficient. The power excursion causes a heatup of the moderator. However, since the power rise is rapid and is followed by an immediate reactor trip, the moderator temperature rise is small. Thus, nuclear power response is primarily a function of the Doppler teragerature coefficient. An increase in temperature due to reduced RCS fl.a: would result in more Doppler feedback, thus reducing the nuclear power exch, r, m as presented in the FSAR which partially compensates for the fle Moction. The FSAR analysis shows that for a reactivity insertion rate of 76 x 10-5 aK/sec, the peak heat flux achieved is 73.1 percent of nominal with a resultant peak fuel average temperature of 760*F. A 2 percent reduction of reactor coolant flow would degrade heat transfer from the fuel by a maximum of 2 percent. Thus, peak fuel and clad temperatures would also increase by a maximum 2 percent, yielding maximum fuel and clad temperatures which are still significantly below fuel melt (5080*F) and zirconium-H O 2 reaction (1800*F) limits. Therefore, the conclusions presented in the FSAR are still valid. Boron Dilution The results of the boron dilution transient will remain unchanged for all modes of operation due to a reduction in reactor coolant ficw. The maximum dilution flow rate, RCS active volumes, and RCS boron concentrations are not impacted by a reduction in flow. Since these parameters determine the amount of time available to the operator to tenainate the dilution event, the results presented in the FSAR remain unchanged. 2
v Loss of Load The loss t' load accident is presented in Section 15.2.7 of the FSAR and can result from either loss of external electrical load or a turbine trip. The reselt of a loss of load is an increase in core power which exceeds the secondary system power extraction, thus, causing an increase in core water tempe,ra ture. A reducticn in RCS flow will result in a more rapid pressure rise tnan that shown in the FSAR. The effect will be minor, however, since the reactor is tripped on high pressurizer pressure. Thus, the time to trip will be decreased, which will result in a lower total energy input to the
. coolant. The analys(s shows a peak pressurizer pressure of 2551 psia. A 2 percent reduction in flow will 'ead to a conservative increase in system
, pressure to 2570 psia. The pressurizer will not fill, and the maximum pressures are within the design limits. Therefore, operation at reduced flow will not violate safety limits following 3 loss of load accident.
Loss of Normal Feedwater/ Station Blackout
. This transient is analyzed to demonstrate that the peak RCS pressure does not exceed al'lowable limits and that the core remains covered with water. These criteria are assured by applying the more stringent requirement that the pressurizer must not be filled with water. The effect of reducing initial core flow would result in an initial more rapid heatup of the RCS. The '
resultant coolant density decrease would increase the volume of water in the pressuri zer.- However, considerable margin to filling the pressurizer is available during the initial portion of the transient. During the long-term portion of the transient, the peak RCS temperature (and resultant peak ' pressurizer water volume) is reached when the heat removal capability of the auxiliary feedWater matches the core decay heat generation. -If 'the RCS flow shortfall is due to higher than anticipated loop resistances, the natural circulation flow would be reduced by an amount proportional to the thermal design flow reduction. The slight reduction in natural circulation at the peak temperature condition would not significantly impair the heat transfer across the steam generator tubes, thus resulting in a similar peak pressurizer water volume. Therefore, the FSAR conclusions remain valid. 3
b s Steamline Break Tne steamline break transient is analyzed at hot zero power, end-of-life conditions for the following cases: Inadvertent opening of a steam dump, safety, or relief valve (Section 15.2.13 of the FSAR) Main steam pipe rupture with and without offsite power available (Section 15.4.2 of the FSAR) A steamline break results in a rapid depressurization of the steam generators and primary side cooldown. This causes a large reactivity insertion due to the presence of a negative moderator temperature coefficient. A reduction in reactor coolant flow will result in a reduction in heat transfer from the fuel to the coolant. Therefore, the reactivity insertion and return to power in the double-ended rupture case for reduced flow conditions would be less limiting than the cases presented in the FSAR. For the double-ended rupture case, the time of safety injection actuation is unaffected by reduced coolant fl ow. This, coupled with a slower return to power would result in a
-c significant reduction in peak average power from the FSAR results. The main
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steam depressurization case is bounded by the double-ended rupture. Since the return to power is less severe and the DNB evaluations remain valid as previously stated, the conclusions presented in the FSAR are still valid for a 2 percent reduction in reactor coolant flow. Rupture of a Main Feedwater Line This transient is analyzed to demonstrate that the peak RCS pressure does not
- exceed allowable limits and that the core remains covered with water. These criteria are assured by applying the more stringent requirement that bulk voiding does not occur at the outlet of the core. The effect of reducing initial core flow would result in an initial more rapid heatup of the reactur coolant system (RCS). However, a considerabl,e margin to hot leg saturation exists during the initial portion of the transient.
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For the case analyzed with offsite power available, the slight reduction in RCS flow would not significantly degrade the heat transfer across the steau .. generator tubes, hence resulting in a similar long term RCS heatup. If the RCS flow shortfall is due to higher than anticipated loop resistances, the natural circulation flow would be reduced by an amount proportional to the thermal design flow reduction for the case analyzed without offsite power available. The slight reduction in natural circulation flow would not significantly degrade the heat transfer from primary to secondary, thus resulting in a similar peak hot leg temperature at the time at which the heat removal capability of the auxiliary feedwater matches the core generated decay heat. Therefore, the FSAR conclusions remain valid. Locked Rotor The FSAR (Section 15.4.4) shows that the most severe locked rotor is an instantaneous seizure of a reactor coolant pump rotor at 100 percent power with three loops operating. Following the incicent, reactor coolant system temperature rises until shortly af ter reactor trip. A reduction in RCS flow will not affect the time to DNB since DNB is conservatively assumed to occur at the beginning of the transient. The flow reduction in the affected loop is so rapid that the time of reactor trip, on low flow, does not change due to the 2 percent reduction in reactor coolant flow. Therefore, the nuclear power and heat flux transients will not change from those presented in the FSAR. However, the reduction in flow will result in slightly higher system press'ures and clad temperatures. The peak RCS pressure reported in the FSAR was 2605 psia. A 2 percent reduction in reactor coolant flow would cause a conservative increase in pressure of 30 psia to 2635 psia, which is still significantly below the pressure at which vessel stress limits are exceeded. The paak clad temperature reported in the FSAR is 1817*F, well below the limit of 2700*F, and shows that a slight increase in this parameter due to reduced RCS flow can be easily accommodated. Therefore, the FSAR conclusions are still valid. Control Rod Ejection . The rod ejection transient is analyzed at full power and hot standby for both beginning and end-of life conditions (Section 15.4.6 of the FSAR). A 5
reduction in core flow will result in a reduction in heat transfer to the coolant, which will increase peak clad and fuel temperatures and peak fuel ,, stored energy. However, all cases have margin to fuel failure limits. The effect of reducing reactor coolant flow is to increase the peak clad temperatures. , The analysis shows that, for the worst case, there is sufficient conservatism in the analysis assumptions and margin in the results such that the peak clad temperature limit (2700*F) is not violated with the reduced flow. The fuel temperatures and peak fuel stored energy will also
' increase slightly due to the 2 percent decrease in reactor coolant flow.
However, there is sufficient margin between the analysis results and the limits to accomodate the effects of the reduced flow. Therefore, the conclusions presented in the FSAR are still valid. LOCA Analysis The impact on ECCS perfonnance of a 2 percent reduction in thermal design (TD) flow has been evaluated for 100 percent power operation for Cycle 1. RCS parameters calculated for 98 percent TD flow exhibit a 0.6*F lower value of Tcold, since steam pressure will be held the same as at 100 percent TD fl ow. The limiting case RCS bilwdown transient at 98 percent TD flow will therefore be more favorable than that previously calculated for 100 percent fl ow. In responding to NRC question 212.103, Westinghouse determined that reducing Tcold produces a benefit in the McGuire 2 limiting case break (CD = 0.6, perfect mixing) calculated peak clad temperature (PCT); that sensitivity study established that a 10*F drop in calculated LOCA PCT results from each 1*F reduction in T cold. Therefore, the foremost impact of a reduced TD flow upon calculated McGuire 2 ECCS performance is an improvement in calculated' PCT of about 6*F. An additional benefit in calculated PCT for McGuire 2 Cycle 1 may be found in the actual as-built fuel pellet average temperature, which is 20*F lower than the generic value for 17x17 fuel utilized in the FSAR LOCA analysis. The 6
sensitivity of calculated PCT to pellet average temperature was detenained for the McGuire 2 limiting case break as [12*F PCT /30*F pellet! temperature] by a , set of computer runs. Therefore, an improvement in calculated PCT of 8*F may be credited to McGuire 2 Cycle 1. Relative to the existing ECCS performance analysis for McGuire 2 Cycle 1, the lower TD flow at an equivalent core power manifests itself in a greater initial fluid enthalpy rise in both the hot assembly and in the core as a whole. The impact of the change.in core enthalpy rise and TD flow was investigated, and it was detennined that DNB occurs 0.02 seconds earlier at the midplane of the hot assembly in a UHISATAN case based upon 100 percent power /98 percent TD flow than in a 100 percent power /100 percent TD flow case for McGuire Unit 2. The earlier DNB time will cause an increase in calculated PCT of about 2*F. In summary, the effect of r2ducing TD flow by 2 percent and taking credit for the as-built fuel parameters is a benefit in calculated PCT of (6*F + 8'F - 2*F) = 12*F for the limiting case break. The existing Cycle 1 ECCS perfonnance analysis for McGuire 2 predicts 12*F margin to the 2200*F regulatory limit. The total margin in calculated PCT exceeds 20*F, which is more than adequate to ensure that McGuire 2 Cycle 1 shows compliance with 10CFR50.46 for 100 percent power operation at 98 percent TD flow. t 7
Technical Specification Changes Appendix A provides the necessary revisions to the Technical Specifications to support operation at the reduced flow for McGuire Unit 2 Cycle 1. The current specs remain applicable for M:Guire Unit 1 Cycle 1. Each Technical . Specification change is discussed below. 2.1 Safety Limits A new reactor core safety limits curve applicable to Unit 2 Cycle 1 is provided. As discussed above, the DNB limits of the figure are unchanged.' However, the Vessel Exit Boiling limits become more restrictive since flow is reduced for a given power. 2.2 Limiting Safety System Settings. The protection system setpoints have been reviewed for the reduced flow. The only setpoints which are impacted by the flow reduction are the Overtemperature aT and Overpower 4T functions. These setpoints are designed to protect the core by tripping the reactor before the core safety limits (Figure 2.1-1) are exceeded. The current setpoint equations have been determined to be adequate if the f(aI) penalty on the Overtemperature aT equation is revised as shown. , 3/4.2.3 RCS Flow Rate and F g, and Bases A new RCS flow vs.1 R , R2 figure is provided for Unit 2 to reflect the reduced flow and the power-flow tradeoff at part power. (The tradeoff is evaluatedinAttachment28.) The rod bow penalty curve for Unit 2 has been changed to reflect the reduced inargin available to partially offset rod bow penalties. i t l 8
_- - - -- - - - , . . -- - - - - . - - ~. r Attachment 2B 1 Current Technical Specification 3.2.3, Figure 3.2-3 limits operation to 190% power whenever RCS flow rate is measured to be between 95% and 100% of the flow rate required for 100% power operation. This provision is based upon trading off a 10% power reduction for a 5% (or less) flow reduction. Attachment 2A discusses a proposed reduction in the RCS flow requirement for 100% power operation; however, the measured RCS flow rate (392,500 gpm) only slightly exceeds the proposed RCS flow requirement (388,880 gpm). Therefore, a slight reduction in the measured RCS flow rate
, in the future would require a reduction in power level. Reducing power level to 490% for only a slight reduction in flow is excessively conservative.
, Therefore, a 1% reduction in power for each 1% flow reduction is proposed., _ As with the current specification, a flow reduction of mnre than 5% would require unit shutdown, per the ACTION statement. The paojosed change would, therefora, allow operation at higher power levels when flow is only slightly reduced; the required power reduction would be more commensurate with the magnitude of the flow reduction. i The proposed 1% power /1% flow tradeoff has been evaluated to ensure that the margin for' Departure from Nucleate Boiling Ratio (DNBR) would be maintained. Generally, the following relationships between core power, flow, and DNBR are i applicable: 1 3 FLOW 1% N'
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(Eq. 1) 3 DNBR 1% 1 3 POWER 1% _, g (Eq. 2)
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, 3 DNBR 1.8% t Thus the relationship between power and flow is: N 4 0.555 . . . (Eq. 3) B FLOW 1.8% l These relationships indicate that the thermal margin to DNB is maintained if power is reduced by 0.56% for_each 1% decrease in RCS flow. The plant-specific relationship for flow and DNBR based on thermal-hydraulic' sensitivity studies
- for the worst case is:
3 flow 2% 1% (Eq. 4)
.s ,
3 DNBR 3.2% 1.6% e b
. ~ Thus, for McGuire Unit 2: 3 power , 1.6% _
-- = 0.88 (Eq. 5)
B flow 1.8% Equation 5 indicates that thermal margin to DNB is maintained if power is reduced by 0.89% for each 1% decrease in RCS flow. In the SER for Amendment No. 9 to the McGuire Unit I license (NPF-9), the NRC Staff reported the results of an independent audit which showed that a 5% , reduction in flow requires a 4.32% reduction in power to maintain thermal margin at the full flow value. Thus: 3: 3: 0.864 (NRC) (Eq. 6) 3 FLOW 5% The proposed one-for-one ratio between power and flow is more conservative than all the relationships noted above. Therefore, thermal margin is maintained. McGuire Unit 2's Technical Specifications limits and accident analyses results have been evaluated to determine the impact of the RCS flow and power reductions. Due to the conservative tradeoff between power and flow, sufficient margin exists to allow plant operation. No other Technical Specification limits require modi-fication, including core limits and instrument setpoints. The impact of operation at 95% power with 95% Thermal Design Flow has been. evaluated for the LOCA analysis. Since RCS parameters calculated for 95% flow will exhibit little change in Tcold, the RCS blowdown transient would be very similar to that predicted in the existing analysis for McGuire Unit 2. At the reduced core power level, the lower flow will not cause a greater initial fluid enthalpy rise in either the hot assembly or in the core as a whole, so no penalty in calculated peaking factor will occur. Overall, it is clear from the sensitivity of ECCS performance to core power that if 95% power is assumed, 95% of Thermal Design Flow can be accomodated at the currently licensed peaking factor of 2.32. Similarly, the tradeoff of 1% power reduction for 1% flow reduction may be made for any Thermal Design Flow shortfall. m
P Attachment 3 Analysis of Significant Hazards Consideration As required by 10 CFR 50.91, this analysis is provided concerning whether the proposed amendments involve significant hazards considerations, as defined by 10 CFR 50.92. One aspect of the proposed amendments would be to reduce the Reactor Coolant System (RCS) flow requirement for McGuire Unit 2 and revise the limits for safety system settings to accomodate the flow reduction. The lower RCS flow would not affect the probability of accidents previously evaluated nor create the possibility of a new or different kind of accident; however, lower RCS flow can have some effect on the consequences of accidents previously evaluated. The effects of lower RCS flow have been evaluated for the accidents discussed in the Final Safety Analysis Report (FSAR), Chapter
- 15. This evaluation (see Attachment 2) has shown that adequate thermal margin to Departure from Nucleate Boiling Ratio (DNBR) would be maintained (i.e.
DNBR greater than 1.30). Non-DNB-limited transients were also evaluated and the results were determined to be within their respective limits. Therefore, operation under this aspect of the proposed amendments would not involve a significant increase in the consequences of accidents previously evaluated. Similarly, because the evaluation showed that the original analysis results are valid for the DNB-limited transients, the safety margins inherent in the DNBR limit of 1.30 (based on the W-3 correlation) are unaffected. Also, the non-DNB-limited transients remain within their respective limits. Therefore, this aspect of the proposed amendments does not involve a significant reduc-tion in a safety margin. The Commission has provided examples of amendments likely to involve no significant hazards considerations (48 FR 14870). One example of actions likely to involve no significant hazards considerations is an amendment which either may result in some increase to the probability or consequenc?s of a previously-analyzed accident or may reduce a safety margin, but where the results of the change are clearly within all acceptable criteria with respect to the system or component specified in the Standard Review Plan. Because the evaluation previously discussed shows that the DNB limit of 1.30 is met (Re: Standard Review Plan Section 4.4, Acceptance Criterion 1) and other design-basis transients would remain within their respective limits, the above example can be applied to this situation. Another aspect of the proposed amendments would involve a requirement to reduce power by 1% for each 1% reduction in RCS flow below the minimum flow required for 100% power. Thermal-hydraulic sensitivity studies have shown that this power / flow tradeoff is conservative with respect to DNB margin. Therefore, this aspect of the proposed amendments would not: (1) involve a significant increase in the probability or consequences of an accident, (2) create the possibility of a new or different kind of an accident, or (3) involve a significant reduction in a safety margin. Based upon the preceding analysis, Duke Power proposes that the proposed amendments do not involve a significant hazards consideration.
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