ML20106A622

Proposed TS 2.0, Safety Limits & Limiting Safety Sys Settings & 3/4.11, Power Distribution Limits

ML20106A622
Person / Time
Site:	Dresden, Quad Cities
Issue date:	09/15/1992
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML17179A463	List: ML20106A622
References
NUDOCS 9209290214
Download: ML20106A622 (85)

Text

- _ _ - _ _ _ _ _ _ _ _

ATTACIIMENT 3 PROPOSED TECHNICAL SPECIFICATIONS Technical Specification 2.0

" SAFETY _ LIMITS AND LIMITING SAFETY SYSTEM SETTINGS" O 00 37 p

PDR

SAFETY LIMITS 2,1 -

2.0 SAFETY L!MITS ANO LIMiTiNu dMETY SYSTt '1 SETTINGS -

2d SAFETY LIMIIS THERMAL POWER. Low Pressure or Low Flow 2.1.A THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.

APPLICABILITY: CPERATIONAL MODE (s) 1 and 2.

ACTION:

With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the teactor vessel-steam dome pressure less than 785 psig or core flow less than 10% of rated flow be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.4.

THERMAL POWER, Hiah Pressure and Hiah Flow 2.1.B The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.05 (Unit 2), or 1.08 (Unit 3), with the reactor vessel steam dome pressure greater than or equal to 785 psig and core flow greater than or equal to 10% of rated flow. During single recirculation loop operation, this MCPR limit shall be increased by 0,01.

APPLICABILITY: OPERATIONAL MODE (s) 1 and 2.

i.

ACTIQNJ With MCPR less than the above app 4 cable limit and the reactor vessel steam dome pressure greater than or equal to 785 psig and core flow greater than or equal to 10% of rated flow, be in l

at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.4.

l DRESDEN - UNITS 2 & 3.

2-1 Amendment No.

i i

4 3

_ _ _ ~

_____._.-m_

SAFETY LIMITS 2.1 2.0 SAFETY LIMITS AND LlMITING SAFETY SYSTEM SETTINGS -

Reactor Coolant System Pressure 2.1.C The reactor coclant system pressure, as measured in the' reactor vessel steam dome, shall not exceed 1345 psig.

APPLICABILITY: OPERATIONAL MODE (s) 1,2,3 and 4.

ACTION:

With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above-1345 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal' to 1345 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6,4.

Reactor Vessel Water Level 2.1.D The reactor vessel water level shall be greater than or equal to twelve inches above the top of the active irradiated fuel.

APPLICABILITY: OPERATIONAL MODE (s) 3,4 and 5.

i ACTION:

I With the reactor vessel water level less than twelve inches above the top of the active irradiated fuel. manually initiate the ECCS to restore the water level, after depressurizing the reactor vessel, if required, and comply with the requirements of Specification 6.4.

i l

DRESDEN - UNITS 2 & 3 2-2, Amendment No.

~ -.

LSSS 2.2.

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS M

LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System (RPS) Instrumentation Setooinut 2.2.A The reactor protection system instrumentation setpoints shall be set t _..sistent with the Trip Setpoint values shown in Table 2.2.A-1.

APPLICABILITY: As shown in Table 3,1.A-1.

ACTION:

With a reactor protection system instrumentation setpoint less conservative than the value shown in the Trip Setpoint column of Table 2.2.A-1, declare the CHANNEL inoperable and apply the applicable ACTION statement requirement of Specification 3.1.A until the CHANNEL is restored to OPERABLE status with its setpoint adjusted consistent with the Trip Setpoint value.

l DRESDEN - UNITS 2 & 3 2-3 Amendment No.

LSSS 2.2 TABLE 2.2. A-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SFTPOINTS Functional Unit lilD_ Setoo:nt

1. Intermediate Range Monitor:

q a.

Neutron Flux - High s120/125 divisions of full scale d

b.

Inoperative NA

2. Average Power Range Monitor:

a.

Setdown Neutron Flux - High s15% of RATED THERMAL POWER b.

Flow Biased Neutron Flux - Hiv

1) Dual Recirculation Loop Operation W

a) Flow Biased s 0.58W

+ 62%

with a maximum of b) High Flow Maximum s120% of RATED THERMAL POWER

2) Single Recirculation Loop Operation a) Flow Biased s 0,58W* + 58.5%

with a maximum of b) High Flow Maximum

$116.5% of RATED THERMAL POWER c.

Fixed Neutron Flux - High s120% of RATED THERMAL POWER d.

Inoperative NA

3. Reactor Vessel Steam Domo Pressure - High s1060 psig

4. Reactor Vessel Water Level - Low 2144 inches above top of active fuel

5. Main Steam Line Isolation Valve - Closure s10% closed

6. Main Steam Line Radiation - High s 3" x normal full power background (without hydrogen addition) a "W* shall be the recirculation loop flow expressed as a percentage of the recirculation loop flow which produces a rated core flow of 98 million Ibs/hr.

b With Unit 2 operatir g above 20% RATED THERMAL POWER and hydrogen being injected into the primary coolant, this Unit 2 setting may be increased to "$3 x full power background (with hydrogen addition)."

DRESDEN - UNITS 2 & 3 24 Amendment No.

LSSS 2.2 TABLE 2.2.A-1 (Continued)

B,E_A_CTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS A

Functional Unit Trio Setooint

7. Drywell Pressure - High s 2 psig

8. Scram Discharge Volume Water Level - High s40.4 gallons (Unit 2) 541 gallons (Unit 3)

9. Turbine Stop Valve - Closure s10% closed

10. Turbine EHC Control Oil Pressura - Low 2900 psig

11. Turbine Control Valve Fast Closure 2460 psig EHC fluid pressure

12. Turbine Condenser Vacuum - Low 221 inches Hg vacuum

13. Reactor Mode Switch Shutdown Position NA

14. Manual Scram NA DRESDEN - UNITS 2 & 3 2-5 Amendment No.

SAFETY LIMITS B 2.1_

BASES 2d SAFETY LIMITS The Specifications in Section 2.1 establish operating parameters to assure that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs). These parameters are based on the -

Safety Limits requirements stated in the Code of Federal Regulations,10 CFR 50.36(c)(1):

" Safety limits for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity."

The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive materials to the environs, Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity limit is set such that no calculated fuel damage would occur as a result of an AOO Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit for the MINIMUM CRITICAL POWER RATIO (MCPR) that represeats a conservative margin relative to the conditions required to maintain fuel cladding integrity.

The fuel cladding is one of the physical boundaries which separate radioactive materials from the environs. The integrity of the fuel cladding is related to its relative freedom _from perforations or cracking. Although some corrosion or use-related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuol cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the protection system safety settings. While fission product migration from cladding perforations is just as measurable as that fiom use-related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration.

Therefore, the fuel cladding integrity Safety Limit is defined with margin to the conditions which would produce onset of transition boi_ ling (MCPR of 1.0). These conditions represent a significant departure from the condition intended by design for planned operation. Therefore, the fuel cladding integrity Safety Limit is established such that no calculated fuel damage shall result from an abnormal operational transient. This is accomplished by selecting a MCPR fuel cladding integrity Safety Limit which assures that during normal operation and AOOs, at least 99.9% of the fuel rods in the core do not experience transition boiling.

Exceeding a Safety Limit is cause for unit shutdown and review by the Nuclear Regulatory Commission (NRC) before resumption of unit operaticn. Operation beyond such a limit may not in itself result in serious consequences but it indicates an operational deficiency subject to regulatory _

review, i

t I

l DRESDEN - UNITS 2 & 3 8 2-1 Amendment No.

SAFETY LIMITS B 2,1 BASES 2.1. A

. THERMAL POWER. Low Pressure or low Flow This fuel cladding integrity Safety Li.ait is established by establishing a limiting condition on core THERMAL POWER developed in the following method. At pressures below 800 psia (~ 785 psig),

the core elevation pressure drop (0% power,0% flow) is greater than 4.56 psi. At low powers and flows, this pressure differentialis maintained in the bypass region of the core. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at law powers and flows will always be greater than 4.56 psi. Analyses show that with a bundle flow of 3

28 x 10 lb/hr, bundle pressure drop is nearly independent of bundle power and has a value of.

3 3.5 psi. Thus, the bundle flow with a 4.56 psi driving head will be greater than 28 x 10 lb/hr.

Full scale ATLAS test data taken'at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. At 25% of RATED THERMAL POWER, the peak powered bundle would have to be operating at 3.86 times the average powered bundle in order to achieve this bundle power. Thus, a core thermal power limit of 25% for reactor pressures below 785 psig is conservative.

241 B THERMAL POWER. Hiah Pressure and Hiah Flow This fuel cladding integrity Safety Limit is set such that no (mechanistic) fue! damage is calculated to occur if the limit is not violated. Since 'he parameters which result in fuel damage are not directly observable during reactor operati m, the thermal and hydraulic conditions resulting in departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not" necessarily result in damage to BWR fuel rods, the critical power ratio (CPR) at which boiling -

transition is calculated to occur has been adopted as a convenient limit. Howeveri the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value 'of the critical power. Therefore, the fuel cladding integrity Safety Limit is defined as~ the CPR in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties.

The margin between a MCPR of -1.0 (onset of transition boiling) and the_ Safety Limit, is-derived from a detailed statistical analysis which considers the uncertainties in monitoring the core operating state, including uncertainty in the critical power correlation. 'Because the transition boiling correlation is based on a significant quantity of practical test data, there is a very high '

confidence that operation of a fuel assembly at the condition where MCPR is equal to the fuel cladding integrity Safety Limit would not produce transition boiling. In addition, during single recirculation loop operation, the MCPR Safety Limit is increased by 0.01 to conservatively account-

- for increased uncertainties in the core flow and TIP measurements.

However, if transition boiling were to occur, cladding perforation would not necessarily be upected. Significant test data accumulated by the NRC and private' organizations indicate that the use of a boiling transition limitation to protect against cladding failure is a very cons _ervative

-=-

DRESDEN - UNITS 2 & 3 8 2-2 Amendment No.

m i

SAFETY LIMITS B 2.1 BASES approach. Much of the data indicates that BWR fuel can survive for an extended peric., in an environment of transition boiling.

2.1.C Reactor Coolant System Preslute The Sekty Limit for the reactor coolant system pressure has been selected such that it is at a pressure below which it can be shown that the integrity of the system is not endangered. The reactor coolant system integrity is an important barrier in the prevention of uncontrolled release of fission products. It is essential that the integrity of this system be protected by establishing a pressure limit to be observed for all operating conditions and whenever there is irradiated fuelin the reactor vessel.

The reactor coolant system pressure Safety Limit of 1345 psig, as measured by the vessel steam space pressure indicator, is equivalent to 1375 psig at the lowest elevation of the reactor vessel.

The 1375 psig value is derived from the design pressu;es of the reactor pressure vessel and coolant system piping. The respective design pressures are 1250 psig at 575oF and 1175 psig at 560 F. The pressure Safety Limit was chosen as the lower of the pressure transients parmitted by the applicable design codes, ASME Boiler and Pressure Vessel Code Section lit for the pressure vessel, and USASI B31.1 Code for the reactor coolant system piping. Ti.e ASME Boiler and Pressure Vessel Code permits pressure transients up to 10% over design pressure (110% x 1250

= 1375 psig), and the USASI Code permits pressure transients up to 20% over design pressure (120% x 1175 = 1410 psig). The Safety Limit pressure of 1375 psig is referenced to the iowest elevation of the reactor vessel. The design pressure for the recirculation suction line piping (1175 psig) was chosen relative to the reactor vessel design pressure. Demonstrating compliance of peak vessel pressure. uith the ASME overpressure protection limit (1375 psig) assures compliance of the suction piping s rith the USASI limit (1410 psig). Evaluation methodology to assure that this Safety Limit pressure is not exceeded for any reload is documented by the specific fuel vandor. The design basis for the reactor pressure vessel makes evident the substantial margin of protection against f ailure at the safety Fssure limit of 1375 psig. The vessel has been designed for a general membrane stress no greater than 26,700 psi at an internal pressure of 1250 psig; this is a factor of 1.5 below the yield strength of 40,100 psi at 575 F. At the pressure limit of 1375 psig, the general membrane stress will only be 29,400 psi, still safely below the yield strength.

The relationships of stress levels to yield strength are comparable for the primary system piping and provides similar margin of protection at the established pressure Safety Limit.

The normal operating pressure of the reactor coolant system is nominally 1000 r %. Both pressure relief and safety relief valves have been installed to keep the reactor vessel peak pressure below 1375 psig. However no credit is taken for relief valves during the postulated full closure of all MSIVs without a direct (valve position switch) scram. Credit, however, is taken for the neutron flux scram. The indirect flux scram and safety valve actuation proside adequate margin below the allowable peak vessel pressure of 1375 psig.

DRESDEN - UNITS 2 & 3 B 2-3 Amendment No.

-..~

' 1 SAFETY LIMITS B 2.1 BASES 2.1 D Reactor Vessel Water Level With fuelin the reactor vessel during periods when the reactor is shutdown, consideration must also be given to water level requirements due to the effect of decay heat. If reactor water level should drop below the top of the active irradiated fuel during this period, the ability to_ remove decay heat is reduced. This reduction in core cooling capability could lead to elevated cladding temperatures and cladding perforation. The core will be cooled sufficiently to prevent cladding melting should the water level be reduced to two-thirds of the core height. The Safety Limit has been established at 12 inches above the top of the active irradiated fuel to provide a point which can be monitored and also provide adequate margin for effective action. The top of active fuelis 360 inches above vessel zero.

t i

l DRESDEN - UNIT 3 2 & 3 B24 Amendment No.

1. y. - -

ym.-.=_-.-.m,.,

,~.

4

'D o o

LSSS B 2.2 I

BASES 2J LIMITING SAFETY SYSIffLSEllL4GS The Specifications in Section 2.2 establish operational sottings for the reactor protection system instrumentation which initiates the automatic protective action at a lovel such that the Safety Limits will not bo excooded. Those sottings are based on the Limiting Safoty System SettinOs requirements stated in the Code of Foderal Ro0ulations,10 CFR 50.3G(c)(1):

" Limiting safety system settings for nuclear reactors bro settings for automatic protectivo devices related to those variablos havin0 significant safety functions. Whore a limitin0 safoty system setting is specified f or a variable on which a safety limit has boon placed, the setting must be so chosen that automatic protectivo action will correct the abnormal situation beforo a safety limit is exceeded. "

22d BRAG 10Lf1D10CliQn_SY110m In11n1!DCulal!9n3tumI 11 E

The Reactor Protection System (RPSI instrumentation setpoints specified in tho tablo are tho values at which the reactor scrams are set for each paramotor. Tho scram settings have boon selected to ensure that the reactor core and reactor coolant system are provented from excooding their Safety Limits during nermal operation and design basis anticipmod operational occurrences and assist in mitigating the consequences of accidents. Conservatism iacorp,.ted into tho transient analysis is documented by each approved fuel vendor. The baser it.r infvual scram settings are discussed in the following para 0raphs.

1.

l_ntermediate Ranoelignitor. Nau1LQHEUx Hinh i

5 The IRM system consists of ei0ht chambers, four in each of the reactor protection system lo0 c CHANNELS. The IRM is a 5 decade,10 range, instrument which cowrs the rango of power lovel between that covered by the SRM and the APRM. The IRM scram setting at 120 of 125 divisions is active in each ran00 of the IRM. For examplo,if the instrument were on Rango 1, the scram setting would be 120 divisions for that ran00: likewise,if the instrument woro on Rango 5, the scram would be 120 divisions on that ran00. Thus, as the IRM is ranged up to accommodato tho increase in power level, the scram settin0 s also f anged up.

i The most si nifican ecurces of reactivity chan0e during the power increase ce due to control rod 0

withdrawal. In order tc ansure that the 19M p avides adequato protection again,n the single rod withdrawal error, a range of rad withdrawal events has boon analyzed. This analysis included starting the event at various power levels. The most severo caso involves an initial c indition in which the reactor is just subcritical and the IRM system is not yet. n scato.

Additional conservatism was taken in this analysis by assuming that the IRM CHANNEL closest to the withdrawn rod is bypassed. The results of this analysis show that the reactor is scrammed and peak power is limited to 1% of rated power, thus mainta.aing MCPR above the fuel cladding integrity Safety Limit. Based on the above anilysis, the IRM provides protection against local Amendment No.

ORESDEN - UNITS 2 & 3 B25 f

_ m,

LSSS B 2.2 DASES control rod withdrawal errors and continuous withdrawal of conaol rods in the sequence and provides backup protection for the APRM.

2.

Averaao Power.RansMonitor For operation at low pressure and low flow during Startup, a reduced power level,i.e., setdown, i

APRM scram setting of 15% of RATED THERMAL POWER providos adequate thermal margin butween the setting and the Safety Limit. The mstgin is adequato to accommodato anticipated maneuvers associated with power plant startup. Effects of increasing pressure at zero of low void content are minor; cold water from sources available during startup are not much colder than that already in the system; temperature coef ficients are small; and, control rod pattores are constrained -

to be uniform by operating procedures backed uo by the rod worth minimizer. Of all possible sources of reactivity input, uniform control rod withdrawalis the most probablo cause of significant power rise. Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change power by a significant percentage of rated power, the rav of power riso is very slow. Generally, the heat flux is in near equilibrium with the fission rate, in an assumed uniform rod withdrawal approach to the scram setting, the rate of power rise is no more than 5% of RATED THERMAL POWER por rninute, and the APRM system would be more than adequate to assure a scram before the power could exceed the Safety Limit. The 15% APRM sotdown scram sotting remains active until the modo switch is placed in the Run position.

1 The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady state conditions, also providos a flow biased neutron flux which roads in-porcent of RATED THERMAL POWER. Because fission chambers provide the basic input signals, the APRM system responds directly to average neutron flux. During transients, the instantaneous rate of heat transfor from the fuel (reactor thermal power)is less than 'ho instantaneous neutron flux due to the time constant of the fuel. During abnormal operational transients, the thermal power of the fuel will be less than that indicated by the neutron flux at the scram setting.

Analyses demonstrate that, with a 120% scram setting for dual recirculation loop operation, or with a'116.5% scram setting for single recirculation loop operation, none of the abnormal operational transients analyzed violates the fuel cladding integrity Safety Limit, and there is a substantial margin from fuel damage. One of the neutron flux scrams is flow dependent untilit reaches the applicable setting where it is " clamped" at its maximum allowed value. The use of the flow referenced noutron flux scram setting provides additional margin beyond the use of a the fixed -

i high flux scram setting alone.

An ir, rease in the APRM scram setting would decrease the margin prosent before the fuel cladding integrity Safety Limit is reached. The APRM scram setting was determined by an analysis of margins required to provido a reasonable range for maneuvering during operation. Reducing this cperating margin would increase the frequency of spurious scrams, which have an adverse effect-on reactor safety because of the resulting thermal stressos. Thus, the APRM scram setting was selected because it provides adequate margin for the fuel cladding integrity Safety Limit, yet allows operating margin that reduces the possibility of unnecessary scrams.

DRESDEN - UNITS 2 & 3 B 2-6 Amendment No J

-,w-

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

~

I LSSS B 2.2 i

BASES During single recirculation loop operation, the normal drive flow relationship is altered as a result of j

reverse flow through the idle loop jot pumps when the active loop recirculation pump speed is above approximately 40% of rated. The core receivos loss flow than would be predicted based upon the dual recirculation loop drive flow to co<o flow relationship, and the APRM flow biased scram sottings must be aftered to continue to provido a reactor scram at a conservativo neutron flux.

l The scram setting must also be adjusted to ensure that the LHGR transient limit is not violated for any power distribut'on. The scram sotting is adjusted in accordance with Specification 3/4.11.B in order to maintain adequate margin for the Safety Limit and yet allow operating margin sufficient to reduce the possibility of an unnecessary shutdown. The adjustment may also be accomplished by increasing the APRM gain. This provides the same degroo ci protection as reducing the scram settings by raising the initial APRM readings closer to the scram settings such that a scram would be rocoived at the samo point in a transient as if the scram settings had boon reducod.

i 3.

Reactor VessgLEigam Domo Progguro - Hiah High pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products. A pressure increase while operatinC will also tend to l

increase the power of the reactor by compressing voids thus adding reactivity. The scram will quickly reduce the neutron flux, counteracting the pressure increase. The scram setting is slightly higher than the operating pressure to permit normal operation without spurious scrams. The scram setting provides for a wide margin to the maximum allowable design pressure and takes into i

account the location of the pressure measuromont (reactor vessel steam space) compared to the highest pressure that occurs in the system during a transient, in compliance with Section lll of the ASME Code, the safety valves must be set to open at no iJphor than 103% of design pressure, and they must limit the reactor pressure to no more than 10% of design pressure. Both the high neutron flux scram and safety valvo actuation are required to provont overpressurizing the reactor pressure vessel and thus, excooding the pressure Saftty Limit. The pressure scram is available as backup protection to the high flux scram.

Analyses arc performed for occh reload to assure that the pressure Safety Limit is not excooded, 4.

Reactor Vessel Water Level Low

. The reactor vessel water level scram sotting was chosen far enough below the normal operatinD level to avoid spurious scrams but high enough above the fuel to assure that there is adequate

~

protection for the fuel cladding integrity and reactor coolant system pressure Safety Limits. Tho scram setting is based on normal operating temperature and pressure conditions because the level instrumentation is density compensated, r

The scram sotting provided is the actual water level which may be differont than the water lovel as l

measured by the instrumentation outside the shroud. The water levelinsido the shroud will DRESDEN UNITS 2 & 3 B 2-7 Amendment No.

i l

-~, _ -

.,m._,

m

.m

~ _ _ _ _.. _

t t

f LSSS B 2.2 l

BASES e

decrease as power is increased to 100% in comparison to the level outside the shroud, to a maximum of seven inches, due to the pressure drop across the steam dryer. Therefore, at 100%

power, on indicated water level of + 8 inches water level may be as low as + 1 inches inside the shroud which corresponds to 144 inches above the top of active fuel and 504 inches above vessel P ro.

i

@tn Steam L_ine Isolation Valve Closure 5.

Automatic isolation of the main steam lines is provided to give protection against rapid reactor depressurization and cooldown of the vessel. When the main steam line isolation valves begin to close, a scram signal provides for reactor shutdown so that high power operation at low reactor i

pressures does not occur. With the scram setting at 10% valve closure (from full open), there is no appreciable increase in neutron flux during normal or inadvertent isolation valvo closure, thus providing protection for the fuel cladding integrity Safety Limit. Operation of the reactor at pressures lower than the MSlV closure setting requires the reactor mode switch to he in the Startup/ Hot Standby position, where protection of the fuel cladding integrity Safety Limit is provided by the IRM and APRM high neutron flux scram signals. Thus, the combination of main steam line low pressure isolation and the isolation valve closure scram with the mode switch in the Run position assures the availability of the neutron flux scram protection over the entire range of applicability of furl cladding integrity Safety Limit.

6.

Main Steam Line Radiation High High radiation levels in the main steam line tunnel above that due to the normal nitrogen and vxygen radioactivity are an indication of leaking fuel. When high radiation is detected, a scram is initiated to mitigate the failure of fuel cladding. The scram setting is high enough above background radiation levels to prevent spurious scrams yet low enough to promptly detect gross failures in the fuel cladding. This setting is determined based on normal full power background (NFPB) radiation levels without hydrogen addition. With the injection of hydrogen into the foodwater for mitigation of intergranular stress corrosion cracking, the full power background levels may be significantly increased. The setting is int.reased based on the new background levels to allow for the injection of hydrogen. This trip function provides an anticipatory scram to limit offsite dose consequences, but is not assumed to occur in the analysis nf any design basis event.

DRESDEN UNITS 2 & 3 B 2-8 Amendment No.

we

,-,ane, w-wn.

>,.m---r.

,.r-mv

,,u----,-,., -

-,--s-n--

ren---n-e---.,-~.r-.e.-+.,~,----.-w.--4s

--m,..w.wa-,-

..,r,-

w

.w-.--,an.w-rw+--m s

,ww-s-,s---

=

LSSS D 2.2 BASES J

7.

Drvwell PJessure Hiah i

J High pressure in tho drywell could indicate a break in the primary pressure boundary systems or a loss of drywell cooling. Therefore, pressure sensing instrumentation is provided as r, backup to the l

water levelinstrumentation. The reactor is scrammed on high pressure in order to minimize the possibility of fuel damage and reduce the amount of energy being added to the coolant and the primary containment. The scram sotting was selected as low as possible without causing spurious

scrams, i

l' 8.

Scram Discharae Volumo Water Leye1Qiab t

The control rod drivo scram system is designed so that all of the water which is dischargod from the reactor by a scram can be accommodated in the dischargo piping. A part of this system is ' '

individual instrument volume for each of the scram dischargo volumos. Those two instrument volumes and their piping can hold in excess of 90 gallons of water and are the low point in the piping, No credit was taken for the instrument volumas in the design of the dischargo piping relativo to the amount of water which must be accommodated during a scram. During normal operations, the scram discharge volumes are empty; however, should either scram discharge volumo accumulate water, the water discharged to the piping from the reactor during a scram may not be accommodated which could result in slow scram times or partial or no control rod insertion.

To preclude this occurrence, loval switi hos have boon installed in both instrument volumes which will alarm and scram the reactor while sufficient volumo romains to accommodate the discharged water. Diverse level sensing methods havo been incorporated into the design and logic of the system to prevent common modo failure. The setting for this anticipatory scram signal has been chosen on the basis of providing sufficient volume romaining to accommodate a scram, even with 5 gpm leakage per drivo into the scram dischargo volume. As indicated above, there is sufficient volume in the piping to accommodate the scram without impairmont of the scram times or the amount of insortion of the control rods.

I 9.

Iur_bine Ston Valvo - C_losure The turbine stop valvo closure scram setting anticipatos the pressure, neutron flux, and heat flux increase that could result from rapid closure of the turbino stop valves. With a scram setting of 10% of valve closure from full open, the resultant increase in surf ace boat flux is limited such that MCPR remains above the fuel cladding integrity Safety Limit, even during the worst case transient

- that assumes the turbino bypass fails to operate.

10.

Turbine EHC Control Oil Pressure - Low The turbine EHC control system operates using high pressure oil. There are st,voral points in this oil system where a loss of oil p. ssure could result in a fast closure of the turbine control valves.

This fast closure of the turbino control valvos is not protected by the turbine control valve fast DRESDEN UNITS 2 & 3 8 2-9 Amendment No.

a-te"ie*N v7'e-est-

-ev vw iv-gi w-gr-ge owne v w 's i-wN-F w

-w we,w eAem

' W-v w,i v e rya ww wie r-yy'yv,aiswp-tveq vm-g+e-g

-erc gg--

c y g-1-

1-ya,,--ge-mmgi-y----ayv-M-pm'yrw-14p-i-gumi-wi: ut'WSuur * '

89"44'-EP$Tr7W""'F"'N9"Ww T'"'r*-#

y-it'tuu-tw-^-e

+e'ver'"**

r i

i LSSS B 2.2 BASES closure scram since failuro of the oil system would not result in the fast closure solonoid valves bein0 actuated. For a turbmo control valve fast closure, the core would bo protected by the APRM and reactor h% pmssure scrams. However, to provido the same mar 0 ns as provided for the i

generator Icad rejection on fast closure of the turbino control valves, a scram has boon added to the tosctor orotection system which sensos failure of control oil pressure to tho turbino control system. Th.9 scram anticipatos the pressure transient which would be caused by imminent control valvo closuro and results in reactor shutdown before any significant increaso in neutron flux occurs. The transient response is very sirnilar to that resulting frorn the turbino control valve fast closuro scram. However, since the control valvos will not start to close until the fluid pressure is approximately 600 psig, the scram on low turbine EHC control oil pressure occurs well beforo turbino control valvo closure begins. The scram setting is hi h enou0h to provide the necessary 0

anticipatory function and low enough to minimize the number of spurious scrams.

LEbiliLCD01rM.VJ1 Lye Fast CLosm 11.

l The turbino control valve f ast closure scram is provided to anticipato the rapid increaso in pressure and noutron flux resulting from f ast closure of the turbino control valvos duo to a load rejection and subsequent failure of the bypass valves; i.0,, MCPh remains above the fuel cladding integrity Safety Limit for this transient. For the load rejection without bypass transient from 100% power, the peak heat flux (and therefore LHGR) increases on the order of 15% which provides a wido mar 0 n to the value correspondin0 to 1% plastic strain of the cladding.

i The scram sotting based on EHC fluid pressure was developed to ensure that the pressure switch is actuated prior to the closure of tho turbino control valves (at approximately 400 psig EHC fluid pressurch yet assure that the system is not actuated unnecessarily due to EHC system pressure transients which may cause EHC system pressure to momentarily decrease.

12.

Lubine t'ondensor Vacuum - Law.

Loss of condonsor vacuum occurs when the condensor can no longer handlo the heat input. Loss of cundenser vacuum initiates a closure of the turbino stop valvos and turbino bypass valves which onminatos the heat input to the condensor, Closure of the turbino stop and bypass valves causes a pressure transient, neutron flux rise and an increase in surfaco heat flux. To provent the fuel cladding integrity Safety Limit from being exceeded if this occurs, a reactor scram occurs on.

turbine stop valvo closure. The turbino stop valve closure scram function alone is adoquate to prevent the fuel cladding integrity Safety Limit from being exceeded, in the event of a turbino trip transient with bypass closure. The condenser low vacuum scram is anticipatory to the stop valve closure scram and causos a scram before the stop valves (and bypass valves) are closed and thus, the resulting transient is less seveie.

DRESDEN - UNITS 2 & 3 8210 Amendment No,

LSSS B 2.2 I

BASES

?

i 13.

Hoacto.tfAodo Switch Shutdown Position

- The reactor modo switch Shutdown position is a redundant CHANNEL to the automatic protectivo instr mentation CHANNEL (s) and providos additional manual reactor scram capability u

14.

Mangalggggfy The manual scram is a rodunda 't CHANNEL to the automatic protective instrumentation CHANNEL (s) and provUes manual reactor scram capability.

r L

i

?

i k

l 1

l DRESDEN UNITS 2 & 3 8211.

Amendment No.

• etar'*I"

.-6*4 1PWT-$$=-it'T" hi4 r *'=iWem-+"@WWW Mwww a utM-e rw'se e-'='r.Pu*sier#-rwiDW'T=+rrcem*=e-+=.aWsh%*"E=*dm

=vWWWw4 9 e

• 4M'**WWF-9 tyW 9 9 91'P-9'N*tfn--

-r=+M-9-l+W4etNM'7-sM-*T'*F--

+ r* tv 9 y v'iw M W@fWt tF

m

. I

\\

l SAFETY LIMITS 2.1 2.0 SAFETY LIMITS ANO LIMITING SAFETY SYSTEM SETTINGS i

a 2d SAFETY LIM _ITS l

THERMAL POWER Low Pressure or Low Flow i

2.1. A THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vossol steam dome pressure loss than 785 psig or core flow loss than 10% of rated flow, i

APPLICA[llLITY: OPERATIONAL MODE (s) 1 and 2.

AC_TlQN; i

With THERMAL POWER excooding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure loss than 785 psig or core flow loss than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and complv with the requirements of Specification 6.4.

THERMAL POWER Hinh Pressure and Hinh Flon 2.1.8 The MINIMUM CRITICAL POWER RATIO (MCPRi shall not be loss than 1.06 with the reactor vessel steam dome pressure greator than 785 psig and coro flow greator than 10% of rated flow. During single recirculation loop operation, this MCPR limit shall no increased by 0.01.

APPLICABILITY: OPERATIONAL MODE (s) 1 and 2.

ACTION:

- With MCPR less than the above applicable limit and the reactor vessel steam domo pressure

- greater than 785 psig and core flow greater than 10% of rated flow, be in at least HOT-SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.4 i

'l OUAD CITIES - UNITS 1 & 2 21 Amendment No.

i vvra m +w--

e v + v-

-++v-r

--+%-

, n v m -r a ~ ~ - -

w w w, --

=---"%

4

--ee-

+ -,. - - -. -

-ve-

---~n-*~ ~.. *-~ ~ + - = =

• SAFETY LIMITS 2.1 4

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS BnacioLCoplanLSystem Pressuro i

2.1.C The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1345 psig, i

l APPLICABJ1 ITY: OPERATIONAL MODE (s) 1,2,3 and 4.

ACllGUI With the reactor coolant system prest,ure, as maasured in the reactor vessel stearn dome, above-1345 psig, be in at least HOT SHUTDOWN with reactor coolant system pressuro less than or equal-to 1345 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification G.4.

Reactor Vessel Water Ley.n]

?,1.D The reactor vessel water level shall be greater than twelve inches above the top of the active irradiated fuel.

APPLICABILITY: OPERATIONAL MODE (s) 3,4 and 5.

l ACTION 1 With the reactor vessel water tevel at or below twelve inches above the top of the active irradiated fuel, manually initiate the ECCS to restore the water level, af ter depressurizing the reactor vessel,if required, and comply with the requirements of Specification 6,4.

OUAD CITIES - UNITS 1 & 2 22 Amendment No.

i-4..

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

..m -..

f..

.._m._.___

.m___.__.___

._ _ _ _ ~._ _ ___.___-._____._--

LSSS 2.2 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2,2 LIMITING SAFETY SYSTEM SETTINGS j

i Reactor Protection System (RPS) Instrumentation Seingjnitt i

2.2.A The roactor protection systens instrumentation sotpoints shall bo sot consistent with the Trip Setpoint values shown in Tablo 2.2.A 1, APPLICABILITY: As shown in Tablo 3.1.A 1.

l ACTION:

With a reactor protection system instrumentation sotpoint less conservative than the value shown -

in the Trip Sotpoint column of Tablo 2.2.A 1, declare the CHANNEL inoperable and apply the applicable ACTION statement requirement of Specification 3.1.A until the CHANNEL is restored to OPERABLE status with its sotpoint adjusted consistent with the Trip Setpoint value.

1 i

l t

l-I' i

OUAD CITIES UNITS 1 & 2 23 Amendmont No.

,,-+vmv,e mw,-~1wwem

-m-r--e-wme~~a--e.,,--n----

---v---- - -. - - -- - - - - - - - - - - - - - - -

LSSS 2.2 TABLE 2.2.A 1 REACTOR PROTECTION SYSTEtAINSTRUMENTATION SETPOINTS Functional Unit Idp_Entnpiel

1. Intermediate Range Monitor:

a.

Neutron Flux + Hi h s120/125 divisions of full scalo 0

'b.

Inoperativo NA

2. Averago Power Range Monitor:

a.

Setdown Neutron Flux - High s15% of RATED THERMAL POWER b.

Flow Biased Neutron Flux High

1) Dual Recirculation Loop Operation a) Flow Biased s o.58W") + 62%, -

with a maximum of b) High Flow Clamped 5120% of RATED THERMAL POWER

2) Single Recirculation Loop Operation I

a) Flow Biased s 0.58W"'.+ 58.5 %,

with a maximum of b)' High Flow Clamped s116.5% of RATED THERMAL POWER c.

Fixed Neutron Flux Hi h s 120% of RATED THERMAL POWER -

0 1

d.

Inoperative NA

3. Reactor Vossel Steam Dome Pressure - High s 1060 psig.

i

4. Reactor Vessel Water Level - Low 2144. inches abovo top of active fuel

't

5. Main Steam Line isolation Valvo - Closure s10% closed-

6. Main Steam Line Radiation - High s 15 x normffull power'_ background I

(without hydrogen addition)

I a -- W shall be the recirculation loop flow expressed as a percentage of the recirculatior, loop flow which produces a rated core flow of 98 million Abs /hr.-

OUAD CITIES _ UNITS l'& 2

-24 Amendment No.

m.

LSSS 2.2 TABLE 2.2.A 1 (Continued)

HEACTOR PROTECTION SYSTEM INSIBUMENTATION SETPOINTS EtinchecaLUnit Trin Setnoi_nt

7. Drywoil Prorsuro High s 2.b psig

8. Scram Dischargo Volt.mo Water level Hi0h:

s40 gallons

9. Turbino Stop Valve - Closure s 10% closed

10. Turbine EHC Control Oil Pressuro. Low

. 2 900 psi 0

11. Turbino Control Valvo Fast Closure a460 psi 0 EHC fluid prost'are 12, Turbino Condensor Vacuum Low a 21 inches Hg vacuum

13. Reactor Mode Switch Shutdown Position NA Y

14. Manual Scram NA QUAD CITIES - UNITS 1 & 2 25 Amendment No.

I

I P

SAFETY LIMITS B 2.1 r

BASES j

2J SAFETY LIMITS The Specificat!ons in Section 2.1 establish operating parameters to assure that specified acceptable fuel do$1 n limits are not exceeded during steady state operation, normal operational 0

transients, and anticipated operational occurrences (AOOs). These paramotors are based on the Safety Limits requirements stated in the Code of Federal Regulations,10 CFR 50.36(c)(1):

" Safety limits for_ nuclear reactors are limits upon important process variables that are found to be necessary in reasonably protect the integrity of cortain of the physical barriors that guard against the uncontrolled release of radioactivity."

The fuel cladding, reactor pressuro vessel and primary system piping are the principal barriers to the release of radioactive materials to the environs. Safety Limits are established to protect the intogrity of those barriers during normal plant operations and anticipated transients. The fuel cladding integrity limit is set such that no calculated fuel damage would occur as a result of an AOO. Because fuel damago is not directly observable, a step back approach is used to establish a Safoty Limit for the MINIMUM CRITICAL POWER RATIO (MCPR) that represents a conservative margin relative to the conditions required to maintain fuel cladding integrity, The fuel cladding is one of the physical boundarios which separate radioactive materials from the environs. The integrity of the fuel cladding is related to its relative freedom from perforations or cracking. Although some corrosion or us9-related cracking may occur during the life of the cladding, fission product migration from this source is incremontally cumulative and continuously measurable, Fuel cladding perforations however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the protection system safety settings. While fission product migration from cladding perforations is just as measurable as that from use-related cracking, the thermally caused cladding perforations signal a threshold beyond-which still greater thermal stresses may cause gtoss rather than ircremental cladding deterioration.

Therefore, the fuel cladding integrity Safety Limit is defined with margin to the conditions which would produce onset c' transition boiling (MCPR of 1.0). These conditions represent a significant departure from the condition intended by design for planned operation. Therefore, the fuel cladding integrity Safety Limit is established such that no calculated fuel damago shall result from an abnormal operational transient. This is accomplished by selecting a MCPR fuel cladding integrity i

Safety Limit which assures that during normal operation and AOOs, at least 99.9% of the fuel rods l

in the core do not experience transition boiling.

Exceeding a Safety Limit is cause for unit shutdown and review by the Nuclear Regulatory-4 l

Commission (NRC) before resumption of unit operation. Operation beyond such a limit may not in itself result in serious consequences but it indicates an operational' deficiency subject to regulatory review.

)

l i

OUAD CITIES - UNITS 1 & 2 8 2-1 Amendment No.

= -. - -.

.. a --,

-- +

a --

., z;

.: w=

SAFETY LIMITS B 2.1 BASES 2dA ltiEfMAL _POWERJg1Y_fttuuro. or Lg.yLfirm This fuel cladding integrity Safety Limit is established by establishin0 a limitin0 condition on core THERMAL POWER developed in the following method. At pressures below 800 psia (~ 785 psio),

the core elevation pressure drop (0% power,0% flow) is greator than 4.56 psi. At low powers and flows, this pressura differentialis maintained in the bypass to0 on of the core. Since the i

pressure drop in the bypass region is essentially all clovation head, the core pressuro drop at low puwers and flows will always be greator than 4.50 psi. Analyses show that with a bundlo flow of 3

28 x 10 lb/hr bundio pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.56 psi driving head will be greator than 25 x 10'Ib/hr.

Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicato that the fuel assembly critical power at this flow is approximately 3.35 MWt. At 25% of RATED THERMAL POWER, the peak poworod bundle would have to be operating at 3.86 times the average powered bundle in order to achieve this bundle power. Thus, a core thermal power limit of 25% for reactor pressures below 785 psig is conservativo.

2d3 THERMAL POWER.lliqh Pressure end Hinh Fig.yt This fuel claddin0 ntegrity Safety Limit is set such that no (mechanistic) fuel damage is calculated i

to occur if the limit is not violated. Since the paramotors which result in fuel damd0o are not directly observablo during reactor operation, the thermal and hydraulic conditions resulting in departure from nucleato boiling have been used to mark the bo0 nnin0 of the region whero fuel i

damage could occur. Although it is recognized that a departure from nucloate boilin0 would not necessarily result in dama00 to BWR fuel rods, the critical power ratio (CPR) at which boilin0 transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in rnonitorin0 the core operatin0 state and in the proceduros used to calculato the critical power result in an uncertainty in the value of the critical power. Thorofore, the fuel cladding integrity Safety Limit is defined as the CPR in the limitin0 uel assembly for which more f

than 99.9% nf the fuel rods in the core are expected to avoid boning transition considerin0 tho l

power distribution within the core and all uncertainties.

t 4

The mar 0 n betwoon a MCPR of 1.0 (onset of transiti3n boilin0) and the Safety Limit,is derived i

from a detailed statistical analysis which considers the uncertainties in'monitorin0 the core operatin0 stato, includin0 uncertainty in the critical power correlation. Because the transition boiling correlation is based on a significant quantity of practical test data, there is a very high confidence that operation of a fuel assembly at the condition where MCPR is equal to the fuel-cladding integrity Safety 1.imit would not produce transition boilin0. In addition, during sin 0 0

-j 1

recirculation loop operation, the MCPR Safety Limit is increased by 0.01 to conservatively account for increased uncertainties in the core flow and VIP measurements.

However,if transition boiling woro to occur, cladding perforation would not necessarily be expected. Si0nificant test data accumulated by the NRC and private organizations indicate that the use of a boiling transition limitation to protect a0ainst claddin0 failure is a very conservative OUAD CITIES - UNITS 1 & 2 B 2-2 Amendment No.

l c__~___,__.-.

22 2.

SAFETY LIMITS B 2.1 BASES approach. Much of the data indicates that BWR fuel can survive for an extended period in an environment of transition boiling.

21C Reactor Coolant Systerr Pressurg The Safety Limit for the reactor coolant system pressure has boon selected such that it is at a pressure below which it can be shown that the integrity of the system is not endangered. The reactor coolant system integrity is an important barrier in the prevention of uncontrolled release of fission products. It is essential that the integrity of this system be protected by ostablishing a -

pressure limit to bo observed for all operating conditions and whenover thoro is irradiated fuelin the reactor vessel.

The reactor coolant system pressuto Safety Limit of 1345 psig, as measured by the vessel steam space pressure indicator, is equivalent to 1375 psig at the lowest olevation of the reactor vessel.

The 1375 psig velue is derived from the design pressoros of the reactor pressure vossol and coolant system piping. The respectivo design pressures are 1250 psig at 575 F and 1175 psig at 500"F. The pressure Safety Limit was chosen as the lower of the pressure transients permitted by the applicable design codos, ASME Boiler and Pressure Vessel Code Section !!! for the pressure vessel, and USASI B31.1 Codo for the reactor coolant system piping. The ASME Boiler and Pressure Vessel Codo permits pressure transients up to 10% over design pressure (110% x 1250

= 1375 psig), and the USASI Codo permits pressure transients up to 20% em design pressure (120% x 1175 = 1410 psig). The Safety Limit pressure of 1375 psig is referenced to the lowest elovation of the reactor vessel. The design pressure for the recirculation suction line piping (1175 psig) was chosen relative to the reacor vessel design pressure. Demonstrating compliance of peak vosso! pressure with the ASME overpressure protection limit (1375 psig) assures compliance of the suction piping with the USASIlimit (1410 psig). Evaluation methodology to assure that this Safety Limit pressure is not exceeded for any reload is documented by the specific fuel vendor. The design basis for the reactor pressure vessel makes evident the substantial margin of protection against failure at the safety prossure limit of 1375 psig. The vessel has been designed for a general membrano stress no greater than 26.700 psi at an internal pressura of 1250 psig; this is a factor of 1.5 below the yield strength of 40,100 psi at 575 F. At the pressure limit of 1375 psig, the general membrano stress will only be 29,400 psi, still safely below the yield strength.

The relationships of stress levels to yield strength are comparable for the primary system piping and providss similar margin of protection at the established pressure Safety Limit.

The normal opersf'ig pressure of the reactor coolant system is nominally 1000 psig. Both, pressure relief and safety relief valves have been installed to keep the reactor vessel peak pressure below 1375 psig. However no credit is taken for rol of valves during the postu!ated full closure of all MSIVs without a direct (valve position switch) scram. Credit, however, is *aken for the neutron flux scram, The indirect flux scram and safety valvo actuation provido adequato margin below the allowable peak vessel pressure of 1375 psig, OUAD CITIES - UNITS 1 & 2 B23 Amendment No.

SAFETY t.lMITS B 2.1 BASES Zla ficactor VencLWMcLLWCl With fuelin the reactor vessel during periods when the reactor is shutdown, consideration must also be given to water level requirernents due to the olioct of decay heat. If reactor water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in core cooling capability could lead to elevated cladding temperatures and cladding perforation. The core will be cooled suf ficiently to prevent cladding melting should the water level be reduced to two thirds of the core hei ht. The Safety Limit has U

been established at 12 inches above the top of the active irradiated fuel to provide a point which can be monitored and also provido adequate margin for effectivo action. The top of active fuelis 360 inches above vessel zero.

QU AD CITIES - UNITS 1 & 2 B24 Amendment No.

LSSS B 2.2 BASES L2 LIMITING SAFETY SYSTEM SETTINGS The Specifications in Section 2.2 ostablish operational sottings for the reactor protection system instrumentation whicn initiates the automatic protective action at a level such that the Safety Limits will not be exceeded. These sottings are based on the Limiting Safety System Settings requirements stated in the Code of Federal Regulations,10 CFR 50.36(c)(1h

" Limiting safety system settings for nuclear _ reactors are settings for automatic protective devices folated to those variables having significant safety functions. Whore a limiting safety system setting is specified for a variable on which a safety limit has boon placod, the setting must be so chosen that automatic protective action will correct the abnormal situation before a ssfoty limit is exceeded. "

L2d Reactor Protection System Instrumentation Setonints The Reactor Protection System (RPS) instrumentation setpoints specified in the table are the values at which the reactor scrams are sot for each paramotor. The scram settings have beor.r.olected to ensure that the reactor core and reactor coolant system are prevented from exceeding their Safety Limits during normal operation and design basis anticipated oporational occurrences and av,ist in mitigating the consequences of accidents. Conservatism incorporated into the transient analysis is documented by each approved fuel vendor. The bases for individual scram sotting: are di cussed -

in the following paragraphs.

1.

Intermediate Ranao Monitor. Neutron Flux Hiah The IRM system consists of eight chambers, four in each of the reactor protection system logic CHANNELS. The IRM is a 5 decado,10 range, instrument which covers the range of power lovel between that covered by the SRM and the APRM. The IRM scram setting at 120 of 125 divi?lons is active in each range of the IRM. For example, if the instrument were on Range 1, the scram setting would be 120 divisions for that rango: likewise,if the instrument woro on Rarge 5, the scram would be 120 divisions on that range. Thus, as the IRM is ranged up to accommodate tho increase in power lovel, the scram setting is also ranged up.

The most significant sources of reactivity chango during the power increase are duo to controt rod withdrawal. in ordts to ensure that the IRM provides adequate protection against the single rod-withdrawal error, a range of rod withdrawal events has been analyzed. This analyais included -

starting the event at various power levels. The most sovere caso involves an initial condition in which the reactor b just subcritical and the IRM system is not yet on scale.

Additional conservatism was taken in this analysis by assuming that the IRM CHANNEL closest to the withdrawn rod is bypassed. The results of this analysis show that the reactor is scrammed and peak power is limited to 1% of rated power, thus maintaining MCPR above the fuel cladding integrity Sefuty Limit. Based on the above analysis, the IRM providos protection against local 3

p QUAD CIT!ES - UNITS 1 & 2 B 2-5 Amendment No.

lt -

=

l LSSS B 2.2 BASES contrcl rod viithdrawal errors and continuous withdrawal of control rods in the sequence and providos backup protection for the APRM.

2.

AY01fl01LEILYt0dDDGo Monitgi For operation at low pressure and low flow during Startup, a reduced power level, i.e., sotdown, APRM scram sotting of 15% of RATED THERMAL POWER providos adequato thermal margin betwoon the setting and the Safety Limit. The margin is adequate to accommodato anticipated maneuvors associated with power plant startup. Effects of increasing pressure at zero or low void content are minor; cold water from sourcos availab'o during startup are not much colder than that already in the system; temperature coef ficients are small; and, control rod patterns are constrained to be uniform by operating procedures backed up by the rod worth minimizer. Of all possible sources of reactivity input, uniform control rod withdrawalis the most probable cause of significant power rise. Becauso the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and becauso soveral rods must be moved to chango power by a significant percentago of rated power, the rato of power riso is very slow, Generally, the heat flux is in near equilibrium with the fission rato, in an assumed uniform rod withdrawal approach to the scram sotting, the rate of power rise is no more than 5% of RATED THERMAL POWER per minuto, and the APRM system would be more than adequate to assure a scram before the power could exceed the Safety Limit. The 15% APRM setdown scram setting remains activo until the modo switch is placed in the Run position.

The ave"go power rango monitoring (APRM) system, which is calibrated using heat balanco data taken dunng steady stato conditions, also provides a flow biased neutron flux which roads in porcent of RATED THERMAL POWER. Because fission chambers provido the basic input signals, the APRM system responds directly to averago neutron flux. During transients, the instantaneous rate of heat transfer from the fuel (reactor thermal power)is less than the instantaneous neutron flux due to the timo constant of the fuel. During abnormal operational transients, the thermal power of the fuel will be loss than that indicated by the noutron flux at the scram setting.

Analyses demonstrato that, with a 120% scram setting for dual recirculation loop operation, or with a 116.5% scram setting for single recirculation loop operation, none of the abnormal operational transients analyzed violates the fuel cladding integrity Safety Limh, and thoto is a substantial margin from fuol damage. One of the noutron flux scrams is flow dependent untilit reachos the applicablo setting where it is " clamped" at its maximum allowed value. The use of the flow referenced neutron flux scram setting providos additional margin beyond the use of a the fixed high flux scram setting alono.

An increase in the APRM scram setting would dect,ase the margin present beforo the fuel cladding integrity St#ty Limit is reached. The APRM scram setting was determined by an analysis of margins requied to provido a reasonable range for maneuvering during operation. Reducing this operating margin would increase the frequency of spurious scrams, which have an adverso effect on reactor safety because of the resulting thermal stressos. Thus, the APRM scram setting was selected because it provides adequate margin for the fuel cladding integrity Safety Limit, yet allows operating margin that reduces the possibility of unnecessary scrams.

QUAD CITIES - UNITS 1 & 2 B26 Atandment No.

I

t i

i LSSS B 2.2 l

BASES l

l During single recirculation loop operation, the normal drive flow relationship is altered as a resu..

reverse flow through the idle loop jet pumps when the active loop recirculation pump speed is above approximately 40% of rated. The core receives less flow than would be predicted based upon the dual recirculation loop drive flow to core flow relationship, and the APRM flow biased scram settings (nust be altered to condnue to provide a reactor scram at a conservative neutron flux.

The scram setting must also be adjusted to ensure that the LHGR transient limit is not violated for any power distribution. The scrarn setting is adjusted in accordance with Specification 3/4.11.B in i

i order to maintain adequate margin for the Safety Limit and yet allow operating margin sufficient to reduce the possibility of an unnecessary shutdown. The adjustment may also be accomplished by increasing the APRM gain. This provides the same degree of protection as reducing the scram settings _by raising the initial APRM readings closer to the scram settings such that a scram would be received at the same point in a transient as if the scram settings had been reduced.

3.

Reactor Vessel SIDEDS.Qt.nofanuro Hiah High pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products. A pressure arease while operating will also tend to increase the power of the reactor by compressing vo;ds thus adding reactivity. The scram will quickly reduce the neutron flux, counteracting the pressure increase. The scram setting is slightly higher than the operating pressure to permit normal operation t eithout spurious scrams. The scram setting provides for a wide margin to the maximum allowable design pressure and takes into.

account the location of the pressure measutoment (reactor vessel steam space) compared to the highest pressure that occurs in tho system during a transient, in compliance with Section ill of the ASME Code, the safety valves must be set to open at no higher than 103% of design pressure, and they must limit the reactor pressure to no more than 110% of design pressure. Both the high neutron flux scram and safety valve actuation are required to prevent overpressurizing the reactor pressure vessel and thus, exceeding the pressure Safety Limit. The pressure scram is available as backup protection to the high flux scram.

Analyses are performed for each reload to assure that the pressure Safety Limit is not exceeded.

4.

Reactor Vessel Water Level Low The reactor vessel water level scram setting was chosen far enough below the normal operating level to avoid spurious scrams but high enough above the fuel to assure that there is adequate protection for the fuel cladding integrity and reactor coolant system pressure Safety Limits. The scram setting is based on normal operating temperature and pressure conditions because the level instrumentation is density compensated.

The scram setting provided is the actual water level which may be different than the water level as.

measured by the instrumentation outside the shroud. The water levelinside the shroud will OUAD CITIES - UNITS 1 & 2 8 2-7 Amendment No.

i l

LSSS B 2.2 BASES decrease as power is increased to 100% in comparison to the level outsiou the shrcud, to a maximum of seven inchos, due to the presauro drop across the steam dryor. Therefore, at 100%

power, an indicated water lovel of + 8 inches watcr levol may be as low as + 1 inches insido the shroud which corresponds to 144 inches above the top of activo fuel ano 504 inches above vessel rnro.

Moji Sloam Lino Is.0ln11pn Valvo - Cloiuta 5.

tL Automatic isolation of the main steam linos is provided to glvo protection against rapid reactor depressurization and cooldown of the vossol. When the main steam line isolation valvos bod n to l

i closo, a scram signal providos for reactor shutdown so that high power operation at low reactor pressures doos not occur. With the scram sotting at 10% valvo closuto (from full open), there is no appreciable increase in noutron flux during normal or inadvertent isolation valvo closure, thus providin0 protection for the fuel claddin0 ntegrity Safety Limit. Operation of the reactor at i

pressures lower than tho MSIV closure sotting requires the reactor modo switch to be in tho Stmtup/ Hot Standby position, whero protection of the fuel claddina intogrity Sciety Limit is provided by the IRM and APRM high noutron ficx scram signals. Thus, the combination of main stoorn lino low pressuro isolation and the isolution valvo closuro scram with the modo switch in the Run position assures the availability of the noutron flux scram protection over tho entito range of applicability of fuel cladding into0rity Safety Limit.

6.

MaiLLS10AULUn(LRh'liadon + Hiall High radiation lovels in the main steam line tunnel above that duo to the normal nitrogen and oxygen radioactivity are an indication of leaking fuel. When hi h radiation is detected, a scram is 0

initiated to miti ato the failure of fuel cladding. Tho scram sotting is high enou0h abovo 0

back round radiation levois to provont spurious scrams yet low enou0h to promptly detect gross 0

failures in the fuel cladding. This settin0 is datormined based on normal full power back round 0

(NFPB) radiation lovels without hydrogon addition. With the injection of hydro 00n into tho feedwater for mitigation of intergranular stress corrosion crackin0, the full power background lovels may be significantly increased. The sotting is suf ficiently high to allow the injection of hydrogon _

without requiring an increase in the sotting. This trip function providos an anticipatory scram to limit offsite dose consequences, but is not assumed to occur in the analysis of any design basis event.

QUAD CITIES UNITS 1 & 2 B28 Amendment No.

I z.-.-.-

LSSS B 2.2 BASEG 7.

Drywell Pressure Hiuh High pressure in the drywell could indicato a break in the primary pressure boundary systems of a loss of drywell cooling. Therefore, pressure sensing instrumoritation is provided as a backup to the water levelinstrumentation. The reactor is scrammed on high pressure in order to minimize tho possibility of fuel damage and reduce the amount of energy being added to the coolant and the 1

primary containment. The scram setting was selected as low as possible without causing spurious scrams.

8.

Scram Discharoe Volume Water t.evel-Hiah The control rod drive scram system is designed so that all of the water which is discharged from the reactor by a scram can be accommodated in the discharge piping. A part of this system is an individual instrument volume for each of the scram dischargo volumes. Those two instrument volumes and their piping can hold in excess of 90 gallons of water and are the low point in the piping. No credit was taken for the instrument volumes in the deaign of the discharge piping relative to the amount of water which must be accommodated during a scram. During normal operations, the scram discharge volumes are empty; however, should either scram dischargo volume accumulate water, the water discharged to the piping frum the reactor during a scram may net be accommodated which could result in slow scram timos or partial or no cor. trol rod insertion.

To preclude this occurrenco, level switches have been installed in both instrument volumes which will alarm and scram the reactor while sufficient volume remains to accommodate the discharged water. Diverse level sensing methods have been incorporated into toa design and logic of the system to prevent common mode f ailure. The setting for this anticipatory scram signal has been chosen on the basis of providing sufficient volume remaining to accommodate a scram, even with 5 gpm leakago per drive into the scram discharge volumo. As indicated above, there is sulficient volume in the piping to accommodate the scram without impairment of the scram times or the amount of insertion of the control rods.

9.

Iuddne Ston Valve - Closure The turbino stop valve closure scram setting anticipates the pressure, neutron flux, and heat flux increase that could result from rapid closure of the turbine stop valves. With a scram setting of 10% of valvo closure from full open, the resultant increase in surface heat flux is limited such that MCPR rernains above the fuel cladding integrity Safety Limit, even during the worst-case transient that assumes the turbine bypass f ails to operate.

10.

Turbino EHC Control Oil Presr - Low The turbine EHC centrol system operates using high pressure oil. There are several points in this oil system whero a loss of oil pressure could result in a fast closure of the turbine control valves.

This fast closure of the turbino control valves is not protected by the turbine control valve fast OUAD CITIES - UNITS 1 & 2 B 2-9 Amendment No.

LSSS B 2.2 BASES clusure scram sinco failure of the oil system would not result in the fast closure solenoid valvos being actuated. For a turbino control valvo fast closure, the coro would be protected by the APRM and reactor high pressure scrams. However, to provido the samo margins as provided for the generator load rejection on fast closure of tho turbino control valvos, a scram has boon added to the reactor proaction system which sensos failure of control oil pressure to the turbino control system. This scram anticipatos the pressure transiont which would be caused by imminent control valvo closure and results in reactor shutdown before any significant increase in neutron flux occurs. The transient responso is very similar to that resulting from the turbino control valve fast closure scram. However, since the control valves will not start to close until the fluid pressuro is approximately 600 psig, the scram on low turbine EHC control oil pressure occurs well before turbino control valvo closuro begins. The scram sotting is high enou0h to provido the necessary anticipatory function and low enou0h to minimize the number of spurious scrams.

11.

IwhioLCpntro! Va!YiLEMLCicSwu The turbino control valvo fast closure scram is provided to anticipato the rapid increase in pressuro and neutron flux resulting from fast closure of the turbino control valvos duo to a load rejection and subsequent failuto of the bypass valves; i.e., MCPR tomains above the fuel claddin0 ntegrity i

Safety Limit for thic transient, For the load rejection without bypass transient from 100% power, the peak hoat flux (and thoroforo LHGRIincreases on the order of 15% which providos a wido marD n to the value correspondin0 to 1% plastic strain of the claddin0 i

The scram setting based on EHC fluid pressure was developed to ensuia that the pressure svutch is actuated prior to the closure of the turbino control valvos (at approximately 400 psig EHC fluid pressure), yet assure that the system is not actuated unnecessarily due to EHC system pressure transients which may causo EHC system pressure to momentarily decrease.

12.

Turbino Condensor Vocwm - Low t.oss of condonsor vacuum occurs when the condonsor can no longer handle the heat input, Loss l

r condan vacuum initiates a closure of the turbino stop valvos and turbine bypass valvos which l

imina~es the heat input to the condensor. Closure of the turbino stop and bypass valvos causes a essure transient, neutron flux riso and an increaso in surface heat flux. To prevent the fuel aladdin 0 integrity Safety Limit from being excouded if this occurs, a roactor scram occurs on turbino stop valvo closuro.. The turbino stop valvo closure scram function alono is adoquate to provent the fuel cladding integrity Safety Limit from being excoodod,in the ovent of a turbino trip i

transient with bypass closuro. The condensor low vacuum scram is anticipatory to the stop valvo closure scram and causes a scram before the stop valvos (and bypass valves) are closed and thus, the re# ting transient is loss sevoro.

i QUAD CITIES - UNITS 1 & 2 B 2-10 Amendment No.

---.-m

~

~- -- -.-.

m

__.-..,.__-,..,_______m~-_

-. ~~-.

-m..-.-.-

m

LSSS B 2.2 DASES J

13.

Huactor Modo Switch Shutdown Position The reactor mode switch Shutdown position is a redundant CHANNEL to the automatic protective instrumentation CHANNEL (s) and provides additional manual reactor scram capability.

14.

Manual Scram The manual scram is a redundant CHANNEL to the automatic protective instrumentation CHANNEL (s) and provides manual reactor scram capability.

QUAD CITIES - UNITS 1 & 2 a 211 Amendment No.

A'lTACllMENT 4 EXISTING TECHNICA'L SPECIFICATIONS Technical Specification 2.0

" SAFETY LIMITS AND LIMITJNG SAFETY SYSTEM SETTMGS"

ATTACHMENT 4 DELETION OF CURRENT TECHNICAL SPECIFICATIONS Tnis technical specification arnendment will replaco the current sections 1.1/2.1 and 1.2/2.2. Fuel Cladding Integrity and Reactor Coolant System, for the Dresden Unit 2 and Unit 3 Technical Specifications. The specifications are replaced in its entirety with revised pages that combine the Unit 2 and Unit 3 specifications.

Delete the followin0 pages:

DPR 19 DPR 25 1/2.11 1/2.11 1/2.12 1/2.1 2 1/2.1 3 1/2.1 3 1/2.1 4 1/2.1 4 1/2.1 6 1/2.1 6 B 1/2.1 0 B 1/2.1 0 B 1/2.1 7 B 1/2.1 7 B 1/2.1 8 B 1/2.18 81/2.19 81/2.19 81/2.110 B 1/2.1 10 B 1/2.1 11 B 1/2.1 11 B 1/2.1 12 B 1/2.1 12 B 1/2.1 13 B 1/2.1 13 B 1/2.1 14 B 1/2.1 14 8 1/2.1 15 B 1/2.1 15 B 1/2.1 16 8 1/2.1 16 B 1/2.1-17 B 1/2.1 17 1/2.2 1 1/2.2 1 B 1/2.2-2 B 1/2.2 2 B 1/2.2 3 0 1/2.2 3 I

B 1/2_2 4 91/22-4 l

l

ATTACHMENT 4 DELETION OF CURRENT TECHNICAL SPECIFICATIONS This tochtilcal specification amendment will replace the current sections 1.1/2.1 and 1.2/2.2, Fuol Cladding integrity and Reactor Coolant System, for the Ovad Citios Unit 1 and Unit 2 Technical Specifications. The specifications are replaced in its entirety with revised pages that combino the Unit 1 and Unit 2 specifications.

Doloto the following pa00s:

1 DPR 29 _j DPR - 30 _.

1.1/2.1 1 1.1/2.1 1 i

1.1/2.1-2 1.1/2.1 2 r

1.1/2.1 3 1.1/2.1 2a 1.1/2.14 1.1/2.1 3 1.1/2.15 1.1/2.1,4 1.1/2.10 1.1/2,16 1.1/2.1 7 1.1/2.1 -0 1.1/2.1 B 1.1/2.1 7 1,1/2.1 9 1.1/2.1 7a 1.1/2.1 10 1.1/2.18 e

1.1/2.1 11 1.1/2.1 -9 1.1/2.1 < 12 1,1/2.1-1o 1.1'2.1 13 1.1/2.1 11 1.1/2.1 14 1.1/2.1 12,

1.1/2.1-16 Figure 2.1 1 '

1.1/2.1 10 Figure 2.13 1.1/2.1 17 1.2/2.2-1 Figure 2.1 1 1.2/2.2 2 f

Figure 2.1 3 1.2/2.2-2n 1.2/2.2 1 1.2/2.2-3 1.2/2.2 2 7

J2/2.23 l

_ L212.z t u

l' l

.-,,,.,,m.,

.. -, ~,.,,. -

_. ~,.

,.,-f

-..,.y.,,

-.w,-.4-,4_~,,

..-,.w.'

A'lTACIIMENT 5 DRESDEN 2/3 DIFFERENCES Technical Specification 2,0

" SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS" l

t l

l l

,8 l-

-.- -. - ~.. --

i ATTACHMENT 5 l

COMPARISON OF DRESDEN UNIT 2 AND UNIT 3 7dCHMCAL SPECIFICATIONS FOR THE IDENTIFICATION OF TECHr$.tCAL DlFFERENCES SECTION 1.1/2.1

" FUEL CLADDING INTEGRITY" Commonwealth Edison has conducted a comparison review of the Dresden Unit 2 and Unit 3 Technical Specifications to identify any technical differences in support of combining the-Technical Specifications into one (e:umont. The intent of the review was not to identify any differences in presentation stylo (0 0. table formats, uso of capital letters, etc.),

punctuation or spellin0 errors, but rather to identify areas which the Technical Specifications are technically or administratively dif ferent, The review of Section 1.1/2.1 " Fuel CladdinD Integrity" reveale' one technical difference:

The MCPR Safety Limit for Unit 2 is different than tho MCPR Safety Limit for Unit

3. This duo to the different resident fuel types in each of the Dresden Units and the dif ferenco is retained in the proposed specifications.

i L

I i

l.

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

=- -

ATTACHMENT 5 COMPARISON OF DRESDEN UNIT 2 AND UNIT 3 1ECHNICAL SPECIFICATIONS FOR THE IDENTIFICATION OF TECHNICAL DIFFERENCES SECTION 1.2/2.2

" REACTOR COOLANT SYSTEM" Commonwealth Edison has conducted a compcrise1 review of seesden Unit 2 and Unit 3 Technical Specifications to identify any *.echnica. Jifferences in support of combining the Technical Specifications into one t.acument. The intent of the review was not to identify any differences in presentation style (e.g. table formats, use of capital letters, etc.),

punctuation, or spelling errors but rather identify areas which the Technical Specificat.;ns arn technically or admisiistratively different.

The review of Section 1.2/2.2 " Reactor Coolant System" did not reveal any technical difforences.

ATI'ACIIMENT 5 QUAD CITIES 1/2 DIFFERENCES Technical Specification 2.0

" SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS" l

ATTACHMENT 5 COMPARISON OF QUAD CITIES UNIT 1 AND UNIT 2 TECHNICAL SPECI.'ICATIONS FOR THE IDENTIFICATION OF TECHNICAL DlFFERENCES SECTION 1.1/2.1

" FUEL CLADDING INTEGRITY" Commonwealth Edison has conducted a comparison review of the Guad Cities Unit 1 and Unit 2 Technical Specifications to identify any technical differences in support of combining the Technical Specifications into one document. The intent of the review was not to identify any differences in presentation style (e.g, table formats. use of capital letters, etc.), punctuation or spelling errors, but rather to identify areas which the Technical Specifications are technically or administratively different.

The review of Section 1.1/2.1 " Fuel Cladding Integrity" revealed the following technical difforences:

The third paragraph, last sontence on page 1,1/2.1-6 (DPR-29) states " Basis of the values derived for this safety limit for each fuel type is documented in Reference 1." The Unit 2 Technical Specifications states, "..is documented in References 1 and 2." NEDO-24259 A (Reference 2 in the current Quad TS) contained information concerning the usa of Mrier fuel. The latest revision of NEDE-24011-P-A contains the information regarding barrier fuel which was previously only contained in NEDO-24259-A. As a result, the reference can be deleted.

The last sentence of paragraph B on page 1.1/2.1-9 (DPR30) states, "As with the scram setting, this may be accomplished by adjusting the APRM gains." This information is not contained in the Unit 1 Tcu,Eyf C,,adi; cations. The Unit 2 Technical Specification information y;;n % retained in the combination since the information is consistent with tis requirements of Limiting Safety System Setting 2.1.B.

~_-

ATTACHMENT 5 COMPARISON OF QUAD CITIES UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS FOR THE IDENTIFICATION OF TECHNICAL. DIFFERENCES SECTION 1.2/2.2

" REACTOR COOLANT SYSTEM" Commonwealth Edison has conducted a comparison review of the Ouad Cities Unit 1 and Unit 2 Technical Specifications to identify any technical differences in support of combining the Technical Specifications into one document. The intent of the review was not to identify any differences in presentation style (e.g. table formats, use of capital letters, etc.), punctuation, or spelling errors but rather identify areas which the Technical Specifications are technically or administratively different.

The revicw of Section 1.2/2.2 " Reactor Coolant System" did not reveal any technical differences.

.i 2

i l

. = -

ATI'ACIIMENT 6 SIGNIFICANT HAZARDS CONSIDERATIONS AND ENVIRONMENTAL ASSESSMENT EVALUATION Technical Specification 2.0

" SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS" l

l l

ATTACHMENT 6 EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION Commonwealth Edison has evaluated this proposed amendment and determined that it involves no significant hazards consideratiun. According to 10 CFR 50.92(c), a proposed amendment to an operating license involves no significant hazards consideation if operation of the f acility, in accordance with the proposed ame.6 ment, would not:

1) Involve a significant increase in the probability or consequences of_ an accident previously evaluated; or

2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or

3) involve a significant reduction in a margin of safety.

Tb proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated because:

The proposed changes to Specifications 1/2.1 and 1/2.2 to delete the prosent Applicability and Objective sections represent administrative changes to format and presentation of material. The proposed changes provide the user with a format that will allow better access to needed intnrmation and provides concise Safety Limit.

Limiting Safety System Settings, Apphcability and Action requirements. The additions of Applicability and Action requirements represent clarification of intended requirements that do not prosently state all required conditions of operability or provide clearly stated Action statements if the requirements are not met. The -

combining of the two sections and the added requirements follow STS guidelines that are in use at many operating BWRs with similar design and operating l

configurations as Dresden and Quad Cities Stations. Operability requirements for Safety Limits have been chosen to reflect only those Operational Modes where the Safety Limits apply. Operability requirements for Limiting Safety System Settings are already stated in other sections of the Technical Specifications, thus reference to the appropriate operability requirement is made rather than repeating the requirement in the Limiting Safcty System Setting Specification.

Deletion of the Power Transient Safety Limit does not impact any safety analyses.

The safety analyses assume the Reactor Protection System (RPS) operates as designed and the reactor scrams when the neutron flux exceeds the limiting safety system setting. The proposed Technict.1 Specifications will continue to provide a highly reliable system to operate as assumed in the safety analyses. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The reactor water levellow scram setpoint is changed (for Quad Cities) to be consistent with other reactor water level setpoints in the Technical Specifications and the STS. The setpoint is equivalent to the current requirement but is expressed as the reactor water level above the top of active fuel.

ATTACHMENT 6 The scram discharge volume scram levelis converted for Dresden Unit 2 and Unit 3 to gallons to be consistent with the Quad Cities Units. The proposed setpoints are consistent with the current specifications. The change in the units does not represent a change in the physical setpo:nt.

The proposed change to delete the APRM Downscale Scram trip function for Quad Cities has been evaluated by Commonwealth Edison and General Electriu and previously approved for Dresden Station. The events of concern with respect to the APRMllRM companion trip are the Control Rod Drop Accident and the low powe Rod Withdrawal Error. The FSAR and reload safety analyses do not credit this scram function in the termination of either of thase events. Since this scram function is not credited in the termination of these events, the elimination of this scram function has no adverse effect of previously evaluated accidents.

The change to the low condenter vacuum scram setpoint kom 23 inches of Hg to 21 inches of Hg is consistent with an identical change made to Quad Cities Units 1 and 2. The low condenser vacuum scram is an anticipatory scram and is not credited in any transient analysis. Thus the reduction in the setpoint will not affect any transient analysis.

The proposed changes do not alter the intent of existing setpoints or accident assumptions and follow existing requirements at other operating BWRs for operability and Action statements. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed changes do not ueate the possibihty of a new or different kind of accident from any previously evaluated because:

The proposed administrative changes to the format and arrangement of material do not affect technical requirements or assumptions of any potential accident and; therefore, cannot create the possibility of a new or different kind of accident from any previously evaluated.

The proposed addition of Applicability and Action requirements enhance the understanding and usability of the Technical Specifications and thus represent an improvement over present specifications. New requirements are modeled after those in use at operating BWRs and do not represent requirements that will adversely affect potential accident analyses or assumptions. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

Deletion of the Power Transient Safety Limit does not involve a change in the design or operation of any systems assumed to eperate in the safety analyses.

Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated, i

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

~ - - -

~. -. - -

- - - ~ ~

f ATTACHMENT 6 The change in the units for the Reactor Water Lovel scram function do not chango any physical plant sotpoints. The setpoint will remain the same but will be expressed as the level above the top of activo fuel. The chango does not create the possibility of a now or different kind of accident, t

The conversion of the Scram Discharge Volume scram sotpoint from inches to gallons does not alter any physical plant setpoints. The setpoint will romain the same but will be expressed in gallons rather than inches. The change will provido consistency between Dresden and Quad Cities.

Tne deletion of the APRM Downscale Scram Trip Function does not introduce any now accident. The limiting accidents, Control Rod Drop, Rod Withdrawal Error, in the operating ro0 ion of transition between the Startup and Run Operational Modos are well understood and ato evaluated in FSAR and reload analysos, Other control-rod initiated events which are le::s limitin0 n this re0 on are subsets of the low i

i power Rod Withdrawal Error event and are bounded by it and the desi n basis 0

Control Rod Drop Accident. General Electric has indicated that, for reactivity insertion mechanisms at very low power, the only offect of the deletion of the APRM downscale scram would be that the initial power lovel could be a few percent lower which would not have a significant offect on the soverity of the -

event, in addition, proper overlap betwoon the IRMs and APRMs is not af focted since the calibration requirements are not being changed.

The change in the low condensor vacuum scram function will not croato the l

- possibility of a now or dif forent kind of accident because the function is not j-recognized in any of the transient analysis. The low condensor vacuum scram function is an anticipatory scram.

i l

The proposed changes do not involve a si nificant reduction in the marD n of safety 0

i I

because:

The proposed administrative chan00s to format, arrangement of material, clarification of requirements and other non technical changes do not affoct any L

safety aspects of the plant and as such can not involve a significant reduction in the mar 0 n of safety.

i The proposed Applicability statements require availability of Safety Limits and-Limiting Safety System SettinOs when required to perform their respective functions. Proposed Actions for Safety Limits allow only 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to be in Hot Shutdown and then reference Specification 6.4 to ensure that proper reports are mado and restart is prohibited until approved by the NRC. Those ' provisions help ensure that present margins are not significantly reduced.

Dolotion of the Power Transient Safety Limit does not impact the mar 0 n assumed -

i in the safety analyses. The safety analysos assume the RPS operatos as designed and the reactor scrams when the neutron flux exceeds the limiting safety system setting. The margins assumed in the design of the RPS and in the safety and

i ATTACHMENT 6 transient analyses calculations have not been revised. Therefore, this change does-not involve a significant reduction in the margin of safety.

The change in units to the Reactor Water Level scram setpoint and the Scram Discharge Volume scram setpoint do not involve a significant reduction in the margin of safety because the changes do not represent a change in the physical setpoints.

The reduction in the Low Condenser Vacuum scram setpoint does not represent a reduction in the margin of safety because the scram is not credited in any transient analysis.

The APRM Downscale Scram Trip Function is not credited in the termination of any FSAR or reload safety analysis event. As such, the elimination of this scram function has no effect on any margin of safety.

l l

{

l-l

-~

ATTACHMENT 6 ENVIRONMENTAL ASSESSMENT STAT EMENT APPLICABILITY REVIEW Ccmmonwealth Edison has evaluated the proposed amendment against the criteria for the identification of licensing and regulatory actions requiring environmental assessmant in accondance with 10 CFR 51.20. It has been determined that the proposed changes meet 1

the criteria for a categorical exclusion as provided under 10 CFR 51.22 (c)(9). This conclusion has been dotormined because the changes requested do not pose significant hazards consideration or do not involve a significant increase in the amounts, and no significant chan0es in the types, of any effluent that may be released offsite. Additionally, this request does not involve a significant increase in individual or cumulative occupational radiation exposure. Therefore, the Environmental Assessment Statement is not applicable for these changes, i

(

l

[

h NITACIIMENT 1:

EXECUTIVE SUWWARY Technical Specification 3/4.I1

" POWER DISTRIBUTION LIMITS" t

4 l

-EXECUTIVE

SUMMARY

The Dresden Technical Specification Upgrade Program (TSUP) was conceptualized in response to lessons learned from the Diagnostic Evaluation Team inspection and the frequent need for Technical Specification interpretations. A comparison study of the Standard Technical Specification (STS), later operating plant's Technical Specifications provisions and Quad Cities Technical Specifications was performed prior to the Dresden TSUP ef fort. The study identified potential improvements in clarifying requirements and requicements which are no longer consistent with current industry practices. The Dreden TSUP will enhance the Quad Cities TSUP currently under review by the NRC. As a result of the inconsistencies in the Quad Cities submittal compared to the Standard Technical Specifications (STS), Dresden's submittal will more closely follow the provisions of STS and in conjunction, Quad Cities will amend their submittal so that Quad Cities and Dresden are identical within equipment and plant design. The format for the Dresden TSUP will remain as a two column layout for human factors considerations. Additionally, chapter rganizations will remain essentially unchanged.

t The TSUP is not intended to be a complete adoption for the STS. Overall, the Dresden custom Technical Specifications provide for the safe operation of the plant and therefore, only an upgrade is deemed necessary.

In response to an NRC recommendation, Quad Cities combined the Unit 1 and Unit 2 Technical Specifications into one document. The Dresden Unit 2 and Unit 3 Technical Specifications will also be combined into one document. To accomplish the combination of the Technical Specifications, a comparison of the individual Technical Spacifications was performed to iderlify any technical differences. The technical differences are identified in the proposed amendment package for each section.

The TSUP was identified as a station top priority and is currently contained in the Dresden Management Action Plan (DMAP). TLo TSUP goal is to provide a better tool to station personnel to implement their responsibilities and to ensure Dresden Station is operated in accordance with current industry practices. The improved Technical Specifications provide for enhanced operation of the plant. The program improves the operator's ability to use the Technical Specifications by more clearly defining the Limiting Conditions for Operation and required actions. The most significant improvement to the specifications is the addition of equipment operability requirements during shutdown conditions.

EXECUTIVE

SUMMARY

(continued)

PROPOSED CHANGES TO TECHNICAL SPECIFICATION SECTIOrd 3/4.11. " POWER DISTRIBUTION LIMITS" The current Dresden and Quad Cities Technical Specifications contain Applicability and Objective statements at ths beginning of most sections. The proposed amendment will delete the " Objective" statement and integrates appropriate applicability statements within the specifications. This provides a clarification of the intended requirements and actions which are required when the specification cannot be met.

The proposed Section 3/4.11 is a new section that results in the consolidation and rearrangemcnt of the power distribution limits. The majority of the proposed specifications are currently contained in section 3/4.5, ECCS Systems. The new specifications are -

adopted from the STS,

=

ATTACIIMENT 2 DESCRIPTION OF CHANGES Technical Specification 3/4.1I

" POWER DISTRIBUTION LIMITS"

)

l i

s l '.

i t

l^

i i

l

ATTACHMENT 2 DESCP.lPTION OF AMENDMENT REQUEST The changes proposed in this amendment request sre made to 1) emptcve the understanding end usability of the present technical specifications,2) incorporate technical improvements, and 3) include some provisions from later operating BWR plar.ts.

GENERIC CHANGES The present Dresden and Quad Cities technical specifications contain appCcability and objective statements at the beginning of most sections. These statements are generic in nature and do not provide any useful information to the user of the technical specifications. The proposed change will delete the objective statement and provide applicability statements within each specification similar to the Standard Technical Specifications (STS). The proposed applicability statement to be included in each -

l specification willinclude the reactor operational modes or other conditions for which the Limiting Condition for Operation (LCO) must be satisfied.

The proposed rearrangement of the power distribution limits from current section 3/4.5 to a new section will provide consistency in presentation of material and prescnt the mcterial in a f ashion consistent with the STS. The addition of the applicable operational modes will-provide readily accessible information concerning when the system is required to be operable and when surveillance requirements must be performed.

The proposed section contains several differences between Dresden and Quad Cities as a-result of the different fuel vendors at the two sites. Dresden uses Siemens Nuclear Power fuel and thus has fuellimits defined by Siemens. Quad Cities uses General Electric fuel and uses thermai limits identical to those presented in the STS.

The proposed changes are consistent with the STS and Generic Letter 88-16. Removal of Cycle Specific Parameters From Technical Specifications. Both Dresden and Quad Cities have the cycle specific parameters in a Core Operating Limits Report (COLR),

SPECIFIC CHANGES Proposed specification 3/4.11.A, Average Planar Linear Heat Generation Rate (APLHGR) is a complete adoption of the STS requirements. Dresden APLHGR limits are a function of bundle average exposure versus average planar exposure for General Electric fuel.

Specification 3.11./ requires that all the APLHGR limits specified in the COLR be met in operational mode 1 -hen thermal power is greater than 25%_of rated thermal power, When the condition s not satisfied the action requires that corrective action be initiated within 15 minutes, the APLHGR va!ues restored within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. or thermal power reduced '

below 25% within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The proposed actions are adopted from STS but are separated for clarification purposes. Surveillance Requirement (SR) 4.11.A.1 requires the values of APLHGR be checked at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> af ter a thermal power increase of at least 15% of rated thermal power, and initially and once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating on a limiting control rod pattem. The proposed specifications implement the current specifications for APLHGR in the Dresden and Quad Cities Technical Specifications but do not require the reactor to be placed in cold

ATTACHlWENT 2 shutdown in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> if the APLHGR limit is not restored within the 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The requirement to monitor APLHGR within 1? hours of a power iacrease is a new requirement based on STS. The requirement to monitor APLHGR while operating on a limiting control tod pattom is a new requirement and is consistent with the STS definition for a limiting control rod pattern. Proposed SR 4.11.A.4 states that the piovisions of specification 4.0.D are not apphcable. The provision allows power to be increased above 25% of rated thermal pawer applicability limit prior to performing the SRs as long as the SR timo limits are met. Without the provision, a hold in the power ascension would have to occur prior to exceeding 25% of rated thermal power to perform the required surveillance.

Proposed specification 3/4.11.B, Average Power Range Monitor (APRM) Setpoints, is adopted from STS with several enhanceme7ts. The enhancements are made to avoid duplicatin0 setpoints within the proposed technical specifications and to more clearly delineate when the actions for the specification are to be implemented. This specification is provided to require the APRM gain or APRM flow biased scram and rod block trip setpoints to be adjusted when operating under conditions of abnormal power peakin0 so that acceptable margin to the fuel cladding integrity limits are maintained. Abnormal power peaking is represented when the Maximum Fraction of Limiting Power Density (MFLPD) is greater than the Fraction of Rated Thermal Power (FRTP) for Quad Cities and for Dresden when the Fuel Design Limiting Ratio For Centerline Melt (FDLRC) is greater than 1.0. To maintain the appropriate margin under conditions of abnormal power peaking, either the APRM gain must be adjusted upward or the flow biased neutron flux upscale scram trip and rod block setpoints be reduced. This is accomplished by multiplying the APRM gain or setpoints by a factor that is representative of the reduction in margin to the fuel cladding inte0rity limits. Adjustment to the scram and rod block setpoint are made by multiplying the setpoint by the inverse of the factor for the APRM gains. This f actor will be less than one and thus cause the setpoints to be lowered to maintain the margin. When the reactor is operating with normal peaking (i.e. FDLRC <

1.0 or MFLPD < FRTP) it is not necessary to modify the APRM flow biased scram or rod block setpoints. The proposed actions are adopted from STS but are separated for clarification purposes. The action requires that if FDLRC is greater than 1.0 for Dresden or MFLPD is greater than FRTP for Quad Cities, that within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> the limit is restored, or the APRM setpoints in 2.2.A and 3.2.E are adjusted by the f actors described above, or the APRM gains are adjusted by the f actor described above. If the action provisions are not met, thennal power is required to be reduced to below 25% of rated thermal power. The proposed SR requires that the value of FDLRC (Dresden) or MFLPD and FRTP (Quad Cities) be checked at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> af ter a thermal power increase of 15% of more, and initially and once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating on a limiting control rod pattom. The SR are modified from STS in accordance with the proposed LCO. Proposed S-R 4.11.B.4 is added to stipulate that the provisions of specification 4.0.D are not applicable. Footnote (a) is adopted from the STS but does not restrict the adjustment of the APRM gains to less than 90% of rated thermal power. The footnote implements current requirements for adjusting APRM gains during operation with abnormal power peaking.

Proposed specification 3/4.11.C, Minimum Critical Power Ratio (MCPR) is adopted from STS. The MCPR is required to be equal to or greater than the limit specified in the COLR in operational mode 1 when thermal powu is greater than 25% of rated thermal power.

- ~

ATTACHMENT 2 The actions for MCPR are adopted from the applicab!c RTS actions and are separated for clarification. Proposed action 3.11.C.1 requires that when the MCPR is less than the applicable limit specified in the COLR, that corrective action be initiated within 15 mhutes, the MCPR restored within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or thermal power be reduced to less than 25% of rated within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The proposed specification implements the requirements of the current spncifications but do not require the reactor to be taken to cold shutdown if the MCPR limit is not restored within the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> time frame. The surveillance requirements are adopted from the STS and are based on a scram insertion time value, t The definition o,.

of t,,,is contained in the COLR. Dresden and Quad Cities use different scram insertion values for t,, and they are identified in the COLR Dresden uses a 90% mean insertion value whereas Quad Cities uses a 20% mean incertion value. The difference is a result of the different fuel vendors at Dresden and Quad Cities. The value of MCPR is required to be determined to be greater than the MCPR limit at least every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> af ter completion of a power increase of at least 15% of rated thermal power and initially and once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating on a limiting control rod pattern. The value used in the development of the MCPR limit is required to be verified within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the completion of specification 4.3.D, Scram insertion times. Proposed SR 4.11.C 4 is added to stipulate that the provisions of apecification 4.0.0 are not applicable.

The limits on Linear Heat Generation Rata are different for Dresden and Quad Cities and are discussed individually due to the dif ferences.

Proposed specification 3/4.11.D, Linear Heat Generation Rate (LHGR) for Quad Cities is adopted from the STS. The LHGR is required to be less than the value specified in the COLR when in operational mode 1 and thermal power is greater than 25% of roted thermal power. When the condition is not satisfied the action requires that correctivo action be initiated within 15 minute ~, the LHGR values restored within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or thermal power reduced below 25% within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The proposed actions arc adopted from LTS but are separated for clarification purposes. SR 4.11.D.1 recuires the values of LHGR be checked at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after a thermal power increase of at least 15% of rated thermal power, and initially and once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when thu aactor is operating on a limiting control rod pattern. The proposed specifications implement the current specifications for LHGR in the Quad Cities Technical Specifications.1 The requirement to monitor LHGR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of a power increase is a new requirement based on STS,-The requirement to monitor LHGR while operating on a limiting control rod pattern is a new requirement and is a result of the adoption of the STS definition for a li!niting control rod pattern. Proposed SR 4.11.D.4 is added to stipulate that the provisions of specification 4.0 D are not applicable.

Proposed specification 3/4.11.D, Steady State Linear Heat Generation Rate (SLHGR) for Dresden is retained from the current specifications but reformatted in accordance with the proposed specifications. The specification requires that the SLHGR values be less than the limits specified in the COLR in operational mode 1 with thermal power greater than 25% of rated thermal power. When the condition is not satisfied the action requires that corrective action be initiated within 15 minutes, the SLHGR values restored within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or thermal power reduced below 25% within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The proposed actions are g adopted from STS but are separated for clarification purposes. SR 4.11.D.1 requires the values of SLHGR be checked at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> af ter a thermal

ATTACHMENT 2 power increase of at least 15% of rated thermal power, and initially and once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating on a limiting control rod pattern. The proposed specifications implement the current specifications for SLHGR in the Dresden Technical Specifications. The requirement to monitor SLHGR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of a power increase is a new requirement based on STS. The requirement to monitor ELHGR while operating on a limiting control rod pattern is a new requirement and is a result of the adoption of the STS definition for a limiting control rod pattern. Proposed SR 4.11.D.4 is added to stipulate that the provisions of specification 4.0.D are not applicable.

Proposed specification 3/4.11.E. Transient Linear Heat Generation Rate (TLHGR) for Dresden is retained from the current specifications. The specification requires that the TLHGR values be lass than the limits specified in the COLR in operational mode 1 with thermal power greater than 25% of rated thermal power. When the condition is not satisfied the action requires that corrective action be initiated within 15 minutes, the -

TLHGR values restored within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or thermal power reduced below 25% within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The proposed actions are adopted from STS but are separated for clarification purposes. SH 4.11.D.1 requires the values of TLHGR be checked at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after a thermal power increase of at least 15% of rated thermal power, and initially and once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating on a limiting control rou pattern. The proposed specifications implement the current specifications for TLHGR in the Dresden Technical Specifications. The requirement to monitor TLHGR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of a power increase is a new requirement based on STS.

The requirement to monitor TLHGR while operating on a limiting control rod pattern is a new requirement and is a result of the adoption of the STS definition for a limiting control rod pattern, Proposed SR 4.11.E.4 is added to stipulate that the provisions of specification 4.0.D are not applicable.

The changes proposed to the Bases for proposed Section 3/4.11 are administrative in.

nature and include the capitalization of terms defined in proposed Section 1.0, Definitions.

l l

[

i

ATI'ACIIMENT 3 PROPOSED TECHNICAL SPECIFICATIONS Technical Specification 3/4.11 "EO_WER DISTRIBUTION LIMITS" l

l l

POWER DISTRIBUTION LIMITS APLHGR 3/4.11.A 3.11 - LIMITING CONDITIONS FOR OPERATION 4.11 - SURVEILLANCE REQUIREMENTS A.

AVERAGE PLANAR LINEAR HEAT A.

AVERAGE PLANAR LINEAR HEAT GENERATION RATE GENERATION RATE All AVERAGE PLANAR LINEAR HEAT The APLHGRs shall be verified to be equal GENERATION RATES (APLHGR) for each to or less than the limits specified in the type of fuel as a function of bundle average CORE OPERATING LIMITS REPORT.

exposure shall not exceed the limits specified in the CORE OPERATING LIMITS 1.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,-

REPORT.

2.

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least APPLICABILITY:

- 15% of RATED THERMAL POWER, and OPERATIONAL MODE 1. when THERMAL 3.

Initially and et Jeast once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> POWER is greater than or equal to 25% of when (to teactor is operating with a RATED THERMAL POWER.

LIMITING CONTROL ROD PATTERN for APLHGR.

e_CI1QN1 4.

The provisions of Specification 4.0.D are not applicable.

With an APLHGR exceeding the limits specified in the CORE OPERATING LIMITS REPORT:

1.

Initiate corrective action within 15 minutes, and 2.

Restore APLHGR to within the required limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

With the provisiens of the ACTION above not met, reduce THERMAL POWER to less than 25% of RATED THERMAL POWT" within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

l l

DRESDEN - UNITS 2 & 3 3/d.11-1 Amendment No.

P_QWER DISTRIBUTION LIMITS APRM Setpoints 3/4.11.B 3.11 LIMITING CONDITIONS FOR OPERATION 4.11 SURVEILLANCE REQUIREMENTS B.

Average Power Range Monitor Setpoints B.

Averego Power Range Monitor Setpoints The Average Power Range Monitor (APRM)

The value of FDLRC shall be verified:

gain or setpoints shall be set such that the FUEL DESIGN LIMillNG RATIO FOR 1.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, CENTERLINE MELT (FDLRC) shall be less than or equal 1.0, 2.

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and APPLICABILITY:

3.

Initially and at Icast once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> OPERATIONAL MODE 1, when THERMAL when the reactor is operating with POWER is greater than or equal to 25% of FDLRC greater than or equal to 1.0.

RATED THERMAL POWER.

4 The provisitm of Specification 4.0.D are not applicabl.4.

ACTION:

With FDLRC greater 1.0, initiate corrective ACTION within 15 minutes and within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> either:

1.

Restore FDLRC to within its limit, or 2.

Adjust the flow biased APRM setpoints specified in Specifications 2.2.A and 3.2.E by 1/FDLRC, or 3.

Adjust

• the APRM cain such that the APRM readings are 2100% of the FRACTION OF RATED THERMAL POWER (FRTP) times FDLRC.

With the provisions of the ACTION above not met, reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

a Provided that the adjusted APRM reading does not exceed 100% of RATED THERMAL POWER and a notice of adjustrnent is posted on the reactor control panel.

DRESDEN - UNITS 2 & 3 3/4.11-2 Amendment No.

POWER DISTRIBUTION LIMITS MCPR 3/4.11.C 3.11 - LIMITING CONDITIONS FOR OPERATION 4.11 - SURVEILLANCE REQUIREMENTS C.

MINIMUM CRITICAL POWER RATIO C.

MINIMUM CRITICAL POWER RATIO The MINIMUM CRITICAL POWER RATIO MCPR, with:

(MCPR) shall be equal to or greater than the MCPR operating limit specified in the CORE 1.

t, = 3.50 prior to performance of the OPERATING LIMITS REPORT.

initial scram time measurements for $e cycle in accordance with Specificat'.on 4.3.D,or APPLICABILITY:

2.

t determined within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the conclusion of each scram time OPERATIONAL.'.; ODE 1, when THERMAL surveillance test required by POWER is greater than or equal to 25% of Specification 4.3.D, RATED THERMAL POWER.

shall be determmed to be equal to or i

I greater than the applicable MCPR limit ACTION:

specified in the CORE OPERATING LIMITS REPORT.

With MCPR less than the applicable MCPR limit as determined for one of the 1.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, conditions specified in the CORE l

OPERATING LIMITS REPORT:

2.

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 1.

Initiate corrective ACTION within 15 15% of RATED THERMAL POWER, and minutes, and 3.

Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 2.

Restore MCPR to within the required when the reactor is operating with a limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

LIMITING CONTROL ROD PATTERN for MCPR.

With the provisions of the ACTION above not met, reduce THERMAL POWER to less 4.

The provisions of Specification 4.0.D than 25% of RATED THERMAL POWEH are not applicable.

within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

l i

i l

i DRESDEN - UNITS 2 & 3 3/4.11-3 Amendment No.-

l

i POWER DISTRIBUTION LIMITS SLHGR 3/4.11.D 3.11 - LIMITING CONDITIONS FOR OPERATION 4.11 SURVEILLANCE REQUIREMENTS D.

STEADY STATE LINEAR HEAT D.

STEADY STATE LINEAR HEAT GENERATION RATE GENERATION RATE The STEADY STATE LINEAR HEAT The SLHGR shall be determined to be equal GENERATION RATE (SLHGR) fo, each type to or less than the limit:

of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits 1.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, specified in the CORE OPERATING LIMITS REPORT.

2.

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> af ter completion of a THERMAL POW 7' increase of at loest 15% of RATED THERMAL POWER, and APPLICABILITY:

3.

Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> OPERATIONAL MODE 1, when THERMAL when the reactor is operating with a POWER is greater than or equal to 25% of LIMITING CONTROL ROD PATTERN for RATED THERMAL POWER.

SLHGR.

4.

The provisions of Specification 4.0 D ACTION:

are not applicable.

With a SLHGR exceeding the limits specified in the CORE OPERATING LIMITS REPORT:

1.

Initiate corrective ACTION within 15 minutes, and 2.

Restore the SLHGR to within the required limit within lr ours.

With the provisions of the ACTION above not met, reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

DRESDEN - UNITS 2 & 3 3/4.11 4 Amendment No.

i POWER DISTRIB11 TION LIMITS TLHGR 3/4.11.E 3.11 - LIMITING CONDITIONS FOR OPERATION 4.11 SURVEILLANCE REQUIREMENTS E.

TRANSIENT LINEAR HEAT GENERATION E.

TRANSIENT LINEAR HEAT GENERATION RATE RATE The TRANSIENT LINEAR HEAT The TLHGR shall be determined to be equal GENERATION RATE (TLHGR) for each type to or less than the limit:

of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits 1.

At least once por 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, specified in the CORE OPERATING LIMITS REPORT.

2.

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> af ter completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and APPLICABILITY:

3.

Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> OPERATIONAL MODE 1, when THERMAL when the reactor is operating with a POWER is greater than or equal to 25% of LIMITING CONTROL ROD PATTERN for RATED THERMAL POWER.

TLHGR.

4.

The provisions of Specification 4.0.D ACTION:

are not applicable.

With a TLHGR exceeding the limits specified in the CORE OPERATING LIMITS REPORT:

1.

Initiate corrective ACTION within 15 minutes, and 2.

Restore the TLHGR to within the required limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

With the provisions of the ACTION above not met, reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

DRESDEN - UNITC 2 & 3 3/4.11-5 Amendment No.

POWER DISTRIBUTION B 3/4.11 BASES 3/4.11. A AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50.46. The specification also assures that fuel rod mechanicalintegrity is maintained durin0 normal and transient operations.

The peak cladding temperature (PCT) following a postulated loss-of coolant accident is primarily a function of the aserage heat generation rate of all the rods of a fuel assembly at any axiallocation and is dependent only secondarily on the rod to-rod power distribution within an assembly. The i

peak clad temperature is calculated assuming a LINEAR HEAT GENERATION RATE (LHGR) for the highest powered rod which is equa. to or less than the design LHGR corrected for densification.

The APLHGR limits specified are equivalent to the LHGR of the highest powered fuel rod assumed in the LOCA analysis divided by its local peaking f actor. A conservative multiplier is applied to the LHGR assumed in the LOCA analysis to account for the uncertainty associated with the measurement of the APLHGR.

The calculational procedure used to establish the maximum APLHGR values uses NRC approved calculational models which are consistent with the requirements of Appendix K of 10 CFR Part 50.

The approved calculational models are listed in Specification 6.6.A,4.

The daily requirement for calculating APLHGR when THERMAL POWER is greater than or equal to-25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The requirement to calculate APLHGR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER ensures thermallimits are met after power distribution shifts while still allotting time for the power distribution to stabilize. The requirement for calculating APLHGR after initially determining a LIMITING CONTROL ROD PATTERN exists ensures that APLHGR will be known following a change in THERMAL POWER or power shape, that could place operation above a thermallimit.

3 /4.11.B APRM SETPOINTS The fuel cladding integrity Safety Limits of Specification 2.1 were based on a power distribution which would yield the design LHGR at RATED THERMAL POWER. The flow biased neutron flux -

high scram setting and control rod bloc!. functions of the APRM instruments for both two recirculation loop operation and t Qle recirculation loop operation must be adjusted to ensure that the MCPR does not become less than the fuel cladding safety limit or that a1% plastic strain does

.not occur in the degraded situation. The scram settings and rod block settings are adjusted in accordance with the formula in this specification when the value of FDLRC indicates a higher peaked power distribution to ensure that an LHGR transient would not be increased in the degraded-condition.

DRESDEN - UNITS 2 & 3 B 3/4.11-1 Amendment No.

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

POWER DISTRIBUTION B 3/4.11

. BASES-3/4.11.C

_ MINIMUM CRITICAL POWER RATIO The required operating limit MCPR at steady state operating conditions as specified in Specification-3.11.C are derived from the established fuel cladding integrity Safety Limit MCPR, and an analysis of abnormal operational transients. For any abnormal operating transient analysif, 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 Limit MCPR at any time dwing the transient -

assuming instrument trip setting given in Specification 2.2.

To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting transients have been ana! red to determine which result in the largest reduction in the CRITICAL POWER RATIO (CPR). The type of transients evaluated were change of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest delta MCPR. When added to the Safety Limit MCPR, the required minimum operating limit MCPR of Specification 3,11.C is obtained and presented in the CORE OPERATING LIMITS REPORT.

The steady state values for MCPR specified were determined using NRC-approved methodology listed in specification 6.6.A.4.

The purpose of the reduced flow MCPR curves specified in the CORE OPERATING LIMITS REPORT are to define MCPR operating limits at other than rated core flow conditions. The reduced flow MCPR curves assure that the Safety Limit MCPR will not bo violated.

Since the transient analysis takes credit for conservatism in the scram speed performance, it must be demonstrated that the spacific scram speed distribution 's consistent with that used in the transient analysis. The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> completion time is acceptable due to the relatively minor changes in t,, expected durir'g the fuel cycle.

At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience indicates that the resulting MCPR value has considerable margin. Thus, the-demonstration of MCPR below this power levelis unnecessary; The daily requirement for calculating MCPR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The requirement for c,alculating MCPR after initially.

determining that a LIMITING CONTROL ROD PATTERN exists ensures that MCPR will be known following a change in THERMAL POWER or power shape, regardless of magnitude, that could place operation above'a thermal limit.

DRESDEN - UNITS 2 & 3 B 3/4.11-2 Amendment No.

4

,,~

.~

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

-. ~. - - - -

POWER DISTRIBUTION B 3/4.11 BASES 3 /4.11.D - SIf.ADY STATE LINEAR HEAT GENERATION RATE This specification assures that the maximum STEADY STATE LINEAR HEAT GENERATION RATE in any fuel rod is less than the design STEADY STATE LINEAR HEAT GENERATION RATE even if fuel pellet densitication is postulated. This provides assurance that the fuel end-of-life steady state criteria are met. The daily requirement for calculating SLHGR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distributions shifts are very slow when there have not been significant power or control rod changes. The requirement to calculate SLHGR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER ensures thermallimits are met after power distribution shifts while still allotting time for the power distribution to stabilize. The requirement for calculating SLHGR af ter initially determining a LIMITING CONTROL ROD PATTERN exists ensures that SLHGR will be known following a change in THERMAL POWER or power shape that could place operation above a thermallimit.

3/4.11.E TRANSIENT LINEAR HEAT GENERATION RATE This specification provides assurance that the fuel will neither experience centerline melt nor exceed 1% plastic strain kr transient overpower events beginning at any power and terminating at 120% of RATED THERMAL POWER. The daily requirement for calculating TLHGR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The requirement to _ calculate TLHGR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER ensures thermallimits are met after power distribution shifts while still allotting time for the power distribution to stabilize. The 2

requirement for calculating TLHGR af ter initially determining a LIMITING CONTROL ROD PATTERN exists ensures that TLHGR will be known following a change in THERMAL POWER or power shape that could place operation above a thermallimit.

DRESDEN - UNITS 2 & 3 8 3/4.11-3 Amendment No.

POWER DISTRIBUTION LIMITS APLHGR 3/4.11.A 3.11 - LIMITING CONDITIONS FOR OPERATION 4.11 - SURVEILLANCE REQUIREMENTS A.

AVERAGE PLAN.AR LINEAR HEAT A.

AVERAGE PLANAR LINEAR HEAT GENERATION RATE GENERATION RATE All AVERAGE PLANAR LINEAR HEAT The APLHGRs shall be verified to be equal GENERATION RATES (APLHGR) for each to or less than the limits specified in the -

type of fuel as a function of AVERAGE CORE OPERATING LIMITS REPORT.

PLANAR EXPOSURE shall not exceed tha limits specified in the CORE OPERATING 1.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, LIMITS REPORT.

2.

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least APPLICABILITY:

15% of RATED THERMAL POWER, and OPERATIONAL MODE 1, when THERMAL 3.

Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> POWER is greater than or equal to 25% of when the reactor is operating with a.

RATED THERMAL POWER.

LIMITING CONTROL ROD PATTERN for APLHGR.

ACTION:

4.

The provisions of Specification 4.0.D are not applicable.

With an APLHGR exceecing the limits specified in the CORE OPERATING LIMITS REPORT:

1.

Initiate corrective ACTION within 15 minutes, and 2.

Restore APLHGR to within the required limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

With the provisions of the ACTION above i

not met, reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, l

l l

OUAD CITIES - UNITS 1 & 2 3/4.11-1 Amendment No.

POWER DISTRIBUTI: 1 LIMITS APRM Sotpoints 3/4.11.0 3.11 LIMITING CONDITIONS FOR OPERATION 4.11 SURVEILL

• NCE REQUIREMENTS w

B.

Avorage Power Range Monitor Sotpoints B.

Average Power Rango Monitor Setpoints The Avorago Power Range Monitor (APRM)

The value of MFLPD shall be verified:

gain or setpoints shall be 'ot such that the MAXIMUM FRACTION OF LIMITING 1.

At least onco por 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, POWER DENSITY (MFLPD) shall be less than or equal to the FRACTION OF RATED 2.

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER (FRTP).

THERMAt. POWER increase of at least 15% of RATED THERMAL POWER, and APPLICABILIIYi 3.

Initially and at least onco per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with OPERAllONAL MODE 1, when THERMAL MFLPD greator than or equal to F*lTP, POWER is greator than or equal to 25% of RATED THERMAL POWER.

4.

The provisions of Specification 4.0.D are not applicable.

ACTION:

With MFLPD Jroater than FRTP, initiato correctivo ACTION within 15 minutes and within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> either:

1.

Rostoro MFLPD to within its limit, or 2.

Adjust the flow biased APRM sotpoints specified in Specifications 2.2.A and 3.2.E by FRTP/MFLPD, er 3.

Adjust the APRM gain such that the APRM readings are 2100% of the

MFLPD, With the provisions of the ACTION above not mot, reduce THERMAL POWER to less than 25% of RATED THERf'AL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

a Provided that the adjusted APRM reading does not exceed 100% of RATED THERMAL POWER and a notice of adjustment is posted on the reactor control panel.

QUAD CITIES - UNITS 1 & 2 3/4.11-2 Amendment No.

POWER DISTRIBUTION LIMITS MCPR 3/4.11.C 3.11 LIMITING CONDITIONS FOR OPERATION 4.11 - SURVEILLANCE REQUIREMENTS tm C.

MINIMUM CRITICAL POWER RATIO C.

MINIMUM CRITICAL POWER RATIO The MINIMUM CRITICAL POWER RATIO MCPR, with:

(MCPR) shall bo equal to or greater than the MCPR operating limit specified in the CORE 1.

t,,, = 0.80 prior to performance of the OPERATING LIMITS REPORT.

initial scram timo measutomonts for the cycle in accordance with Specification 4.3.D, or APPLICABILITY:

2 t,,, dotormined within 72 houes of the conclusion of each scram time OPERATIONAL MODE 1, when THERMAL surveillance test required by POWER is greator than or equal to 25% of Specification 4.3.D, RATED THERMAL POWER.

shall be determined to bo equal to or greator than the applicablo MCPR operating ACTION:

limit specified in the CORE OPERATING LIMITS REPORT.

With MCPR loss than the applicablo MCPR operating limit an determined for nno of the 1.

At least once por 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, conditions specified in the CORE OPERATING LIMITS REPORT:

2.

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a l

THERMAL POWER increase of at least 1.

Initiato corrective ACTION within 15 15% of RATED THERMAL POWER, and 3.

Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 2.

Rostoro MCPR to within the required when the reactor is operating with a f

limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

LIMITING CONTROL ROD PATTERN for MCPR.

With the provisions of the ACTION above not met, reduce THERMAL POWER to loss 4.

The provisions of Specification 4.0 D than 25% of RATED THERMAL POWER are not applicable.

within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

i OUAD CITIES - UNITS 1 & 2 -

3/4.11 3 Amendment No.

. u.

n.._,_-

..u..

EQ1yER DISTRIBUTION LIMITS LHGR 3/4.11.D I

l 3.11 LIMITING CONDITIONS FOR OPERATION 4.11 - SURVEILLANCE REQUIREMENTS l

D.

LINEAR HEAT GENERATION RATE D.

LINEAR HEAT GENERATION RATE The LINEAR HEAT GENERATION RATE The LHGR shall be dotormined to be equal (LHGR) for each type of fuel shall not to or loss than the limit:

exceed the limits specified in the CORE OPERATING LIMITS REPORT.

1.

At least onco por 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 2.

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> af ter completion of a AfTLICABILITY:

THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and OPERATIONAL MODE 1, when THERMAL POWER is greater than or equal to 25% of 3.

Initially and at least once por 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> RATED THEPMAL POWER.

when the reactor is operating with a LIMITING CON'ROL ROD PATTERN for LHGR.

ACIlDB 4.

The provisions of Specification 4.0.D With a LHGR excooding the limits specified are not applicable, in the CORE OPERATING LIMilS REPORT:

1.

Initiato correctivo ACTION with n 15 minutes, and 2.

Rostore the LhtaR to within the required limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

With the provisions of the ACTION above not met, reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

QUAD CITIES - UNITS 1 & 2 3/4.11 4 Amendment No.

1 POWER DISTRIBUTION LIMITS B 3/4.11 i

BASES 3/4.11. A AVERAGE PLANAR LINEAR HEAT _ GENERATION RAIE This specification assures that the peak cladding temperature following the postulated design basis loss-of. coolant accident will not exceed the limit specified in 10 CFR 50.46. The specification also assures that fuel rod mechanicalintegrity is maintained during normal and transient operations.

l The peak cladding temperaturo (PCT) following a postulated loss-of coolant accident is primarily a function of the average heat generation rato of all the rods of a fuel assembly at any axiallocation and is dependent only secondarily on the rod to-rod power distribution within an assembly. The peak clad temperature is calculated assuming a LINEAR HEAT GENERATION RATE (LHGR) for the highest powered rod which is equal to or less than the design LHGR corrected for densification.

The APLHGR limits specified are equivalent to the LHGR of the highest powered fuel rod assumed in the LOCA analysis divided by its local peaking factor. A conservative multiplier is applied to tho i

l LHGR assumed in the LOCA analysis to account for the uncertainty associated with the measutomont of the APLHGR.

The calculate tal procedure used to establish the maximum APLHGR values uses NRC approved i

l calculational models which are consistent with the requirements of Appendix K of 10 CFR Part 50.

l The approved calculational models are listed in Specification 6.6.A.4.

The daily requiromcnt for calculating APLHGR when THERMAL POWER is greator than or equal to 25% of RATED THERMAL POWER is sufficient since power distribution shi'ts are very slow when l.

thero have not boon significant power or control rod changes. T*, requircmont to calculato APLHGR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the completion of a THERMAL FJWR increase of at least 15% of RATED THERMAL POWER onsures thermallimits are mot after power Jistribution shifts while still allotting timo for the power distribution to stabilize. The requirement for calculating APLHGR after initially determining a LIMITING CONTROL ROD PATTERN exists ensures that APLHGR will be known following a change in THERMAL POWER or power shape, that cvuld place operation above a thermallimit.

i i

l 3/4.11.8 APRM SETPOINTS l

The fuci cladding integrity Safety Limits of Specification 2.1 were based on a' power distribution which would yield the design LHGR at RATED THERMAL POWER. The flow biased neutron flux -

high scram sotting and control rod block functions of the APRM instruments for both two recirculation loop operation and singlo recirculation loop operation must be adjusted to ensure that I

the MCPR does not becomo less than the fuel cladding safety limit or that 2: 1% plastic strain does not occur in the degraded situation. The scra" m.ttings and rod block settings are adjusted in accordance with the formula in this specificath... when the value of MFLPD indicatos a higher peaked power distribution to ensure that an LHGR transient would not be increased in the degraded l

j' condition.

-QUAD CITIES - UNITS 1 & 2 B 3/4.11 1 Amendment No.

. - ~

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

e

,..,-~

l POWER DISTRIBUTION LIMITS B 3/4.11 BASES l

3/4.11.C MINIMUM CRITICAL POWER RATLQ The required operating limit MCPR at steady stato operating conditions as specified in Specification 3.11.C are derived from the established fuel cladding integrity Safety Limit MCPR, and an analysis of abnormal operational tremsients. For any abnormal operating transient analysis ovaluation 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 Safoty Limit MCPR at any timo during the transient assuming instrument trip setting given iri Specification 2.2.

l To assure that the fuel cladding integrity Safety Limit is not excooded during any anticipated abnormal operational transient, the most limiting transients havo boon analyzed to determine which result in the largest reduction in the CRITICAL POWER RATIO (CPR). The type of transients ovaluated woro change of flow, increase in pressure and power, positive reactivity insertion, and

~

coolant temperature decrease. The limiting transient yields the largest delta MCPR. When added l

to the Safety Limit MCPR, the required minimum operating limit MCPR of Specification 3.11.C is obtained and presented in the CORE OPERATING LIMITS REPORT.

Tho steady stato values for MCPR specified woro determined using NRC approved methodology listed in specification 6.6.A.4.

The purpose of the MCPR multiplicative factor specified in the CORE OPERATING LIMITS REPORT is to defino MCPR operating limits at other than rated core flow conditions. At loss than 100% of l

rated flow, the requireo MCI'R is the product of the MCPR and the off rated flow MCPR multiplier factor. The MCPR multiplier assures that the Safety Limit MCPR will not be violated.

Since the transient analysis takes credit for conservmism in the scram spood performance, it must be demonstrat9d that the specific scram spood distribution is consistent with that used in the transient analysis. Tho 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> completion timo is acceptable due to the relatively minor changes in t,,, expected during the fuel cycle.

At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER, the reactor-will be operating at minimum recirculation purnp speed and the moderator void content will be very small. For all designated control rod pattems which may be employed at this point, operating plant experience indicates that the resulting MCPR value has considerable marpin. Thus, the demonstration of MCPR below this power levelis unnocessary. The daily requirement for -

calculating MCPR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient r. ice power distribution shifts are very slow when there have not been significant power or control rod changes. The requiremont for calculating MCPR af ter initially dotormining that a LIMITING CONTROL ROD PATTERN oxists ensures that MCPR will be known following a change in THERMAL POWER or power shape, regardless of magnitude, that could place operation above a thermallimit.

L QUAD CITIES - UNITG i & 2 B 3/4.112 Amendmont No.

.i..a'u.1

.,m a,,w-asmue.--.-

wawaerne e emm---e wsa.a6g--n-,.ea.-wm--_-e sum neNap,,.

,.e-gp~w W==

,-wo L

14Wer*="d'IteNt'-%'+'""t+"L' tl' hT"'T fr-e w

re

=='v'rw" rw v

t.-r-me--oh a--

t

i l

i POWER DISTRIDUTION LIMITS B 3/4.11 BASES 3L4d1&

LLtdAR HLA1.SE!KfMILQ1LBAIE This specification assures that the LINEAR HEAT GENERATION RATE (LHGR)in any fuel rod is loss than the design linear heat generation even if fuel pollet densification is postulated. The daily requirement for calculating LHGR when THERMAL POWER is greator than or equal to 25% of RATED THERMAL POWER is cuf hcient sinco power distributions shif ts are very slow when there have not been si nificant powr' u 1 trol md changes. The requirement to calculato LHGR within 0

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> af ter the completion 's"

'X K">ER increase of at least 15% of RATED

( m

'er power distribution shif ts whilo still allotting THERMAL POWER onsures the tm wa e !.

timo for the power distribution to N ij - Too,2ositomont for calcubtin0 LHGR af ter initially determinin0 a LIMITING CONTROL ncs AtTERN exists ensures that LHGR will be known f ollomng a change in THERMA 1. POWEP or power shapo that could place operation above a thormal limit.

QUAD CITIES - UNITS 1 & 2 B 3/4.11-3 Amendment No.

N1'I'ACllMICNT 4 EXISTING TECHNICAL SPECIFICATIONS Technical Specification 3/4.11 "POWliR DISTRIBUTION IJMILS"

(

m._.... _ _ _

_ ~. _ _

_. ~._..

F s

ATTACHMENT 4 7

DELETION OF CURRENT TECHNICAL SPECIFICATIONS I

This tochtucal specification amendment will creato a now section 3/4,11 that will replace j

soveral specifications in the current section 3.5/4.5, ECCS Systems, for the Dresden Unit l

2 and Unit 3 Technical Specifications. Sections 3.5/4.5 will be replaced in its entirety with revised pages that combino the Unit 2 and Unit 3 specifications when the upgraded section 3/4,5 is developed and therefore, no pages ato being doloted with this amendrnent.

a 4

s I

i

.-....,._,_,.__....m.-..-m-.-.,-.___,_.--_

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

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

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

ATTACHMENT 4 DELETION OF CURRENT TECHNICAL SPECIFICATIONS This technical specification amendment will creato a now section 3/4.11 that will replaco soveral specifications in thu curiont section 3.0/4.5, ECCS Systems, for the Quad Cities Unit 1 and Unit 2 Technical Specifications. Sections 3.5/4.5 will be replaced in its entirety with revised pa00s that combino the Unit 1 and Unit 2 specifications when the upgraded

]

section 3/4.5 is developed and therefore, no pagos are being doloted with this amendment.

i e

i 5

l 1

t l

A

+

t 9

e=,-m..--en

-e-w,,,-e.

i.-,-----,--r,-,---.w.---...,.,;-y

..w-,-m,,, -.,,,

.--4 m..w,,..-~w,..-.-,.-,,--%,,..v.v%-wr,s,-m,-,---.%-wy.a-

--.w.,

wmt

NITACllMENT 5 DRESDFsN 2/3 DIFFERENCES Technical Specification 3/4,i1 "POWIIILDISIRIB_UTION LIMITS"

ATTACHMENT 5 t

COMPARISON OF DRESDEN UNIT 2 AND UNIT 3 TECHNICAL SPECIFICATIONS FOR THE IDENTIFICATION OF TECHNICAL DIFFERENCES SECTION 3/4.5 ECCS SYSTEMS" I

Commonwealth Edison has conducted a comparison review of the Dresden Unit 2 and Unit 3 Technical Specifications to identify any technical differences in support of combinin0 the Technical Specifications into one document. The intent of the revioW was not to identify any dif ferences in presentation style (e.g. table formats, use of capitalletters, etc.) or punctuation but rather to identify areas in which the Technical Specifications are technically different.

The review of Section 3.5/4.5 "ECCS System" Sections 1,-J, K, and L did not reveal any technicLt d;fferences.

t 3

i b

f'

' ]

m' T

r

-~=ew^^1

"'e-'-*

w -' t'r-c'r""

F N='

• • F'-

t

-"***uss' m'v--

'--**m=*

"-'=--

E-'+-'r-'

- - - - - - - *e'-e d

~

l A'I'l'ACIIMEN'I' 5 QUAD CITIES 1/2 DIFFERENCES Technical Specification 3/4.11

" POWER DISTRIBJUIlON IJMITS"

I ATTACHMENT 5 COMPARISON OF QUAD CITIES UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS FOR THE

- i IDENTIFICATION OF TECHNICAL DIFFERENCES SECTION 3.5/4.5 "ECCS SYSTEM" Commonwealth Edison has conducted a comparison review of the Quad Cities Unit 1 and Unit 2 Technical Specifications to identify any technical differences in support of combining the Technical Specifications into one document. The intent of the review was not to identify any differences in presentation style (e.g. table fortnats, use of capital letters, etc.), punctuat!on or spellin0 enors, but rather to identify areas which the Technical Specifications are technically or administratively dif mrent.

The review of Section 3.5/4.5 "ECCS System" Sections I, J, and K revealed the following technical dif ferences:

1.

Unit 1 Page 3.5/4.520: The Aveta00 Planar Linear Heat Generation Rate (APLHGR) also serves a secondary function which is to assure fuel tod mechanical integrity.

Unit 2 Page 3.5/4.5-13 and 14: Power operation with LHGRs at or below those specified in the CORE OPERATING LIMITS REPORT assures that the peak cladding temperature following a postulated loss-of-coolant accident w'> not exceed the 2200 *F limit. Those values represent limits for operation to ensure conformance with 10 CFR 50 Appendix K only if they are more limitin0 than other design parameter. The maximnm average planar LHGRs specified in the CORE OPERATING LIMITS REPORT at higher-exposures result in a peak cladding temperature of less than 2200 F.

However, the maximum average planar LHGRs are specified in the CORE OPERATING LIMITS REPORT as limits because conformance calculations have not been performed to justify operation at LHGRs in excess of those shown.

The Unit 2 Bases materialis adopted in theory by adopting the Bases material presented in the Standard Technical Specifications.

2.

Unit 1 Page 3.5/4.5-21: The MCPR Operating Limit re!!ects an increase of:

0.03 over the most limiting transient to allow continued operation with one feedwater heater out of service.

Unit 2: Unit 2 does not have this paragraph.

i

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

.co,--

ATTACHMENT 5 The material presented in the Unit 1 Bases represents a cycle specific added that is currently contained in the core operating limits report. Therefore, the materialis not adopted in accordance with the STS.

3.

Unit 2 Pago 3.5/4.5-14a: This page contains the information related to the ODYN option B determination of tho scram time input for the MCPR limit.

The information is not adopted in the proposed specification in accordance with the Standard Technical Specifications. The information is contained partially in the core operating limits report and within the technical manuals.

A b

1

xcracnuiwr 6 SIGNIFICANT HAZARDS t

^

CONSIDERATIONS AND ENVIRONMENTAL ASSESSMENT EVALUATION Technical Specification 3/4.11 "ROWE3_ DISTRIBUTION LIMD3"

________.m.___,___

ATTACHMENT G EVAL.UATION OF SIGNIFICANT HAZARDS CONSIDERATION Commonwealth EJison has evaluated this proposed amendmont and determined that it involves no significant hazards consideration. Accordin0 to 10 CFR 50.92(c), a proposed amendment to an operating licenso involves no si0nificant hazards consideration if operation of the f acility, in accordance with the proposed amendment, would not:

1)

Involve a significant increase in the probability or consequences of an accident previously evaluated; or 4

2)

Croato the possibility of a new or different kind of accident from any accident previously evaluated; or i

3)

Involvo a significant reduction in a margin of safety, l

The proposed changes do not involvo a si nificant increase in the probability or 0

consequences of an accident previously evaluated because:

In General, the proposed chan00s represent the conversion of current requiroments to a more generic format, or the addition of requirements which are based on the t

current safety analysis. Implementation of these chan0es will provide increased i

reliability of equipment assumed to operato in the current safety analysis, or provido continued assurance that specified parameters tomain within their acceptance 4

limits, and as such, will not si nificantly increase the probability or consequences of 0

a previously evaluated accident.

Some of the proposed changes represent minar curtailments of the current requirements which are based on generic guidance or previously approved provisions for other stations. These proposed changes are consistent with the current safety analyses and have benn previously determined to represent sufficient requirements for the assurance of reliability of equiprnent assumed to operate in the safety analysis, or provido continued assurance that specified paramotors romain within ",eir acceptance limits. As such, these changes will not si0nificantly l

increase the probability or consequencer, of a previously evaluated accident.

(

The Generic Changes to the technical specifications involve administrative chan00s to format and arrangement of the m_aterial. As such, these changes cannot involve -

a si nificant increase in the probabildy or consequences of an accident previously I

0 evaluated.

i i

The current specifications require the reactor to be placed in cold shutdown when a thermallimit was exceeded and not rostored within the allotted 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, but the-proposed specifications require the reactor to be less than 25% of rated thermal power if this condition occurrod. The chanao eliminates a shutdown and requires the power level to be reduced to the point that the limits are no longer applicable.

l l

w N

gm-

+-m5

>mes t g-ar -

w-ma iet ow v,aa.mm=--a-es yr--o-va-evg-a ww+,W--

n e p e w -- mm:wmm.ae.wwe-wm-e4--r--.--e sw wsw--,g.m,

-a,-e m-wauma----

mw+&

F-rwma-k e <w-eime--et--

e--

h-me -armt gr u w - M-Tr g* pr Frir'1*

~ _ _ _ _

)

i i

ATTACHMENT C 1

Thereforo, the chango will not increase the probability or consequences of an l

accident.

i l

Croato the possibility of a now or different kind of accidt at from any previously ovaluated l

because:

l i

in general, the proposed chan00s represent the conversion of current requirements j

to a more generic format, or the addition of requirements which are based on tho l

current safety analysis. Others reprosent minor curtailmonts of the curront i

requirements which are based on generic guidance or previously approved i

provisions ior other stations. Thoso chan00s do not involvo revisons to the desi n 0

of the station. Some of the chan00s may involvo revision in the operation of the l

station; however, those chan00s provido naditional rostrictions which are in i

accordance with the current safety analysos, or are to provido for additionhl testin0 or surveillance which will not introduce now failure mechanisms beyond thoso already considered in the current safety analysom Thoroforo, thoto changes will not croato the possibility of a now or different kind of accident from any accident previously evaluated.

Since the Generic Chan0es proposed to the technical specifications are administrative in naturo, they cannot create the possibility of a now or different kind of accident from any previously evaluated.

The toquirement to reduce thermal power to less than 25% of rated thortnal power l

rather than place tha reactor in cold shutdown will not creato a now or diilotent kind of accident because the thermallimits are not required in operational modo 1 t

when thermal power is loss than 25% of rated thermal power, involvo a si nificant reduction in the mar 0 n of sufoty becauso:

0 i

In general, the proposed changes represent the conversion of current requiromonts to a moro Generic format, or the addition of requirements which are based on the curront safety analysis. Others represent minor curtailments of the current requirements which are based on generic Guidance or previously approved provisions for other stations. Some of the latter individual items may introduco minor reductions in the mar 0 n of safety when compared to the curront i

requirements. However, other individual chan00s are the adoption of now requirements which will provido significant enhancement of the reliability of tho equipment assumed to operato in the safety analysis, or provido enhanced assuranco that specified paramotors remain within their acceptance limits Thoso enhancements compensate for the individual minor reductions, such that taken together, the proposed changes will not si nificantly reduce the mar 0in of safety.

0 The Genoric Chan00s proposed in this amendment request are administrativo in

[

nature and, as such, do not involve a reduction in the rnar0 n of safety, i

l

~2 c.

...u u..

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

.m.

-,a

..._.. _ _ _._ __ _.. _ _ _ _ _ _ _. _. _. _ _ _. _ _ _. _ _ _ _ _ _ _. _ _.... _. m i

ATT ACHMENT 6 ENVIRONMENTAL ASSESSMENT STATEMENT APPLICABILITY REVIEW Cominonwealth Edison has ovaluated the proposed amendment against the critoria for the identification of licensin0 and regulatory actions requiring onvironmental assessment in i

accordance with 10 CFR 51.20. It has boon deterrnined that the proposed changes meet the critoria for a categorical exclusion as provided under 10 CFR 51.22 (c)(9). This conclusion has boon determined because the changes requested do not poso significant hazards consideration or do not involve a significant increase in the amounts,'and no significant changes in the typos, of any of fluent that inay be released of(sito. Additionally, this toquest does not involvo a significant increase in individual or cumulative occupational radiation exposuro. Thorofore, the Environmental Assessment Staternent is not applicable for those changes.

-l 1

i I

t g,

r~,,,v,,--e.w,vwr,.,-nen w-w+,,mm,4---e-,

a~-

..-ne,www,-e e-, - -, _

w..-mn.

,e,e,-w--

wa s

.e

--w w r-,,n.,n,,,-n..,

,,.wv.,na-

<,wwe

-,+v.

--~r.e-nse

A'ITACilMENT 7 GENERIC LETTER 87-09 IMPLEMENTATION Technical Specification 3/4.11 "PDIVEILDISJJllBUTION LIMITS" l

l

J ATTACHMENT 7 APPLICATION OF GENERIC LETTER 87 00 I

REVISION TO SPECIFICATION 3.0.D The Dresdon/Ouad Cities Technical Specification Up0rado Program has implernented the recommendations of Genoric Letter 87 09. Included in those recommendations was a revision to Standard Technical 1:ipecification 3.0.4 for which theso stations had no j

corrosponding rostriction. Under the pioposed Specification, entry into an operational modo or other specified condition is permitted under compliance with the Action requirements. Indicated below is the method of implomontation for this recommendation for each Action requirement in this packa00.

t i

PROPOSED APPL.

CONT OPS IN TECH SPEC ACTION MODES APP.COND?

' CAT-CLARIFICATION i

3.11. A 1, > 2 5 %

2 hrs NO Must reduce to < 25%

3.11.B 1

1, > 25%

0 hrs NO Must reduce to < 25%

2 1, > 25%

6 hrs NO Must reduce to < 25%

3 1, > 25%

G hrs NO Must reduco to < 25%

3.11.C 1

1, > 2 5 %

2 hrs NO Must reduce to < 25%

I i

3.11.D 1, > 2 5 %

2 hrs NO Must reduce to < 25%

3.11.E 1, > 25%

2 hrs NO lust reduce to < 25%

(Crosdon) i 4

l l

P 1,1,

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

--a-

NITACllMENT I EXECUTIVE

SUMMARY

Technical Specification 3/4.12 "SEECIAL TEST EXGPlJDES" l

l

EXECUTIVE

SUMMARY

The Dresden Technical Specification Upgrade Program (TSUP) was conceptualized in response to lessons learned from the Diagnostic Evaluation Team inspection and the frequent need for Technical Specihcation interpretations. A cornparison study of the Standard Technical Specificatien (STS), later operating plant's Technical Specifications provisions and Quad Cities Technical Specifications was performed prior to the Dresden TSUP ef f ort. The study identified potential improvernents in clarif ymg requirements and requirements which are no longt r consistent with current industry practices. The Dresden TSUP will enhance the Ouad Cihes TSUP currently under review by the NRC. As a result of the inconsistencies in tbo Ot,ad Cities submittal compared to the Standard Technical Specifications (STS), Dresden's submittal will more closely follow the provisions of STS and in conjunction, Quad Cities will amend their submittal so that Ouad Cities and Dresden are identical witt,.a equipment and plant desi n. The format for the Dresden TSUP will 0

remain as a two column layout for human factors considerations. Additionally, chapter organizations will remain essentially unchanged.

The TSUP is not intended to be a complete adopt:on for the STS. Overall, the Dresden custom Technical Specifications provide for the safe operation of the plant and therefore, only an upgrade is deemed necessary.

In response to an NRC recommendation, Quad Cities combined the Unit 1 and Unit 2 Technical Specifications into one document. The Dresden Unit 2 and Unit 3 Technical E

Specifications will also be combined into one document. To accomplish the combination of the Units' Technical Specification, a comparison of the Unit 2 and Unit 3 Technical Specification was performed to identify any technical dif ferences. The technical dif ferences are identified in the proposed amendment packa00 for each section.

The TSUP was identified as a station top priority and is currently contained in the Dresden Management Action Plan (DMAP). The TSUP goal is to provide a better tool to station personnel to implement their responsibilities and to ensure Dresden Station is operated in accordance with current industiy practices. The improved Technical Specifications provide f or enhanced operation of the plant. The program improves the operator's ability to use the Technical Specibcations by more clearly defining the Limitin0 Conditions for Operation and required actions. The most signihcant improvement to the speci" cations is the addition of equipment operability requirements during shutdown conditions.

l l

f EXECUTIVE

SUMMARY

(continued)

PROPOSED CHANGES TO TECHNICAL SPECIFICATION SECTION 3/4.12, "SPECIAL TEST EXCEPTIONS" The current Dresden and Quad Cities Technical Specifications contain Applicability and Objective statomonts at the beginning of most sections. The proposed amendment will doloto the "Objectivo" statement and integratos appropriato applicability statements within the specifications. This providos a clarification of the intended requirements and actions which are required when the specification cannot be met.

Proposed Section 3/4.12.A allows the primary containment integrity specifications to be suspended for low power physics testing.

Proposed Section 3/4.12.8 delineates the req w 9ments needed during shutdown margin demonstrations.

k m a w

-- e e.

r

~

e 1

c.

9

ATI'ACIIMENT 2 l

DESCRIPTION OF CHANGES Technical Specification 3/4.12 "SEEClAL_T_EST EXCEPTIONS" i

ATTACHMENT 2 DESCRIPTION OF PROPOSED AMENDMENT The changes proposed in this amendment request are made to improve the understanding and usability of the present technical specifications, and to incorporato the technical improvements from the Standard Technical Specifications (STS).

The present Dresden and Quad Cities Technical Specifications contain Applicability and Objectivo statements at the bo0 nning of most sections. Those statements are generic in i

nature an" do not provido any usefulinformation to the user of the technical specificatlans. The proposed change will dolote *.ho Objectivo statement and provido Applicability statements within each specification based on the STS. The proposed Applicability statomont to be included in each specification willincludo the applicable operational modos or other conditions for which the Limiting Condition for Operation (LCO) must be satisfied.

The STS action provisions which delineate a specification 3.0.4 exception are not incorporated into the proposed specifications. The incorporation of the Generic Lotter 87-09 chan00 to STS specification 3.0.4 (Dresden and Quad Cities proposed 3.0.D specification) requires that each action be independently evaluated for applicability of the new specification. These evaluations are provided in Attachment 7.

Proposed section 3/4.12 is a new section but contains provisions that are currently allowed by the Technical Specifications.

SPECIFIC CHANGES Section 3/4.12, Primary Containment Inte0rity The proposed specification implements the currer" provisions contained in specification 3.7.A.2 of the Dresden and Quad Cities Technical Specifications in accordance with STS guidelines. Tho specification allnws the primary containment integrity requirements to be suspended for the purpose of performing low power physics tests with thermal power less than 1 % of rated therrnal power and the reactor coolant temperature is less than 212*F.

The temperature requirement is 212*F versus the STS requirement of 200 F for consistency with the proposed operational modes defined in Section 1.0 and the current specifications. The current technical specifications require the reactor power to be less than 5 MWt (~0.2%). The proposed specification uses '% for consistency with other plants and STS. The proposed applicability is operattonal mode 2 durin0 ow oower l

physics test. The proposed action requires the reactor modo switch to be placed in the shutdown position if thermal power is nised ebove the 1% l 901 or the reactor coolant temperature becomes Greater than 212"F. The proposed Surveillanco Requirement (SR)is to verify thermal power and reactor coolsrit tan peratura are within the limits at least once per hour during low power physics tests.

ATTACHMENT 2 Section 3/4.12.B. Shutdown Margin Demonstrations The proposed specification implements the current provisions contained in specification 3.3.B.3.b of the Dresden and Quad Cities Technical specifications. The proposed specification allows the mode switch interlocks and control rod position provisions contained in specifications 3.10.A and 3.10.C to be suspended for the purpose of performing shutdown marD n demonstrations provided tho sourco rango mon 4 ors are i

operablo, the control rod sequence is verified by either the rod worth minimizer or a second qualified individual, the rod out-notch overrido function is not used, and no other coro alterations are in progress. The proposed applicability is operational mode 5 during shutdown mar 0 n demonstrations. The proposed action requires the reactor mode switch i

to be placed in the shutdown or refuel position if any of the requirements are not mot.

The proposed SRs require that within 30 minutes prior to and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during the performance of the shutdown margin demonstration that the sourco ran00 monitors are operable, the control rod sequence is being enforced, and that no other coro alterations are in progress.

The remainder of the STS section 12 specifications are not adopted because the specifications are only applicablo during the Startup Test Program.

The Bases for Section 3/4.12 are implemented in accordance with the proposed specifications.

ATI'ACllMENT 3 PROPOSED TECHNICAL SPECIFICATIONS Technical Specification 3/4.12 "SEECIAL TEST EXCEETIONS"

SPECIAL TEST EXCEPTIONS PCI 3/4,12.A 3.12 LIMITING CONDITIONS FOR OPERATION 4.12 SURVEILLANCE REQUIREMENTS A.

PRIMAPY CONTAINMENT INTEGRITY A.

PRIMARY CONTAINMENT INTEGRITY The provisions of Specifications 3.7 A, The THERMAL POWER and reactor coolant 3.7 E and 3.10.A and Tablo 12 may be temperature shall be verified to be within suspended to permit the reactor pressure the limits at least once por hour during low vessel closuro head and the drywell head to power PHW!CS TESTS.

be tomoved and the primary containment air lock docrs to be open when the reactor mode switch is in the Sttrtup position during low power PHYSICS TESTS with THERMAL POWER less than 1% of RATED THERMAL POWER and roactor coolant temperature less than 212'F.

APPLICABILITY:

7 OPERATIONAL MODE 2, during low power PHYSICS TESTS, ACTION:

With THERMAL POWER greater than or equal to 1% of RATED THERMAL POWER or with the reactor coolant temperature greater than or equal to 212'F,immediately place the reactor mode switch in the Shutdown position.

DRESDEN - UNITS 2 & 3 3/4.12 1 Amendment No.

SPECI AL TEST EXCEPTIONS SOM 3/4.12.B 3.12 - LIMITING CONDITIONS FOR OPERATION 4.12 - SURVElLLANCE REQUIREMENTS B.

SHUTDOWN MARGIN Demone +tions B.

SHUTDOWN MARGIN Demonstrations The provisions of Specifications 3.10.A and Within 30 minutes prior to and at least once 3.10.C and Table 1-2 may ba suspended to per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during the po'formance of a permit the reactor mode switch to be in the SHUTDOWN MARGIN demonstration, verify Startup position and to allow more than one that; control rod to be withdrawn for SHUTDOWN MARGIN demonstration, 1.

The source range monitors are provided that at least th: following OPERABLE with the RPS circuitry requirements are satisfied.

" shorting links" removed per Specification 3.10.B, i

1.

The source range monitors are OPERABLE with the R?S circuitry 2.

The rod wonn minimizer in OPERABLE

" shorting links" removed per with the required program per Specification 3.10.D.

Specification 3.3.L or a second licensed operator or other technically qualified 2.

The red worth minimizer is OPERABLE individual is present and verifies per Specification 3.3.L and is compliance *.vith the SHUTDOWN programmed for the SHUTDOWN MARGIN demonstration procedures, MARGIN demonstration, or and conformance with the SHUTDOWN MARGIN demonstration procedure is 3.

No other CORE ALTERATION (s) are in verified by a second licensed operator progress.

or other technically qualified individual.

3.

The " rod-out-notch-override" control shall not be used during out-of-sequence movement of the control

~

rods.

4.

No other CORE ALTERATION (s) are in progress.

AfPLICABILITY:

OPE 7.ATIONAL MODE 5, during SHUTDOWN MARGIN demonstrations.

ACTION:

With the requirements of the above specification not satisfied,immediately placo the reactor mode switch in the Shutdown or Refuel position.

DRESDEN - UNITS 2 & 3 3/4.12-2 Amendment No.

l

SPECIAL TEST EXCEPTIONS - B 3/4.12 BASES s

,n.

3/4.12. A PRIMARY CONTAINMENT INTEGRITY The requirement for PRIMARY CONTAINMENT INTEGRITY is not applicable during the period when open vessel tests are being performed during the low power PHYSICS TESTS. Low power PHYSICS TESTS during OPERATIONAL MODE 2 may be required to be performed while still maintaining access to the primary containment and reactor pressure vessel. Additional requirements during these tests to restrict reactor power and reactor coolant temperature provide protection against potential conditions which could require primary containment or reactor coolant pressure boundary integrit/.

3/4.12.B SHUTDOWN MARGIN Demonstrations Performence of SHUTDOWN MARGIN demonstrations with the vessel head removed requires additional restrictions in order to ensure that criticality does not occur. These additional restrictions are specified in this LCO. SHUTDOWN MARGIN tests may be performed while in OPERATIONAL MODE 2 in accordance with Table 1-2 without meeting this Special Test Exception.

For SHUTDOWN IO9 GIN demonstrations performed while in OPERATIONAL MODE 5, additional requirements must ue met to ensure that adequate protection against potential reactivity excursions is available, Because multiple control rods will be withdrawn and the reactor will potentially become critical, the approved control rod withdrawal sequence must be enforced by the RWM, or must be verified by a second licensed operator or other technically qualified individual.

To provide additional protection against inadvertent criticality, control rod withdrawals that are "out-rsf-sequence", i.e., do not conform to the P nked Position Withdrawal Sequence, must be made in individual notched withdrawal mode to minimize the potential reactivity insertion associated with each movement. Because the reactor vessel head may be removed during these tests, no other CORE ALTERATION (s) may be in progress. This Special Test Exception then allows changing the Table 1-2 reactor mode switch position requirements to include the Startup or Hot Standby position such that the SHUTDOWN MARGIN demonstrations may be performed while in OPERATIONAL MODE 5.

i u?.ESDEN - UNITS 2 & 3 B 3/4.12-1 Amendment No.

l-

SPECIAL TEST EXCEPTIONS PCI 3/4.12.A 3.12 - LIMITING CONDITIONS FOR OPERATION 4.12 - SURVEILLANCE REQUIREMENTS A.

PRIMARY CONTAINMENT INTEGRITY A.

PRIMARY CONTAINMENT INTEGRITY The provisions of Specifications 3.7.A, The THERMAL POWER and reactor coolant 3.7.E and 3.10.A and Table 1-2 may be temperature shall be verified to be within suspended to permit the reactor pressure the limits at least once per hour during low vessel closure heal,.

the drywell head to power PHYSICS TESTS.

be removed and the pri. nary containment air lock doors to be open when the reactor modo switch is in the Startup position during low power PHYSICS TESTS with THERMAL POWER less than 1% of RATED THERMAL POWER and reactor coolant temperature less than 212 F.

APPLLCABILITY:

OPERATIONAL MODE 2, during low power PHYSICS TESTS.

ACTION:

With THERMAL POWER greater than or equal to 1% of RATED THERMAL POWER or with the reactor coolant temperature greater than or equal to 212 F, immediately place the reactor mode switch in the Shutdown position.

Amendment No.

QUAD CITIES UNITS 1 & 2 3/4.12-1 I

-. - = -.

SPECIAL TEST EXCEPTIC 4 SDM 3/4.12.B 3.12 - LIMITING CONDITIONS FOR OPERATION 4.12 - SURVEILLANCE REQUIREMENTS B.

SHUTDOWN MARGIN Demonstrations B.

SHUTDOWN MARGIN Demonstrations The provisions of Specifications 3.10.A and Within 30 minutes prior to and at least once -

3.10.C and Table 1-2 may be suspended to per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during the performance of a permit the reactor mode switch to be in the SHUTDOWN MARGIN demonstration, verify _

- Startup position and to allow more than one that; control rod to be withdrawn for SHUTDOWN MARGIN demonstration, 1.

The source range monitors are provided that at least the following OPERABLE with the RPS circuitry requirements are satisfied.

" shorting links" removed per Specification 3.10.B, 1,

The source range monitors are OPERABLE with the RPS circuitry 2.

The rod worth minimirer is OPERABLE

" shorting links" removed per with the required program per Specification 3.10.B.

Specification 3.3.i. or a second licensed operator or other technicaliy qualified 2.

The rod worth minimizer is OPERABLE individual is present and verifies -

per Specification 3.3.L and is compliance with the SHUTDOWN programmed for the SHUTDOWN MARGIN demonstration procedures, MARGIN demonstration, or and-conformance with the SHUTDOWN MARGIN demonstration procedure is 3.

No other CORE ALTERATION (s) are in verified by a second licensed operator

progress, or other technically qualified individual.

3.

The " rod-out-notch-override" control shall not be used during out-of-sequence movement of the control rods.

4.

No other CORE ALTERATION (s) are in progress.

APPLICABILITY:

l' OPERATIONAL MODE 5, during SHUTDCWN MARGIN demonstrations.

ACTION:

l With the require?, ants of the above specification not xtisfied,immediately place _the reactor mode switch in the Shutdown or Refuel g,sition.

QUAD CITIES - UNITS 1 & 2 3/4.12-2 Amendment No.

SPECIAL TEST EXCEPTIONS. B 3/4.12 BASES 3/4.12. A PRIMARY CONTAINMENT INTEGRITY The requirement for PRIMARY CONTAINMENT INTEGRITY is not applicable during the period when open vessel tests' are being performed during the low power PHYSICS TESTS. Low power PHYSICS TESTS during OPERATIONAL MODE 2 may be required to be performed while still maintaining access to the primary containment and reactor pressure vessel. Additional requirements during these tests to restrict reactor power and reactor coolaat temperature provide protection against potential conditions which could require primary containment or reactor coolant pressure boundary integrity.

3/4.12.Q SHUTDOWN MARGIN Demonstrations Performance of SHUTDOWN MARGIN demonstrations with the vessel head removed requires additional restrictions m order to ensure that criticality does not occur. These additional restrictions are specified in this LCO. SHUTDOWN MARGIN tests may be performed while in OPERATIONAL MODE 2 in accordance with Table 1-2 without meeting this Special Test Exception.

For SHUTDOWN MARGIN demonstrations performed while in OPERATIONAL MODE 5, additional requirements must be met to ensure that adequate protection against potential reactivity excursions is available. Because multiple control rods will be withdrawn and the reactor will potentially become critical, the approved control rod withdrawal sequence must be enforced by the RWM, or must be verified by a second licensed operator or other techr.ically qualified individual.

To provide additional protection against inadvertent criticality, control rod withdrawals that are "out-of-sequence", i.e., do not conform to the Banked Position Withdrawal Sequence, must be made in individual notched withdrawal mode to minimize the potential reactivity insertion

-associated with each movement. Because the reactor vessel head may be removed during these tests, no other CORE ALTERATION (s) may be in progress. This Special Test Exception then allows changing the Table 1-2 reactor mode switch position requirements to include the Startup or Hot Standby position such that the SHUTDOWN MARGIN demonstrations may be performed while in OPERATIONAL MODE 5.

QUAD CITIES - UNITS 1 & 2 B 3/4.12-1 Amendment No.

4

A'ITACIIMENT 4 EXISTING TECHNICAL SPECIFICATIONS Technical Specification 3/4.12 "SPECIAL TEST EXCEPTIONS" E

~ _. _ _ _... _ _ _ _ _ _ _ _ _ _.. _ _ _ _ _ _ _ _ _ _

ATTACHMENT 4 DELETION OF CURRENT TECHNICAL SPECIFICATIONS This technical specification amendrnent is a new section for the Dresden Unit 2 and Unit 3 Technical Specifications. Therefore, no technical specifications are being deleted with this amendment.

N

. -. ~.

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

.. - - ~... -

i ATTACHMENT 4

-4 DELETION OF CURRENT TECHNICAL SPECIFICATIONS'.

' This technical specification amendment is a new section for the Ouad Cities Unit 1 and

. Unit 2 Technical Specifications. Therefore, no technical specifications are being deleted with this amendment.

)

s

. l b

4

.i d'

b 9

d'

~

ATI'ACIIMENT 5 DRESDEN 2/3 DIFFERENCES Technical Specification 3/4.12 "SPECIAL TEST EXCEPTIONS" e

i 1

I' f.

_ ~ _.

1 ATTACHMENT 5 COMPARISON OF DRESDEN UNIT 2 AND UNIT 3 TECHNICAL SPECIFICATIONS FOR THE IDENTIFICATION OF TECHNICAL DIFFERENCES SECTION 3/4.12 "SPECIAL TEST EXCEPTIONS" Commonwealth Edison has conducted a comparison review of the Dresden Unit 2 and Unit 3 Technical Specifications to identify any technical differences in support of combining the Technical Specifications into one document, The intent of the review was not to identify any differences in presentation style (e.g table formats, use of capitalletters, etc.),

punctuation or spelling errors, but rather to identify areas which the Technical Specifications are technically or administratively dif forent.

Proposed section 3/4.12, Special Test Exceptions is a new section and thorofore, does not contain any technical dif ferences.

A'ITACllMENT 5

-QUAD CITIES 1/2 DIFFERENCES Technical Specification 3/4.12 "SPECIA_L TEST EXf_EPTIONS" 4

ATTACHMENT 5 i

COMPARISON OF QUAD CITIES UNIT 1 AND UNIT 2 TECHNICAL SPECIFICATIONS FOR THE IDENTIFICATION OF TECHNICAL DIFFERENCES SECTION 3/4.12 "5PECIAL TEST EXCEPTIONS" Commonwealth Edison has conducted a comparison review of the Quad Cities Unit 1 and Unit 2 Technical Specifications to identify any technical differences in support of combining the Technical Specifications into one document. The intent of the review was not to identify any differences in presentation style (e.0. table formats, use of capital letters, etc.), punctuation or spelling errors, but rathe to identify areas which the Technical Specifications are technically or administratively dif ferent.

Proposed section 3/4.12, Special Test Exceptions is a now section and therefore, does not contain any technical dif ferences.

ra.-

2

---

__._.__------,________________r________

NITACIIMENT 6 SIGNIFICANT HAZARDS CONSIDERATIONS AND ENVIRONMENTAL ASSESSMENT EVALUATION Technical Specification 3/4.12 "SPECIAL TEST E. CEPTION_S" X

l ATTACHMENT 6 EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION Commonwealth Edison has evaluated this proposed amendment and determined that it involves no significant hazards consideration. According to 10 CFR 50.92(c). a proposed amendment to an operating license involves no significant hazards consideration if operation of the f acility, in accordance with the proposed amendment, would not:

1)

Involve a signif; cant increase in the probability or consequences of an accident previously evaluated; or 2)

Create the possibility of a new or different kind of accident from any accident previously evaluated; or 3)

Involve a significant reduction in a margin of safety.

The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated because:

The proposed Specification 3/4.12 is a new section which will provide the user with a format that will allow better access to needed information and provide concise Applicability and Action requirements. The additions of Applicability and Action requirements represent clarification of intenced requirements that do not presently state all required conditions of operability or provide clearly stated Action statements if the requirements are not met. The combining of the two sections and the added requirements follow Standard Technical Specifications (STS) guidelines that are in use at many operating BWRs with similar design and operating configurations as Dresden and Quad Cities Stations.

The proposed Section 3/4.12 involves the relocation of present requirements into one section identical to STS provisions. The changes also implement the Applicability and Action provisions of the STS and later operating BWR plants that have been evaluated and found acceptable for use at Dresden and Quad Cities.

Present Surveillance Requirements are replaced, where applicable, with proven STS guidelines that are being used at plants with a system similar to that at Dresden and Quad Cities. The changes in the present Surveillance Requirements add testing requirements that are not presently in the Dresden and Quad Cities technical specifications. The prcposed changes do not affect accident assumptions other than a minor increase in the initial power level (~ 0.2% to 1 %) and as such, do not involve a significant increase in the probability of an accident previously evaluated.

The proposed specifications add additional requirements to specifications currently contained in the Technical Specifications. Since the proposed changes to the Technical Specifications implement requirements that have been demonstrated to provide acceptable operability provisions at other facilities with a design similar to that at Dresden and Quad Cities, the proposed changes do not significantly increase the consequences of an accident previously evaluated.

L

- ATTACHMENT G The proposed changes do not create the possibility of a new or different kind of accident

[

from any previously evaluated because:

The proposed administrative chan00s to the format and arrangement of material do not affect technical requirements or assumptions of any potential accident and; therefore, cannot create the possibility of a new or different kind of accident from any previously evaluated.

The proposed addition of Applicability and Action requirements enhanca the understanding and usability of the Technical Specifications and thus represent an improvement over pmsent specifications. New requirements are modeled af ter those in use at operating BWRs and do not represent requirements that will adversely affect potential accident analysos or assumptions. Therefore, the proposed chan0es do not create the possibility of a new or different kind of accident from any previously evaluated.

The proposed changes do not involve a significant reductivn in the margin of safety because:

The proposed administrative changes to format, arrangement of material, clarificatian of requirements and other non technical chan0es do not affect any safety aspects of the plant and as such can not involve a significant reduction in i

the mar 0 n of safety.

In addition, the commission has provided Guidance concerning the application of standards for determining whether significant hazards consideration exists by providing certain-examples (51 FR 7751) of amendments that are considered not likely to involve significant hazards considerations. Commonwealth Edison has reviewed the proposed changes against these examples and believes that the proposed changes fall within the scope of -

example (ii) "a change that constitutes an additional limitation, restriction, or control not presently included in the technical specifications" The proposed amendment does not involve a significant relaxation of the criteria used to establish safety limits, a si nificant relaxation of the bases for the limiting safety system 0

settin0s or a significant relaxation of the bases for the limiting conditions for operations.

Therefore, based on the guidance provided in the Federal Register and the criteria established in 10 CFR 50.92(c), the proposed change does not constitute a significant hazards consideration.

4

,-,,y

.w.~

---

,...-,.--.-..-e>

- -- - - - - - - - + -

. ~..

ATTACHMENT 6 ENVIRONMENTAL ASSESSMENT STATEMENT APPLICABILITY REVIEW Commonweat'\\ Edison has evaluated the proposed amendment against the criterie for the identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.20. It has been determined that the proposed changes meet the criteria for a categorical exclusion as provided under 10 CFR 51.22 (c)(9). This conclusion has been determined because the changes requested do not pose significant nazards consideration or do not involve a significant increase in the amot nts, and no significant changes in the types, of any effluent that may be released offsite. Additionally, this request does not involve a significant increase in individual or cumulative occupational radiation exposure. Therefore, the Environmental Assessment Statement is not applicable for these changes.

NITACllMENT 7 GENERIC LETTER 87-09 IMPLEMENTATION Technical Specification 3/4.12 "SPECIAL TEST EXCEPTIONS" l

l l

\\-

i I

ATTACHMENT 7 APPLICATION OF GENERIC LETTER 87 09 REVISION TO SPECIFICATION 3.0.D The Dresden/Ouad Cities Technical Specification Upgrade Program has implemented the recommendations of Generic Letter 87-09. Included in these recommendations was a revision to Standcrd Technical Specification 3.0.4 (Dresden and Quad Cities proposed 3.0.D specification) for which these stations had no correspondin0 restriction. Under the proposed Specification, entry into an operational mode or other specified condition is permitted under compliance with the Action requirements. Indicated below is the method of implementation for this recommendation for each Action requirement in this package.

PROPOSED APPL.

CONT. OPS IN TECH SPEC ACTION MODES APP. COND?

CAT CLARIFICATION 3.12. A

2. durin0 No No immediately place the low power reactor modo switch in the Physics Shutdown position Tests 3.12.B 5, during No No immediately place the Shutdown reactor mode switch in the Margin Shutdown or Refuel position

.__