ML20116N134

Proposed Tech Specs,Consisting of Suppl to Change Request 170,modifying TS 15.3.10, Control Rod & Power Distribution Limits & TS 15.4.1, Operational Safety Review

ML20116N134
Person / Time
Site:	Point Beach
Issue date:	08/15/1996
From:	WISCONSIN ELECTRIC POWER CO.
To:
Shared Package
ML20116N132	List: ML20116N131 ML20116N134
References
NUDOCS 9608210257
Download: ML20116N134 (26)

Text

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i 15.3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS Applicability.

i' Applies to the operation of the control rods and to core power distribution limits.

Objective

)

To insure (1) core subcriticality after a reactor trip, (2) a limit on potential reactivity insertions from a hypothetical rod cluster control assembly (RCCA) ejection, and (3) an acceptable core power ,

distribution during power operation.

Snecification A. Ennh In=rtien Li-9: SHUTDOWN MARGIN

31. The shutdown margin shall exceed the applicable value as shown in Figure 15.3.10-2 under all steady-state operating conditions from 350 F to full power. An ene ption to i : tuck RCCA ecmpenent Ofi chutdce margin requirement is permitted for physics tests,If the shutdown margin is less than the applicable _yalue of FigureJ5.3.10-2. within 15 minutes initiate boration to restore the shutdown marein. l Q. Encept for physic ::: ';;g shutdown margin of at least 1% Ak/k shall be maintained when the reactor coolant temperature is less than 350 F. If the shutdown margin is less than this limit. within 15 minutes initiate boration to restore the shutdown marein.

5. During any app = ch te criticality, ene pt for physic te:;ts,i: critical red positien cha!!

net be !cw : San Se in=r.icn limit for z:= pcwer. That is, ifie centre! red: " ere wiid=vn in nc=c! =quene: /ii na cSer recetivity change, de := ter would not be l critical =ti! 1 centrc! banks were abcv: ie in=r.icn !!mit  !

1 B. Pcwer Distr %tien L! ::: ROD OPERABILITY AND BANK ALIGNMENT LIMITS

1. During power and low nower coeration. all shutdown and control rods shall be onerable.

with all individual indicated rad nositions within twelve stens of their bank demand position. excent when the bank demand position is <30 steps or >_215 stens. In this case.

all individual indicated rod positions shall be within 24 steps of their bank demand l nosition.

l If an RCCA does not step in upon demand, up to six hours is allowed to determine whether the problem with stepping is an electrical problem. If the problem cannot be resolved within six hours, the RCCA shall be amumed declared inoperable until it has been verified that it will step in or would drop upon demand.

9608210257 960815  :

PDR ADOCK 05000266 l P PDR Unit 1 - Amendment No. 15.3.10-1 Unit 2 - Amendment No.

a. Rod Onerability Requirements (1) If one rod is determined to be untrippable. perfom1 the following actions:

1 (a) Within one hour verify that the shutdown marein exceeds the i anolicable value as shown in Ficure 15.3.10-2:

dR {

(b) Within one hour restore the shutdown margin by boration.

OR '

(c) Within six hours be in hot shutdown. l l

(2) If sustained oower operation with an untrippable rod is desired perform the following actions:

(a) Within one hour verify that the shutdown margin exceeds the  ;

applicable value as shown in Figure 15.3.10-2: OR within one j hour restore the shutdown margin by boration: 1 AND I (b) Within six hours. adiust the insertion limits to reflect the wonh j of the untrippable rod. l (c) If the above actions and associated comnletion times are not met. be in hot shutdowm within six hours.

(3) If more than one rod is determined to be untrippable. perform the following actions:

(a) Within one hour verify that the shutdown margin exceeds the apnlicable value as shown in Figure 15.3.10-2: OR within one hour restore the shutdown margin by boration; AND (b) Within six hours be in hot shutdown.

b. Rod Bank Alignment Limits (1) Ifit has been determined that one rod is not within alignment limits. and the indicated misalignment is not being caused by malfunctioning rod position indication. within oJ1.e hour restore the rod to within alignment limits: OR pr1 form the following actions:

(a) Within one hour verify that the shutdown margin exceeds the applicable value as shown in Figure 15.3.10-2: OR within one hour restore the shutdown margin by boration:

AND Unit 1 - Amendment No. 15.3.10-2 Unit 2 - Amendment No.

(b) Within eight hours reduce thermal power to <;75 percent of rated thermal power:

AND <

(c) Verif.y that the shutdown marein exceeds the an.nlicable value as l

shown in Figure 15.3.10-2 once per twelve hours AND (d) Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> verify that measured values of Fe(Z) are within i limits; AND l (e) Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> verify that F(is within limits: l 1

I (f) If the above actions and associated comnletion times are not met. be in hot shutdown within the following six hours.

(g) In order to subsequently increase thermal power above 75 percent of rated thermal power with the existing rod misalignment. perform an analysis to detennine the hot channel factors and the resulting allowable power level in accordance with TS 15.3.10.E.

(2) Ifit has been determined that more than one rod is not within alignment limits and the misalignments are not being caused by malfunctioning rod position indication. nerform the following actions:

(a) Within one hour verify that the shutdown marcin exceeds the applicable value as shown in Fjgure 15.3.10-25 OR within one l hour restore the shutdown margin by boration:

AND (b) Be in hot shutdown within six hours.

C. 1 coerab!e Red C!unte- Cor rol A.scer bly 'RCC A N a

1. An RCCA cha!! be ec=idered inope=b!: if cne er more of the fc!!cv ing occu=

a. He RCCA de:S no &cp upon ::=cval af static =.j gripper ci! vc!! ge.

b. Se RCCA does =t step in properly -h= the proper vc!! ge nequ=ces are app!!:d ic the centre! red &iv: mech =i = co!!:. It ch !! th= be amumed

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c. If the b=k de==d pecition is greater th= cr equa! to 215 r. ps, c , !:= th= cr eq=1 to 30 3:eps, =d the red perition indienter ch=ne! hc= a mi=lignment frc= the b=h de==d p=ition ef 21 steps, the RCCA cha!! be amumed incpemb!e utilit ha 5:= :::: d te verify the it dee::tep properly.

Unit 1 - Amendment No. 15.3.10-3 Unit 2 - Amendment N,o.

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I C. ROD POSITION INDICATION  !

1 1

NOTE: Separate entry into TS 15.3.10.C.1.a. b. or e is allowed for each inoperable rad I position indicator and each bank of demand position indication. I

1. During nower operation >10 percent of rated thermal power. the rod position indication i

system and the bank demand nositionindication system shall be operable.

a. If one or more rod nosition indicators (RPI) are determined to be inonerable.

nerform the following actions:

(1) Within eight hours verify the position of the rods with inoperable RPIs by l usine movable incore detectors:

AND l (2) Once per shift check the nosition of the rods with inoperable RPIs by using excore detectors. or thermocouples. or movable incore detectors:

(3) If the above actions and associated completion times are not met. perform the actions in accordance with TS 15.3.10.B.I.b.

b. If one or more rods with inoperable RPIs have been moved in excess of 24 steos in one direction since the last determination of the rod's nosition. nerform the followine actions:

Unit 1 - Amendment No. 15.3.10-4 Unit 2 - Amendment No.

. - ~ . . .-

i 1 l

(1) Within four hours check the nosition of the rods with inocerable RPIs by using excore detectors. or thermocounles. or movable incore detectors:

(2) If the above action and associated completion time is not met. perform the actions in accordance with TS 15.3.10A1.b.

c. If bank demand position indication. for one or more banks. is determined to be inoperable. perform the following actions:

l (1) Once per shift verify that all RPIs for the affected banks are onerable: '

AND .

_ (2) Once per shift verify that the most withdrawn rod and the least withdrawn t rod of the affected banks are s12 steos anart. excent when the bank demand position is s30 stens or >215 stens. in this case. once _cer shift verify that the most withdrawn rod and the least withdrawn rod of the affected banks '

are <24 steps anart:

I 1

(3) If the above actions and associated comnletion times are not met. nerform .

l I the actions in accordance with TS 15.3.10.B.1.b. l n

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Unit 1 - Amendment No. 15.3.10-5 Unit 2 - Amendment No.

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2. For operation be:== 10% pc=r =d Rated Pe=r, Se p=ition of Se RCCA( ) l vc!S de f !!:d red p=ition indie::c ch:=e!(c) vi!! he ch=hed indi=ct!y hy ec= l

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b. Per cpem:icn be!cv 10% cf Rated Pe=r, nc specia! men!!cring i =qu!=d. ]

AD. BANK INSERTION LIMITS

- l 1

1

1. When the reactor is critical, ==p: for physic :=:: =d :=:=1 =d ==i= , the l shutdown banks shall be fully withdrawn.' Fully withdrawn is defined as a bank position equal to or greater than 225 steps. This definition is applicable to shutdown and control

}

banks. l l

1 If this condition is not me.Lperform the following actions: '

l a. Within one hour verify that the shutdown margin exceeds the applicable value as shown in Ficure 15.3.10-2: OR within one hour restore the shutdown marnin b.y i

boration:

l 1

AND l b. Within six hours fully withdraw the shutdown banks.

1 I

c. If the above actions and associated comnletion times are not met. be in hot 1

shutdown within the followine six hours.

l 2. When the reactor is critical, the control banks shall be inserted no further than the limits i

shown by the lines on Figure 15.3.10-1. E=eptic= :0 de i=ertion !!mit are pe=9ted for physi = :=t: =d :=tre! red ==ises* If this condition is not met. perform the l followine actions:

a. Within one hour verify that the shutdown margin exceeds the annlicable value as l

shown in Figure 15.3.10-2: OR within one hour restore the shutdown margin by l boration:

AND l b. Within six hours restore the control banks to within limits.

c. If the above actions and associated comoletion times are not met. be in hot shutdown within the followine six hours.

BE. POWER DISTRIBUTION LIMITS

1. Ilot Channel Factors 1

Unit 1 - Amendment No. 15.3.10-6 Unit 2 - Amendment No.

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,_ hot channel factors def'med in the basis must shall meet the following limits:

4 Fo(Z)s (2*50) p x K(Z) for P > 0.5 F9(Z) s 5.00 x K(Z) for Pso.5 ,

l F% < l.70 x [l + 0.3 (1-P)] I i

Where P is the fraction of full power at which the core is operating, K(Z) is the i function in Figure 15.3.10 3 and Z is the core height location of Fq. -

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1

b. If Fe(Z) exceeds the limit of Speci6 cation 15.3.10.E.1.a. within 6fteen minutes redtice thermal power until Fe(Z) limits are satis 6ed:

~

1 l

l (1) After thermal power has been reduced in accordance with Sneci6 cation 15.3.10.E.1.b. nerform the followine actions:

l (a) Within eight hours reduce the full power Power Range Neutron Flux -

l Hi_ch trio set.ooints b.y an amount ea.uivalent to the nower reduction i reauired in Soccification 15.3.10.E.1.b:

AND 4

4 l

Unit 1 - Amendment No. 15.3.10-7 Unit 2 - Amendment No.

6 &

fb) Within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> reduce Overoower and Overtemperature AT trip setooints by an amount equivalent to the nower reduction required in l . , Snecification 15.3.10.E.1.h;  !

AND (c) Verify that Fe(Z) will be within limits fc r the increased power level orior to increisine any setooints that have been reduced and thermal l cower above the limit specified in Specification 15.3.10.E.1.b.

! (d) If the above actions and associated completion times are not met. be in low power operation within the following six hours. I

c. If FU3n exceeds the limit of Specification 15.3.10.E.1.a. within four hours reduce thernial power to restore F0 g ot within limits OR nerform the following actions:

(1) Within four hours reduce thermal power to <50 nercent rated thermal power: ,

AND 1 i

(2) Within eight hours reduce the full power Power Range Neutron Flux -liigh '

trip setpoints to <58 percent rated thermal power. ,

i l

l l

In addition to the above actions. the following actions shall also be performed during the subseauent power escalation if Fag , had exceeded the limit of i Specification 15.3.10.E.1.a:

(31 Verify that FMg si within limits within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s: l AND (4) Verify that FEgi si within limits prior to thermal power exceeding 50 percent of rated thermiTpower:

AND l

. (5) Verify that FMg! is within limits nrior to thermal nower exceeding 75 nercent I

of rated thermiTpower:

AND (6) Verify _thaLEE m is within limits within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching >95 percent of rated therm _ 'vm_

(7) If the above actions and associated comnletion times are not met. be in hot l shutdown within the following six hours.

2. Axial Flux Difference a The indicated ! ! flux difference (AFD):ha!! be .aintained vithin the 21!c.ved

, cpemtional space defined by Figure 15.3.101 except during phys!= !=ts. The

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. . - . . . . . ..... . . . . . . ~.

equa! to 85% cf Rated Pc.ver and the !: nits of Specificatien 15.3.10.B.I .a are 4

catisfied. During suspencica of the specification, the ther:=! pc.ver sha!! be determined ic be !:m than or equal to 85% of rated therma! pcvccr at ! =t enee per r

Unit 1 - Amendment No.

15.3.10-8 Unit 2 - Amendment No.

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i NOTE- The axial Dux difference shall be considered outside limits when two or more operable excore channels indicate that axial Hux difference is outside I hmits. i

a. Durine power operation with thermal power '250 nercent of rated thermal power.

the axial Hux difference shall be maintained within the limits snecified in Ficure ~ ~

15.3.10 4. 1 (1) If the axial Cux difference is not within limits. within 15 minutes restore to within limits. If this action and associated completion time is not mrt perform the following actions:

(a) Reduce thermal nower until the axial Dux difference is within limits:

OR (b) Within three hours reduce thermal nower to <50 n. ercent of rated .

j thermal nower.

b. Ifit is necessary to restrict thermal nower to <50 percent of rated thermal power.

within the next four hours reduce the Power Rance Neutron Flux - Hich Trin setpoints to <55 nercent.

Unit 1 - Amendment No. 15.3.10 9 Unit 2 - Amendment No.

i

. # j l

1 l

l l l r ,

c. If the alarms used to monitor the axial flux difference are rendered inoperable, l verify thalthe. axial Hux difference is within limits for each onerable excore  !

channel once within one hour and every hour thereaften  !

l l

3. Ouadrant Power Tilt c fm

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a. During power oneration with thermal power greater than 50 nercent of rated thermal power. the indicated quadrant power tilt shall not exceed 2 percent. If this condition is not met. perform the following actions:

l l (1) Within two hours. reduce thermal power >2 percent from rated thermal power for each I percent ofindicated ouadrant power tilt: j AND i (2) Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and once ner seven days thereafter. verify that Fn_(Z) and E g are within the limits of Snecification 15.3.10.E.1.a:

AND (3) Upon comnletion of Sneci6 cation 15.3.10.E.3.a(2). calibrate the excore detectors. This action shall be comnleted n. rior to increasin_e thermal n. ower .

above the limit imposed by Specification 15.3.10.E.3.a(1):

AND (4) Verify that Fn(Z) and Fh are within the limits of Specification

~

J 5.3.10.E.1.a within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching rated thermal nower. or within i Unit 1 - Amendment No. 15.3.10-10 1

, Unit 2 - Amendment No.

I

1 l

. l l

. 1 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after increasine thermal nower above the limit imnosed by Specification 15.3.10.E.3.a(1). 1 (5) If the above actions and associated completion times are not met. within the following four hours reduce thermal power to s50 percent of rated thennal DGEL

b. If no quadrant power tilt alarms are available. within twelve hours and every twelve hours thereafler. verify that quadrant power tilt is within limits by performing calculations.

l l

eg. When one power range channel is inoperable and thermal power is greater than 75% ofrated thermal power, the quadrant pcycer tilt cha!! be confirmed a'; l acceptable within twelve hours and every twelve hours thereafter. verify that quadrant pg.wer tilt is within limits by use of the movable incore detectors at-least l ence per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. I F. AT-POWER PIIYSICS TESTS EXCEPTIONS

1. During the performance of at-power nhysics tests. the requirements of:

Specification 15.3.10.B. " Rod Operability and Bank Alignment Limits" Specification 15.3.10.D. " Bank Insertion Limits" Spscification 15.3.10.E.2. " Axial Flux Difference" Specification 15.3.10.E.3. "Ouadrant Power Till" are suspended. provided:

a. Thermal pmver is maintained <85 percent of rated thermal nower:

AND

b. Power Range Neutron Flux - High Trip setnoints are set at a maximum setting of 90 percent of rated thermal power:

2. Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to the initiation of nhysics tests. verify that Power Rance Neutron Flux - Iligh Trip setnoints are <90 percent of rated thermal nower.

3. If the shutdown margin is not within the limits of Specification 15.3.10. A.I . within 15 minutes initiate boration to restore the shutdown marein. AND within one hour susnend physics tests excentions.

4. If thermal power exceeds 85 percent of rated thermal power. within one hour reduce thermal power to <85 percent of rated thermal power. OR within one hour suspend physics tests exceptions Unit 1 - Amendment No. 15.3.10-11 l Unit 2 - Amendment No.

t

1 i

l e

5. If the Power Range Neutron Flux -liigh Trin setpoints are greater than 90 percent of l rated thermal nower. within one hour restore the Power Range Neutron Flux - Iligh Trip setpoints to <90 percent of rated thermal power. OR within one hour suspend nhysics 1ests excepliant l

6. Every hour. while at-power physics tests are in progress. verify that thermal power is <85 percent of rated thermal power.

7. At least once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. verifv F 7 (Z) and Fh are within the reauired limits.

G. LOW POWER PHYSICS TESTS EXCEPTIONS

1. During the nerformance oflow nower nhysics t sts. the recuirements of:

Specification 15.3.10.B. " Rod Operability and Bank Alignment Limits" Snecification 15.3.10.D. " Bank Insertion Limits" I Specification 15.3.10.E. " Power Distribution Limits" are suspended. provided the lowest RCS loop average temnerature is greater than the I minimum temnerature for criticality.

2. If the shutdown margin is not within the limits of Specification 15.3.10.A. within 15 minutes initiate boration to restore the shutdown marcin. AND within one hour susnend physics tests excentions.

3. If nower is not within limits. onen the rector trin breakers immediatelv.

4. Iflowest RCS loop average temperature is less than the minimum temnerature for criticality. within 15 minutes restore lowest RCS loon average temperature to within l limits. OR within 30 minutes be suberitical. l Ey. RCCA DROP TIMES

1. At operating !:mperature and full flow, With RCS temperature greater than the minimum temnerature for criticality and with both reactor coolant pumns running. the drop time of each RCCA shall be no greater than 2.2 seconds from the loss of stationary gripper coil voltage to dashpot entry. If this condition is not met. nerform the following actions:

a. If the reactor is critical. declare the rod untrippable:

OR I b. If the reactor is suberitical. maintain the reactor suberitical Basis Insertion Limits and Shutdown Marcin Unit 1 - Amendment No. 15.3.10-12 Unit 2 - Amendment No.

l l

l The-reaetivity centrc! concept is that reactivity change: accompanying changes in reacter pcwer are i

cempen=:ed by centrol red =ctica Reactivity change weinted with xenen, =marium, fuel depletica and large chang = in reac:c ;ce! ant :empe=ture (cpe= ting ::mpe=:ure to ce!d shutdown) are ecmpen=:ed by-chang = in the =!uble bcron concent=tica During power operation, the shutdown banks are fully withdrawn. Fully withdrawn is defined as a bank demand position equal to or greater than 225 steps. Evaluation has shown that positioning control rods at 225 steps, or greater, has a negligible effect on core power distributions and peaking factors. Due to the low reactivity worth in this region of the core and the fact that, at 225 steps, control rods are only inserted one step into the active fuel region of the core, positioning rods at this position or higher has minimal effect. This position is varied, based on a predetermined schedule, in order to minimize wear of the guide card in the guide tub = cf the RCCNs. RCCA's from the guide Glfill The control rod insertion limits provide for achieving hot shutdown by reactor trip at any time and assume the highest worth control rod remains fully withdrawn. A 10% margin in reactivity worth of the control rods is included to assure meeting the assumptions used in the accident analysis. See A reactor trip occurring during power operation vi!! put places the reactor into the hot shutdown eondition. In addition, the insertion limits provide a limit on the maximum inserted rod worth in the unlikely event of a hypothetical rod ejection and provide for acceptable nuclear peaking factors. The specified control rod insertion limits take into account the effects of fuel densification. The rods are withdrawn in the sequence of A, B, C, D with overlap between banks. The overlap between l successive control banks is provided to compensate for the low differential rod worth near the top and bottom of the core.

When the insertion limits are observed and the control rod banks are above the solid lines shown on Figure 15.3.10-1, the shutdown requirement is met. The maximum shutdown margin requirement I occurs at end of core life and is based on the value used in analysis of the hypothetical steam break  ;

accident. Figure 15.3.10-2 shows the shutdown margin equivalent to 2.77% reactivity at end-of-life l with respect to an uncontrolled cooldown. All other accident analyses assume 1% or greater reactivity shutdown margin. Shutdown margin calculations include the effects of axial power distribution. One-mav The accident analyses assume no change in core poisoning due to xenon, samarium or soluble boron.

If the shutdown margin requirements are not met. boration must be initiated promptly. Fifteen minutes is an adequate neriod of time for an operator to correctly align and start the required svstems and components. It is assumed that boration will be continued until shutdown margin requirements are met.

Bod Operability Requirements and Bank Alignment Limits The operability (e.g. trippability) of the shutdown and contro.1 rods is an initial assumption in all safety analyses that take credit for rod insertion upon reactor trin. Maximum rod misalignment is also an initial assumption in the safety analyses that directiv affect core nower distributions and Unit 1 - Amendment No. 15.3.10-13 Unit 2 - Amendment No.

)

assumptions of available shutdown margin. A rod cluster control assembly (RCCAT shall be considered operable if the RCCA drons upon removal of stationary gripper coil voltage.

Mechanical or electrical failures may cause a control rod to become inonerable or to become misaligned from its groun. Control rod inoperability or misalignment may cause increased power peaking due to the asymmetric reactivity distribution. This will also cause a red.uction in the total available rod worth for reactor shutdown. Therefore. control rod alienment and onerability are

~

related to core operation in design power peaking limits and the core design requirement of a minimum shutdown margin.

An incpe=b!: red impc= dditic=! dem'mt en the Ope = tem. 'Ile permirible ==b= cf incpemb!: centre! rc& is !!mited to one in ord= to !!=it the magnitude cf the opemting burden. From operating experience to date, an RCCA which steps in properly will drop when a trip signal occurs because the only force acting to drive the rod in is gravity. When it has been determined that a rod does not drop, extra duidc= == gin I g Ined by bemtica er by adjusting the in=-tion limit tc aseeurt for the venh of the incpemb!: :=tre! =d the shutdown margin calculation will need to include the worth of the inoperable control rod. Further experience indicates that control rods which do not step are usually affected by electrical problems. That is, normally the problem is in the rod control cabinets. If cpembi!!!y :=nct be =::c=d, the RCCA vi!! be dec!=ed incpemb!: =d corrective actic= := be id= te ecmp==t for the :=cciated = duction in dutdov n ==gm If thee i =c= th= = RCCA affected, = =de!y duider =uld be sta-ted. Sud = evolutica

uld have to be p=fc=ed in a de!!be=t mann= "ithcut =due p=== en the ope = ting peme=e! be == cf the ===! techniqu= to be u=d to a=c=modate the reactivity ch=g

=cciated ith the dutdcyn Rod cluster control assemblies (RCCAs). or rods. are moved by their control rod drive mechanisms 4

(CRDMs). Each CRDM moves its RCCA one step (approximatelv 5/8 inch) at a time. but at varying rates (steps per minute) depending on the signal output from the Rod Cor. trol System.

. The RCCAs are divided among control banks and shutdown banks. A groun consists of two or more RCCAs that are electrically naralleled to step simultaneousiv. A bank of RCCAs consists of one or two c_rouns that are moved in sta_c_ cered fashion. but always within one sten of each other. Each unit has four control banks and two shutdown banks.

When one or more rods are determined to be untrippable. there is a possibility that the required 4 shutdown marcin may be adversely affected. Under these conditions. it is imnortant to determine the shutdown margin. and ifit is less than the required value. initiate beration until the required shutdown margin is restored. The one-hour time limit is adequate for determining the shutdown margin and. if necessarv. for restoring the shutdown margin by boration. In this situation. shutdown margin verification must include the worth of the untrippable rod. as well as a rod of maximum worth.

4 l If the untrippable reds cannot be restored to an operable condition. the plant must be niaced in a condition where the LCQteguirements are not applicable. To achieve this status. the unit must be 1

placed in hot shutdown within six hours. This allows this nlant condition to be reached in an orderly manner. without challenging any plant systems.

Unit 1 - Amendment No. 15.3.10-14 Unit 2 - Amerdment No.

Limits on control rod alignment have been established and all rod nositions are monitored and controlled during nower operation to ensure that the nower distribution and reactivity limits dermed by the desien nower neakine and shutdown marein limits are nreserved.

1 If the misalignment condition cannot be readily corrected. thermal nower will be adjusted so that hot channel factors are maintained. and so that the requirements on shutdown margin and e_iected rod worth are preserved. Continued oneration of the reactor with a misaligned control rod is allowed if Eo(Z) and F?g are verified to be within their limits. When a control rod is misaligned. the assumptions tTat are used to determine the rod insertion limits. axial flux difference limits. and quadrant power tilt limits are not preserved. Therefore. the limits may not preserve the design peaking factors and FJZ) and Fag must be verified directiv by incore mapping.

Unon detection of a notential problem concerning one or more rods. a maximum of six hours is provided for troubleshooting activities. Immediately unon determining that one or more rods is inoperable. the anplicable actions in TS 15.3.10.B shall be performed. If after six hours. an operability determination has not yet been made. the rod (s) shall be declared inonerable and the annlicable actions in TS 15.3.10.B shall be nerformed.

Mi=!!gned RCCAS Rod Position Indication During power operation at greater than ten nercent of rated thermal power. the rod position indication system and the bank demand nosition indication system are required to be onerable. These systems are reauired to be onerable because the nosition of rods must be determined in order to ensure that rod alignment and insertion limits are being satisfied. Rod position accuracy is essential during power onerations. Power peaking. elected rod worth. or shutdown margin limits may be violated in the event of a design basis accident with rods onerating. undetected. outside of their required limits.

The various control rod banks (shutdown banks and control banks, A, B, C, and D) are each to be moved as a bank; that is, with all rods in the bank within one step (5/8 inch) of the bank position.

Direct information on rod position indication is provided by two methods: A digital count of actuating pulses which shows the demand position of the banks and a linear position indicator (LVDT) which indicates the actual rod position. The rod position indicator channel has a demon-strated accuracy of 5% of span (*11.5 steps). Therefore, an analysis has been performed to show that a misalignment of 24 steps cannot cause design hot channel factors to be exceeded. A single fully misaligned RCCA, that is, an RCCA 230 steps out of alignment with its bank, does not result in i exceeding core limits in steady-state operation at power levels less than or equal to rated power. In other words, a single dropped RCCA is allowable from a core power distribution viewpoint. If the misalignment condition cannot be readily corrected, the specified reduction in power to 75% will insure that design margins to core limits will be maintained under both steady-state and anticipated transient conditions. The eight (8) hour permissible limit on rod misalignment at rated power is short with respect to the probability of an independent accident.

Because the rod position indicator system may have a 12 step error when a misalignment of 24 steps is occurring, the Specification allows only an indiceted misalignment of 12 steps. However, when the Unit 1 - Amendment No. 15.3.10-15 Unit 2 - Amendment No.

bank demand position is greater than or equal to 215 steps, or, less than or equal to 30 steps, the consequences of a misalignment are much less severe. The differential worth of an individual RCCA is less, and the resultant perturbation on power distributions is less than when the bank is in its high differential worth region. At the top and bottom of the core, an indicated 24 step misalignment may be representing an actual misalignment of 36 steps.

The failure of an LVDT in itself does not reduce the shutdown capability of the rods, but it does reduce the operator's capability for determining the position of that rod by direct means. The operator has available to him the excore detector recordings, incore thermocouple readings and periodic incore flux traces for indirectly determining rod position and flux tilts should the rod with the inoperable LVDT become malpositioned. The excore and incore instrumentation will not necessarily recognize a misalignment of 24 steps because the concomitant increase in power density will normally be less than 1% for a 24 step misalignment. The excore and incore instrumentation will, however, detect any rod misalignment which is sufficient to cause a significant increase in hot channel factors and/or any significant loss in shutdown capability. The increased surveillance of the core if one or more rod position indicator channels is out-of-service serves to guard against any significant loss in shutdown margin or margin to core thermal limits.

The history of malpositioned RCCA's indicates that in nearly all such cases, the malpositioning occurred during bank movement. Checking rod position after bank motion exceeds 24 steps will verify that the RCCA with the inoperable LVDT is moving properly with its bank and the bank step counter. Malpositioning of an RCCA in a stationary bank is very rare, and ifit does occur, it is usually gross slippage which will be seen by external detectors. Should it go undetected, the time between the rod position checks performed every shift is short with respect to the probability of occurrence of another independent undetected situation which would further reduce the shutdown capability of the rods.

Any combination of misaligned rods below 10% rated power will not exceed the design limits. For this reason, it is not necessary to check the position of rods with inoperable LVDT's below 10%

power; plus, the incore instrumentation is not effective for determining rod position until the power level is above approximately 5%.

Power Distribution During nower operation. the global power distribution is limited by TS 15.3.10.E.2. " Axial Flux Difference." and TS 15.3.10.E.3. "Ouadrant Power Tilt." which are directiv and continuous)v .

measured process variables. These specifications. along with TS 15.3J 0.D. " Bank Insertion l_imits."

maintain the core limits on nower distributions on a continuous basis.

The purnose of the limits on the values of Fe(Z). the height denendent heat flux hot channel factor. is to limit the local peak nower density. The 61ue of Fe(Z) varies along the axial height (Z) of the core.

Eo(Z) is defined as the maximum local fuel rod linear power density divided by the average fuel rod liAcar_po_wer density. assuming nominal fuel pellet and fuel rod _ dimensions. Therefore. F g (Zlis a measure of the neak fuel pellet power within the reactor core.

Unit 1 - Amendment No. 15.3.10-16 Unit 2 - Amendment No.

l i

l l

Ee(Z) varies with fuel loading patterns. :ontrol bank insertion. fuel burnun. and chances in axial

'nower-distribution. Fn(Z)is measured neriodically usine the incore detector system.'These

' ~

measurements are gen"erally taken with the core at or near steady state conditions.

The purnose of the limits on FMf. the nuclear enthalpy rise hot channel factor. is to ensure that the fuel dnign criteria are not exceTed and the accident analysis assumntions remain valid. The desien limits on local and integrated fuel rod peak power density are exnressed in terms of hot channel l factors. Control of the core power distribution with respect to these factors ensures that local l conditions in the fuel rods and coolant channels do not challenge core integrity at any location during {

either normal operation or a nostulated accident analyzed in the safety analyses. l

' N F an, Nuclear Enthalpy Rise Hot Channel Factor,is defined as the ratio of the integral oflmear N

power along a fuel rod to the average fuel rod power. Imposed limits pertain to the maximum F 6n jn the core, that is the fuel rod with the highest integrated power. It should be noted that F"an is based on an integral and is used as such in the DNB calculations. Local heat flux is obtained by using hot channel and adjacent channel explicit power shapes which take into account variations in horizontal (x-y) power shapes throughout the core. Thus, the horizontal power shape at the point of maximum heat flux is not necessarily directly related to F"au-l EMai si sensitive to fuel loading patterns. bank insertion. and fuel burnup. FM , typically increases wiificontrol bank insertion and typically decreases with fuel burnun. l EM3 , is not directiv measurable but is inferred from a nower distribution map obtained with the movable incore detector system. Snecifically. the results of the three dimensional power distribution 1

map are analyzed by a computer to determine F Mi. This factor is calculated at least monthly. -

llowever. during nower operation. the global poder distribution is monitored by TS 15.3.10.E.? l

" Axial Flux Difference." and TS 15.3.10.E.3. "Ouadrant Power Tilt." which address directly and continuousiv measured orocess variables.

Design criteria have been chosen "chich are consistent "ith-the fue! integrity ana!yses. Rene relate 4e fissica gas release, pe!!et temperature and cladding mechanica! properties. A! e the minimum DNBR ia4hecere must not be ! = than the limit DNBR in nc=al operation er in short term 4ransient:

In-addition to the above, the pech linear power density must net exceed the limiting h"dft values which resu!: from the large break !c= cf ecolant accident analysis based upon the ECCS acceptance eriteria !imit of 2200"F His is required to meet the initial conditions azumed for !c= cf coelaat accidert To aid in specifying the limits en power distribution, the fc!!cv.ing het channel factors are defined:

F9(Z), Height Denender+ Heat F!ux Het Chnnne! Facte- is defmed as the !ccal heat flux en the Surface of a fue! red ;! core e! Vation 3 d!V!ded by the average fue! red hea! [!uX, al!cW!Cg [cr manufacturing to!:rances en fue! pe!!ets and reds. Impeced limits pertain to the maxinmm-F9(7,}4n the core.

Unit 1 - Amendment No. 15.3.10-17 Unit 2 - Amendment No.

F'q, Fng!=er:ng H=t Flux Het Chrnnel Fact =, is d:E=d = Se c!!cvc=ce un heat flux required for

==ufac =ing ic!::=ce. He = gin =ing fact = n!!c= for !ccal variatic= in :=ichment, pe!!::

de=ity =d diamet=, :.=f=: = cf Se fuel red =d eccetricity of Se gap bet ve= pe!!:t =d c!ad.

ComH=d :::! ica!!y, Se =: efvect in n f=t= cf 1.03 to be app!!:d ic fuel red :=f= h=: Sux For normal opemtion, it is no =ce=arj to me==e See qu= tit?=. !=:=d i It has been determined that, provided the following conditions are observed, the hot channel factor limits will be met:

1. Control rods in a single bank move together with no individual rod insertion differing by more than 24 steps from the bank demand position, when the bank demand position is between 30 steps and 215 steps. A misalignment of 36 steps is allowed when the bank position is less than or equal to 30 steps, or, when the bank position is greater than or equal to 215 steps, due to the small worth and consequential effects of an individual rod misalignment.

2. Control rod banks are sequenced with overlapping banks as described in Figure 15.3.10-1.

3. He fu!! !=gi e Gontrol bank insertion limits are not violated.

4. Axial power distribution control procedures, which are given in terms of flux difference control and control bank insertion limits, are observed. Flux difference refers to the difference in signals between the top and bottom halves of two-section excore neutron detectors. The flux difference is a measure of the axial offset which is defined as the difference in normalized power between the top and bottom halves of the core.

N The permitted relaxation of F 3n allows radial power shape changes with rod insertion to the insertion limits. It has been determined that provided the above four conditions 16:cugh i are observed, these hot channel factor limits are met. In Specification 15.3.10.Bg.1.a, F q is arbitrarily limited for p s 0.5 (= cept f= !cv/ pc;ver phy != := s).

An upper bound envelope of 2.50 times the normalized peaking factor axial dependence of Figure 15.3.10-3 consistent with the Technical Specifications on power distribution control as given in Section 15.3.10 was used in the large and small break LOCA analyses. The envelope was determined based on allowable power density distributions at full power restricted to axial flux difference (AI) values consistent with those in Specification 15.3.10.Bg.2.

The results of the analyses based on this upper bound envelope indicate a peak clad temperature of less than the 2200 F limit. When an F q measurement is taken, both experimental error and manufacturing tolerance must be a!!cwed f= taken into account. Five percent is the appropriate allowance for a full core map taken with the moveable incore detector flux mapping system and three percent is the appropriate allowance for manufacturing tolerance. In the design limit of F"3n, there is eight percent allowance for uncertainties which means that normal operation of the core is expected to result in a design F"3n s 1.70/1.08. The logic behind the larger uncertainty in this case is that gs follows:

Unit 1 - Amendment No. 15.3.10-18 Unit 2 - Amendment No.

L, .

j l

[ (a) Normal perturbations in the radial power shape (i.e., rod misaligmnent) affect F"ui, in most j cases without necessarily affecting Fn.

l (b) While the operator has a direct influence on F n through movement of rods, and can limit it to I the desired value, he has no direct control over F*ai. )

i (c) An error in the predictions for radial power shape which may be detected during startup physics  ;

tests can be compensated for in Fq by tighter axial control; but compensation for F% is less readily available.

"' hen measurement of Ph is ken, experiment:1 e rc must be !!cwed for and four percent i: Se appropriate !!cwanee for a fu!! core =cp takrii Se mcVeab!: incere detecter flux mapping l cystem '

Measurements of the hot channel factors are required as part of startup physics tests, at least each full power month operation, and whenever abnormal power distribution conditions require a reduction of core power to a level based upon measured hot channel factors. The incore map taken following j initial loading provides confirmation of the basic nuclear design bases including proper fuel loading patterns. The periodic monthly incore mapping provides additional assurance that the nuclear design bases remain inwiolate and identify operational ,

anomalies which would, otherwise, affect these bases. I The measurrd hot channel factors are increased as follows:

(a) The measurement of total neaking factor. Fo"S!!. shall be increased by three percent to account for manufacturing tolerance and further increased by five cercent to account for measurement )

error.

(b) The measurement of enthalpy rise hot channel factor. F( shall be increased by four percent to account for mensurement error.

Axial Power Distribution The limits on axial flux difference (AFD) assure that the axial power distribution is maintained such l

that the F n (Z) upper bound envelope of q F " times the normalized axial peaking factor [K(Z)] is not exceeded during either normal operation or in the event of xenon redistribution following power changes. This ensures that the power distributions assumed in the large and small break LOCA analyses will bound those that occur during plant operation.

Provisions for monitoring the AFD on an automatic basis are derived from the plant process i computer through the AFD monitor alarm. The computer determines the AFD for each of the operable excore channels and provides a computer alarm if the AFD for at least 2 of 4 or 2 of 3 operable excore channels are outside the AFD limits and the reactor power is greater than 50 percent of Rated Power.

Unit 1 - Amendment No. 15.3.10-19 Unit 2 - Amendment No.

Ouadrant Tilt Ihe_ quadrant. tilt limit ensures that the gross radial power distribution remains consistent with the design values used in the safety analyses. Precise radial power distribution measurements are made during startup testing.after refueling. and periodically during power operatioA The poet density at any point in the core must be limited so that the fuel desien criteria are maintained. Together. specifications associated with axial flux difference. auadrant tilt and control rod insertion limits nrovide limits on nrocess variables that characterize and control the three dimensional nower distribution of the reactor core. Control of these variables ensures that the core operates within the fuel design criteria and that the power distribution remains within the bounds used in_the safety analyses.

The excore detectors are somewhat insensitive to disturbances near the core center or on the major axes. It is therefore possible that a five percent tilt might actually be present in the core when the 1 excore detectors respond with a two percent indicated quadrant tilt. On the other hand, they are overly responsive to disturbances near the periphery on the 45 axes.

Tilt restrictions are not applicable during the startup and initial testing of a reload core which may I have an inherent tilt. During this time sufficient testing is performed at reduced power to verify that l the hot channel factor limits are met and the nuclear channels are properly aligned. The excore detectors are normally aligned indicating no quadrant power tilt because they are used to alarm on a rapidly developing tilt. Tilts which develop slowly are more accurately and readily discerned by incore measurements. The excore detectors serve as the prime indication of a quadrant power tilt. If a channel fails, is out-of-service for testing, or is unreliable, two hours is a short time with respect to the probability of an unsafe quadrant power tilt developing. Two hours gives the operating personnel sufficient time to have the problem investigated and/or put into operation one of several possible i alternative methods of determining tilt. l Physics Tests Exceptions The primary purnose of the at-nower and low nower physics tests is to permit relaxations of existing specifications to allow performance ofinstrumentation calibration tests and special physics tests. The at-power specification allows selected control rods and chutdown rods to be positions outside their specified alignment and insertion limits lo conduct physics tested at power. The nower level is limited to s85 percent of rated thermal power and the power range neutron flux trip sclpoint is set at maximum of 90 percent of rated thermal power. Operation with thermal nower s85 nercent of rated thermal nower during nhysics tests provides an acceptable thermal margin when one or more of the applicable specifications is not being met. The Power Range Neutron Flux - High trip setpoint is reduced so that a similar marein exists between the steady-state condition and the trin setnoint that exists during normal operation at rated thermal power.

The low nower specification allows selected control and shutdown rods to be positioned outside of their snecified alienment and insertion limits to conduct nhysics tests at low power. If nower exceeds two percent. as indicated by nuclear instrumentation. during the performance oflow power nhysics Unit 1 - Amendment No. 15.3.10 20 Unit 2 - Amendment No.

1 1

tests. the only acceptable action is to onen the reactor trip breakers to prevent oneration of the reactor bevond its design limits. Immediately opening the reactor trip breakers will shut down the reactor and orevent operation of the reactor outside ofits design limits. If the RCS lowest loop average  ;

lemperature falls below the minimum temperature for criticality. the temperature should be restored '

within 15 minutes because coeration with the reactor critical and temocrature below the minimum temperature for criticality could violate the assumptions for accidents analyzed in the safety analyses.

If the temperature cannot be restored withm 15 minutes. the plant must be made suberitical within an additional 15 minutes. This action wi 1 place the plant in a safe condition in an orderly manner without challenging plant systems.

1 l

1 i

Unit 1 - Amendment No. 15.3.10-21 Unit 2 - Amendment No.

. . - . - . . . . . - . - - . . . ~ - . - . ~ - - -

5  !

TABLE 15.4.1-2

, MINIMUM FREOUENCIFS FOR EOUIPMENT AND SAMPLING TESTS IC11 Freauency

1. Reactor. Coolant Samples Gross Beta-gamma 5/ week")

activity (excluding tritium) i

Tritium activity Monthly l

) Radiochemical E Semiannually ")

, Determination i

j Isotopic Analysis for ' Every two weeks")

Dose Equivalent I-131 Concentration

Isotopic Analysis for a.) Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> L lodine including I-131, whenever the specific

[ I-133, and I-135 activity exceeds 1.0 Ci/.

4 gram Dose Equivalent I-131 1

or 100/E Ci/ gram.")

J b.) One sample between 2 and 6 l hours following a thermal power i change exceeding 15% of rated i power in a one-hour period.

J

) Chloride Concentration 5/ week")

l Diss. Oxygen Conc. 5/ week") -

Fluoride Cone. Weekly -

i

2. Reactor Coolant Boron Boron Concentration Twice/ week 1

i 3. Refueling Water Storage Boron Concentration Weekly")

l Tank Water Sample i

1 l ' 4.~ Boric Acid Tanks - Boron Concentration Twice/ week and after l each BAST concentration

'. change when they are -

being relied upon as a

[!i source of borated water.

5. Spray Additive Tank NaOH Concentration Monthly

6. Accumulator Boron Concentration Monthly Unit 1 - Amendment No.

Unit 2 - Amendment No. Page lof 44

. . 1 TABLE 15.4.1-2 (Continued)

Issi Freauency

7. Spent Fuel Pit a) Boron Concentration Monthly b) Water Level . i Verification Weekly

8. Secondary Coolant Gross Beta-gamma Weekly (6)

Activity or gamma isotopic analysis  ;

j lodine concentration Weekly when gross Beta-gamma activity equals or exceeds j

~

1.2 pCi/cc (6) l

9. Control Rods a) Rod drop times of all Each refueling or J fulllength rods0) after maintenance that could affect I proper functioning ")

b) Rodworth measurement Following each refueling shutdown ,

prior to commencing power operation

10. Control Rod Partial movement of Every 2 weeks ('8) i all rods  ;

11. Pressurizer Safety Valves Set point Every five years (") '

i

12. Main Steam Safety Valves Set Point Every five years (")

1

13. Containment isolation Trip Functioning Each refueling shutdown i i

14. Refueling System Interlocks Functioning Each refueling shutdown l

15. Service Water Syvem Functioning Each refueling shutdown

16. Primary System Leakage Evaluate Monthly (6)

17. Diesel Fuel Supply Fuel inventory Daily

18. Turbine Stop and Governor Functioning Annually (6)

Valves 19.- Low Pressure Turbine Visual and magnetic Every five years Rotor 1nspection "I particle orliquid penetrant

20. Boric Acid System Storage Tank and Daily @)

piping temperatures 2 temperature required by Table 15.3.2-1 l

l Unit 1 - Amendment No.

Unit 2 - Amendment No. Page 20f 45

TABLE 15.4.1-2 (Continued)

P Inst Freauency

21. ' PORV Block Valves a. Complete Valve Cycle Quarterly ()

. b. Open position check Every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (")

22. Integrity of Post Accident Evaluate Each refueling Recovery Systems Outside cycle Containment

23. Containment Purge Supply Verify valves are Monthly (')

and Exhaust Isolation locked closed Valves

24. Reactor Trip Breakers a. Verify independent Monthly (')

operability of automatic shunt and undervoltage trip functions.

b. Verify independent Each refueling operability of man- shutdown ual trip to shunt and undervoltage trip functions.

25. Reactor Trip Bypass a. Verify operability Prior to Breakers of the undervoltage breaker use trip function.

b. Verify operability Each refueling of the shunt trip shutdown functions.

c. Verify operability Each refueling of the manual trip shutdown to undervoltage trip functions.

26. 120 VAC Vital Instr. Verify Energized (12) Shiftly Bus Power

27. Power Operated Relief Operate UO Each shutdownos)

Valves (PORVs),

PORV Solenoid Air Control Valves, and Air System Check 28.- Atmospheric Steam Dumps Complete valve cycle Quarterly

29. Crossover Steam Dump System Verify operability of Quanerly each steam dump valve.

o Unit 1 - Amendment No.

Unit 2 - Amendment No. Page 3of41

TABLE 15.4.1-2 (Continued)

30. Pressurizer Heaters Verify that 100 KW of Quarterly heaters are available.

31. CVCS Charging Pumps Verify rability Quarterly pumps

32. Potential Dilution in Verify operability of Prior to placing plant in Progress Alarm alarm. cold shutdown.

33. Core Power Distribution Perfonn power distribu- MonthlyE tion maps usin_e movable incare detector system to confinn hot channel factors.

34. Shutdown Margin Perform shutdown margin Daily @ q calculation. 4 (1) Required only during periods of power operation.

(2) E determination will be started when the gross activity analysis of a filtered sample indicates 210 Ci/cc and will be redetermined if the primary coolant gross radioactivity of a filtered sample increases by more than 10pCihc.

(3) Drop test shall be conducted af rated reactor coolant flow. Rods shall be dropped under both cold and hot condition, but cold d op tests need not be timed.

(4) Drop tests will be conducted in the hot condition for rods on which maintenance was perfonned.

(5) As accessible without disassembly of rotor.

(6) Not required during periods of refueling shutdown.

(7) At least once per week during periods of refueling shutdown.

(8) At least three times per week (with maximum time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> between samples) during periods of refueling shutdown.

(9) Not required during periods of cold or refueling shutdown, but must be performed prior to exceeding 200 F if it has not been performed during the previous surveillance period.-

(10) Sample to be taken after a minimum of 2 EFPD and 20 days power operation since the reactor was last suberitical for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or longer.

(11) An approximately equal number of valves shall be tested each refueling outage such that all valves will be tested within a five year period. If any valve fails its tests, an additional number of valves equal to the number originally tested shall be tested. If any of the additional tested valves fail, all remaining valves shall be tested.

(12) The specified buses shall be determined energized in the required manner at least once per shift by verifying correct static transfer switch alignment and indicated voltage on the buses.

(13) Not required if the block valve is shut to isolate a PORV that is inoperable for reasons other than excessive seat leakage.

(14) Only applicable when the overpressure mitigation system is in service.

(15) Required to be performed only if conditions will be established, as defined in Specification 15.3.15, where the PORVs are used for low temperature overpressure protection. The test must be performed prior to establishing these conditions.

Unit 1 - Amendment No.

Unit 2 - Amendment No. Page 4of 41

, TABLE 15.4.1-2 (Continued)

(16) Test valve operation in accordance with the inservice test requirements of the ASME Boiler and Pressure Vessel Code,Section XI.

(17). Operability of charging pumps is verified by ensuring that the pumps develop the required flowrate, as specified by the In-Service Test Program.

(18) Not required to be performed if the reactor is suberitical.

(19) Required only when the B AST(s) are relied upon as a source of borated water.

(20) Perform during nower operation at efTective full power monthly intervals. Following a refueling shutdown a power distribution map shall be nerformed prior to exceeding 908/t of rated thermal oower.

(21) Only anolicable during low-oower ohysics testing.

Unit 1 - Amendment No.

Unit 2 - Amendment No. Page Sof44

- - _ _ - _ _ _ _ _ .