ML20085H012

Amends 134,128,155 & 151 to Licenses DPR-19,DPR-25,DPR-29 & DPR-30,respectively,re Proposed Upgrade of Sections 2.0,3/4.11 & 3/4.12 of Dresden & Quad Cities TS

ML20085H012
Person / Time
Site:	Dresden, Quad Cities
Issue date:	06/13/1995
From:	Pulsifer R, Stang J NRC (Affiliation Not Assigned)
To:	COMMONWEALTH EDISON CO.
Shared Package
ML20085H017	List: ML20085H012 ML20085H030
References
NUDOCS 9506200445
Download: ML20085H012 (63)

Text

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UNITED STATES j

j NUCLEAR REGULATORY COMMISSION 1

t WASHINGTON, D.C. 20es6 0001

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COMMONWEALTH EDISON COMPANY AND IOWA-ILLINOIS GAS AND ELECTRIC COMPANY i

DOCKET NO. 50-254 OVAD CITIES NUCLEAR POWER STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 155 License No. DPR-29 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Commonwealth Edison Company (the licensee) dated September 15, 1992, as supplemented by letter dated April 21, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health ar;d safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; j

D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

j 2.

Accordingly, the license is amended by changes to the Technical Spectfications as indicated in the attachment to this license amendment, and paragraph 3.B. of Facility Operating License No. DPR-29 is hereby amended to read as follows:

u 9506200445 950613 PDR ADOCK 05000237 P

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c-0 2-B.

JEhnical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.155, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance and shall be implemented no later than June 30, 1996.

FOR THE NUCLEAR REGULATORY COMMISSION k

s Robert M. Pulsifer, Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: June 13, 1995

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M ero UNITED STATES y

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,j NUCLEAR REGULATORY COMMISSION t

WASHINGTON, D.C. 2006H001

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COMMONWEALTH EDISON COMPANY b80 IOWA-ILLIN0IS GAS AND ELECTRIC COMPANY DOCKET NO. 50-265 OVAD CITIES NUCLEAR POWER STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 151 License No DPR-30 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Commonwealth Edison Company (the licensee) dated September 15, 1992, as supplemented by letter dated April 21, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that sudi activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, i

and paragraph 3.B. of Facility Operating License No. DPR-30 is hereby 1

amended to read as follows:

j

' i B.

Technical Specifications The Technical Specifications contained in Appendices A and B, as

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revised through Amendment No.151, are hereby incorporated in the l

license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance and shall be implemented no later t' _: lune 30, 1996.

t FOR THE NUCLEAR REGULATORY COMMISSION l

)

r Robert M. Pulsifer, Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

June 13, 1995

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ATTACHMENT TO LICENSE AMENDMENT NO.155 AND 151 FACILITY OPERATING LICENSE N05. OPR-29 AND DPR-30 DOCKET N05. 50-254 AND 50-265 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by the captioned amendment number.

UNIT 1 UNIT 2 REMOVE REMOVE INSERT 1.1/2.1-1 1.1/2.1-1 2-1 1.1/2.1-2 1.1/2.1-2 2-2 1.1/2.1-3 1.1/2.1-2a 2-3 1.1/2.1-4 1.1/2.1-3 2-4 1.1/2.1-5 1.1/2.1-4 2-5 1.1/2.1-6 1.1/2.1-5 8 2-1 1.1/2.1-7 1.1/2.1-6 B 2-2 1.1/2.1-8 1.1/2.1-7 8 2-3 1.1/2.1-9 1.1/2.1-7a B 2-4 1.1/2.1-10 1.1/2.1-8 B 2-5 1.1/2.1-11 1.1/2.1-9 B 2-6 1.1/2.1-12 1.1/2.1-10 B 2-7 1.1/2.1-13 1.1/2.1-11 B 2-8 1.1/2.1-14 1.1/2.1-12 B 2-9 1.1/2.1-15 B 2-10 1.1/2.1-16 B 2-11 1.1/2.1-17 Figure 2.1-1 Figure 2.1-1 Figure 2.1-2 Figure 2.1-2 Figure 2.1-3 Figure 2.1-3 1.2/2.2-1 1.2/2.2-1 1.2/2.2-2 1.2/2.2-2 1.2/2.2-3 1.2/2.2-2a 1.2/2.2-4 1.2/2.2-3 3/4.11-1 3/4.11-2 3/4.11-3 3/4.11-4 B 3/4.11-1 B 3/4.11-2 B 3/4.11-3 3/4.12-1 3/4.12-2 B 3/4.12-1

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SAFETY LIMITS 2.1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS 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: OPERATIONAL MODE (s) 1 and 2.

ACTION:

With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor 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.

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l THERMAL POWER, High Pressure and Hiah Flow 2.1.B The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.07 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 ACTION:

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With MCPR less than the above applicable 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 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.

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QUAD CITIES - UNITS 1 & 2 2-1 Amendment Nos. 155, 151 l

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SAFETY LIMITS 2.1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS Reactor Coolant System Pressure 2.1.C The reactor coolant 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 twelve inches above the top of the active irradiated fuel, i

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

ACTION:

With the reactor vessel water level 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.

QUAD CITIES - UNITS 1 & 2 2-2 Amendment Nos. 155, 151

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LSSS 2.2

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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System (RPS) Instrumentation Setpoints 2.2.A The reactor protection system instrumentation setpoints shall be set consistent 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.

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QUAD CITIES - UNITS 1 & 2 2-3 Amendment Nos. 155, 151

LSSS 2.2 TABLE 2.2.A-1 i

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS r

Functional Unit Trio Setpoint l

1. Intermediate Range Monitor:

i a.

Neutron Flux - High s120/125 divisions of full scale b.

Inoperative NA

2. Average Power Range Monitor:

a.

Setdown Neutron Flux - High

$15% of RATED THERMAL POWER b.

Flow Biased Neutron Flux - High i

1) Dual Recirculation Loop Operation a) Flow Biased sO.58W'd + 62%,

with a maximum of b) High Flow Clamped

$120% of RATED THERMAL POWER

2) Single Recirculation Loop Operation a) Flow Biased sO.58Wid + 58.5%,

with a maximum of b) High Flow Clamped

$116.5% of RATED THERMAL POWER c.

Fixed Neutron Flux - High 5120% of RATED THERMAL POWER d.

Inoperative NA

3. Reactor Vessel Steam Dome Pressure - High

$1060 psig

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

5. Main Steam Line Isolation Valve - Closure

$10% closed

6. Main Steam Line Radiation - High 515 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.

QUAD CITIES - UNITS 1 & 2 24 Amendment Nos. 155, 151

[i LSSS 2.2 TABLE 2.2.A-1 (Continued)

.l REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS l

l Functional Unit Trio Setooint j

7. Drywell Pressure - High s2.5 psig l

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8. Scram Discharge Volume Water Level-High:

540 gallons

9. Turbine Stop Valve - Closure 510% closed

10. Turbine EHC Cor. trol Oil Pressure. Low 2900 psig

11. Turbine Contro! 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

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l OUAD CITIES - UNITS 1 & 2 2-5 Amendment Nos. 155, 151

SAFETY LIMITS B 2.1 BASES 2.1 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 represents 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. 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 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 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 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 t 3sumption 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.

QUAD CITIES - UNITS 1 & 2 B 2-1 Amendment Nos. 155, 151

SAFETY LIMITS B 2.1 BASES 2.1. A THERMAL POWER, Low Pressure or Low Flow This fuel cladding integrity Safety Limit 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 (O% power,0% flow) is greater than 4.56 psi. At low powers and flows, this pressure differential is 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 low powers and flows will always be greater than 4.56 psi. Analyses show that with a bundle flow of 28 x 10'lb/hr, bundle 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 greater than 28 x 103 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.

2.1.B THERMAL POWER, Hinh Pressure and High Flow This fuel cladding integrity Safety Limit is set such that no (mechanistic) fuel damage is cal.:ulated to occur if the limit is not violated. Since the parameters which result in fuel damage are, r.ot directly observable during reactor operation, 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. However, 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 Lirnit 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 transitiori 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 expected. Significant test data accumulated by the NRC and private organizations indicate that the use of a boiling transition limitation to protect against cladding f ailure is a very conservative QUAD CITIES - UNITS 1 & 2 B 2-2 Amendment Nos.155,151

SAFETY LIMITS B 2.1

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

2.1.C Reactor Coolant System Pressure The Safety 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 fuel in the reactor vessel.

1 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 pressures of the reactor pressure vessel and coolant system piping. The respective design pressures are 1250 psig at 575 F and 1175 psig at 560'F. The pressure Safety Limit was chosen as the lower of the pressure transients permitted by the applicable design codes, ASME Boiler and Pressure Vessel Code Section ill for the pressure vessel, and USASI B31.1 Code for the reactor coolant system piping. The 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 lowest 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 with the ASME overpressure protection limit (1375 psig) assures compliance of the suction piping with 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 vendor.

The design basis for the reactor pressure vessel makes evident the substantial margin of protection against failure at the safety pressure 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 compareble 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 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 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 provide adequate margin below the allowable peak vessel pressure of 1375 psig.

QUAD CITIES - UNITS 1 & 2 B 2-3 Amendment Nos. 155, 151

1 SAFETY LIMITS B 2.1 BASES 2.1.D Reactor Vessel Water Level With fuel in the reactor vessel during periods when the reactor is shutdown, consideration must f

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.

1 QUAD CITIES - UNITS 1 & 2 B24 Amendment Nos. 155, 151

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

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

2.2.A Reactor Protection System instrumentation Setpoints The Reactor Protection System (RPS) instrument:nion setpoints specified in the table are the values at which the reactor scrams are set for each r;arameter. The scram settings have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and assist in mitigating the consequences of accidents. Conservatism incorporated into the transient analysis is documented by each approved fuel vendor. The bases for individual scram settings are discussed in the following paragraphs.

1.

Intermediate Range 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 decade,10 range, instrument which covers the range of power level between that covered by the SRM and the APRM. The IRM scram setting at 120 of 125 divisions 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 range; likewise, if the instrument were on Range 5, the scram would be 120 divisions on that range. Thus, as the IRM is ranged up to accommodate the increase in power level, the scram setting is also ranged up.

The most significant sources of reactivity change during the power increase are due to control rod withdrawal. In order to ensure that the IRM provides adequate protection against the single rod withdrawal error, a range of rod withdrawal events has been analyzed. This analysis included starting the event at various power levels. The most severe case involves an initial condition in which the reactor is 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 Safety Limit. Based on the above analysis, the IRM provides protection against local OUAD CITIES - UNITS 1 & 2 B 2-5 Amendment Nos. 155, 151

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

2.

Averaae Power Ranae Monitor For operation at low pressure and low flow during Startup, a reduced power level, i.e., setdown, APRM scram setting of 15% of RATED THERMAL POWER provides adequate thermal margin between the setting and the Safety Limit. The margin is adequate to accommodate anticipated maneuvers associated with power plant startup. Effects of increasing pressure at zero or low void content are minor; cold water from sources available during startup are not much colder than that stready in the system; temperature coefficients 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 withdrawal is the most probable 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 rate of power rise 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 per minute, 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 active until the mode switch is placed in the Run position.

The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady-state conditions, also provides a flow biased neutron flux which reads in percent 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 transfer from the fuel (reactor thermal power) is less than the 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 until it reaches the applicable setting where it is " clamped" at its maximum allowed value. The use of the flow referenced neutron flux scram setting provides additional margin beyond the use of a the fixed high flux scram setting alone.

An increase in the APRM scram setting would decrease the margin present before the fuel cladding integrity Safety Limit is reached. The APRM scram setting was determined by an analysis of margins required to provide a reasonable range for maneuvering during operation. Reducing this operating margin would increase the frequency of spurious scrams, which have an adverse effect on reactor safety because of the resulting thermal stresses. 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.

Amendment Nos. 155, 151 OUAD CITIES - UNITS 1 & 2 B 2-6 i

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LSSS B 2.2 BASES During single recirculation loop operation, the normal drive flow relationship is altered as a result of 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 must be altered to continue 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 scram setting is adjusted in accordance with Specification 3/4.11.8 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 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 Steam Dome Pressure - High 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 operating will also tend to 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 account the location of the pressure measurement (reactor vessel steam space) compared to the highest pressure that occurs in the system during a transient, in compliance with Section lit 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 Amendment Nos.155,151 OUAD CITIES - UNITS 1 & 2 B 2-7 l

LSSS B 2.2

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BASES 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, an 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 zero.

5.

Main Steam Line Isolation Valve - Closure 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 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 valve closure, thus providing protection for the fuel cladding integrity Safety Limit. Operation of the reactor at pressures lower than the MSIV closure setting requires the reactor mode switch to be 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 l

Run position assures the availability of the neutron flux scram protection over the entire range of l

applicability of fuel cladding integrity Safety Limit.

l l

6.

Main Steam Line Radiation - Hiah l

High radiation levels in the main steam line tunnel above that due to the normal nitrogen and oxygen radioactivity are an indication of leaking fuel. When high radiation is detected, a scram is initiated to mitigate the f ailure 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 f

(NFPB) radiation levels without hydrogen addition. With the injection of hydrogen into the l

j feedwater for mitigation of intergranular stress corrosion cracking, the full power background levels may be significantly increased. The setting is sufficiently high to allow the injection of hydrogen without requiring an increase in the setting. This trip function provides an anticipatory scram to limit offsite dose consequences, but is not assumed to occur in the analysis of any design basis event.

I QUAD CITIES - UNITS 1 & 2 B 2-8 Amendment Nos. 155, 151 I

l i

1

\\

LSSS B 2.2 BASES 7.

Drywell Pressure - Hiah High pressure in the drywell could indicate a break in the primary pressure boundary systems or a loss of drywell cooling. Therefore, pressure sensing instrumentation is provided as a backup to the water level instrumentation. 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 setting was selected as low as possible without causing spurious scrams.

8.

Scram Discharae Volume Water Level-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 discharge volumes. These 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 design 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 discharge volume 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, 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 the design and logic of the system to prevent common mode failure. 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 leakage per drive into the scram discharge volume. As indicated above, there is sufficient volume in the piping to accommodate the scram without impairment of the scram times or the amount of insertion of the control rods.

9.

Turbine Stop Valve - Closure The turbine 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 valve closure from full open, the resultant increase in surface heat flux is limited such that MCPR remains above the fuel cladding integrity Safety Limit, even during the worst-case transient that assumes the turbine bypass f ails to operate.

10.

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

This fast closure of the turbine control valves is not protected by the turbine control valve fast OUAD CITIES - UNITS 1 & 2 B 2-9 Amendment Nos.155,151

LSSS B 2.2 BASES closure scram since failure of the oil system would not result in the fast closure solenoid valves being actuated. For a turbine control valve fast closure, the core would be protected by the APRM and reactor high pressure scrams. However, to provide the same margins as provided for the generator load rejection on fast closure of the turbine control valves, a scram has been added to the reactor protection system which senses failure of control oil pressure to the turbine control system. This scram anticipates the pressure transient which would be caused by imminent control valve closure and results in reactor shutdown before any significant increase in neutron flux occurs. The transient response is very similar to that resulting from the turbine control valve fast closure scram. However, since the control valves will not start to close until the fluid pressure is approximately 600 psig, the scram on low turbine EHC control oil pressure occurs well before turbine control valve closure begins. The scram setting is high enough to provide the necessary anticipatory function and low enough to minimize the number of spurious scrams.

11.

Turbine Control Valve Fast Closure The turbine control valve fast closure scram is provided to anticipate the rapid increase in pressure and neutron flux resulting from fast closure of the turbine control valves due to a load rejection and subsequent failure of the bypass valves; i.e., MCPR remains above the fuel cladding integrity i

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 wide margin to the value corresponding to 1% plastic strain of the cladding.

The scram setting based on EHC fluid pressure was developed to ensure that the pressure switch is actuated prior to the closure of the turbine control valves (at approximately 400 psig EHC fluid pressure), 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.

Turbine Condenser Vacuum - Low Loss of condenser vacuum occurs when the condenser can no longer handle the heat input. Loss of condenser vacuum initiates a closure of the turbine stop valves and turbine bypass valves which eliminates the heat input to the condenser. Closure of the turbine stop and bypass valves causes a pressure transient, neutron flux rise and an increase in surface heat flux. To prevent the fuel cladding integrity Safety Limit from being exceeded if this occurs, a reactor scram occurs on turbine stop valve closure. The turbine stop valve closure scram function alone is adequate to j

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

QUAD CITIES - UNITS 1 & 2 B 2-10 Amendment Nos. 155, 151

LSSS B 2.2 BASES 13.

Reactor Mode 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 maneal scram is a redundant CHANNEL to the automatic protective instrumentation CHANNEL (s) and provides manual reactor scram capability.

1 l

I QUAD CITIES - UNITS 1 & 2 B 2-11 Amendment Nos. 155, 151 i

l

POWER DISTRIBUTION LIMITS AFLHGR 3/4.11.A 3.11 - L8MITING CONDITIONS FOR OPERATION 4.11 - SURVEILLANCE REQUIREMENTS i

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 AVERAGE CORE OPERATING LIMITS REPORT.

PLANAR EXPOSURE shall not exceed the 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 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 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 3/4.11-1 Amendment Nos. 155, 151

POWER DISTRIBUTION LIMITS APRM Satpoints 3/4.11.B i

3.11 - LIMITING CONDITIONS FOR OPERATION 4.11 - SURVEILLANCE REQUIREMENTS B.

Average Power Range Monitor Setpoints B.

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

The value of MFLPD shall be verified:

gain or setpoints shall be set such that the MAXIMUM FRACTION OF LIMITING 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 />, 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).

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 /> when the reactor is operating with OPERATIONAL MODE 1, when THERMAL MFLPD greater than or equal to FRTP.

POWER is greater than or equal to 25% of RATED THERMAL POWER.

4.

The provisions of Specification 4.0.D i

are not applicable, j

ACTION:

With MFLPD greater than FRTP, 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 MFLPD to within its limit, or 2.

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

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

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 />, i

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 panet.

QUAD CITIES - UNITS 1 & 2 3/4.11-2 Amendment Nos. 155, 151

l 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,,, = 0.86 prior to performance of the OPERATING LIMITS REPORT.

initial scram time measurements for the cycle in accordance with Specification 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 MODE 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 determined to be equal to or greater than the applicable MCPR operating ACTION:

limit specified in the CORE OPERATING l

LIMITS REPORT.

l With MCPR less than the applicable MCPR operating 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 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 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 />.

1 1

4 1

QUAD CITIES - UNITS 1 & 2 3/4.11-3 Amendment Nos.155,.151 i

POWER DISTRIBUTION LIMITS LHGR 3/4.11.D 3.11 - LIMITING CONDITIONS FOR OPERATION

'4.11 - SURVEILLANCE REQUIREMENTS D.

LINE.\\R HEAT GENERATION RATE D.

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

exceed the limits specified in the CORE OPERATING LIMITS REPORT.

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 />, 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 APPLICABILITY:

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 per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> RATED THERMAL PO'NER.

when the reactor is operating with a LIMITING CONTROL ROD PATTERN for LHGR.

ACTION:

4.

The provisions of Specification 4.0.D With a LHGR exceeding the limits specified are not applicable.

in the CORE OPERATING LIMITS REPORT:

1.

Initiate corrective ACTION within 15 minutes, and 2.

Restore the LHGR 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 Nos. 155, 151

POWER DISTRIBUTION LIMITS B 3/4.11 BASES l

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 mechanical integrity is maintained during normal and transient operations.

The peak cladding temperature (PCT) following a postulated loss-of coolant accident is primarily a function of the average 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 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 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 ut 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 thermal limit.

3/4.11.8 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 contro rod block functions of the APRM instruments for both two recirculation loop operation and single recirculation loop operation must be adjusted to ensure that the MCPR does not become less than the fuel cladding safety limit or that 21% 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 th? value of MFLPD indicates a higher peaked power distribution to ensure that an LHGR transient would not be increased in the degraded condition.

QUAD CITIES - UNITS 1 & 2 B 3/4.11-1 Amendment Nos. 155,.151

~

POWER DISTRIBUTION LIMITS 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 analysis evaluation with the initial condition of the reactor being at the steady state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during 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 analyzed 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 MCPR multiplicative factor specified in the CORE OPERATING LIMITS REPORT is to define MCPR operating limits at other than rated core flow conditions. At less than 100% of rated flow, the required MCPR 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 conservatism in the scram speed performance, it must be demonstrated that the specific scram speed distribution is 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 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 pump speed and the moderator void content will b

• 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 calculating 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.

QUAD CITIES - UNITS 1 & 2 B 3/4.11-2 Amendment Nos. 155, 151

~.

. = _.

POWER DISTRIBUTION LIMITS B 3/4.11 BASES 3/4.11.D LINEAR HEAT GENERATION RATE This specification assures that the LINEAR HEAT GENERATION RATE (LHGR) in any fuel rod is less than the design linear heat generation even if fuel pellet densification is postulated. The daily requirement for calculating LHGR 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 LHGR 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 thermal limits are met after power distribution shifts while still allotting time for the power distribution to stabilize. The requirement for calculating LHGR after initially determining a LIMITING CONTROL ROD PATTERN exists ensures that LHGR will be known following a change in THERMAL POWER or power shape that could place operation above a thermal limit.

QUAD CITIES - UNITS 1 & 2 B 3/4.11-3 Amendment Nos. 155, 151

4 I

SPECIAL TEST EXCEPTIONS PCI 3/4.12.A

)

3.12 - LIMITING CONDITIONS FOR OPERATION 4.12 - SURVEILLANCE REQUIREMENTS t

A.

PRIMARY CONTAINMENT INTEGRITY A.

PRIMARY CONTAINMENT INTEGRITY r

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 head and the drywell head to power PHYSICS TESTS.

be removed and the primary containment air lock doors to be open when the reactor mode switch is in the Startup position during low power PHYSICS TESTS with 1

THERMAL POWER less than 1% of RATED i

THERMAL POWER and reactor coolant l

temperature less than 212*F.

APPLICABILITY:

l 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 i

greater than or equal to 212'F, immediately place the reactor mode switch in the Shutdown position.

QUAD CITIES - UNITS 1 & 2 3/4.12-1 Amendment Nos. 155, 151 k

SPECIAL TEST EXCEPTIONS 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 i

3.10.C and Table 1-2 may be suspended to 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 t

permit the reactor rnode switch to be in the of a SHUTDOWN MARGIN demonstration, Startup position and to allow more than one verify 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 i

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 minimizer is OPERABLE

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

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

The rod worth minimizer is OPERABLE individualis 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 i

MARGIN demonstrat;on procedure is 3.

No other CORE ALTERATION (s) are in i

verified by a second licensed operator progress.

or other technically qualified individual.

f i

3.

The " rod-out-notch-override" control l

shall not be used during out-of-sequence movement of the control f

rods.

4.

No other CORE ALTERATION (s) are in progress.

APPLICABILITY:

OPERATIONAL MODE 5, during i

SHUTDOWN MARGIN demonstrations.

j ACTION:

With the requirements of the above l

specification not satisfied, immediately j

place the reactor mode switch in the Shutdown or Refuel position.

QUAD CITIES - UNITS 1 & 2 3/4.12-2 Amendment Nos. 155, 151

1 4

SPECIAL TEST EXCEPTIONS B 3/4.12 BASES 3/4.12. A PRIMARY CONTAINMENT INTEGRITY The requirement for PRIMARY CONTAINMENT INTEGRITY is not applicable ouring 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 containmcnt 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 integrity.

3/4.12.B SHUTDOWN MARGIN Demonstrations Performance 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 MARGIN demonstrations performed while in OPERATIONAL MODE 5, additional requirements must be met to ensure that adequate protection against potential reactivity j

excursions is available. Because multiple control rods will be withdrawn and the reactor will i

l 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 J

"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. Becauce 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 3/4.121 Amendment Nos. 155. 151

_ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _.