ML20133K934

Proposed Tech Spec Changes,Allowing Reactor Operation W/ Single Recirculation Loop in Svc

ML20133K934
Person / Time
Site:	FitzPatrick
Issue date:	10/11/1985
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20133K922	List: ML20133K920 ML20133K927 ML20133K934
References
JPN-85-74, NUDOCS 8510220352
Download: ML20133K934 (48)

Text

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i ATTACHMENT I TO JPN-85-74

{

PROPOSED TECHNICAL SPECIFICATION CHANGES RELATED TO SINGLE RECIRC'JLATION ~

LOOP OPERATION

'. (JPTS-85-007)

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NEN YORK POWER AU'.TORITY JAMES A. FITZPATRICK NUCLEAlf POWER PLhNT DOCKET NO. 50-333 DPR-59 3

JAFNPP e

TABLE OF CONTENTS (cont'd) k M F. Minimum Emergency Core Cooling System F. 122 Availability C. Malatenance of Filled Discharge Pipe C. 122 H. Average Planar Linear Heat Generation H. 123 Rate (APLNGR)

I. Linear Heat Generation Rate (LHCR) I. 124 J. Thermal Hydraulle stability J. 124a SURVEILLANCE LIMITING CONDITIONS FOR OPERATION REQUIREMENTS 3.6 Reactor Coolant System 4.6 A. Thermal Limitations A. 136 B. Pressurization Temperature B. 137 C. Coolant Chemistry C. 139 D. Coolant Leakage D. 141 E. Safety and Safety /mollef Valves E. 142a F. Structural Integrity F. 144 C. Jet Pumps C. 144 H. DELETED l I. Shock Suppressors (Snubbers) I. 145b I 3.7 Containment Systems 4.7 165 A. Primary containment A. 165 B. Standby Cas Treatment System B. 181 C. Secondary Containment C. 184 D. Primary Containment Isolation Valves D. 185 3.8 Miscellaneous Radioactive Material Sources 4.8 214 3.9 Auxiliary Electrical Systems 4.9 215 A. Normal and Reserve AC Power Systems A. 215 B. Emergency AC Power System B. 216

c. Diesel Fuel C. 218 D. Diesel-Generator Operability D. 220 E. Station Batterles E. 221 F. LPCI MOV Independent Power Supplies F. 222a 3.10 Core Alterations 4.10 227 A. Refueling Interlocks A. 227 B. Core Monitoring B. 230 C. Spent Fuel Storage Pool Water Level C. 231 D. Control Rod and Control Rod Drive Mainter.ence D. 231 3.11 Additional Safety Related Plant Capabilities 4.11 237 A. Main Control Room Ventilation A. 237 B. Crescent Area Ventilation B. 239 C. Battery Room Ventilation C. 239 Amendaent No. [ [ 11

e JAFNP'P LIST OF FIGURES rimure Till* P_u*

i 3.1-1 Ranual Flow Control 47a Operating Limit NCPR versus 3.1-2 47b 4.1-1 Craphic Aid in the Selection of an Adequate Interval 48 j Between Tests l

4.2-1 Test Interval vs. Probability of System Unavailability 87 I 3.4-1 Sodium Pentaborate solution of System Volume-Concentration 110 ,

Requirements  !

3.4-2 Saturation Temperature of Sodium Pentaborate Solution 111

3.5-1 Thermal Power and Core Flow Limits of Specifications l

3.5.J.1 and 3.5.J.2 134 I 3.5-6 (Deleted) 135d l 3.5 ' (Deleted) 135e

]

3.5-8 (Deleted) 135f [

l 3.5-9 NAPLHCR Versus Planar Average Esposure Reload 4 j P8DRB284L 135g l 3.5-10 MAPLHGR Versus Planar Average Esposure Reloads 4 & 5, P8DRB299 135h ,

i i 3.5-11 NAPLHGR Versus Planar Average Exposure Reload 6 BP8DRS299 1351 1 3.6-1 Reactor Vessel Thermal Pressurizatloc Limitations 163 l

4.6-1 Chloride Stress Corrosion Test Results at 500'r 164 ,

i 6.1-1 ' Management Organization Chart 259  ;

6.2-1 Plant Staff Organization 260 1

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! Amendment No. [ [ . [ [ ,)/, [ , /

vil l

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, _ . _ , , . _ _ _ , _ _ , _ , _ _____..___,__.__.,_m._,_.,_,.,

_ , . , ____.,,,____m, , _ _ _ _ _ _ . , _ , _ _ _ _ ,

, __,__,,____,,___.__,,____.,v,m,- _

_ , _ -,-. .m. - , . _ _, _ _ , , , .

a o JAFMPP 1.1 FUEL CLADDIWC INTEGRITY 2.1 FUEL CLADDIEC INTEGRITY Applicability: Applicability:

The Safety Limits established to preserve the fuel The Limiting safety Eystem Settings apply to trip cladding integrity apply to those variables which settings of the instruments and devices which are monitor the fuel thermal behavior, provided to prevent the fuel cladding integrity safety Limits from being exceeded.

Objective: Objective:

The objective of the Safety Limits is to establish The objective of the Limiting Safety Systesa settings limits below which the integrity of the fuel cladding is to define the level of the process variablas at is preserved. which automatic protective action is initiated to prevent the fuel cladding integrity Safety Limits from being exceeded.

Specifications: Speelfications:

Reactor Pressure > 785 psia and Core Flow > 10%

A. A. Trio Settinas of Rated The limiting safety system trip settings shall be .

The existence of a minimum critical power ratio as specified below:

, (MCPR) less than 1.07 shall constitute violation of the fuel cladding integrity safety limit. 1. Neutron Fluz Trio Settinas hereafter called the Safety Limit. An MCPR Limit I of 1.08 shall apply during single-loop operation, s. IRM - The IRN flux scram setting shall be set at $120/125 of full scale.

l Amendment No. J 4 g , % f(

7 2

s .

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

A.1.b. APRM Flux Scram Trip Settina (Refuel or Start & Hot Standby Mode)

APE 1 - The APRM flux scram setting shall be 1 15 percent of rated neutron flux with the Reactor Mod; Switch in Startup/ Hot Standby or Refuel.

B. Core Thermal Power Limit (Reactor Pressure $ 735 psig) (1) Flow Eaferenced N.utron Flux Scram Trip Setting When the reactor pressure is < 785 pais or core flow is less than 10% of rated, the core thermal When the Mode Switch is in the RUN power shall not exceed 25 percent of rated position, the AP;Gt flow referenced flus thermal power. scram trip setting shall be:

\

j C. Power Transient 3 10.66 W + 54% ror two loop operation ~

cc:

To ensure that the Safety Limit established in S 1(0.66 W + 54% - 0.666W) for single Specification 1.1.A and 1.1.B is not exceeded, loop cperation

~

each required scram shall be initiated by its where:

expected scram signal. The Safety Limit shall be assumed to be exceeded when scram is accomplished S= Setting in percent of rated

! by a means other than the expected scram signal. thermal power (2436 MVT)

W= kocirculation flow in percent. of rated AW= Difference between two Isop and l single loop effective drive flow at the same core flow. ( AW = 0 for two loop operation. A W for single loop operation is to be determined upon implementation of single loop l operation.)

Amendment No.) (, g , [ f[

g

a .

, JAFNPP 1

2.1 (coat'd)

I For no combination of recirculation flow rate and core thermal Power shall the APEN flux scram trip setting be allowed to exceed 117% of rated thermal i Power.

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5 Amendment No.

Sa

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

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

D. Reactor Water Level (Hot or Cold Shutdown In the event of operation with a maximum fraction conditions) of limiting power density (NFLPD) greater than the fraction of rated power (FRP), the setting Whenever the reactor is in the shutdown condition shall be modified as followst with irradiated fuel in the reactor vessel, the water level shall not be less than that corres- S 1(0.66 W + 54%)(FRP/MFLPD) ponding to 18 inches above the Top of Active Fuel when it is seated in the core, for two loop operation or, S 1(0.66W+54%-0.66aW)(FRP/MFLPD) for single loop operation.

Where:

FRP = fraction of rated thermal power (2436 MWt)

NFLPD = maximum fraction of limiting power density where the limiting power' density is 13.4 KW/ft.

The ratio of FRP 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 Mode Switch is in the RUN position, the APRM fixed high flux scram trip setting shall be:

S <120% Power 1

4 Amendment No. [. . [ [. [

9

O .

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

, A.1.d APRM Rod Block Setting The APRM Pod block trip setting shall be:

S < (0.66 W + 42%)

for two loop operation cr.

. S ,< (0.66 W + 42% - 0.66 a W) for single loop operation, where:

.S = Rod block setting in percent of thermal power (2346 IGdt).

W= Loop recirculation flow rate in percent of rated.

A W. Difference between two loop and single loop effective drive flow at the same t

core flow.

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 1(0.66 W + 42%)(FRP/MFLPD) for two loop operation or, 1

i S< (0.66 W + 42% - 0.66aW)(FRP/MFLPD) for single loop operation.

where:

FRp = fraction of rated thermal power (2436 MWt)

Amendment No. , , [, [ [, [

10

- - - - - - - - - - - - - - . - -n..a,-su _ -. _ _ _ - , m. -. - --_.- u a _.L__- _ _ _ . u_ m .- . - - - _=

_ ik - .4----* ..e. A. u m .. . - i+--

JAFWPP 2.1 (cont'd)

MFLPD = maximum fraction of limiting power I

density where the limiting power density is 13.4 KW/ft.

. The ratio of FRP to NFLPD 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.

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i Amendment No.

10s

j JAFNPP 1

i 1.1 BASES I 1.1 FUEL CLADDING INTEGRITY ,

The fuel cladding integrity limit is set such that no elevated clad temperature and the possibility of 1 calculated fuel damage would occur as a result of an clad failure. However, the existence of critical abnormal operational transient. Because fuel damage power, or boiling transition, is not a directly is not directly observable, a step-back approach is observable parameter in an operating reactor.

used to establish a safety Limit such that the mini- Therefore, the margin to boiling transition is i aun critical power ratio (MCPR) is no less than 1.07. calculated from plant operating parameters such I NCPR > 1.07 represents a conservative margin relative as core power, core flow, feedwater temperature, to the conditions required to maintain fuel cladding and core power distribution. The margin for each

integrity. The fuel cladding is one of the physical fuel assembly is characterized by the critical 1 barriers which separate radioactive materials from power ratio (CPR) which is the ratio of the

! the environs. The integrity of this cladding barrier bundle power which would produce onset of transi-

! is related to its relative freedom from perforations tion boiling divided by the actual bundle power.

{ or cracking. Although some corrosion or use related The minimum value of this ratio for any bundle in j cracking may occur during the life of the cladding, the core is the minimum critical power ratio

{ fission product migration from this source is incre- (MCPR). It is assumed that the plant operation j mentally cumulative and continuously measurable, is controlled to the nominal protective setpoints j Fuel cladding perforations, however, can result from via the instrumented variable, i.e., the oper-thermal stresses which occur from reactor operation ating domain. The current load line limit significantly above design conditions and the protec- analysis contains the current operating domain j tion system safety settings. While fission product map. The Safety Limit (MCPR of 1.07) has

migration from cladding perforation is just as sufficient conservatism to assure that in the i measurable as that from use related cracking, the event of an abnormal operational transient thermally . caused cladding perforations signal a initiated from the MCPR operating conditions in i threshold, beyond which still greater thermal specification 3.1.B. more than 99.9% of the fuel j stresses may cause gross rather than incremontal rods in the core are expected to avoid boiling

cladding deterioration. Therefore, the fuel cladding transition. The MCPR fuel cladding safety limit

! Safety Limit is defined with margin to the conditions is increased by 0.01 for single-loop operation as j which would produce onset of transition boiling, (MCPR discussed in Reference 2. The margin between j of 1.0). These conditions represent a significant NCPR of 1.0 (onset of transition boiling) and the 1 departure from the condition intended by design for Safety Limit is derived from a detailed statisti-

! planned operation. cal analysis considering all of the uncertainties i in monitoring the core operating state including l A. Reactor Pressure > 785 psit and Core Flow > 10% the uncertainty in the boiling transition corre-l of Rated lation as described in Reference 1. The uncer-tanties employed in deriving the Safety Limit are Onset of transition boiling results in a decrease l in heat transfer from the clad and, therefore, Amendment No. f f f f , f )$

J 12 4

JAFNPP 1.1 BASES (Cont'd)

C. Power Transient Plant safety analyses have shown that the scrans Safety Limit at 18 in. above the top of the fuel caused by exceeding any safety system setting will Provides adequate margin. This level will be assure that the Safety Limit of 1.1.A or 1.1.8 continuously monitored whenever the recirculation will not be exceeded. Scram times are checked pumps are not operating.

periodically to assure the insertion times are adequate. The thermal power transient resulting E. References when a scram is accomplished other than by the expected scram signal (e.g., scram from neutron 1. General Electric BWR Thermal Analysis Basis i flux following closure of the main turbine stop (CETAB) Date, Correlation and Design valves) does not necessarily cause fuel damage. Application, NEDO 10958 and NEDE 10958.

However, for this specification a Safety Limit violation will be assumed when a scram is only 2. FitzPatrick Nuclear Power Plant Single-Loop accomplished by means of a backup feature of the Operation. NEDO 24281, August 1980.

plant design. The concept of not approaching a Safety Limit provided scram signals are operable 3. Generic Reload Fuel Application, NEDE 24011 -

is supported by the extensive plant safety P-A and Appendices.

analysis.

D. Reactor Water Level (Hot or Cold Shutdown Condition)

During periods when the reactor is shut down, consideration must also be given to water level requirements due to the effect of decay heat. If reactor water level should drop below the top of the active fuel during this time, the ability to cool the core is reduced. This reduction in core cooling capability could lead to elevated cladding temperatures and clad perforation. The core will be cooled sufficiently to prevent clad

• i melting should the water level be reduced to two-thirds the core height. Establishment of the Amendment No.

14

JAFNPP i

j aaSES ,

1 2.1 FUgL CLADDING INTEGRITY

! The abnormal operational transients applicable to The most limiting transients have been analyzed

, operation of the FitzPatrick Unit have been ana- to determine which result in the largest reduc-lyzed throughout the spectrum of planned operating tion in CRITICAL POWER RATIO. The type of tran-

[ conditions up to the thermal power condition 2535 sients evaluated ween increase in pressure and

IRdt. The analyses were based upon plant operation power, positive reactivity insertion, and coolant

! in accordance with the operating map given in the temperature decrease. The limiting transient current load line limit analysis. In addition, yields the largest delta MCPR. When added to the 4

2436 is the licensed maximum power level of Fitz- Safety Limit, the required operating limit MCPR l Patrick, and this represents the maximum steady- of Specification 3.1.B is obtained.

! state power which shall not knowingly be exceeded.

} The evaluation of a given transient begins with

The transient analyses performed for each reload the system initial parameters shown in the cur-

) are given in Reference 2. Models and model rent reload analysis and Reference 2 that are

conservatism are also described in this input to a core dynamic behavior transient com-

! reference. As discussed in Reference 4, the core puter program described in References 1 and 3.

j wide transient . analysis for one recirculation The output of these programs along with the 4 peep operation is conservatively bounded by initial MCPR form the input for the further

, two-loop operation analysis, and the flow- analyses of the thermally limited bundle with a i dependent rod block and scram setpoint equations single channel transient thermal hydraulic code.

are adjusted for one-pump operation. The principal result of the evaluation is the i reduction in MCPR caused by the transient. +

1 Fuel cladding integrity is assured by the oper- s l sting limit MCPR's for steady state conditions i given in Specification 3.1.B. These operating l limit MCPR's are derived from the established fuel cladding integrity Safety Limit, and an

, analysis of abnormal operational transients. For l any abnormal operating transient analysis evalu-stica with the initial condition of the reactor

! being at the steady state operating limit, it is l required that the resulting MCPR does not decrease '

! below the Safety Limit MCPR at any time during

! the transient.

\

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.pp,y i

amendment j 15  ;

l

JAFNPP 2.1 BASES (Cont'd) .

C. References

1. Linford, R.B., " Analytical Methods of Plant Transient Evaluations for the General Electric Boiling Water Reactor", NEDO-10802, Feb., 1973

2. " General Electric Fuel Application", NEDE 24011-P-A (Approved revision number applicable at time that reload fuel . analyses are per-formed).

3. " Qualification of the One-Dimensional Core "

Transient Model for Boiling Water Reactors",

NEDO-24154. October, 1978.

4. FitzPatrick Nuclear Power Plant Single-Loop I Operation, NEDO-24281. August, 1980.

c.

Amendment No. f8f, p(

20 (Next page is 23)

JAFNPP 3.1 (CONTINUED)

MCPR Operating Limit for incremental C. MCPE shall be determined daily during reactor cycle Coro Average Exposure power operation at L 25% of rated thermal power and following any change in power level or dis-At RBM Hi-trip BOC to EOC-2GWD/t to EOC-1GWD/t tribution that would cause operation with a level setting EOC-2GWD/t EOC-1GWD/t to EOC limiting control rod pattern as described in the bases for Specification 3.3.B.5.

S = .66W + 39% 1.24 1.29 1.31 D. When it is determined that a channel has failed S = .66W + 40% 1.27 1.29 1.31 in the unsafe condition, the other RPS channels that monitor the same variable shall be function-S = .66W + 41% 1.27 1.29 1.31 ally tested 1sumediately before the trip system containing the failure is tripped. The trip S = .66W + 42% 1.29 1.29 1.31 system containing the unsafe failure may be placed in the untripped condition during the S = .66W + 43% 1.30 1.30 1.31 period in which surveillance testing is being performed on the other RPS channels.

S = .66W + 44% 1.34 1.34 1.34 E. Verification of the limits set forth in speci-During - single loop operation, the operating limit fication 3.1.B shall be performed as follows:

~

MCPR shall be increased by 0.01 from that in the table above to reflect the increase in safety limit 1. The averags scram time to notch position 38 MCPR. (See Specification 1.1.A) shall be:

AYEI B

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

n n

= Ni i N1 I AVE / i 1-1 1-1 where: n = number of surveillance tests perfermed to date in the cycle. W1 = number of active rods measured in Amendment No p(, [. [ [

31

JAFNPP TABLE 3.1-1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REOUIREMENT Minimum No. Modes in Which Total Number of Operable Trip Level Function Must be of Instrument Instrument Trip Function Setting Operable Channels Pro- Action Channels vided by Design (1) per Trip Refuel Startup Run for Both Trip System (1) (6) Systems (16) 1 Mode Switch in X X X 1 Mode Switch A Shutdown (4 Sections) 1 Manual Scram I X 2 Instrument A Channels 3 IRM High Flux 1120/125of I I 8 Instrument A full scale Channels 3 IRM Inoperative X X 8 Instrument A Channels 2 APRM Neutron Flux- 115% Power X X 6 Instrument A Startup(15) Channels 2 APRM Flow Referenced Si(0.66W+54%)(FRP/MFLPD) I 6 Instrument A or B Neutron Flux (Not to Channels exceed 117%) (12)(13)

(14)(17) 2 APRK Fixed High 1120% Power I 6 Instrument A or B Neutron Flux (14) Channels 2 APRM Inoperative (10) X X X 6 Instrument A or B Channels AmendmentNo.)#,[,[, [ 41

I JAFNPP l TABLE 3.1-1 (cont'd) l REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT NOTES OF TABLE 3.1-1 (cont'd)

14. The APRN flow biased high neutron flux signal is fed through a time constant circuit of approximately 6 seconds. Tne APRM fixed high neutron flux singal does not incorporate the time constant, but responds directly to instantaneous neutron flux.

15. This Average Power Range Monitor scram function is fixed point and is increased when the reactor mode switch is place in the Run position.

16.* During the proposed Hydrogen Addition Test, the normal background radiation level will increase by approximately a factor of 5 for peak hydrogen concentration. Therefors, prior to performance of the test, the Main Steam Line Radiation Monitor Trip Level Setpoint will ipe raised to <_ three times the increased radiation levels. The test will be conducted at power levels > 80% of normal rated power. During controlled power reduction, the setpoint will be readjusted prior to going below 20% rated power without the setpoint change, control rod withdrawal will be prohibited until the necessary trip setpoint adjustment is made, i

17. This APRM Flow Referenced Scram setting is applicable to two loop operation. For one loop operation this setting becomes S < (0.66W+54%-0.666W)(FRP/MFLPD) where AW = Difference between two-loop and single-loop effective drive flow at the same core flow.

• This specification is in effect only during Operating cycle 7.

] AmendmentNo.g,[,)>d i 43a l

1 l

JAFWPP TABLE 3.2-3 INSTRUMENTATION THAT INITIATES CONTROL ROD BLOCKS Minimum no.

of Operable .._ Total Number of Instrument Instrument Trip Level Setting Instrument Channels Action Channels Per Provided by Design Trip System for Both Channels APRM Upscale (Flow Blased) sjl(0.66W+42%)(FRP/MFLPD)(11) 6 Inst. Channels (1) l2 2 APRM Upscale (Start-up Mode) j$12% 6 Inst. Channels (1) 2 APRM Downscale 2*2.5indicatedon 6 Inst. Channels (1) scale 1 (6) Rod Block Monitor Sj[0.66W+K(8)(12) 2 Inst. Channels (1) i (Flew Biased) 1 (6) Rod Block Monitor 2*2.5indicatedon 2 Inst. Channels (1)

(Downscale) scale 3 IRM Downscale (2) 2"2%offullscale 8 Inst. Channels (1) 3 IRM Detector not in (7) 8 Inst. Channels (1) 3 tart-up Position 3 IRN Upscale j[86.4% of full scale 8 Inst. Channels (1) 2 (4) SRM Detector not in (3) 4 Inst. Channels (1)

Start-up Position 2 (4) (5) SRM Upscale j[105 counts /sec 4 Inst. Channels (1) 1 Scram Discharge Instrument ji26.0 gallons per 2 Inst. Channels (9) (10)

. Volume High. Water Level instrument volume NOTES FOR TABLE 3.2-3

1. For the Start-up and Run positions of the Reactor Mode Selector Switch, there shall be two operab)e or tripped trip systems for each function. The SRM and IRN block need not be operable in run mode, and AmendmentNo.)(,g.[

72

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

JAFNPP TABLE 3.2-3 (Cont'd)

N J_NSTRUMENTATION THAT INITIATES CONTROL ROD BLOCKS NOTES FOR TABLE 3.2-3

11. This is the APRM Rod Block line for two loop operation. For single loop operation this line is S < (0.66W+42LO.66 A W)(FRP/MFLPD) .

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

12. This is the RBN Rod Block line for two loap operation. For single loop operation this line is S < (0.66W+K-0.66aW) where:

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

k Amendment No.

74

l JAFNPP 3.5 (cont'd) 4.5 (cont'd) condition, that pump shall be considered inoper- 2. Following any period . where the LPCI subsys-able for purposes satisfying Specifications tems or core spray subsystems have not been 3.5.A. 3.5.C. and 3.5.E. required to be operable, the discharge piping of. the inoperable system shall be H. Average Planar Linear H_ eat Generation Rate vented from the high point prior to the

,(_APLHCR }, return of the system to service.

The APLHCR for each type of fuel as a function of 3. Whenever the HPCI, RCIC, or Core Spray System average plante exposure shall not exceed the is lined up to take suction from the condon-limiting valuo shown in Figures 3.5-9 through sate storage tank, the discharge piping of 3.5-11 for two locp operation. For single loop the HPCI, RCIC, and Core Spray shall be operation these values are reduced by multiplying vented from the high point of the system, by 0.84 (see Specification 3.5.K. Reference 1). and water flow observed on a monthly basis.

If anytime during reactor power operation greater ,

than 25% of cated power it is determined that the 4. The level switches located on the Core Spray limiting value for APLHCR is being exceeded, and RHR System discharge piptog high points action shall then be initiated within 15 minutes which monitor these lines to insure they are to restore operation to within the prescribed full shall be functionally tested each month.

limits. If the APIMGR is not returned to within the prescribed limits within two (2) hours, an H. Average Planar Linear Heat Generation Rate orderly reacter powsr reduction shall be (APLHCR) commenced irsmediately, The reactor power shall be reduced to less than 25% or rated power within The APLHCR for each type of fuel as a function of the next four hours, or until the APLHCR is average planar exposure shall be determined delly returned to within the prescribed limits. during reactor operation at h 25% rated thermal power, i

Amendment No. d f , pl. /

J,%

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

J. Thermal Hydraulic-Stability J. Thermal Hydraulic Stability

1. Whenever the reactor is in the startup or 1. Establish baseline APRM and LPRM neutron flus run modes, two Reactor Coolant System noise values within 2. hours of entering the recirculation loops shall be in operation, region for which monitoring is required with: unless baselining has been performed since the last refueling outage. Detector levels A

a. Total core flow greater than or and C of one LPRM string per core octant plus equal to 45 percent of rated, or detectors A and C of one LPRM string in the center of the core should be monitored.

b. Thermal power less than or equal to the limit specified in Figure 3.5-1 (Line A).

except as specified in Specifications 3.5.J.2 and 3.5.J.3.

2. With two Reactor Coolant System recircula-tion loops in operation an6 total core flow less than 45 perc( at of tated, and thermal power greater than the limit specified in Figure 3.5-1 (Line A); or with one Reactor Coolant System loop , operating and thermal power greater than the limit specified in Figure 3.5-1 (Line A):

a. Determine the APRM and LPRM noise levels:

1. Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after reaching steady-state within the regions of Figure 3.5-1 where monitoring is required, and at least once per 8 hours '

thereafter; and Amendment No. [

124a

JAFNPP 3.5 (cont'd)

2. W Rhin 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after completing an increase in thermal power of 5 percent or more of.cated thermal power,

b. If the APRM and LPRM neutron flur noise levels are greater than 5 percent and greater than three times their established bcseline noise levels, initiate corrective action within 15 minutes to restore the noise levels to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, by incrossing core flow and/or reducing thermal power.

3. If during single-loop operation, ' core thermal power is greater than the limit defined by line A of Figure 3.5-1, and core flow is less than 39 percent. Immediately initiate corrective action to restore core thermal power and/or core flow to within the limits, specified in Figure 3.5-1, by increasing core flow and/or initiating an orderly reduction of core thermal power by inserting control rods.

4. The requirements applicable to single-loop operation in Specifications 1.1.A. 2.1.A.

3.1.A, 3.1.B. 3.2.C and 3.5.H shall be in effect within 8 hoors following the removal of one recirculation loop from service, or the reactor shall be placed in the hot shutdown condition.

Amendment No.

124b

JAFNPP

.3.5 (cont'd)

5. During resumption of two-loop operation following a period of single-loop operation, the discharge valve of the low-speed pump shall not be opened unless the speed of the faster pump is less than 50 percent of it's rated speed.

6. With no Reactor Coolant System Recirculation loop in service, the reactor shall be placed in Hot Shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

j i

Amendment No.

124c i

. ._.-m---- .. _ _ _ m -._ .e_.- __

_ _ . _ _ - - - - - -- . . f .

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

JAFNPP 3.5 BASES (cont'd) l requirements for the emergency diesel generators, generation rate is sufficient to assure that calculated temperatures are within the .10 CFR 50 C. Malatenance of Filled Discharae Pipe Appendix K limit. The limiting value for APLHCR

, is shown in Figure 3.5-9 through 3.5-11. The

! If the discharge piping of the core spray, LPCI, reduction factor for single loop operation for RCIC, and HPCI are not filled, a water hassner can the above curves is 0.84. The derivation of this develop in this piping when the pump (s) are factor can be found in Specification 3.5.K.

started. To minimize damage to the discharge Reference 1.-

piping and to ensure added margin in the opera-tion of these systems, this technical specifica- I. Linear Heat Generation Rate (LNGR) tion requires the discharge lines to be filled whenever the system is required to be operable. This specification assures that the linear heat If a discharge pipe is not filled, the pumps the generation rate in any rod is less than the supply that line must be assumed to be inoperable design linear heat generation.

for technical specification purposes. However, if a water hamuner were to occur, the system would The LHCR shall be checked daily during reactor still perform its design function. operation at d 25% rated thermal- power to deter-mine if fuel burnup, or control rod movement, has H. Averaae planar Linear Heat Generation Rate caused changes in power distribution. For LHCR (APLNGR) to be a limiting value below 25% rated thermal Power, the ratio of local LHGR to average LNGR This specification assures that the peak cladding would have to be greater than 10 which is pre-temperature following the postuinted design basis cluded by a considerable margin when employing loss-of-coolant accident will not exceed the limit any permissible control rod pattern.

specified in 10 CFR 50 Appendix K.

The peak cladding temperate e following a postulated loss-of-coolant accident is primarily i 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 distri-bution within c. fuel assembly ' affect the calcu-I lated peak clad temperature by less than i 20'F relative to the peak temperature for a typical -

fuel design, the limit on the average linear heat Amendment No. f, ]$,)MI 4

JAFNPP 3.5 BASES (cont'd)

J. Thermal Hydraulle Stability Operation in certain regions of the power vs.

flow curve have been identified as having a high potential. for thermal hydeaulic instability (Figure 3.5-1). These regions aro located in the high power / low flow area of the curve and can be encountered during startup, shutdown, rod sequence exchange or recirculation pump trip.

Operation in these regions is associated with higher than normal neutron flux noise levels.

Increased awareness of LPRM and APRM signal noise when operating in these regions will identify instability and allow operator action to correct the problem. The neutron flux noise level, thermal power and core flow lisaits are prescribed in accordance with the recommendations of General Electric Service Information Letter No. 380, Revision 1 "BWR Cara Thermal Hydraulle Stability", dated February 10, 1984.

Requiring the discharge valve of the lower speed -

loop to remain closed until the speed of the faster pump is below 50 percent of its rated speed provides assurance when going from one to two pump operation that excessive vibration of the jet pump risers will not occur.

K. References

1. "FitzPatrick Nuclear Power Plant Single-Loop Operation", NEDO-24281, August 1980.

AmendmentNo.[.

Figure 3.5-1 Thermal Power and Core Flow Limits of Specifications 3.5.J.1, 3.5.J.2 and 3.5.J.3 70 Stability Stability Monitoring mnitoring (APRM and LPRM) Required Stability Monitoring (APFM and

, For Single Loop 60 - (APRM and LPRM) Required LPFM)

During Two-Loop Operation Required Operation g

w During B

Single and Line A I w o-Loop 8 50 - .

Operation gj Single-Loop Operation g Prohibited , g m

S.  ; I 40 -

y w l b 35 -

l I

l

$ 30 -

g E I w I

I I

w 20 Stability Mor!itoring Not Required g -

g i

o l l

I l

10 -

I I

l I

I

. t 0 ,i i,i ii i i ii i i i 8 30 40 45 50 60 70 CORE FLOW (PERCENT RATED)

Amendment No. [ , [ ,JPI, f4 134

JAFNPP Figure 3.5-9 13-Reload 4 P8DRB284H a -

22 12-

=t M h, m 12 -

5f N-o?

E E 10 *

$c 5$

S Ji xe a 9-5*

, , , i . . . .

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

Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)

Versus Planar Average Exposure For single-loop operation, these MAPLHGR

Reference:

NEDO-21662-2 I values are multiplied by 0.84. (As amended August 1981 )

Amendment No. Jb(' 135g

JAFNPP Figure 3.5-10 13" Reloads 4 & 5 P8DRB299 a

g7 12-

• R u3

$5

.3 a:

"$ 11-81 E-o8 5'@ 10-

$c

>o

<g S2 av

• • @ 9-z 5e I I I I 3 3 8 3 5 10 15 20 25 30 35 40 Planar Average Exposure (GWD/t)

Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)

~

Versus Planar Average Exposure For single-loop operation, these MAPLHGR

Reference:

NEDO-21662-2 I values are multiplied by 0.84. (As amended August 1981 )

Amendment No. g , g 135h

JAFNPP 4 Figure 3.5-11 l }-

Reload 6 BP8DFD299 a

gg 12-m u3E

@ d!

b5 m '

"E 11-

$d

%2 4 5-o3

$M 10-

$c NSo NO

!!8 9' 55 s

I a s s e 4 5 10 15 20 25 30 35 40 Planar Average Exposure (GWD/t)

Maxirum Average Planar Linear Heat Generation Rate (MAPLHGR)

Versus Planar Average Exposure For single-loop operation, these MAPLHGR

Reference:

NEDO-21662-2 I values are multiplied by 0.84. (As amenced December 1984)

Acendment No. JN?' 1351

-m

JAFNPP 4.6 (cont'd)

1. The two recirculation loops have a flow

, imbalance of 15 percent or more when the 1

pumps are operated at the same speed.

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

3. The diffuser to lower plenur. differential pressure reading on an individual jet pump varies from the average of all jet pump differential pressures by more than 10 percent.

A. Whenever the reactor is in the startup/ hot '

standby or run modes, and there is one loop recirculation flow, jet ptesp operability shall be verified as follows.

a. Baseline enadings will be taken had operating characteristics- for the following parameters established:

1. Jet Pump Loop Flev end Recircul,4tica '

Pump Speed for' the c; crating isop.

l

2. Individual Jet Pump percent differen'.lal pressures for all jet pum;n.

b. In111elly, and defly thereafter, jet pwap ops ubility will be verified by accuring that the following de rot occur sinnitersously:

• AmendmentNo.[ '

b r v --

e JAFNPP 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 ludividual jet pump percent differential pressure to the loop's average jet pump percent differential press tre does not vary from the ir.itially established value by more than 20 percent.

Amendment No. )d,jVI 145a

.. - - , . . , - . . , _ . -,-.--,.,..--,,.v. . . - -. - ,. .---..-.. . , -- -. _- ., -...

• O JAFNPP LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 3.6 (cont'd) 4.6 (cont'd) t 3.6.I Shock Suppressors (Snubbers) 4.6.I Shock Suppressors (Snubbers)

Applicability Applicab111tv Applies to the operational status of the Applies tio the periodic testinst, requirement shock suppressors (snubbers). for the shock suppressors (snubbers).

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

Prevent unrestrained pipe motion under dynamic loads as might occur during an earthquake or savare transient, and Allow normal thermal motion during startup and shutdown. Specification Specification Each snubber shall be demonstrated operable by performance of the following augmented

1. During all modes of operation except Cold inservice inspection program.

Shutdown and Refueling, all snubbers which are required to protect the primary 1. Snubbers shall be visually inspected in coolant system or any other safety related accordance with the following schedule:

system or component shall be operable.

During Cold Shutdown or Refueling mode of No. Inoperable Snubbers Subsequent Visual operation, only those snubbers shall be per Inspection Period Inspection Period *#

operable which are on systems that are required to be operable in these modes. 0 18 months i 25%

1 12 months 1 25%

2. With one or more snubbers inoperable. 2 6 months i 25%

within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during normal operation. 3,4 124 days i 25%

or within 7 days during Cold Shutdown or 5,6,7 62 days 1 25%

Refueling mode of operation for systems 8 or more 31 days 2 25%

• The inspection interval may not be extended more then one step at the time.

145b

JAFNPP 3.6 (cont'd) 4.6 (cont'd) which are required to be operable in these # The snubbers may be categorized into two modes, cceplete one of the following: groups: Those accessible and those inaccessible during reactor operation. Each

a. replace or restore the inoperabic group may be inspoeted independently in snubber (s) to operable status or, accordance with the above schedule, i b. declare the supported system inoperable 2. Visual inspection shall verify (1) that

and follow the appropriate limiting there are no visible indications of damage condition for operation statement for or impaired OPERABILITY, (2) attachmonts to that system or, the foundation or supporting structure are secure, and (3) in those locations where

c. perform an engineering evaluation to snubber movements can be manually induced demonstrate the inoperable snubber is without disconnecting the snubber, that the unnecessary to assure operability of.the snubber has freedom of movement and-is not ,

system or to meet the design criteria of frozen up. Snubbers which appear i

the system, and remove the snubber from . inoperable as a result of visual

. the system. inspections may be determined OPERABLE for the purpose of establishing the next visual

3. With one or more snubbers found inoperable, inspection interval, providing that (1) the within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> perform a visual inspection cause of the rejection is clearly of the supported component (s) associated established and remedied for that with the inoperable snubber (s) and document particular snubber and for other snubbers the results. For all modes of operation that may be generically susceptible; and except Cold Shutdown and Refueling, within (2) the affected snubber is functionally 14 days complete an engineering evaluation tested in the as found condition and as per Specification 4.6.I.6 to ensure that determined OPERABLE per Specifications the inoperable snubber (s) has not adversely 4.6.I.7 or 4.6.I.8, as applicable.

affected the supported component (s). For Hydraulic snubbers which have lost Cold Shutdown or Refueling mode, this sufficient fluid to potentially cause evaluation shall be completed within 30 uncovering of the fluid reservoir-to- ,

days. snubber valve assembly port or bottoming of '

the fluid reservoir piston with the snubber i

AmendmentNo.g,[

145c 1

JAFNPP 3.6 (cont'd) 4.6 (cont'd) in the fully extended position shall be functionally' tested to determine operability.

3. Once each operating cycle, 10% of each type.

of snubbers shall be functionally tested for operability, either in place or in a bench test. For each unit and subsequent unit that does not meet the requirements of 4.6.I.7 or 4.6.I.8, an additional 10% of that type of snubber shall be functionally tested until no more failures are found, or all units have been tested.

4. The eipresentative sample selected for functio.ially testing shall include the various configurations, operating environments and the range of size and capacity of snubbers. At least 25% of the l snubbers in the representative sample shall include snubbers from the following three categories:

a. The first snubber away from reactor vessel nozzle.

b. Snubbers within 5 feet of heavy

equipment (valve, pump, turbine, motor, etc.).

c. Snubbers within 10 feet of the discharge from a safety relief valve.

Amendment No. g l

145d

JAFNPP 3.6 (cont'd) 4.6 (cont'd)

In addition to the regular sample, snubbers which failed the previous functional test shall be reteeted during the next test period. If a spare snubber has been installed in place of a failed snubber, then both the failed snubber (if it- is repaired and installed in anothee position) and the spare snubber shall be rotested.

Test results of these snubbers may not be included for the re-sampling.

5. If any snubber selected for functional I testing either fails to lockup or fails to move, i.e. is frozen in place, the cause will be evaluated and if due to manufacturer or design deficiency, snubbers of the same design subject to the same defect shall be functionally tested. This testing requirement shall be independent of the requirements stated above for snubbers not meeting the functional t e s '. acceptance criteria.

6. For the snubber (s) found inoperable, an engineering evaluation shall be performed on the components which are supported by the snubber (s). The purpose of this engineering evaluation shall be to determine if the components supported by the snubber (s) were adversely affected by the inoperability of the snubber (s) in order to ensure that the supported components remain capable of meeting the designed service requirements.

AmendmentNo.[ 145e

JAFNPP 3.6 (cont'd) 4.6 (cont'd)

7. The hydraulic se.Sber functional test shall verify that;

a. Activation (restraining action) is achieved within the specified range of velocity or acceleration in both tension and compression.

b. Snubber bleed, or release rate. where required, is within the specified range in compression or tension. For snubbers specifically required not to displace

' under continuous load, the ability of the snubber to withstand load without displacement shall be verified.

8. The mechanical snubber functionel test shall verify that:

a. The force that initiates free movement of the snubber rod in either tension 'or compression is less than the specified maximum drag force. Drag force shall not have increased more than 50% since the last functional test.

b. Activation (restraining action) is achieved within the specified range of velocity or acceleration in both tension and compression.

l Amendment No. k 145f

JAFNPP 3.6 (cont'd) 4.6 (cont'd)

c. Snubber release rate, where required, is within the specified range in compression or tension. For snubbers specifically required not to displace under continuous load, the ability of the snubber to withstand load without l displacement shall be verified.

9. Snubber Service Life Monitoring A record of the service . life of each snubber, whose failure could adversely affect the primary coolant or other safety-related system, the date at which the designated service life commences, and the installation and maintenance records on which the designated service life is based shall be- maintained as required by specification 6.10.B.13.

At least once per operating cycle, the  ;

insta11etion and maintenance records for each snubber, whose failure could adversely affect the primary coolant or other safety related system, shall be reviewed to verify that the indicated service life has not been exceeded or will not be exceeded prior i

5 to the next sheeduled snubber service life review. If the indicated service life will be exceeded prior to the next scheduled snubber service life review, the snubber service life shall be reevaluated or the snubber shall be replaced or reconditioned so as to extend its service life beyond the date of the next schedule service life review. This reevaluation, replacement or reconditioning shall be indicated in the records.

Amendment No. g 1453

JAFNpP 3.6 and 4.6 BASES (cont'd)

(

! would provide a leckage path past the core thus reducing the core flow rate. The reverse flow through the inactive jet pump would still be indicated by a positive differential pressure but the net effect would be a slight decrease (3 percent to 6 percent) in the total core flow measured. This decrease, together with the Icap flow increase, would result in a lack of corto-lation between measured and derived core flow rate. Finally, the affected jet pump diffuser differential pressure signal would be reduced because the backflow would be less than the normal forward flow.

A nozzle-riser system failure could also generate the coincident failure of a jet pump body; however, the converse is not true. The lack of any substantial stress in the jet pump body makes failure impossible without an initial nozzle-riser system failure.

Surveillance tests are performed to verify jet pump operability. Significant changes in either:

(1) the relationship between loop flow and recirculation pump speed, or (2) individual jet pump differential pressure compared to average jet pump differential pressure, sie used to detect degraded jet pump performance.

n Amendment No. J/

155

6 JAFNPP l

3.6 and 4.6 BASES (cont'd) 4 H. (DELETED) followed. As an alternative to snubber repair or replacement an engineering evaluation may be I. Shock Suppressors performed: to demonstrate that the inoperable snubber is unnecessary to assure operability of Snubbers are designed to. prevent unrestrained pipe the system or to meet the design crit'eria of the notion under dynamic loads as might occur during system; and, to remove the snubber from the an earthquake or severe transient, while allowing -

system. With one or more snubbers found normal thermal motion during startup and shutdown. Inoperable, within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> a visual inspection The consequence of an inoperable snubber is an shall be performed on the supported component (s) i increase in the probability of structural damage associated with the inoperable snubber (s) and the to piping as a result of a seismic or other. event results shall be documented. For all modes of' initiating dynamic loads. It is therefore operation except Cold Shutdown and Refueling, required that all snubbers required to protect within 14 days an engineering evaluation shall be

, the primary coolant system or any other safety performed to ensure tht the inoperable snubber (s)

system or component be operable during reactor has not adversely affected. the supported operation. Snubbers excluded from this component (s). For Cold Shutdown or refueling inspection program are those installed on mode, this evaluation shall be completed within

non-safety related system and then only if their 30 days. A period of 7 days Ms been selected

failure or failure of the sys.en on which they for repale or replacement of the inoperable l are installed would have no adverse effect on any snubber during cold shutdown or refueling mode of safety-related system. Because the snubber operation becuase in these modes the relative protection is required only during low probability of structural damage to the piping probability events, a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (for systems would be lower due to lower values of

. normal operation) or 7 days (for cold shutdown or total stresses on the piping systems. In case a

! refueling mode of operation) is allowed for shutdown is required, the allowance .of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> repairs or replacement of the snubber prior to to reach a cold shutdown condition will permit an i taking any other action. Following the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> orderly shutdown consistent with standard I

(or 7 day) period, the supported system must be operating procedures.

declared inoperable and the Limiting Condition of operation statement for the supported system i

AmendmentNo.[,[

j 156 l

ATTACHMENT II TO JPN-85'y I

s i

SAFETY EVALUATION FOR TECHNICAL SPECIFICATION CHANGES RELATED TO SINGLE RECIRCULATION LOOP OPERATION (JPTS-85-OO7) l i

[

F NEW YORK POWER AUTHORITY I

JAMES A. FITZPATRICK NUCLEAR POWER PLANT  !

DOCKET NO. 50-333  !

DPR-59 i

I. Description of Proposed Chances The proposed changes to the FitzPatrick Technical

• Specifications will permit reactor operation with a single recirculation loop.

The changes proposed also incorporate the recommendations described in General Electric's Service Information Letter No. 380, Revision 1. "BWR Core Thermal-Hydraulic Stability" (Reference 3) into the FitzPatrick Technical Specifications as the Authority committed in Reference 8.

Specifically, the following changes are proposed:

a. Page 11 (Table of Contents) is updated to show the renumbering of page 145a to 145b, Specification 3.6.I, Shock Suppressors (Snubbers). It also reflects the deletion of Specification 3.6.H (Jet Pump Flow Mismatch).

b. Page vii, (List of Figures) has been revised to show the addition of Figure 3.5-1 on page 134.

c. Page 7, Specification 1.1.A. add the following sentence:

~

"An MCPR limit of 1.06 shall apply during single-loop operation." This increased MCPR limit accounts for the rise in uncertainty in core flow measurement and TIP reading uncertainty during single-loop operation.

d. Page 8, Specification 2.1.A.I.c, add an equation for single-loop operation flux scram trip setting and define the

[i W term in the new equation. The new equation accounts for the difference in drive flow for the same core flow for two-loop versus single-loop operation.

e. Page 8a, a new page containing text previously on page 8. No new or changed material is included on page 8a.

f. Page 9, Specification 2.1.A.l.c, add the equation for single-loop operation flux scram trip setting for operation with a Maximum Fraction of Limiting Power Density (MFLPD) greater than the Fraction.of Rated Power (FRP). This equation accounts for the difference in drive flow for the same core flow for two-loop versus single-loop operation.

g. Page 10, Specification 2.1.A.1.d, add equations for single-loop operation APRM rod block trip setting and define [i W term in both equations. These equations account for the difference in drive flow for the same core flow in two-loop versuc single-loop operation,

h. Page loa, a new page containing text previously on page

10. No new or changed material is included on page 10a.

, i. Pago 12 Spocification A. cdd tho contence: "Th2 MCPR fuol cicdding ocfoty lioit is increcocd by 0.01 for single loop operation as discussed in Reference 2."

This addresses the same factor discussed in the change to page 7.

j. Page 14, Specification 1.1.E (Bases), add Reference numbers 2 and 3: "2. FitzPatrick Nuclear Power Plant Single-Loop Operation NEDO-24281 August 1980, Revision 1" and "3. Generic Reload Fuel Application, NEDE 240ll-P-A and Appendices".

k. Page 15 Specification 2.1 (Bases), add paragraph pertaining to reload transient analysis. The paragraph states that the core wide transient for single-loop operation is conservatively bounded by two-loop operation.

1. Page 20 Specification 2.1.C (Bases), add Reference number 4: "4. FitzPatrick Nuclear Power Plant )

Single-Loop Operation, NEDO-24281 August 1980".

l

m. Page 31, Specification 3.1, add Note following table l "MCPR Operating Limit for Incremental Cycle Core Average l Exposure". This note reflects the increase in safety l limit MCPR as described in changes made on page 7.

i n. Page 41. Table 3.1-1, add Note (17) following equation for APRM Flow Referenced Neutron Flux. See page 43a for a description of Note (17).

Also, corrects typographical error introduced in Amendment No. 87. In the column titled " Modes in Nhich Function Must be Operable", for item "APRM Flow Referenced Neutron Fluxr, an "X" was erroneously moved from the "Run" subcolumnJto the " Refuel" subcolumn.

This trip function is required only during the "Run" l mode.

( o. Page 43a, Notes of Table 3.1-1, add Note (17). This I note reflects the addition of the equation for.

calculating the APRM flow referenced scram setting for single-loop operation. _This is the same equation added on page 8.

p. Page 72 Table 3.2-3, add Note (11) after APRM upscale (Flow Biased); add Note (12) after Rod Block Monitor (Flow Biased). Page 74 describes Notes (11) and (12).

q. Page 74, Notes for Table 3.2-3, add Notes (11) and (12). Note (11) reflects the addition of the equation for calculating the APRM Rod Block Line for single-loop' operation. Note (12) reflects the addition of the equation for calculating the RBM Rod Block line for single-loop operation.

r. Paga 123 Spscification 3.5-H cd

. phroso: "for two-loop oparation.d the folicwingFor single-loop oporation thsse valucs cro reducsd by cultiplying by 0.84 (see Reference 1, Specification 3.5.K)". This reflects the comparison of single-loop to two-loop determination'of MAPLHGR. The reduction factor used reflects the calculated peak cladding temperature for the limiting break size for single-loop operation versus two-loop operation.

, s. Page 124a, 124b and 124c Specification 3.5.J (Thermal-Hydraulic Stability) specify limiting conditions for operation and surveillance requirements for single-loop operation. This Specification provides parameters for operation such as: core flow versus thermal power, and APRM and LPRM noise levels. These specifications implement the recommendation of General Electric's Service Information Letter (SIL) No. 380 (Reference No.

3).

t Specification 4.5.J.1 on page 124a establishes surveillance requirements to establish baseline APRM and LPRM neutron flux noise values for each refueling.

t. Page 130, Specification 3.5.H. add the following sentence at the.end of paragraph: "The reduction factor for single-loop operation for the above curve is 0.84.

The derivation of this factor can be found in Specification 3.5.K. Reference 1." This is the same factor added to page 123.

, u. Page 131, Specification k. (References) add Reference No. 1: "FitzPatrick Nuclear Power Plant Single Loop Operation, NEDO-24281, August 1980." Also add new last i

paragraph to Specification J (Thermal Hydraulic Stability); a similar paragraph was previously in the B.4SES associated with Specification 3.6.H (p. 156).

v. Page 134, Figure 3.5-1 added " Thermal Power and Core Flow Limits of Specifications 3.5.J.1 and 3.5.J.2".

This figure provides a g'uide for monitoring requirements set forth in the above specifications,

w. Pages 135g through 1351, (Figures 3.5-9 through 3.5-11),

add the following note: "For single-loop operation, these MAPLHGR values should be multiplied by 0.84."

l This is the same reduction factor discussed on page l 123. A typographical error is also corrected on these pages.

xc Pages 145 and 145a, Specification 4.6.G, add Specification 4.6.G.A. This Specification establishes surveillance requirements for jet pumps during single-loop operation. The requirements are for monitoring jet pump loop flow versus recirculation pump speed; and, the ratio of individual jet pump differential pressure to the loop average jet ~ pump differential pressure. Add new page 145a to accommodate addition.

l l

Ssetion 3.6.H (Jot Purp Flcw Miecatch) on page 145 was i deleted. Sections 3.6.H.1 and 4.6.H.1 applied to the l LPCI loop selection logic. Technical Specification Amendment No. 8 (Reference 11) authorized the removal of these RHR logic circuits: Amendment No. 14 (Reference 12, pages 11 and 12) of associated Safety Evaluation authorized the deletion of these sections. Section l

3.6.H.2 has been replaced by Specification 3.5.J.4 on

! Page 124b.

y. Page 145b, a new page containing text previously on page 145a. No new or changed material is included on page 145b.

z. Page 145c, a new page containing text previously on page 145b. No new or changed material is included on page 145c.

aa. Page 145d, a new page containing text previously on page i 145c. No new or changed text is included on page 145d.

bb. Page 145e, a new page containing text previously on page 145d. No new or changed text is included on page 145e.

cc. Page 145f, a new page containing text previously on page 145e. No new or changed text is included on page 145f.

dd. Page 145g, a new page containing text previously on page 145f. No new or changed text is included on page 145g.

ee. Pages 155 and 156, delete BASES associated with Sections 3.6.H and 4.6.H (Jet Pump Flow Mismatch). The last paragraph of this Section on p. 156 has been moved to the BASES associated with Sections 3.5 and 4.5 on page 131.

Also, a new paragraph has been added on page 155 describing the surveillance requirements for assuring jet pump operability during single loop operation.

II. Purpose of the Proposed Chances The proposed changes are necessary to permit power operation ,

with one reactor recirculation loop out of service.

1 III. Impact of the Proposed Chances The reactor recirculation system provides coolant flow through the core. The capability of operating at reduced power with a single recirculation loop in service is '

desirable from a plant availability / outage planning '

standpoint, in the event one loop is inoperative. The analyses reported in Reference 2 demonstrate that these r

chcagas to paroit reactor operation with only a single recirculation loop will not decrease overall plant safety.

During periods of single-loop operation, the MCPR (Maximum Critical Power Ratio) safety limit is increased to account  ;

for greater core flow measurement and Traversing Incore  ;

Probe (TIP) reading uncertainties. This MCPR increase of ,

0.01 preserves the margin to boiling transition established  ;

in Specification 1.1.A of the Technical Specification Bases.  ;

Single-loop operation results in a maximum power output 20 to 30 percent less than that attainable during two-loop operation. Because of this reduction in maximum power, the  !

consequences of core-wide, abnormal operational transients will be less severe. The reload analysis results, previously transmitted, establish an. upper limit on the  ;

thermal and over-pressure consequences for uingle-loop i. l operation. The flow-biased RBM (Rod Block Monitor) setpoint l changes described above assure that the thermal conrequences  ;

of a rod withdrawal error during single-loop operation will i be bounded by the two-loop analyses included with the Authority's reload application. A reduction factor will be applied to each MAPLHGR (Maximum Average Planar Linear Heat Generation Rate) curve as a result of differences in loss-of-coolant accident (LOCA) for single-loop versus l two-loop operation. These reduction factors will assure that a peak clad temperature of 2200 degrees Fahrenheit [

(Appendix K to 10 CFR 50) is not exceeded during a LOCA. i The most significant contribution to a decreased MAPLHGR l during single-loop operation is the very short. time to boiling transition assumed (0.1 second versus approximately  !

10 seconds assumed in the two-loop analysis.)- This [

conservative assumption accounts for decreased forced i circulation during recirculation pump coastdown in the early stages of a LOCA.

The proposed changes include monitoring of jet pump  !

operability when in single-loop operation. This is l necessary because of the different hydraulic characteristics  !

of the reactor during single-loop operation. {

i Changes related to monitoring of LPRM and APRM noise levels i when the reactor is operated at high power and low flow [

guard against reactor operation in an unstable mode. i Unstable operation is identifiable by high noise levels on  !

APRM and LPRM indicators. In the event that high noise is i detected, operator action will minimize the duration of any L instability, thereby increasing the safety of the plant. [

The changes associated with increased monitoring are based [

upon guidance suggested by the General Electric Co. in their Service Information Letter No. 380, Revision 1 (Reference 3).

The Commission has provided guidance concerning the i application of the standards for making a "no significant i hazard considerations" determination by providing certain ,

h i

oxceplee in th9 Fcderal Register (F.R.) Vol. 48, No. 67 dated April 6, 1984, page 14870. The proposed change matches example (iv) which states:

"A relief Tranted upon demonstration of acceptable operation from an operating restriction that was  :

imposed because acceptable operation was not yet demonstrated. This assumes that the operating restriction and the criteria to be applied to a request for relief have been established in a prior review and that it is justified in a satisfactory way'that the criteria have been met."

The proposed change is similar to this example in that single-loop operation had not been allowed for any length of time due to the potential for thermal hydraulic instability. Reference 3 outlines tests performed by General Electric'which identified parameters to monitor thermal hydraulic stability.

These parameters have been incorporated in the proposed change. Similar parameters have been included in submittals made by other utilities.

This proposed amendment to the FitzPatrick Technical Specifications can be classified as not likely to involve a significant hazards consideration since this proposed amendment, as per 10 CPR 50.92, does not:

(1) involve a significant increase in the probability or consequences of an accident previously evaluated because the number of operational recirculation loops does not affect the outcome or probability of most accident scenarios. These exceptions have been analyzed in Reference 2. The least stable thermal hydraulic condition (which' results from tripping both recirculation pumps) is unaffected by single-loop operation. Revised MCPR fuel cladding integrity limit and scram setpoint equations further assure that the probability or consequences are not significantly increased. The  :

proposed changes also incorporate changes to  !

improve thermal-hydraulic stability during l two-loop operation. Similar restrictions imposed  :

during single-loop operation will further assure  !

that the probability or consequences of an  !

accident are not significantly increased.  !

i (2) create the possibility of a new or different kind i of accident from any accident previously evaluated

~

because the transients identified'as part of our l analysis are conservatively bounded by similar i transients during two-loop operation. Abnormal l operating transients which could be initiated i because of single-loop operation have been l analyzed in either the FitzPatrick FSAR or  !

NEDO-24281 (Reference 2).

l l \

l [

! t l

I

e (3) involve a significant reduction in a margin of safety because an analysis has been completed

. (Reference 2) considering the changes required to implement the requested changes. To maintain the margin of safety present during two-loop operation, a 0.01 incremental increase in the MCPR fuel cladding integrity safety limit is imposed during single-loop operation. This adjustment accounts for increased uncertainties in total core flow and TIP readings. No other increase in this limit is necessary because core-wide transients are bounded by the rated power / flow analyses performed for each fuel cycle. Scram setpoint equations have also been altered to ,

consider the differences associated with 1 single-loop operation. Operation with one recirculation loop results in a maximum power output twenty to thirty percent below that attainable during two-loop operation. Therefore, the consequences of abnormal transients occurring during single-loop operation are~significantly less severe than those during two-loop operation.

IV. Implementation of the Chances Implementation of the changes, as proposed, will not impact the ALARA or Fire Protection Programs at FitzPatrick, nor will the changes impact the environment.

V. Conclusion The incorporation of these changes:

a) will not change 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 of an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report; c) will not reduce the margin of safety as defined in j the basis for any Technical Specifications; d) does not constitute an unreviewed safety question

as defined in 10 CFR 50.59; and e) involves no significant hazard considerations, as defined in 10 CFR 50.92.

(

l

VI. References

1. James A. FitzPatrick Nuclear Power Plant Final Safety Analysis Report (FSAR), Rev. 1. July, 1983.

F

2. "FitzPatrick Nuclear Power Plant Single Loop Operation",

General Electric Co., NEDO-24281, August 1980.

3. General Electric Co. Service Information Letter (SIL)

No. 380. Revision 1, February 10, 1984, "BWR Core Thermal-Hydraulic Stability."

4. USNRC letter, H. I. Abelson to J. P. Bayne, dated November 16, 1983 regarding Notice of Consideration of Issuance of Amendment for Single-Loop Operation.

5. NYPA letter, J. P. Bayne to D. B. Vassallo, dated July 25, 1983 (JPN-83-71) regarding Proposed Changes to the Technical Specifications related to Single-Loop Operation.

6. USNRC letter, T. A. Ippolito to G. T. Berry, dated August 24, 1980 regarding meeting scheduled for September 9, 1981 on Single-Loop Operation at power levels above fifty percent.

7. PASNY letter, J. P. Bayne to D. B. Vassallo, dated January 27, 1983 (JPN-83-08) regarding proposed changes to Technical Specifications related to Single-Loop Operation.

4

8. NYPA letter, J. P. Bayne to D. B. Vassallo, dated April 8, 1985 (JPN-85-27) regarding Reload 6/ Cycle 7 -

Thermal Hydraulic Instability.

9. NRC letter, D. B. Vassallo to J. P. Bayne, dated March 26, 1985 regarding Reload 6/ Cycle ? - Thermal Hydraulic Instability.

10. PASNY letter, J. P. Bayne to T. A. Ippolito, dated December 29, 1982 (JPN-82-92) regarding Proposed Technical Specificatioh Changes Related to Single Loop Operation,

11. NRC letter, Robert W. Reid to G. T. Berry, dated January 15, 1976 regarding Amendment No. 8 authorizes modifications to improve LPCI function; removal of LPCI recirculation loop selection logic.

12. NRC letter, Robert N. Reid to G. T. Berry, dated March 12, 1976 regarding Amendment No. 14 authorizes operation with modifications authorized in Amendment No. 8.

13. General Electric Co. Service Information Letter (SIL)

No. 330, June 9, 1980, " Jet Pump Beam Cracks"; Appendix A - BWR 3/4 Jet Pump Performance Monitoring".