ML20138L401

Proposed Tech Specs Section 2.1.B Re Thermal Power,Section 3/4.11 Re Power Distribution Limits,Section 3/4.6 Re Primary Sys Boundary,Section 5.3 Re Reactor Core & Section 6.9 Re Reporting Requirements

ML20138L401
Person / Time
Site:	Quad Cities
Issue date:	02/17/1997
From:	COMMONWEALTH EDISON CO.
To:
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ML20138L399	List: ML20138L396 ML20138L401
References
NUDOCS 9702240118
Download: ML20138L401 (7)

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,- SAFETY UMITS B 2.1

. BASES .

r l Lid THERMAL POWER, Low Pressure or Low Flow ,

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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),  ;

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

28 x 108 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 108 lb/hr.

Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel-

cssembly critical power at this flow is approximately 3.35 MWt. At 25% of RATED THERMAL

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l POWER, the peak powered bundle would have to be operating at 3.86 times the average powered  :

l bundle in o'rder to achieve this bundle powen Thus, a core thermal power limit of 25_% kr -Mar i j

pressures below 785 psig is conservative.

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! L1,,& THERMAL POWER. Hioh Pressure and Hioh Flow 4/ 6 UN O 3 '

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! This fuel cladding integrity Safety Umit is set such that no (mechanistic) fuel damage is calculated '

! to occur if the limit is not violated. Since the parameters which result in fuel damage are not

! directly observable during reactor operation, the thermal and hydraulic conditions resulting in j' d:parture 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 nicessarily result in damage to BWR fuel rods, the critical power ratio (CPR) at which boiling I l trcnsition 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 I critical power result in an uncertainty in the value of the critical power. Therefore, the fueg

! c!rdding integrity Safety Limit is defined = $: 0"". F *- "9;; S:: ::::-d:/ " %!:.. more I

than 99.9% of the fuel rods in the core are expected to avoid boiling transitiort ::.;#:4 the i

j pswer distribution within the. core and all uncertainties. 4 74ti 7,v /e/

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

!' confidence that operation of a fuel assembly at the condition where MCPR is equal to the fuel ci:dding integrity Sefety Limit would not produce transition boiling. In addition, during single

'l j r: circulation loop operation, the MCPR Safety Limit is increased by 0.01 to conservatively account '

far increased uncertainties in the core flow and TIP measurements.

i Hswever, if transition boiling were to occur, cladding perforation would not necessarily be i cxpected. Significant test data accumulated by the NRC and private organizations indicate that the u:e of a boiling transition limitation to protect against cladding failure is a very conservative ,

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QUAD CITIES - UNITS 1 & 2 B22 Amendment Nos. mam  !

$ 9702240118 970217

' PDR ADOCK 05000254 ',

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PRIMARY SYSTEM BOUNDARY B 3/4.6 BASES' I

3/4.6.E Safetv Valves 5()( y Nolo/oq OGNn 4e.r'16 Ao.r g, ,g.

3/4.8.F _ Relief Valves gac/ y c/g, The American Society of Mechanical Engineers (ASME) Boiler a Pressure Vessel Code requires tha reactor pressure vessel be protected from overpressure du g upset conditions by self-cetuated safety valves. As part of the nuclear pressure relie system, the size and number of .

. safety valves are selected such that peak pressure in the n clear system will not exceed the ASME  ;

Code limits for the reactor coolant pressure boundary. overpressure protection system must  !

, accommodate the most severe pressurization transie .

aluations have determined that the most j severe transient is the closure of all the main steam line isolation valves followed by a reactor ,

scram on high neutron flux. The analvuis results demonstrate that the design safety valve capacity is capable of maintaining reactor pressure below the ASME Code limit of 110% of the reactor

. pr:ssure vessei design pressure.

Thn relief valve function is not assumed to operate in response to any accident, but are provided to remove the generated steam flow upon turbine stop valve closure coincident with failure of the turbine bypass system. The relief vcive opening pressure settings are sufficiently low to prevent tha need for safety valve actuation following such a transiertt. -

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Erch of the five relief valves discharge to the suppression chamber via a dedicated relief valve  !

' discharge line. Steam remaining in the relief valve discharge line following closure can condense,  ;

cr ting a vacuum which may draw suppression pool water up into the discharge line. This .

condition is normally alleviated by the vacuum breakers; however, subsequent actuation in the l i

presence of an elevated water leg can result in unacceptably'high thrust loads on the discharge piping. To prevent this, the relief valves have been designed to ensure that each valve which ci:ses .will remain closed until the normal water level in the relief valve discharge line is restored.

l Tha opening and closing setpoints are set such that all pressure induced subsequent actuation are limited to the two lowest set valves. These two valves are equipped with additional logic which

) functions in conjunction with the setpoints to inhibit valve reopening during the elevated water leg duration time following each closure. -

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jL Each safety / relief valve is equipped with diverse position indicators which monitor the tailpipe ccoustic vibration and temperature. Either of these provide sufficient indication of. safety / relief  !

valve position for normal operation. -

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3/4.6.G Leakaae Detection Systems i

l Tha RCS leakage detection systems required by this specification are provided to monitor and d:;tret leakage from the reactor coolant pressure boundary. Limits on leakage from the reactor l

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lint pressure boundary are required so that appropriate action can be taken before the integrity cf the reactor coolant pressure boundary is impaired. I.eakage detection systems for the reactor

}  ; c;: lint system are provided to alert the operators when leakage rates above the normal background levels are detected and also to supply , quantitative measurement of leakage rates.

QUAD CITIEG - Ul!!TO 1 & 2 g m ,3 ,r$g,hestu, n. . f. ur n .

POWER DISTRIBUTION UMITS B 3/4.11

, BASES 3/4.11. A AVERAGE PLANAR UNEAR HEAT GENERATION RATE Gt M/ I ~

This specnication assures that the peak cladding temperature following the postulated de 1::ss-of-coolant accident will not exceed the limit specified in 10 CFR 50.46. The specification cssures 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 function of the average heat generation rate of all the rods of a feel assembly at any axialloc and is dependent only secondarily on the rod-to-rod power distribution within an assemb peak clad temperature is calculated assuming a UNEAR HEAT GENERATION RATE (LHGR) fo 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 assu in the LOCA analysis divided by its local peaking factor. A conservative multiplier is applie f HGR assumed in the LOCA analysis to account for the uncertainty associated with the measurement of the A LHGR. ryg gj

- The calculational proceddre used to esfEBiish me ....m;...um~APLHGR values uses NRC a

,criculational models which are consistent with the requirements of Appendix K of 10 CFR Part The approved calculational models are listed in Specification 6.9.

The daily requirement for calculating APLHGR when THERMAL POWER is greater than o 25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow w 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%

I RATED THERMAL POWER ensures thermal limits are met after power distribution shifts while cllotting time for the power distribution to stabilize. The requirement for calculating APLHGR afte i

initially determining a UMITING CONTROL ROD PATTERN exists ensures that APLHGR will be known following a change in THERMAL POWER or power shape, that could place operation a ermal limit.

3/4.11.B -Am.vi 5c , < CWTO TA%A/S/E#7 /./A/E44 #MMM NME wh ala i il d u high scram setting and control rod block functions of the APRM instruments for both two j r2 circulation loop operation and single recirculation loop operation must be adjusted to

ths MCPR does not become less than the fuel cladding safety limit or that 21% plastic str not occur in the degraded situation. The scram settings and rod block settings are adjusted i creased in thecccordance with the formula in this specificatio praked power distritsution to ensure that an LHGR transient would not be ir digraded condition.

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

4 171 & 47 a

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REACTOR CORE 5.3 5.0 DESIGN FEATURES

. . l 5.3 REACTOR CORE i f{, ,,3e,,[/,'eJ p og foMai

u,de, rxis or wl" b""' ,

Fuel Assemblies -

i 5.3:A The reactor core shall contain 724 fuel assem s. Each assembly consists of a matrix of Zircaloy clad fuel rods with an ini omposition of natural or slightly enriched uranium dioxide as fuel material.:_.2.; 9 Umited substitutions of zf rcal y hy, in accordance with NRC-approved applications of fuel rod i

" configurations, may be used. Fuel assemblies shall be limited to those fuel dec!gne  !

2ino ~~ that have been analyzed with applicable NRC staff-approved codes and methods, and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be  !

placed in non-limiting core regions.

Control Rod Assemblies 5.3.8 The reactor core shall conteln 177 cruciform shaped control rod assemblies. The control material shall be boron carbide powder (B C) and/or hafnium metal. The '

control rod assembly shall have a nominal axial absorber length of 143 inches.  ;

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QUAD CITIES - UNITS 1 & 2 5-5 Amendment Nos.

s J

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Rep;rting Requirements 6.9

, ADMINISTRATIVE CONTROLS k'*

. (3) Commonwealth Edison Topical Report NFSR-OO85, Supplement 1, -

" Benchmark of BWR Nuclear Design Methods - Quad Cities Gamma Seas.

Comparisons," (latest approved revision).

fs sad '

g (4) Commonwealth Edison Topical Report NFSR OO85', Supplement 2,

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• Benchmark of BWR Nuclear Design Methods - Neutronic Licensing Analyses," (latest approved revision).

c.

The core operating limits shall be determined so thet all applicable limit thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nucle such as shutdown margin, and transient and accident analysis limits) oj analysis are met. The CORE OPERATING UMITS REPORT, including any m 1 cycle revisions or supplements thereto, shall be provided upon issuance, for elI reioad cycle, to the NRC Document Control Desk with copies to the Regiol Administrator and Resident inspector.

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6.9.B Special Reports Office within the time period specified for each report.Spe I

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l QUAD CITIES - UNITS 1 & 2 6 16 Amendment Nos.

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Insert #11 i

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. (5) Advanced Nuclear Fuels Methodology for Boiling Water Reactors, XN-NF-80-19(P)(A), Volume 1, Supplement 3, Supplement 3 Appendix F, and Supplement 4, Advanced Nuclear Fuels Corporation, November 1990.  !

(6) Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC '

Methodology to BWR Reloads, XN-NF-80-19(P)(A), Volume 4, Revision 1, Exxon - i Nuclear Company, June 1986. i (7) Exxon Nuclear Methodology for Boiling Water Reactors THERMEX: Thermal Limits Methodology Summary Description, XN-NF-80-19(P)(A), Volume 3, Revision 2, Exxon Nuclear Company, January 1987.

j!

(8). Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis, XN-NF-80-19(P)(A) Volume 1 and Supplements 1 and 2, Exxon i Nuclear Company, March 1983. *

(9) Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, XN-NF-85-67(P)(A) Revision 1, Exxon Nuclear Company, September 1986.  !

h (10) Qualification of Exxon Nuclear Fuel for Extended Burnap Supplement 1: Extended i Burnup Qualification of ENC 9x9 BWR Fuel, XN-NF-82-06(P)(A) Supplement 1, Revision 2, Advanced Nuclear Fuels Corporation, May 1988.

(11) Advanced Nuclear Fuels Corporation Generic Mechanical Design for Advanced Nuclear Fuels 9x9-IX and 9x9-9X BWR Reload Fuel, ANF-89-014(P)(A), Revision 1 l

, and Supplements I and 2, Advanced Nuclear Fuels Corporation, October 1991.

(12) Generic Mechanical Design Criteria for BWR Fuel Designs, ANF-89-98(P)(A) l Revision 1, and Revision 1 Supplement 1, Advanced Nuclear Fuels Corporation, May  ;

1995. '

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.(13) Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors, .

XN-NF-79-71(P)(A), Revision 2 Supplements 1,2, and 3, Exxon Nuclear Company, 1 March 1986.

(14) ANFB Critical Power Correlation, ANF-1125(P)(A) and Supplements I and 2,

' Advanced Nuclear Fuels Corporation, April 1990.

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Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water licactors/ Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors: Methodology for Analysis of Assembly Channel Bowing Effects /NRC Correspondence, ANF-524(P)(A),' Revision 2, Supplement.1 Revision 2, Supplement 2, Advanced Nuclear Fuels Corporation, November 1990.

(16) COTRANSA 2: A Computer Program for Boiling Water Reactor Transient Analyses, ANF-913(P)(A) Volume 1 Revision I and Volume 1 Supplements 2,3, and 4, Advanced Nuclear Fuels Corporation, August 1990.

(17) Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model, ANF-91-048(P)(A), Advanced Nuclear Fuels Corporation, January 1993.

,(18) Commonwealth Edison Topical Report NFSR-Ob91," Benchmark'of CASMO/MICROBURN BWR Nuclear Design Methods," Revision 0, Supplements 1 and 2, December 1991, March 1992, and May 1992, respectively; SER letter dated March 22,1993.

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