ML20032F028

Proposed Changes to Tech Specs Re Reload 4/Cycle 5.Safety Evaluation Encl

ML20032F028
Person / Time
Site:	FitzPatrick
Issue date:	11/18/1981
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20032F017	List: ML20032F016 ML20032F022 ML20032F028 ML20033A002 ML20033A009
References
NUDOCS 8111230623
Download: ML20032F028 (42)

Text

f 4

n

't ATTACHMENT I PROPOSED TECHNICAL SPECIFICATION CHANGE RELATED TO RELOAD 4/ CYCLE 5 POWER AUTHORITY OF T E STATE OF NEW YORh JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 DPR-59 l

t i

0111230623 811118 PDR ADOCK 05000333 P

PDR x----

r o

JAFNPP e

LIST OF FIGURES Ficture Title Page 1.1-1 APPM Flow Bias Scram Relationship to Normal Operating 23 Conditions 3.1-1 Manual Flow Control 4~a 3.1-2 Operating Limit FCPR versus t-47b 4.1-1 Graphical Aid in the Selection of an Adequate Interval 48 Between Tests 4.2-1 Test Interval vs. Probability of Syst m Unavailability 87 3.4-1 Sodium Pentaborate Solution Volume-Concentration 110 Requirments 3.4-2 Saturation Teriperature of Sodium Pentaborate Solution 111 1

3.5-3 MAPIEGR Versus Planar Average Exposure 135a Reload 1, 8D274L 3.5-4 MAPIEGR Versus Planar Average Exposure 135b Reload 1, 8D274H 3.5-5 MAPIEGR Versus Planar Average Exposure 135c Reload 2, 8DRB265L 3.5-6 MAPIEGR Versus Planar Average Exposure 135d Reload 2, 8DRB283 3.5-7 MAPIEGR Versus Planar Average Exposure 135e Reload 3, P8DRB265L 3.5-8 MAPIEGR Versus P' mar Average Exposure 135f Reload 3, P8DRB28 3 3.5-9 MAPIEGR Versus Planar Average Exposure 135g Reload 4,P8DRB284L 3.5-10 MAPIEGR Versus Planar Average Exposure 135h Reload 4,P8DRB299 3.6-1 Reactor Vessel 'Ihermal Pressurization Limitations 163 4.6-1 Chloride Stress Corrosion Test Pesults at 500*F 164 6.1-1 Manag ment Organization Chart 259 6.2-1 Plant Staff Organization 260 Anendnunt No. M, j [, [

Y V11

JAENPP surveillance tests, checks, calibrations, and V.

Electrically Disarmed Control Ibd examinations shal.1 te performed within the specified surveillance intervals. Wese intervals To disarm a rod drive electrically, the four-may be adjusted + 25 percent. %e interval as amphenol type plug connectors are removed pertaining to instrunent and electric surveillance frm the drive insert and witMrawal shall never exceed one operating cycle. In cases solenoids rendering the rod incapable of where the elapsed interval ins exceeded 100 per-witMrawal. m is procedure is equivalent cent of the specified interval, the nort surveil-to valving out the drive and is preferred.

lance interval shall cmmence at the end of the Electrical disaming does not eliminate original specified interval.

position indication.

U.

% ermal Parameters W.

liigh Pressure Water Fire Protection Systm 1.

Minimtra critical power ratio (MCPR)-Ratio We Iligh Pressure Water Fire Protect. ion of that power in a fuel assmbly which is System consists of: a water source and calculated to cause sme point in that fuel punps; and distribution systs piping with assmbly to experience boiling transition associated post indicator valves (isolation to the actual assm bly operating power as valves). Such valves include the yard calculated by application of the GEXL hydrant curb valves and the first valve correlation (Reference NEDE-10958),

ahead of the water flow alarm device on each sprinkler or water spray subsystem.

2.

Fraction of Limiting Power Density - We ratio of the linear heat generation rate X.

Staggered Test Basis (DiGR) existing at a given location to the design UiGR. %e design U1GR is 13.4 KW/ft A Staggered Test Basis shall consist of:

for 8x8, 8x8R and P8x8R bundles, a.

A test schedule for a systems, sub-3.

Maxinun Fraction of Limiting Power Density-systems, trains or other designated

%e Maximum Fraction of Limiting Power cmponents obtained by dividing the Density (MFLPD) is the highest value exist-specified test interval into n equal ing in the core of the Fraction of Limiting subintervals.

Power Density (FLPD).

b.

We testing of one s';sts, subsysts, 4.

Transition Boiling - Transition boiling means train or other designated cmponent the boiling region between nucleate and film at the beginning of each subinterval.

boiling. Transition boiling is the region in which both nucleate and film boiling occur intermittently with neither type being cm-pletely stable.

i+nu din:nt No. /

6

JAFNPP l

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

D.

Reactor Water Invel (Hot or Cbld In the event of operation with a maximtra Shutdown Condition) fraction of limiting power density (MPLPD) greater than the fraction of rated power Whenever the reactor is in the shutdown (FRP), the setting shall be modified as condition with irradiated fuel in the follows:

reactor vessel, the water level shall not be less than that corresponding to S < (0.66 W + 54%)

FRP 18 in. (-146.5 in. indicated level)

IFLPD above the top of the active fuel when it is seated in the core.

where:

FRP = fraction of rated thermal power (2436 bMt)

BFLPD = maximtzn fraction of lirrdting power density where the limiting power density is 13.4 KW/ft for 8x8, Sx8R, and P8x8R fuel.

'Ihe ratio of M@ to MFLPD shall be set equal to 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 High Neutron Flux Scram Trip Setting When the Made Switch is in the PUN position, the APRI fixed high flux scram trip setting shall be:

S{120% Power Amendment No.

9

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

A.l.d.

APIN Pod Block Trip Setting

'Ihe APFM Pod block trip setting shall be:

S i 0.66 W + 42%

where:

S = Ibd block setting in percent of thermal power (2436 FWt)

W = Loop recirculation flow rate in percent of rated (rated loop rpirculation flow rate o.quals (34.2 x 10 lb/hr))

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

S { (0.66 W + 42%)

FRP FFIPD,

g where:

FRP = fraction of rated thermal power (2436 FMt)

MFLPD = 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 FRP to WLPD shall be set equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which g, g g

g g7,

JAFrPP 1.1 (cont'd) provided at the beginring of each fuel cycle.

B. (bre %ermal Pcwr Limit (Reactor Pressure Because the boiling transition correlation

<785 psig) is based on a large quantity of full scale i

data there is a very high confidence that At pressures below 785 psig tM core elevation

{

operation of fuel assmbiv at the Safety pressure drop (0 power, O flow) is greater Limit would not produce wiling transition.

than 4.56 psi. At low powers and flows this hus, although it is not required to establish pressure differential is maintained in the i

the safety limit, additional margin exists bypass region of the core. Since the pres-i between the Safety Limit and the actual sure drop in the bypass region is essentially l

decurrence of loss of cladding integrity.

all elevation head, the core pressure drop l

at low powers and flows will always be greater l

Howevur, if boiling transition were to occur, than 4.56 psi. 3 Analyses show that with a l

clad perforation would not be expected. Cladding flow of 28 x 10 lbs/hr bundle flow, bundle tmperatures would increase to approximately pressure drop is nearly independent of bundle 1100'F which is below the perforation temper-power and has a value of 3.5 psi. nus, the ature of the cladding material. Wis has been bundh flow with a 4.56 psi driving head will 3

verified by tests in the General Electric Test be greater than 28 x 10 lbs/hr. Full scale Beactor (gel'R) where fuel similar in design ATU\\S test data taken at pressures frcm 0 to FitzPatrick operated above the critical heat psig to 785 psig indicate that the fuel as-flux for a significant period of time (30 min-scanbly critical power at this flow is approx-utes) without clad perforation.

imately 3.35 MWt. With the design peaking factors this corresponds to a core thermal l

If reactor pressure should ever exceed 1400 psia power of more than 50%. %us, a core thermal during normal power operating (the limit of power limit of 25% for reactor pressures applicability of the boiling transition corre-below 785 psig is conservative.

lation) it would be asstuned that the fuel cladding integrity Safety Limit has been violated.

In addition to the boiling transition limit (Safety Limit) operation is constrained to a maximin UiGR = 13.4 kw/ft for 8x8, 8x8R, and P8x8R fuel. At 100% power, this lirrit is reached with a maxinzn fraction of limiting power density l

(MFIPD) equal to 1.0.

In the event of opera-l tion with a MFLPD greater than the fraction i

of rated power (FRP), the APRM scram and rod block settings shall be adjusted as required in Specifications 2.1.A.l.c and 2.1.A.l.d.

Amendment No.

)[, M

l JAFNPP ansFs l

l 2.1 FUEL CIADDIM IN1EGRI'lY

'Ihe abnormal operational transients appli-

'Ihe nest limiting transients have been cable to operation of the FitzPatrick Unit analyzed to determine which result in the have been analyzed throughout the spectrum largest reduction in CRITICAL PCNER PATIO.

t of planned operating conditions up to the

'1he type of transients evaluated were in-j thermal power condition of 2535 MWt. 'Ihe crease in pressure and power, positive analyses were based upon plant operation in reactivity insertion, and coolant tarper-accordance with the operating map given in ature decrease. 'Ihe limiting transient Figure 3.7-1 of the ESAR. In addition, 2436 yields the largest delta MCPR. When added is the licensed max 2 mum power level of Fitz-to the Safety Limit, the required operat-Patrick, and this represents the maxinum ing limit MCPR of Specification 3.1.B is steady-state power which shall not knowingly obtained.

be exceeded.

'1he evaluation of a given transient begins Fuel clMAing integrity is assured by the with the systen initial parameters shown in operating limit FCPR's for steady state the current reload analysis and re.ierence conditions given in Specification 3.1.B.

2 that are inptt to a core dynamic behavior

'Ihese operating limit MCPR's are derived transient carputer program described in frun the established fuel ciadding integ-references 1 and 3.

'Ihe output of these rity Safety Limit, and an analysis of abnor-programs along with the initial MCPR form mal operational transients. For any abnor-the input for the further analyses of the mal operating transient analysis evaluation thermally limited bundle with a single with the initial condition of the reactor channel transient thermal hydraulic code.

being at the steady state operating limit,

'Ihe principal result of the evaluation is it is required that the resulting FCPR the reduction in MCPR caused by the tran-does not decrease below the Safety Lintit sient.

MCPR at any time durino the transient.

JWendment No. /I 15

- n ~

- -. ~ -

~

. - - - - - _ = _ _ - - _ _

JAFNPP 2.1 BASEF (cont'd)

C.

References 1.

Linford, R.

B., " Analytical fiethods of Plant Transient Evaluations for the l

General Electric Boiling Water Reactor",

N G O-10802, Feb., 1973.

2.

" General Electric Fuel Application" NEDE 240ll-P-A (Approved revision ntmber applicable at time that reload fuel analyses are performed).

3.

" Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors" NEDO-24154, October, 1978 Amendment No. g 20 (Next page is 23)

1.2 and 2.2 BASIE JAFNPP L e reactor xolant pressure boundary ANSI Code permits pressure transients up to integrity is an inportant barrier in the 20 percent over the design pressure (120% x l

prevention of uncontrolled release of 1,150 - 1,380 psig). W e safety limit fission products. It is essential that pressure of 1,375 psig is referenced to the the ii W ;ity of this boundary be pro-lowest elevation of the Ibactor Coolant tected by establislung a pressure limit Systan.

to be observed for all operating condi-tions and whenever there is irradiated fuel in the reactor vessel.

%e pressure safety limit of 1,325 psig W e current reload analysis shows that the l

as measured by the vessel steam space main steam isolation valve closure transient, pressure indicator is equivalent to with flux scram, is the most severe ever;t

1. 1,375 psig at the lowest elevation of sulting directly in a reactor coolant bReactor Coolant Systs. %e 1,375 systs pressure increase. %e reactor psig value is derived frm the dcqign vessel pressure code limit of 1,375 psig, l

pressures of the reactor pressure given in FSAR Section 4.2, is above the l

vessel and reactor coolant systm peak pressure produced by the event above.

l piping. Se respective design pressures

@ us, the pressure safety limit (1,375 psig) are 1250 psig at 575*F for the reactor is well above the peak pressure that can vessel, 1148 psig at 568 F for the re-result frm reasonably expected cwerpressure circulation suctin piping and 1274 psig transients. (See current reload analysis for l

at 575'F for the discharge piping. %e the curve produced by this analysis.) Reactor J

l pressure safety limit was chosen as the pressure is continuous'y indicated in the l

lower of the pressure transients permitted control rom during operation.

by the applical)q design codes: 1965 ASME

?

Boiler and Pressure Vessel Oode, Section A safety limit is applied to the Besidual III for pressure' vessel and 1969 ANSI Heat Rmoval Systs (141RS) when it is operating B31.1 Oode Tur the reactor coolant systs in the shutdown cooling mode. When operatic piping. S e ASME Boiler and Pressure in the shutdown cooling mode, the RHRS is Vessel Code permits pressure transients included in the r W r coolant syst s.

up to 10 percent over design pressure (110% x 1,250 -1,375 psig), and the W e nunerical distribution of safety / relief valve setpoints shown in 2.2.1.B (2 @ 1090 psi, l

2 01105 psi, 7 @ 1140 psi) is justified by anal-l yses described in the General Electric report IIIO-24129-1, Suppl e ent 1, and assures that the structural acceptance criteria set forth in the Mark I containment Short Term Program are satisfied.

AT.raliiu.nt No. g 29

3.1 LIMITING CINDITIONS FOR OPERATIQi 4.1 SURVEILIANCE REQUIREMEITIS 3.1 REACIOR PROPDCTION SYS'IEM 4.1 RFACIOR PRDrfrTIO1 SYS M 1 Applicability:

Applicability:

Aoplies to the instrunentation and associated Ipplies to the surveillance of the instru-devices which initiate the reactor scram.

nentation and associated devices which initiate reactor scram.

Objective:

Objective:

'Ib assure the operability of the Reactor Protection Systm.

'Ib specify the tyIn of frequency of surveil-lance to be applic.1 to the protection l

Specification:

instrumentation.

A.

'Ihe setpoints, minimtzn ntsnber of trip systms, Specification:

mininun ntznber of instrunent channels that nust be operable for each position of the reactor A.

Instrumentation systens shall be node switch shall be as shown on Table 3.1-1.

functionally tested and calibrated as

'Ihe design system response time frm the opening indicated in Tables 4.1-1 and 4.1-2 of the sensor contact to and including the respectively, opening of the trip actuator contacts shall f

not exceed 50 msec.

B.

Maximum Fraction of Limiting Power I

Density (MFIPD)

B.

Mininun Critical Power Ratio (ITPR)

'Ihe MFLPD shall be determined daily during During reactor power operation at rated power reactor power operation at > 25% rated and flow, the MCPR operating limits shall thermal power and the APFM high flux scram not be less than those shown below:

and Ibd Block trip settings adjusted if necessary as required by Specifications 1.

When surveillance requirment 2.1.A.l.c and 2.1.A.l.d, respectively.

4.1.E is met ( %g itb)

Amendment No.

30

3.1 (Cont'd)

JAFNPP

}CPR Operating Limit for Incremental C.

MCPR shall be determined daily during Cycle Core Average Exposure reactor power operation at > 25% of rated tier-l mal power and following any change in power Fuel Type BOC to EOC-lGWD/t to level or distribution that would cause 1GWD /t before EOC DOC operationwith a limiting control rod

{

pattern as described in the bases for At RBM trip level setting S = 0.66 W + 39%

Specification 3.3.B.5.

8x8 1.22 1.23 D.

When it is detennined that a channel has 8x8R 1.22 1.23 failed in the unsafe condition, the P8x8R 1.22 1.25 other RPS channels that monitor the sane variable shall be functionally At RBM trip level setting S = 0.66W + 40%

tested imnediately before the trip systcm containing the failure is tripped.

8x8 1.24 1.24

'Ihe trip systcm <x>ntaining the unsafe 8x8R 1.24 1.24 failure may be placed in the untripped P8x8R 1.24 1.25 condition during tha period in which surveillance testing is being perfonned At RBM trip level setting S = 0.66 W + 41%

on the other RPS channels l

l 8x8 1.27 1.27 E.

Verification of the limits set forth 8x8R 1.27 1.27 in specification 3.1.B.

shall be performed P8x8R 1.27 1.27 as follows:

At RBM trip level setting S = 0.66 W + 42%

1.

The average scram time to notch position 38 shall be: T NS AVE 8x8R 1.31 1.31 2.

The average scram time to notch P8x8R 1.31 1.31 position 38 is determined as follows:

n n

TAVE =

Ni Ti Ni i

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

number of active rods measured in 31

JAFNPP 2.

If requirenent 4.1.E.1 is not met (i.e. Tg the i th surveillance, and pi =

(T AVE) then the Operating Limit EPR verage scram tune to notch values (as a function of t) are as given in p sition, 38 of all rods Figure 3.1-2 measured in the ith surveillance

test, Where T = ( Ug -t )/ (te,7Q B

3.

'Ihe adjusted analysis mean scram ard tg= the average scram time to notch position 38 as defined in speci-time is calculated as follows:

fication 4.1.E.2, T = the adjusted analysis mean scram N

g 1

time as defined in specification U (sec)= A, +1.65 tr 4.1.E.3, g

t = the scram time to notch position

[N 4

i 38 as defined in specification 1,y 3.3.C.1

• Note:

Should the operating limit E PR

)

I obtained fran this figure be less than the operating limit where x = mean of the distribution MCPR found in Specification 3.1.B.1 for the average scram insertion time to notch for the applicable RBM trip level setting then specification 3.1.B.1 position 38 = 0.723 sec.

t shall apply.

0"= stardant deviation of the distribution for average scram insertion time to notch position 38=0.054 sec.

If anytime during reactor operation greater than N,= the total ntrmr of active 25% of rated power it is determined that the limit-rods neasural in specifi-ing value for MCPR is being exceeded, action shall cation 4.3.C.1 then be initiated within fifteen (15) minutes to -

restore operation to within the prescribed limits.

'Ihe ntrnber of rods to be scram tested If the E PR is not returned to within the prescribed and the test intervals are given in limits within two (2' hours, an orderly reactor specification 4.3.C.

power reduction shall be comenced innediately.

'Ihe reactor power shall be reduced to less than 25%

l of rated power within the next four hours, or until the MCPR is returned to within the proscribed limits.

For core flows other than rated, the MCPR operating limit shall be multiplied by the appropriate kg is as shown in figure 3.1.1.

Amendhent No. g 31a

JAFNPP 3.1 BASES (cont'd)

Turbine control valves fast closure initiates a scram based on pressure switches sensing electro-hydraulic control (EHC) system oil pressure. The switches are located between fast closure solenoids and the disc dtmp valves, ard are set relative (500 (, P ( 850 psig) to the normal (EHC) oil pressure of 1,600 psig so that based on the small syst s volume, they can rapidly detect valve closure or loss of hydraulic pressure.

l I

The requirement that the I m's be inserted I

in the core when the AP m's read 2.5 l

indicated on the scale in the start-up and refuel nodes assures that there is i

proper overlap in the neutron nonitoring system functions and thus, that adequate coverage is provided for all ranges of reactor operation B.

The limiting transient which detennines the required steady state MCPR lir. lit l

depends on cycle exposure. The operating l

limit MCPR values as determined frcm the transient analysis in the current reload subnittal for various core exposures are given in Specification 3.1.B.

The EOG performance analysis assumed reactor I

operation will be limited to MCPR = 1.20, as described in NEDO-21662-2,1he Technical Specifications limit operation of the reactor to the more conservative MCPR based on consid-eration of the limiting transient as given in Specification 3.1.B.

Funt:rds.:nt No. f8f 35

JAFNPP

'IABIE 3.1-1 (cont'd)

PEACIOR PPPfirrIOT SYSTEM (SCPAM) DETPINENTATION REDUIRENENT NOITS OF TABIE 3.1-1 (cont'd)

C.

High Flux IBM D.

Scram Discharge Volume High IcVel E.

APPM 15% Power Trip 7.

Not required to be operable when primary containment integrity is not required.

8.

Not required to be operable when the reactor pressure vessel head is not bolted to the vessel.

9.

'Ihe APFM downscale trip is automatically bypassed when the IPM Instrunentation is operable and not high.

10. An APPM will be considered operable if there are at least 2 LPM inputs per level and at least 11 IPPM inputs of the normal carplanent.

11. See Section 2.1.A.1.

12.

'Ihis equation will t a used in the event of operation with a maximum fraction of limiting power density (METED) greater than the fraction of rated pwer (FRP).

where:

FRP

= Fraction of rated thermal power (2436 FMt) l METED

= 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 FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less l

then the design value of 1.0, in which case the actual operating value will be used 6

W

= Ioop Recirculation flow in percent of rated (rated is 34.2 x 10 lbM)

S

= Scran setting in percent of initial

13. '1he Average Power Range Monitor scram function is varied (Figure 1.1-1) as a function of recirculation loop flow (W). 'Ihe trip setting of this function nast be maintained in accordance with Specification 2.1.A.1.c.

Igredient No. 37 43

JADPP l

l Figure 3.1-2 Operating Limiting EPR vs. 't' (sec. 4.1.E. )

l 1

l 1.40 l

}

}

I i

j 1.35 i

x 1

4 U

%e a

c4

.b l

4 1.30 9,9

1. s#g i

.i

1. #s e

p *#

e (0.60, 1.251)

(0.77, 1.251) j 1.25 BB i

ps%

61 i

t0 l'

BE gaC (0.70, 1.242)

B%

g -1 t

8x8 \\ (0.18, 1.210) 1.20 l

j i

i i

i i

i i

i i

i 0

.1

.2

.3

.4

.5

.6

.7

.8

.9 1.0 option B

=0 cption A

=1 Amendment No.

47b

--._ _ -- - ~._. _ _ _

J4FNPP 3.3 (cont'd) 4.3 (cont'd) 2.

'ihe average of the scram insertion 2.

At 16-week intervals, 10 percent of l

times for the three fastest operable the operable control rod drives shall control rods of all groups of four be scram timed above 950 psig.

When-control rods in a two-by-two array ever such scram time measv vants are shall be no greater than-.

made, an evaluation shal'

,. made to l

provide reasonable assu.ance that Control Rod Average Sciom proper control rod drive performance Notch Position Insertion Time is being maintained.

i Observed (Sec)

{

i j

46 0.361 38 0.977 24 2.112 04 3.764

{

l Amendment No. 30, 49 96

JAFNPP 3.3 and 4.3 BASES (cont'd) his syst s backs up the operator who I

withdraws control rods according to rods have been withdrawn (e.g., groups and written sequences. The specified re-A34, it is dm onstrated that the Group ch strictions with one channel out of 1

made for the control drives is enforced. This service conservatively assure that d m onstration is made by performing the hardware '

fuel damage will not occur due tc rod j

furetional test sequence. %e Group 110tch re-withdrawal errors when this condition straints are autmatically rmoved above 20% power.

exists.

l During reactor shutdown, similar surveillance A limiting control rod pattern is a pattern l

checks shall be made with regard to rod group which results in the core being on a themal availability as soon as autmatic initiation of hydraulic limit (i.e., KPR limits as sfrun the RSCB occurs and subsequently at appropriate in specification 3.1. B ). During use of stages of the control rod insertion.

such patterns, it is judged that testing of the RBM Syst m prior to withdrawal of 4.

% e Source Range Monitor (SPM) Systm performs no such rods to assure its operability will autmatic safety systs function; i.e.,

it has no assure that inproper withdraw does not scram function.

It does provide the operator with occur.

It is the responsibility of the a visual indication of neutron level. The con-Reactor Analyst to identify these limit-sequences of reactivity accidents are functions of ing patterns and the designated rods either the initial neutron flux. We requirment of at when the patterns are initially established l

least 3 counts per sec. assures that any transient, or as they develop due to the occurrence should it occur, begins at or above the initial of inoperable control rods in other than value of 10-8 of rated power used in the analyses limiting patterns. Other qualified of transient cold conr"tions. One operable SPM personnel may perform this function.

channel would be adequate to monitor the approach to criticality using hmogeneous patterns of C.

Scram Insertion Times squattered control rod withdrawal. A minimum of l

two operable SPM's are provided as an added conservatism.

We Control Rod System is designated to bring the reactor subcritical at a rate fast 5.

We Fod Block Monitor (RBM) is designed to auto-enough to prevent fuel damage; i.e., to matically prevent fuel damage in the event of prevent the EPR frm becming less than erroneous rod withdrawal frm locations of the Safety Limit. Scram insertion high power density during high power level time test criteria of Section 3.3.C.1 operation. 'Iwo channels are provided, and were used to generate the generic scram one of these may be bypassed from the console reactivity curve shown in NEDE-240ll-P-A.

for maintenance and/or testing. Tripping of

% is generic curve was used in analysis of one of the channels will block erroneous rod non-pressurization transients to determine withdrawal soon enough to prevent fuel damge.

KPR limits. %erefore, the required pro-tection is provided.

[

Isis A sst No.

JAFNPP 3.5 (cont'd) 4.5 (cont'd) condition, that ptztp shall be considered 2.

Following any period where the LPCI inoperable for purposes satisfying Speci-subsystes or core spray subsystms fications 3.5.A, 3.5.C, and 3.5.E.

have not been required to be oper<tle, the discharge piping of the inoperable systm shall be vented frm the high H.

Average Planar Linear Heat Generation Rate point prior to the return of the (APUlGR) systs to service.

'Ihe APUlGR for each type of fuel as a 3.

Whenever the HPCI, ICIC, or Core function of average planar exposure shall Spray Systs is lined up to take not exceed the limiting value shown in suction from the condensate storage l

Figures 3.5.3 through 3.5.10.If anytime tank, the discharge piping of the during reactor power operation greater HPCI, ICIC, and Core Spray shall than 25% of rated power it is determined be vented frm the high point of that the limiting value for APUlGR is the systs, and water flow observed being exceeded, action shall then be on a monthly basis.

initiated within 15 minutes to restore operation to within the prescribed limits.

4.

'Ihe level switches located on the If the APIllGR is not returned to within Core Spray and RIIR Syst m discharge the prescribed limits within two (2) hours, piping high points which monitor an orderly reactor power reduction shall be these lines to insure they are full ccanenced inmediately. 'Ihe reactor power shall be functionally tested each shall be reduced to less than 25% of rated month.

(

power within the next four hours, or until the APIliGR is returned to within the pre-H. Average planar Linear Heat Generation Rate scribed limits.

(APUIGR)

'Ihe APUlGR for each type of fuel as a function of average planar exposure shall be determined daily during reactor operation at > 25% rated thermal power.

7mendment No. g 123 C

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

I.

Linear lieat Generation Rate (U1GR) l

'Ihe linear heat generation rate (UlGR) of any I. Linear lieat Generation Pate (UlGR) rod in. any fuel assernbly at any axial location shall not exceed the maxinun allowable UlGR of

'Ihe UlGR shall be checked daily during l

I 13.4 KW/ft for 8x8, 8x8R and P8x8R bundles.

reactor operation at 1 25% thermal 73er.

l If anytime during reactor power operation greater than 25% of rated power it is determined that the limiting value for UIGR is being exceeded, action shall then be initiated within 15 minutes to re-store operation to within the prescribed limits.

If the IIIGR is not returned to within the pre-scribed limits within two (2) hours, an orderly reactor power reduction shall be comenced ume-diately. 'Ihe reactor power shall be reduced to less than 25% of rated power within the next four leurs, or until the UIGR is returned to within the prescribed limits.

l l

1 l

l l

l l

Amendment No. 38/

l 124 l

l

JAFNPP 3.5 BASES (cont'd) requirments for the mergency diesel generators.

are within the 10 CFR 50 Appendix K limit.

We limiting value for APUIGR is shown in G.

Maintenance of Filled Discharge Pipe Figure 3.5.3 through 3.5-10.

l If the discharge piping of the core spray,12CI, I.

Linear lleat Generation Pate (UlGR) l BCIC, and HPCI are not filled, a water hanmer can develop in this piping when the punp(s) are Wis specification assures that the linear started. To minimize damage to the discharge heat generation rate in any rod is less than piping and to ensure added margin in the operation the design linear heat generation.

of these systems, this technical specification requires the discharge lines tc be filled when-

%e UIGR shall be checked daily during reactor l ever the system is required to be operable.

If operation atJ_25% power to determine if fuel burn-a discharge pipe is not filled, the pumps that up, or control rod movment has caused changes in supply that line must be assumed to t;e inoperable power distribution. For UIGR to be a limit-for technical specification purposes. However, ing value below 25% rated thermal power, if a water hanmer were to occur, the systan the ratio of local UlGR to average UlGR would would still perform its design function.

have to be greater than 10 which is precitxled by a considerable margin when m ploying any H.

Average Planar Linear Heat Generation Pate (APUlGR)

Nrmissible control rod pattern.

mis specification assures that the peak cladding tmperature following the postulated design basis

(

loss-of-coolant accident will not exceed the i

limit specified in 10 CFR 50 Appendix K.

We peak cladding tsperature following a postu-lated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is only dependent secondarily on the rod to rod power distribution within an assembly. Since expected local variations in power distribution within a fuel assm bly affect the calculated peak clad tsperature by less than + 20*F relative to the peak tsperature for a typical fuel design, the limit on the average linear heat generation rate is suf-ficient to assure that calculated tmperatures Amendment No. MI 130

_. i, b

i iL l'

il i

il l1

)

k na lb y

l lan P

o P

i 1

N t

3 F

n 1

A e

J tn i

eg ap s

i h

T

(

41

.oN tnem dnemA fI1llllll ll1l

(This page intentionally blank)

I Amendment No. K g, [,

l

( W s page intentionally blank)

I 13s i

%e m. p, g g i

l i

I 1

JAFNPP Fig 3.5-3 13 -

a

@9 Reload 1, 8D274L

c u N

43

$U 12 -

u5 ecj ll -

"G on

>o 10 -

<.aa S*

.58 N$

9-1 I

I I

I I

I I

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

Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) i Versus Planar Average Exposure

Reference:

NEDO-21662-2 1

(As Ammended August 1981) i 135a Amendment No. g

2d) e1 2d8 6 n9 6 e1 1 m 2 mt

- As O

u 0

Dsg l

4 eau N( A ec

)

R n

H 5

G e

4 3

H r

l 7

L e

2D P

f 8

A e

M R

1

(

da 0

e 3

t o

I l

a

)

e t

R R

/

e D

n r W

o u G

i s

t o

(

5 a p I

2 e r x r

e E u

n s

e e o

G g p

a x

t r

b 0 E a e 53 2

e v I

1 P

e H A P

g N

a r r F

r a a A

4 e

e n J

v n a 5

A i

l 3

5 L P I

1 r 9

a r s i

n a u F

a n s l

a r P

l e

P V 0

I 1

eg arevA 1

5 mu m

i 0

x 3

a M

oN t

3 2

1 0

9 n

1 1

1 1

e m

d uuN3d $ 5o.I Gu5 $ g-*

8$

n e

%oc 4*u $' um

" oEu$< $55E m

A

Ji D E P Fig 3.5-5 l

13 ~

a

@g Ibload 2 8DRB265L

c u N

be 3

@d 12 -

• M Ao La b em 11 -

m~o ou tn e m en La

$o 10 -

<-a S*

Eo

';l5 mo 9-E 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 (MAPLNGR)

Versus Planar Average Exposure

Reference:

NEDO-21662-2 (As Ammended August 1981)

Amendment No. 43 135c

JAFNPP Fig 3.5-6 13 'I a

@g Beload 2 8DRB283 muN k 3r EU 12 c

.a st:

AOu5 m a.

• g 11 c

c. -

W4 mm m su L4

$o 10

<wa BC Eo aa X0 cc 9

lE I

l I

I I

I I

i 5

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

Maximum Average Planar Linear lleat Generation Rate (MAPLIIGR)

Versus Planar Average Exposure

Reference:

NEDO-21662-2 (As Ammended August 1981)

Amendment No.

43 135d

JAFtFP Fig 3.5-7 13 -

a$g Ibload 3 P8DRB265L muu)

EU 12 -

c.s a:

10 u5 n a.

c6 5^

11 -

"' d ou E'E u

s 8 lo-

$2

.S a xe yo 9~

s 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 Heat Generation Rate (MAPLIIGR)

Versus Planar Average Exposure 9

Reference:

NEDO-21662-2 (As Ammended August 1981)

Amendment No. 49 135e

l l

JAFNPP Fig 3.5-8 13 -

a

$U Ibload 3 P8DRB283

c u N

k3 EU 12 -

l c

l

• e4 C$

30 La b mm j# h 11 m~v GU cn m m cu 14

$o 10 -

4g

$2 Eo e4 C X0 mo 9-lc I

I I

I I

I I

1 5

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

Maximum Average Planar Linear Heat Generation Rate (MAPLIIGR)

Versus Planar Average Exposure

Reference:

NEDO-21662-2 (As Ammended August 1981) 135f Amendment No. 49

JAFr@P Figure 3.5-9 13 ~"

Peload 4 P8DRB284!!

ou

c uu)

$5 12 -

u5 11 -

eB tr eo e cc u

10 -

4y B*

28 sa 9-x I

I I

I

]

l l

i 5

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

Maximum Average Planar Linear Heat Generation Rate (MAPLilGR)

Versus Planar Average Exposure

Reference:

NEDO-21662-2 (As Ammended August 1981) 135g Amendment No.

l l

JAFNPP Figure 3.5-10 13 -

f[

Peload 4 P8DRB299 N

u3

$U 12 -

c

.H M

.-1 0 u5" a g 11 -

mg N

ou CP C 4%

uoa

>o 10 -

sC **4a Ee ou Eo

.e4 c XG mo 9

C 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 Heat Generation Rate (MAPLilGR)

Versus Planar Average Exposure

Reference:

NEDO-21662-2 (As Ammended August 1981)

Amendment No.

135h

)

l JAFNPD 3.7 (cont'd) 4.7 (cont'd) 9.

Primary Contairinent Atrosphere 9.

Primary Contairstent Atmosphere Fbnitoring Instruments Monitoring Instrtunents a.

Instrunentation shall be a.

Primary containment atmos-phere shall be continuously functionally tested and monitored for hydrogen and calibrated as specified in Table 4.7-1.

oxygen when the containment integrity is required.

B.

Standby Gas Treatment System B. Standby Gas Treatment Systcm 1.

Standby Gas Treatment System 1.

Except as specified in 3.7.B.2 surveillance shall be performed below, both circuits of the as indicated below:

btardby Gas Treatment Systan shall be operable at all times when

a. At least once per operating secondary containment integrity cycle, it shall be denonstrated is required.

that:

(1.) Pressure drop across the ccenbined high-efficiency and charcoal filters is less than 5.7 in. of water at 6,000 scfm and (2.) Each 39 KW heater shall dissipate greater than 29KW of electric power as calcu-lated by the following ex-pression:

P = h EI Wnere: P = Dissipated Electrical.,

Power; E = Measured line-to-line

,/

__m_

im m u uit No.

,p 181 voltage in volts (P1S); I = Aver-age measured phase current in anperes (141S)

JAFNPP 1

5.0 DESIGN FEATURES B.

'Ihe reactor core contains 137

~

cruciform-shaped control rods 5.1 SITE as described in Section 3.4 of the FSAR.

A. 'Ihe James A. FitzPatrick Nuclear Power Plant is located on the PASNY 5.3 RFNJIOR PRESSURE VESSEL portion of the Nine Mile Point site, approximately 3,000 ft. cast of the

'Ihe teactor pressure vessel is as l

Nine Mi.le Point Nuclear Station, Unit 1.

described in Table 4.2-1 and 4.2-2 The INP-JAF site is on Lake Ontario of the FSAR. The applicable design in Oswego v0untry, New York, approxi-codes are described in Section 4.2 mately 7 miles northeast of Oswego.

of the FSAR.

The plant is located at coordinates north 4,819, 545. 012 m, east 386, 268.'345 m, 5.4 COPEAltNENT 4

on the Universal Transverse Mercator A.

The principal design parameters System.

and characteristics for the B. The nearest point on the propecy primary containment are given in line fran the reactor building and Table 5.2-1 of the FSAR.

any points of potential gaseous effluents, with the exception of the B.

The secondary containment is as lake shoreline, is located at the described in Section 5.3 and the northeast corner of the property.

applicable codes are as described This distance is approximately in Section 12.4 of the FSAR.

3,200 ft. and is the radius of the i

exclusion areas as defined in 10 CPR C.

Penetrations of the primary con-100.3.

tairrnent and piping passing through such penetrations are designed in 5.2 REACIOR accordance with standards set forth in Section 5.2 of the FSAR.

A. The reactor core consists of not 5.5 FUEL STORtV;E more than 560 fuel assanblies. For A.

'Ihe new fuel storage facility design l

the current cycle three fuel types are present in the core:

criteria are to maintain a Keff dry 8x8, 8x8R and P8x8R. These fuel 40.90 and flooded < 0.95.

types are described in NEDO-240ll.

Canpliance shall be verified prior to The 8x8 fuel has 63 fuel rods and introduction of any new fuel design 1 water rod, and the 8x8R and to this facility.

P8x8R fuel have 62 fuel rods and '. water i

rods.

Isiealisit No. )$ [ K, [

245

\\

l 1

i i

l-ATTACHMENT II i

t SAFETY EVALUATION RELATED TO RELOAD 4/ CYCLE 5 i

i t

l 1

4 i

l i

1 i

POWER AUTHORITY OF THE STATE OF NEW YORK JAMES A.

FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 DPR-59

'mw-

_ e

-ww-- - -vwwww -

1 l

1 Description of Change Section I A.

Maxiumum Average Planar Linear Heat Generation Rate Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) data for two new fuel types is provided in Figures 3.5-9 and 3.5-10.

Additionally, the MAPLHGR data for all fuel types remaining in the JAF core (Figures 3.5-3 thru 3.5-8) has been extended from a maximum nodal exposure of 30 GWD/short ton to 40 GWD/short ton.

This new MAPLHGR data was developed by General Electric as a revision to the General Electric Report NEDO-21662-2 (Attachment III).

The methods used to generate the new MA?LHGR limits are the same as those used in the original LOCA Analysis which was presented in NEDO-21662-2.

The effect of enhanced fission gas release at high exposures on these extended MAPLHGR limits was discussed in a letter dated May 6, 1981 from R. E. Engel of General Electric to T.

A.

Ippolito.

In this letter, General Electric demonstrates that adequate margin to the 2200 F peak clad temperature (PCT) limit exists when General Electric's evaluation of the fission gas correction factor is applied.

B.

Power Spiking The requirement to check linear heat generation rate (LHGR) as a function of core height has been elminated from Section 3.5.I (p.124) and 4.5.I (p.124), and from the Bases 3.5.I (p.130).

References related to fuel densification (p.131) have been deleted since they are not referred in the body of Appendix A.

i I

The requirement to check LHGR as a function of core height originated with the axially varying power spiking penalty on LHGR which was instituted to account for the effect of fuel densification.

In Amendment No. 49 to the JAF Technical Specifications, the power spiking penalty for 8 x 8 and 8 x 8R fuels was eliminated in accordance with the NRC safety evaluation issued June 9, 1978 to General Electric [ Reference (f)].

Since the Cycle 5 core will consist only of 8 x 8 fuel types, this specification no longer applies.

C.

Control Rod Drive Scram Surveillance Frequency The proposed modification would change the control rod drive scram surveillance frequency from the present 15%

every eight weeks, to what it was prior to the issuance of Amendment No. 30, namely 10% every sixteen weeks.

D.

ODYN Pressurization Transient Analysis Section 3.1.B containing Operating Limit MCFRs has been rewritten; Specification 4.1.E concerning control rod and scram time test surveillance has been added and the Bases, (Sections 2.1 and 4.3.C) have been ammended as a result of the use of the ODYN computer program (in lieu of REDY) to calculate the plant response to pressurization transients.

The amended MCPR values found in Section 3.1.B are contained in the General Electric report, " Supplemental Reload Licensing Submittal for James A.

FitzPatrick I

Nuclear Power Plant - Reload 4," Attachment IV.

In a letter [ Reference (g)) dated November 4, 1981, the Commission informed licensees that the analysis of pressurization transients in reload cores would require the use of the ODYN computer program.

A safety evaluation

i issued under a letter [ Reference (h)] dated Janaury 29, 1981 detailed implementation procedures of the j

ODYN code.

The changes made to Section 3.1.B are in accordance with the generic statistical adjustment factor approach (generally known as ODYN option "B") given in the supplementary safety evaluation attached to the January 29th letter.

This generic statistical implementation method requires a new surveillance requirement on control rod scram time of the type l

contained in the new Section 4.1.E.

In this test, periodic control rod scram time test data is checked against the statistical scram time test data used in the ODYN calculation; if the surveillance test data shows that the scram time exceeds that used in the analysis an adjustment is made to the operating limit MCPR as described in the revised Section 3.1.B.

The revisions to the Bases in Sections 2.1 and 4.3.C were made to account for the revised ODYN transient methodology.

The previous descriptions were particular to REDY methodology and these revised sections contain a more general description applicable to the methodology used rather than the particular computer program used.

E.

Test of Standby Gas Treatment Heaters Section 4.7.B.l.a.2 (p. 181) is revised to require that the standby gas treatment heaters dissipate a minimum of 29 KW of electrical power.

This power level will insure a maximum heater outlet relative humidity of 70 percent at 6000 scfm.

Therefore, the existing and proposed surveillance requirements are equivalent forms of the same physical condition.

Section II - Purpose of Change A.

Maximum Average Planar Heat Generation Rate The MAPLHGR curves are being extended in anticipation of exceeding the present 30 GWD/short ton limit in Cycle 5.

Based on current predictions of Cycle 5 energy content, 4

the maximum cumulative nodal exposure at end of Cycle 5 1

will be approximately 36 GWD/short ton.

}

B.

Power Spiking 4

l The purpose of this change is to eliminate a specification-1 j

which no longer applies.

l C.

Control Rod Drive Scram Surveillance Testing i

i The increased surveillance now in effect was initiated by l

Amendment No. 30 to the Technical Specifications issued on September 16, 1977 and was intended to protect the drive mechanisms from possible accumulation of corrosion product particulates from the carbon steel system piping.

However, a General Electric Company report entitled, j

" Evaluation of the Effect of Corrosion Particles on Control Rod Drive Operation at the James A. FitzPatrick Nuclear Power Plant" [ Reference (c)] indicates that the presence of corrosion particles does not affect the reliability of the scram function of the control rod drive system, including operation without the use of CRD exhaust header filter.

The requested modification will improve

{

CRD system reliability by removing the additional duty on j

the drive imposed by the present Technical Specification 4

i L

l surveillance requirement, and will increase the capacity factor, since such testing causes reduction in reactor i

power.

D.

ODYN Pressurization Transient Analysis These changes are required to reflect the use of the ODYN computer program for pressurization transient analysis.

The new surveillance requirement on control rod scram times allows the use of ODYN option "B" MCPR limits which are less restrictive than the deterministic optien "A" limits.

E.

Test of Standby Gas Treatment System Heaters The Authority originally noted ifi OR 78-16 (and UER 78-10) that the Standby Gas Treatment System heater controls did not operate properly due to a failed transformer.

The NRC reported, in Inspection Report No. 78-08, that the Authority intended to either replace the failed components or modify the heater control circuits in such a manner that the heater will operate any time the Standby Gas Treatment System operates.

This is acceptable since the humidity controls and heater are intended to maintain humidity at the charcoal filters below 70% relative humidity.

Full time operation of the heater will maintain relative humidity belou 70%.

In support of the proposed modification and proposed Technical Specification amendment, calculations have been performed which demonstrate that dicsipation of 29 KW or more will maintain the relative humidity at the charcoal filter inlet below 70%.

This modification will further restore the systen, to proper operation condition.

i 1

Section III - Impact of Change 1

A.

Maximum Average Planar Heat Generation Rate The extended MAPLHGR limits are an amendment to General Electric report NEDO-21662-2, " Loss-of-Coolant Accident Analysis Report for James A. FitzPatrick Nuclear Power Plant (Lead Plant)" which was submitted to the Commission in a letter from George T. Berry to Robert W. Reid, Chief of Operating Reactor Branch, on July 29, 1977 [ Reference (d) }.

This-report forms the basis of the ECCS analysis for the FitzPatrick plant and SER issued by the Commission on September 16, 1977.

Thus the proposed modifications will not alter the conclusions reached in the FSAR accident analysis and SER since they are based on the same plant response calculated in NEDO-21662-2 report and have been calculated using the same approach as the previous MAPLHGR.

B.

Power Spiking Since these specifications no longer apply to the fuel types present in the cycle 5 core, the elimination of this specification is necessary to avoid confusion over its applicatility.

C.

Control Rod Drive Scram Surveillance Frequency The General Electric report [ Reference (c)] provides technical justification for the proposed change to the Technical Specifications.

Since the presence of corrosion particles does not affect the CRD scram function l

reliability, and since a decrease in the duty requirement upon the drives will improve their reliability, the requested change will improve the reliability of the system.

il

. - - _ _ - - ~, _ ~. - - - _,,

,------,-.,,,-,n-,---

D.

ODYN Pressurization Transient Analysis Since the methodology employed is consistent eith that accepted in the referenced safety evaluation [ Reference (h)], there is no degradation in the margin to the safety limit MCPR.

4 i

E.

Standby Gas Trreatment Heaters Surveillance Requirements The incorporation of this alternate method for ensuring the relative humidity at the Standby Gas Treatment charcoal filters, and the completion of the associated modifications will close NRC Office of Inspection and Enforcement, open item No. 81-14-02.

Section IV - Implementation of the Modification i

The modifications as proposed will not impact the ALARA or Fire I

Protection Program at James A.

FitzPatrick Nuclear Power Plant.

1 Section V - Conclusion The incorporation of these modifications:

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.

l

.. ~.

_. = - -. - -.. -

Section VI - References (a)

JAF FSAR l

(b)

JAF SER (c)

General Electric Company Report, " Evaluation of the Effect of Corrosion Particles on Control Rod Drive Operation at i

f the James A.

FitzPatrick Nuclear Power Plant."

i (d)

General Electric Company Report, " Loss-of-Coolant Accident Analysis Report for James A. FitzPatrick Nuclear Power Plant (Lead Plant)," NEDO-2166 2-2, July 1977.

1 1

i l

(e)

Letter R. E. Engel to T. A.

Ippolito, Extension of i

Emergency Core Cooling System Performance Limits, May 6, f

1981.

(f)

Letter D. G. Eisenhut to R. L. Gridley, June 9, 1978

{

(contains Commission SER on Power Spiking and i

Densification in 8 x 8 Fuels).

(g)

Letter D. G. Eisenhut to BWR Licensees, November 4, 1980 (ODYN Implementation).

l (h)

Letter D. G. Eisenhut to BWR Licensee, January 29, 1981 (ODYN SER and Supplementary SER).

m._.._.,

._.____.___.m-_._--.

_.J