ML20064D813

Proposed Tech Specs Re Single Recirculation Loop Operation. Safety Evaluation Encl

ML20064D813
Person / Time
Site:	FitzPatrick
Issue date:	12/29/1982
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20064D779	List: ML20064D775 ML20064D808 ML20064D813
References
NUDOCS 8301040757
Download: ML20064D813 (33)

Text

.

, a. .

ATTACHMENT I PROPOSED TECHNICAL SPECIFICATION CHANGES RELATED TO SINGLE RECIRCULATION LOOP OPERATION I

POWER AUTHORITY OF THE STATE OF NEW YORK JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 8301040757 821229 PDR ADOCK 05000333 p PDR

J- .

1.1 FUEL CLADDING INTEGRITY 2.1 FUEL CLADDING INTEGRITY Applicability: . -.

Applicability:

The Safety Limits established to preserve the fuel cladding integrity apply to those The Limiting Safety System Settings apply

. to trip settings of the instruments and variatyles which monitor the fuel thermal devices which are provided to prevent the

"^#* #*

fuel cladding integrity Safety Limits from .

Objective: being exceeded.

. . . . Objective:

• The obj.ective of the Safety Limits is to establish limits below which the integrity . .

of the fuel cladding is preserved. The objective of the Limiting Safety System Settings is to define the level of the process

..^. variables at which automatic protective action is initiated to prevent the fuel cladding A. Reactor Pressure > 785 psig and Core Flow n gri y aey im rm e ng xceeded.

3>10% of Rated .

Specificaitons The existence of a minimum critical power rati A. Trip Settings (MCPR) less than 1.07 during two recirculation loop operation shall constitute violation of . . .

the fuel cladding integrity safety limit, hereafter The limiting safety system trip settings

• ^ "" "E * **

called the Safety Limit. An MCPR Limit of 1.08 shall apply during single-loop operation.

1. Neutron Flux Trip Settings .

A. IRM - The IRM flux scram setting shall be set at f 120/125 of full scale.

Amendment No. [,[,[,p/ 7

JAFNPD 2.1 (cont'd) 1.1 (cont'd) *

b. APRM Flux Scram Trip Setting (Refuel or H. Core Thermal Power Limit (Reactor Pressure Start & Ilot Standby Mode) 1 785 psig)

APRM - The APRM flux scram setting shall be When the reactor pressure is i 785 psig or i IS percent of rated neutron flux, with core flow is less than 10% of rated, the the Reactor Mode Switch in Startup/Ilot Standby core thermal power shall not exceed 25 or Refuel.

percent of rated thermal power.

c. APRM Flux Scram Trip Settings (Run Mode) .

C. Power Transient (1) Flow Referenced Neutron Flux Scram Trip ,-

To ensure that the Safety Limit established Setting ,

in Specification 1.1.A. and 1.1.B is not exceeded, each required scram shall be When the Mode Switch is in the RUN position, initiated by its expected scram signal. the APRM flow referenced flux scram trip The Safety Limit shall be assumed to be setting shall be exceeded when scram is accomplished by .

a means other than the expected scram S I(0.66W + 540 for two-loop operation or:

signal S i (0.66W +54% - 0.6f4W) for single-loop operation.

where:

S = Setting in percent of rated thennal power (2436 MWT)

W = Inop recirculation flow rate in percent of rated (rated loop re-circulation flow rate equals 34.2 x 10 6 lb/hr)

AW = Difference between two-loop and single-loop effective drive flow at the same core flow.

= 0 for two recirculation loop operation.

= for one recirculation loop operation. (to be determined upon implementation of single loop operation)

For no combination of loop recirculation flow rate and core thermal power shall the APRM flux g

scram trip setting be allowed to exceed 117% of Amendment No. rated thermal power.

l L

y

~

JM11PP 1.1 (cont 'd) 2.1 (cont'd)

n. 14 actor Water Ie.: vel (llot or Gold In the event of operation with a traxinnsn i;hiiLlown O):viltloiiF~

~

fraction of limitiry gewer derisity (MFLPD) greater than tJe fractien of ratoil power h1wriviver llae react or is in tie shutdown O'ItP) , tie set tiry slull le stolifial as anslition with Iraailiatal fuel in tim follows:

s cactor vessel, tlns wat et level shall , ,

s u it twa less than (lat. n>tlengosslity to FRP for two-loop S(- (0.66W + 54%)

111 in. (-14 6.'i in. helicatol level) MFLPD , operation or, alove tie tip of tlw act_iv . fuel wlen it is ~

1. "9 seatal in the cor e. S( (0.66W + 54% - 0.666 W) f1FLPD., loop operat10_ 0 where:

FRP = fraction of rated thermal power (2436 MWt )

!!FLPD = maximum fraction of limiting power density where the limiting power density is 13.4 KW/ft for 8X8, 8X8R and P8X8R fuel.

'Ihe ratio of FTIP to MFIM) shall t;e set erpral to 1.0 unless tie actual operating value is less than the design value of 1.0, in which case tie actual operatirn value will le usal.

(2) Fixed Illgh lieutron Flux Scram Trip Settlig When the Itxle Switch is in the 10N position, tle APlel fixal high flux scrian trip settire stall tur Merubient lb. /( g 9 Sf120%Ihser

1.1 (cont'd) JAFt3PP 2.1 (cont'd)

- A.1.d APRM Rod Block Trip Setting The APRM Rod block trip setting shall be:

S $('

O .66W + 42%) for two-loop operation or:

Si (0.66W + 42% - 0.666W) for single-loop operation. .

where:

S = Rod block setting in percent of thermal power (2436 MWT)

W = Loop recirculation flow rate in percent of rated (rated loop recirculation flow rate equals (34.2 x 106 lb/hr) 6W = Difference between two-loop and single-loop effective drive flow at the same core flow.

= 0 for two recirculation loop operation.'

= for one recirculation loop operation.

(to be determined upon implementation of single-loop operation.) ,

In the event of operation with a maximum fraction limiting power density (MFLPD) greater than the fraction of rated power (FRP), the setting shall, be modified as follows:

S -( (0.66W + 42%) MFLPD 1- fr

, FRP S I (0.66W + 42% - 0.66AW) P "9

MFLPD w -

operation.

where:

MFLPD = maximum fraction of limiting power density where the limiting power density is 13.4 KW/ft for 8X8, 8X8R and P8X8R Fuel.

FRP = fraction of rated thermal power (2436 MWT)

! The ratio of FRP to MFLPD shall be set equal to j Amendment No. J4, )0~, 38, f4

, 10 1.0 unless the actual operating value is less than the design value of 1.0, in which case the the actual operating value will be used.

JAFi1PP e

1.1 IIAS ES A. Reactor Ptesnure > 785 psig and Core Flow >

107. o f Psat ed 1.1 FUEL CIADDI!1G IlfrECRITY Cnset of t ransition boiling results in a de-The fuel cladding integrity limit is set such crease in heat transfer from the clad and, -

that no calculated fuel damage would occur as t he re for e , elevated clad temperature and the a result of an abnotwal operational transient. poisibility of clad failure. However, the Hecause fuel damage is not directly observ- existence of critical power, or boiling trans-

, able, a step-back approach is used to establish ition, is not a directly observable parameter a Safet y I.init such that the minimum critical in an operating reactor. Therefore, the mar-power ratio (!!CPR) is no less than 1.07. MCPR > gin to boiling transition is calculated from 1.07 represents a conservative margin relative plant operating parameters such as core power, to the conlit.fons required to maintain fuel core flow, feedwater temperature, and core cladding integrity. The fuel cladding is one power distribution. The margin for each fuel of the physical barriers which separate radio- assembly is characterized by the critical power active materials f ro n the envirens. The in- ratio (CPR) which is the ratio of the bundle tegrity of this cladding barrier is related to power which would produce on: et of transition its relative freedom from perforations or boiling divided by the actual bundle power.

cracking. ^lthough some corrosion or use re-The minimum value of this ratio for any bundle lated cracking may occur during the life of in the core is the minimum critical power ratio the cladding, fission product inigration from (11CPR) . It is assumed that the plant operation this :ource is incrementally cinnulative and is controlled to the nominal protective set-continuously measurable. Fuel cladding, per- points via the instrtanented variables, i.e.,

forations, however, can result from thermal nomal plant operation presented on Figure st resses winich occur frem reactor operation 1.1-1 by the nominal expected flow control significantly above design conditions and the line. The Safety Limit (MCPR of 1.07) has protection system safety settings. While sufficient conservatism to assure that in the fission product migration from cladding per- event of an abnormat operational transient foration is just as measurable as that from initiated from the MCPR operating limits speci-use rela'.ed cracking, the thermally caused fled for the normal cperating conditions in speci-cladding perforatione signal a threshold, be- fication 3.1.1, more than 99.9' of the fuel rods in yond which still greater themal stresses may the coye are expected to avoid boiling transi-cause gross rather than incremental cladding Lion. The MCPR fuel cladding safety limit is increased deterioration. 'l h e re fo re , the fuel cladding by 0.01 for single loop operation as discussed in Safety I.imit is defined with margin to the Reference 2. The margin between MCPR of 1.0 (onset of conditions which would produce onnet of trans- transition boiling) and the Safety Limit is derived ition boiling, (MCPR o f L .0) . These conditions from a detailed statistical analysis considering represent a significant departure from the condition intended by design for plaiuied all of the uncertainties in monitoring the core operating state including uncertainty in the boiling transition "P"f8tiO"-

l 12 correlation as described in Reference 1. The uncertainties employed in deriving the Safety Limit are Amendment tio . f,1[,[,f,4[

JAFNPP 1.1 BASES (Cont'd.)

C. Power Transient

-v Plant safety analyses have shown that die scrams safety limit at 18 in. above the top of the caused by exceeding any safety system setting fuel provides adequate margin. This level ,

will assure that the Safety Limit of 1.1.A or will be continuously nonitered wtwnever tim 1.1.B will not be exceeded. Scram tines are recirculation pmps are not operating.

checked periodically to assure the insertion tines are adequate. 'Ibe thermal power trans- E. References ient resulting when a scram is accmplished other than by the expected scram signal (e.g., 1. Generic Ibload Fuel Application scram from neutron flux following closure of General Electric BWR 1hermal Analysis the main turbine stop valves) does not neces- Basis (GETPAB) Data, Correlation and Design sarily cause fuel damage. Ilowever, for this Application, NEDO 10958 and NEDE 10958.

specification a Safety Limit violation will be NEDE - 240ll-P-A and Appendices assuned when a scram is only accmplished by neans of a backup feature of the plant design. 2. FitzPatrick Nuclear Power, Plant Single-loop

'lhe concept of not approaching a Safety Limit Operation NEDO - 24281, August 1980 provided scram signals are operable is sup-ported by the extensive plant safety analysis. 3. Generic Peload Fuel Application, NEDE-240ll-D. Ibactor Water IcVel (llot or Cold Shutdown Condition)

During periods when the reactor is shut down, consideration must also be given to water level requirments due to the effect of decay heat. If reactor water level should drop below the top of the active fuel during this Line, the ability to cool the core is reduced.

'lhis reduction in core cooling capability could lead to elevated cladding taperatures and clad perforation. 'lhe core will be cooled sufficiently to prevent clad nelting should the water level be reduced to two-thirds the core height. Establishment of the 14 Anendnent !b. ,14

! JAFtIPP BASFS 1

2.1 FUCL CIl\DDIfKi ItMI3RITY

%e abnontal operational transients appli- evaluation with the initial condition of tle ,

cable to operation of the FitzPatrick Unit reactor being at the steady state operating -

have been analyzed throughout the spectrum limit, it is required that the resulting ICPR of plannal operating conditions up to the does not decrease below the Safety Limit thermal power condition of 2535 FHt. %e ICPR at any time during the transient.

analyses were based epon plant operation in accordance with the operating map given in The nest limiting transients have been l Figure 3.7-1 of the FSAR. In addition, 2436 analyzed to determine which result in the is the licensed maxinun power level of Fitz- largest reduction in CRITICAL POWER RATIO.

Patrick, and this represents the maxinun The type of transients evaluated were in-steady-state power which shall not knowingly crease in pressure and power, positive be exceedal. reactivity insertion, and coolant tenper-ature decrease. %e limiting transient The transient analyses performed for yields the largest delta FCPR. bhen added each reload are given in Reference 2. to the Safety Limit, the required operat-Fkxlels and nrxlel conservatism are ing limit FCPR of Specification 3.1.B is also describal in this reference. obtained.

As discussal in Reference 4, the core wide transient analysis for one The evaluation of a given transient begins rceirculation punp operation is with the systen initial parameters shown conservatively bounded by two-loop in the current reload analysis and refer-operation analysis, and the flow- ence 2, that are input to a core dynamic dependent rod block and scram behavior transient couputer program des-setpoint equations are adjusted cribed in references 1 and 3. The output for one-pump operation. of these programs along with the initial FCPR fonn the input for the further analyses Fuel cladding integrity is assured by of the thermally limited bundle with a the operating limit PCPR's for steady single channel transient thermal hydraulic state conditions given in Specification code. The principal result of the evaluation 3.1.B. These operating limit ICPR' is the reduction in fCPR caused by the derived fran the establishal fuel cladd- transient.

ing integrity Safety Limit, and an analysis of abnonnal operational transients. For any abnonnal operating transient analysis Amendment No. M [ 15

JAI 18'P

_ 2.1 ists (oct'd)

c. Iki ferers:es

1. 1.inford, R. II. , "Arolytical ik2tinia of l>lant Trarmient Evaluat ions for tjas Gesutral Electric Ik>illerj Hater thaactor",

t113n-10802, Fel)., 1973.

2. "Gescral Electric itel Agplication" tul0 240ll-P-A (Alptovett revision nimider atplicalile at Lisies Liut reload

fuel azulyses ar e turfornul) .

1. "(pia 1ifIcat ion of Ilui rse-1)iirenslonal o> e Transient nalet for Inilirvj thter Iteactors" 1u2 0-24154, Octolxer, 1978

4. "FitzPatrick tiuclear Power Plant Single-Loop Operation "t1EDO-24281, August 1980.

Ariesablesnt tb. f(, g 20 (tkut gwje is 23)

l 1.1 ( Con t. ' d ) 3pgpp

):114 geratlig I.imit for I:crurent2il C. FCP11 shall le determient daily duriswJ g cle Oore Avesoge Ex[osure teactor [wer geration at ) 25% of ratal tier-inal [mer an! following any change in power imel *1'yle linC to IIC-lGvlD/ t to level or distrilmtica that would cause IGUD /t lefore LIC IIE gerationwith a limitirvj cogitrol rul tuttern as descrlin) in the bases for At IdH trip level set tiivj S = 0.66 W 4 39% Specification 3.3.H.S.

l oxu 1.22 1.23 D. When it is determined that a channel has ex0H 1.22 1.23 fallal in tjie unsafe corylition, the PilxuH 1.22 1.25 other ItPS ohanneI s that monitor the sane variable shall le furctionally At IGH trip level settiss S = 0.66W t 40% testal insiultately imfore the trip systan containire the failure is trippal.

uxu 1.24 1.24 'the trip systun containirvj the unsafe UxHH 1.24 1.24 failure may le placal in the untripped Pux0H 1.24 ,

1.25 etnulition durire lle gerlod in which surveillance testing is teing terfornal At lati trip level settiivj S = 0.66 W 4 41% on the other ItPS clunnels Ux8 1.27 1.27 E. Verification of the limits set forth 8xult 1.27 1.27 in specification 3.1.B. shall be perfornal Puxult 1.27 1.27 as follows:

At la*t t rip level set tiivj S = 0.66 W t 42% 1. 'the average scram tine tn notch

[usition 30 shall bet t', ito Oxun 1.31 1.31 2. ' lie average scram tine to notch Ptixtill 1.31 1.31 gosition 30 is determiini as follows:

n For one recirculation loop operation the MCPR T 4g =: y[ T1 pg limits at rated flow are 0.01 higher than 93 g the comparable two-loop values. where: n = nimier of surveillance testa gerformal to date in the cycle, Ni =

ntmber of active rods neasural in 31 Anenilmentflo,pf

Y X h W K:MM&XGr.127iGCtZMW5C'.MMK4;7'70L~..s. .. . .. - -

E j t i

+

s e

I*

~

W.

'; F r '.

i; g

e I,

I m

I',

.:?

4

!$a.

.~

g

~

bd e

4 r.

Q g

eT r ,4

JAFNPP Fi93.5-4

$ 13 - Peload 1 8D274fl g _

n, oo g :H O N a 43

~

$5 c

12 -

'd IU O

. AO x

4A d4 c$

d^ 11 -

a n ~

0 04 wc d C4 4 '

I s5

< .a 10 -

0 ~~

h[ ~

Ra _

xe -

3o 9- -

• • /,

. /

I

~

l I I I I l 7 -

-c; -

35 '

~

15 20 25 30 40 5 .

10---Planar Average Exposure -(GWD/t) .

Maximum Average Planar Linear lleat Generation Rate (MAPLilGR)'

' Versus Planar Average Exposure

Reference:

NEDO-21662-2 MAPIllGR (As Ammended 2, .-

Fbr single recirculation loop operation fthese . August ~1981) valdes should be aultiplied 0.84.

h 135B -

/

JAFrlPP Fig 3.5-5 c.,

N U 13 - .

o m- Ibload 2 8DRB265L a oa a,

g lC H O N p 4 3 rt 5di 12 -

-, a o -r4 %

. ae K- u5 a a.

c$

5^

N n. -

11 -

0 04 CD (0 (O M M

s8

,c .~

10 -

v en N s il

.a c x0 50 4

9-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 IIeat Generation Rate (MAPLilGR)

' Versus Planar Average Exposure

Reference:

NEDO-21662-2 (As Ammended For single recirculation loop operation,these MAP 1J1GR August 1981) values should be multiplied 0.84.

135c

JAFt3PP Fig 3.5-6

$ 13 - -

in a Reload 2 8DRB283

{ @g g xv cs N

, a u3 '

l w,

a 12 -

1 1

5 .a m

. AO a:

HA b

Sh 5 11 -

k "J op wn om L4 sS

< .a 10 - ,

o a5 oa ee

.a a xe 50 9 -

m I I I i 1 I I I 5 10 15 ,

20 25 30 35 40 Planar Average Exposure (GWD/t)

Maximum Average Planar Linear lleat Generation Rate (MAPLilGR)

' Versus Planar Average Exposure

Reference:

NEDO-21662-2

~

MAPUlGR (As Ammended For single recirculation loop operation,these August 1981) values should be multiplied 0.84.

135d

JAFilPP Fig 3.5--7

$ 13 -

a *'

Peload 3 P8DRB2651, g @g 3 =u o N y 4 'S

" $U 12 -

. C

.a a:

.U AO

=

hA

, m a.

Cf j e< 11 -

\, c. -

oO wa n a; W

$o

<a 10 - ,

r>

Em ou Eo aC l Mo l 20 9- .

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

Maximum Average Planar Linear lleat Generation Rate (MAPLilGRJ Versus Planar Average Exposure

Reference:

NEDO-21662-2 MAPIJ1GR (As Ammended For single mcirculat ion loop operation,these August 1981) values should be multiplied 0.84.

135e

i JAFtWP Fig 3.5-8

$ 13 ~

c U a c- Reload 3 P8DRB283

c. oU g =u Q \

a L 3

"., $U c

12 -

o a a:

. Ao

=

\ W4 t o c. -

A' C$

d^ 11 m .

v oo tp c c a:

k4

$o az a 10 - ,

a Ei o au Eo aa Xo 50 9-I I l l l l l l 5 10 15 20 25 30 35 40 Planar Average Exposure (GWD/t)

Maximum Average Planar Linear lleat Generation Rate (MAPLilGR)

Versus Planar Average Exposure

Reference:

NEDO-21662-2 For single recirculation loop operation,these MAPIllGR (As Ammended August 1981) values should in nultiplied 0.84.

135f

JAFilPP Figure 3.5-9 N 13 -

g y_ Peload 4 P8DRB28411

n. eo n xv o N a W3 ~

n @5 12 -

y 5 a:

. AG x

, ha 0%

cd nE 11 -

a m -

e a4

a. m n o:

W s8

<.8 10 - ,

a ed aW eo

.a c xe so a

9-l I I I I I I i s 10 is 20 25 30 35 40 Planar Average

• Exposure (GWD/t)

Maximum Average Planar Linear lleat Generation Rate (MAPLilGR)

Versus Planar Average Exposure

Reference:

NEDO-21662-2 (As Ammended Ebr single recirculation loop operation,these MAPUlGR August 1981) values should be multiplied 0.84.

135g

JAFt1PP Figure 3.5-10 a

13 -

n "c -

ct ou Reload 4 P8DRB299 y =v 0 N 3 " ?, r.e

$C a

12 _

o" '"4 %

. AO x

kA

• d G. .

• d^

C 1 1 _._

n. -

e ou ci c CZ W

C) C g 10 - ,

a En au Eo wc x c>

c0 9_

4 I I I l l l l 1 5 10 15 20 25 30 35 40 Planar Average Exposure (GWD/t)

Maximum Average Planar Linear lleat Generation Rate (MAPLilGR) i Versus Planar Average Exposure

Reference:

NEDO-21662-2 For single recirculation loop operation, these tjApulGR (As Ammended values should be multiplied 0.84. August 1981) i 135h

JAFNPP 4.6 (cont' d) 3.6 (con t ' d)

F. Structural _Integ r ity F. Structural Integrity The structural integrity of the 1. The nondestructive inspections reactor coolant system shall be listed in Table 4.6.1 shall maintained at the level required be performed as specified. The by the original acceptance standards results obtained from compliance throughout the life of the Plant. with this specification will be evaluated after 5 years and the conclusions of this evaluation will be reviewed with the AEC.

2. An augmented in-service inspection program is required for those high stressed circumferential piping joints in the main steam and feed- -

water lines larger than 4 inches in diameter, where no restraint against pipe whip is provided.

The augmented in-service inspection program shall consist of 100 percent inspection of the welds in place of the 25 percent inspection per '

inspection interval required by Section XI of the 1970 Edition of the ASME Boiler and Pressure Vessel Code.

G. Jet PDmDE G. Jet Pumps Whenever the Reactor is in the startup/ 1. Whenever there is two-loop recirculation l hot standby or run modes, all jet pumps flow with the reactor in the startup/ hot shall be operable. If it is determined standby or run modes, jet pump operability that a jet pump is not operable, the shall be checked daily by verifying that.

Reactor shall be placed in a cold the following conditions do not occur condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. simultaneously:

144 Amendment No.

J AFill'l' 3.6 (con t ' d) 4.6 (cont' d)

a. The two recirculation loops have a l-flow imbalance of 15 percent or more when the pump are operated at the -same speed.

b. The indicated value of core flow rate l varies from the value derived from loop flow measurements by more than 10 percent.

c. The diffuser to lower plenum differ- l ential pressure reading on an individual jet pump varies from the average of all jet pump differential pressures by more

.than 10 percent.

2. Whenever the reactor is in the startup/ hot j

standby or run modes, and there is one loop J recirculation flow, jet pump operability I shall be verified as follows:

a. Baseline readings will be taken and operating characteristics for the following parameters established: .

1. Jet Pump Loop Flow and Recircula-tion Pump Speed for the operating-loop.

2. Individual Jet Pump percent differential pressures for all jet pumps.

b. Initially, and daily thereafter, jet l

pump operability will be verified by assuring that the following do not occur simultaneously:

145 Amendment No.

JAFNPP

,3.6 (cont ' d) ,

4.6 (cont ' d)

1. The ratio of jet pump loop flow to recirculation pump speed for ,

the operating loop does not vary from the initially established value by more than 10 percent.

2. The ratio of individual jet pump percent differential pressure to the loop's average jet pump percent differential pressure does not vary from the initially established value by more than 20 percent.

11 . Jet _Eump Flow Mismatch H. Jet Pump Flow Mismatach

1. When both recirculation pumps are in .

1. Recirculation pump speeds shall be checked -

steady state operation, the speed of and logged at least once/ day.

the faster pump may not exceed 122 percent the speed of the slower pump when core power is 80 percent or more of rated power, or 135 percent the speed of the slower pump when core power is below 80 percent of rated power.

2. Following one-pump operation, the '

discharge valve of the low speed pump may not be opened unless the speed of the faster pump is less than 50 percent of its rated speed.

I 145a l Amendment flo.

3.6 (cont'd) JAFt1PP

3. Operation with a single recirculatien loop is permitted with the designated adjustments for: APRM rod block and scram setpoints (Technical Specifications 2.1.A.I.c, 2.1.A.1.d, and Tables 3.1-1 and

3. 2-3) ; RBM setpoint, Table 3.2-3; MCPR fuel cladding integrity safety limit and operating limits (Tech. Specs. 1.1.A and 3.1.B, rc.spectively) ; and MAPLilGR (Tech.

Spec. 3. 5.11) . .

Amendment rio . 145b

JAFripp j 3.6 (cont'd) 4.6 (cont'd) 1 1.6.1 Shock Suppressors (Snubbers) 4.6.1 Shock Suppressors (Snubber) -

Applicability Applicability

$ Applies to the operational status of the Applies to the periodic testing requirement for i shock suppressors (snubbers). the hydraulic shock suppressors (snubbers),

t Objective Objective .

To assure the capability of the snubbers to: To assure the operability of the snubbers to perform their intended functions.

Prevent unrestrained pipe motion under dynamic loads as might occur during an earthquake or severe transient, and Specification j Allow normal thermal motion during The following surveillance requirements apply j startup and shutdown, to all hydraulic snubbers listed in Table 3.6-1.

t i Specification 1. All hydraulic snubbers whose sent material has been demonstrated by operating

1. During all modes operation except Cold experience, lab testing or analysis to be l Shutdown and Refuel, all snubbers compatible with the operating environment which are required to protect the primary shall be visually inspected. This inspec- -

i coolant system or any other safety related tion shall include, but not necessarily l system or component shall be operable be limited to, inspection of the hydraulic

• l except as noted in 3.6.1r9 through 3.6.1.5 fluid reservoir, fluid connections, and

} below. These safety related snubbers linkage connections to the piping and j are listed in Table 3.6-1. anchor to verify snubber operability I in accordance with the following

! schedule:

Number of Snutbers Found Inoperable During Inspection or During Inspection Interval

) Mext Required 1

j Inspection Interval

' O 18 months + 25%

l 1 12 months i 25%

Amendment tio. g 145 2 6 mcnths i 25%

I'. Shock Suppressors (Snubbe rs) (Cont'd) JAFMPP 1. Shock Suppressors (Snubbers) (cont'd) 3,4 124 days i 25%

5,6,7 62 days i 25%

)8 31 days + 25% -

2. From and after the time that a snubber ,

is determined to bc: inoperable, continued The required inspection interval shall not reactor operation is permissible only be lengthened more than one step at a time.

during the succeeding 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> unless the snubber is sooner made operable or Snubbers may be categorized in two groups, replaced. " accessible" or " inaccessible" based on their accessibility for inspection during

3. If the requirements of 3.6.1.1 and reactor operation. These two groups may 3.6.1.2 cannot be met, an orderly shut- be inspected independently according to down shal.1 be initiated and the reactor the above schedule. f shall be in a cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 2. All hydraulic snubbers whose seal materials are other than ethylene propylene or other

4. If a snubber is determined to be in- material that has not been demonstrated to operable while the reactor is in the be compatible with the operating ' environment, '

shutdown or refuel mode, the snubber shall be visually inspected for operability shall be made operable or replaced prior every 31 days.

to. reactor startup.

3. Once each refueling cycle, a representative

5. Snubbers may be added to safety related systems sample of 10 hydraulic snubbers or piston without prior License Amendment to Table 3.6.1 approximately 10% of the hydraulic snubbers whichever is less, shallbe functionally provided that a revision to Table 3.6.1 is l included with the next License Amendment tested for operability including verificat%on of proper movement, lock up and request.

bleed. For each unit and subsequent unit found inoperable, an additional 10% or ten hydraulic snubbers shall be so tested until no more failures are found or all units have been tested. Snubbers of rated capcity greater than 50,000 lbs. need not be functionally tested.

145d l Amendment No. Ja'

l ATTACHMENT II SAFETY EVALUATION RELATED TO SINGLE RECIRCULATION LOOP OPERATION b

POWER AUTHORITY OF THE STATE OF NEW YORK JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333

Ssction I - Description of the Changes The proposed amendment allows operation of the reactor with one recirculation loop out of service at reduced power. Adjustments are made to the flow biased APRM Scram and Rod Block lines, as well as the flow biased Rod Block Monitor line to account for the change in core flow versus re-circulation flow behavior during single-loop operation. The MCPR fuel clad-ding integrity safety limit is raised by 0.01 during single-loop operation for increased uncertainties in total core flow and TIP reading. A l

reduction factor is applied to each MAPLHGR curve to account for the difference in LOCA analysis of one loop versus two-loop operation.

i The largest contribution to a decrease in MAPLHGR for one-loop operation I is the very short time to onset of boiling transition used, 0.1 sec versus approximately 10 sec in the two-loop analysis. This conservative 1 assumption is a result of the decrease in forced circulation provided by recirculation pump coast down during the early stages of LOCA.

A description of analyses referred to in this section can be found in the General Electric report NEDO-24281 (as amended) entitled, "FitzPatrick Nuclear Power Plant Single-Loop Operation". Copies of this report are included in the submittal to the NRC.

Surveillance testing of the operating recirculation loop will include verification of jet pump operability by monitoring jet pump instruments, and comparing the readings with baseline readings. These baseline readings are taken each time single loop recirculation is initiated.

Section II - Purpose of the Changes The current Technical Specifications for FitzPatrick do not allow plant operation for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with one recirculation loop out of service. However, the capability of operating at reduced power with a single recirculation loop is highly desirable from a plant availability standpoint in the event maintenance on a recirculation pump or other component renders one loop inoperative.

Section III - Impact of the Changes l The changes to the Technical Specifications outlined in Section I of this l evaluation ensure that the consequences of plant transients and accidents analyzed in the FSAR are unaffected during the single-loop operation, and do not alter conclusions reached in the FSAR and SER accident i analysis.

l The reduced power in single-loop operations ensures that pressurization transients are less severe as seen in Figure 3-1 of NEDO-24281. MAPLHGR reduction factors accountfor the decrease in blowdown heat transfer during a LOCA and maintain the fuel within the 2200 F peak clad temperature limit required by Appendix K of 10 CFR 50.

l Section IV - Implementation of the Changes These changes, as proposed, will not impact the ALARA or Fire Protection Programs at JAF. These changes will not impact the environment.

-2 - .

Section V-- Conclusion The incorporation of these changes: a) will not increase the probability nor the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the Safety Analysis Report; b) will not increase the possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis' Report; and c) will not reduce the margin of safety as defined in the basis for any Technical Specification, and d) does not constitute an unreviewed safety question.

Section VI - References (a,) JAF FSAR (b) JAF SER (c) General Electric Company Report NEDO-24281, "FitzPatrick Nuclear Power Plant Single-Loop Operation'.

(d) General Electric Company Service Information Letter (SIL) No.

• 330, " Jet Pump Beam Cracks", June 9, 1980 1

- - _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _}