ML18058B390

Proposed TS Table 3.23-2, Radiation Peaking Factor Limits for Cycle 11.

ML18058B390
Person / Time
Site:	Palisades
Issue date:	01/29/1993
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
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ML18058B389	List: ML18058B388 ML18058B390 ML18058B391 ML18058B392
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NUDOCS 9302040159
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ATTACHMENT 1 Consumers Power Company Palisades Plant Docket 50-255 CYCLE 11 TECHNICAL SPECIFICATIONS CHANGE Proposed Pages January 29, 1993 6 Pages 9302040159 930129 PDR ADOCK 05000255 p PDR

2.0 BASIS - Safet~imits and Limiting Safety Syst~Settinqs 2.1 Basis - Reactor Core Safety limit To maintain the integrity of the fuel cladding and prevent fission product release, it is necessary to prevent overheating of the cladding under normal operating conditions. This is accomplished by operating within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is large enough so that the clad surface temperature is only slightly greater than the coolant temperature. The upper boundary of the nucleate boiling regime is termed "departure from nucleate boiling" (DNB).

At this point, there is a sharp reduction of the heat transfer coefficient, which would result in high-claddin~ temperatures and the possibility of cladding failure. Although DNB is not an observable parameter during reactor operation, the observable parameters of thermal power, primary coolant flow, temperature and pressure, can be related to DNB through the use of a DNB Correlation. DNB Correlations have been developed to predict DNB and the location of DNB for axially uniform and nonuniform heat flux distributions. The local DNB ratio (DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core location to the actual heat flux, is indicative of the margin to DNB. The minimum value of the DNBR, during steady-state operation, normal operational transients, and anticipated transients is limited to DNB correlation safety limit. A DNBR equal to the DNB correlation safety limit corresponds to a 95% probability at a 95% confidence level that DNB will not occur which is considered an appropriate margin to DNB for all operating conditions.

The reactor protective system is designed to prevent any anticipated combination of transient conditions for primary coolant system temperature, pressure and thermal power level that would result in a DNBR of less than the DNB correlation safety limit. The Palisades safety analyses uses two DNB correlations. The XNB correlation discussed in References I and 2 determines the safety limit for those fuel assemblies initially loaded prior to Cycle 9. The ANFP correlation discussed in References 4 and 5 determines the safety limit for those fuel assemblies initially loaded in Cycle 9 and later. Fuel assemblies initially loaded prior to Cycle 9 are of a different construction than later assemblies which utilize a High Thermal Performance design.

The minimum DNBR analyses are in accordance with Reference 6.

References 1 XN-NF-62l(P)(A), Rev 1 2 XN-NF-709 3 Updated FSAR, Section 14.1.

4 ANF-1224 (P)(A), May 1989 5 ANF-89-192(P), January 1990 6 XN-NF-82-2l{A), Revision 1 Amendment No. pJ, ~p, JJ~, JpJ, J~~'

B 2-1

2.0 BASIS - Safet~imits and Limiting Safety Syst~Settings 2.3 Basis - Limiting Safety System Settings (continued)

5. Low Steam Generator Water Level - The low steam generator water level reactor trip protects against the loss of feed-water flow accidents and assures that the design pressure of the primary coolant system will not be exceeded. The specified set point assures that there will be sufficient water inventory in the steam generator at the time of trip to allow a safe and orderly plant shutdown and to prevent ~team generator dryout assuming minimum auxiliary feedwater capacity. 1 The setting listed in Table 2.3.1 assures that the heat transfer surface (tubes) is covered with water when the reactor is critical.

6. Low Steam Generator Pressure - A reactor trip on low steam generator secondary pressure is provided to protect against an excessive rate of heat extraction from the steam generators and subsequent cooldown of the primary coolant. The setting of 500 psia is sufficiently below the rated load operating point of 739 psia so as not to interfere with normal operation, but still high enough to provide the required protection in the event of excessi~ely high steam flow. This setting was used in the accident analysis.' 1

7. Containment High Pressure - A reactor trip on containment high pressure is provided to assure that the reactor is shutdown beforie the initiation of the safety injection system and containment spray. 71 References 1 EMF-92-178, Table 15.0.7-1 2 Updated FSAR, Section 7.2.3.3.

3 EMF-92-178, Section 15.0.7-1 4 XN-NF-86-9l(P) 5 ANF-90-078, Section 15.1.5 6 ANF-87-150(NP), Volume 2, Section 15.2.7 7 Updated FSAR, Section 7.2.3.9.

8 ANF-90-078, Section 15.2.1 Amendment No$), ~i, ))~, )$7, )~~'

B 2-5

3.1 PRIMARY COOLANT SYSTEM (Cont'd)

Basis (Cont'd) measurement; +/-0.06 for ASI measurement; +/-50 psi for pressurizer pressure;

+/-7°F for inlet temperature; and 3% measurement and 3% bypass for core flow. In addition, transient biases were included in the derivation of the following equation for limiting reactor inlet temperature:

Tinlet ~ 542.99 + .0580(P-2060) + O.OOOOl(P-2060)**2 + l.125{W-138) -

.0205(W-138)**2 .

The limits of validity of this equation are:

1800 < pre§sure < 2200 psia 100.0-x 10 ~Vessel Flow~ 150 x 10 6 lb/h ASI as shown in Figure 3.0 With measured primary coolant system flow rates > 150 M lbm/hr, limiting the maximum allowed inlet temperature to the T1 10 LCO at (1

150 M lbm/hr increases the margin to DNB for higher PCS flow rates 141

• The Axial Shape Index alarm channel is being used to monitor the ASI to ensure that the assumed axial power profiles used in the development of the inlet temperature LCO bound measured axial power profiles. The signal representing core power (Q) is the auctioneered higher of the neutron flux power and the Delta-T power. The measured ASI calculated from the excore detector signals and adjusted for shape annealing {Y 1) and the core power constitute an ordered pair {Q,Y 1). An alarm signal is activated before the ordered pair exceed the boundaries specified in Figure 3.0.

The requirement that the steam generator temperature be ~ the PCS temperature when forced circulation is initiated in the PCS ensures that an energy addition caused by heat transferred from the secondary system to the PCS will not occur. This requirement applies only to the initiation of forced circulation (the start of the first primary coolant pump) when the PCS cold leg temperature is < 430°F. However, analysis (Reference 6) shows that under limited conditions when the Shutdown Cooling System is isolated from the PCS, forced circulation may be initiated when the steam generator temperature is higher than the PCS cold leg temperature.

References 1 Updated FSAR, Section 14.3.2.

2 Updated FSAR, Section 4.3.7.

3 Deleted 4 EMF-92-178 Section 15.0.7.1 5 ANF-90-078 6 Consumers Power Company Engineering Analysis EA-A-NL-89-14-1 3-3 Amendment No. pJ, ~J, JJ7 JJ~, JpJ, Jp~, J$7, J~p,

3.12 MODERATOR TEMPERATURE COEFFICIENT OF REACTIVITY Applicability Applies to the moderator temperature coefficient of reactivity for the core.

Objective To specify a limit for the positive moderator coefficient.

Specifications The moderator \emperature coefficient (MTC) shall be less positive than +0.5 x 10- Ap/°F at ~ 2% of rated power.

Bases The limitations on moderator temperature coefficient (MTC) are provided to ensure that the assumptions used in the safety analysi s 111 remain valid.

Reference (1) EMF-92-178, Section 15.0.5 3-67 Amendment No. lJ~, 1$7, J~p, (next page is 3-69)

TABLE 3.23-1 LINEAR HEAT RATE LIMITS No. of Fuel Rods Assembly 208 216 Peak Rod 15.28 kW/ft 15.28 kW/ft TABLE 3.23-2 RADIAL PEAKING FACTOR LIMITS, FL Peaking Factor No. of Fuel Rods in Assembly 208 216 216 216 Reload M Reload N Reload 0 Assembly FAr 1.48 1.57 1.66 1. 76 Peak Rod FTr 1.92 1.92 1.92 2.04 TABLE 3.23-3 POWER DISTRIBUTION MEASUREMENT UNCERTAINTY FACTORS LHR/Peaki:ng Factor Measurement Measurement Measurement Parameter Uncerta i nty<a> Uncertai nt/b> Uncerta i nty<c>

LHR 0.0623 0.0664 0.0795 FAr 0.0401 0.0490 0.0695 FTr 0.0455 0.0526 0. 0722 (a) Measurement uncertainty for reload cores using all fresh incore detectors.

(b) Measurement uncertainty for reload cores u~ing a mixture of fresh and once-burned incore detectors.

(c) Measurement uncertainty when quadrant power tilt, as determined using incore measurements and an incore analysis computer program< 6 >, exceeds 2.8% but is less than or equal to 5%.

3-107 Amendment No.  %~, JJ~, J~~' J~~'

POWER DISTRIBUTION LIMITS 3.23.2 RADIAL PEAKING FACTORS LIMITING CONDITION FOR OPERATION The radial peaking factors Fe, and F~ shall be less than or equal to the value in Table 3.23-2 times the fo1lowing ~uantity. The quantity is [1.0 + 0.3 {l -

P)] for P ~ .5 and the quantity is 1.15 for P < .5. P is the core thermal power in fraction of rated power.

APPLICABILITY: Power operation above 25% of rated power.

ACTION:

1. For P < 50% of rated with any radial peaking factor exceeding its limit, be in at least hot shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

2. For P ~ 50% of rated with any radial peaking factor exceeding its limit, reduce thermal power within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to less than the lowest value of:

F

[1 - 3.33( r - 1) ] x Rated Power F

L Where Fr is the measured value of either F~, or FT and FL is the corresponding limit from Table 3.23-2. r Basis The limitations on F~, and F~ are provided to ensure that assumptions used in the analysis for establishiny DNB margin, LHR and the thermal margin/low-pressure and variable high-power trip set points remain valid during operation.

Data from the incore detectors are used for determining the measured radial peaking factors. The periodic surveillance requirements for determining the measured radial peaking factors provide assurance that they remain within prescribed limits. Determining the measured radial peaking factors after each fuel loading prior to exceeding 50% of rated power provides additional assurance that the core is properly loaded.

To ensure that the design margin of safety is maintained, the determination of radial peaking factors takes into account the appropriate measurement uncertainty factors 111 given in Table 3.23-3 References (1) FSAR Section 3.3.2.5 3-111 Amendment No.~~' JJ~, J~l, J~~' J~~'

ATTACHMENT 2 Consumers Power Company Palisades Plant Docket 50-255 CYCLE 11 TECHNICAL SPECIFICATIONS CHANGE Existing Pages Marked to Show Change January 29, 1993 6 Pages I

__J

2.0 BASIS - Safe~Limits and Limiting Safety Sys~ Settings 2.1 Basis - Reactor Core Safety limit To maintain the inte~rity of the fuel cladding and prevent fission product release, it is necessary to prevent overheating of the cladding under normal operating conditions. This is accomplished by operating within the nucleate boiling regime of heat transfer, wherein the heat transfer coefficient is large enough so that the clad surface temperature is only slightly greater than the coolant temperature. The upper boundary of the nucleate boiling regime is termed "departure from nucleate boiling" (DNB}. At this point, there is a sharp reduction of the heat transfer coefficient, which would result in high-cladding temperatures and the possibility of cladding failure. Although DNB is not an observable parameter during reactor operation, the observable parameters of thermal power, primary coolant flow, temperature and pressure, can be related to DNB through the use of a DNB Correlation.

DNB Correlations have been developed to predict DNB and the location of DNB for axially uniform and nonuniform heat flux distributions. The local DNB ratio (DNBR), defined as the ratio of the heat flux that would cause DNB at a particular core location to the actual heat flux, is indicative of the margin to DNB. The minimum value of the DNBR, during steady-state operation, normal operational transients, and anticipated transients is limited to DNB correlation safety limit. A DNBR equal to the DNB correlat;on safety limit corresponds to a 95% probability at a 95% confidence level that DNB will not occur which is considered an appropriate margin to DNB for all operating conditions.

The reactor protective system is ~e~i~ried to p~~vent-any anticipated combination of transient conditions for primary coolant system temperature, pressure and thermal power level that would result in a DNBR of less than the DNB correlation safety limit. The Palisades safety analyses uses two DNB correlations. fhe XNB correlat-ton-

.,,,,...._---:~,u*1-eretts-s-ed-+n-Re-fe.renees-l-and-2-de-t&rm-i*nes-the-s-a-~~

-fue.1-a.s-&eme+i-e-s-i-n-i-t-i-a-1-1-y-l-oaded-i-n-C-ye--l-e--&-r-lhe-ANtPr-eor-re-l*a-t+oR-

• d-i-seu~s-ea-4 n References 4 anEl- determ-i-nes-=the safety limit for those fuel assemhlies ini-tially loaded in-Cycle 9 and later. -Fue-1 as;emblies

-initially~ed in Cycle 8 are of-a-d-i-f.f-erent construction than later assemblies which ut~er-mal Per--ffmanee-de.s-1-gn-.-

The minimum DNBR analyses are in accordance with Reference 6.

References 1 XN-NF-62l(P)(A), Rev 1 2 XN-NF-709.

3 Updated FSARi~ Section 14. 1.

4 ANF-1224 (PJ{A), May 1989 5 ANF-89-19~CPlt January 1990 6 XN-NF-82-21(AJ, Revision l

-:.::-=-~~--~~~~--~~~~~~~~-- --~-~---

' The XNB correlation discussed in References 1 and 2 determines the safety limit for those fuel assemblies initially loaded prior to Cycle 9. The .A~FP correlation discussed in References 4 and 5 determines the safety l1m1t :or those fuel assemblies initially loaded in Cycle 9 and later: Fuel assemblies

• initially loaded prior to Cycle 9 are of different construct1~n than later  : 17 , tso-

, assemblies which utilize a High Thermal Performance design. -~ S, 1992 B 2-1

2.3 Basis - Limiting Safety System Settings (continued)

5. Low Steam Generator Water Level - The low steam generator water level reactor trip protects against the loss of feed-water flow accidents and assures that the design pressure of the primary .

coolant system will not be exceeded. The specified set point assures that there will be sufficient water inventory in the steam generator at the time of trip to allow a safe and orderly plant shutdown and to prevent steam~enerator dryout assuming minimum auxiliary feedwater capacity.'

The setting listed in Table 2.3.1 assures that the heat transfer surface (tubes) is covered with water when the re*ctor is critical.

6. Low Steam Generator Pressure - A reactor trip on low steam ~enerator secondary pressure is provided to protect against an excessive rate of heat extraction from the steam generators and subsequent cooldown of the primary coolant. The setting of 500 psia is sufficiently below the rated load operating point of 739 psia so as not to interfere with normal operation, but still high enough to provide the required protection in the event of excessi~~ly high steam flow.

This setting was used in the accident analysis .

. 7. Containment High Pressure - A reactor trip on *containment high pressure is provided to assure that the reactor is shutdown before the in~tiation of the safety injection system and containment spray. 1 1 References

&"IF-9:2-178 1 EHF 91 176, Table 15.0.7-1 2 U dated FSAR, Section 7.2.3.3.

1£/VIF-'l;/.-li8--'-=-3-,..,.~~;...o;~.::u:.-, Section 15. O* 7-1 4 XN-NF-86-9l(P) 5 ANF-90-078, Section 15.1.5 6 ANF-87-150CNP), Volume 2, Section 15.2.7 7 Updated FSAR, Sect ion 7. 2. 3. 9.

8 ANF-90-078, Section 15.2.1

.:~*. -. -

B 2-5

3.1 e

PRIMARY COOLANT SYSTEM (Cont'd) 1Wl! {Cont'd}

measurement; +/-0.06 for ASI measurement; +/-50 psi for pressurizer pressure; +/-7°F for- inlet temperature; and 3% measurement and 3% bypass for core fl ow. In addition, transient biases were included in the derivation of the following equation for limiting reactor inlet temperature: _

T~. s 542.99 + .0580(P-2060} + O.OOOOl(P-2060)**2 + l.125(W-138) -

.0205(W-138)**2 The limits of validity of this equation are:

1800 s pressure s 2200 psia 100.0 x 108 s Vessel Flow s 150 x 108 lb/h ASI as shown in Figure 3.0 With measured primary coolant system flow rates > 150 Mlbm/hr, limiting the maximum allowed inlet temperature to the Trnr* LCO at 150 Mlbm/hr increases the margin to DNB for higher PCS flow rates.(~)aJJ The Axial Shape Index alarm channel is being used to monitor the ASI to ensure that the assumed axial power profiles used in the development of the inlet temperature LCO bound measured axial power profiles. The signal representing core power {Q)- is the auctioneered higher of the neutron flux power and the Delta-T power. The measured ASI calculated from the excore detector signals and adjusted for shape annealing {Y,) and the core power constitute an ordered pair {Q,Y 1). An alarm signal is activated before the ordered pair exceed the boundaries specified in Figure 3.0.

The requirement that the steam generator temperature be s the PCS temperature when forced circulation is initiated in the PCS ensures that an energy addition caused by-- heat transferred from the secondary system to the PCS wi 11 not occur. This requirement applies only to the initiation of forced _

circulation (the start of- the*. first primary coolant pump} when the PCS cold leg temperature is < 43o*f. However, analysis {Reference 6) shows that under limited conditions when.the- Shutdown- Cooling System-is isolated from the PCS, forced circulation may be initiated when the steam generator temperature is higher than the PCS cold leg temperature.

References (1) Updated FSAR, Section 14. 3. 2.

{2) Updated FSAR, Section 4.3.7.

{3) Deleted E/4F-'l~-l78

{4) EMF 91 176 Section 15.0.7.1 (5) ANF-90-0JS (6) Consumers Power Company Engineering Analysis EA-A-NL-89-14-1 3 Amendment No. '/,'J, J'J, 'J'J1 'J'JJ, 'J1'J, J1"ft, f}l, --+4"3'

- March 2.7, 1992

3.12 MODERATOR TEMPERATURE COEFFICIENT OF REACTIVITY Applicability Applies core. to the moderator temperature coefficient of reactivity for the Objective To specify a limit for the positive moderator coefficient.

Specifications The moder~tor temperature coefficient (MTC) shall be less positive than

+0.5 x 10- Ap/°F at s 2% of rated power.

Bases The limitations on moderator temperature coefficient fMTC) ~re provided to ensure that the assumptions used in the safety ana ysis 11 remain valid.

Reference E/i?F-</2.-178 (1) EMF 91 176, Section 15.0.5 3-67 Amendment No. JJJ, Jil, ~

March 27, 1992 (next page is 3-69}

TABLE 3.23-1 LINEAR HEAT RATE LIMITS No. of Fuel Rods Assembly 208 216 Peak Rod 15. 28 kW/ft 15.28 kW/ft TABLE 3.23-2 RADIAL PEAKING FACTOR LIMITS, FL I TABLE 3.23-3 POWER DISTRIBUTION MEASUREMENT UNCERTAINTY FACTORS LHR/Peaki ng Factor Measurement 111 Measurement Measurement Parameter Uncerta i*n*t.x!. ..

Uncerta i nty 1b1 Uncerta i nty'cl LHR 0.0623::*:  :.. ~ **:

0.0664 . 0.0795 J:::.:\o., ~~:,

FAr 0.040P'* 0.0490 0.0695 FTr 0.0455 0.0526 0.0722

-~-~~-::--.

Peaking Factor No. of Fuel Rods in Assembly

• - -- - - - --- - - ~ -...,.,_

208 216 216 2*16 Reload M Reload N Reload 0 1.66 1.76 Assembly F~ 1.48 1.57 1.92 2.04 Peak Rod F~ 1.92 1.92

- 3-107 Amendment No. ~J5,

~

JJJ, Jft'/J, --1-<<-

Apr i l 3, 1992

• ~

e e POWER DISTRIBUTION LIMITS 3.23.2 RADIAL PEAKING FACTORS

. '* ~ *~ ****~" .

LIMITING CONDITION FOR OPERATION The *;~d*~~*l*, ~e~k*~-~g factors FA, and FT ~~a 11 be less than or equal to the value in Table 3.23-2 times fhe following quantity. The quantity is [l.O +

0.3 (1 - P)] for P ~ .5 and the quantity is 1.15 for P < .5. P is the core thermal power in fraction of rated power.

APPLICABILITY: Power operation above 25% of rated power.

ACTION:

1. For P < 50% of rated with any radial peaking factor exceeding its limit, be in at least hot shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

2. For P ~ 50% of rated with any radial peaking factbr exceeding its limit, reduce thermal power within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to less than the lowest value of:

F

[l - 3.33( r - 1) ] x Rated Power

-r L

Where Fr is the measured value of either F~, or F~ and FL is the corresponding limit from Table 3.23-2.

Bas i.s The limitations on Fe, and F~ are provided to ensure that assumptions used in the analysis for establishiny DNB margin, LHR and the thermal mar~in/low pressure and variable high-power trip set points remain valid during operation. Data from the incore detectors are used for determining the measured radial peaking factors. The periodic surveillance requirements for determining the measured radial peaking factors provide assurance that they remain within prescribed limits. Determining the measured radial peaktng factors after each fuel loading prior to exceeding 50% of rated power provides additional assurance that the core is ptoperly loaded.

Th~ IAc:A Ri"i"::tl*1~~ ~IJl'll'IArh thf!r.iwHal ~akiAa fRetAr Hm~.+c< in Tanl~ 3.?A :>. (

To ensure that the design margin of safety is maintained, the determination radial peaking factors takes into account the appropriate measurement uncertainty factors 111 given in Table 3.23-3 oft References (1) FSAR Section 3.3.2.5 t

. 3-111 Amendment No.~~, ))~, ),7, )~ -l-4-4-'"

ft.gril*~. 1992