ML18057B350

Proposed Tech Spec for Cycle 10 Operation Including Total Radial Peaking Factor,Primary Coolant Sys,Safety Injection & Shutdown Cooling Sys & ex-core Power Distribution Monitoring Sys

ML18057B350
Person / Time
Site:	Palisades
Issue date:	11/01/1991
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML18057B349	List: ML18057B348 ML18057B350 ML18057B351 ML18057B352
References
NUDOCS 9111080147
Download: ML18057B350 (29)

Text

ATTACHMENT 1 Consumers Power Company Palisades Plant Docket 50-255 CYCLE 10 TECHNICAL SPECIFICATIONS CHANGE REQUEST PROPOSED TECHNICAL SPECIFICATIONS PAGES November 1, 1991 12 Pages

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 23 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.

Certain 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.

Shutdown Boron Concentrations Boron concentration sufficient to provide Keff ~ 0.98 with all control rods in the core and the highest worth control rod fully withdrawn.

Refueling Boron Concentration Boron concentration of coolant at least 1720 ppm (corresponding to a shutdown margin of at least 53 Ap with all control rods withdrawn}.

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

Assembly Radial Peaking 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.

Total Radial Peaking Factor - F; The maximum product of the ratio of individual assembly power to core average assembly power times the highe~t 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. $1, ~$, ~~' ~7, iJ~, 1$7

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 2.3.1.

The TM/LP trip system monitors core power, reactor coola.nt maximum inlet temperature, (Tin), core coolant system pressure and axial shape index.

The low pressure trip limit (Pvar> is calculated using the following equation.

Pvar = 2012(QA}(QR1) + 17.0(Tin) - 9493 where:

QR1 =

0.412(Q) + 0.588 Q ~ 1.0 Q = core power Q,

Q > 1.0 rated power ASI =

O when Q < 0.625 QA = -0.720(ASI) + 1.028 when

-0.628 ~ ASI < -0.100

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

= +0.375(ASI) + 0.925 when +0.200 ~ ASI ~ +0.565 The calculated limit (Pva) is then compared to a fixed low pressure trip limit (p~inf. The auctioneered highest of these signals becomes the trip limit (Ptr}R).

Ptrip is compared to the measured reactor coolant pressure l~) and a trip signal is generated when P is less than or equal to Ptrip" A pre-trip alarm is also generated when P is less than or equa1 to the pre-trip setting Ptrip + AP.

2-4 Amendment No. 11~, J~l

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 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.ca>

7.

Containment High Pressure - A reactor trip on containment high pressure is provided to assure that the reactor is shutdown before the initiation of the safety injection system and containment spray.'10>

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 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. Special operating precautions will be in effect during these tests in accordance with approved written testing procedures.

At reactor power levels below 10*11 of rated power, the thermal margin/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 is not required because the low steam generator pressure will not.*allow a severe reactor cooldown, should a steam line break occur during these tests.

References (1)

EMF-91-176, Table 15.0.7-1 (2) deleted (3)

Updated FSAR, Section 7.2.3.3.

(4)

EMF-91-176, Section 15.0.7.1 (5)

XN-NF-86-9l(P)

(6) deleted (7) deleted (8)

ANF-90-078, Section 15.1.5 (9)

ANF-87-lSO(NP), Volume 2, Section 15.2.7 (10)

Updated FSAR, Section 7.2.3.9.

(11)

ANF-90-078, Section 15.2.1 2-9 Amendment No. ~J, JJ~l/13:

J

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 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 106 s Vessel Flow s 150 x 106 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 T1nle LCO at 150 M lbm/hr increases the margin to DNB 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 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,Y1).

An alarm signal is activate~ 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 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-91-176 Section 15.0.7.1 (5)

ANF-90-078 (6)

Consumers Power Company Engineering Analysh EA-A-NL-89-14-1 3-3 Amendment No. ~J, JJ, JJ1 JJJ, J~J, J~~' J~l

3.1 PRIMARY COOLANT SYSTEM (continued)

ASI Limit for Tinlet function L

Q)

~

1.1 1

0 0.9

a.

lJ (l) a.a

+i

((]

0::

4-0 0.7 c

0.6 0

+i u

o.s

((]

L LL 0.4 0.3 Unacceptable Operations Acceptable Operations 0.2.___.___.._~__..~~-'-~

............ ~~....._~~~--"~~_._~_._~~_,__~_.

-0.6

-0.S

-0.""

-0.3

-0.:2

-0.1 0

0.1 0.:2 0.3 D.4 0.5 Break Points:

-0.550, 0.250

-0.300, 0.700

-0. 080' 1. 000

-0. 080, 1. 065

+O. 400, 1. 065

+0.400, 0.250 Axial Shape Index FIGURE 3-0 3-3a Amendment No. ~' 1)~/, J~l,

3.3 EMERGENCY CORE COOLING SYSTEM Applicability Applies to the operating status of the emergency core cooling system.

Ob.iective To assure operability of equipment required to remove decay heat from the core in either emergency or normal shutdown situations.

Specifications Safety Iniection and Shutdown Cooling 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 4o*F.

b.

All four Safety Injection tanks are operable and pressurized to at least 200 psig with a tank liquid level of at least 174 inches and a maximum level of 200 inches with a boron concentration of at least 1720

c.

d.

e.

f.

g.

h.

i.

j.

ppm but not more than 2500 ppm.

One low-pressure Safety Injection pump is operable on each bus.

One high-pressure Safety Injection pump is operable on each bus.

Both shutdown heat exchangers and both component cooling heat exchangers are operable.

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

All valves, pipjng and interlocks associated with the above components and required to function during accident conditions are operable.

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

The Safety Injection bott-le motor-operated isolat-ion valves shall be opened with the electric power supply to the valve motor disconnected.

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

. 3-29 Amendment No. ~J, 7~, J~J, J~~

3.3 EMERGENCY CORE COOLING SYSTEM (Continued)

c.

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

Basis The normal procedure for starting the reactor is, first, to heat the primary coolant to near operating temperature by running the primary coolant pumps.

The reactor is then made crit~nal by withdrawing control rods and diluting boron in the primary coolant.

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 boron 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 so*F.

Heating steam is provided to maintain the tank above 4o*F to prevent freezing.

The 1.43% boron (2500 ppm) solution will not precipitate out above 32*f. The source of steam during normal plant operation 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 durin~ an accident and are based on values used for the acci~ent analyses.

The min1mum 174-inch level corresponds to a volume ~f 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 operaoility within a relatively short time.

For a-'single component to be inoperable does not negate the ability of the syste11'to perform its function, but it reduces the redundancy provided in the reactor design and thereby limits the 3-31 Amendment No. 117, 1$~

3.11.2 POWER DISTRIBUTION INSTRUMENTATION EXCORE POWER DISTRIBUTION MONITORING SYSTEM LIMITING CONDITION FOR OPERATION Basis (Contd)

Surveillance requirements ensure that the instruments are calibrated to agree with the incore measurements and that the target AO is based on the current operating conditions. Updating the Excore Monitoring APL ensures that the core LHR limits are protected within the +/-0.05 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:

APL ""

[

LHR(Z)rs 2

---

] x Rated Power< >

Where:

(1)

(2)

(3)

(4)

(5)

LHR(Z)Max X V { Z) X 1. 02 Min LHR(Z)rs is the limiting LHR vs Core Height (from Section 3.23.1),

LHR(Z)Ma~.i~ the measured peak LHR including uncertainties vs Core Me1ght, V(Z) is the function (shown in Figure 3.11-1),

The factor of 1.02 is an allowance for the effects of upburn, 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 quantity in brackets is greater than one, the APL shall be the rated power level.

References (1)

XN-NF-80-47 (2)

EMF-91-177 3-66b Amendment No. JJ, ~J, JJJ

• Corrected (next page is 3-66d)

3.12 MODERATOR TEMPERATURE COEFFICIENT OF REACTIVITY App 1 icabi l i ty Applies to the moderator temperature coefficient of reactivity for the core.

Ob.ject i ve To specify a limit for the positive moderator coefficient.

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

+0.5 x lo* Ap/°F at ~ 2% of rated power.

Bases The limitations on moderator temperature coefficient CMTC),ftre provided to ensure that the assumptions used in the safety analysis remain valid.

Reference (1)

EMF-91-176, Section 15.0.5 3-67 Amendment No. ii~, i~l (next page is 3-69)

J

Peak Rod Peaking Factor Assembvl F _A Peak Rod F.. T TABLE 3.23-1 LINEAR HEAT RATE LIMITS No. of Fuel Rods Assembly 208 216 15.28 KW/ft 15.28 kW/ft TABLE 3.23-2 RADIAL PEAKING FACTOR LIMITS, FL No.

208 1.48

1. 92 3-107 of Fuel Rods in Assembly 216 (Reload Mand earlier) 1.57

1. 92 Amendment No. ~~, JJ~, 137 February 20, 1991 216 1.66 i

1. 92 i

POWER DISTRIBUTION LIMITS 3.23.1 LINEAR HEAT RATE (LHRl LIMITING CONDITION FOR OPERATION Basis (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 account a measurement uncertainty factor of 1.10, an engineering uncertainty factor of 1.03, a thermal power measurement uncertainty factor of 1.02 and allowance for quadrant tilt.

References (1)

EMF-91-177 (2)

(Deleted)

(3)

(Deleted)

(4)

XN-NF-80-47 3-105 Amendment No. ~a, iia (next page is 3-107)

POWER DISTRIBUTION LIMITS 3.23.2 RADIAL PEAKING FACTORS LIMITING CONDITION FOR OPERATION The radial p~aking factors FA, and F~ shall be less than or equal to the value in Table 3.23-2 times fhe following quantity.

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 ih 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:

Basis F

[l - 3.33{ r - 1) ] x Rated Power T

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

The limitations on F~, and F~ are provided to ensure that assumptions 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 with*in 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 factor limits in Table 3.23-2.

3-111 Amendment No. ~J, llJ, l~7

ATTACHMENT 2 Consumers Power Company Palisades Plant Docket 50-255 CYCLE 10 TECHNICAL SPECIFICATIONS CHANGE REQUEST MARKED UP EXISTING PAGES November 1, 1991 13 Pages

I l. 1

'-'i B£ACTOB OPERATING CONDITIONS (Contd)

Lgw Pa**r 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.

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

Sbutdgwn Bocgn Concentratigns Boron concentration sufficient to provide k,,, s 0.98 with all control rods in the core and tbt highest worth control rod fully withdrawn.

Befy1linq Bgcgn Conc1ntratign Boron concentration of coolant at least 1720 ppm (corresponding to a shutdown margin of at least SI ~P with all control cods withdrawn).

Oy1dc1nt pgw1r Tilt The difference between nuclear power in any core quadrant and the average 1n all Qij&drants.

Asstmbly Radial Peaking Factor - F!..

The assembly radial peaking factor 1~ tht maximum ratio of individua 1 fuel assembly power to core average assembly power integrated over the total core height, including tilt.

Tgtal ' '

I ;.,.,.Bad111 peak1 ng Factqr - F!

The maximum product of the ratio of indiv~dual assembly powe~ to c:r~

av1r191 asslllbly power times the highest local peaK1ng factor 1ntJiated over the total core h1igbt.,,includ1n9 tilt: Local

. J DMk1ng 1s fined u the max1mum ratio of 1iA1 Jl 29 an ind1v1d*.:

fuel ~~to,, tseebly av1rag1 rod power.

f"CJw tr' 1-2 Amendment No. JJ,,J, 91, 97, ::a.~

f1lrta111 ZO, l 99+-

I I

2. 3

-*~

L.r~, r.~'.UG S'FE~.Y s sr-w SET~*

~-*c*~

-**. -~-

_,.. =

Y

*.. '.NGS -

yR Apel j c:tbi 1 j ty This spec1f1catton applies to reactor trip settings and bypasses for 1nstru111nt channels.

Qb1est1ye To provide for automatic protective action 1n the event that the principal process variables approach a safety limit.

Spec: 1f i c:at ion Tht reactor prottcttve system trip setting limits and tht permissible bypasses for the instrument channels shall bt as stated in Table 2.3.1.

Tht TM/LP trip system monitors cart power, reactor coolant maximum inltt t1mp1raturt, (T.~), cart coolant system pressure and axial shape index.

The low pressure trip limit (Pw.,.) is calculated using the following equation.

P ** r

• 201Z(QA) (QR,) + 17.O(T, 11 )

- 9493 where:

QR,

• 0.412(Q).+ 0.588 Q

• cor1 pgwtr rated power

-0.720(ASI) + 1.02!J -0.628. 'ASI < -0.100

• -0.333(ASI) + 1.067

-0.100 'ASI < +O.ZOO

• +0.375(ASI) + 0.925

+0.200 ' ASI ' +0.565

--Jnill when

• l 0 HCS A

---

==--"..... A.Sr= o

'"""1-1,,, a c

'-'. o z.s The calculated 11*1t (P ** r) is then compared to a fixed low pressure trip 11*1t (p., 11). The auctioneered highest of these signals becomes the trip limtt (P,,.,,).

P,,.., ts compared to the measured reactor coolant pressure (P) and a trip stgnal ts generated when P 1 s less than or equal to P,...,. A pre-trip alarm is also generated when P ts less than or 1qua1 to th* pre-trtp setting '*r** + *

z.. 4 Amendment No. 119. ~

,. 1 ee, **;i

"'

• 1 ISll J

~

z J

~!~!T!~G SAFETY SYSTE~ SETTrNGS -

qE~CTOR pqgr~~r:vE svsT~~ (CJn::i

.a.ull (Contd) 6.

Low Steam Generator Pressyre

• A reactor trip on low steim g1nerator secondary pressure is provided to protect igiinst an excessive rite 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 is 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.'"

7.

Containmtnt High Pressyre *A reactor trip on containment high pressure is provided to assure that the reactor is shutdown befor1 the initiation of the safety injection system ind containment spray.< 101

8.

Low p9w1r physics Testing

• For low power physics tests, certain tests will require the reactor to b* critical at low temperature (2260.F) ind low pressure (2415 psia).

For these certain tests only, the thermal margin/low pressure, primary coolant flow and low stea111 generator pressure trips may bt 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 approved written testing procedures.

At reactor power levels below 10*

11 of rated power, the ~htrmal margin/low-pressure trip and low flow tr1p are not required to prevent fue 1 rod therma 1 1 ; mi ts-fro11 be 1 ng 1xc1edtd. The 1 ow steam generator pressure trip is not required because the low steam generator pressure will not allow a severe reactor cooldown, should a steam lint break occur during th1s1 tests.

References

~ M,:. C//* 17~

(1)

-'NF II 871, Table 15.0.7-1 (2) d1l1ted (3)

UO:d~!U.SU, Section 7.2.3.3.*

(4) 115iiiiW&~"Sect1on 15.0.7el (5)

XN-NF-86-ll(P)

(6) deleted (7) deleted (8).

MF-90-071, Section 15.1.5 (9)*~

MF-17-llO(NP), Volume 2, Section 15.2.7 (lQJ> Updated FSAR, Section 7.2.3.9.

( 11)'

MF-I0-071, Stet 1 on 15. 2.1 J

I 2-9 Amendment No. JJ, JIS. ~

-Fe9PWiP¥ ao, 1991

3.

pq;~ARY COOLANT SYST~M 'Cort'j)

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 1~

the derivation of the following equation for limiting reactor inlet temperature:

T,~ 1 u ~ 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 101 s Vessel Flow s 150 x 101 lb/h t,...,

ASl as shown in Figure 3.0 With measured primary coolant system flow rates > 150 M lbm/hr, i

1 imiting the maximum allowed inlet temperature to the Tr..,.. LCO at 150 M lbm/hr increases the margin to ONB for higher PCS flow rates. t-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 Oelta-T power. The measured ASI calculated from the excore detector signals and adjusted for shape ann11ling (Yr) and the cart power constitute an ordered pair (Q,Yr>*

An 1lan11 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 secondarl' 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, analysts (R1f1r1nc1 6) shows that under limited conditions when the Shutdown Cooling System is isolated from th* PCS, forced circulatton.. 1 be initiated when the steam generator temperature is higher than the PCS cold leg temperature.

R1fsnnc1a (l) Updated FSAR, Section 14.3.2.

(I) Updated FSAA, Section 4.3.7.

( 3) DI 1 tted s. m F - 9'1 - /7 ftJ;

( 4) :'8;7 g 7 1 Stet i on 1 S. 0.}. 1

(

(5) ANF-I0-071,

(6) Consumers Power Colft1)any Engin11rtng Analysts EA-A-NL-89-14-1 3-3 Amendment No. JJ, JJ, 111 11*, 1J1, 11*. ~

eiPUll' 26, i SS

• a:

IJ

E 0 a. *.&

0 IJ a:

I a:

Ill

~

0

• .JD z

0 -

~

f ***

lllPtilPll&t:

. mDllllM f IGURC: 3-0 ftSI LCO rOR TLnlel fUNCTION ORCftlC rolNJS I. -0.550, 0.25

2. -Q.JOO, 0.7

]. -0.080, 1.0

i. *O.iOO, I.0 0.0 0.2 0.t fiX I tll s11111*c I NOCX

~-c

3.1 PRIMARY COOLANT SYSTEM (continued)

ASI Limit for Tintet function 1.1 1

L

~

0 0.9 a.*

'O G.>

0.8

+' &

'+-0 O.?

c:

o. 6 0 *-+' u o.s IO L

u.

0.4 0.3 Unacceptable Operations Acceptable Operations 0.2 L-....J.-...J-~-J..~~.i.......~.....i.....~-J..~~~~.....&......~-J..~~..&..-~-L.~--J

-0.6

-0.5

-0.~

-0.3

-0.2

-0.1 0

0.1 0.2 0.3 C.'4 0.5 Break Points:

-0.550, 0.250

-0.300, 0.700

-0. 080' 1. 000

-0.080, 1.065

+O. 400, 1. 065

+0.400, 0.250 Axial Shape Index FIGURE 3-0 3-3a Amendment No. W, 1/.1'/, J.'/17,

.J

3.3 EMERGENCY CQRE COOLING SYSTEM Aoplicabiliti Applies to the operating status of the emergency core cooling system.

Objective To assure operability of eQuipment required to remove decay heat from the core in either emergency or normal shutdown situations.

Specjfjcations Safety Iniection and Shytdown Cooling Systems 3.3.l 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~ J

b.

c.

ppm at a temperature not less than 40* F.

'%Sa:'

All four Safety Injection tanks are operable and pressurized to at {

least 200 psig with a tank liquid level of at least 174 inches ana a maximum level of 200 inches with a boron concentration of at 1 east 1720 ppm but not more than...2.0eO ppm.

I Z,S'?)O Ont low-pressure Safety Injection pump is operable on each bus.

d.

One high-pressure Safety Injection pump is operable on each bus.

e.

Both shutdown heat exchangers and both component cooling heat exchangers art operable.

f.

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

g.

All valves, piping and interlocks associated with the above components and required to function during accident conditions ar~

operable.

h.

t.

j.

The Low-P1"essur1 Safety Injection Fiow Control Valve CV-3006 sha'.

• be opened and disabled (by isolating the air supply) to prevent spurious closure.

The Safety Injection bottle motor-optrattd isolation valves sha1' be opened with the electric power supply to the valve motor disconnected.

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

3-29 Amendment No.

~J, 7*,

1~1. ~

~ilaP'Wli j 15, l 39 :

3.3 EMERGENCY CORE COOLING SYSTEM (Continued)

c.

.6.lill If Specification a. and b. cannot be met, an orderly shutdown shall be 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 />.

The normal procedure for starting the reactor is, first, to heat the primary coolant to near operating temperature by running the primary coolant pumps.

The reactor is then made criticr:l by withdrawing control rods and diluting boron in the primary coolant. ii 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 tt... t'h'1th11'"1v.n'1 ~ 1720 ppm bor This is sufficient boron concentration to provide a 5%

~~IJWn margin with all control rods withdrawn and a new core at a temperature of 60° F.

/.'I~ 0/ 0

J.500 d

Heating steam is ~ded-t~ --~ain the tank above 40°F to prevent a-n a.11Y1~1,,,.""'freezing. The Jl"t,~~~n (~p~) solution will not precipitate out

_ / J.{l)Oo.'flT1J above 32°F.

The source of steam during. normal pl ant operation is

• "t'*

r*

extraction steam line in the turbine cycle.

The limits for the safety injection tank pressure and volume assure the reguired amount of water injection during an accident and are based on values used for the accident analyses.

The minimum 174-inch level t corresponds to a volume of 1040 ftl and the maximum 200-inch level corresponds to a volume of 1176 ftl.

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 perfo,llld while the reactor is operat~ng in the power ran~e. If a c** ts found to be inoperable, 1t will be possible 1n most cases to. e.

• repairs and restore the system to full operability within a relat y short time.

For a single component to be inoperable does no~

negate the ability of the system to perform its function, but it reduces the redundancy provided in the reactor design and thereby limits the 3-31

3.11.2 EXCOR! POWER D!STR!BUT!ON ~ON!TOR!NG SYS~

t!~!TING CONDITION FOR OPERATION Balis (Coned)

Surv1illanc1 r1quir1m1nc:1 1nsur1 chat th1 insc:rum1nc:s ar1 calibrac1d co agr11 with th1 incor1 m1aaur1m1nt1 and ~hac c:h1 carg1c AO i1 ba1ed on chi curr1nc: op1raciag conditiona.

Updatin~ c:h1 Excor1 Monitoring APL enaur1s ehac th1 core LHR limits ar1 protected within th1 !0.0S band on AO.

Thi APL considers LOCA based LHR limits, and factors ar1 included to account for changes in radial power shape and LHR limits over th1 calibration interval.

Thi APL is d1c1n11in1d froa the follavin1:

LHll(Z)TS (2)

APL * ( LH1l(Z)Max x V(Z) x l.02 ]Min x Rated Pov1r Wh1r1:

(l) LRll(Z)TS ii th* limitin1 LHR VI Cori R1i1ht (from Section 3.23.11, (2) LHll(Z)M i1 the m1a1ur1d peak ~Hll includin1 uac1rtainti11 VI Corea**i1ht, (3) V(Z) i1 the function (shown in Figure 3.11-1),

(4) 'nle factor of 1.02 is an allowance for the effects of upburn,

~

(5) TI\\1 quantity in brackets 11 the ainimua value for the 1ntir1 core at any el1vatioa (1xcludin1 the top and botto. 10% of core) con1id1r1n1 limits for peak roda. If the quantity in

~

bracket* ii ar**t*r than on*. th* A1t *hall b1 th* rated paver level.

R1f 1r11ac11

( 1) llMf-I0-4 7 (2) --~II 117 Eff\\F- '11 -177 3-66b

• Corr1ct1d (next page is 3-66d)

TSP1088-0181-Nt.04 Allendment No. JI, fl, +te-N1 u cab1t 13, 1989*

I

3 : 2 Apel i cabi 1 i ty Applies to the moderator temperature coefficient of reactivity foe tht core.

Object1ye To specify a limit for the positive moderator coefficient.

Specifications The moderator temperature coefficient (MTC) shill be less posit1vt thin +0.5 x io** 'p/°F &t s 21 of rited power.

Buis The limitations on moder&toc temptr&ture coefficient (MTC) are provided to ensure that the assumptions used in the s1fety 1n11ysis 0 l remain val id.

B1f1c1nc1 (1)

ANF=98=879-, Section 15.0.S EMF* ~11-11(,

3-67 (next page is 3-69)

Allendlllnt No. JZI.~

F1br~1r, 2Q 1 199 1 I

Peak Rod Peaking Factor Assembly FrA Peak Rod FrT TABLE 3.23-1 LINEAR HEAT RATE UMITS No. of Fuel Rods Assembly 208 216 15.28 KW/ft 15.28 kW/ft TABLE 3.23-2 RADIAL PEAKING FACTOR LIMITS, FL 208 1.48 1.92 No. of Fuel Rods in Assembly 216 216 (Reload M and earlier) 1.57 1.92 3-107 1.66 1.92 Amendment No.

-~' JJ~, +a-t-February 29, 199.l..

J

?OWER D[57R:3~::c~ ::~::s

3. 23. t t!NEAR HEAT RAT'E (tHR) tI~!T!NG CONDITION :OR OPERATION Basis (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 ch* minimum of lO d1c1ccors per quadrant are suf f icienc co maintain ad1quac1 surveillance of che core power distribution co detect signif icanc changes until chi monicorin1 systems are returned co service.

To 1nsur1 chat chi design margin of safety is maintained, chi determination of both chi incore alarm s1cpoinc1 and the APL cakes into account a measurement unc~rtaincy factor of l. LO, an 1ngin1erin1 uncertainty factor of l.03, a thermal power m1a1ureaent unc1rtaincy factor of l.02 and allowance for quadrant tilt.

R1hr1nc11 (1)

MIF II 107 fEMi=- 17 7 (2)

(D1l1t1d)

(3)

(D1l1t1d)

(4)

XN-NF-80-47 3-105 (n1xc P*i* is 3-107)

TSPL088-0L81-NL04

.Aaendaenc No *.,, ~

M1v1*~ct 13, 1988

=cwE~ Q(STRrsur;oN LIMITS 3.23.2 BADIAL PEAKING FACTORS LIMITING CONDITION FOR OPERATION The radial ptak1ng factors FA, and FT shall be less than or equal to the value in Tablt 3.23-2 times the follbwing quantity.

The quantity is [l.O ~

0.3 (1 - P)] for P 2.5 and the quantity is 1.15 for P <.s. Pis the core thermal power in fraction of rated power.

APPLICABILITY:

,ACTION:

Power operation above 2S~ of rated power.

l.

For P < SOI of rated with any radial peaking factor exceeding its limit, be 1n 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 2 SOI of rated with any radial peaking factor exceeding its limit, reduct thermal power within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to ltss than the lowest value of:

Wi1 F

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

L Where Fr 1s the meuured valut of either F~. or F~ and FL is the corresponding limit from Ta~le 3.23-2

• The limitations on F~. and F~ art provided to ensure that assumptions used in.z, the analysis for establishing DNB margin, LHR and the thermal margin/low-pressure and variable high-power trip set points rtmain valid during operation. Data frot1 the incore detectors art used for determining the measured radial p11ktng factors.

The periodic surveillance rtquirtmtnts 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 SOI of rated power provides additional assurance that the core is properly loaded.

The ra~111,.111t111 t1 lt*tt*~ to thou val u11.... ~ 1" tt.e L8EA 1n1ly11 s.

s; nee,~. I er* *0 *h<<s ts Ji *tt s t bt *19" 1 hd* 0 r rad I., ptak I 119' Tille 3 I 23 2 explicitly '11tai11 the11 limits.

The. LOCA a.n"lys1s S'-'PPorts ~~c.

j1;,.1f.J' i 11 To.. bi*.... 3. 2. 3-7-.,

3-111 Amendment No.**, JI8.~

EtD,Wl'Y 2Q, i;;;

ATTACHMENT 3 Consumers Power Company Palisades Plant Docket 50-255 CYCLE 10 TECHNICAL SPECIFICATIONS CHANGE REQUEST SIEMENS NUCLEAR POWER CORPORATION REPORT PALISADES CYCLE 10: DISPOSITION AND ANALYSIS OF SRP CHAPTER 15 EVENTS (EMF-91-176)

November 1, 1991

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SIEMENS Palisades Cycle 1 O: Disposition And Analysis Of Standard Review Plan Chapter 15 Events October 1991 1

Siemens Nuclear Power Corporation I

EMF-91-176