ML20071K682

Proposed Tech Specs 3.1-1,3.1-2,4.1-1,4.2-1,3.4-1,3.4-2, 3.5-6 to 3.5-10,3.6-1,4.6-1,6.1-1 & 6.2-1 Re Reload 5/Cycle 6

ML20071K682
Person / Time
Site:	FitzPatrick
Issue date:	05/25/1983
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20071K581	List: ML20071K576 ML20071K630 ML20071K682 ML20071K736
References
NUDOCS 8305270364
Download: ML20071K682 (38)

Text

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Attachment I PROPOSED TECHNICAL SPECIFICATIONS REGARDING RELOAD 5/ CYCLE 6 I

JAMES A. FITZPATRICK NUCLEAR POWER PLANT POWER AUTHORITY OF THE STitTE OF NEW YORK Docket No. 50-333 I

B305270364 830525 p DR ADOCM 05000333 PDR

4

, LIST OF FIGURES 1

4 FIGURE TITLE PAGE 3.1-1 Manual Flow Control 47a 3.1-2 Operating Limit MCPR versus 'I" 47b 4.1-1 Graphical Aid in the Selection of an Adequate Interval 48

~

', , Between Tests ,

4.2-1 Test Interval vs. Probability of System Unavailability 87 3.4-l' Sodium Pentaborate Solution Volume-Concentration 110 Req 2irements 4 3.4-2 Saturation Temperatura of Sodiun Pentaborato Solution 111 k

3.5-5 MAILHGR Versus Planar Averaga Exposure 135 d 1 Ralpaa 2, BDRB283 h 3.5-7 MAPLHGR Versus Planar Average Exposure 135 e Eeload 3, P8DRB265L 3.5-8 MAPUIGR Versu:s Planar Tverage Exposure 135 f

, Reload 3, PSDEB28 3 1

.' . 5- 9 '1APUIGR Verrus Planar Average Exposure 135 g i Ralcad 4, P8DRB234L 3.5-10 MAPLIIGR Versus Planar Average Exposure 135h Reloads 4&5, P8DRD299 f l-l 3.6-1 Reactor Vessel Thermal Pressurization Limitations 163 4.6-1 Chloride Stress Corrosion Test Results at 500 F 164

6.1-1 Management Organization Chart 259 6.2-1 Plant Staff Organization 260 l

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l AmendmentNo./,[, ,g4, vii

JM11PP t i

i surveillance tests, checks, calibrations, an) V. Electrically Disar'med Control Itx1 examinations shall be performal within tim s[ecifiul surveillance intervals. %cse intervalu 'Ib disans a m3 drive electrically, the four rey le adjustal + 25 percent. %e interval a53 nrgbenol t.ype plug connectors are raioved fortainite to initnsient anl electric surveillance fron the drive insert and witivirawal

! t>lall never exceal one operating cycle. In cases solenoids reniering the mi incapable'of utere tic clapsal interval has excealed 100 per- wittuiravnl. Wis procedure is equivalent cenL of the speciflul interval, the next survell- to valvityj out (Jie drive and is preferral.

lance interval shall comunce at the en) of the . Electsical disarming does not eliminate original ugccifiul interval. position irxtication.

II. 'llenel Paraneters W. I l lgh Pressure Wter Fire Protection Systan

1. Mininaan critical power ratio (MCPR)-Hatio %e Ili 9h Pressure Wter Fire Protection

of tlat tower in a fuel assmbly which in Sy

. tan ccanists ofI a water source an]

calculatal to cause axxie point in tint fuel pnips; Mv1 distritution systan piping with assaibly to experience boillig transition asociatal post inlicator valves (isolation to the actual assmbly operatire guer aa valves). Such valves include the yard calculatal by application of tJie GEXL hydrant curb valves an1 the first valve i

correlat.lon (Reference 13EDE-10958) . aliead of t!ie water flow alarm device on cach sprinkler or water spray subsyston.

2. Praction of Limitinj Power Density - %e rat.io of the linear heat generation rate X S_tacyJered Test Basis *

(tilGit) existire at a given location to the design IllGR. %e design IllGR is 13.4 IGi/tt. A Staggered Test Basis shall consist of:

a. A test schedule for a systans, sub-

3. Maxinasa Fraction of Limitire Power Density- systans, trains or other designated

' lim Wxinaan Fraction of Limitire Power caiponents obtained by dividing the Density (MPIPD) is the highest value exist- specified test interval into n equal inn in tim core of tin Fraction of Limitity mhintervals.

tbwer Density (FIPD) .

I b. %e testiry of one systan, subsystan,

! 4. Transition Dolliro - Transition boillan arcans train or other designated caiponent

! tiv: Inilinj region between nucleate ant film at the beginning of each subinterval.

i Inillry. Transition boilire is the reJ i on in which both suicleate an] film telling occur Y. Rated Recirculation Flow intermittently with neither type leinj mn-pletely stable. Wat drive flow which produces a core flow of 77.0 x 106 lb/hr.

hiershient ID..A(( g 6

_ _ _. _ ___--__ __ _ .-._._ _ _ _ _. _ _ _ . . _ _ . . . _ _ _ . . . _ . . _ . _ . . _ . _ , _ . . _ _ ~ . . . _ -

i  : .

1 JAFilPP -

j Z. Top of Active Fuel

The Top of Active Fue'1,.' corresponding .;

1 to the top of the enriched fuel column of each fuel' bundle, is located 352.5 inches above vessel j;

4 zero, which is the lowest point in j the inside bottom of the. reactor vessel. (See General Electric drawing ,

No. 919D690BD.)

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Amendment tio. 6a '.

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JAFNPP 1.1 (cont'd) 2.1 (cont'd)

D. Reactor Water Level (flot or Cold) In the event of operation with a Shutdown Conditions) density (MFLPD) greater than the "9 E '

Whenever the reactor is in the shut- fraction of rated power (FRP), the down condition with irradiated fuel setting shall be modified as follows:

in the reactor vessel, the water level shall not be less than that corresponding S $[(0.66 W + 54%) x FRP MFLPD l to 18 inches abovethe Top of Active Fuel when it is seated in the core. " ~

Where:

FRP = fraction.of rated thermal power (2436 MWt)

MFLPD = maximum fraction of limitin9 power density where the limiting power density is 13.4 KW/ft. l The ratio of FRP to'MFLPD shall be set equal tc 1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual operating value will be used.

(2) Fixed liigh Neutron Flux Scram Trip Setting Wlien the Mode Switch is in the-RUN position, the APRM fixed high flux scram trip setting shall be:

S fil20% Power Amendment No. 14 , .3 0', ,4 3', 64' 9

.e.

1 l JAl1&P 1.1 (cont'd) 2.1 (cont'd) l 4 A.l.d. APIN Ibd Block 'frip Setting i

' lim APIM Ibd block trip setting shall ims l 8$0.66'W+42%

Were:

1 I G = led block setting in tercent of Germal power (2436 mt)

W = Icop recirculation flow rate in percent of ratal In the event of operation with a maximun fractio 1

limiting power density (ef!PD) greater than tie i

fraction of ratal power (FRP), the setting stall i be modified as follows:

S(- (0.66 W + 42%)  : FIP u MFIED dere FIP = fraction of rated thermal power (2436 mt)

IfIFD = maxinnan fraction of limiting power density Were the limiting power density is 13.4 IM/ft.

'lte ratio of FIO to MPLPD shall be set eqtml i

Amendment No. I J , )(I, df , p<[, 7[, 10 to 1.0 unless tie actual operating value is less than tie design value of 1.0, in Mitch case tim actual operating value will be used.

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JAFNPP ,

1.1 (cont'd) provided at the beginning of each At 100% . power, this limit is reached with fuel cycle. Because the boiling a maximun fraction of limiting power density -

transition correlation is based (MPLPD) equal to 1.0. In'the event of on a large quantity of full scrtle operation with a MFLPD^ greater than the fraction data there is a very high confidence of rated power (FRP), the'APRM scram and rod that operation of fuel assembly at block settings shall be adjusted as required the Safety Limit would not prodLee in specifications 2.1.A.1.c and 2.1.A.l.d.

boiling transition. Thus, although it is not required to establish t he S. Cece Thermal Power Limit (Reactor' Pressure safety limit, additional margin (785 psiq),-'

exists between the Safety Limit and the actual occurrence of loss At pressures below 785 psig the core of cladding integrity. e)evation pressure 4.56 is greater,.than drop,(0 power,At low Opower)s psi.

flow Howerver, if boiling transition were and flows this pressure differential is to occur, clad perforation would not raintained in the bypass region of the core.

be expected. Cladding tempcratures Since the pressure drop in the bypass region would increase to approximately 1100 F is essentially all elevation head, the core which is below the perforatior, pressure drop at low powers and flows will temperature of the cladding material. always be greater than 4.56 psi,3 Analys es This has been. verified by tests in show that with a flow of 28 x 10 lbs/hr the General Electric Test Reactor bundle flow, bundle pressure drop is nearly (GETR) where fuel similar in design independent of bundle power and has a value to FitzPatrick operated above the of 3.5 psi. Thus, the bundle flow with critical heat flux for a significant a 4.56 psi griving head will be greater period of time (30 minutes) without than 23 x'10 lbs/hr. Full scall ATLAS clad perforation. tect data taken at pressures fram 0 psig.

to 785 psig indicate that the fuel assembly If reac'_or pressure should ever excaed critical power at this flow is approximately 1400 psia during normal power operation 3.35 MWt. With the design peaking factors (the limit of applicability of the this corresponds to a core thermal power of-boiling transition correlation) it more than 50%. ,Thus,'a core thermal power would be assumed that the fuel clad- limit.of 25% for reactor pressures below dungintegrity Safety Limit has been 735 psig is conservative.

violated.

In addition to the boiling transition limit (Safety Limit) operation is constrained to a maximum LHGR of l 13.4 kw /f t.

Amendment No. }I, M , M, M, fM. 13

s JAFrPP 3.1 (CONTINUED)

MCPR Operating Limit for Incremental C. MCPR shall be determined daily during Cycle Core Average Exposure reactor power operation at? 25% of rated thermal power and following any change in power level or distribution At RBM lli-trip DOC to EOC-2GWD/t to EOC-1GWD/t tha t- would cause operation with a limiting level setting EOC-2GWD/t EOC-lGWD/t to F1C controi rod pattern as described in the bases for Specification 3.3.B.S.

S= .66W + 39% 1.21 1.25 1.29 D. When it is determined that a channel has failed in the u.nsafe condition, the S= .66W + 40% 1.22 1.25 1.29 other RPS channels that monitor the same variable shall be functionally S= .66W + 41% 1.24 1.25 1.29 tested immediately before the trip I system containing the failure is tripped.

S= .66W + 42% 1.25 1.25 1.29 The trip system containing the unsafe

1. ailure may be placed in the untripped S= .66W + 43% 1.27 1.27 3.29 candition during the period in which surveillance testing is being performed S= .66W + 44% 1.33 1.33 1.33 on the other RPS channels.

E. Veri fication of the limits set forth in specification 3.1.B shall be performed as follows:

1. The average scram time to notch position 38 shall be: 77 AVE (- t;B

2. The average scram time to notch position 38 is determined as follows:

AVE " / ,

iel i=1 where: n = number of surveillance tests performed to date in the cycle, Ni =

number of active rods measured in Amendment No. ,%'4 31

JMis?P

'2. If rapiironent 4.1.E.1 la not net (i.e. T B-i (T AVE) tien tim Operating I.imit MCPit tie ith surveillance, and Ti =

average scram time to notch valtais(as a function of't') are as given in position 38 of all rods l Figuru 3.1 neasurell in tim ith surveillance Where t = ( tagg 4 , ( TA ~T6 ^

! arsi g= tle average scram tine to notch 3. 'the adjusted analysis nean scrasa position 30 as deflini in speci- time is calculated as follows:

i fication 4.1.E.2,

'l', = tie adjustal analysis nean scraia N tine as defiral in specification 1 4.1.E.3, Ug(sec)=~ A H .65 F j the scram t.ine to notch position -

'l4 = 30 as deflint in specification N l g,y i 1

3.3.C.1

':bte: Sinald tim operating limit MCPR 1 chtalral fmn t.his figure im i less than tie operating limit where 34.= mean of the distribution l FCPR futuvl in Specification 3.1.B.1 for the average scrani for lle agplicable illet trip level insertion time to notch setting tien specification 3.1.B.1 position 38 = 0.723 sec.

shall apply. .

0 = standartl deviation of the distribution for average scram insertion time to

- notch position 38=0.054 sec.

l If anytiac during reactor operation greater tien i 25% of ratal tower it is determined tint the limit- N,= the total ntsober of active l ing value for MCPR is icing exceeded, action shall r da neasured in s[mcifi-I tlen le initiatal within fif teen (15) minutes to. cation 4.3.C.1

! restore operation to within the prescribed limits.

l If the MCPR is not returint to within the preceribed iho rumber of rods to be scram tested limitu within two (2) tours, an onterly reactor and tie, test intervals are given in power ralaction shall be comenced inmediately, specification 4.3.C.

'ite reactor gower shall be reduced to less tlan 25%

of ratal tower within tim next four intra, or until tim HCPit is returrnt to within the proscrital limits, Ebr core ilows oller Llan ratal, the MCPR operat_irvj ,

limit shall be nultiplial by tie agpropriate kg in au slown in figure 3.141.

hie whient 130. JI5'.[ 31a

JAFNPP TABIE 3.T-1 (cont'd)

REACIOR PIUTECTION SYSTD1TSCFAM) T@le.'mTIO1 REQtIIREMENT tbtes of Table 3.1-1 (cont'd)

C. Iligh Flux IIM 9 D. Scram Discharge Voltme fligh Invel when any control ml in a control cell containing fuel is not fully inserted.

E. APRM 15% Power Trip

7. Not required to be operable when primary containnent integrity is not required.

8. Not required to be operable when the teactor pressure vessel head is not tolted to the vessel.

9. We APR1 downscale trip is automatically bypassed when the IBM Instnmentation is operable and not high.

10. An APIN will be considered operable if there are at least 2 IPFM 11 puts per level and at least 11 LPRM inputs of the nonnal ccmplement.

11. See Section 2.1.A.l.

12. This equation will be used in the event of operation with a maximum fraction of, limiting power density (MFLPD) greater than the fraction of rated pcA;er (FRP) .

Where: FRP = Fraction of Rated % ental Pov.er (2436 P49t)

MFLPD = Maxinum Fraction of Lintiting Pwcr Ibnsity wherc the limiting power density is 13.4 KW/ft. l We ratio of FRP to MFLPD shall i;e et equal to 3.0 unless the actual operating value is less than the design value of 1,0, .in which case the actual operating value will be used.

W = Loop Recirculation Flow in percent of rated S = Scram Setting in percent of Initial

13. %e Average Power Range Monitor scram function is varied as a tunction of recirculation flow (W) .

%e trip setting of this function nust be maintair.od in anrrdance with Specification 2.1.A.l.c.

Amendment No. ,49,pI,,6(,g,f9',J2' 43

Figure 3.1-2 Operating Limit MCPR Versus T (defined in Section 3.1.B. 2)

FOR ALL FUEL TYPES 1.40 _ (1,1.40)_,.l.40 (1,1.37)-

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/ -

"'g 1.35 6 @C / -'

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g ( 1,1. 3 2) y /'/ /

1. O Ej 1.30 (0.2,1.204) ./ _

y' ( 0 . 6 6 7,1. F'/ )

e -

Z

$ SOC '2 5

m 1.25 (0,1.25) - 1.25 (0.6,1.236) to EOC-2 1.20 1.20 (0,1.20) l l  !  ;  ;  ;  ;

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0

> . E Amendment No. 47b

This page deleted Amendnent No. 47 c

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Amendment No. 47d

JAFNPP TABLE 3.2-3 (Cont'd)'

INSTRUMENTATION THAT INITIATES CONTROL ROD BLOCKS NOTES FOR TABLE 3.2-3 (Cont'd)

The APRM and RBM rod blocks need not be operable in start-up mode. From and l after the time it is found that the first column cannot be met for one of the two trip systems, this condition may exist for up to seven days provided that during that time the operable system is functionally tested immediately and daily thereafter; if this condition lasts longer than seven days, the system shall be tripped.

From and after the time it is found that the first column.

cannot be met for both trip systems, the systems shall be tripped.

2.

IRM downscale is bypassed when it is on its lowest range.

3.

This function is bypassed when the count is 1 100 cps.

4. One of the four SRM inputs may be bypassed.

5.

This SRM function is bypassed when the IRM range switches are on range 8 or above.

6. The trip is bypassed when the reactor power is i 30%.

7. This function is bypassed when the Mode Switch is placed in Run.

8. S = Rod Block Monitor Setting in percent of initial.

W = Recirculation flow in percent of rated K = Intercept values of 39%, 40%, 41%, 42%, 43% and 44% can be used with appropriate MCPR Limits from Section 3.1.B. l 9.

When the reactor is suberitical and the reactor water temperature is less than 212 F, the control rod block is required to be operable only if any control rod in a control cell containing fuel is not fully inserted.

10.

When the control rod block function associated with scram discharge' instrument volume high water level is not operable when required to be operable, the trip system shall be tripped.

Amendment No. f d, )VI, 7I 73

JAMIPP 3.5 (cont'd) 4.5 (cont'd) corvlition, that psip shall be consideral 2. Ebliowing any period where the IFCI inoperable for purloses satisfyirvj Speci- subsystans or core spray subsystans fications 3.5 A, 3.5.C, arvi 3.5.E. have not teen rcquired to be operable, tie discharge pJping of the inoperable systan shall be vented fmn the high II. Average Planar I.inear IIcat Generation Itate teint prior to the return of tle (APillGit) systan to service.

%e APINGIt for each type of fuel as a 3. menever the IRCI, ICIC, or Core ftuction of average planar exposure shall Spray Systan is lined up to take act exceal tic limiting valte shown in suction frun tle condensate storage l Pigures 3.5-6 th!UFJh 3.5-10.I f anytine tank, tim dischartje piping of the durirvj reactor power operation greater IIPCI, ICIC, and Core Spray shall tlan 25% of ratal power it is detennined te vented fran the high point of that tie limiting value for APIllGR la the systan, aryl water flow observed Leisyj exccatal, action stall then be on a nalthly basis.

initiatul within 15 minutes to restore operation to within the prescribal limits. 4. 'lle Icvel switches located on the '

If the APillGR is not returned to within Cbre Spray arvi MIR Systan discharge the prescrital limits within two (2) hours, piping high points which monitor an onlerly reactor power rattotion shall be these lines to insure they are full comencal inmediately. '1he reactor power stall be functionally tested each stall be relucal to less than 25% of ratal ponth, power within tic next four hours, or until the APIllGil is returnal to within the pre- 11. Average Planar Linear lleat Generation Itate scriiul IJmits. . (APIJIGR)

W e APIllGR for each type of fuel as a l

function of average planar exposure shall te detennined daily during reactor j operation at > 251 rated tiennal power.

hneruhent tio. jff,fpI 123

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emk m

JAIM P -

3.5 (cont'd) 4.5 (cont'd)

I. I.incar lleat Generation Rate (IllGR)

' lim linear leat generation rate (IllGR) of any I. Linear lleat Generation Rate (tJIGR) ml in any fuel assembly at any axial location stall not exceed tie maxirman allowable IllGR of 'Ihe IJIGR shall be checked daily during l ,

13.4 IM/f t. reactor operation at t 25% rated thermal l power.

If anytine during reactor power operation greater tlan 251 of ratal power it is detennined that tim limiting value for IllGR is bolivj exceeded, action stall tien le initiatal within 15 minutes to re -

store operation to within tim prescribed limits.

If tle 111GR is not returned to within tim pre-scrital limits within tuo (2) Intra, an onierly reactor tower reduction shall be comenced inne-diately. '1he reactor power shall le rahx:ed to less Llan 25% of rated power within tie next four lairs, or until the IllGit is returnal to within tim prescrital limits.

uiemiientm.pr.#

124

JAIMPP 3.5 MSES (cont'd) -

rerpsiraients for the aienjency diesel generators. are within the 10 CPR 50 Appendix K limit.

1he limiting value for APUlGR is shown in G. Maintenance of Fillal Discharge Pipe Figure 3.5-6 th m yjh 3.5-10. l If tie dischanje pipiryj of the core spray, UCI, I. Linear lleat Generation Rate (UlGR)

ICIC, aint l110I are not filled, a water hamner '

can develop in this pipirvj when the ptmp(s) are 'this specification assures that the linear startal. 'lu minimize danage to the dischartje heat generation rate in any mi is less than piplayJ arvl to ensure anklal manjin in the operation the design linear heat generation.

of these systans, this technical specification requires the dischanje lines to be filled when- 'Ihe UiGR shall be checked daily during reactor ever the systun is recpairal to be operable. If operation at > 25% rated thennal power to l a disclarge pipe is not filled, the ptmps that aetentune if fuel burnup, or control roa movanent, sulply that line nust in asstsial to be inoperable has caused changes in power distribution. For

, for technical specification purposes. Ilowever, UlGR to'be a limiting value below 25% rated j if a water hamner were to occur, the systan thermal power, the ratio of local U1GR to

! would still gerform its design function. average UlGR would have to be greater than 10

) which is precluded by a considerable margin

11. Average Planar Linear IIcat Generation Rate (APulGR) when a1 playing any permissible control roa pattern.

'this specification assures that the peak cladding tuiperature following the postulatal design basis loss-of-coolant accident will not exceed the limit specifial in 10 CPR 50 Appendix K.

'lic peak cladding taiperature following a postu-lated loss-of-coolant accident is primarily a function of the average leat generation rate of all the rods of a fuel assaibly at any axial I location and is only dependent secondarily on l

tim ml to ml tower distribution within an

{ assoibly. Since expecial local variations in l power distritution within a fuel. assutbly affect tle calculatal peak clad taiperature by less tlan + 20*F relative to tJie peak taiperature for a typical fuel design, the limit on tim average linear heat generation rate is suf-ficient to assure Llat calculated taiperatures AmendmentNo.fd -130-l

This page deleted 1

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l A'nendment No. [, 135a l

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'Ihis page deleted

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Amendment No. 30, 64 135b

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'Ihis page deleted Anundment No. , 135c

JAENPP i

j Figure 3.5-10 13 -

43 Reloads 4 & 5 P80HB299 l NU a

12-d$ -

.5 as c

8,8 11 -

A e'

i e s>

os s e t4

$a

>0 10 -

4 ,j Sn

.b $

xe as o 9_

E i

I I I I I I I I 5 10 15 20 25 30 35 40 Planar Average Exposure (GWD/t)

Maximum Average Planar Linear llent Generation Rate (MAPLilGR)

Versus Planar Average Exposure

Reference:

NEDO-21662-2 (As Ammended August 1981) 4 Amendment No. f/[ 135h

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JAENPP 5.0 IESIG1 ETAlinES B. h reactor core contains 137 crucifonn-shaped control rods 5.1 SI'IE as described in Section 3.4 of the FSAR.

A. h James A. FitzPatrick Naclear Power Plant is located on the PASNy 5.3 RFK* LOR PRESSUIE VISSEIj Ix>rtion of the Nine Mile Point site, agproximately 3,000 f t, east of tim h reactor pressure vessel is as Nine Mile Point Nuclear Station, Unit 1. described in 1*able 4.2-1 and 4.2-2

'lhe INP-JAF site is on lake Ontario of the FSAR. h agplicable design in Osmgo Country, New York, atproxi- codes are described in Section 4.2 mately 7 miles northeast of Oswgo. of the FSAR.

'1he plant is located at coordinates north 4,819, 545.012 m, east 386, 968.945 m, 5.4 00frAltNENP

- on tim Universal Transverse Marcator . .

Syston. A. h principal design parameters.

arul characteristics for the B. W nearest point on the property primary containnent are given in lism frun the reactor building and Table 5.2-1 of- tle FSAR.

any p,oints of potential gaseous effluents, with t]m exception of the B. 'the socorslary containnent is as lake shoreline, is located at tim described in Section 5.3 and the northeast corner of tim property, applicable codes are as described

'this distance is atproximately in Section 12.4 of tle FSAR.

3,200 f t. arvi is tim radius of the exclusion areas as defined in 10 CFR C. Penetrations of the primary con-100.3. tainnent and piping passing through such penetrations are designed in 5.2 IrXnOR accx>rdance with standards set forth in Section 5.2 of the FSAR.

5.5 EUEI, S'IORNE A. The reactor core consists of not

! nore than 560 fuel assenblies. For l the current cycle, tw> fuel types A. h new fuel storage facility design

are present in the core

8x8R and criteria are to maintain a 1(eff dEY l P8x8R. h se fuel types are des- 40.90 and flooded 4 0.95.

cribed in NEDO-240ll. Both 8x8R Qupliance shall be verified prior to and P8x8R fuel types have 62 fuel introduction of any new ftel design rods and 2 water rods. to this facili,ty.

1 Ancrusnentun..pgjg y, yf pt 245 l _ __ _

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o JAFNPP /

5.5 (cont'd) .

B. The spent fuel storage pool is designed to maintain Kefg less than 0.95 under all conditions as described in the Authority's application for spent fuel storage nodification transmitted to the NRC July 26, 1978. In order to assure that the criterion is met, new fuel will be limited to an axial loading of 16.28 cyn U-235/ axial cm or equivalent. (Fbr the present

• fuel design, described in NEDO-240ll, this axial loading is equivalent to an average lattice enrichment of 3.3 w/o U-235.) The ntmber of spent fuel assemblies stored in the spent fuel pool shall not exceed 2244.

5.6 Seismic Design

'Ihe reactor building and all engineered safeguards are designed on a basis of dynamic analysis using acceleration response spectrum curves uhich are normalized to a ground notion of 0.08 g for the Operating Basis Earth-quake and 0.15 g for the Design Basis Earthquake.

Mendment No. Js, yI 246 l

Attachment II PROPOSED TECHNICAL SPECIFICATIONS REGARDING RELOAD 5/ CYCLE 6 JAMES A. FITZPATRICK NUCLEAR POWER PLANT POWER AUTHORITY OF THE STATE OF NEW YORK Docket No. 50-333 l

[

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l I. Description of the Changes ,

1 All changes described herein refer to Appendix A of the j FitzPatrick Technical Specifications. Most of the changes l involve deletions associated with discharged fuel types and i new transient analyses for the forthcoming Reload 5/ Cycle 6 core. This core will consist of 548 P8x8R bundles and 12 8x8R bundles. The core will contain- no 8x8 bundles.

Specific changes are described below.

i

l. In the List of Figures on page vii, references to Figures 3.1-2a, 3.1-2b and 3.1-2c are deleted. These figures, specifying MCPR Operating Limit versus. scram time ratio, j_ are incorporated into a single new Figure 3.1-2 reflecting Cycle 6 transient analyses for all fuel types. References

to Figures 3.5-3, 3.5-4 and 3.5-5, specifying thermal limits

for discharged fuel types, are also deleted.

2. All references to specific fuel types are deleted on pages i

6, 9, 10, 13, 43 and 124. These references occur in sections discussing thermal limits, which will be core-wide limits for Cycle 6.

3. On a new page 6a, a definition for Top of Active Fuel is added to the definitions section, since there are numerous references to this term in the Technical Specifications.

I i 4. In Section 3.1.B on page 31, the Table of MCPR Operating

{ Limits for Incremental Cycle Core Average Exposure

! is revised to reflect transient analyses for the Reload S/

i Cycle 6 core. Again, all references to specific fuel types l are deleted, since the P8x8R limits are core-wide.

Furthermore, MCPR limits at two' additional Rod Block

, Monitor Trip Level Settings (S=. 66W + 43% and S=. 66W + 44%)

l are added to the table.

1 l 5. On pages 31a, 123 and 130, references to deleted figures are removed.

l 6. In section 1.1.D on page 9, the parenthetical reference to indicated water level is removed. Thus, water level remains

! defined in relation to the Top of Active Fuel to avoid confusion with other level indications in the reactor vessel.

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7. In the first line of the notes on page 73, a typographical a error, reading RRM, is corrected to read RBM, for Rod Block Monitor.

4 8. On pages 124 and 130, references to core power are prefaced

{_ with the words " rated thermal," to clarify the references.

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9. In the Notes for Table 3.1-1 on page 43, a' statement is added to note 6D, " Scram Discharge Volume High Level." The note should read, " Scram Discharge Volume High Level when any control rod in a control cell containing fuel is not fully inserted." The latter part of this statement was inadvertently omitted in the retyping and processing of page 43 for Amend-ment No. 69. >

10. A new Figure 3.1-2 on page 47b replaces Figures 3.1-2a b and c, which specify Operating Limit MCPR versus scram time ratio. Since limits for P8x8R fuel can be applied to the entire core, only a single Figure, 3.1-2, is re-quired.

11. In the Notes for Table 3.2-3 on page 73, two intercept values, 43% and 44%, are added to those specified in the definition for K. These values reflect the two additional Rod Block Monitor Trip Level Settings described in No. 4 above.

12. Figures 3.5-3, 3.5-4 and 3.5-5 on pages 135a, b and c, respectively, are deleted. These figures specify Maximum Average Planar Linear Heat Generation Rate versus Planar Average Exposure for those fuel types which are being discharged prior to Cycle 6.

13 Figure 3.5-10, specifying Maximum Average Planar Linear Heat Generation Rate versus' Planar Average Exposure, is relabeled to include fuel added for Reload 5. The fuel added in both Reload 4 and Reload 5 is of type P8DRB299.

14. On page 245,'the reference to discharged 8x8 fuel is deleted.

15. Lastly, in Section 5.5.B on page 246, a_ parenthetical sentence is added to the description of axial loading limits for the spent fuel storage pool. In order to clarify axial loading requirements, the statement specifies the average enrichment (3.3 w/o U-235) that corresponds to the maximum allowed axial loading in the pool.

II. Purpose of the Changes The proposed changes revise the Appendix A Technical Specifi-cations to account for the discharge of old fuel and the addi-tion of new fuel for the Reload 5/ Cycle 6 core. As noted above, most of the changes simply involve deletions of references to specific fuel types due to the discharge of all 8x8 fuel bundles.

Substantive changes are described below.

e

1. MTR Operating Limits

'Ihe Cycle _6 core will consist of 548 P8x8R bundles and 12 8x8R l bundles. Since all remaining 8x8R bundles will be placed on the periphery of the core, where local power levels will never be limiting vis 'a-vis all.other. radial locations, Minimum Critical Power Ratio (MCPR) operating limits for P8x8R fuel are applied cn a core-wide basis. As noted in the supporting General Electric hmmt (Attachment III and Reference 1), both initial MCPR's and MCPR's calculated for postulated transients at all exposures are, for P8x8R fuel, greater than or equal to ttose for 8x8R fuel.

Hence, MCPR operating limits for P8x8R fuel are limiting for Cycle 6 at all exposures.

2. Operating Limit MCPR Versus Scram Time Ratio (t)

. As noted in Section I above, Figure 3.1-2 replaces Figures 3.1-2a, b, and c due to the applicability of P8x8R MCPR limits to the entire core. For Cycle 6, the new figure contains curves for three different exposure ranges (BOC-1 to EOC, EOC-2 to EOC-1, BOC to EOC-2).

3. . Rod Block Monitor Trip Ievel Settings As noted in Section I above, MCPR operating limits were specified at two aMitional Bod Block Monitor (REM) Hi-Trip

, Ievel Settings:

S = .66W + 43% ; and S = .66W + 44%

Where: S = RBM Hi-Trip Ievel Setting, and W = total recirculation loop flow

'Ihe RM is designed to prevent fuel damage from a rod with-drawal er;.or transient by amparing local power levels around a selected rod to average powers in the core. Bod movements are blocked if local power exceeds average power by a preset margin.

'Ihe incorporation of two aMitional RM Hi-Trip Ievel Settings in the Technical Specifications gives operators more latitude for performing rod pulls while maintaining existing safety margins.

At any given time, only one of the allowed Hi-Trip Ievel Settings

is progu===d into the plant's process conputer and the corresponding

!. MCPR operating limit applied.

i

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i' As noted on page 4 of the accompanying General Electric document (Attachment III and Reference 1), the two.

additional RBM Hi-Trip Level Settings, corresponding to rod block readings of 109% and 110%, allow longer rod withdrawals than did the previous maximum setting of 108%. Consequently, to guarantee that the MCPR safety-limit of 1.07 is not transcended for any rod withdrawal error, MCPR operating limits at these settings must be maintained at the higher values shown in the proposed  ;

new table of MCPR Operating Limits for Incremental Cycle i Core Average Exposure in Section 3.1.B.  !

Since transient analyses have been perf.ormed for the two additional RBM Hi-Trip Level Settings and appropriately higher MCPR operating limits established for them, opera-tion under either of the two additional settings would provide the same conservatism and safety margins as would operation under any of the existing settings. ',

4. MAPLHGR Versus Planar Average Exposure Figure 3.5-10, specifying maximum Average Planar Linear Heat Generation Rate versus Planar Average Exposure, applies to P8DRB299 fuel. Since this is the fuel type added to the core in both Reload 4 and Reload 5, the same curve is used to represent both reloads.

5. Spent Fuel Pool Averace Lattice Enrichment The parenthetical statement added to page 246, section 5.5B, is designed to clarify spent fuel pool axial loading requirements by specifying an average enrichment correspond-

, ing to the axial loading limit.

III. Impact of the Changes Because existing safety limits and conservatisms are maintained with the incorporation of the above changes, approval of the proposed amendment will not have negative safety implications.

The two additional Rod-Block Monitor Trip Level Settings will potentially allow more latitude in control rod withdrawal maneuvers while maintaining existing core safety margins.

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IV. Implementation of the Changes Implementation of the changes, as proposed, will not impact the ALARA'or Fire Protection programs at FitzPatrick.

Moreover, the changes will not impact the environment.

V. Conclusion The incorporation of these changes: a) will not increase.

the probability or the consequences of an accident or malfunction of equipment important to safety as evaluated previously in the Safety Analysis Report; b) will not increase the possibility of an accident or malfunction of a type other than that evaluated previously in the Safety Analysis Report; c) will not reduce the margin of safety.

as defined in the basis for any Technical Specification; d) does not constitute an unreviewed safety question, and e) involves no Significant Hazards Considerations,as defined in 10 CFR 50. 92.

VI. References

1. " Supplemental Reload Licensing Submittal for James A.

FitzPatrick Nuclear Power Plant Reload 5," General Electric report Y1003J01A56, Rev. O, March 1983.

2. " General Electric Standard Application for Reactor Fuel (GESTAR) , "NEDE-240ll-P-A-4 , January 1982.

3. " Loss-of-Coolant Accident Analysis Report for James A. FitzPatrick Nuclear Power Plant (Lead Plant) ,"

July 1977, NEDO-21662 (as amended).

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