ML20042E632
| ML20042E632 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 04/20/1990 |
| From: | POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20042E626 | List: |
| References | |
| NUDOCS 9004260233 | |
| Download: ML20042E632 (40) | |
Text
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ATTACHMENT I PROPOSED TECHNICAL SPECIFICATION CHANGES REGARDING
- REMOVAL OF CYCLE SPECIFIC PARAMETER UMITS JPTS 88-020 I
New York Power Authority _
JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333
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.- i-JAFNPP J t l 1
- UST OF FIGURES -
Figure Title Page ]
L 4.1 1 Graphic Ald in the Selection of an Adequate Interval Between Tests 48 4.2 1 Test interval vs. Probability of System Unavailability - 87 3.4 1 Sodium Pentaborate Solution 34.7 B 10 Atom % Enriched Volume- 110 Concentration Requirements 3.4 2 Saturation Temperature of Enriched Sodium Pentaborate Solution 111
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' x l- 3.5 1 Thermal Power and Core Flow Umits of Specifications 3.5.J.1,3.5.J.2 and 134 3.5.J.3 l
3.6 1 Reactor Vessel Pressure Temperature Umits' 163 4.61 Chlorlds Stress Corrosion Test Results at 500 F 164 -
L 6.1 1 (Deleted) 259 '
I 6.2-1 (Deleted) 260 r l
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5 l- Amendment No, M, g,4Ef, pdf,7(,74',86,96,195,1#3,148,1#f,1M,187 vil a
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JAFNPP 1.0 (cont'd) V. Electrically Disarmed Contiot Rod sarm a M h MMy,tMw @%pe @g surveillance tests, checks, calibrations, and examinations shall m __ . connectors are removed from the drive insert and withdrawal be pouv_ m__m_d within the specified surveillance intervals. These intervals may be adjusted 25 percent. The interval as r the M '@dWMh procedure is equivalent to valving out the drive and is preferred.
pertaining to instrument and electric surveillance shall never M e ion.
exceed one operating cycle. In cases where the elapsed interval has exceeded 100 percent of the specified interval, the next W. High Pressure Water Fire Protection System surveillance interval shall commence at the and of the original N @ Pressure h Fire Prde W m d a sW intental. water source and pumps; and disenbuhon system piping with U. Thermal Parameters associated post indicator valves (isolabon valves). Such valves include the yard hydrant curb valves and the first valve ahead of
- 1. Minimum critical power ratio (MCPR)- Minimum value of the water flow alarm device on each sprinkler or water spray l
the ratio of that power in a fuel assembly which is subsystem.
calculated to cause some point in that fuel essoireAy to X ' Stwed Test Basis expenence boiling transition to the actual essois eiy operating power for all fuel assemblies in the core. A Staggered Test Bases shall consist of:
l
- 2. Fraction of Limiting Power Density - The ratio of the linear a. A test schedule for *n" systems, subsystems, trains generat ate (LHGR) existing at a given location t a h designmed 6 WW h dividing the specmed test interval into "n" equal
- 3. Maximum Fraction of Umiting Power Density - The subentervals.
Maximum Fraction of Umiting Power Denssty (MFLPD) is b. The testing of one system, subsystem, train or other ue me m Fract'on i d component at the bcsivdiig of each
- 4. Transition Boiling - Transition boiling means the boiling region between nucleate and film boiling. Transition Y. Rated Recirculation Flow boiling is the region in whech both nucleate and film boiling That drive flow which produces a core flow of 77.0 x 10 8 lb/hr.
occur intermettently with neither type being completely stable.
Amendment No. 46,64,7E,74,199, #
6
JAFNPP AD. Core Operating Umits Report (COLR) op of Active M . This report is the plant-specific document that provides the core The Top of Active Fuel, corresponding to the top of the enriched operating limits for the current operating cycle. These cycle-fuel column of each fuel bundle, is located 352.5 inches above specific operating Imts shall be deterisiiried for each reload vessel zero, which is the lowest point in the inside bottom of the cycle in accordance with Specificahon 6.9.A.4. Plant operation within these operating limits is addressed in indmdual Tecin i; cal reactor vessel. (See General Electric drawing No. 919D690BD.)
AA. Rod Density W
- Rod density is the number of control rod notches inserted expressed as a fraction of the total number of control rod notches. - All rods fully inserted is a condition representing 100 percent rod density.
AB. Purge-Purging Purge or Purging is the controlled process of discharging air or gas from a confinement in such a manner that replacement air or p
gas is required to purify the confinement.
AC. Venting Venting is the controlled process of releasing air or gas from a confinement in such a manner that replacerrsit air or gas is not provided or required.
Amendment No. 7gp3 6a
. - _ . a.~--.=-
JAFNPP . ;
1.1 (cont'd) 2.1 (cont'd)
A.
1.
- b. APRM Flux Scram Trip Settog (Refuel or Start & Hot B. Core Thermal Power Umit (Reactor Presstne <785 psig) Standty Mode)
When the reactor pressure is <785 psig or core flow is less than APRM - The APRM flux scram setting shall be < 15 or equal to 10% of rated, the core thermal power shall not . percent of rated neutron flux with the Reactor Mode exceed 25 percent of rated thermal power. Switch in Startup/ Hot Standby or Refuel.
C. Power Transient c. APRM Flux Scram Trip Settings (Run Mode)
To ensure that the Safety Umit established in Specification 1.1.A g). - Tdp and 1.1.B is not exceeded, each required scram shall be initiated by its expected scram signal. The Safety Umit shall be assumed . .
to be exceeded when scram is accomplished by a means other When the Mode Switch is in the RUN possbon, than the expected scram segnal. the APRM flow referenced flux scram trip .
settog shall be less than or equal to the limit :
specified in Table 3.1-1. This setting shall be aqusted during sogle loop operation when required by Sf=cificai;06 3.5.J.
For no c&ntination of recirculabon flow rate -
. and core thermal power shall the APRM flux -
scram trip setting be allowed to exceed 117%
'of rated therntal power.
Amendment No. 1 / 36,43' 71,96,134 8'
. -.- _- = - . - -. = -- -- -. --- = = - - - - - - -
JAFNPP .
1.1 (cont'd) 2.1 (cont'd)
D. Reactor Water Level (Hot or Cold Shutdown Conditions) (2) Fixed High Neutron Flux Scram Trip Setting Whenever the reactor is in the shutdown condition with irradiated When the Mode Switch is in the RUN position, the fuel in the reactor vessel, the water level shall not be less than APRM fixed high 22x scram trip setting shall be:
that corresponding to 18 inches above the Top of Active Fuel S <120% Power when it is seated in the core.
equal to the limit specified in Table 3.2-3. This settog shall -
be adjusted dunng smgie loop operation when required by .
SpMahan 35.J.
Amendment No J4,;)6,4a, fii4,74,98, IFJ 9
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JAFNPP -
1.1 BASES 1.1 FUEL CLADDING INTEGRITY A. Reactor Pressure >785 psig and Core Flow >10% of Rated
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The fuel cladding integrity limit is set such that no calculated Onset of transition boshng results in a decrease in heat transfer fuel damage would occur as a result of an abnormal from the clad and, therefore,' elevated clad temperature and the operational transient. Because fuel diariage is not directly possibility of clad failure. However, the existence of critical observable, a step-back approach is used to establish a Safety power, or boihng transition, is not a directly obseivable l Umit minimum critical power ratio (MCPR). This Safety Umit parameter in an operating reactor. Therefore, the margin to represents a conservative margin relative to the cosiditions bosimg transition is calculated from plant operating parameters required to maintain fuel cladding integnty. The fuel cladding is such as core power, core flow, feedwater temperature, and one of the physical bamers which separate radioactive core power distnbution. The margin for each fuel assembly is materials from the environs. The integnty of this cladding characterized by the entica' power ratio (CPR) wtich is the
- barrier is related to its relative freedom from perforations or ratio of the bundle power which would produce onset of
, cracking. Although some corrosion or use related cracking transition boiling divided by the actual buncSe power. The may occur dunng the life of the cladding, fission product minimum value of this ratio for any bundle in the core is the :
, migration from this source is incrementally cumulative and _ minimum critical power ratio (MCPR). It is assumed that the continuously measurable. Fuel claddkig perforations, however, ' plant operation is controlled to the nominal protective setposiih, ,
can result from thermal stresses which occur from reactor via the instrumented variable, i.e., the operating domain.' The -
operation significantly above design conditions and the current load lino limit analysis contains the current operating protection system safety settings. While fission product domam map. The Safety Umit MCPR has sufficient migration from cladding perforation is just as measurable as
_l conservatism to assure that in the event of an abnormal that from use related cracking, the thermally caused cladding operational transsent initiated from the MCPR operating limit in perforations signal a threshold, beyond which still greater the Core Operahng Umsts Report, more than 99.9% of the fuel l<
thermal stresses may cause gross rather than incremental . rods in the core are expected to avoid bolhng transstion. The cladding deterioration. Therefore, the fuel cladding Safety Umit MCPR fuel claddog safety limit is increased by 0.01 for sogle-is defined with margin to the conditions which would produce loop operation as discussed in Reference 2. The margin onset of transition boiling, (MCPR of 1.0). These conditions between MCPR of 1.0 (onset of transition boshng) and the represent a significant departure from the condition intended Safety Umit is denved from a detailed statistical analysis by design for planned operation. considering all of the uncertanties in monstonng the core
, operating state including the uncertainty in the boshng transition -
correlation as (msuibed in Reference 1. The uncertanties employed in deriving the Safety Umit are i
I Amendment No.)f,1Q1,30,47,74 96,1#7,j8'l 12 i
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JAFNPP -
1.1 (cont'd) l provided in Reference 1. Because the boiling transition At 100% power, t'iis limit is reached with a maximum fraction of correlation is based on a large quantity of full scale data there is limiting power density (MFLPD) equal to 1.00. In the event of l a very high confidence that operation of fuel assaiMAy at the operation with a MFLPD greater than the frachon of rated power Safety Umit would not produce boeling transition. Thus, although (FRP), the APRM scram and rod block settmgs shall be adjusted it is not required to establish the safety limit, additional margin as specified in Tables 3.1-1 and 32-3 raipoC:dy. l exists between the Safety Umit and the actual occurrence of loss B. Core Thermal Power Umit (Reactor Pressure <785 psig) of dadding integrity.
At pesswes WM @ h cae he pesswe &op b However, if boiling transition were to occur, dad perforation W" '89'"" A would not be expected. Cladding temperatures would increase - p was W h,2 pesswe dop W to h hh to approximately 1100"F which is below the perforation pesswe & W @ h cae. W h M temperature of the dadding material. This has been venfied by fm p ww ange d 1-5 m,h h h d tests in the General Electric Test Reactor (GETR) where fuel mg 28 x W/A M h 6 h h similar in design to FitzPatnck operated above the critical heat pswa NW h h W region W h W flux for a segnificant penod of time (30 minutes) without clad channel. The pressure differential is primarily a result of changes P8'I" -
in the elevation pressure drop due to the density d'fference e
if reactor pressure should ever exceed 1400 psia dunng normal between the boshng water in the fuel channel and the non-boshng power operation (the limit of applicability of the boshng transition water in the bypass region. Full scale ATI.AS test data taken at correlation) it would be assumed that the fuel dadding integrity indicate that the fuel sssaisaiy Safety Umit has been violated. pressures critical powerfrom at 28 x0 10"Ib to 785 y/hr is approximately 3.35 MWt. With h W W fadas,26 to a cae W in addition to the boiling transition limit (Safety Umit), operation p ww d mwe h K % a cae M poww M d is constrained by the maximum LHGR identified in the Core 25% fw reacts pesswes N M Was conservat,m Operating Umits Report.
Amendment No.1f,2Q6,46,64,74,199,14 13
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JAFNPP ..
1.1 BASES (Cont'd) E. References C. Power Transient 1. General Electnc Standard App;k,rd;cm for Reactor Fuel, Plant safety analyses have shown that the scrams caused by NEDE-24011-P, latest approved revision and amandinents.-
exceeding any safety system setting will assure that the Safety 2. FitzPatnck Nuclear Power Plant Single-Loop Operation, Umit of 1.1.A or 1.1.B will not be exceeded. Scram times are NEDO 24281, August 1980.
, checked periodically to assure the insertion times are adequate.
The thermal power transient resulting when a scram is accca sp ished other than by the expected scram signal (e.g.,
scram from neutron flux following closure of the main turtune stop valves) does not necessarily cause fuel dc= cey3. However, for this specification a Safety Umit violation will be assumed when a scram is only accomplished by means of a backup feature of the plant design. The concept of not approaching a Safety Umit provided scram signals are operable is supported by the extensive plant safety analysis.
D. Reactor Water Level (Hot or Cold Shutdown Condition) i During penods when the reactor is shut down, conside ation 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 meltirg should the water level be reduced to two-thirds the core height.
Establisiiment of the Se'ety Umit at 181.1 above the top of the fuel provides adequate margin. This level will.be continuously monitored whenever the recirculation pumps are not operating. .I
-l Amendment No.1/,ptr q 14
i JAFNPP -
BASES 2.1 FUEL CLADDING INTEGRITY The most linutmg transsents have been analyzed to determine The nruing operational transients appikahle to operation of which result in the largest reduchon in CRITICAL POWER RATIO.
the FitzPatnck Unit have been analyzed throughout the spectrum The type of transients evaluated were increase in pressure and of planned operating eciditions up to the thermal power power, posstive reactivity insertson, and coolant temperature condibon of 2436 MWt.. The analyses were based upon plant decrease. The limstng transient yields the largest delta MCPR.
l When added to the Safety Umst, the required operating limit operation in accordance with the operating map given in the currerd load line limit analysis, in addition,2436 MWt is the MCPR in the Core Operating Umsts Report is obtained.
l licensed maximum power level of FitzPatrick, and this represents The evduation of a given transsent begens with the systen initid h h sWee powsM WW W & parameters shown in the current reload er.ey.;. and Reference 2 exW. that are input to the core dynamsc behavior transient computer The transient analyses performed for each reload are given in _ programs described in Reference 2. The output of these Reference 2. Models and model conservatism are also programs along with the initial MCPR form the input for the desenbed in this reference. As discussed in Reference 4, the further analyses of the thermally limited bunde with a sogle core wide transient analysis for one recirculation pump operation channel transsent thermal hydraulic code. The pnncipal result of is conservatively bounded by two-loop operation analysis, and the evaluation is the reduction in MCPR caused by the transient.
the flow-dependent rod block and scram setpoint equations are adjusted for one-pump operation.
Fuel cladding integrity is assured by the applicable operating limit MCPR for steady state conditions geven in the Core .
Operating Umits Report (COLR). These operating firnit MCPR's are derived from the established fuel cladding integrity Safety Umit, and an analysis of abnormal operational transients. For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady state operating limit, it is required that the resulting MCPR does not decrease below the Safety Umit MCPR at any time during the transient. -
Amendment No. 41f,64,7#,9ff 15
- ___- - - - =- = - - - -
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1 JAFNPP .
2.1 BASES (Cont'd) A. Trip Settings l The MCPR operating limits in the COLR are conservatively assumed to The bases for indmdual trip settings are discussed in tfm exist pnor to initiation of the transeents. follomng paragraphs.
This choice of using conservee wlues of controlling parameters and inetsahng transsents at the design power level, produces more Wrest;c 1- %T@ W answers than would result by usmg expected values of control parameters and weJy&g at higher power levels. "
84 % T@ W Steady-state operation without forced recirculation is not pomutted. The ,
analysis to support operation at various power and flow relahonstups IRM is a N 'NNmh has considered operation with esther one or two recerculahon pumps. range d power M h WW covered by #m in summary- SRM and the APRM. The 5 dar* are covered by the IRM by means of a range switch and the 5
. The abnormal operational transients were analyzed to the decadan are broken dcen into 10 ranges, each hcensed manomum powerlevel. being one-half of a darada in size. The IRM scram trip sening of 120 dvisions is active in each range of .
. The licensed e powerlevelis M MWt. the IRM. For exarnple,if tie instrument were on
. Analyses of transsents employ adequately conservative values of Range 1, Wie scram sothng would be a 120 divisions the controlling reactor parameters. for that range; likewise, if the instrument were on
. The #4/.;Oai procedures now used result in a more logscal range 5,#m scrarn would be 2 h on #wt g range. Thus,as #m Mis ranged up b answer than the altemabwe method of assuming a higher sic.i.g accomnsdale 9m h en power imi, We scram power in congunchon with the expected values for the g ,,,, trip sethng is also ranged up. The most signelicant sources of reactmey change dunng the power increase are due to control rod withdrawal. For insequence control rod withdrawal, the rate of change of power is slow enough due to the physical limstahon of withdramng control rods, that heat flux is in equihbnum with the neutron flux and an IRM scram would result in a reactor shutdown well before any Salsty Limit is exceeded.
Amendment No.1/,18',21,X 16 I .
JAFNPP from the MCPR operating limits specified in the Core 2.1 BASES (Cont'd)
Operating brmts Report.
- c. APRM Flux Scram Trip Setting (Run Mode) (cont'd) d. APRM Rod Block Trip Settog rated power. This reduced flow referenced trip setpomt Reactw m W @e varW m W rods d resuNin an ear #w scram dunng slow W or by varymg the recucuishon flow rate. The APRM transients, such as the loss of 80"F feedwater heating system provides a control rod block to prevent rod event, than would result with the 120% fixed high neutron gg , g flux scram trip. The lower flow referenced scram setpoint AWh an added M d prh therefore decreases the s6venty (ACPR) of a slow thermal APRM h This rod block hip Wwhichis transsent and allows lower Operating Lirmts if such a automahcaNy W with recrculdson loop Dow rde' transient is the Iwalmg abnormal operational transiert prevents an increase in the reactor power level to during a certam exposwe irWavd in the cycle. excesssve values due to control wmufrawd. The Sow The APRM fixed high neutron flux segnal does not vanable trip sottog parallels that of the APRM Scram and mcorporWe the time constant, but responds directly to provides marge to scram, assummg a steady-state instantaneous neutron flux. This scram setpost scrams operation at the trip settmg, over the entire recrculahon the reactor dunng fast power increase transsents if credit is flow range. The actual power distribuhon in the core is not taken for a direct (puM scram, and also serves to established by cw control rod sequences and is scram the reactor if credit is not taken for the flow momtored contwamusly by the in-core LPRM system. As referenced scram. with the APRM scram trip settmg, the APRM rod block trip The scram trip settog must be adjusted to ensure that the Wis @ W Wthe rnaxwnurn kachon d limiting power donosty exceeds the frachon of rated power, M hansiaW peak is not increased for any combination of maxwnum frachon of fimstmg power denssty (MFLPD) y scram swang, viis may be .44,iv.
fnargin. As e h by agushng the and reactor core thermal power. The scram setting is garn. g l adjusted as sperXied in Table 3.1-1 when the MFLPD is Ww the kactonypy (% Ms aqustment may be a,wsyumou by either (1) reduch.g
- 2. Reactor Water Low Level Scram Trip Settog the APRM scram and rod block settings or (2) adjusting The reactor low water levssi scram is set at a point which will the indicated APRM segnal to reRect the high peaking assure that the water level used in the Bases for the Safety Umit is mantamed. The scram setpout is based on normal operating condibon.
tanperdwe and pmsswe N W thM Analyses of the limiting trisrica..Ms show that no scram W ss denssty wsyuM adjustment is required to assure that the MCPR will be greater than the Safety Limit when the trans6ent is initiated Amendment No. 3Ef,179 18
JAFNPP 2.1 BASES (Cont'd)
C. References
- 1. (Deleted)
- 2. " General Electric Standard Appbcation for Reactor Fuel *,
NEDE 24011-P-A (Approved rension number appbcable at time that reload fuel analyses are perfamed).
l 0. (Deleted)
- 4. FitzPatrick Nuclear Power Plant Sogle-Loop Operation, NEDO-24281, August,1980.
Amerdient No. 93,#4,jl6 20-(Next page is 23)
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JAFNPP -
4.1 SURVEILLANCE REQU!REMENTS 3.1 UMITING CONDITIONS FOR OPERATION 3.1 REACTOR PROTECTION SYSTEM 4.1 REACTOR PROTECTION SYSTEM Applicarnif: W'p:
Apphes to the instrumentation md associated demces wtuch initiate the Apphes to the surveillarce of the instrumentabon and associated reactor scram. devices which wwbate reactor scram.
Ob!ective: Obrectmr To assure the operability d the Reactor Protechen System. To speedy the type of frequency of sunellance to be applied to the protecheninstrumentarm Specirication: SMI A. The setpoents, mamum number of trip systems, mamum A. Instrumentation systems shaN be functionaNy it,sied and number of instrument channels that must be operable for each calibrated as indmated in Tables 4.1-1 and 4.1-2 respectively, poist;cri of the reactor mode switch shall be as showm on Table 3.1-1. The desegn system respciise time from the opening of the sensor contact to and including the opening of the trip actuator contacts shaN not exceed 50 meec.
- 8. Maximum Fraction of Limshng Power Densely (MFLPD)
B. Minimum Critical Power Ratio (MCPR) hMMMN%Wem Dunng reactor power operation, the MCPR Ofm ainig limit shall operation at >25% rated thermd power and the APRM high flux not be less than that spMTmd in the Core Operating Limits - scram and Rod Block trip seinnsac%ustedif necessary as Report. specified in the Core Operating Limds Report. l
- 1. During Reactor power operation with core flow less than 100% of rated, the MCPR operating limit shall be mulhplied by the apfnupiiate Kg as specified in the Core Operating Lsmits Report.
Amendment No. 46,64,86,}09 3 01
_-_-.___.___m_- ---_ =
- JAFNPP -
3.1 (cont'd) 4.1 (cont'd)
! 2. If anytwne during reactor operation at greater than 25% of C. MCPR shaN be deteraned daily dunng reactor power operation rated power it is determned that the operating limit MCPR at >25% of rated thermal power and follomng any change in is bemg exceeded, achon shall then be inibated withm power level or distribubon that would cause operation with a fifteen (15) mmutes to restore operation to vnthm the limshng control rod pettom as described in the bares for
, presenbod Imts. If the MCPR is not retumed to withm the Speedicahon 3.3.B.S.
p %@ ,an mM re D. When it is determmed that a channel has faded in the unsafe power reduccan shall begn u' nmediately. 7% reactor power shall be reduced to less than 25% of rated power A h RPS h M e h m hours a undlthe MCPRis rehaned to vanable shaN be funchonaNy tested W W the trip
[,the next F -
system contanng the failure is tnpped. The trip system contammg the unsafe failure may be placed in the untnpped condihon dunng the penod in whsch survedlance testog is bemg performed on the other RPS channels.
! E. Venficahon of the MCPR operating limits shaN be performed as speedied in the Core Operating 1.imsts Report.
f Amendment No. C/,74, M,96,9tf,191, ty/
31
JAFNPP -
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I 3.1 BASES (cont'd) l Turbine control valves fast closures initiates a scrarn based on pressure switches sensing electrohydraulic control (EHC) systern oil prestoa. The switches are located between fast closure solenoids and the disc dump valves, and are set relabve (500 < P<850 psig) to the normal (EHC) oil pressure of I 1,600 psig so that based on the small system volume, they can rapidly detect valm c!osure or loss of hydraulic pressure.
The requirement that the IRtts be inserted in the core when the APRlWs read 2.5 indicated on the scale in #ie start-up and I refuel modes assures that there is proper overlap in the neutron morntonng system funcibs and thus, that adequate coverage !
j is provided for all ranges of reactor operabon.
B. The limming transsent witch determenes the required steady
- state MCPR limit depends on cycle exposure. The operating limit MCPR values as determred from the transient analysis in the current reload submittal for vanous core exposures are j i speedied in the Core Operating Umits Report (COLR).
l The ECCS pasivin.ance analyses assumed reactor operation i
will be limited to MCPR = 1.20, as desenbed in NEDO-21662 ,
and NEDC-31317P. The Ted m Ed Specificat;vis limit operation of the reactor to the more conservative MCPR based on consideration of the limiting transeent as spec;Ted in the COLR.
i- ,
l i Amendment No. 41f,64, }09 35 1
.- _ . ~ . . , . . . . , . . . . - - - . _ . _ _~ - . . . . _ ___. _ _ _ _ _ _ _ _ _ - _ _ _ _ _.. . _ - _ _ . _ _ - _ _ . -
4 JAFNPP TABLE 3.1-1 REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIREMENT Minimum No. ModeinWhich Funchon
- of Operable Must be Operable Total Number of
- instrument instrument Channels Channels Per Refuel Startup Run Provided by Design Action Trip System (1) Trip Function Trip Level Setting (6)(16)' for Both Trip Systems (1) 1 Mode Switchin X X X 1 Mode Switch A Shutdown (4 Selechons) 1 ManualScram X X X 2 Instrument Channels A 3 IRM High Flux < 120/125 of X X 8 Instiument Channels A full scale 3 IRM inoperative X X dinstrument Channels A 2 APRM Neutron Flux- < 15% Power X X 6 instrument Channels A Startup (15) l 2 APRM Flow Referenced (12) X 6 lastrument ChanntJs A or B Neutron Flux (Not to exceed l 117 %) (13)(14) 2 APRM Fixed High < 120% Power X 6 Instrument Channels A or B Neutron Flux (14) 2 APRMInoperative (10) X. X X 6 Instrumerd Civ=Ms A or B '
t j Amendment No. 18,30,46, R, IFT,96, J24 41
. , . . - . , ,- -, . - . . . . ~ - _ - - _ - - - _- -- - . - - - - -
l..II ..f JAFNPP TABLE 3.1-1 (cont'd)
REACTOR PROTECTION SYSTEM (SCRAM) INSTRUMENTATION REQUIRERENT NOTES OF TABLE 3.1 (cont'd)
C. High Flux IRM.
D. Scram Discharge Volume High Level when any control rod in a control cell contarwng fuel is not fully inserted.
E. APRM 15% Power Trip.
- 7. Not required to be operable when pnmary contammert integnty is not required.
- 8. Not required to be operable when the reactor pressure vessel head is not bolted to the vessel.
- 9. The APRM downscale trip is automabcally bypassed when the IRM instrumentahon is operable and not high.
- 10. An APRM will be conssdered operable if there are at least 2 LPRM w* puts per level and at least 11 LPRM inputs of the normal complement.
- 11. See Sechon 2.1.A.1.
l 12. The APRM Flow Referenced Neutron Flux Scram settmg shall be less than or equal to the limit specelled in the Core Operating tirruts Report.
- 13. The Average Power Range Monitor scram funchon is varied as a funchon of recirculation flow (W). The trip setting of this funchon must be mantamed as specified in the Core Operating Umits Report.
~
- 14. The APRM flow biased high neutron flux signalis fed through a time constant circuit of approxanately 6 seconds. The APRM fixed high
. neutron Bux segnal does not incorporate the time constant, but responds directly to instantaneous neutron flux.
- 15. This Average Power Range Monitor scram funchon is fixed point and is increased when the reactor mode switch is place in :he Run p61;cn.
- 16. *Dunng the proposed Hydrogen Addihon Test, the norma Mckground radiahon level will increase by approxunately a factor of 5 for peak hydrogen concentration.- Therefore, prior to performance of the test, the Main Steam Line Radiahon Monitor Trip Level Setpoint will be raised to < three twnes the increased radiation levels. The test will be conducted at power levels > 80% of normd rated power. Dunng controlled power reduchon, the setpomt will be readjusted pr5 to gomg below 20% rated power without the setpowd change, control rod withdrawal will be prohibited until the necessary trip seipckd ac$ustment is made.
- This specification is in effect only during Operating Cycle 7.
Amendment No. 46,lii( $14, Of, f/9,74, /4,1p9 43
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JAFNPP TABLE 32-3 INSTRUMENTATION THAT INITIATES CONTROL ROD ELCCKS '
Minimum No. 7 of Operable Total Number of instrument instrument Ctwies Channels Per Prowded by Design Trip System Instrument Trip Level Setting for Both Ctwies Action l 2 APRM Upscale (Flow Beased) (8) 6 inst. Ctanes (1) i 2 APRM Upscale (Start-up Mode) 512% 6 Inst. Channels (1) 2 APRM Downscale > 2.5 indicated on scale 6 Inst.Ctwies (1) l 1(6) Rod Block Monitor (Flow Besed) (8) 2 Inst.Ctanes (1) 1(6) Rod Block Monitor (Downscale) > 2.5 indicated on x. ale 2 Inst. Channels (1) 4 3 IRM Downscale (2) > 2% of full scale 8 Inst. Channels (1) 3 IRM Detector nef in Sta:t-up Position (7) 8 Inst.Ctw -es (1) 3 IRM Upscale 586.4% of full scale 8 Inst.Ctaves-(1) 2(4) SRM Detector not in Start-up Position (3) 4 inst. Channels (1)
SRM Upscale 5 2 (4)(5) 110 counts /sec 4 Inst. Channels (t) 1 Scram Discharge Instrument 526.0 gallons per 2 Inst. Channels (9)(10)
Volume High Water Level m
' strument volume NOTES FOR TABLE 32-3
- 1. For the Start-up and Run positions of the Reactor Mode Selector Switch, there shall be two operable or inppe.1 trip systems for each function.'
The SRM and IRM block need not be operable in run mode, and Amendment No. 44,(iF_, M, Sif 72
6 . . ..
JAFNPP TABLE 32-3 (Cont'd)
INSTRUMENTATION THAT INiTMTES CONTROL ROD BLOCKS NOTES FOR TABLE 32.-3 the APRM and RBM rod blocks need not be operable in star.-up rnode. Frorn and after the time it is found that the first column cannot be met for one of the two trip systems, this condition may exist for up to seven days pronded that during that time the operable system is functionally tested immediately and daily thereaner; if this condibon lasts longer than seven days, the system shall be tripped. From md aner the time it is found that the first column cannot be met for both trip systems, the systems shall be inpped.
- 2. IRM downscale is bypassed when it is on its lowest range.
- 3. This function is bypassed when the count rate is > 100 cps.
- 4. One of the four SRM inputs may be byprosed.
- 6. The trip is bypassed when the reactor power is < 305
- 7. This function is bypassed when the Mode Switch is placed in Run.
- 8. The Flow Biased APRM Upscale and Rod Block Monitor trip level setpoint shall be less than or equal to the limit specif.ed in the Core Operating Umits Report.
- 9. When the reactor is subcritical and the reactor water temperature is less than 212*F, the control rod block is required to be operable only if any control rod in a control cell contairung fuel is not fully inserted.
- 10. When one of the instruments associated with scram discharge instrument voim high water rod blocks is not operable, the trip system shall be tripped.
Amendment No. 46,OF,72)$
73
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JAFNPP 33 and 43 BASES (cont'd)
- 5. The Rod Block Monitor (RBM) is designed to automatically C. Scraminsertion Tirnes prevent fuel dimT,imy, in the event of erroneous rod withdrawal N Control Rod W is %d to % h re kom d@mM%@mM subenbcal at a rate fast enough to prevent fuel damage; i.e., to operation. Two channels are provided, and one of these may h MCPR kom W b then h W M be bypassed fmm the console for mantenance and/or teshng. Scram insertson time test cntena of Sechon 3.3.C.1 were used Trinping of one of the channels will block erroneous rod to generate the generic scram reachwity curve shown in Wai. val soon enough to W W W. NEDE-24011-P-A. This generic curve was used in analyses of This system backs up the operator who withdraws control rods rmpressurization transsents to determme MCPR limits.
according to written sequences. The specslied restnctions with Therefore, the requwed protechon is provided.
one channel out of service conservatsvely assure that fuel e to @ Wawd was & % The numencal values @W to the specified scram performance se based on the analyses of data from other A hmiting control rod pattem is a pattem which results in the BWR's with control rod dnves the same as those on JAFNPP.
core bemg on a thermal hydraulic limit (e.g., MCPR limit). The occurrence of scram times withm the Imts, but l
Dunng as of such pattems,it is judged that testeg of the RBM m,;;. ,i;f onger l than the average, should be viewed as an System prior to withdrawal of such rods to assure its operabilrty 66da p FN & W W h will assure that w' nproper withdraw does not occur. It is the .g g g, g N " 'U Y exceeds eight, the allowable nt nber of inoperable rods.
pattems and the W.e ed rods esther when the pattems are
! initially estabhshed or as they develop due to the occurrence of inoperable control rods in other tbsn hmsting pattens.
l Amendment No.14 1K ;Pi, 36, 42I, M, 55, 6T, Jii6 102
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JAFMPP -
3.5 (cont'd) 4.5 (cont'd) i cordtion, that pump shall be cons 4wed inoperable for 2. Follomng any penod where the LPCI subsystems or core i purooses of satisfying Spedfications 3.5.A. 3.5.C, and spray subsystems hewe not been mantamed in a filled ,
3.5.E. cordborg the discharge ppng of the affected subsystem ;
shall be vented from the high point of the system and H. Average Planar Unear Heat Generation Rate (APLHGR) water flow observed.
During power operation, the APLHGR for each type of fuel ar. a 3. Whenever the HPCI or RCIC System is lined up to take function of axiallocation and average planar exposure shad be suchon from the condensate storage tank, the discharge within limits based on epphcable /J'LHGR limit values which ppng of the HPCI or RCIC shall be vented from the high :
have been approved for the respective fuel and lathce types. point of the system, and water flow observed on a r . ithly l' These values are specir,ed in the Core Operating Limits Report. bases. I If anybme dunng reactor power operation greater than 25% of i rated power it is deter.wned that the hmetmg value for APLHGR is
- 4. h M m W W Ne % W RHR bemg exceedert, acbon shall then be inibated within 15 mmutes S@
gn .m W %@ Ammh to restore operation to within the prescribed hmsts. If the m fullM h W W M APLHGR is not retumed to withm the prescribed limits withm two 1 (2) hours, an orderly reactor power reduction shall be H. Average Planar Linear Heat Generation Rate (APLHGR) coiiniw-ceu immediately. The reactor power snalt be reduced to less than 25% of rated power withm the next four hours, or The APLHGR for each type of fuel as a funchon of average until the APLMGR is retumed to withm the presuibed limits. planar exposure shall be determmed daily dunng reactor operation at >25% rated thermal power.
i l
Amendment No. Aff,6#,7A,86,96,1A19,1 W,132,136 f
- 123
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.a JAFNPP -
i 3.5 (cont d) 45 (cont d) 3 1.
Unear Heat Generation Rate (LHGR) 1.
Linear Heat Generation Rate (LHGR)
The knear heat generation rate (LHGR) of any rod in any fuel The LHGR shall be deterrruned daily dunng reactor operation at ass My at any axiallocahon shall not exceed the maximurn >25% rated thermal power.
l akwable LHGR specified in the Core Operating Linits Report.
1 if anyhme during reactor power operation greater than 25% of rated power it is determned itial the limshng value for LHGR is bemg exceeded, achon sb31 then be ir:stiated wittwn 15 miruAes to restore operation to w'diin the presenbed lirruts. If the LHGR is not retumed to withm the presenbed limsts within two (2) hours, an orderly react x power reduction shall be cuini&Ced ,
- immediately. The rearAor power shall be reduced to less than !
25% of rated power y ithin the next four hours, or until the LHGR is retumed to withe: the presenbed Imts.
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Amendment No. GT6(,74,30$
124
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f JAFNPP -
- . 3,.5 BASES (cont'd) requirements for the emergency diesel generators. the calculated peak clad temperature by less than + 2&F G. r to h M temperature 6 a W W W,me Maintenance of Filled Discharge Pipe, limit on the average linear heat generation rate is sullicsent to i If the discharge piping of the core spray, LPCI, RCIC, and HPCI assure that calculated temperatures are within the 10 CFR 50 are not filled, a water tw.v ner can develop in this pipmg when Appendix K limit. The limdmg values for APLHGR are specelied
- the pump {s) are started. To mmmze damage to the discharge in the Core Operating Limits Report. Dunng Smgle Loop ;
piping and to ensure added margin in the operation of these Operation a multiplier is applied to these values. The derivation i systems, this technical specslication requires the discharge of this muhiplier can be found in Bases 3.5.K Fieference 1.
lines to be filled whenever the system is seguired to be
,, g%gg operabie. If a discharge pipe is not filled, the pumps the syply that line must be assumed to be inoperable for techrucal This specdicahon assures hat the linear heat generation rate in specification purposes. Hca,if a water hammer were to any rod is less then the desegn linear heat generation.
- occur, the system M stNI W " s design A function.
The INGR sher be dedoed daily dunng reactor operation at i H. Average Planar Unear Heat Generation Rate (APLNGR) 25% rated thermal power to determine if fuel bumup, or control 4 This specification assures that the peak claddng temperature m d==gm power dstnbuhon. For i the postulated W h WW W LHGR to be a limitmg value below 25% rated thermal power, will not exceed the limit w__ _u, ,in 10 CFR 50 Appendix K. the ratio of local LNGR to average LHGR would have to be i greater than 10 whs::h is precluded by a conssderable mwgin
- The peak clad 6 y temperJture following a postulated loss-of- when employmg any permissible control rod pattem. ,
coolant accideid is primarily a function of the average heat generation rate of all the rods of a fuel asseii4Ay at any axial location and is only dependsid secondarily on the rod to rod power distribution within an assembly. Since expected local l variations in power distribution within a fuel assembly affect i
i i
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Amendment No. C/,J4, %,96,146,)4 130 I
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v JAFNPP -
5.0 DESIGN FEATURES 5.3 REACTOR PREdSUFtE VF~.dSEL The reactor pressure vessel is as desenbod in Table 42-1 and
., A. The James A. FitzPatnck Nuclear Power Plant is located on the 42-2 d h N N ap W codes are W h
~
PASNY portion of the Nine Mile Pcint site, approximately 3,000 42 m N ft. east of the Nine Mile Poird Nuclear Station, Unit 1. The NPP-5.4 CONTAINMENT JAF sito is on Lake Ontano in Oswego Country, New York, approximately 7 mdes northeast of Oswego. The plant is located at coordinates north 4,819,545.012 m, east 386,968.945 m, on pnmary containment are gben in Table 52-1 of the FSAR.
the N Tre h System.
B. The nearest point on the property line from the reactor budding B. h % N h as Win W 53 and the applicable codes are as described in Section 12.4 of the and any ponts of potential gaseous effluents, with the exception of the lake shoreimo, is located at the northeast comer of the property. This distance is appro-imately 3,200 ft. and is the C. Penetrations of the primery contamment and paping passmg radius of the exclusion areas as defined in 10 CFR 100.3. through such penetrahons are dessgned in accordance with standards set forth in Sect.on 52 of the FSAR.
52 REACTOR 5.5 FUELSTORAGE l A. The reactor core consists of not more than 560 fuel assembhes.
Each Eisso.iniy shall conssst of a matrix of Zircaloy clad fuel rods A. The new fuel storage facshty dessgn criteria are to mantain a K, with an initial composshon of shghtly ennched urarwum dioxide dry <0.90 and flooded <0.95. Comphance shall be venfied prior (UO2) as fuel material. Fuel assembbes shall be limited to those to introduchon of any new fuel desegn to this facshty.
fuel %w approved by the NRC staff for use in BWRs.
B. The reactor core contains 137 crucifomi-shaped control rods as described in Sechon 3.4 of the FSAR.
Amendment No. 36. M. W,04,66,7f,109, Iff 245
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JAFNPP (A) ROUTINE REPORTS (Continued)
- 4. CORE OPERATING UMITS REPORT
- a. Core operating limits shall be established prior to startup from each reload cycle, or prior to any remaininD Portion of a reload cycle for the following:
. The Average Planar Unear Heat Generation Rates (APLHGR) of Specification 3.5.H; e
The Minimum Critical Power Ratio (MCPR) and MCPR low flow adjustment factor, K, , of Specifications 3.1.B and 4.1.E; e The Unear Heat Generation Rate (LHGR) of Specification 3.5.l; e The Reactor Protection System (RPS) APRM flow biased trip satings of Table 3.11; and e
The flow biased APRM and Rod Block Monitor (RBM) rod block settings of Table 3.2 3.
and shall be documented in the Core Operating Umits Report (COLR).
- b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC as described in:
- 1.
- General Electric Standard Application for Reactor Fuel," NEDE-24011 P, latest approved version and amendments.
- 2. " James A. FitzPatrick Nuclear Power Plant SAFER /GESTR LOCA Loss-of Coolant Accident Analysis," NEDC 31317P, October,1986 including latest errata and addenda.
- 3. " Loss-of Coolant Accident Analysis for James A. FitzPatrick Nuclear Power Plant,' NEDO 21662 2, July,1977 including latest errata and addenda,
- c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, ECCS limlts, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met,
- d. The COLR, including any mid-cycle revisions or supplements thereto, shall be provided, upon issuaace for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident inspector.
Amendment No. l 254-c
JAFNPP (THESE PAGES INTENTIONAlt.Y BLANK) i l
Amendment No. 32,110 254 d thru 254-t i
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ATTACHMENT 11 l l
SAFET/ EVALUATION FOR PROPOSED TECHNT5KL EPTCIPicITIDFCITANGES REGARDING REMOVAL OF CYCLE SPECIFIC PARAMETER LIMITS JPTS-88-020 i
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New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 i
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Attachment il SAFETY EVALUATION Page 1 of 10
- 1. DESCRIPTION OF THE PROPOSED CHANGES The proposed changes to the James A. FitzFatrick Technical Specifications are contained in Attachment I and are described below, in addition to these changed pages, all text from the following pages has been either relocated or deleted, and these pages should be removed from the Technical Specifications: Ba,10a,31a,43a,47a,47b,47c,47d,135a, 135b,135c,135d,135e,135f,135g,135h,1351,135j,135k, and 1351.
Page vii, Ust of Figures Figures 3.1 1 and 3.12 are deleted and the pages combined. ,
Figures 3.5 3 through 1.514 are deleted and the pages combined. l Page 6, Specifications 1.0.U.1 and 2 Insert " Minimum value of tho' to the definition of Minimum Critical Power Ratio.
Replace 'as calculated by application of the GEXL correlation (Reference NEDE 10958)* j with, "for all fuel assemblies in the core.'
Delete "The design LHGR is 14.4 KW/ft for GE8x8EB fuel and 13.4 KW/ft for the remainder."
Page 6a, Specifications 1.0.AD A new definitionis added to read:
AD. Core Operating Umits Report (COLR)
This report is the plant specific document that provides the core operating limits for the current operating cycle. These cycle specific operating limits shall be determined for each reload cycle in accordance with Specification ,
6.9.A.4. Plant operation within these operating limits. Is addressed in individual Technical Specifications, i
Page 8, Specification 2.1.A.1.c.(1)
The scram trip setting formula and remainder of the column are replaced with, 'less than I or equal to the lims Soecmed in Table 3.11. This setting shall be adjusted durirag sing %
loop operation when requkad by Specification 3.5.J.*
Page 8a, Specification 2.1.A.1.c.(1) (cont'd)
Relocate this specification to page 8 and remove this page.
Page 9, Specification 2.1.A.1.c.(1) (cont'd)
Delete this specification in its entirety.
i t
4 e - - - - . , , - , -, , , , , , ,- .- r, ,,r- --e ,,
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Attachmoni11 SAFETY EVALUATION Page 2 of 10 P_aoe 10, Specification 2.1.A.1.d The Rod Block trip setting formula and remainder of the column are replaced with, "less than or equal 19 the limit specified in Table 3.2 3. This setting shall be adjusted during single loop operation when required by Specification 3.5.J.*
This specification is relocated to page 9 and page 10 is now intentionally blank.
Page 12, Bases 1.1 In the second sentence, delete 'such that the* and end this sentence after *(MCPR).*
In the third sentence replace 'MCPR > 1.04* with "This Safety Umit.'
Bases 1.1.A Replace *(MCPR of 1.04)' with *MCPR.*
Replace 'MCPR operating conditions in specification 3.1.B' with, "MCPR operating limit in the Core Operating Umits Report."
Page 13, Bases 1.1.A FIRST PARAGRAPH Replace "at the beginning of each fuel cycle
- with, 'in Reference 1.*
FOURTH PARAGRAPH i
Replace 'to a maximum LHGR of 14.4 KW/ft for GE8x8EB fuel and 13.4 KW/ft for the '
remainder' with, 'by the maximum LHGR identified in the Core Operating Umits Report."
FIFTH PARAGRAPH Insert an additional *0" into *1.0" to make the number of significant figures consistent with the specifications.
Replace ' required in specifications 2.1.A.1.c and 2.1.A.1,d" with, "specified in Tables 3.1 1 and 3.2 3 respectively."
Page 14, Bases 1.1.E Replace Reference 1. with, ' General Electric Standard Application for Reactor Fuel, NEDE 24011 P. latest approved revision and amendments '
Delete Reference 3.
Page 15. Bases 2.1 FIRST PARI. GRAPH Replace *2535' with "of 2436.*
i insert *MWt* after *2436.*
THIRD PARAGRAPH Insert " applicable" before " operating Umit."
Attachment 11 SAFETY EVALUATION Page 3 of 10 Replace 'MCPR's" with *MCPR.*
Replace
- Specification 3.1.B" with,7he Core Operating Umits Reoort COLR.* ,
FOURTH PARAGRAPH Replace
- Specification 3.1.B' with, "the Core Operating Umits Report."
FIFTH PARAGRAPH Replace 'a' with "the,' replace ' program
- with " programs," and replace " References 1 and 3" with " Reference 2.*
Paos 16, Bases 2.1 (cont'd)
FIRST PARAGRAPH Replace 'of specification 3.1.B' with. *in t ". 70LR.*
THIRD PARAGRAPH Replace "will not be' with *is not.*
In subparagraph (i), replace 'a power level of 2535 MWl' with, "the licensed maximum power level."
Replace the roman numerals of subparagraphs I through Iv with bullets (.).
Pace 18, Bases 2.1.A.1.c THIRD PARAGRAPH Replace 'in accordance with the formula in Specification 2.1.A.1.c" with, 'as specified in Table 3.1 1."
FOURTH PARAGRAPH i Replace 'provided in Specification 3.1.B' with, *specified in the Core Operating Umits Report.'
Page 20, Bases 2.1.C Delete References 1. and 3.
In Reference 2, replace ' Fuel Application
- with, " Standard Application for Reactor Fuel."
Page 30f, Specification 3.1.B and 3.1.B.1 Replace "shown below:' am she subsequent specification 3.1.B.1 with, "specified in the Core Operating Umits Report."
Specification 4.1.B Replace " required by Specifications 2.1.A.1.c and 2.1.A.1.d, respectively' .Ath "specified in the Core Operating Umits Report.'
]
I Page 31, Specification 3.1 B.1 (cont'd) and 3.1.B.2 Delete these specifications in their entirety.
Attachment 11 ;
SAFETY EVALUATION i Page 4 of 10 l 4.1.E Rep'sce Elis specification and all subparagraphs with, " Verification of the MCPR operating limits Ghall be performed as specified in the Core Operating Umits Report."
Page 31a, Spec!fications 3.1.B.2 (cont'd),3.1.B.3,4 and 5, and 4.1.E.3 Deleted the note associatcd with Specification 3.1.B.2.
3.1.B.3 l Delete this spocification in its entirety. -
3.1.B.4 Replace "shown in Figure 3.1.1* with 'specified in the Core Operating Umits Report."
Renumber this speelfication 3.1.B.1 and relocate to page 30f.
3.1.B.5 Replace ' limiting value for" with, ' operating Umit."
Renumber this specificatica 3.1.B.2 and relocate to page 31.
4.1.E.3 Delete this specification in its entirety.
Remove page 31a from the specifications.
Page 35, Bases 3.1.B >
Replace 'given in Specification 3.1.B' with, "specified in the Core Operating Umits Report ,
(COLR)." '
Replace "given in Specification 3.1.B' with, *specified in the COLR.*
Page 41, Table 3.1 1 Replace the Trip Level Setting formula for the APRM Flow Refnrenced Neutron Flux Scram -
with *(12)."
Delete the reference to notes (12) and (17) from the Trip Function for the APRM Flow Referenced Neutron Flux Scram.
1 Page 43, Notes of Table 3.1 1 (cont'd)
Replace note 12 with the following:
- 12. The APRM Flow Referenced Neutron Flux Scram sett:ng shall be less than or equal to the limit specified in the Core Operating Umits Report. <
In note 13, replace 'in accordance with Specification 2.1.A.1.c" with "as specified in the Core Operating Umits Report."
Page 43a, Notes of Table 3.1 1 (cont'd)
Delete note 17 in its entirety.
o * *-
Attachment 11 d SAFETY EVALUATION ;
Page 5 of 10 I
Relocate notes 14 through 16 to, page 43 and remove page 43a from the Technical ;
Specifications. l Pages 47a,47b,47c and 47d, Figures 3.1 1 and 3.12 Delete both figures and remove these four pages.
Page 72, Table 3.2-3 f
Replace the Trip level Setting formulas for both Flow Blased APRM Upscale and Rod "
Block Monitor Control Rod Blocks with, *(8)."
i Page 73, Notes for Table 3.2 3 Replace note 8 with the following:
- 8. The Flow Blased APRM Upscale and Rod Block Monitor trip level setpoint shall be less than or equal to the limit specified in the Core Operating Umits Report.
Page 74, Notes for Table 3.2 3 (cont'd)
Delete notes 11 and 12 in the r entirety, l
Remove the headings from this page and insert "This Page Intentionally Blank." ;
Page 102, 3.3 and 4.3 Bases,6B.5 THIRD PARAGRAPH Replace 'i.e., MCPR limits as shown in Specification 3.1.B' with, "e.g., MCPR limit.'
Page 123, Specification 3.5.H Replace the'second and third sentences with, "These values are specified in the Core Operating Umits Report.'
l In specifications 4.5.G.2 and 3 on this page, restore the Amendment 132 changes inadvertently deleted by Amendment 134.
Page 124, Specification 3.5.1 Replace "of 14.4 KW/ft for GE8x8EB fuel and 13.4 KW/ft for the remainder of the fuel" with, 'specified in the Core Operating Umits Report.' '
Specification 4.5.1 Replace " checked
- with " determined" to accurately reflect that the LHGR is a calculated value, not an instrument reading.
Page 130, Bases 3.5.H SECOND PARAGRAPH
Attachment 11 SAFETY EVALUATION Page 6 of 10 in the third sentenos, replace *given in Figures 3.511 through 3.514" with, *specified in the Core Operating Umits Report."
Delete the fourth and fifth sentences in their entirety.
Replace the sixth sentence through the word *during" with, *A multip!!er is applied to these values during.'
Page 135a through 1351, Figures 3.5-3 throuch 3.514 Delete these figures and remove these twelve pages.
Page 245, Specification 5.2.A Delete the second sentence through the end of this Specification. In their place insert,
- Each assembly shall consist at a matrix of Zircaloy clad fuel rods with an initial composition of slightly enriched uranium dioxide (UO2 ) as fuel material. Fuel assemblies shall be limited to those fuel designs approved by the NRC staff for use in BWRs."
Page 254-c, Specification 6.9.A.4 Insert e new Specification 6.9.A.4 on a new page 254 c. The text of this specification is given in Attachment 1.
Page 254 c thru 254 f Renumber this page *254 d thru P54 f" to support the change described above, ll. PURPOSE OF THE PROPOSED CHANGES The purpose of the proposed Technical Spacification changes is to remove cycle specific parameter limits in accordance with the guidance provided by the NRC in Generic Letter 8816 (Reference 1). Use of the Generic 1.stter 8816 alternative consists of three separate ;
actions to modify the Technical Specifications: '
- 1) The addition of a definition of a formal report that ircludes the values of cycle-speelfic parameter limits that have been established using an NRC-approved methodology and consistent with all applicable limits of the safety analysis. At FitzPatrick, the report will be titled, " Core Operating Umits Report."
- 2) The addition of an administrative reporting requirement to submit the Core Operating Umit Report to the NRC for information.
- 3) The modification of individual Technical Specifc' ations to note that cycle specific parameters shall be maintained within the limits provided in the Core Operating ;
Umits Report. '
The proposed Technical Specification changes are responsive to industry and NRC efforts !
to improve Technical Specifications, reduce the administrative burden on the NRC and the !
New York Power Authority, and permit future reloads to be accomplished without license l
amendments. The proposed changes are consistent with those discussed previously
~
I-Attachment ll BAFETY EVALUATION Page 7 of 10 between the NRC and General Electric Co. as described in Reference 2 and telephone conversations between the Authority and NRC staff.
The following Technical Specification parameters have been identified as cycle specific limits that can be relocated to the Core Operating Umits Report:
- 1) Operating Umit Minimum Critical Power Ratio (MCPR);
- 2) Flow Dependent MCPR Umits;
- 3) Maximum Average Planar Unear Heat Generation Rate (MAPLHGR);
- 4) Unear Heat Generation Rate (LHGR); and
In addition, discussions contained in the Technical Specification Bases associated with the above parameters which are cycle specific are modified in accordance with the guidance of Generic Letter 8816.
The Authority is implementing these Generic Letter 8816 changes during the current Reload 9/ Cycle 10 refueling outage. The Authority is preparing a Core Operating Umits Report (COLR) to support the reloaded core. The Cycle 10 COLR will be provided to the NRC upon issuance, but no later than at the startup of Cycle 10 as required by the proposed Technical Specifications. Cycle 10 wl'l commence with the Cycle 10 speelfic limits in a PORC and SRC reviewed COLR. At no time will the core be operated without the cycle-specific limits in either the Technical Specifications or the COLR.
As part of this Technical Specification amendment, an addnional change is also proposed.
The bases for Specification 2.1 on pages 15 and 16 state that the abnormal operational transients were analyzed at a power of 2535 MWt, corresponding to 104 percent of the licensed maximum power level of 2436 MWt. However, the NRC has approved the GE transient analysis methods designated GEMINI methods, which use the nomhal (100%)
power level in transient analyses. Consequently, the Bases to Specification 2.1 are modified to state that transient analyses are performed at 100 percent power (the maximum licensed power level) consistent with the NRC approval given in Reference 3.
This method of transient analysis was approved for FitzPatrick Cycle 8 operation in Amendment 109 to the Technical Specifications (Reference 4).
Additional changes are being proposed as requested by the NRC staff. The APRM flow biased scram and rod block setpoint formulas are being deleted from Specifications 2.1.A.1.c and 2.1.A.1.d (on pages 8,9, and 10) and the associated Bases. These formulas will be replaced with references to Tables 3.11 and 3.2 3 respectively. Table 3.11 '
provides the requirements for reactor protection system instrumentation which includes the APRM flow biased scram. Table 3.2 3 provides the requirements for control rod block instrumentation which includes the APRM flow biased rod block. These tables are also being revised to relocate these formulas into the COLR. In addition, Bases 1.1 and 1.1.A '
on page 12 are revised to delete the numeric value of the Safety Umit MCPR. This Safety Umit value is contained in Specification 1.1.A on page 7 and need not be repeated in the Bases. '
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I
' Attachment 11 !
SAFETY EVALUATION Page 8 of 10 111. IMPACT OF THE PROPOSED CHANGES' A. Generic Letter 8816 Changes The current method of controlling reactor physics parameters to assure conformance with 10 CFR 50.3G is to specify the values determined to be within specified acceptance criteria, usually the limits of the safety analysis, using an approved calculation ,
methodology. The proposed Technical Specification changes maintain control of the values of cycle specific parameters and assure conformance to _10 CFR 50.36 by specifying the approved calculation methodology and approved acceptance criteria. The Core Operating Umits Report documents the specific values of parameter limits that are determined using these methods and that meet the acceptance criteria. The Technical ,
Specifications continue to require that operation will remain withln limits, and that required . ;
remedial actions are taken if the limits are not met. ,
The Core Operating Umits Report for each cycle, and any necessary mid cycle revisions, l will be provided to the NRC for information. This report, the associated Nuclear Safety ;
Evaluation, and all revisions thereto will be reviewed by both PORC and SRC in !
accordance with Technical Specifications 6.5.1(E) and 6.5.2.7. This will provide a similar i level of quality assurance and document control for the Core Operating Umits Report as i a for the Technical Specifications. This will ensure that the proper operating limits are being enforced. ;
B. Technical Specification Bases 2.1 Change This change updates the Bases to reflect the power level used in the FitzPatrick transient analyses. With the introduction of the approved GEMINI methods, transient analyses are l performed at the 100 percent power level. Previously, analyses were performed at a >
- power level in excess of 100 percent to account for uncertainties in power level measurement as required by Regulatory Guide 1.49. However, with GEMINI methods,
, power level measurement uncertainty is accounted for instead by increasing the MCPR calculated with the GEMINI methods instead of the posver level as used previously. The NRC has generically approved this method of accounting for power level measurement i uncertainty in Reference 3 and has approved its use at FitzPatrick in Reference 4.
IV. EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION Operation of the FitzPatnok plant in accordance with the proposed Amendment would not involve a significant hazards consideration as defined in 10 CFR 50.92 since it would not:
l 1. Involve a algnificant increase in the probability or consequences of an ,
l accident previously evaluated.
I A. Generic 1.stter 8816 Changes:
The proposed amendment merely moves cycle specific parameter limits from l
the Technical Specifications to the Core Operating Umits Report. NRC ,
approved methodologies will continue to be used as the basis for establishing those limits. The establishment of these limits in accordance with NRC-
- approved methodology and the incorporation of these results into the Core Operating Umits Report will ensure that proper steps have been taken to :
establish the values of these limits. Furthermore, the submittal of the Core
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Attachment il SAFETY EVALUATION ;
Page 9 of 10 I Operating Umits Report to the NRC will allow the staff to continue to trend the values of these limits.
B. Technical Specification Bases 2.1 Change-The use of 100 percent power in the analysis of abnormal operational transients using GEMINI methods has been reviewed and approved previously a by the NRC for both generic and FitzPatrick specific application (see- 1 References 3 and 4). Power level measurement uncertainties are accounted ,
for adequately in the MCPR Operating Umit, and the level of confidence that i the MCPR Safety Umit will not be violated as a result of a transient is not j 4 reduced. j
- 2. create the poselbility of a new or different kind of accident from any i accident previously evaluated. 1 No safety related equipment, function, or plant operation will be aftered as a result of the proposed changes. The changes do not create any new accident mode. The level of document control and quality assurance app!!ed to the q preparation and use of the Core Operating Umits Report will be equivalent to ;
that applied to Technical Specifications.
l . 3. Involve a significant reduction in a margin of safety. i j A. Generic Letter 8816 Changes:
The proposed changes are administrative in nature and do not impact the .c operation of the plant in a manner that will reduce the margin of safety. The proposed amendment still requires operation within the limits determined using NRC approved methods, and that appropriate remedial actions be taken . i; if the limits are violated. ,
l B. Technical Specification Bases 2.1 Change:
' (
The MCPR Operating Umit continues to be determined using an approved i methodology that conservatively accounts for power level measurement uncertaintles. The same criterion for acceptable operation is ma'ntained; that i
is,99.9 percent of all fuel rods will not enter bolling transition in the _ event of the limiting transient. Therefore, the margin of safety is not reduced.
i- V. IMPLEMENTATION OF THE PROPOSED CHANGE Implementation of the proposed changes will not impact the ALARA or Fire Protection l Programs at the FitzPatrick plant, nor will the changes impact the environment.
- VI. CONCLUSION i The change, as proposed, does not constitute an unreviewed safety question as defined in
- 10 CFR 50.59. That is, it: -
! a. Will not change the prooability nor the consequences of an accident or malfunction of '
l equipment important to safety as previously evaluated in the Safety Analysis Report; k
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,, a e SAFETY EVALUATION Page 10 of 10
- b. will not increase the possibility of an accid 1mt or malfunction of a different type from any previously evaluated in the Safety Analysis Report; c, will not reduce the margin of safety as defined in the basis for any technical
. specification; and
- d. Involves no significant hazards consideration, as defined in 10 CFR 50.92.
Vll. REFERENCES
- 1. NRC Generic Letter 8818, " Removal of Cycle-Specific Parameter Umits from Technical Specifications," dated October 4,1988.-
- 2. GE letter, J. S. Charnley to M. W. Hodges (NRC), " Acceptance implementation of Generic Letter 8816,* dated August 8,1989.
- 3. NRC letter, G. C.1.ainas to J. S. Chamley (GE), " Acceptance for. Referencing of Ucensing Topical Repo:t NEDE 24011 P A, 'GE Generic Ucensing . Reload Report,' Supplement to Amendment 11,* dated March 22,1986.
. 4. . NRC letter, H.1. Abelson to J. C. Brons (NYPA), " Amendment 109 to Technical -
Specifications," dated April 3,1987.
_ 5. James A. FitzPatrick Nuclear Power. Plant Updated Final Safety Analysic-Report.-
. 6. James A. FitzPatrick Nuclear Power Plant Safety Evaluation' Report (SER),
dated November 20,1972, and Supplements.
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