ML20091E578

Amend 143 to License DPR-20,revising Plant TS in Support of Cycle 10 Operations

ML20091E578
Person / Time
Site:	Palisades
Issue date:	03/27/1992
From:	Marsh L Office of Nuclear Reactor Regulation
To:
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ML20091E580	List: ML20091E578 ML20091E592
References
NUDOCS 9204140274
Download: ML20091E578 (15)

Text

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M UNITED STATES n--

I NUCLEAR REGULATORY COMMISSION

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CONSUMERS POWER COMPANY QQGET NO. 50-255 PALISADES PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 143 License No. OPR.l.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Consumers Power Company (the licensec) dated November 1,1991, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),-

and the Comission's rules and regulations set forth in 10 CFR Chapter I; 1

B. --

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the commission;.

-C.

There is reasonable assurance -(i) that the activities authorized by -

this amendment can be conducted without endangering the health and safety of the public;-and (ii) that such. activities will be conducted in conpliance with the Commission's-regulations; D.-

The issuance of this amendant will not be inimical to the common defense and security or to the health and-safety of the.public; E.

The issuance of this amendment is in accordance with=10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

.2.

Accordingly, the license is amended by changes.to the Technical Specifications as indicated-in.the attachment to the license amendment and Paragraph 2.C.2 of Facility Operating License No. DPR-20 is hereby amended to read as follows:

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- Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.143, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION E D.

/' L. B. Marsh, Director Project Directorate 111-1 Division of Reactor Projects !!!/IV/V Office of Nuclear Reactor Regulation

Attachment:

Changes to the Techr.ical Specifications Date of Issuance: March 27,1992

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____._____.________.______-______:.___.__.__.__.__m

ATTACHMENT TO LICENSE AMEN 0 MENT NO. 1a1 FACILITY OPERATING-LICENSL" NO. OPR-29 DOCKET NO. 50-255 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages.

T5e revised pages are identified by the amendment number and contain marginal lines indicating the area of change.

RE}Myl INSERT l-2 1-2 2-4 2-4 2-9 2-9 3-3 3-3 3-2a 3-3a 3-29 3-29 3-31 3-31 3-66b 3-66b 3-67 3-67 3-105 3-105 3-107 3-107 3-111 3-111

1.1 REACTOR OPERATING CONDITIONS (Contd)

Low Power Physics Testing Testingperformedunderapprovedwrittenprocedurestodetermine contro5 red 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 5.**F and 415 psia to 2150 psia, respectively.

Certain deviations frr...

mal 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.

Shutdown Boron Concentrationi Doron concentration sufficient to provide X,n s 0.98 with all control rods in the core and the highest worth control rod fully withdrawn.

Refuelina Boron Concentration Boron concentration of coolant at least 1720 ppm (corresponding to a shutdown margin of at least 5% Ap with all control rods withdrawr.).

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

Assembly Radial Peakina Factor - F" 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.

TotalRadialPeakinaFactor-Fl l

The maximum product of the ratio of individual assembly power to core average assembly power times the highest local peaking factor integrated over the total core height, including tilt.

Local peaking is defined as the maximum ratio of an individual fuel rod power to the assembly average rod power.

1-2 Amendment No. JJ, f), Ef, E7, JJE, JJ7,143

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2.3 LIMITING SAFETY SYSTEM SETTINGS - REACTOR PROTECTIVE SYSTEM

'Anolicability

^ '

This specification applies to reactor trip settings and bypasses for instrument channels.

Ob.iectiya To~ provide-for automatic protective action in the event that the principal process variables approach a safety limit.

Snecification The reactor _ protective system trip setting limits and the permissible bypasses for-the instrument channels shall be as stated in Table

~ 2. 3.1. -

The TM/LP trip system monitors core power, reactor coolant maximum inlet temperature, (T ), core coolant system

. in pressure and axial shape'index. The low pressure trip limit (P,,,) is-calculated using the:following equation.

P,,, =_2012(QA)(QR ) '+ 17.0(T ) - 9493 g

in 4

where:-

QR, =

0.412(Q) + 0.588 Q s 1.0

_Q = core oower Q

Q > 1.0 rated power

=

ASI.=-

0 when Q < 0.0625.

QA: = -0.720(ASI) + 1.028_when. -0.628 s ASI < -0.100

-0.333(ASI) + 1.067 when -0.100 s ASI < +0.200-

+0.375(ASI)_.+>0.925 when +0.200 s ASI s +0.565 The calculated limit (P ) is then compared to a fixed low

. pressure trip limit (p i) limit (P,'h)).

The auctioneered highest of these

.x signals-becomes the'trlp.-

P is compared to the measured reactor coolant' pressure and,'Itripsignalis-generated when P is lessLthan or equal to P,i. A pre-trip alarm s

isalsogeneratedwhenPis-lessthanorequaItothepre-trip setting P,,,,~+ AP.'

2-4 Amendment No. JJE, 47, 143 i

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'2.3 tfMITING SMETV SYSVEM SETTINGS REACTOR PROTECTIVE SYSTFM (Contd)

Basis (Contd) 6.

Low St,eam Generator Pressure A reactor trip on low steam generator secondary pressure is providad 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 excessiv was used in the accident analysis,ely high steam flow.This setting

)

7.

Containment Hioh Pressure A reactor trip on containment high pressure is provided to assure tnat the reactor is' shutdown before the initiation of the safety injoction system and containment spray. N 8.

Low Power Physics Testing - For low power physics tests, certain tests will require the reactor to be critical at low temperature (2260*F) and low pressure (2415 psia)

For these certrin 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. Special operating precautions will be in effect during these tests in accordance with approvp% of d

written testing procedures. At reactor power levels below 10' rated power, the thermal margin / low pressure tv.p and low flow trip are not required to prevent fuel rod thermal limits from being exceeded.

The low steam generator pressure trip is not required because the low steam generator pressure will not-allow a severe reactor cooldown, shculd a steam line break occur during these tests.

References (1)

EMF 91-176, Table 15.0.7-1 l

(2) deleted (3)

Updated FSAR Section 7.I.3.3.

(4)

EMF-91-176, Section 15.0.7.1 l

(5)

XN NF-86 91(P)

(6) deleted (7) deleted (8)

ANF-90 078, Section 15.1.5 (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 1

2-9 Amendment No. 31, 118 0 ))

143

3.1 PRIMARY COOLANT SYSTEM (Cont'd) hsis (Cont'd) measurement; 20.06 for ASI measurement; !50 psi for pressurizer pressure; 17'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:

Tw,, s 542.99 +.0580(P 2060) + 0.00001(P-2060)**2 + 1.125(W-138) -

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

1800spregsures2200 psia 100.0 x 10 s Vessel Flow s 150 x 10' 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 T LCO at s

150 M lbm/hr increases the margin to ONB for, higher PCS flow rates.

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 po er. 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 boundarie,). An alarm signal is activated before the ordered pair exceed the s specified in Figure 3.0.

The requirement that the steam generator temperature be 5 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.

Ints 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-91 176 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. H, H, Jn 118, In, Jn, n1 143

3.1 PRIHARY COOLANT SYSTg3 (continued)

-P r

ASI Limit for T,1 function ini t,

1.2

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Unacceptable 3

Operations 1

b 0.s

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E Acceptable Operations o

c 0.6 S

h 0.5 E

0.4 0.3 o

gg e

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t 0.8

-Q S

-Q 4

.Q.3

-0.3 0.1 0

0.1 0.2 0,3 0.4 05 Axial Shape Index Break Points:

-0.550, 0.250

-0.300, 0.700

-0.080, 1.000

-0.080, 1.065

+0.400, 1.065

+0.400, 0.250 FIGURE 3-0 3-3a Amendment No. W, IJA/, JJ7,143 i

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r 3.3 EMERGENCY CORE COOLING SYSTEM l

Aeolicability

^

Applies to the operating status of the emergency core cooling system;-

Obiective To assure operability of_ equipment required to remove decay heat from the core in either emergency or_ normal shutdown situations.

.Soecifications Safety Iniection and Shutdown Cooline Systems 3.3.1'The reactor shall not be made critical, except for low-temperature physics tests, unless all of the following conditions are met:

a.

The SIRW tank contains not less than 250,000 gallons of water with a boron concentration of at least 1720 ppm but not more than'2500 ppm _at a temperature not-less than 40*F.-

b.

All four Safety injection tanks'are operable and pressurized to at-least.200 psig with a tank l_iquid level _of at least 174 inches and a maximum level of 200 inches with a boron-concentration of at least 1720 ppm but not more than 2500 ppm.

One low-pressure Safety Injtetion pump is1 operable on each bus.

c.

'd.

One high pressure Safety injection-pump it operable on eash bus.

Both shutdown heat exchangers and both component cooling heat e.

exchangers are-operable.

f.

Piping and valves shall be operable to provide two flow paths from the SIRW tank to the primary cooling system.

'g..

- All valves, pip.ing' and interlocks associated with the above components:

and required to function-during accident conditions are operable.

Lh.

- The 1.ow Pressure Safety Injection Flow Control Valve CV 3006 shall= be opened and disabled-(by isolating the air supply) to prevent spurious closure.

s 1.-

. The Safety Injection bottle motor-operated isolation val'ves-shall be opened with the electric power supply to the valve motor disconnected.

j.

The Safety Injection miniflow valves CV-3027 and.3056 shall be opened with'HS 3027~and S056 positions to maintain them open.

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3-29 Amendment No. JJ, 7f, JpJ, J35 143

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m 3.3 EMEDGENCY CORE COOLING SYSTEM (Continued)

If Specification a. and b. cannot be met, an orderly shutdown shall be c,

initiated and the reactor shall be in hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and cold shutdown within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

E1111 The normal proceoure for starting the reactor is, first to heat the primar coolant to near (perating temperature by running the prlmary coolant pumps.y The reactor is then made crit boronintheprimarycoolant.galbywithdrawingcontrolrodsanddiluting With this mode of start up, the energy stored in the primary coolant during the approach to criticality is substantially equal to that during power operation and, therefore, all engineered safety features and auxiliary cooling systems are required to be fully operable.

During low temperature physics tests, there is a negligible amount of stored energy in the primary coolant; therefore, an accident comparable in severity' to the design basis accident is not possible and the engineered safeguards systems are not required.

The SIRW tank contains a minimum of 250,000 gallons of water containing a minimum of 1720 ppm boren and a maximum of 2500 ppm.

This is sufficient boron concentration to provide a 5% shutdown margin with all control rods withdrawn and a new core at a temperature of 60'F.

Heatin The 1.g steam is provided to maintain the tank above 40'F to prevent freezing.

43% boron (2500 ppm) solutian will not precipitate out above 32'F.

The

-source of steam during normal plant operaticn is extraction steam line in the [

turbine cycle.

The limits for the safety injection tank pressure and volume assure the required amount of water injection during an accident and are based on valuas usedfortheaccigentanalyses.

The minimum 174-inch level corresponds to a volume pf 1040 ft and the maximum 200-inch level corresponds to a volume of 1176 ft Prior to the time the reactor is brought critical, the valving of the safety injection system must be checked for correct alignment and appropriate valves locked.

Since the system is used for shutdown cooling, the valving will be changed and must be properly aligned prior to start-up of the reactor.

The operable status of the various systems and components is to be demonstrated by periodic tests. A large fraction of these tests will be performed while the reactor is operating in the power range.

If a component is found to be inoperable, it will be possible in most cases to effect repairs and restore the system to full operability within a relatively short time.

For a single component to be inoperable does not negate the ability of the systes to perform its function, but it reduces the redundancy prpided in the reactor design and thereby limits the 3-31 Channe No. 1 Amendment No. JJ7, J B 143

POWER DISTRIBUTION INSTRUMENTATION 3.11,2 EXCORE POWER DISTRIBUTION MONITORING SYSTEM LIMITING CONDITION FOR OPEoATION Easis (Contd)

Surveillance reouirements ensure that the instruments are calibrated to acree with ne i m 4 e m re:en u r : +nat tna u rget A0 is casec on :ne :.rr46t operating conditions.

Updatin the Excore Monitoring APL ensures that the core LHR limits are protected ithin the 20.05 band on AO.

The APL considers LOCA based LHR limits, and factors are included to account for changes in radial power shape and LHR limits over the calibration interval.

The APL is determined from the following:

LHR(Z),,

APL =

[

ta)

LHR(Z),,, x V(Z) x 1,02,s] x Rated Power Where:

(1)

LHR(Z)n is the limiting LHR vs Core Height (from Section 3.23.1),

(2)

LHR(Z vs Cor),e,' Height,is the measured peak LHR including uncertainties (3)

V(Z) is the function (shown in Figure 3,11 1),

(4)

The factor of 1,02 is an allowance for the effects of upburn, (5)

The quantity in brackets is the minimum value for the entire core at any elevation (excluding the top and bottom 10% of core) considering limits for peak rods.

If the quantit in brackets is greater than one, the APL shall be the rated power leve,

References

(

(1)

XN NF-80 47 (2)

EMF 91 177

{

l 3 66b Amendment No, JJ, JJ, JIS

• Corrected (next page is 3-66d) u

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3,12 MODERATOR TEMPERATURE COEFFICIENT-0F' REACTIVITY Aeolicability Applies to the moderator temperature coefficient of reactivity for the core.-

-0biective

To specify a limit for the positive moderator coefficient.

-Snecifications The modera 9n

+0.5 x.10', tor temperature coefficient (MTC) shall be less positise Ap/'F at 5 2% of rated power.

11181 The limitations on moderator temperature coefficient.MTC

'toensurethattheassumption;usedinthesafetyana'lystgfreprovided remain

. valid.

Reference-

.(1)

EMF-91-176, Section 15.0.5 i-.

Y 3 67 L

1 Amendment No.. 118,117'.

(nextpageis3-69)

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POWER DISTRIBUTION tiMITS 3.23.1.

LINEAR HEAT RATE (LHRi tiMITING CONOTTION FOP ODERATION E1111 (Contd)

The time interval of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and the minimum of 10 detectors per quadrant are sufficient to maintain adequate surveillance of the core power distribution to detect significant changes until the monitoring systems are returned to service.

To ensure that the design margin of safety is maintained, the determination of both the incore alarm setpoints and the APL takes into.cc; gat : measurement uncertainty factor of 1.10 an engineering.

uncertainty factor of 1.03, a thermal power measuremen,t uncertainty factor of 1.02 and ail:wance for quadrant tilt.

References (1)

EMF-91-177

[

(2)

(Deleted)

(3)

(Deleted)

(4)

XN NF 80-47 3-105 Amendment No. 65, JJB,143 (next page is 3-107)

TABLE 3.23 1

-LINEAR HEAT RATE LIMITS I

No. of Fuel Rods Assembly 208 216 Peak Red 15.28 KW/ f t 15.28 kW/ft TABLE 3.23 2 RADIALPEAKINGFACTORLIMITS,4 E.

Peaking Factor No. of Fuel Reds in Assembly J

208 216 216 (Reload M and earlier)

Assembyl F,"

1.48 1.57 1.'66 l

Peak Rod F.'

l.92 1.92 1.92

,l 3-107 Amendment No. JJ, Ilj, 137, 143

4 POWER DISTRIBUTION LIMITS 3.23.2 RADIAL PEAKING FACTORS LIMITINGCONQTIONJOROPERATION A

T TheradialpeakingfactorsF[hefollowingquantity

, and F shall be less than or equal to the value in Table 3.23 2 times The quantity i. 't.0 +

0.3 (1 - P)) for P 1.5 and the quantity is 1.15 for P <.5.

P is he vre themal 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 ct 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 2 50% of rated with any radi, 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:

[1

- 3.33( r - 1) ) x Rated Power L

A T

Where F is the measured value of either F, or F and F 5

r L

is the corresponding limit from Table 3.23 2.

Basis A

ThelimitationsonF,andFfareprovidedtoensurethatassumptionsusedin 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 LOCA analysis supports the radial peaking facter limits in Table 3.23-2.

3-111 Amendment No. $$, JJS, J37,1/,3

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