ML20024G442
| ML20024G442 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 03/12/1975 |
| From: | NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20024G439 | List: |
| References | |
| NUDOCS 9102120469 | |
| Download: ML20024G442 (25) | |
Text
{{#Wiki_filter:- D. Irrned tate - Irrnediate means that the required acti ,n wilI be initiated as soon as practicable censidering the safe operati<m of the unit and the importance of the required action. ' F Instrument Puretional Test - An instrument functienal test means the injection of a sirulated signal into the primary sensor to verify the proper in-trument channel response, alam, and/or initiating action. F. Instrument Calibration - An instrument calibration means the adjustment of an instrument signal output so that it corresponds, within acceptable range. accuracy, end response time to a known value (s) of the parameter which the instrument menitors. Calibration shall encompass the entire instrument including actuatien, alarm or trip. Ranpore time is not part of the routine instrument calibration but will be checked once per cycle. G. Limiting Conditions for Operation (LCO) - The limiting conditions for operation specify the minimum acceptable levels of system perfomance necessary to assure safe startup and operation of the facility. When these conditicas are met, the plant can be operated safely and abnomal situat ans can be safely controlled. H. Limiting Safety System Setting (ISSS) - The limiting sefety system settings are settings on instru-mantation which initiate the automatic pro'.ective action at a level such that the safety limits will not be exceeded. The region between the safety limit and these settings represents margin with normal
~
operation lying below these settings. e margin has been established so that with proper operation of the instrumentation, the safety limits will never be exceeded.
- 1. Minimum Critical Power Ratio (MCPR) - The minimum critical power ratio is the value of critical power ratio associated with the most limiting assembly in the reactor core. Critical power ratio (CPR) is the ratio of that power in a fuel assembly which is calculated by the GEZL correlation to cause some point in the assembly to experience boiling transition to the actual asse-bly operating power.
J. Mode - The reactor mode is that which is established by the mode-selector switch. K. Operable - A system or component shall be considered operable when it is capable of perfoming its intended function in its required manner. L. Operating - Operating means that a system or component is performing its required functions in its required manner. M. Operating Cycle - Interval between the end of one refueling outage and the end of the next subsequent refueling outage. 2 1.0 REV Q2120469750312 p ADOCK 0500C,263 PDR_ .-
2.0 SAIT.TY LTSIITS LI?fiTING SAFET. SYSTDi SETTINGS 2 I Fig'L CIADDII!G INTJZ:"ITY 2.3 JLTL CLADDING INTEGRITY Appijenbility: t3pplicability: Applies to the interrelated variables App 1!cs to trip settings of the instruments and associated with fual themal behavior, devices which are provided to prevent the reactor system safety limits frcu being exceeded. Objective: Objective: To define the level of the process variables To establish limits belev which the at which automatic protective action is integrity of the fuel cladding is preserved. initiated to prevent the safety limits from bcIng exceeded. Spec lfication: Specifiention: Core Thermal Pwer Limit (Reactor The finiting safety system settings shall be as A. specified belev: Pressure > 800 Psin and (bre Flow is
> 107. of Rated)
A. Neutron Flux Scram When the reactor pressure is > 800 Psia APRM -- The APRM flux scram trip setting and core flow is > 107. of rated, the 1.
< xisteece of a minimum critical power shall be as shown in Figure 2.3.1 unless the combination of power and peak heat t atio (!'CPR) less than 1.06 shall cen- flux is above the applicable curve in Figu c stitute violation of the fuel cladding integrity safetf limit. 2.3.2. When the combination of power and reak heat flux is above the curve in Figure 2.3.2, the scrm setting (S) is given by:
l 6 2.1/2.3 REV
j' - l
- f
, 2.0 SAFETY LIMITS LIMITING SnrETY SYSTEM SETTINGS l ' 1 i 4 S = 486,000 P (7x7 fuel) i i X i S = 425,000 P (8x8 fuel) I B. Core Thermal Pcwcr Limit (Reactor X ! Where: Presnure di 800 Psia or Core i Flow -~< 107. of Rated) P = perr.ent of rated power
- X = pe9k heat flux - (grU/IIR/Fr ) 2 :
When the reactor pressure is $s 800 Psia shall be used. ' i or core flow is < IU7. of rated, the ~~ i core thernal power shnl1 not exceed 257. l 2. IRM--Flux Scrari setting shall be CE 20% I of rnted thermal power. of rated neutron flux.. B. APRM Rod Block - The APRM rod block setting shall be as shown in Figure 2.3.1 un'ess the
- combination of power and peak heat flux is !
C. Power Transients above the applicable curve in Figure 2.3.2. When the combination of power and peak flux is t To insure that the safety ILnit established nbove the curve in Figure 2.3.2, the rod block trip in Specification 2.1.A is not exceeded, setting (RB) is given by: i each required scram shall be initiated by ~ l its primary source signal as indicated by n3 - 437.400 P (7x7 fuel) , i the plant process computer. X ' RB = 382,400 P (8x8 fuel)-
.X 4
i vhere: P.= percent of rated power X = peak heat flux (BIV/PR/FT 2) 4 shall be used. , i ', 2.1/2.3 c. Reactor Low Water Level Scram setting shall t be 2i 10'6 eboze the top .of the active fuel. i 7
+
2.0 SAFFTY LIMITS LIMIT"!G .FETY SYSTD! SETTING D. Reactor Vater Invei (3hutdown Condition) l D. Tanctor Lw Lcw Water Level ECCS initiation shall be 1 6'6" .5 6'10" above the tcp of Witenever the reactor is in the shutdwn t he active fuel. condition with irraciated fuel in the reactor vessel, the water level shall not be less than that corresponding to 12 inches above the top of the active fuel when it is seated in the core. This level sh.11. be con-tInueusly monitored whe -c.er the recirculation pinnps are not eperstit4
- 5. Turbine Control Vol'e Fast Closure Scram i hall initiate upt,a loss of pressure at the veccleration relay with turbine first st. age Pressure > 307..
F. Turbine Step Valve Scram shall be S 107.
; valve closure from full open with turbine i first stage pressure 2 307..
G. Itain Steamline Isolation Valve Closure c.czam shall be $ 107. valve closure from rull open.
, 2.1/2.3 8 i
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4 n ases: 2.1 The fuel c1 Mding integrity limit is set such that no cal < ulated fuct der:: age would occur as a result of n aburrmal operatictal transient. Because fuel da tage is net directly observable, a step-back approach i. i used to establish a Safety Limit such that the MCPR !? no lesc than 1.06. MC7R> 1.06 represents a con-servative margin relative to the con (itions .cquired te maintain fuel cladding integrity. ne fuel cimidi g is one of the physical barricts which separate radioactive materials from the environs. The integrity C this cladding barrier is related to its reistive freetem from perforations or cracking. Although some corresion or use related cracking nay occur during the life of the cledding, fission product migration t :, this source is incrementally curulative and continuous ly rwasurable. Puel cladding perforations, however. can result frem themal stresses which occur froci re,ctor operation significantly above design conditio~ and the protection system safety settings. While fission product migration from cladding perforation le just as measurabic as that from use related cracking, the themally caused cladding perforations signal 4 threshold, beyond which still greater thermal stresser may cause gross rather than increnental cladding deterioration. Werefore, the fuel cladding Safety Limit is defined with margin to the conditions which wculd produce onset of transition boiling. (MCPR of 1.0). These conditions represent a significant departure fro:a the condition intended by design for planned operation. The concept of !"2R, as used in the G" TAB /GEXL critical power analysis, is discussed in Reference 1. A. Core Themal Pcwer Limit (Reactor Pressure > 800 psia and Core Flow > 107. of Rated.) Onset of transition boiling results in a decrease in heat transfer frem the clad and, therefore, elevated clad te=peratere and the possibility of clad failure. H wever, the existence of critical power, or boiling transitica, is not a directly observable parameter in an operating reactor. Therefore, the margin to boilins: transition is calculated from plant operating parameters such as core power, core ficw, feedvater temperature, and ccre power distribution. The margin for each fuct assembly is characterized by the critical power eatio (CPR) which is the ratio of the bundle power d ich would nreduce onset of j transition boiling divided by the actual bundle pruer. The minimum value of thit ratio for any burdle in the core is the mininum critical power ratio (?*CPR). It is assu wd that the plant operation is controlled to the nominal protective setpoints via the instrumented variables. Tha Safety Limit (T.S.2.1.A) has sufficient conservaticm to assure that in the event of an abnomat operational transient initiated from the Operating MCPR Limit (T.S.3.11.C) note than 99.97. of the fuel - rmis in the core re expected to avoid boiling transition. The : argin between MCPR of 1.0 (onset
^
2.1 BASES 13 RE"
1 Joses Centinued: l cf transition boiling) and the Safety Linit is derived frc t a detailed statistical analysis consider-l ing all of the uncertatntics in e enitorine; the core op. rating state including uncertainty in the bot iing t ransition correlati r as described in Rt *rc~ e 1. The u- certainties creleyed in deriving the Safety Lirit are provided at the beginning of ench fuel cycle. Because the boiling . ranai cion correlation is based on , large quantity of full scale data, there is a very high confiden e that creration of , fuel assenb!y at the MCPR Safety Linit would not produce boiling transition. Thus, although it is not required to establish the Safety Limit, additional marr '.n exists between the Safety Limit and the actual occurrc ce of loss of cladding integrity. IIcuever, if boiling transition were to occur, clad perforation would not be expected. Claddirg te peratures would increase to approximately 1100 r which is belc-: the perforation temperature of th? cladding material. This has been verified by teste in the General Electric Test Reactor (GETR) whera fuci sinitar in design to Menticello operated abova the boiling transfrion for a significant period of time (30 ninutes) without cled perforation. If reactor pressure should ever exceed 1400 psia during normal power operating (the limit of applicability of the boiling transition correlation) it vould be assumed that the fuel cladding integrity Safety Limit has been violated. In addition to the 1TPR Safety Limit, operation is constrained to a maxinn IRCR of 17.5 kw/ft for 7x7 fuel and 13.4 kv/ft for 8x8 fuel. At 100% pcwer this limit is reached with a maximum total peaking factor of 3.08 for 7x7 fuel or 3.04 for 8x8 fuel. For the case of the maxinn total peaking
- factor exceeding design, operation is permitted only at less than 100% of rated thermal power and only with reduced AFRM scram and rod block settings as required by specifications 2.3.A.1 and 2.3.B.
B. Core Thernal Power Limit (Reactor Pressure 6 800 psia or Core F1cw 6107, of Rated) At pressure below C00 psia, the core elevation pressure drop (0 pcver, O flow) is greater than 4.56 psi. At low powers and all core flows, this pressure differentici is maintained in the bypass region of the core. 2.I BASES 14 REY
i
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i 1 t TAses Continued: Since the ;,ressure drop in thn bypass region is esrentially all elevation head, the core pressure drop at Icw powers and all fi ns vill always be greater than 4.56 psi. Analyses shcw that with a ' bundle ficw of 28x103 lbs/hr, bundle pressure drop is raarly independent of bundle power and has a l 28x10 3 alue of 3.5 psi. Therefore, due to the 4.56 psi driving hesd, the bundle flow will be greater tha 2 , lbs/hr irrespective of total core flow and independent of bundle power for the range of of bundle powers of concern. Full scale ATIAS test data taken at pressures from 14.7 psia to 800 .{ ' psia indicate that the fuel assembly critical pcver at 28x103 lbs/hr is approximately 3.35 m't. With the design peaking factors this corresponds to a core themal pcuer of more than 50"4. Thus, a core themal power limit of 25% for reactor pressures below 800 psia or core flow less than IfT/. is conservative. , t l j C. i Power Transient Plant safety analyses have shown that the scramr initiated by exceeding safety system setting will assure that the Safety Limit of 2.1. A or 2.1.B vill not be exceeded. Control red scram times and safety svstem settings are checked periodically to assure that a scram will proceed as analyzed. As a further check, the plant process computer will be used as a fast data- , acquisition system, when available during a scram, to verify that the scram was initiated by the prin.ary source signal. The corquter is nomally availabic for this function. However, it is i recogniecd that the plant may operate without the computer in service, in which event the con- ' fimatory data vill not be available and the vertification specified by 2.1.C will not be required. 4 The thermal pcver transient resulting when a scram is accomplished other than by the expected scram signal (e.g., scram from neutron flux following clcsure of the main turbine stop ralves) does not necessarily cause fuel damage. For this specification, when a scram is only acccuplished by means of a backup feature of the plant design, a specific analysis is required to detemine whether or ' not a Safety Limit has been violated. The concept of net approaching a Safety T.intt, providing scram signals are operable, is supported by the extensive plant safety analysis. t l D. Reactor Vater Level (Shutdown Condition) During periods when the reactor is shut down, consideration ! nust also be given to water level requirements due to the effect of decay heat. If reactor water level shocid drop belcv 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 vill be cooled sufficiently to prevent clad melting should the water level be reduced to two-thirds the core height. Establishment of the safety limit at 12 tnches above the top af the fuel provides adequate margin. This level will be continuously monitored when-l ever the recirculation pumps are not operating. l 2.1 BASE 3 15 '
4 1~ r I'.-3 I l [. i i i Banas Centitmed: I i , _R_. ? erences i i 1. . General Electric BWR Thert al Analysis Basis (CETAP'. Dat a, Correlatian and Design Applicatien,- 1 l NTM 1093'3. l i-i l l! l 2.1 BASES 16 { FEV . l _ _ _ _ . . _ , , . . - - . - - . - - - - _ , ~ . - - , . - - - - - ---- - - - - - - - - - - - - - - - - - ~-------J
I i Bases: 2.3 The abnormal operational transients applicable to oper, tion of the MonMccllo Unit have been analyzed throughout the spectrs= of planned operating conditior up to the the.rs.1ptver icvel of 1670 N t. The analyses were based upon plant operation in accordance with the ope m :c map given in Figure 3-2-3 of the FSAR. The licensed maximum pcuer level 1670 Nt represents the r a.-irei steady-state power which shaii not knowingly be exceeded. Conservatism is incorporated in the transient analyses in estimating the controlling factors, such as void reactivity coefficient, control rod scram worth, scra:- delay tune, peaking factors, and axial power shapes. These factors are selected conservatively with respect to their effect on the applicable transient results as determined by the current analysis model. This transient model, evolved over many years, has been substantiated in operation as a conservative tool for evaluating reactor dynamic per-formance. Results obtained from a General Electric boiling water reactor have been compared with predictions nde by the model. The cor:parisons and requits are summarized in Reference 1. The absolute value of the void reactivity coefficient used in the analysis is conservatively estimated to be about 237. greater than the nominal maximum value expected to occur during the core lifetime. The Doppler reacti-ity feedback coefficient has conservatively been derated to 907. of the expected value. ! The scram worth used has been derated to be equivalent to approximately 807. of the total scram worth of the control rods. The scram delay time and rate of rod insertion assumed by the analyses are conservatively set equal to the longest delay and slowest insertion rate acceptable by Technical Specifications. The effect of scram worth, scram delay time and rod insertion rate, all conservatively applied, are of greatest significance in the early portion of the negative reactivity insertion. The rapid insertion of I negative reactivity is assured by the time requirements for 57. and 207. insertion. The early portion of } the scram stroke accomplishes the desired effect by inserting sufficient negative reactivity to turn the transient around. The times for 507. and 907. insertion are given to assure proper completion of the s expected performance in the earlier pcrtion of the transient, and to establish tAe ultimate fully shutdown steady-state condition. 4 t 2.3 MSES 17 REV
B ues Centinued: i For snalyses of the thermal conscquences of the transients , the Operating MCPR Limit (T.S.3.11.C) is conservatively assumed to exist prior to initiation of the transients. This choice of using conservative values of controlling parameters and initiating transients at the desien ! pcver level, produces more pessimistic ansvers than vc,1d result by using expected values of control l parameters and analyzing at higher power levels. Deviations froct as-left settings of setpoints are expected due to inherent instrument error, operator setting error, drift of the setpoint, etc. A11cvable deviations are assigned to the limiting safety system settings for this reason. The effect of settings being at their allowable deviation extreme is minimal with respect to that of the conservatisms discucsed above. Although the operator vill set the setpoints within the trip settings specified, the actuni values of the various setpoints can vary frca the specified trip setting by the ellowable devirtion. ' A violation of this specification is assumed to occur only when a device is knowingly set outside of the limiting trip setting or when a sufficient number of devices have been affected by any means such that the automatic function is incapable of preventing a safety li=it from being exceeded while in a reactor mode in which the specified function rust be operable. Sections 3.1 and 3.2 list the reactor modes in which the functions listed above are required. A. Neutron Flux Scram The average power range monitoring (APRM) system, which is calibrated using beat balance data taken during steady state conditions, reads in percent of rated thermal power (1670 N t). i Because fission chambers provide the basic luput signals, the AFEM systesa responds directly to averste neutron flux. During transients, the instantaneous rate of heat transfer fram the fuel (reactor thermal pcver) is less than the instantaneous neutron f 2 due to the time constant of the fuel. Therefore, during abnormal operational transients, the thernal pcwer of the frel will be less than 2.3 FASES : IB l FE ' e I
I i h i i Bases Continued: ' that indicated by the neutron flux at the scram setting. Analyses demonstrate that, with a 1207. scram trip setting, none of the abnormal operation,1 transients analyzed violate the fuel Safety ' Lirrit and there is a substantial margin from fuel ', mage. Therefore, the use of flow referenced scram trip provides even additional nargin. An increase in the APRM scram trip setting would decrease the margin present before the fuel claddin~ integrity Safety Limit is reached. The APRM scr.r trip setting was determined 5 r an snalysis of margins raquired to provide a reasonable range for maneuvering during operation. Reducing this operating margin would increase the frequency of murious scrams which have an adverse effect on reactor safety because of the resulting thermal stresses. ""hus , the A-PFli scram trip setting wa., selected because it provides adequate margin for the fuel cladding integrity Safety Limit yet ' allows operating margin that reduces the possibility of unnecessary scrams. Therefore, it is intended to ultimately replace (with prior NRC approval) the automatic flow referenced scram with a fixed 120 percent scram setting. ( The scram trip setting must be adjusted to ensure that the UIGR transient peak is not increased for any combination of maximum total peaking factor an3 reactor core thett:a1 power. The scram setting ; is adjusted in accordance with the formula in Specification 2.3.A.I.when the maximum total pe sking factor is greater than design. If the APRM scram setting should require a change due to an i abnormal peaking condition, it will be done by increasing the APRM gain and thus reducing the slope and intercept point of the flow referenced scram curve by the reciprocal of the APRM gain t change. Analyses of the limiting transients show that no scram adjustment is required to assure that the MCPR Safety Limit (T.S.2.1.A) is not exceeded when the transient is initiated fract the ; Operating MCPR Limit (T.S.3.11.C). . 4 For operstion in the startup mode while the reacter is at low pressure, the IRM scram setting of [ 207. of rated power provides adequate thermal margin between the setpoint and the safety limit, 257. of rated. The margin is adequate to accommodate anticipated maneuvers associated with p wer plant startup. Effects of increasing pressure at zero or low void content are minor, cold water from sources available during startup is not much colder than that aircady in the syscem, i 2.3 BASES 19 RW
D es ccntinued: terperature coefficients are mall, and control re3 pot terns are constrained to be uniform by oparating procodures backed up by the rod worth einimirei. Worth of individual rods is very Icv in a unifom rod pattern. Thus, of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant pmrer rise. Eccause the flux distribution associated with unifom rod withdrawals does not involce high local peaks, and because several rods
- 2:st be coved to change power by a significant percentrge of rated power, the rate of power rise is very slow. Generslly, the heat flux is in near equilibrium with the fission rate. In an assumed uniform rod withdrwal approach to the scram level, the rate of power rise is no more than 57. of rated power per cinute, and the IRM system would ba note than adequate to assure a scram before the power could exceed the safety limit. The IRM scran remains active until the mode s.ritch is placed in the run position. This switch occurs when reac* or pressure is greater than 850 psig.
The analysis to support 7peration at various power and flow relatienships has considered operation with either one or two recirculation pu=ps. During steady-state operation with one recirculation pu.p operating the equalizer line shall be open. Analysis of transiests from this operatins: con-dition are less severe than the same transients frms the two pump operation. The operator will set the APRM neutron flux trip setting no greater than that shcun in Figure 2.3.1. 11c=.rever, the actual setpoint can be as much as 37. greater than that shown on Figure 2.3.1 for recirculation driving flows less than 507. of design and 27. greater than that shown for recirculatior. driving flows greater than 50~ of design due to the deviations discussed on page 18. B. AFFli Control Rod Block Trips Reactor power level nay be varied by moving control rods er by 3 varying the recirculation flow rate. The APRM system provides a control red block to prevent rod withdrawal beyond a given point at constant recirculation flow rate, and thus to protect against the condition of a MCFR less than the Safety TAnf t (T.S.2.1.A). This rod block trip setting, which is automatically varied with recirculation loop ficv rate, prevents an increase in the reactor i I power level to excessive values due to control rod withdrawal. The ficv variable trip setting provides substantial margin frem fuel damage, assu ing a steady-state operation at the trip setting, l 2.3 BASES 20 RE'
Bases Cptinued: over the entire recirculation flow range. The mar:;in to the Saf ty Limit increases as tha f1cv decreases for the specified trip setting versus fl<v relationship; therefere, the worst case MCPR which ccald occur during stesdy-state operation ir at '087. of rated thermal power because of tbc APRM rod block trip setting. He actual p'.rer dis trib"tlon in the core is established by specified control rod sequences and is monitored c~ntinuously by the in-core LFPl! system. When the maxist total peaking factor exceeds the desir" value, the red block setting is adjusted in accordance with the formula in Specification 2.3. .. If the AFFJ1 red block setting should requi*e a change due te an abnormal peaking condition, it <ill be done by increasing the APRM gain and thus reducing the slope and intercept point of the flew referenced rod block curve by the reciprocal of the APRM gain change. The operator will set the APPJ1 rod block trip settings no greater than that shown in Figure 2.3.1. Hewever, the cetual setcoint can be as much as 3% c,reater than that shcun on Figure 2..i.1 for recirculation driving flows less than SW, of design and 27., greater than that shcwn for recirculation driving flows greater than 50~. of design due to the deviations disenssed on Fage 18. C. Reactor Low Water Level Scram The reactor low water level scram is set at a point which will assure that the water level used in the bases for the safety limit is main' vined. The operator will set the Icw water level trip setting no lower then 10'6" above the top of the active fuel. IIowever, the actual setpoint can be as much as 6 inches Iwer due to the deviations discussed on page 18. D. Re9ctor Low Low Water Level ECCS Initiation Trip Foint The emergency core cooling subsystcms are designed to provide sufficient cooling to the core to dissipate the energy associated with the loss of coolant accident and to limit fuel clad temperature to well below the clad melting teaperature to assure that core geometry remains intact and ta linit any clad metal-water reaction to less than 17.. The design of the ECCS ccaponents to meet the above criterion was dependent on three previously set parmteters: the maximum break size, the low water level scram setpoint, and the ECCS initiation set-poirt. To Icer the setpoint for initiation of the ECCS could prevent the ECCS components from 2.3 BASES 21 REV
i i Ba<:es Continued: l & cting their criterion. To : aire the FCCS initi,tien setpoint would be in a safe directien, but it would reduce the margin established to prevent act"atic1 of the ECCS during normal operation or during normally expected trancients. n e operator will set the Icv low water level ECCS initiation trip setting 2 6'6" $ 6'10" abmre the ' top of the active fuel. Ilowever, the actual setpnint can be as ruch as 3 iaches Icwer than the 6'6" setpoint and 3 inches greater than the 6'10" retpoint due to the deviations discussed on page 1". E. Turbine Control Valve Fast Closute Scram The turbine control valve fast closure scram is prcvided to anticipate the rapid increase in pressure and neutron flux resulting from fast closure of the turbine centrol valves dae to a load rejection and subsequent failure of the bypass. This transieit is less severe than the turbine stop valve closur with bypass failure and therefore adequate nargin exists. F. Turbine Stcp Valve Scram The turbine stop valve closure scram trip anticipates the pressure, neutrc, flux and heat flux increase that could result from rapid closure of tha turbine stop valves. With a , scram trip setting of 510% of valve closure free full open, the resultant increase in surface heat i flux is limited such that MCPR remains above the Sefety Limit (T.S.2.1.A) even during the worst case transient that assumes the turbine bypass is closed. G. Main Stea t Line Isolation V lve Closure Scram The main steam line isolation valve closure scram anticipates the pressure and flux transients which occur during normal or inadvertent isolation c lonure. With the scram set at 10% valve closure there is no increase in neutron flux. II. Reactor Coolant Lw Pressure Initiates Main Steam Isolation Valve Closure The Icw pressure isolatic of the main stems lines at 850 psig was provided tn give protection against rapid reactor depressurimtion and the resulting rapid cooldein of the vessel. /4 vantage was taken of the scram feature which occurs when the main stea line isolation valves are closed to previde for reactcr shut ' wn so that
- high power operation at I m reactor pressure does net occur, thus providing protection a the fuel cladding integrity safety li-it. Operation of the reactor at pressures I wer ~:han 850 psig requires 2.3 BASES 2
REV
i L i i a > i Bases Continued-
- l I that the reactor exxle switch S in the startup position w* tere pre
- action of the fuel cladding integrity safety limit is provided by the IRM high neutron flux scr:ca. IIc, the combination of
, nain steam line Icv pressure isolation sud 1~ on tion valve closure scram assures de availability
- of the neutron scram protection over the cut - a re- s;e of npplicability of the fuei cladding in+-str it- !
safety limit. [ t The operator will set this pressure trip at greater than or equal to 850 psig. Hwever, the actupI trip setting can be as much as 10 psi I wer due to the deviations discussed on page 18. 1
}
. References t i
- 1. Linford, R. B., " Analytical Methods of Plant Transictit Evaluations for the General Electric Boiling Woter Reactor," tTEO-10802, Feb. , 1973.
i 4 I 4 h i l 2.3 bases - 2n M r i
. ~
7* es Centinued: 3J Mc FFCI ned/or RCIC high fm and temperature inst runentation is previded to detect a break in the in I and ~r RCIC piping. Trippirg of this instrtmentatirn r(sults in actration of HFCI and/or RCIC isolati m va h <-s ; i.e., Group 4 and/or Group 5 valvas. %c trip rettings of 200 F and 1507. of HFCI and 307e of RCIC design ficw and valve closure times are such ' hat the core will not be uncovered and fission proa tc reicase sill not exceed 10 CFR 100 guidelires. The instrumentation which initiates ECCS action is nrranged in a dual bus system. As for other vital in-trumentation arranged in this fashion the Specification preserves the effectiveness of the system even during periods tdien maintenance or testing is being perfemed. The control rod block functions are provided to prevent excessive control rod withdrawal so that NCFR remains above the Safety Limit (T.S.2.1.A) . D e trip logic for this function is 1 out of n; e . st . . any trip on one of the six AFRM's, eight IRM's, or four SRM's will result in a rod block. The einice-instrument channel requirements for the IRM and REM .ay be reduced by one for a short period of time to allow for maintenance, testing, or calibration. See Section 7.3 FSAR. The APRM red block trip is referenced to fice and prevents a significant reduction in MCPR especially during operation at reduced ficv. The AFRM provide gross core protection; i.e., limits the gross cc:e power increase from withdrawal of control rods in the norral withdrawal sequence. The trips are set so that MCFR is maintained greater than the Safety Limit. The REM provides local protection of the core; i.e., the prevention of critical pcuer in a local regic-- of the core, for a single rod withdrawal error from a limiting control rod pattern. The trip point is I referenced to itsr.i. The worst case single control rod withdrawal error bes been analyzed and the results shc4r that with the specified trip settings rod withdrawal is blocked at ICFR greater than the Safety Limit, thus allowing adequate margin. Below 607. power, MCFR remains above the Safety Limit for the worst case withdrawal of a single control rod without rod block action, thus below this level it is
! not required. This subject is discussed in General Electric IE"A Thermal Analysis Basis (CETAB):
Data, Correlation and Design Application, NEDO-10958. Requiring at least half of the normal LFRM inputs from each Icvel to be operable assures that the REM response will be adequate to prevent rod withdrawal errors. The IRM rod block function provides local as well as gross core protection. The scaling arrangement is such that trip setting is less than a factor of 10 above the indicated level. Analysis of the worst l case acci. dent results in rod block action before MCFR approaches the Safety Limit (T.S.2.1.A). A dcvrscale indication of an APRM or IRM is an indication the instrument has failed or the instrument is not sensitive enough. In either case the instrument will not respond to changes in control rod motion and thus control rod motien is prevented. The downscale trips are set at 3/125 s.f full scale. 3.2/4.2 U
; REV
Bases continued 3.3 and 4.3: consequences of reactivity occidents are functi~'s of the initial neutron flux. The require-ment of at least 3 counts per second assures that any transient, shculd it occut, begins at or above the initial value of 10' cf rated p~ee used in the analyses of transients from cold conditions. One operable SP11 channel would be edequate to monitor the approach to criticality using homogeneous patterns cf scattered control red vithdrawal. A minirn.:m of two operable SPMs are provided as an a&'ed conservatism.
- 5. The consequences of a red block monitor failure have been evaluated. These evaluations show that during reactor operation with certain limiting control rod patterns, the withdrawal of a designated single control rod could result in om or more fuel rods with PCFR's belev the Safety Limit (T.S.2.1.A). During use of such patterns, it is judged that testing of the RMi system prior to viihdrawal of such rods to assure its operability vill assure that improper withdrawal does not occur. It is the responsibility of the Engineer, Nuclear, to identify these liniting patterns and the designated rods either d en the patterns arc initially established or as they develop due to the occurrence of inoperabir rods in other than limit ing patterna.
C. Scrn t Insertion Times The control rod system is designed to brit.g the reactor subcritical at a rate fast enough to prevent fuel damage; i.e., to prevent the MCFR from becoming less than the Safety Limit (T.S.2.1.A). This requires the negative reactivity insertion in any local region of the core and in the overall core to be equivalent to at least the scram reactivity curve used in the transient analysis. The required average scram times for three control rods in all two by two arrays and the required average scram times for all control rods are based on inserting this a-cunt of negative reactivity at the specified rate locally and in the overall core. Under these conditions, the thermal limits are never reached during the transients requiring control rod scram. The limiting operational transient is that resulting from a turbine stoo valve closure with failure of the turbine bypass system. Analysis of this transient shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given in he above Specification, provide the required protection, and MCPR remains above the Safety Limit (T.i.2.1.A). In the analytical treatment of the transients, 290 milliseconds are a11 cued between a neutron sensor reaching the scram point and the start of motion of the control rods. 3.3/4.3 FASES B5 REV f
1.f JJWT_pigjfrGIT!ONS FOR OPEM1'jNS '.0 SUWEILIANgEREQUIRDimTS _ _ . , 3.11 UACICM ITEL ASSCIBLTES 4.11 REACTOR PUEL ASSmBLIES Applicability Applicability The Liciting Corditions for t*peration The Surv"I11ance Requirments apply to associe.ed with the fuel rods apply to , the pa ra eters which moniter the fuel those parameters which c oriNr the fel ' rod opera ting conditiens. rod opnrating conditiens. Objective Objective The objective of the Limitirn Conditf ors The objective of the Surveillance Requirrents for Operation is to assure the perfor- is to specify the type and frequency of surveil-mance of the fuel rods. lance to be applied to the fuel rods. Specifications Specifications ' A. Average Planar Linear Hoat Genera- A. Average Planar Linear Heat Genera-tion Rate (APLHGR) tion Rate (APLHGR) During steady state power operation, 1. The AFLHCR for each type of lue1 as a the APLHGR for each type of fuel as function of average planer exposure 6all a function of average planar exposure be detemined daily during reactor operation shall not aceed the li:niting value at 2t 257. rated themal power. shown in Figures 3.11-1. If at any , tir.2 it is deterciined that the limit- 2. Zienever the plant technical staff determines inn value for APLHGR is being exceed- that more frequent surveillance of AFLUGR ed, action chall be taken innediately is necescary, it shall specify an augmented to restore operation to within the surveillance program cer:anensurate with prescribed limits. l reactor conditions. 3.11/4.11 gg, ,
3." LT;t1 TING CONDI110 TIS FOR OFFA UION 4.0 StT.*EIL1x:CE REQUIRHtE.TS B. Loc 11 Utr:r B. toc:11 IFGR During steady state power operation, the linear 1. The local LHGR as a function of core % , ~,h t heat generation rate (U!GR) of any rod in any shat.' be checked daily durieg reacte - fuel asser,bly at any axial location shall not ope t e' ion a t > 257. o f ra ted the mal e 'c r. exceed the maximum allowabic U GR as calculated by the foi towing equation: 2. Ebenever the plant technical staff ( .a t-ines
~
.' that more frequent surveillance of h :21 UIGR _
< UlGR 1- bP ) 'L UlGR is necessary, it shall specify mt max d ( p/ max LT augmented surveillance prograri corrner tra te f
UtGR _. with reactor conditions. d = Design LIIGR
= 17.5 kw/ft for 7x7 fuel
=
13.4 kw/ft for 8x8 fuel
/ = Maximum power spiking penalty
\,$j) p max
= 0.076 for 7x7 fuel
= 0.021 for 8x8 fuel LT= Total core length = 12 ft L- Axial position above bottom core If at any time it is determined that the limiting value of UlGR is being exceeded, action shall be taken ic nediately to restore operation to within prescribed limits.
3.11/4.11 169 C REV
.a
. o
_3. 0,J.IM1!'!!G CB DITITS FOR OPERA M f
' .C SURVEILIE% EQUIREMDTS i C. Minimum Critical Pover Ratio (MCPR) C. Minimum C.ritical Power Ratio (MCPR)
[ During cready state pacer cre-ation, 1 MCrn chat 1 be checked daily the Ope , ting MCPR Limit shall be during reactor power operation at 11.41 f or Sx8 fuel and 21.'1 for 22 257 rated thermal pover. 7x7 fuel et rated pwer and flou. For cor. flows other than rated the 2.14tenevar the plant technical staff deterr; "' ~! Operati ; MCPR Limit shall b< the that rore frequent surveillance of MCPR i lj abcs e va?ue ratitiplied by Kr. ,
~
necessary, it shall specify an augmented where K y is given by Figure 3.11.2. surveillance program enensurate with re- .or if ett any tine it is detentred that
,I cenditions.
the limiting value of MCPR is being exceeded, action shall be taken imediately to restore operationao j '. within prescribed limits. ! y ' e i 9 ig r l r 'l s
. i 3.11/4.11 189 D 4 REV l
Ba <.es 3.11 - A, rr . , i. g n rInnar Lirear Tiaat Go" vation Rate (APU4G" l t This specification assures that the peak cladd 5g temperature fr 11owing the postulated design .
, basis less-of-coolant acci4nt will not exceed .5c lirit cpecified in the 10CFR50, Appendix K.
The pee cladding temperatere following a poet" lated lect-of-cea ant accident ts prirarily a functi m of the average her t generation rate or all *he rods of a fuel assembly at any :'xial location and is only depen hnt secondarily on *Se red to rod pmer distributio, within en as semi' ly. Since expected Iccal variations in rmer distribution within a fuct assembly affect the calculated peak cladding temperature by 1ers thnn i 20 F relative to the peak temperature for a typical fuel design, the limit on the average linear heat generation rate ir. sufficient to assure that calculated temperatures are within the 10CFR50 Appendix K limit. The liniting value for APU1GR is given by this specification. It is recocnized that APU1GR is a calculated para =eter that is not continually monitored and ala-d directly during core power distribution changes. If at the time of the calculation it is fcund r"it the limits are being exceeded, there is always an action which vill return the average planar Ric to within prescribed limits, r.amely power reduction. Under most circumstances, this will not be the 'aly alterrative. Wenever the limit is exceeded the monitored value will be documented and available for review, audit and inspection of plant operations. The only way to violate the Limiting CondMton for Operat ion is to knowingly allow operation beyond the prescribed limits without taking the necem q
- action to restore the average planar UIGR to within prescribed limits.
B. Local UIGE ! This specification assures that the linear heat generation rate in any rod is less than the design I l linear heat generation if fuel pellet densification is postulated. The power spike penalty specified is based on the analysis presented in Section 3.2.1 of Reference 1 c.nd in References 2 and 3, and assumes a linearly increasing variation and axial gaps between core bottom and top and assures vi-h a 957. confidence, that no rare than one fuel rei exceeds the design linear heat generation rate due to power spiking. It is recognized that UIGR is a calculated parameter that is not continually monitored and alar c ' directly during core power-distribution changes. If at the time of the calibration it is found i that the limits are being acceded, there is always an action which will return the UIGR to with prescribed limits, namely pasar reduction. L'nder most circumstances, this will not be the only alterritive. Wenever the limit is exceeded the nenitored value vill bc documented and available l for review, audit and inspection of plant operations. The only way to violate the Limiting Cond ton for OperatEon is to knowingly allow operation beyond the prescribed limits without taking the necesmtv , actior to restore the U1GR to within prescribed linits. , 3.11 EASES 169 E e.,,
Emes '.11 (entinuedi C. Mini"nrn re itic.,I Tower Ratin CiCPR) The ERCS evaluation presented in Reference 4 am-,ad the steady state MCPR prior te the pestehted iocs of coolant accident to be 1.19 fcr all fuel tyi a. The Operating itCPR Limit oi 1.41 for 8x8 fuel and 1.33 for 7x7 fuel is Jeter:nined from the analysis of transients discu3ned in Oases Sections 2 1 and 2.3. By mintrining an operating MCPR abcvc these limit s, the Sa fety Limit of 1.06 (T.S. 2.1. A) applicable to all fuel types is maintaine ' ir. the event of the most limiting abncmal operational tran .icnt. For operation with less than rated core flow the Operating MCPR Lirit is adjus ed by multiplyinc, the above limit by K . f Reference 5 discusses how the transient analysis done at rated conditions cacompasses the reduend flow situation when the proper Kr factor is applied. It is recogni ed that 11CPR is a calculated parameter that is not continually monitored and alarmed directly during core power distribution and thermal-hydraulic changes. If at the time of the evaluation it is found that the limits are being sceeded, there is always an action which will return the MCPR to within prescribed limits, namely power reduction. Under most circunstances, this will not be the only alternative. Whenever the limit is exceeded the monito =d value will be documented and available for review, audit and inspection of plant operations. We only way to violate the Limiting Condition for Operation is to knowingly allow operation bey-d the prescribed limits without takir.g the necessary action to restore the MCPR to within prescrib ' linits. Re ferences l' 1. " Fuel Densification Effects in Cencral Electric Boiling Water Reactor Fuel," Supplements 6, 7, and 8, NEDH-10735, August, 1973.
- 2. Supplement 1 to Technical Report on Densification of General Electric Reactor Fuels, December 14, 1974 (USAEC Regulatory Staff)
- 3. Cemnunication: V A liocre to I S Mitchell, " Modified CE Model for Fuel Densification,"
Docket 50-321, March 27, 1974.
- 4. "Monticello Nucicar Generating Plant Loss-Of-Coolant Accident Analysis Confonnance with 10 CFR 50 Appendix K, August 1974," L 0 Mayer (USP) to J F O' Leary, August 20, 1974.
- 5. " General Electric BWR 1hermal Analysis P"is (CETAB): Data, Correlatii a and Design
, Arolication," NEDO-10958, November,1973.
3.11 BASES 189r REV a
Bancs 4.11
^
I The APU TR. UN and MCPR chall be checked daily to det ermine if fuel burnup, or control rod novcocrt has causm! changes in power distributien. Since changes due to burnup cre sIcw, and only a few control rods are removed daily, a daily check of power distribu*lon is adequcte. For a limitir. value to occur bela- 2T. of rated themal power, ou enreasonably large peaking factor w"Id be required which is not the cas, ror crerating control rod requences. At certain tirm during plant startups nud power changee tbc plant technten1 staff may detemine that sur"ciliance of APUiGR, MGR and/or IEPR is necessary mere frequently than daily. Because the necessity for such an au;pnented surveillance program is a functic, of a number of interrelated parameters, a reason.,Ue program can only be detemined on a case-by-case basis by the plant technical staff. The check of APUIGE. MGR and I'CPR will normally be dere using the plant process coeputer. In the event that the creputer is unavailable, the check will consist of either a manual calerIntion or a ce=parison of existing core condit tms to those existing at the time of a previous check to det emice if a significant change has occurred. J r 4 4.11 BASES 189G nrv
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NRC L.sTRIBUTION FOR PART 50 DOCKE. AATERI AL (TEMPORARY FORM) FILE: FROM: . Northern States Pover DATE OF DOC DATE REC'D LTR TWX RPT OTHER [i""3P li U" 3-12-75 3-15-75 XXXX ORlG CC OTHER SENT AEC PDR XX TO: Mr Gianbusso 3 signed SENT LOCAL PDR XX
' CLASS UNCLASS PROPINFO INPUT NO CYS REC'D DOCKET NO:
XXXXXX XXXXXXXXXX 3 50-263 DESCRIPTION: ENCLOSURES: Ltr notarized 3-12-75...trans the followini: : Amdt to OL/ change to Tech Specs: Consisting of the ddoptation of the GETAB.GEXL heat flux correlation. . . . .(40 cys enc 1 rec 'd)
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PLANT NAME: Monticello FOR ACTION /lNFORMATION 3.is_7s e,h 5 BUTLE7 (L) SCHWENCE R (L) ZIEMANN (L) REG AN (E) W/ Copies W/ Copies / W/6 Copies W/ Copics CLARK (LI STOLZ (L) DICKER (E) LE AR (L) W/ Copim W/ Coniet W/ Conirn W/ Cocies PARR (L) VAbMLLU (L) ruwOn UWtc) bt'c Lb W/ Copics W/ Copics W/ Copies W/ Copies KNIEL (L) PURPLE (L) YOUNGB LOOD (E) W/ Copies W/ Copies W/ Copies W/ Copies INTERNAL DISTRIBUTION TECH REVIEW DENTON LIC ASST AAIND y SCHROEDER GRIMES ,,R. DIGGS (L) BRAITMAN 70GC, ROOM P 50GA MACCARY G AMMILL H. GE ARIN (L) SALTZMAN GOSSICK/ST AF F KNIGHT K ASTN E R E. GOULBOURNE (L) MELTZ f CASE PAWLICKl BALLARD P. KREUTZER (E) GIAMBUSSO SHAO SPANG LE R J. LEE (L) PLANS BOYD STELLO M. MAIGRET (L) MCDONALD MOORE (L) HOUSTON _ ENVIRO- S. REED (E) CHAPMAN DEYOUNG (L) NOVAK MULLER M. SERVICE (L) /*DUBE (Ltr) SKOVHOLT (L) ROSS DICKER S. SHEPPARD (L) M. COUPE GOLLER (L) (i tr) IPPOLITO KNIGHTON M. SLATER (E) PETERSON f P. CO L LINS TEDESCO YOUNGBLOOD H. SMITH (L) HARTFIELO (2) DENISE LONG REGAN S. TEETS (L) KLECKER BEG OPR LAIN AS OJECT LDR G. WILLI AMS (E) EISENHUT
/ FILE & REGION (2) BENAROYA / van V. WILSON (L) WIGGlh ON T.R. WI LSON VOLLMER H A R LESS R. INGRAM (L) 0 Tod "58^
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