ML18057A496

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Proposed Tech Specs Re Cycle 9 Operations
ML18057A496
Person / Time
Site: Palisades Entergy icon.png
Issue date: 09/19/1990
From:
CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
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ML18057A494 List:
References
NUDOCS 9010050121
Download: ML18057A496 (36)
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ATTACHMENT 1 Consumers Power Company Palisades Plant Docket 50-255 CYCLE 9 TECHNICAL SPECIFICATIONS CHANGE REQUEST

  • PAGE CHANGES September 19, 1990 TSP0990-0073-NL04

1.1 REACTOR OPERATING CONDITIONS (Contd)

Low Power Physics Testing Testing performed under approved written procedures to determine control rod worths and other core nuclear properties. Reactor power during these tests shall not exceed 2% of rated power, not including decay heat and primary system temperature and pressure shall be in the range of 260°F to 538°F and 415 psia to 2150-psia, respectively.

Certa~n deviations from normal operating practice which are necessary to enable performing some of these tests are permitted in accordance with the specific provisions therefore in these Technical Specifications ..

Refueling.Boron Concentration Boron concentration of coolant at least *1720 ppm (corr~sponding .to a shutdown margin of at least 5% /:J.p with all control rods w!_t~h~rawn).

Quadrant Power Tilt The difference between nuclear power in any core quadrant and the average in all quadrants.

Assembly Radiai Peaking Factor FA r

The assembly radial peaking factor is the maximum ratio of individual fuel assembly power. to core average assembly power integrated over the total core height, including tilt.

./

I I

I I

Total Interior Rod Radial Peaking Factor - FT r I The-maximum product of the* ratio of individual assembly po...,er to core average assembly power times the highest local peaking 1*

  • factor integrated over the total core height including tilt. I Local peaking is defined as the maximum ratio of the power in an I individual fuel rod to assembly average rod power. I 1-2 Amendment No. 11, a1, $a, $7, llS TSP0990-0073-NL04

_J

  • 2.1 SAFETY LIMITS

- REACTOR CORE (Contd)

DNB will not occur whfch is considered an- appropr-iat=e 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 DNB correlations used in the Palisades safety analysis are listed in the following table.

Safety References Name Limit Correlation Applicability XNB 1.17 1 2 ANFP 1.154 4 5 The MDNBR analyses are perforrpe? in accordance with Reference 6. I References (1) XN-NF-621(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 I 2-2 Amendment No. 11, <<1, 11S TSP0990-0073-NL04

e.

2.3 LIMITING SAFETY SYSTEM SETTINGS - REACTOR PROTECTIVE SYSTEM Applicability This specification applies to reactor trip settings and bypasses for instrument channels.

Objective To provide for automatic protective action in the event that the principal process variables approach a safety limit.

Specification The reactor protective system trip setting limits and the permissible bypasses for the instrument channels shall be as stated in Table ~.3.1 .

The TM/LP trip system monitors core power, reactor coolant maximun inlet *t~mperature, (Tin)' core coolant system pressure and axial shape index. The low pressure trip limit (P ) is calculated using the var following. equation.

I . '

Pvar =)201~2.(QA)(QR 1 ) + 17.0(Tin) - 9493 I wher~:

l'*".

QRl ~ ~o. 412 (q> + o.588 Q ~ 1.0 Q core power

=t6* . Q > LO rated power QA -0. 720 (ASI) + 1 ~028 -0.628 < ASI < -0.100 I

-0.333(ASI) + 1.067 .c.0.100 < ASI < t().200 I

+0.375(ASI) + 0.925 +0.200 < ASI < +o.565 I

  • 1.085 when Q < 0.0625

.The calculated limit (P . ) is then compared to a fixed low pressure var . .

  • trip .limit (Pmin). The auctioneered highest of these signals b.ecomes*

the tri~ liiliit (Ptrip). Ptrip is compared to the measured reactor coolant P.ressure (P) and a trip signal is generated when P is less than o.r equal to Pt . . A pre-trip alarm is also generated when P rip is less than or equal to the pre-trip setting P + ll.P.

trip

  • TSP0990-0073-NL04 2-4 Amendment No. tt$

2.3 LIMITING SAFETY SYSTEM SETTINGS - REACTOR PROTECTIVE SYSTEM (Contd)

Basis The reactor protective system consists of four instrument channels to monitor selected plant conditions which will cause a reactor trip if any of these conditions deviate from a preselected operating range to*

the degree that a safety limit may be reached.

1. Variable High Power - The variable high power trip (VHPT) is incorporated in the rea~tor protection system to provide a reactor trip for transients exhibiting a core power increase starting from any initial power level (such as the boron dilution transient).

The VHPT system provides a trip setpoint no more than .a predetermined amount above the indicate.d core power. Operator action is required to increase the setpoint as core power is increased; the setpoint is.automatically decreased as core power decreases. Provisions have been made to select different set points for three pump and four pump operations.

During normal plant operation with all primary coolant pumps operating, reactor trip is initiated when the reactor power level.

reaches 106.5% of indi~ated rat~d p6wer. Adding to this the possible variation in trip point due to calibration and instrument errors, the maximum actual steady state power at which a trip would.be actuated is 115%, which was used for the purpose of I safety analysis. (l)

2. Primary Coolant System Low Flow - A reactor trip is provided to

~ protect the core against ~NB should the coolant flow suddenly decrease significantly.

3 Flow in each of the four coolant loop$

~s determined from a measu~emeti.t of pressure drop from inlet to outlet of the steam gene"r-ators. The total flow thro,ugh the reactor core is measured by summing *the loop pressure drops across the steam generators and correlating this pressure sum with the pump calibration flow curves. The percent of normal core flow is shown in the. fo~lowing table:

4 Pumps 100.0%

3 Pumps 74. 7%

During four-pump operation, the low-flow trip setting of 95%

insures that the reactor cannot operate when the flow rate is less tht~)93%' of the nominal value considering instrument errors.

  • 2-6 Amendment No it, 11$

TSP0990-0073-NL04

2.3 LIMITING SAFETY SYSTEM SETTINGS - REACTOR PROTECTIVE SYSTEM (Contd)

Basis (Contd)

6. Low Steam Generator Pressure -.A reactor trip on low steam generator secondary pressure is provided to protect against an excessive rate of heat extracti.on 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 excessively high steam flow. This setting was used in the accident analysis. ( 8 )
7. Containment High Pressure - A reactor trip on containment high pressure is provided to assure that the reactor is shut down before
8. Low Power Physics Testing - For low power physics tests, certain tests will require the reactor to be critical at low temperature

(> 260°F) and low pressure(> 415 psia). For these certa-in tests only, the thermal margin/low-pressure, primary coolant flow and low steam*generator pressure trips may be bypassed in order that reactor power can be increased for improved qata acquisition.

    • Special o~erating precautions will be in effect during these tests in accordance with approved written testing procedures. At.reactor power levels below 10 -1 %. of rated. power, the thermal margin I low-pressure trip and low flow trip are not required to prevent fuel rod thermal limits from being exceeded. The low steam generator pressure trip i~ not required because the low steam generator I

pressure will not. allow a severe reactor cooldown~ should a steam line break occur.during these tests.*

References (1) ANF.;_90-078, Table 15.0.7-1 I

.(2) deleted' (3) Updated*FSAR, Section 7.2.3.3.

(4) ANF-90-078, Section 15.0.7-1 I (5) XN-NF-86-91(P)

(6) _deleted (7) deleted (8) ANF-90-078, Section 15 .. 1.S I (9) ANF-87-150(NP), Volume 2, Section 15.2.7 (10) Updated FSAR, Section 7.2.3.9;**

(11) ANF-90-078, Section 15.2.1 I I

2-9 Amendment No 'J,1, HS TSP0990-0073-NL04

3 .1 PRIMARY COOLANT SYSTEM Applicability_

Applies to the operable status of the primary coolant system.

Objective To specify certain conditions of the primary coolant system which must be met to assure safe reactor operation.

Specifications 3 .1.1 Operable Components

a. At .least one primary coolant pump or one shutdown cooling pump with a flow rate greater than or equal to 2810 gpm shall be in operation whenever a change is being made in the boron concentration of tpe primary coolant and the plant is operating in cold shutdown or above, except during an emergency loss of coolant flow situation. Under these circumstances, the bor_on concentration may be increased with no primary coolant pumps or shutdown cooling pumps running.

-,* b. Four-primary coolant pumps shall be in operation whenever the reactor is operated above hot shutdown, with the following exception:

Bef.ore removing a pump from service, thermal power shall be reduced as specified in Table 2.3.1 and appropriate corrective action implemented. With one pump out of service, return the pump to service within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (return to four-pump operation) or* be in hot .shutdown (or below) with the reactor tripped (from the C-06 panel, opening the.42-01 and 42-02 circuit breakers) within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Start-up (above hot shutdown) with less than four pumps is not-permitted and power operation with less than three pumps is not permitted.

c. The measured four primary coolant pumps operating reactor 6

vessel flow shall be 140.7 x 10 lb/hr or greater;*when I corrected to 532°F.

d. Both steam generators shall be capable of performing their heat transfer function whenever the average temperature of the

- primary coolant is above 325°F.

e. Maximum primary system pressure differentials shall not exceed the following:

(1) Deleted 3-lb Amendment No it, S$, ttS, 119 TSP0990-0073-NL04

3 .1 PRIMARY COOLANT SYSTEM (Continued)

-* 3 .1.1 Operable Components (Continued)

(2) Hydrostatic tests shall be cortaucted in accordance- with applicable paragraphs of Section XI ASME Boiler &

Pressure Vessel Code (1974). Such tests shall be conducted with sufficient pressure on the secondary side of the steam generators to restrict primary to secondary pressure differential to a maximum of 1380 psi. Maximum hydrostatic test pressure shall not exceed 1.1 Po plus 50 psi where Po is nominal operating pressure *

.(3) Primary side leak tests shall be conducted at normal operating pressure. The temperature shall be consistent with applicable fracture toughness criteria for ferritic materials and shall be selected such that the differential pressure across the steam generator tubes is not greater than 1380 psi.

(4) Maximum secondary hydrostatic test pressure shall not

  • exceed 1250 psia. A minimum temperature of 100°F is required. Only ten cycles are permitted.

(5) Maximum secondary leak test pressure shall not exceed 1000 psia. A minimum temperature of 100°F is required.

(6) In performing the tests identified in 3.1.1.e(4) and

3. 1. 1. e (5), above, the secondary pressure shall n'ot
  • exceed the primary pressure by more than 350 psi *

. f. Nominal primary system operation pressure shall not exceed*

. 2100 psia.

g. The reactor inlet temperature (indica~ed) shall not exceed the value given by the following equation at steady st~te power operation:

Ti 1 t ~ 542.99 +.. 0580(P-2060) + O.OOOOl(P-2060)**2 + 1.125(W-138) - I n e 0.0205(W-138)**2 I Where: T inlet p

= reactor inlet temperature in F 0

= nominal operating pressure in psia 6

W= total recirculating mass flow in 10 lb/h corrected to the operating temperat.ure conditions.

When* the ASI exceeds the limits specified in Figure 3.0, within 15 minutes, initiate corrective actions to restore the ASI to the acceptable region. Restore the ASI to acceptable values within one hour or be at less than 70% of rated power within the following two hours.

If the measured primary coolant system flow rate is greater than 150 M lbm/hr, the maximum inlet temperature shall be I less than or equal to the TI LCO at 150 M lbm/hr. I n 1 et 3-lc Amendment No Zt, jt, Sj, lt7, llS TSP0990-0073-NL04

3.1 PRIMARY COOLANT SYSTEM (contd)

Basis (Contd)

The FSAR safety analysis was performed ass'um:ing four.primary-coolant pumps were operating for accidents that occur during reactor operation. Therefore, reactor startup above hot shutdown is not*

permitted unless all four primary coolant pumps are operating.

Operation with three primary coolan.t pumps is permitt~d. for a limited time to allow the restart of a stopped pump or for reactor.internals vibration monitoring and testing.

Requiring the plant to be in hot shutdown with the reactor tripped from the C-06 panel, opening the 42-01 and 42-02 circuit breakers, assures an inadvertent rod bank withdrawal will not be initiated by the control room operator. Both steam generators are required to be operable whenever the temperature of the primary coolant is greater than the design temperature of the shutdown cooling system to a~sure a redundant hear removal system "for the reactor *.

Calculations* have been performed to demons.trate that a pressure differential of .1380 psi( 3 ) can be withstood by a tube uniformily thinned. to 36% of its original nominal wall thickness (64% degradation), while maintaining:

(1) A factor ~f safety of thre~ between the actual pressure differential and the pressure differential required to cause bursting._

(2) Stresses within the yield stress for Inconel 600 at operating*temperature.

I .

Accepta~le stresses during accident conditions.

Secondary side hydrostatic and leak testing requirements are consistent with ASME BPV Section XI (1971). The differential maintains stresses in the steam generator tube walls within code allowable stresses.

The minimum temperature of 100°F for pressurizing the steam generator secondary side is set by the NDTT of the manway cover of+ 40°F; The transient analyses were*performeg assuming a vessel* flow at hot zero power (532°F) of 140.7 x 10 lb/hr minus 6% of account I for flow measurement uncertainty and core flow bypass. A DNB

-*analysis* was performed in- a parametric fashion to determine the core inlet temperature as a function of pres~ure and flow-for*

which the minimum DNBR is.equal to 1.17. This analysis includes the following uncertainties and allowances: 2% of rated power for power 3-2 Amendment No t0, $1, llS, tit TSP0990-0073-NL04

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 derivatio~ ?f the following equation for limiting reactor *inlet temperature:

4 f ' *  !'

!inlet S1542.99 + .0580(P-2060) + O.OOOOl(P~2060)**2 + l.~25(W-13~) I

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

1800 S Pregsure s.2200 psia x lb*/h 100.0 x 10 S Vessel Flow S 150 106 I ASI as shown,in Figure 3.0 With measured primary coolant system flow rates > 150 M lbm/hr, I limiting the maximum allowed i~let temperature to th'e Tinlet LCO at; 150 M lbm/hr increases the margin to DNB for higher pcs flow rates. I The ~ial Shape Index alarm channel is being used to monit9r the ASI t<?. 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 calculat~d.from the excore detecfor signals and 1

adj~s~ed fbr shape annealing (YI) and *the core power constitute an ordered pair (_Q,YI). 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 th~

PCS temp~ratur~ when forced circulation is initiated in the PCS ensures that ari 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 c*itculation (the sta?rt of the first primary coolant pump) when the PCS cold leg temperature is < 430°F. However, analysis (Reference 6) s_q9~s that under limited conditions when the Shutdown C.ooling 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 t __Updated FSAR, Section 14. 3. 2.

(2) Updated FSAR, Sectfori 4 .3. 7. -

(3) Palisades 1983/1984 Steam Generator Evaluation and Repair Program Report, Section 4, April 19, 1984 (4) ANF-90-078, Section 15.0.7.1 I (5) ANF-90-078 I (6) Consumers Power Company ~ngineering Analysis EA-A-NL-89-14-1 3-3 Amendment No n' $1" 117' HS' tit TSP0990-0073-NL04

~*

(/)

"d 0

FIGURE 3-0

'°

'° 0 RSI LCO fOR TLnlet FUNCTION I

0 0

.....i VJ

~

0 UNACCLPTABLE

.I:'

OPE:RftllONS

. J.00 3.

a:::

w 3:

0 a... 0.85 Cl w

f-4 a: BREAK POINTS VJ a:::

I t.....

VJ Ill 0 0.70 1. -0.550, 0.25 z 2. -0.300, 0.7 0

f-4 3. -0.080, 1.0

(_)

a: 1. +0.100, 1.0 a:::

t..... 0.55 ACCEPTABLE OPERATIONS D.iD 1

D. 25 L---L.-.____.,_.__..____.___..___.___ __,.___._____.~~--L-- ~--.........._____,___-.L--....L--- ._._~-~ .. - .L-----'- - . -'- *---

-D. 6. -D.i -0.2 0.0 0.2 0.i 0.6 AXIAL SHAPE INDEX

3.10 3.10.6 CONTROL ROD AND POWER DISTRIBUTION LIMITS (Contd)

Shutdown Rod Limits

a. All shutdown rods shall be withdrawn before any regulating _

rods are withdrawn.

b. The shutdown rods shall not be withdrawn until normal water level is established in the pressu_rizer.
c. The shutdown rods shall not be inserted below their exercise limit until all* regulating rods are inserted.

3.10.7 Low Power Physics Testing Sections 3.10.1.a, 3.10.1.b, 3.10.3, 3.10.4.b, 3.10.5 and 3.10.6 may be deviated from during low power physics testing and CRDM exercises if necessary to perform a test but only for the time necessary to perform the test.

3.10.8 Center Control Rod Misalignment The requirements of Specifications 3.10.4.1, 3.10.4.a, and 3.10.5 may be suspended during the performance of physics tests to determine the isothermal temperature coefficient and power coefficient provided tha~ only the center control rod is misaligned and the limits of Specification 3.23 are maintained.

Basis Sufficient control rods shall.be withdrawn at all times to assure that the reactivity decrease from a reactor trip provides adequate

,shutdown margin. The available worth of withdrawn rods must

'include the reactivity defect of power and the failure of the withdrawn*rod of highest worth to insert. The requirement for a shutdown margin of 2 .0% in reactivity with 4-pump o.peration, and of 3.75% in reactivity with less than 4-pump operation, is consistent with the assumptions used in the analysis of accident conditions (including steam line break) as reported in Reference 1 and additional analysis. Requiring the boron j concentration to be at cold *shutdown boron concentration at less than hot shutdown assures adequate shutdown margin exists to ensure a return to power does not occur if an unanticipated cooldown accident occurs. This requirement applies to normal operating situations and not during emergency conditions.where it is necessary to perform operations to mitigate the consequences of an accident. By imposing a minimum shutdown cooling pump flow rate of 2810 gpm, sufficient time is provided for the _opexator to terminate a boron dilution under asymmetric conditions. For operation with no primary coolant pumps operating and a recirculating flow rate less than 2810 gpm the increased shutdown margin and controls on charging pump operability or alternately the surveillance of the charging pumps will ensure that the acceptance criteriflYor. an inadvertent boron dilution event will not be violated. The change in insertion limit I with reactor power shown on Figure 3-6 insures that the shutdown 3 Amendment No it, $4, $7, ~S. ttS TSP0990-0073-NL04

3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS (Continued)

Basis (Continued) margin requirements for 4-pump operation is met a~ all power levels. The 2.5-second drop time specified for the c~yyrol ro.ds is the drop time used in the transient analysis. I The insertion of part-length rods into the core, except for rod exercises or physics tests, is not permitted since it has been demonstrated on other CE plants that design power distribution envelopes can, under some circumstances, be violated by using part-length rods. Further information may justify their use.

  • Part-length rod insertion is permitted for physics tests, since resulting power distributions are closely monitored under test conditions. Part-length rod insertion for rod exercises (approximately 6 inches) is permitted since this amount of insertion has an insignificant effect on power distribution.

For a control rod misaligned up to 8 inches from the remainder of the banks, hot ch.annel factors will be well within design limits. If a control rod is misaligned by more than 8 inches, the maximum reactor power will be reduced so that hot channel factors, shutdown margin and ejected rod worth limits are met.

If in-core detectors are not available to measure power distribution and rod misalignments >8 inches exist, then.

reactor pow.er must not exceed 75% of rated power to insure that hot channel conditions are met .

  • Continued operation with that rod fully inserted will only be permitted if the hot channel factors, shutdown margin and ejected rod worth limits are satisfied.

In the event a withdrawn control rod cannot be tripped, shutdown margin requirements will be main~ained by increasing the boron concentration by an amount equivalent in reactivity to that control rod. The deviations permitted by Specification 3.10.7 are required in order that the control rod worth values used in the reactor physics calculations, the plant safety analysis, and the Technical Specifications can be verified. These deviations will only be in effect for the time period required for the test being performed. rhe testing interval during which these deviations will be in effect will be kept to a minimum and special operating precautions will be in effect during these deviations in accordance with approved written testing procedures~*

  • TSP0990-0073-NL04 3-63 Amendment No. it, ~J, j7, ~$, ¥1$

3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS (Continued)

Basis (Continued)

Violation of the power dependent insertion limits, when it is necessary to rapidly reduce power to avoid or .minimize a situation harmful to plant personnel or equipment, is acceptable due to the brief period of time that such a violation would be expected to exist, and due to the fact that it is unlikely that

  • core operating limits such as thermal margin and shutdown margin would be violated as a result of the rapid rod insertion. Core thermal margin will actually increase as a result of the rapid rod insertion. In addition, the required shutdown margin will most likely not be violated as a result of the rapid rod insertion because present power dependent insertion limits result in shutdown margin in excess of that required by the safety analysis.

References (1) ANF-90-078 '** I I

I

  • 3-64 Amendment No. ~S. 118 November 15, 1988 TSP0990-0073-NL04

3.12 MODERATOR TEMPERATURE COEFFICIENT OF REACTIVITY

  • Applicabili-ty-Applies to the moderator temperature coefficient of reactivity for the core.

Objective To specify a limit for the positive moderator coefficient.

Specifications The moderator temperature coefficien~ (MTC) shall be less 4

positive than +o.5 x 10- t::,p/°F at < 2% of rated power:.,

Bases The limitations on moderator temperature coefficient (MTC) are provided to ensure that the assumpt.ions used in the safety analysis (l) remain valid.

Reference (l)ANF-90-078, Section 15.0.5 I 3-67 Amendment No. llS (next page is 3-69)

TSP0990-0073-NL04

.J!..

e

~

TABLE 3.23-1

--* LINEAR HEAT RATE LIMITS No. of Fuel Rods in Asiie'mbly t

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 Assembly FAr 1.48 1.57 I Peak Rod *T Fr 1.92 1. 92 I TABLE 3.23-3 POWER DISTRIBUTION MEASUREMENT UNCERTAINTY FACTORS LHR/Peaking Factor Measuremenf ) Measurementb) Measurement(c)

Parameter Uncertainty a . Uncertainty Uncertainty LHR 0.0623 0.0664 . 0.0795 FA 0.0401 0.0490 0.0695 r

FT 0.0455 0.0526 0. 0722 I r*

(a) Measurement uncertainty for reload cores using all fresh incore detectors.

(b) Measurement uncertainty for reload cores using a mixture of fresh and once-burned incore detectors *

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

3-107 Amendment No. ~$, tlS, TSP0990-0073-NL04

.l, POWER DISTRIBUTION LIMITS 3.23.2 RAD*IAL PEAKING FACTORS LIMITING CONDITION FOR OPERATION The radial peaking factors ~, and FT shall be less than or equal to r r the value in Table 3.23-2 times the following quantity. The quantity is [1.0 + 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 r~ted 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 />.

~. 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

[l - 3.33 .L.£ - 1) ] x Rated Power FL

  • Where F.r is the measured value of either ~,

is the corresponding limit from Table 3.23-2.

Basis r

or FTr and F L /-

The. limitations on FA, and FT are provided to ensure that assemptions I r r used in the analysis for establishing 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 *

. The radial peaking is limited to those values used in the LOCA analysis. I Since the LOCA analys.is limits the magnnude of radial, peaking, Table 3. 23-2 /

explicitly contains these limits. --/

3-111 Amendment No. ~S. ttS TSP0990-0073-NL04

  • 4.19 4.19.2 POWER DISTRIBUTION LIMITS RADIAL PEAKING FACTORS SURVEILLANCE REQUIREMENTS

. ~ T 4.19.2.1 The measured radial peaking factors (r-, and F ) I obtained by using the incore detectionrsystem,rshall be determined to be less than or equal to the values stated in the LCO at the following intervals:

a After each fuel loading prior to operation above 50% of rated power, and l:?* At least once per week of power operation

  • 4-84 Amendment No ~S. 118 November 15, 1988 TSP0990-0073-NL04

ATTACHMENT 2 Consumers Power Company Palisades Plant Docket 50-255 cYCLE 9 TECHNICAL SPECIFICATIONS CHANGE REQUEST MARKED UP PAGES September 19, 1990

  • TSP0990-0073-NL04

1.1 REACTOR OPER.AT'INC CON'D!T!ONS (Contd) tow Power Physics Testing Testing perfonned under approved lo(t"itten procedures to determine control rod worths and ocher core nuclear properties. Reaccor power during these tests shall not exceed 2% of raced power, noc including decay heat and primary system temperature and pressure shall be in che range of 260 9 F to SJ8 9 F and 415 psia to 2150 psia, respectively.

,, Certain deviations fr01ll normal operating practice which are necessary to enable performing some of these tescs are pennitced in accordance vtch ~he specific provisions therefore in these Technical Spec1f icac1ons.

Shutdown Boron Concentrations Boron concen,tration sufficient co provide keff ~ 0.98 with all control rods in the core and the highest vortfi control rod fully withdrawn.

Refueling Boron Concentratio~

Boron concentration of coolant at least 1720 ppm (corresponding to a shutdown margin of at least 5% Ap "'1th a~l control .rod~ withdrawn).

Quadrant Pover Tilt The difference between nuclear pover,in any core quadrant and the

.;tverage in all quadrants.

Assembly Radial Peaking Factor - F A r

The assembly radial peaking factor is the maximum*ratio of individual fuel assembly power to core a~erage assembly power integrated over the total core height, including tilt.

I on that assembly's I I ~

The maximum product of the ratio of individual asaembly power to

.\,_

core aver*s* aasembly power times the hishest ..ia.cerictr' local peaking factor integrated over th* total core heiaht including tilt. l..oc.u/

I

  • f.eoJ:iho 1$ d~1~-e..O a..o +k ,..,,.,...a..,,,;....... v,....,,...... 1-0110 o.6+1..4... pow.JV\.'~~

l'°"d'*v;oue\ ~v-f\ \-cd-to t1S~:i- .bl '1 a..v~cic.-4... >--o~ Po......,~.

TSP1088-0181-NL04 Amendment No. 1t, IJ, JI, J7, -H Nofember lJ, 1~88

  • 7: 1 a FROM CPCO PALISADES PAGE.006 Z.1 SAP'!:TY t.IMITS - ltF.ACTOR CORE (Contd)

DNB ~111 not occur which is conaidered an appropriate margin to DNB for all operatin1 conditions.

!he reactor p~otective system is dea1gned to prevent any anticipated combination of tranaient conditions for primary coolant system temperature, preseure and thermal power level that would result in a DNBR of lea* thaa the DNB correlation tafety limit. The ONB correlations use<l in the Pal1sad** aafety analysis are liated in th* followin1 table.

Safety, Reterenc:ea Limit Correlation Applicability

  • xn i.11 *1 2 ANFP 1. 1.54 , . 4 5

~The MCJNliR. unalys,f5 o..v-c..pt.r~ormcc9  ; ""' l&. C. (O rclcu--. l~ U.* *~ h f?.. f,S-erenc.<L. ~

""' Reterenc**

(1) XN-NF~2l(P)(A), ~ev l .*

(2) XN-NF-709 (3) Updated FSAR, S*ct1on 14.1.

(4) AMF-1224 (P) (A), May 1989 (5) ANF~89-192 (P), January 1990 t_

(6) X,N-1'11*:~ 8 l ~ '2. /(A)) R e..vi"51of) ~ I

-rh~ 9.f;.,~* ~1u:~ Odt p1>ep&o=~o1 -Po eJ,l,,S-

  • pet6-e.. C>~ "'1f'i'iG ed 6' e e El
  • e\ ": *6 =t=-0 C hon~<BO rrno ~To ~-11S- pq~-. ~v~

~of ~a??C-DCJ p~os.J) d oY"> f /31 jc/o

); o-p--e.. j, ~ o...p p J-tW-1' cl 0--n cl ; s .s {/ ~d '.

Amendment No. JI, IJ, ltl .

\\

2.3 LIMITING SAFETY SYSTEM SETTINGS

  • REACTOR PROTECTIVE SYSTEM Applicability This specification applies to reactor trip settings and bypasses for instrument channels.
  • Objective To provide for automatic protective action in the event that the principal process variables approach a safety limit.

Specification The reactor protective system trip setting limits and 'the permissible bypasses for the instrument channels shall be as stated in Table 2.3.l.

The TM/LP 'trip system monitors core power, reactor coolant maximun i~let temper~tur_e, (Tin), core coolant system pressure and axial shape index. The low pressure trip limit (~var) is calculated using the following equation.

2 o /. 2

  • I ?. 0 4 ~ q3 Pvar * .l.56l,?(QA)(QR 1) +~(Tin) -~563.~ #f I where:

QR 1

  • 0.412(Q) + 0.588 Q S 1.0 Q
  • core power
  • Q Q > 1.0 rated power {
  • 0 *? ZO - O
  • lol8 - O
  • I oo QA * -Q......i'1(ASI) + ~ t.IJ21 -0.~ < ASI < -0 *.,L.H-

-o. 333 * -a...Hi(ASI) + ~ l 0 oe?7~<

+O.J7S * ..0 2i6(ASI) + Q..i&40*'flS

  • 1.085 when Q < 0.0625

.~!

  • 200 ASI < +o.J,.iiO.l.00 ASI ~ +o.)j44()*S'S '~1

. I

- *I oo

'nle calculated limit (Pvar> is then compared to a fixed lov pressure trip liait (Pmin). The auctioneered highest of these signals becomes the trip liait

. * (Pt r p}.* Pt r i p is compared to the measured reactor 1

coolant preasure (P) and a trip signal is generated when P, is less than or>equal to P i

  • A pre-trip alarm is also generated when P tr p is. lesa than or equal to the pre-trip setting Ptrip +AP.

*-** Amendment No. -H-&-

Nov. .he* U 1 1988 -

\7-2.J LIMITINC SAFETY SYSTEM SETTINGS - REAC"t'OR PROTECTIVE SYSTEM (Coned)

Basis The reactor protective system consists of four instrument channels co monitor selected plan~ conditions which will cause a reactor trip if any of these conditions deviate from a preselected operating range co che degree that a safety limit may be reached.

1. Variable High Power - The variable high power trip (VliPT) is incorporated i~ the reactor protection system to provid~ a reactor trip for transients exhibiting a core power increase starting from any initial power level (such as the boron dilution transient).

The VHPT system provides a trip setpoint no more than a predetermined amount above the indicated'core power. Operator action is required to increase the setpoint as *core power is increased; the setpoint is automatically decreased as core power decreases. Provisions have been made to select different sec points for three pump and four pump operations.

During nol"11lal plant operation with all primary coolant pumps operating, reac~or trip is initiated when the reactor power level reaches 106.5% of indicated rated.power. Adding to this the

  • possible variation in trip point due to.calibration and instrument errors, the maximum actual steady state power at which a trip would be actuated is j..l.-~ which was used for the purpose of (1) ~...

safety analysis. /I 5 ~

2. Primary Coolant System Low.Flow - A reactor trip is provided to 1:....

protect the core aga1nsb9Nll should the coolant flow suddenly decrease significantly. Flow in each of the four coolant loops ~

is determined from a measurement of pressure drop from inlet to outlet of the steam generators. The total flow through the reactor core ie meaaured by summing the loop pressure drop*s across the ateaa geaeratora and correlating this presaura sum with the pua, calibration flov curves. The percent of normal core flow is abOVD in th* follovtna table: 4:.

4 Pumps 100.0%

3 Pumpa 74. 7%

During four-pump operation, th* low-flow trip setting of 95%

insures that the reactor cannot operate when the f lov rate is lesa tht!>93% of the nominal value considering instrument errors. +

2-6 Amendment No Jl, tte-Neveabn U, l988 TSPlOE

LIMITING SAFF.TY SYSTEM SETTINGS - REACTOR PROTECTIVE Basis {Contd)

6. tow Steam Generator Pressure - A r.eactor t:'rip on low steam

~enerator secondary pressure is provide4 co protect against: an SYSTE~ (Contd) 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 excessively high steam flow. This setting was used in the accident analysis. (S)

7. Containment High Pressure - A reactor trip on containment high pressure is provided to assure that the reactor is shut down before.

the inHoftion of the safety* injection system and containment t

spray.

8. Low P.ower Physics Testing - For low power physics tests* certain tests will require the reactor to be critical at low temperature
  • (> 260.F) and low pressu.re {> 415 psia). For these certain tests .,

. only. the thermal margin/low-pressure. primary coolant flow and ~

low steam generator pressure trips may be bypassed in order that reactor power can be increased for improved data acquisition.

I Special operating precautions will be in effect during th~se te~~s in accordance with approved written testing procedures. At reactor

. ' . -l ,. ' '

power levels below 10 % of '.raced power*, the thermal inargin/low-pressure trip and lov flov trip are not required to prevent fuel -f rod* thermal limits from being exceeded. The low steam gener'ator pressur~ trip is not required because the low steam generator pressure vill not allow* a severe reactor cooldown.- should -~ steam line break occur during these tests, References A"1F-'7c -078

,I (l) ,AffP-tt-i"SOtNPT;-Voiume--2', Table 15.0. 7-1 (2) deleted i:.

(3) Updated PSAll, Section 7.2.3.3 * * . ==-- ~

(4) AMF 87=1SO(lfP), Volam~:~aa 15:3~-':>o -orv_; I (S)

(6)

(7)

XJl-Nl'-86-91 (P)

  • deleted claleted .

,Se<..tl O 11 15.o.?-/

  • *5 \ c j..__:

(8)

  • NI' 17 181 Seelien J ..a Al'.lF OIB) "'::><ch~ I ..... I (9) dl'-87-15.0(NP), -Volume 2, Section lS.2. 7 ~
  • ~*-

(10) Updated FSA!l, Section 7.2.3.9.

( 11) .ANP 87=lSO(Nr) 1 \'elume-2-. ection 15. 2. l ~::.:;....-.::.::-\~-=- /

(12) Al'h-87-l"O(HP)' '!elwa~l-,-See-Mon-l...>.-0-.-1-. I#

1 Amendment No 11, -+te-

-Nouaaber lS, 1988

?SP1088-0l81-Nt.04

'v 3.1 PRIMARY COOLANT SYSTEM Applicability Applies to the operable status of the primary coolant system.

Objective To specify 'certain conditions of the primary coolant system which must be met to assure safe reacto~ 6peration.

Specifications

3. l. l Operable Components
a. At least one primary coolant pump or one shutdown cooling pump with a flow rate greater than or equal co 2810. gpm shall be in operation whenever a change is being made in the boron concentration of the primary coolant and the plant is operating in cold shutdown or above, except during an emergency loss of coolant flow situation. Under these circumstances, the boron concentration may be increased with no primary coolant*pumps or*shutdown. cooling* pumps running,'
b. Four primary coolant pumps shclll be,in operation whenever the reactor is operated ab.ove hot shutdown, with the
  • following exception:
  • Before removing a pump from service, thermal power shall be.

reduced as specified in Table 2.J.l and appropriate corrective action implemented. With one pump ~ut of service, return.the pump to service within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (return to four-pump operation) or be in hoc shutdown (or below) with the reactor tripped (from the C-06 panel, ope~ing the 42-01 and 42-02 circuit breakers) within the next 12 hou.rs. Start-up' (above hot shutdown) with less th~n four pumps is not permitted and power operation with less than* three pumps ~s. not permitted*.

c. The measured four ~~....._~c=:_:o~olant pumps operating reactor vessel flow sha b/hr or greater, when corrected to 53 ° * "
d. Both steam generatQrs shal be c~pable of performing their heat trans.fer function whenever the average temperature of the primary coolant is above 325°F.
e. Maximum primary. system pressure diffe,rentials shall not exceed the following:
  • (l) Deleted 3-lb Amendment No*Jf, SJ, !ZS.~

~eeembet 12, 1988 TSP1088-0181-NL04

iJ

\.S

3. l PRIMARY COOLANT SYSTEM (Continued)
3. l. l Operable Components (Continued)

(2) Hydrostatic tests shall be conducted in accordance with applicable paragraphs of Section XI ASME Boiler & I Pressure Vessel Code (1974). Such tests shall be conducted with sufficient pressure on the secondary side of the steam generators to restrict primary to secondary pressure differential co a maximum of 1380 psi. Maximum hydrostatic test pressure shall not exceed 1.1 Po plus 50 psi where Po is nominal operating pressure.

(3) Primary side leak tests shall be conducted at noI'1!1al operating pressure. The temperature shall be consistent with applicable fracture toughness criteria for f erricic materials and shall be selected such chat the differential pressure across the steam generator cubes is not greater than 1380 psi.

(4) Maximum secondary hydrostatic test pressure shall not exceed 1250 psia.

  • A minimum temperature of 100°F is required. Only ten cycles are permitted.

(5) Maximum secondary leak test .pressure shall not exceed 1000 psia. A minimum temperature of 100°F is required.

(6) In performing the tests identified in 3.l.l.e(4) and 3.1.1.e(S), above, the secondary pressure shall not exceed the primary pressure by more than 350 psi.

f, Nominal primary system operation pressure shall not exceed 2100 psia *.

g. The reactor inlet temperature (indicated) shall not exceed the value given by the following equation at steady state power operation:
  • S'fZ , qq . OSiO Ti l t S ~ .J. + ._os..rs'(P-2060) I n e ~(W-J..le?**2 .

- 0.02.0S 131 0

/l Where: Tinlet

  • reactor inlet temperature in F P
  • nominal operating pressure in psia W
  • total recirculating mass flow in 10 6 lb/h corrected to the operating temperature conditions.

When the ASI exceeds the limits specified in Figure 3.0, within 15 minutes, initiate corrective actions to restore the ASI to the acceptable region. RestQre the* ASI to accep.table values within one hour or be at less than 70% of rated power within the following two hours.

If the measured primary coolant system flow rate is greater than ..µoM lbm/hr, the maximum inlet temperature shall be lesslt?an or equal to the TI l LCO at ~3<f M lbm/hr.

1 n et 150

/?0 3-lc Amendment No 11, Jl, JJ, 117.

TSP1088-0181-NL04

\'4,

'.3. l ~RIMARY COO'L\NT SYSTEM (contd)

~ (Coned) rhe FSAR safety an alysfis was dperformehd assumingd fo ur primary coolant ~

pumps were operat 1 n~ or ac~ 1 ents t at occur ur 1 ng reactor ~

operation. Therefore, reactor startup above hot shutdown is not permitted unless all four primary coolant pumps are operating:

Operation with three primary coolant pumps is permitted for a limited time to allow the ~estart of a stopped pump or for reactor i: ~rnals vibration monitoring and testing.

Requiring ~~e plant to be in hot shutdown with the reactor tripped from the C-06 panel, opening the 42-01 and 42-02 circuit breakers, assures an inadvertent rod bank withdrawal will not be initiated by the control room operator. Both steam generators are required to be operable wheneve~ the temperature of the primary coolant ~s greater than the design temperature of the shutdown cooling system co assure a redundant hear removal syst~m for the reactor.

r.alculations have beer. perf onned to demonstrate chat a pressure differential of 1380 psi(J) can be withstood by a tube unifortnily thinned to 36% of its original nominal wall thickness (64% degradation), while maintaining:

(1) A factor of safety of three between the actual pressure differential and the press~re differential required to cause bursting.

(2)

(3)

Stresses within the yield stress operating temperature.

.f~r Inconel 600 at Acceptable stresses during accident conditions.

Secondary side hydrostatic and leak testing requirements are consistent with ASME BPV Section XI (1971). The differential maintains stresses in the steam generator cube walls within code allowable stresses.

The minimum temperature of 100°F for pressurizing the steam generator secondary side is set by the NDTl' of the manway_ cover of + 40°F.

Th* transient analyses ssuming a vessel flow at hot zero power (532°F) lb/hr minus.6% of account /~

for flow measurement uncertainty and core flow bypass. A DNB analy1ia was perfol'1!1ed in a parametric fashion co determine the core inlet temperature as a function of pressure and flow for which the minimum DNBR is equal co 1.17. This analysis includes the following uncertainties and allowances: 2% of rated power for power

  • TSP0889-0101-MD01-NL0' 3-2 Amendment No

. ..tp'FU 26, UCJ6 2~,

Jt, 111, 1'7t--

r.t

\~

3.1 PRIMARY COOLANT SYSTEM (Cont'd)

Basis (Cont'd) measurement; :0.06 for ASI measurement~ :SO psi*for pressurizer pressure; :7°F for inlet temperature; an~ 3% measurement and 3%

bypass for core flow. In addition, transient biases were included in temperature: 4 .

the derivatio~ 7f t~e following equation for limiting reactor inlet 5'12. 'l'I -1 o. oseo. *ocJoo/ I* 1zr 13'0 T inlet ~ ~ + .~(P-2060) + O.~(P-2060)**2 + .l. L7l(W-¢) -4* I

. o-ttJ7 (W-1.,ld') **2 '?\/

  • *C~O~ /~fi The limits of validity of this equation are:

1800 ~ Pre~sure ~ 2200 psia *

  • 6 100 .0 x 10 S Vessel Flow S ;.a6 10 lb/h ASI as shown in Figure 3.0 1:;;-v .

,.;-"

With measured primary coolant system flow races > ,..l-3~M ll~m./hr,  %. /

limiting the maximum allowed .inlet temperature to the Tinlet LCO at b?> _;..ac('M' lbm/hr increases the margin to DNB for higher PCS flow races. /

The Axial Shape Index alarm channel is being used to monitor the ASI to ensure that the assumed axial power profiles used in the

.

  • developm~nt of the inlet temperature LCO bound measured axial power profiles. The signal representing core power (Q) is the aucti~~eered higher of the neutron flux power and the Delta-T power.

The measured ASI calculated from the excore detector signals and adjusted for sh~pe annealing (YI) and.the core power constitute an ordered pair (Q,Y!). An alar:m s~gnal is activated before the ordeted pair exceed the boundarie~ specified in Figute J!O.

The requirement that the steam generator temperature be S the

~CS temperature when forced circulation is initiated in the PCS ensures that an energy addition caused by heat transferred from the seconda~y system to the PCS will not occur. This requirement applies only to the initiation of forced circulation (the stare 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 hi.gher than the PCS cold lea temperature.

Reference*

(1) Updated FSAR., Section 14.3.2.

(2) Updated FSAll, Section 4.3.7.

(3) Palisades 1983/1984 Steam Ce~or Evaluation, and Repair Program Report, Sec ti on Y,-April 19, 19816--. .

(4) lcNP 87 tSe(MP), TJolWD~, Section 15.0.7.l. A~*f"-9v-<;;7~ fo..-..1 (5) AHP 88-108" ,At-i(- *9u -0~7E::, I (6) Consumers Power Company Engineering Analysis EA-A-NL-89-14-1 T J-3 Amendment Noll, J!, 117. l!S,-+3+-

-Apttl 26, 1990 TSP0889-0101-MD01-N'L04

FIGURE 3-0 ASI LCD FOR TLnlel FUNCTION 1.15 UNRCCEPTRBLE OPERRTIONS 1.00 3 1 Q:

w 3:

0 0.. o.as Ul Cl w

I .....

a:

lN Q:

BREAK.POINTS f ~

0.70 0

2 1." -0.550,, '0.25 z 2. -o.3oo; o.7 0

u

3. -0.080, 1.0 a: 1. +0.100; 1.0 Q:
  • ~ 0.55 ACCEPTABLE OPERATIONS O.iO e o.25_oL.6____.L~~--...J.o.-.. ~--:- SH~~E

' *--L.. . _.__._ ....___ _.___...J.___~ -

~_J___.._ __ ~_.__

~---L- _.... ---

-- 0.2

-0.2 AX I AL I NOEX O.i 0.6

J.10 CONTROL ROD AN'D POWER DISTRIBUTION t!~!TS (Contd)

J.10.6 Shutdown Rod Limits

a. All shutdown rods shall be withdraWT\ before any regulating rods are withdrawn.
b. The shutdown rods shall not be withdravn until no't"tnal water level is established in the pressurizer.

c:

  • The shutdown rods shall not be inserted below their exercise limit until all regulating rods are inserted.

Low Power Physics Testing Sections J.10.l.a, J.10.l.b, J.10.J, 3.10.4.b, J.10.S and J.10.6 t may be deviated from during low power* physics testing and CRDM exercises if necessary to perform a test.but only for the time necessary to perform the test.

J.10.8 Center Control Rod Misalignment Th* requirements of Specifications J.10.4.l, J.10.4.a, and J.10.S may be suspended during the performance of physics tests "to determine the isothermal temperature coefficient and power coefficient provided that only the center control rod is misaligned and the limits of Specification J.23 are maintained.

Basis Sufficient control rods shall be withdrawn at all times to assure that the reactivity decrease from a reactor trip provides adequate shutdown D1argin.

  • The available worth of wit.hdravn rods must include the reactivity defect of power and the failure of the *
  • withdrawn rod of highest worth to insert. The requirement for a shutdown margin of 2.0% in reactivity with 4-pump operation, and of J.75% in reactivity with less than 4-pump operation,* is consistent With the assumptions.used in the analysis of accident conditions (inciuding steam line break) as reported in ,.

Reference l~and: addition*! analysis. Requiring the boron concentration to be at cold shutdoVll boron concentration at . ,,/~

less thanhot shutdOV!l assures adeq~ate shutdown-margin exists to ensure a return to power does not occur if an unanticipated cooldovn accident occura. Th_ia requirement applies to normal operating situation* and not during emergency conditions where it i* neceaaary to perfot'11l operations to mitigate the couequencea of an accident. By imposing a minimum shutdown coolSaa pump flov race of 2810 Sl)m, suf~icient time is provided for die operator to tenainat* a boron dilution under asymmetric condiC101l8. For operation with no p~imary coolant pumps operating and a recirculat1na flov rate 1*1* ch&D 2810 gpm the increased shutdown margin and . controla on charging PWll>> operability or

  • alteruately th* surveillance of Che char1in1 pump* will ensure that Che acc'!ptance criter*1f jor an inadvertent boron dilution event will not: be violated. 11 The change in i~aertion limit with reactor power shown on Figure 3-6 insures that the.shutdown J-61 Amendment No Jl, JI, J7, U, H-8.

November 15, 1988 TSP1088-0181-Nt.04

  • 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS (Continued).

!!!..!..! (Continued) margin requirements for 4-pump operation is met at all power ldvels. The'2.5-second drop time specified for the c~~~rol rods ts the drop time used in the transient analysis. 1 /~

The insertion of pare-length rods into the core, except for rod exercises. or physics cests, is not permitted since 1c, has been demonstrated on othe~ CE planes that design power d1str1b~t1on envelopes can, under ~ome circumstances, be violated by using pare-length rods. Further information may justify their use.

Pare-length rod insertion is permitted for physics tests, since resulting power distributions are closely monitored under test conditions. Pare-length rod insertion for rod exercises.

(approximately 6 inches) is petillitted since this amount of insertion has an insignificant ~ffect on power distributior\.

For a control* rod misaligned up to 8 inches fr01ll the. remainder of ~he banks, hot channel factors will be well within design limits. If a control rod is misaligned by more than 8 inches, the .. maximum reactor power will.be reduced so that hot 'channel:

factors, shutdown margin and ejected rod worth limits are met.

If in-core detectors are not available to measure*power distribution and rod misalignments >8 inches exist, then reactor power muscnoc exceed 75% of rated power to insure that hot channel conditions are met.

Continued operation wich* that rod fully inserted will only be permitted if the hot channel factors, shutdown margin and ejected rod worth l,imits are satisfied.* ... *

. In the event a withdrawn control rod cannot be tripped, shutdown margin requirements will be ma~ntained by increasing the boron concentration by an amount equivalent in reactivity.to that co_ntrol rod. The deviations permitted by Specification 3.10. 7 are required in order that the control rod worth values used in the reactor phy*ica calculations, the plant safety analysis, and the Technical Specifications can be verified. These deviations will only be iu effect for the time period required for the test be1D1 perforaed. The testing interval during which these deviation* will be in ef f ecc will be kept to a minimum and special operatilll precaution* will be in effect during these deviations in accordance with approved Vl"itten testing procedures.

3-63 Amendment No. ll. l:J. J7, U, -H:-s-

. TSP1088-0l81-N1.04

J,

. 3 .10 CONTROL ROD AND POWER DISTRIBUTION tIMITS (Continued)

Basis (Continued) .

  • Violation of che power dependent in~ertion limits, when ic is ,

necessary co rapidly reduce power co avoid or minimize a situation ~armful co plant personnel or equipment, is acceptable due co che brief period of cime chac such a violacion would be expected co exist. and due co the face chac ic is unlikely chac core operating limits such as chennal margin and shucdown margin would be violated as a result of the rapid rod insercion. Core thermal margin will actually increase as a resulc of che rapid rod insertion. In addition, the required shucdovn margin will most likely not be violated as a result of che rapid rod insertion because present power dependenc insertion limits result in shutdown margin in excess of chat required by the safety analysis.

  • References (1) XN-NF. 77 t-8 At-J F - 9 o - C>7 E'J I (2) AN~~oium~ I

( J) AN-'F-=88w 108- I 3-64 Amendmenf' v ..

  • TSP1088*0181-Nt.04

3.12 MODERATOR TEMPE~TURE COEFF!CIEN'l' OF REACTIVITY Applicabilicy Applies co che moderacor cemperacure coefficienc ~f reaccivicy for che core.

Objeccive To specify a limit.* for the positive moderator coef ficienc.

Specif !cations The moderator temperature coefficient (MTC) shall be less

. -4 positive than +o.5.x 10 t.o/°F.at < 2% of raced power.

Bases The limitations on moderator temperature coefficient (MTC),

  • are provided co ensure chat the assumptions used in che safety analysis (l) remain valid.

Reference (l)ANF-8-~o.{NP), 'Vctlume-h-Section 15.0.5 ANF*-90..:. c 1 B 3-67 Amendment No.~

Nevember l,, 1988 s 3-69)

TSP1088-0l81-Nl.04

.e -----

_ TABLE 3.23-1 LINEAR HE;AT RATE LIMITS No. of Fuel Rods in Ass.einoly 208 216 TABLE 3.23-2

~AOIAL PEAKING FACTOR LIMITS,FL Peaking Factor No. of Fuel, Rods in Assembly 208 216 Assembly .1.48  :-5& /. s. 7 I

-+/--=-ffl' I. 9:i.. * .1. 73- /, 9L- /

TABLE *3:23-3 POWER DISTRIBUTION MEASUREMENT UNCERTAINTY FACTORS

LHR/Peaking Factor Measuremenf ) Measur~men.fb) Measurement'( )

Parameter Uncertainty a Uncertainty Uncertainty c LHR 0.0623 0.0664 0.0795 FA 0.0401 0.0490 0.0695 r I I F4' 0.0455 0.0526 0.-0722 r

(a) Measurement uncertainty for reload cores using all fresh incore detectors.

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

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

3-107 Amendment No. fS, 118,

POWER DISTRIBUTION LIMITS

  • 3.23.2 RADIAL PEAKINC FACTORS tIMIT!NC CONDITION FOR OPERATION The radial peaking factors r", and I

~shall be less than or equal to /~

. r r the value in Table J.23-2 times the following 'quantity. The quantity is [1.0 + O.J (1 - P)) for P ! .5 and che q~ant1cy 1s 1.15 for p < .5. P is the core che?'11lal 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 ita 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 ltiit. 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:*

[l - 3.33 (Fr - 1) x Rated Power FL* '*

I

  • r
  • Where *r r is the measured value of either ~, or F~H and FL I ¥£_

is the corresponding limit from Table 3.23-2.

Basis The limitations on ~' and F~ are provided to ensure that assumptions uaed~in the analy*i* for* eetablishing DNB margin, LHR aud che thennal margin/lov-preaaure and variable high*p*ower trip set points remain vali.d during operation. Data fraa th* incore detectors are used for determining the meaaured radial peaking factors. The periodic surveillance requirements for detenainiug the measured radial peaking factors provide assurance.that they r ...ill within prescribed limits. Detentining the meaaured radial peakilll factors after each fuel loading prior to exceeding 50% of rated power provide* additional assurance that the core is properly loaded.

'lht...ro..cliq,,I Pe..ct~in9 is Ji'rf1it:.e..P -to -bho.se.. va.lv~S u~eJ iVl.bhlloGA I

~r)al'JSis. $1.nle.. t.h~ LOCA a..\'\ll.t~.s1s /imito th<?.ma.9n1.t;ud-e. I D~ r~J..iCA-1 Pta.K:1t"\9) /able, 3.23-z ~-x:p1,*cd:;J~ c. oni-a1ns t.hebe. l JI h'1 / t§. . I I J-111 Amendment No. '8, -tt-s-Newem~er 15, 1988 TSP1088-0181-ML04

4.19 POWER DISTRIBUTION LIMITS 4.19.2 RADIAL PEAKING FACTORS SURVEILLANCE REQUIREMENTS 4.19.2.l The measured radial peaking factors (rA, and F()f(;/' -/ ~

obtained by using the incore detectionrsystem,rshall be determined to be less than or equal to the values stated in the LCO at the following intervals:

a After each fuel loading prior to operation above 50% of rated power, and

b. At least once per week of power operation.

f 4-84 Amendment No U, -rta--

Ngvuaber 15, 1988 "TSP1088-0181-NL04

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