Text

Amend 40 to License DPR-18,authorizing Tech Specs Re Control Rod Position Indication & Control Rod Misalignment
	ML17258A991
Person / Time
Site:	Ginna
Issue date:	04/17/1981
From:	Crutchfield D; Office of Nuclear Reactor Regulation
To:	;
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ML17258A992	List: ML17258A991; ML17258A993; ML17258A994;
References
	NUDOCS 8104300527
	Download: ML17258A991 (26)
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 ROCHESTER GAS AND ELECTRIC CORPORATION DOCKET NO. 50-244 R.

E.

GINNA NUCLEAR POllER PLANT AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No. 40 License No.

DPR-18 1.

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

A.

The application for amendment by Rochester Gas and Electric Company (the licensee) dated August 29, 1980 (transmitted by letter dated September 3, 1980),

complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission'.s rules and regulations set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance

{i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities wi 11 be conducted in compliance with the Commission's regulations; D.

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

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

810.43005@/

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and by changing paragraph 2.C(2) of Provisional Operating License No.

DPR-18 to read as follows:

(2)

Technical S ecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.

4O

, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

April 17, 1981 ennss M. Crutchfield, C

ef Operating Reactors Branch b'5 Division of Licensing

ATTACHMENT TO LICENSE AMENDMENT NO. 40 PROVISIONAL OPERATING LICENSE NO. DPR-18 DOCKET NO. 50-244 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages.

The revised pages contain the captioned amendment number and marginal lines which indicate the area of changes.

REMOVE

3. 5-4 3.5-4a 3.10-3 3.10-5 3.10-6 3.10-7 3.10-8 3.10-8a 3.10-8b 3.10-8c 3.10-9 3.10-10 Fig. 3.10-1 Fig. 3.10-2 Fig. 3.10-3 4.1-5 4.1-6 4.1-8 4.1-9 4.1-10 INSERT

3. 5-4 3.5-4a 3.10-3 3.10-5 3.10-6 3.10-7 3.10-8*

3.10-8a 3.10-8b*

3.10-8c*

3.10-9 3.10-10 3.10-11 3.10-12 3.10-13 3.10-14 4.1-5 4.1-6 4.1-8 4.1-9*

4.1-10 These pages are included for pagination purposes only.

NO.

I'IINC'I'IONAI. IINl'I' l.

'fllrbinc Trip 12.

SLe;lm I'low Feellw;lt.cr fI ow nl I sllla Lch w I Lh I 0 SLU> IIII Gene rlILor I.cve I NO.

OF CIIANNI'.I.S 2/loop NO.

OF CIIANNEI,S

'fO TRJP 1/loop 3

HIN.

OPERABT,I'.

CIIANNELS I/loop 4

>HIN.

DEOleEE OF III;DUNDANCY I/loop 5

PERH I SSA13I,E BYPASS CONDITIONS 6

OPERA'I'OI< ACTION lF CONDITIONS OF COI.UNPIN 3 OR 5 CANNOT BE PIET Haintain 50% of rated power Haintain hot sllutdown 13.

I.o l.o Steam Genera-Lol. Water I.i vel 3/loop 2/ J.oop 2/I.oop 1/loop Haintain hot shuLdown 14.

IInllervolt,age 4

VV Bus 2/bus 1/bus I/bus Haintain hot.

shUtdown 15.

UnllerI'rcllucncy 4 KV Bus 2/hus 1/bus (bot.h busses) 1/bus HainLain hot sIIUL(lown 16.

(Jua(lr;IUL powe I.

Lilt Inonitor.

(upper f. lower.

cx-core >>cutro>>

<leLertor..s) or.

I>og I nil1 vlllua I upper I

lower.

ion chamber currents once/hr h al'Ler a Ioall change of 10% or al'Ler 30" of control rod motion I'Ia I.n L Il n II0t shut. down

HO.

FUNC'I'10NAI UH LT 17.

Circ>>lating WaLer I'.ood 1 rotect

> on NO.

OF CIIAHHI'.I.S 2

NO.

OF CIIANNE],S TO TRlP 3

IllN.

OPI'.RAIIT.I'.

CIIANNEI.S 1lfN.

I)ECREE OF REI)UNI)ANCY 5

PERlll SSAIII,E 1)YPASS CONI)ITIOHS 6

OPERATOR ACTION IF CONDIT10NS OF COLUNN 3 OR 5 CANNOT IIE I'lET 4J V>

I A.

Screenhouse II ~

Co>Idol>so Power. operation may be continued for a period of up to 7

days with I channel inoperable or for a period of 24 hrs.

w-ith Lwo channels inoperabl>>.

Pover operation may be continued for a period of up to 7

days with 1 channel i>>operable or for a period of 24 hrs.

with Lwo channels inoperable.

HO'I'E I:

When bJock co>>>lition exists, u>ainLain nor>>>;>1 operatior>

F.I'.

=

F>>l I Power HoL A>>l ical>le c>

+

0 ll If a I'>>>>etio>>:>I unit is operating with the minim>>m operable channels>

Lhe n>>u>ber of cha>u>eis Lo trip the reactor vill be column 3 less column 4.

. A ch;>>u>el is co>)si<lered operable with 1 out of 2 logic or 2 out of 3 logic.

average power tilt ratio shall be determined once a day by at least one of the following means:

a.

Movable detectors b.

Core-exit thermocouples

3. 10. 2. 2 Power distribution limits are expressed as hot channel factors.

At all times, except during low power physics tests the hot channel acto must meet the following lim's:

Fg (Z)

= (2. 32/P) *K(Z) for P o5 Fg (Z)

= 4. 64*K(Z)

N Fa.rj

=

2 '2

.56P for P ~.5 for P

~.75 F~HN

= 1.80 for P

~.75 where P is the fraction of rated power at wt.ich the co e is operating, K(Z) is the func"ion given by Figu e 3.10-3, and Z is the height in the core.

The measured FqN shall be increased by tnree percen to yield F<. Iz the measured Fg or F~hN exceeds the limiting value, with due allowance for measurement error, the max'mum allowable reactor power level and the Nuclear Ove power Trip set point, shall be reduced one percent for each percent whi'ch F~P or Fg exceeds the limiting value, whichever is more restrictive.

Zf the hot channel actors cannot be reduced below the limiting values within one day, the Overpower hT trip setpoint and the Overtemperature aT trip setpoint"~hall be simila lv reduced.

3.10.2.3 Except for physics tests, if the quadrant to average power tilt ratio, exceedsl. 02 but is less than 1. 12

then within two ho rs

a.

Correct the situation, or b.

Determine by measurement the hot channel factors, and apply Specification 3.10. 2.2, or c.

Limit power to 75% of rated power, 3.10-3 Amendment No.+ 40

3.10.3 3.10.3.1 3.10.3.2 3.10.4 3.10.4.1 3.10.4.2 3.10.4.3 3.10.4.3.1 3.10.4.3.2 Control Rod Dro Time While critical, the individual full length (shutdown and control) rod drop time from the fully withdrawn position shall be less than or equal to 1.8 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with:

a.

T greater than or equal to 540'F, and avg b.

All reactor coolant pumps operating.

With the drop time of any full length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to criticality.

1 d,~h While critical, and except for physics testing, all full length (shutdown and control) rods shall be operable and positioned within i 12 steps (indicated position) of their group step counter demand position.

With any full length rod inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untripable, determine that the shutdown margin requirement of Specification 3.10.1.1 is satisfied within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in hot shutdown within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

With one full length rod inoperable due to causes other than addressed by 3.10.4.2,

above, or misaligned from its group step counter demand position by more than i 12 steps (indicated position), operation may continue provided that within one hour either:

The rod is restored to operable status within the above alignment requirements, or The rod is declared inoperable and the shutdown margin requirement of Specification 3.10.1.1 is satisfied.

Ope ations may then continue provided either:

a

~

The remainder of the rods in the group inoperable rod are aligned to within t of the inoperable rod within one hour, maintaining the limit of Specif'cation or with the 12 steps wh'e 3.10.1.3; b.

The power level is reduced to less than or equal to 75% of rated power within the next one hour, and the high neutron flux trip setpoint is reduced to less tnan or equal to 85% rated power within the next 3.10-5 Amendment No. 40

four hours (total of six hours) and the following evaluations are performed:

(i)

The shutdown margin requirement of Speci-fication 3.10.1.1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

(ii)

A power distribution map is obtained from

~e movable incore detectors and F (Z) and F5H are verified to be within thei9 limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

(iii)

A reevaluation of each accident analysis of Table 3.10-1 is performed within 5 days; this reevaluation shall confirm that the

'reviously analyzed results of these accidents remain valid for the duration of operation under these conditions.

C.

if power has been restricted in accordance with (b)

above, then following completion of the evaluation identified in (b), the power level and high neutron flux trip setpoint may be r adjusted based on the results of the evaluation provided the shutdown margin requirement of Specification 3.10.1.1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

3. 10.4

~ 4 3.10.5 3.10.5.1 3.10.5.2 With two or more full length rods inoperable or misaligned from the group step counter demand position by more than 2 12 steps (indicated position),

be in hot shutdown within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Control Rod Position indication Sv tems While critical, the analog rod positior. indication system and the step counters shall be operable and capable of determining the control rod positions within 2 12 steps.

With a maximum of one analog,rod position indicator'er bank inoperable either:

a

~

Determine the position of th'e non-indicat'ng rod(s) indirectly bv the movable incore detectors at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and immediately after any motion of the non-indicatina rod which exceeds 24 steps in one direction since the last determination of the rod's position, or b.

Reduce the power to less than 50/ of rated power within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

3.10.5,3 With a maximum of one step counter per bark inoperable either:

3.10-6 Amendment No.~ 40

a.

Verify that all analog rod position indicators for the affected bank are operable and that the most withdrawn rod and the least withdrawn rod of the bank are within a maximum of 12 steps of each other at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or b.

Reduce the power to less than 50% of rated power within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

Basis:

The reactivity control concept is that reactivity changes accompanying changes in reactor power are compensated by con+rol rod motion.

Reactivity ch'anges associated with xenon,

samarium, fuel depletion, and large changes in reactor coolant temperature (operating temperature to cold shutdown) are compensated by changes in the soluble boron concentration.

During power operation, the shutdown groups are fully withdrawn and control. o reactor power is by the control groups.

A reactor trip occurring during power operation will put the reactor into the hot shutdown condition.

The control rod insertion limits provide for achieving hot shutdown by reactor trip at any time, assuming the highest worth control rod remains fully withdrawn with sufficient margins t~1peet the assumptions used in the accident analysis.

ln addition, they 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 lines shown on Figure 3.10-1 meet the shutdown requirement.

The maximum shutdown margin requirement occurs at end-of-cycle life and is based on the value used in analysis of the hypothetical steam break accident.

Early in cycle life, less shutdown margin is requi ed, and Figure 3.10-2 shows the shutdown margin equivalent to that which is required at end-of-life with respect to an uncontrolled cooldown.

All other accident analyses are based on 1Ã reactivity shutdown margin.

An upper bound envelope of 2.32 times the normalized peaking factor axial dependence of Figure 3.10-3 has been determined from extensive analyses considering operating maneuvers consistent with the Technical Specifications on power distribution control as given in Section 3.10.

The results of the loss of coolant accident analyses based on this upper bound envelope demonstrate compliance with the Final Acceptance Criteria limit for Emergency Core Cooling Systems.

Amendment No. 34,~ 40 3.10-7

When an F~ measurement is taken, both experimental error and manufacturing tolerance must be allowed for.

Five percent is the appropriate allowance for a full core map taken with the movable incore detector flux mapping system and three percent is the appropriate allowance for manufacturing tolerance.

When a

measurement of F is taken, experimental error must be allowed for aN 4 percent is the appropriate allowance for a full core map with the movable incore detector flux mapping system.

Neasurements of the hot channel factors are reauired as part of startup physics tests, at least each full power month of operation, and whenever abnormal power distribution conditions require a reduction of core power to a level based on measured hot channel factors.

The incore map taken following initial loading provides confirmation of the basic nuclear Amendment Ho.

40 3.10-8

design bases including proper fuel loading pattern.

The periodic incore mapping provides additional assurance that the nuclear design bases remain inviolate and identifies operational anomolies which might, otherwise, affect these bases.

For normal operation, it is not necessary to measure these quantities.

Instead it has been determined that, provided certa'n conditions are observed, the hot channel factor limits will be met; these conditions are as follows:

1. Control rods in a single bank move together with no individual rod insertion differing by more than 15 inches from the bank demand position.

2. Control rod banks are sequenced with overlapping banks as described in Specification 3.10.

3.

The full length control bank insertion limits are not violated.

4. Axial power distribution limits which are given in terms of flux difference limits and control bank insertion limits are observed.

Flux difference is qT q> as defined in Specification 2.3.1.2d.

The permitted relaxation in F& with reduced power allows radial power shape changes with rod insertion to the insertion limits.

Xt has, been determined that provided the above conditions 1 through-4-are-observed,

~

these hot channel factors limits are met.

Xn specification 3.10 F~ is arbitrarily limited for P<0.5 (except for low power physics tests).

The limits on axial power distribution re-ferred to above are designed to minimize the effects of xenon redistribution on the axial power distribution during load-follow maneuvers.

Basi-cally, control of flux difference is required to limit the difference between the current value of Flux Difference

(>I) and a reference value which corresponds to the full power equilibrium value of Axial Offset (Axial Offset = ~X/fractional power).

The reference value of flux difference varies with power level and burnup but expressed as axial offset it varies primarily with burnup.

The technical specifications, on power distribution assure that the F~ upper bound envelone of 2.32 times Figure 3.10-3 is not exceeded and xenon distributions are not developed which, at a

later time, could cause greater local power peaking even though the flux difference is then within the limits.

3.10-8a Amendment No.~ 40

The target (or reference) value of flux difference is determined as follows.

At any time that equilibrium xenon conditions have been established, the indicated flux difference is noted with part length rods withdrawn from the core and with control Bank D more than 190 steps withdrawn.

This value, divided by the fraction of full power at which the core was operating is the full power value of the target flux difference.

Values for all other core power levels are obtained by multiplying the full power value by the fractional power.

Since the indicated equilibrium value was noted, no allow-ances for excore detector error are necessary and indicated deviation of

+

5 percent sI is permitted from the indicated reference value.

During periods where extensive load following is required, it may be impossible to establish the required core conditions fo measuring the target flux difference every month.

For this reason, two methods are permissible for updating the target flux difference.

Strict control of the flux difference (and rod position) is not as neces'sary during part, power operation.

This is because xenon distribution control at, part power is not as significant as the control at full power and allowance has been made in predicting the heat flux peaking factors for less strict control at part'ower; Strict control of the flux difference is not possible during certain physics tests, control rod exercises, or during the required periodic excore calibration which require larger flux differences than permitted.

Therefore, the specifications on power distribu-tion are not applicable during physics tests, control rod exercises, or excore calibrations; this is acceptable due to the extremely low probability of a significant accident occurring during these opera-tions.

Excore calibration includes that period of time necessary to return to equilibrium oper'ating conditions.

In some instances of rapid plant power reduction'utomatic rod mot'on will cause the flux difference to deviate from the target band when the reduced power level is reached.

This does not necessarily affect the xenon distribution sufficiently to change the envelope of peaking factors which can be reached on a subsequent return to full power within the target

band, however to simplify the specification, a limita-tion of one hour in any period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is placed on operation outside the band.

This ensures that the resulting xenon distributions are not significantly

3. 10-8b Amendment No.~ 40

gi,5.Ã

\\

V I

r w(

e'g V

s t

/

~

s

~"

differen from Chose resulting from <<eration within..,,:--i the targeC'band.

The instantaneous consequence of-'.

".;"-.'eing outside the band,'rovided,rod'nsertion'limits'"'.'.-:"I

-I are observed, is not worse than 'a 10 percent increment, in peaking factor for flux diffexence in the range.::."---<

+14 percent to -14 percent'(+ll'percent to -11 percent..'ndicated) increasing by +1 percent of each 2 percent decrease in rated power.

Therefore, while the deviation.

exists the power level is limited.to 90='erce'nt or-.

lower depending on the indicated flux difference..

w n

If, for any reason, flux difference is not controlled within the

+

5 percent band for as long a period as one::-

hour, then xenon distributions may be significantly cnanged and opexation at 50 percent is required to pro-tect against potentially more severe consequences of some accidents.

As discussed

above, the essence of the limits is to maintain the xenon distribution in the core as close'.".,'I to the equilibrium full power condition as possible.

This is accomplished, without part length rods, by using the chemical volume control system to position the full length control rods to produce the required indication flux difference.

The effect of exceeding the flux difference band't; or

.below half power is approximately half -as great as iC would be at 900 of rated power, where the effect of deviation has'been evaluated.

~

The reason for xequiring hourly logging is to pxovide-'...',.',

continued surveillance of the flux difference if the normal alarm functions are out of service. 'It is intended that this surveillance would be temporary until 'the alarm functions are restored.

The quadrant power tilt ratio limit assures that the I

radial power distribution satisfies the design values used in the power capability analysis.

Radial power distribution measurements are made during startup

~

'esting and periodically during power operation.

The limit of 1.02 at which corrective action is required'"

provides DilB and linear heat generation rate protection with x-y plane power tilts.

A limiting tilt of 1;025 can be tolerated befoxe the margin'or uncerCainity in F< is depleted.

Therefore, the limi:ting tilt has bien set as 1.02.

To avoid unnecessary power changes, the.

operator is allowed two hours in which to verify the tilt reading and/or Co determine and correct the cause of the tilt.

'Should this action verify a tilt in excess of 1.02 which remains uncorrected, the margin for uncertainty in F" and F

is reinstated by reducing the power by 2% fo9 each f5rcent of tilt above 1.0, in accordance with the 2 to 1 ratio above, or as required by the restriction on peaking factors.

t 3.'10-8c ue Amendment No./ py qp

The two hours in 3.10.2.3 are acceptable since complete rod misalignment (full-length control rod l2 feet out of alignment with its bank) does not result in exceeding core safety'limits in steady state operation. at rated power and is short with respect to pr'obability of an independent accident.

If instead of determining the hot channel factors, the operator decides to reduce

power, the specified 75%'power maintains the design margin to core safety limits for up to a 1.12 power tilt, using the 2 to 1

ratio.

Reducing the overpower trip set point ensures that the protection system basis is maintained for sustained plant operation.

A tilt ratio of 1.12 or more is indicative of a serious performance anomaly and a plant shutdown is prudent.

The maximum rod drop time restriction is consistent with the assumed rod drop time used in the safety analyses.

Measurement with T greater than or equal to 540'F and with both PeHctor coolant pumps operating ensures that the measured drop times will be representative of insertion times experienced during a reactor trip at operating conditions.

The various control rod banks (shutdown banks, control banks A, B, C,

and D.are each to be moved as a bank; that is, with all rods in the bank with'n one step (5/8 inch) of the bank position.

Position indication is provided by two methods:

a digital count of actuation pulses which shows the demand posit'r. of the banks and a linear position indj.gator (LVDT) which indicates the actual rod position.

These are known as the step counters and analog rod position indication, respectively.

Operability of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits.

The 12 step (7.5 inches) per-missible indicated misalignment e..sures that the 15 inch misalignment assumed in the safety analysis is met.

The action statements which permit limited variations from the basic requirements are accompanied bv additional rest ictions which ensure that the original design criteria are met.

Misalignment of a rod requires measurement of peaking factors or a restriction in power; either of these restrictions provide assurance of fuel rod integrity during cont'nued operation.

In Amendment No. ~

40 3.10-9

C

addition, those safety analyses affected by a misaligned rod are reevaluated to confirm that the results remain valid during future operation.

References:

(1)

Technical Supplement Accompanying Application to Increase Power - Section 14 (2)

FSAR, Section 7.3 Amendment No. P3,40 3.10-10

100 I

80 kJ UI 60 O

~e 40 20 0

FIGURE 3. 10-1 CONTROL R00 INSERTION LPllTS VERSUS CORE POMER fOR QOL TllROUGH EOL

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~I 3.10-13 Amendment No.

,40

Table 3.10-1 ACCIDENT ANALYSIS RE UIRING REEVALUATION IN THE EVENT OF AN INOPERABLE CONTROL ROD Rod Insertion Characteristics Rod Misalignment Loss of Reactor Coolant From Small Ruptured Pipes Or From Cracks In Large Pipes Which Actuates The Emergency Core Cooling System Rod Withdrawal At Full Power Major Reactor Coolant System Pipe Ruptures (Loss Of Coolant Accident)

Steam Line Break Rod Ejection 3.10-14 Amendment No. 40

I

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TABLE 4. 1-1 III N I LIUH FREQUI'.HCIES FOR CIIECKS, CAI'I13RAT10NS ANI)

TEST OV INSTRUNENT CIIANNEI,S

(.I<a r>>1<'. I I)user i>>t. i orr Nrrcle<<r I'ower R<<rrge Clreck S

II'<(3)

Ca I ibra te O(1)

Q"<(3)

Test D/W(2) (4)

P(2) (5)

Remarks

1) Ileat. balance calcul;rtion"'":<

2) Sigrral t.o QT; hist<Ie acLion (permissive, rod stop, trips)

3) Upper 6 lower clraml>ers for axial offset.:.
4) Iligh setpoint
(<
109% of rated power)
5) Low setpoint.
(< 25% of rat.ed power) 2.
H<<clear lnt.ermedi at.e S(1)
Itarrge 3.
Nrrclear Source Range S(1) 4.
It<'.<<ctor'oolant Ten<pe r<<Lu re N.A.
P(2)
P(2)
N(1)
1) Once/shift wlren in service
2) Log level; bistable action (pernrissiv<.,
rod stop, trip)
1) Orrce/shifL wlren irr service
2) I)istable <<etio<< (alarm, t.rip)
1) Overtemperature-Delta T
2) Overpower -,1)el ta T WA
<n
<1<
CL EC)
B 6 c+ 0 5.
Reactor CoolanL Vlow 6.
I'ress>>river W<<Ler I eve I 7.
Pressurizer Pressure S
le 8.
4 Kv Volt<<ge f Vl'e<lrl<<tl<'y N.A Ie Reactor Prot.ect.ion circuits only 9.
Analog Ro<l I'ositiorr S(1,2)
1) With sLep co<>>rters
2) I.og arralog roti posit.ions each 4 Irours wire>> ro<l deviation monit.or is out of service 13y mearrs of tire movable in-core detector syst<m.
Hot re<I<rired during lrot, cold, or refueling shutdown but as soon as possible after return to power.
TABI.E 4. 1-1 (CONTlHUI'.I))
Chan>>el Descri>>Lio>>
10.
Rod Position Ba>>k Cou>>L<.rs Check S(1,2)
Ca libra Le TesL Remarks
1) With analog rod posi Lion
2) I.og analog rod positions each 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months when rod deviation monitor is out. of service 11.
Steam Generator l.evel S
12. Chargi>>g I'Iow N.A.
R I:3. Resid>>ai.
II<'.aL Removal H.A.
Pump I'Iow
14. Iloric Acid Ta>>k I,evel D
N.A.
Bubbler tube rodded weekly
15. Refueli>>g Water SLorage H.A.
'I'a>>k I.evel 16.
Volume Co>>Lrol Ta>>k H.A.
I.eve 1 le R
17. Reactor Co>>tai<>>nenl Press>>re R
1) Isolation Valve sig>>a1 18.
Radi'<Lio>> No>>i,tori>>g D
Sys Le<<<
1<3. Iloric Acid Co>>Lrol H.A.
R N.A.
20.
Co>>tai>>menL Drai>>
S>><l!I<
H.A.
I.ev<'
R 2 I.
Va I ve T<!mperat>>re Interlocks R
22.
I'u<<<p-V < lve 1>> Le r lock R.
23. Turbine 'I'rip Set-Poi>>L N.A.
N.A.
NBA.
1) Block trip 24.
Acc<m<>>lalor I.evel and Press>>re
Y.,
~
TABLE 4.1-2 HINIHUH FREQUENCIES FOR EQUIPHENT P%) SAHPLING TESTS Test
~Fee uenc FSAR Section Reference 1.
Reactor Coolant Samples Gross Radioactivity Concentration (beta-gamma) 3 times/weekly and at least every third day (1) (7) 2.
Reactor Coolant Boron Radio-chemical (2)(4)
E Determination (2)
Tritium Concentration Chloride and Fluoride Oxygen Gross Radioiodine Concentration Boron concentration Honthly (6)
Honthly (6)
Meekly (6) 3 times/week and at least every third day 5 times/week and at least every second day except when below 250 F
Meekly (3) (6)
Meekly 3.
Refueling 'Water Storage Tank Mater Sample Boron concentration Meekly 4.
Boric Acid Tank Boron concentration Twice/week 5.
Control Rods Rod drop times of all full length rods After vessel head removal and at least once per 18 months (8) 6.
Full Length Control Rod Hovement of at least 10 steps in any one direction for any rod not fully inserted
,"lonthly 7.
Pressurizer Safety Valves 8.
Hain Steam Safety Valves Set point Set point Each Refueling shutdown Each Refueling shutdown 10
4. 1-8 Amendment No. g,~~
eo
Table 4.1-2 (Continued) 9.
Containment Isolation Tr 3.p Test Functioning Fre uenc Each Refueling Shutdown FSAR Section Reference 10.
Refueling System Interlocks ll.
Service Water System Functioning Functioning Prior to Refueling Operations Each Refueling Shutdown
9. 4.5 9.5.5 12.
13.
~
14.
Accumulator Fire Protection Pump and Power Supply Spray Additive Tank Functioning Monthly NaOH Concent.
Monthly Boron Bi-Monthly Concentration 9.5.5 16.
Diesel Fuel Supply Fuel Inventory 15.
Primary System Leakage Evaluate Daily Daily 8.2.3 17.
Spent Fuel Pit Boron Concentration Monthly 9.5.5 18.
Secondary Coolant Samples Gross activity 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (5) (6) 19.
Circulating Water Flood Protection Equipment Calibrate Each Refueling Shutdown Notes:
(1)
A gross radioactivity analysis shall consist of the quantitative measurement of the total radioactivity of the primary coolant in units pCi/gm.
The total primary coolant activity shall be the sum of the degassed beta-gamma activity and the total of all identified gaseous activities 15 minutes after the primary system is sampled.
Whenever the gross radioactivity concentration exceeds 10% of the limit specified in the Specification 3.1.4.l.a or increases by
4. 1-9 Amendment No..M. 23, 40
10 p Ci/gm from the previous measured level, the sampling frequency shall be increased to a minimum of once/day until a steady activity level is established.
(2)
A radiochemical analysis shall consist of the quantitative measurement of the activity for each radionuclide which is identified in the primary coolant 15 minutes after the primary system is sampled.
The activities for the individual isotopes shall be used in the determination of E.
A radio-chemical analysis and calculation of E and iodine isotopic activity shall be performed if the measured gross activity changes by more than 10 p Ci/gm from the previous measured level.
(3)
In addition to the weekly measurement, the radioiodine concentration shall be determined if the measured gross radioactivity concentration changes by more than 10 p Ci/gm from the previous measured level.
(4)
Iodine isotopic activities shall be weighted to give equivalent I-131 activity.
(5)
An isotopic analysis or DOSE EQUIVArENT I-131 concentration is required at least monthly whenever the gross activity determination indicates iodine concentration greater than 10% of the allowable limit but only once per 6 months when-ever the gross activity determination indicates iodine con-centration below 10% of the allowable limit.
(6)
Not required during a cold or refueling shutdown.
(7)
During a cold or refueling shutdown, primary coolant Gross Radioactivity will be determined weekly.
(8)
Also required for specifically affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop time of those specific rods.
4.1-10 Amendment gn. g, +
40

ML17258A991

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