ML20148N481

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Proposed Tech Specs,Supporting Operation of Unit During Cycle 10 by Changing MAPLHGR Limits for Reload Fuel & Adjustment of Min Critical Ratio
ML20148N481
Person / Time
Site: Quad Cities Constellation icon.png
Issue date: 03/28/1988
From:
COMMONWEALTH EDISON CO.
To:
Shared Package
ML20148N405 List:
References
4409K, NUDOCS 8804070035
Download: ML20148N481 (26)
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Text

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r ATTACM Wif 3 ,

(cont'd) l B. QUAD CITIES UNIT 2 CYCLE 10 LICENSE E .

l TECHNICAL GPECIFICATIj;3! CHANGES License page 3 Technical Specification pages: 1.1/2.1-1 1.1/2.1-4 1.1/2.1-7 Fig. 2.1-3 3.2/4.2-14 3.2/4.2-14a l 3.3/4.3-5 l 3.5/4.5-5 l 3.5/4.5-10  !

3.5/4.5-12 3.5/4.5-14 [

3.5/3.5-14e 3.5/3.5-14b ,

Fig. 3.5-1 Sheet 1 ,

Fig. 3.5-1 Sheet 2 I Fig. 3.5-1 Sheet 3 Fig. 3.5-1 Sheet 4 Fig. 3.5-1 Sheet 5 Fig. 3.5-2  ;

3.6/4.6-5a 1.6/4.6-13a ,

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4409K

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8804070035 880328 PDR ADOCK 05000265 P PDR ,

OPR-30 F 4  ;

4 C. Pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time, any byproduce, source and special nuclear materials as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring l equipment calibration, and as fission detectors in amounts required; Am 36 0. Pursuant to the Act and 10 CFR Part 30, 40 and 70, to receive, ,

2/03/77 possess and use in amounts as required any byproduct, source and i special nuclear materials without restriction to chemical or physical form, for sample analysis or instrument and equipment calibration or associated with radioactive apparatus or components; ,

Am. 41 1/30/78 E. Pursuant to the Act and 10 CFR Parts 30 and 70, possess, but not '

separate, such byproduct and special nuclear materials as may be produced by the operation of Quad Cities Nuclear Power Station, Unit Nos. I and 2.

1

3. This license shall be deemed to contain and is subject to the F conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30. Section 40.41 of Part .

40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

A. Maximum Power Level Commonwealth Edison is authorized to operated Quad Cites unit No. '

2 at power levels not in excess of 2511 megawatts (thermal).

Am. 99 B. Technical Specifications 8/06/88 The Technical Specifications contained in Appendices A and B as revised through Amendment No. 99 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.  ;

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i 09268

00AD CITIE$

i DPR-30 1.1/2.1 Ft)EL CLADDING INTE6RI?Y $

$AFETY LIMIT t!MITING $AFETY $YSTEM SETTING

  1. pplicability: Applicability:

The tafety limits established to preserve The limiting safety system settings apply

the fuel cladding integrity apply to to trip settings of the instruments and .

those variables which monitor the fuel devices which are provided to prevent the I thermal behavior. fuel c %dding integrity safety limits from being exceeded, i Objective: Objective:

The objective of the safety limits is to The objective of the limiting safety sys-establish limits below which the integ+ tem settings is to define the level of rity of thw fuel cladding is preserved. the process variables at which automatic protective action is initiated to pre-vent the fuel cladding integrity safety limits from being exceeded.

SPLCIFICATIONS 1 A. Reactor Pressure > 800 psig and Core A. Neutron Flux Trip settings Flow > 10% of Rated 1

The existence of a minimum critical The limiting safety system trip set- ,

power ratio (MCPR) less than 1.04 tings shall be as specified below:

shall constitute violation of the ,

fuel cladding integrity safety limit. 1. APRM Flux Scram Trip $etting F (Run Mode)

8. Core Thermal Power Limit (Reactor Pressure 1 800 psig) When the reacto* mode switch is in the Run position the APRM When the reactor pressure is 1 800 flux scram setting shall be as l psig or core flov is less than 10% of shown in Figure 2.1.1 and shall 4

rated the core thermal power shall be:

i not exceed 25% of rated thermal power.

$ i (.!8WD + 62)

C. Power Transient with a maximum selpoint of 120%

1. The neutron flux shall not for core flow equal to 98 x i exceed the scram setting estab- 10' lb/hr and greater. -

Itshed in Specirteation 2.1A for Ionger than 1.5 seconds as indt- where [

f cated by the process computer.  ;

$4 setting in percent of rated t When the process computer is out power  ;

2.

of service. this safety simit shall be assumed to be exceeded WD = percent of drive flow if the neutron flux exceeds the required to produce a rated core i scram setting estabitshed by flow of 98 million 1b/hr. In  ;

< Specification 2.1,A and a con- the event of operation with a l trol rod scram does not occur, maximum fraction of limiting 6 3'

power density (NFLPD) greater [

than the fraction of rated power e (FRP). the setting shall be f modified as follows:

EM i 51 (.58WD + 62 ( MFLPD ] [

1 l i i

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7 t9258 1.1/2.1-1 Amen Dent No. k j

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Qua0 CITIES DPR-30 1,1 $AFETY LIMIT BA$!$

The fuel cladding integrity limit is set such that no calculated fuel damage would occur as a result of an abnorral operational transient. Because fuel damage is not directly observabic, a step-back approach is used to establish a safety limit such that the einimum critical power ratio (MCPR) is no less than the fuel cladding integrity safety limit MCPR > the fuel cladding integrity safety limit represents a conservative nurgin relattve to the conditions required to maintain fuel cladding integrity, the fuel claddtng is one of the physical boundaries which separate radioactive materials from the environs. The integrity of the fuel cladding is related to its relative treedom from perforations or cracking.

Although sane corroston or use-related cracking may occur during the life of the cladding, fission product migratton f rom this source is increnentally cumulative an/

continuously measurable. Fuel cladding perforaticis, 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 perforation 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 safety limit is defined with margin to the conditions which would produce onset of transition boiling (MCPR of 1.0). These conditions represent a signific. ant departure from the condition intended by design for planned operation. Therefore, the fuel cladding integrity saf ety limit is established such that no calculated fuel damage shall result f rom an abnormal operational transient. Basis of the values derived for this safety Itmit for each fuel type is documented in Reference 1 and 2.

A. Reactor Pressure > 800 psig and Core Flow > 10% of Rated Onset of transition boiling results in a decrease in heat transfer from the cladding and therefore elevated cladding temperature and the possibility of cladding failure. However, the existence of critical power, or boiling transttien is not a directly observable parameter in an operating reactor.

Therefore, the margin to boiling transition is calculated from plant operating parameters such as core power, core flow, feedwater temperature, and core power dtstribution, The margin for each fuel assenbly 15 characterized by the critical power ratto (CPR), which is the ratio for the bundle power which would produce onset of transition boiling divided by the actual bundle power.

The minimum value of this ratio for any bundle in the core is the minimum critical power ratio (MCPR). It is assumed that the plant operation is controlled to the nomtnal protective setpotnts via the instrumented variables (Figure 2.1-3).

- The MCPR fuel cladding integrity saf ety limit has suf ficient conservatism to assure that in the event of an abnormal operational transient initiated from the normal operation condition, more than 99.9% of the fuel rods in the core are expected to avoid boiling transition. The margin between HCPR of 1.0 (onset of transition boiling) and the safety limit, is derived f rom a detailed statistical analysis considering all of the uncertainties in monitorir,g the core operating state, including uncertainty in the bolling transition correlation (see e.g., Reference 1). Because the botling transition correlation is based on a large quantity of full-scale data, there is a very high confidence that operation of a fuel assembly at the condition of HCPR -

the fuel cladding integrity safety Itmit would not produce boiling transition.

However, if boiling transition dere to occur, Cladding perforation would lot be espected. Cladding temperature would increase to approximately 1100*F, which is below the perforation temperature of the cladding material. This had been verified by tests in the General Electric fest Reactor (GETR), where similar fuel operated acove the critical heat flux for a signi,'icant period of time (30 minutes) without cladding perforatien.

If reactor pressure should ever exceed 1400 psia during normal power operation (the limit of applicability of the botling transition correlation), it would be assuned that the fusl Cladding integrity saf ety limit has been violated.

In addition to the boiling transition limit (MCPR) operation 15 constrained to a mautmum LHGR of 13.4 kw/f t f or fuel typ,4 P8u8R and BP8a8R, and 14.4 kw/f t l for fuel types GE8x8E and GE8u8E8. Inis constraint 15 established by I 5pectftcation 3.5.J. to provide adequate safety margin to it plastic strain for abnorpul operating transients initiated f rom high power conditions.

Specification 2.1.A.1 provides for equivalent safety margin for transients initiated from lower poner conditions by adjusting the APRM flow-biased scram setting by the ratto of FRP/MFtro.

09568 1.1/2.1-4 Amendment No.

QUA0 C8VMS OPR-30

> 2.1 LIMITING SAFETY SYSTEM SETTING BASES i

- The abnormal operational transients applicable to operation of the units have been analyzed throughout the spectrum of planned operating conditions in

accordance with Regulatory Guide 1.49. In addition, 2511 MHt is the licensed maximum steady-state power level of the units. This maximum steady-state power level will never knowingly be exceeded.

5 Conservatism incorporated into the transient analysis is documented in References I and 2. Transient analyses are initiated at the conditions given in

these References.

The scram delay-time and rate of rod insertion allowed by the analyses are conservatively set equal to the longest delay and slowest insertion rate acceptable by technical specifications. The effects of scram worth, scram delay time, and rod insertion rate, all conservatively applied, are of greatest significance in the early portion of the negative reactivtty insertion. The rapid insertion of negative reactivity is assured by the time requirements for 5% and 20% insertion. By the time the rods are 60% inserted, approximately 4 dollars of negative reactivity have been inserted, which strongly turns the transient and accomplishes the desired effect. The times for 50% and 90%

insertion are given to assure proper completion of the expected performance in the earlier portion of the transient, and to establish the ultimate fully shutdown steady-state condition.

The MCPR operating limit is, however, aojusted to account for the statistical variation of measured scram times as discussed in Reference 2 and the bases of Specification 3.5.K.

Steady-state operation without forced ccirculation will not be permitted except during startup testing. The analysis to support operation at various power and flow relationships has considered operation with either one or two recirculaticq pumps.

The bases for individual trip settings are discussed in the following paragraphs.

For analyses of the thermal consequences of the transients, the MCPR's stated in Paragraph 3.5.K as the limiting condition of operation bound those which are conservatively assumed to exist prior to initiation of the transients.

A. Neutron Flux Trip Settings

1. APRM Flux Scram trip Setting (Run Mode)

The average power range monitoring (APRM) system, which is calibrated using heat balance data taken during steady-state conditions, reads in percent of rated thermal power. Because fission chamoers provide the basis input signals, the APRM system responds directly to average neutron flux. During transients the instantaneous rate of heat transfer / rom the fuel (reactor thermal power) is less than the ins'antaneous neutron flux due to the time constant of the fuel.

0926B 1.1/2.1-7 Amendment No.

5 DPR- 30 160 _

A?M BACKUF *SCRAM- - * - * - - ~ * * - * - + - - - * - - ~ + = = ==" -*-****-

120 -

- AFM RCD BLOCK ,

LINE (0.58VD + 30) /

" (100,87) (100,108)**

100 s'

AFM 5 CRAM e m LINE (0.58WD + 62 ,e

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Cll '

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80 , '

% p

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  • NATURAL ,'

S CIRCULATION NOMINAL, CON 5fANT XENON g tg3g / 100/100 POWER /ft0V LINE

g. 60 -

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/ $ Operating Region Supported g / Sy N.E.D.C. - 24167 and u N.E.D.C. 22192 j

H *0perating on Stagle Loop or Natural Circulation is o

60 -

Limited per Tech. Spoo.

U 3.6.N.3 and 2.1.A.4

, 201 FVMF

/ SPEED LINE **0peration at greater than rated core flow is supported by NEDC-31449

,0 RATED CONDITIONS 70VtR 2$11 MWth CORE TLOW 98 K1be/MR 1 I '

1 1 0 -

60 40 100 120 O 20 60 W C011 FLOV Raft (10F RATED)

T TICURE 2.1-3

($CNDIATIC)

Amendment No. AFM Tt0V 31A5 SCRAM RILAT10N5N1P TO NOL%L CPERATING CONDITIONS

. QUAO-CITIES

. OPR-30 TABLE 3.2-1 l INSTRUMENTATION THAT INITIATES R00 BLOCK Minimum Number of Operable or Tripped Instrume't channels per e frio Systqst,Ul Instrument Trin Level StLM ng 2 APRM upscale (flow bias)(7} 1(0.58WD + 50) ERf_ (2)

MFLPD 2 APRM upscJle (Refuel and 112/125 full scale

$tartup/ Hot Standby mode) 2 APRM downscaleI73 13/125 full scale 1

Rodblocgnitorupscale(flow 10.65WD , 43( ]( 0]

bias) 1 Rod block monitor downscaleI7} 13/125 full scale 3 1RM downscale(3) (8] 13/125 full scale 3 IRH upscaleI8I 1108/125 full scale 2(5) $RMdetector(gtinStartup 12 feet below core centerline position 3 IRMdetectorggtinStartup 12 feet below core centerline

! position I J 2[5] (6) $RM upscale 1105 counts /sec 2(Il SRM downscale (93 110 2 counts /see 1 (per bank) High water level in scram i 25 gallons (per bank) discharge volume ($DV) i 50V high water level scram NA trip bypassed j

5.Q.Lai

1. For the $tartup/Het Standby and Run positions of the reactor mode selector switch, there shall be two operable or tripped trip systems for each ,

function except the $RM rod blocirs. IRM upscale and IRM downscale need not be operable in the Run position. APRM downscale. APRM upscale (flow biased).

< and RSM downscale need not be operable in the Startup/ Hot Standby mode. The RBM upscale need not be operable at less than 301 rated thermal power. One channel may be bypassed above 30% rated thermal power provided that a limiting control rod pattern does not exist. For systems with more than one channel per trip system if the first colupn cannot be met for one of the l

two trip systems this condition may exist for up to 7 days provided that j during that time the operable system is functionally tested intnediately and J daily thereaf ttr; if this condition lasts longer than 7 days the system i shall be tripped. If the first column cannot be met for both trip systems.

l the systems shall be tripped.

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l 09256 3.2/4.2-14 Amenenent No.

QUAD-Cl?IES s

OPR-30

2. W D is the percent of drive flow required to produce a rated core flow of 98 million 1b/hr. Trip level setting is in percent of rated power (2511 MWt).
3. IRM cownscale may be bypassed when it is on its lowest range.
4. This function is bypassed when the count rate is 1100 cos.
5. One of the four SRM inputs may be bypassed.
6. This SRM function may be bypassed in the high IRM ranges (ranges 8. 9. and 10) when the IRN upscale rod block is operable.
7. Not required to be operable when performing low power physics tests at atmospheric pressure during or after refueling at power levels not to exceed 5 MWt.
8. This IRM function occurs when the reactor mode switch is in the Refuel or Startup/ Hot Standby position.
9. This trip is bypassed when the SRM is fully inserted.
10. The trip level setting shall be a maxifmam of 108% for core flow equal to 98 x 106 lb/hr and greater. l l

3.2/4.2 14a 0154H/00682 .nrinent No.

  • QUAD-C878E5 OPR-30  !

t sidered inoperable, fully provide reasonable assurance i i

inserted into the core. and that proper control rod drive electrically disarmed. performance is being maintained. The results of

5. If the overall average of the measurements performed on the 20% insertion scram time data control rod drives shall be f generated to date in the current submitted in the annual cycle exceeds 0.o8 seconds, the l operating report to the NRC.

MCPR operating limit must be modified as required by 5. The cycle cumulative mean scram Specification 3.5.K. time for 201 insertton will be determined immediately following the testing required in 6 Spectfications 4.3.C.1 and

+

4.3.C.2 and the MCPR operating limit adjusted it necessary, as required by Specification 3.5.K.

i O. Control Rod Accumulators D. Control Rod Accumulators i At all reactor operating pressures, a Once a shift. check the status of the rod accumulator may be inoperable pre:sure and level alarms for each provided that no other control rod in accumulator, the nine-rod square array around that rod has:

1. An inoperable accumulator.
2. A directional control valve electrically disarmed while in 2 ,

nonfully inserted position, or

3. A scram insertion greater than  ;

maximum permissible insertion i time.

If a control rod with an inoperable

accumulator is inserted full-in and its directiceal control valves are electrically disarmed, it shall not be considered to have an inoperable accumulator, and the rod block asso-  !

Ciated with that inoperable accuno-1ator may be bypassed.

E. Reactivity Anomalies E. Reactivity Anomalies i

i The reactivity equivalent of the dif- During the startup test program and ference between the actual critical startups following refueling outages, rod configuration and the espected the critical rod configurations will t configuration during power operation be compared to the expected configur-shall not exceed it A k. If this ations at selected operating condi-limit is esceeded, the reactor shall tions. These canpartsons will be be shutdown until the cause has been used as base data for reactivity ,

determined end corrective actions monitoring during subsequent power have been taken. In accordance with operation throughout the fuel cycle.

Specification 6.6. the NRC shall be At spectfic power operating condi-notified of this reportable occur. tions the critical rod configuration rence within 24 hcurs. will be compared to the configuration 1 expected based upon appropriately corrected past data. This comparison will be made at least every equiva-J 1ent full power month.

j F. Economic Gent'ation Control System F. Economic Generation Control System Operation of the unit with the eco- Prior to entering EGC and once per I nomic generation control system with shift while operating in EGC. the ECC

! automatic flow control shall be per- operatinc parameters will be reviewed

' missible only in the range of 651 to for acceptability.  ;

i 100% of rated core flow, with reactor power above 20%. i 09258 3.3/4.3 5 Amenoment No.  ;

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QUAD-C2T!ss DPR-30 provided that during such 7 days operable immediately. The RCIC all active components of the system shall be demonstrated to automatic pressure relief be operable daily thereafter.

subsystems, the core spray Daily demonstration of the subsystems. LPCI mode of the RMR automatic pressure relief system, and the RCIC system are subsystem operability is not operable. required provided that two feedwater pumps are operating at levels above 300 NWeg and one

3. If the requirements of feedwater pump is operating as specification 3.5.C cannot be normally required with one met, an orderly shutdown shall additional feedwater pump -

be initiated, and the reactor operable at power levels less pressure shall be reduced to 90 than 300 MWe.

psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

L. Automatic pressure Rollef Subsystems D. Automatic pressure Relief Subsystams Surveillence of the automatic

1. The automatie pressure relief pressure reliaf subsystem shall be subsystem shall be operable performed as follows:

whenever the reactor pressure is greater than 90 psig, irradiated 1. The following surveillance shall fuel is lu the reactor vessel be carried out on a six-month and prior to reactor startup surveillance intervals from a cold condition,

s. With the reactor at pressure each relief valve shall be
2. From and af ter the date that two l manually opened. Relief of the five relief valves of the valve opening shall be automatic pressure relief verified by a compensating subsystem are made or found to l turbine bypass valve or be inoperable when the reactor control valve closure.

is pressurized above 90 pois l with irradiated fuel in the reactor vessel. reactor 2. A logic system functional test operation is permissible only shall be performed each durir.g the succeeding 7 days refueling outage, unless repairs are made and provided that durir.g such time the HPCI subsystem is operable. 3. A simulated automatic initiation which opens all pilot valves shall be performed each

3. If the requirements of Speelfi- refueling outage, cation 3.5.D cannot be met. an orderly shutdown shall be initi-ated and the reactor pressure 4. When it is determined that two shall be reduced to 90 pois relief valves of the automatic within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. pressure relief subsystem are inoperable. the NPCI shall be demonstrated to be operable immediately.

l 09258 3.5/4.5-5 Amendment do.

QUAO-CITIES OPR-30 within the prescribed limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. the reactor shall be brought to the cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. $urveillance and corresponding action shall continue untti reactor operation is within the prescribed limits. Maximum allowable LHG4 is 13.4 kw/f t, for fuel types P8x8R and SP3x8R. For fuel types GE8x8E and GE8u8E8 the maximum allowable LHGR is 14.4 kw/ft.

K. Minimum Critical Power Ratio (MCPR) K. Minimum Critical Power Ratio (MCPR)

During steady-state operation at The MCPR shall be determined daily during rated core flow. MCPR shall be steady-state power operation above 25% of greater than or equal to: rated therrul power.

1.30 for TAVE i 0.68 sec l 1.35 for T Ayg 10.86 sec O 278 TAyg + 1.111 ,

for 0.68 sec i T Ayg i .86 sec d

where TAyg a mean 20s teram insertion the for all surve111ance data from 4 .

specification 4.3.C which has been generated in the

, current cycle.

For core flows other than rated, these nominal values of MCPR shall be increased by a f actor of kg where kr is as shown in Figure 3.5.2. If any time during operation it is determined by normal surveillince that the limiting value for MCPR is being exceeded, action shall be inittated within 15 mtnotes to restore operation to within the prescribed limits. If the steady-state MCPR is not returned to within the prescribed limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the reactor shall be brought to the cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveillance and corresponding action shall continue until reactor operation is within the prescribed Itmits.

09258 3.5/4.5-10 AmendFent No.

QUAD-C! TIES OPR-30 Based on the fact that when on3 lo:p of the contatbeent Cooling mode of the RHR system becomes inoperable, cnly onn system remains, which is tested daily, a 7-day repair period was specified.

C. High-Pressure Coolant Injection The high-pressure coolant injection subsystem is provided to adequately cool the core for all pipe breaks smaller than those for which the LPCI rode of the RHR system or Core Spray subsystems can protect the core.

The HPCI meets this requirement without the use of offsite electrical power. For the pipe breaks f or which the HPCI is intended to functiofe.

the core never uncovers and is continuously cooled thus no cladding damage occurs (reference 5AR 5ection 6.2.5.3). The repair times for the limiting conditions of operation were set considering the use of the HPCI as part of the isolation Cooling system.

D. Automatic Pressure Relief The relief valves of the automatic pressure relief subsystems are a backup to the HPCI subsystem. Tney enable the core spray subsystem and l LPCI mode of the RHR system to provide protection against the small pipe break in the event o' HPCI failure by depressurizing the reactor vessel rapidly enough to actuate the core spray subsystems and LPCI l mode of the RHR system. Tre core spray subsystem and/or the LPCI mode of the RHR system provide suf ficient flow of coolant to limit fuel cladding temperatures to less than 2200'F. to assure that core geometry remains intact, to limit the core wide clad retal-water reaction to less than it, and to limit the calculated local metal-water reaction to less than 171.

Analyses have shown that only four of the five valves in the autonutic depressurization system are required to operate. Loss of one of the relief valves does not significantly affect the pressure relieving capability. therefore continued optration is acceptable. Loss of two relief valves signif tcantly reduces the pressure relief capability of the A05: thus, a 7 day repair period is socctfied with the HPCI available, and a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> repatr period with the HPCI unavailable.

L. RCIC The RCIC systen is provided to supply continuous nakeup water to the reactor core when the reactor is isolate 1 from the turbine and when the fetewater system ts not available. Under these conditions the pumping capactty of the RCIC system is sufficient to maintain the water level above the core without any other water system in operation. If the water level in the reactor vessel decreases to the RCIC initiation level, the system automatically starts. The system may also be manually initiated at any time.

The HPCI system provides an alternate method of supplying makeup water to the reactor should the normal feedwater become unavailaole.

Theref ore, the specification calls for an operability check of the HPCI system should the RCIC system be found to be inoperable.

F. Emergtney Cooling Availability The purpose of Specification 3.5.F is to assure a minimum of core cooling equipacnt ts available at all times. If. for esample, one core spray were out of servtce and the diesel which powered the opposite core spray were out of service, only two RHR pumps would be available.

Likewise if two RHR pumps were out of service and two RHR service water pumps on the opposite side were also out of service no containment cooling would be available. It is during the refueling outages that major ruintenancs is performed and during such tine that all low-pressure core cooling systems may be out of service. This specification provides that should this occur, no work will be perf ormed on the primary system which could lead to draining the vessel. This work would include work on certain control rod drive components and rectrculation systems. Thus, the specification precludes the events which could requira core cooling. Specif ic ation 3.9 must also be consulted to determine other requirerents for the diesel generators.

Quad-Cities Units 1 and 2 share certain process systems such as the nukeup demineralizers and the ratwaste system and also some saf ety systems such as the standby gas treatment system, batteries, and 0921B 3.5/4.5-12 Amensment 4o.

I 1

_ __ .__ ________________________________J

s QUAD-CIT!!$ t

  • OPR-30 [

shown on Figure 3.5-1 45 limits because conformance calculations have not been performed to justify operation at LHGR's in excess of those shown.

J. Local LHGR This spectftcation assures that the maximum lineer heat generation rate  ;

in any rod is less than the destgn linear heat-generation rate even if '

fuel pellet dentification is postulated. The power spike penalty is

  • i discussad in Reference 2 and assumes a linearly increasing variation in a axial gaps between core bottom and top and assures with 95% confidence that no more than one fuel rod exceeds the design LHGR due to power spiking. No penalty is required en $pecification 3.5.L because it has

.been accounted for in the reload transient analyses by increasing the  !

calculated peak LNGR by 2.2%.

K. Minimum Critical Power Ratio (MCPR)

The steady state values for MCPR specified in this specification were

i. selected to provide margin to acconenodate transients and uncertainties in monitoring the core operating state as well as uncertainties in the critical power correlation itself. These values also assure that operation will be such that the initial conditton assumed for the LOCA analysis plus two percent for uncertatnty is satisfied. For any of the special set of transtents or disturbances caused by single operator 4 error or single equipment malfunction, it is required that design analyses intttaltted at this steady-state operating limit yield a MCPR of not less than that specified in Spectfteation 1.1.A at any time during the transient, assuming instrument trip settings given in tpecification 2.1. Fcr analysts of the thermal consequences of these

, transients, the value of MCPR stated in this specification for the Ilmiting condition of operation bounds the initial value of MCPR l i

assumed to exist prior to the initiation of the transients. This initial condition, which is used in the transient analyses, will ,

preclude violation of the fuel r.1 adding integrity safety Itmit.

Assumptions and methods used in calculating the required steady state r MCPR limit for each reload cycle are documented in References 2 and 4 l The results apply with increased conservatism while operating with 1 MCPR's greater than specified.

1' '

The most limiting transients with respect to MCPR are generally:

a) Rod withdrawal error j b) Load rejection or turbine trip without bypass e

c) Loss of feedwater heater Several factors influence which of these transients results in the largest reduction in critical power ratio such as the specific fuel loading, esposure, and feel type. The current cycle's reload licensing

, analyses specifies the limiting transients for a given exposure l

increment for each fuel type. The values specified as the Limiting

- Condition of Operation are conservatively chosen to bound the most

! restrictive over the entire cycle for each fuel type.

The need te adjust the MCPR operating limit as a function of scram time arises from the stattstical approach used in the implementation of the 00VN computer code for analyzing rapid pressuritation events. Generic statistics) analyses were performed for plant groupings of similar design which considered the stattstical vartation in several parameters (initial power level CR0 scram insertion time, and model uncertainty). These analyses (which are described further tn aeference

4) produced generic Stattstical Adjustment Factors which have been applied to plant and cycle specific 00VN results to yield ops. Sting limits which provide a 95% probability with 951 confidence that the limiting pressurization event util not cause MCPR to fall below the fuel cladding integrity safety limit.

09258 3.5/4.5-14 Amenoment No.

l

QUAD-CIT 8ES OPR-30 As a result of this 95/95 approach, the average 20% insertion scram time must be monitored to assure compliance with the assumed statistical distribution. If the mean value on a cycle cumulative (running average) basis were to exceed a 5% significance level compared to the distribution assumed in the 00YN statistical analyses, the MCPR limit must be increased linearly (as a function of the mean 20% scram time) to a more conservative value which reflects an NRC determined uncertainty penalty of 4.4%. This penalty is applied to the plant specific ODYN results (i.e. without statistical adjustment) for the limiting single failure pressurization event occurring at the limiting point in the cycle, it is not applied in full until the mean of all current cycle 20% scram times reaches the 0.90 secs value of Specification 3.3.C.I. In practice, however, the requirements of 3.3.C.1 would most likely be reached (i.e. Individual data set average

> .90 secs) and the required actions taken (3.3.C.2) well before the running average exceeds 0.90 secs.

The 5% significance level is defined in Reference 4 as:

n TB = p + 1.65 (Nj/ { Nj)l/2 0 11 where:

p - Mean value for statistical scram time distribution to 20%

inserted a = standard deviation of above distribution N) = number of rods tested at BOC (all operable rods) n

{N) = total number of operable rods tested in the current cycle 11 The value for TB used in Specification 3.5.k is 0.68 secs which is l conservative for the following reasons:

a) For simplicity in formulating and implementing the 1.C0, a n

conservative value for { Nj of 708 (i.e. 4x177) was used.

11 This represents one full core data set at BOC plus 6 half core data sets. At the maximum frequency allowed by Specification 4.3.C.2 (16 week intervals) this is equivalent to 24 operating months. Thai h , a cycle length was assumed which is longer than any past or contenolated refueling interval and the number of rods tested was maximizeo in order to simplify and conservatively reduce the criteria hr the scram time at which MCPR penal 12ation is necessary, b) The values of p and a were also chosen conservatively based on the dropout of the position 39 RPIS switch, since pos. 38.4 is the precise point at which 20% insertion is reached. As a result Specification 3.5.k initiates the linear MCPR penalty at a slightly lower value T ave. This also produces the full 4.4%

penalty at 0.86 secs whicn would occur sooner than the required value of 0.90 secs.

0926B 3.5/4.5-14a Amendment No.

QUAD-CITIES DPR-30 For core flow rates less than rated, the steady state MCPR is increased by the formula given in the specification. This ensures that the MCPR will be maintained greater than that specified in Specification 1.1.A even in the event that the motor-generator set speed controller causes the scoop tube positioner for the fluid coupler to move to the maximum speed position.

References

1. "Quad Cities Nuclear Power Station Units 1 & 2 SAFER /CESfR-LOCA Loss of Coolant Accident Analysis" NEDC-31345P.*
2. "Generic Reload Fuel Application " NEDE-24011-P-A**
3. I. M. Jacobs and P. W. Marriott, GE Topical Report APED 5736, "Guidelines for Determining Safe Test Intervals and Repair Times for Engineered Safeguards " April, 1969.
4. "Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors" General Electric Co. Licensing Topical Report NEDO 24154 Vols. I and II and NEDE-24154 Vol. III as supplemented by letter dated September 5, 1980 from R.H. Buchholz (GE) to P. S. Check (NRC).

l

  • Approved revision at time of plant operation.
    • Approved revision number at time reload fuel analyses are performed.

l 4409K 3.5/4.5-14b Amendment No.

I a

MAPLHGR VS. Average P10ncr Exposure fuel Type BP80RB282

.:r:-

. .c Ill ::D ZZ 12.0 .-- -n- 'E----._-_<_a

3. .

3_;__.

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k sc

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1 8.5 0 10.000 20.000 J0.000 40.000 50.000 Average Planer Exposure (ufd/St)

WAPLHCR VS. Aver 0ge P10nct Exposure fuel Type BP80RB283H 12.5 ..__. _

b 12.c j  %

u,,',

' s

/ ',,

11.5 '-~h

.(_

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x11.c I A

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= N 3.5 k- ,

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8.5 - - -

0 10.000 20.000 30.000 40.000 50.000 i AveragePlanorEsposare(Wud/St) i i

I Figure 3.5-1 '

I Sheet 1 of 5

uAPlilGR VS. Avgrage Plonor Exposure {

Fuel Type P80G8263L i t

L 12.5 i - .._ .. . .._

l t

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WAPLHGR VS. Average Plonor Exposure j

fuel Type P8008298 l

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Figure 3.5-1 l Sheet 2 of 5 '

. _ _ . . . . . . _ _ _ _ - .- . = _ - . - .

WAPLHCR VS. Average Plonor Exposure fuel Type BP80R8299 12.3 q, 1

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AveragePlanarExposure(Wed/St) l 1

UAPLHCR VS. Avercge Planar Exposure i

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I j

WAPlilCR VS. Average Planor Exposure l

i i . se fuel Type 80316A l

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WAPLHCR VS. Average Planer Exposure fuel Type 8D3000 ,

l i

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Figure 3.5-1 Sheet 4 of 5

WAPLHGR VS. Average Planor Exposure l  ;

fuel Types P80R8265H, BP80RB265H I i

f .

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i i

l i

l Figure 3.5-1 Sheet 5 of 5 .

h I

f

i I

l i 1.4 FOR FLONS GREATER THAN 100%fK = 1.0 12

\ AUTOMATIC FLOW CONTR0t N NN

= e5 g,g A

{ w w MANUAL FLOW CONTROL SCOOP-TUBE SET-POINT CALIBRATION I = 102.5% u 1.0 107.0%

112.0%

117.01 I I 40 50 50 70 SO 90 SOS 30 CORE FLOW 1

," 2 D $

a-"

i

~

l t

QUAD CITIES DPR-30 1-i

3. Prior to Single Loop Operation for more than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, the following l l

restrictions are required:

a. The MCPR Safety Limit shall be increased by 0.01. (T.S. 1.1A); ,
b. The MCPR Operating Limit shall be increased by 0.01 (T.S.

3.5.K);

c. The flow biased APRM Scram and I Rod Block Setpoints shall be reduced by 3.5% to read as ,

follows:

1 T.S. 2.1.A.1, S S .58WD + 58.5 [

T.S. 2.1.A.1;*

S 1 (.58WD + 58.5) FRP/MFLPD T.S. 2.1.B S 1 58WD + 46.5 r

T.S. 2.1.B;* i r

S 1 (.58WD + 46.5) FRP/MFLPD T.S. 3.2.C (Table 3.2-3);*

l APRM upscale s (.58WD + 46.5) i f

FRP/MFLPD i

i

d. The flow biased RBM Rod Block l [

. setpoints shall be reduced by 4.0% to read as follows:

T.S. 3.2.C (Table 3.2-3); RBM  !

Upscale s .65WD + 39 l

e. The suction valve in the idle l loop shall be closed and electrically isolated except when the idle loop is being [

prepared for return to service.  :

f I

0925B 3.6/4.6-5a Amendment No.

1 I i

i

i .

l QUAD-ClT!ES OPR-30

. The licensee's analyses indicate that above 80% power the loop select logic l could lot be expected to function at a speed differential of 15%. Below 80%

power, the loop select logic would not be expected to function at a speed differential of 20%. This specification provides a margin of 5% in pump speed

, differential before a problem could arise. If the reactor is operating on one pump, the loop select logic trips that pump before making the loop selection.

Analyses have been performed which support indefinite single loop operation provided the appropriate restrictions are implemented within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The MCPR l Safety Limit has been increased by 0.01 to account for core flow and TIP reading oncertainties which are used in the statistical analysis of the safety ilmit.

! The HCPR Operating Limit has also been increased by 0.01 to maintain the same i

margt.1 to the safety limit as during Dual Loop operation.

The flow biased scram and rod block setpoints are reduced to account for uncertainties associated with backflow through the idle jet pumps when the operating recirculation pump is above 20 - 40% of rated speed. This assures that the flow biased trips and blocks occur at conservative neutron flux levels for a given core flow.

The closure of the suction valve in the idle loop prevents the loss of LPCI flow through the idle recirculation pump into the downcomer.

a 1

l i

i i

I 1

0926B 3,6/4.6-13a Amendment No.

e e CENERAL ELECTRIC C0MPAN7

&EFIDAVIT I, Rudolph Villa, being duly sworn, depose and state as follows:

1. I am Manager, Consulting Services, General Electric Company, and t have been delegated the function of reviewing the information I

described in paragraph 2 which is sought to be withheld and have been authorized to apply for its withholding.

2. The information sought to be withheld is contained in "Quad Cities Nuclear Power Station Units 16 2, SAFER /CESTR-LOCA Loss of Coolant Accident Analysis," NEDC 31345P, Revision 1, dated January 1988.
3. In designating material as proprietary, General Electric utilizes the definition of proprietary information and trade secrets set forth in the American Law Institute's Restatement of Torts, Section 757. This definition provides:

"A trade secret may consist of any formula, pattern, device or compilation of information which is used in one's business and which gives him an opportunity to obtain an advantage over competitors who do not know or use it.... A substantial element of recrecy must exist, so that, except by the use of improper means, there would be difficulty in acquiring information. . . . Some factors to be considered in determining whether given information is one's trade secret are: (1) the extent to which the information is known outside of his business; (2) the extent to which it is known by employees and others involved in his business; (3) the extent of measures taken by him to guard the secrecy of the information; (4) the value of the informatior. to him and to his competitors; (5) the amount of effort or money expanded by him in developing the information; (6) the ease or difficulty with the which the information could be properly acquired or duplicated by others."

4 Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method or appara;us where prevention of its use by General Electric's competitors without license from Genera). Electric constitutes a coepetitise economic advantage over other companies;

- _ _ _ _ _ _ _ _ - - . _ _ _ _ _ . _ _ . _ _ _ _ _. _ _ . - _ _ _ _ _ _ _ _ _ _ _ _ . - _ _ -a

4

b. Information consisting of supportin5 data and analyses, including test data, relative to a process, method or apparatus, the application of which provide a competitive economic advantage, e.g., by optimization or improved marketability;
c. Information which if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manvfacture, shipment, installation, assurance of quality or licensing of a similar product;
d. Information which reveals cost or price information, production capacities, budget levels or commercial strategies of General Electric, its customers or suppliers; '
e. Information which reveals aspects of past, present or future General Electric customer funded development. plans and programs of potential commercial valuo to General Electric:
f. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection;
g. Information which General Electric must treat as proprietary according to agreements with other parties.
5. Initial approval of proprietary treatment of a document is typically made by the Subsection manager of the originating component, the person who is most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within the Company is limited on a "need to know" basis and such documents are clearly identified as proprietary.
6. The procedure for approval of external release of such a document typically requires review by the Subsection Manager, Project Manager, Principal Scientist or other equivalent authority, by the Subsection Manager of the cognizant Marketing function (or delegate) and by the Legal Operation for technical content, co=petitive effect and determination of the accuracy of the proprietary designation in accordance with the standards enumerated above. Disclosures outside General Electric are generally limited to regulatory bodies, customers and potential customers ar.d their agents, suppliers and licensees then only with appropriate protection by applicable regulatory provisions or proprietary agreements.
7. The document mentioned in paragraph 2 above has been evaluated in "

accordance with the above criteria and procedures and has been found to contain information which is proprietary and which is custoinarily held in confidence by General Electric.

l o . . ,

I

8. The information to the best of my knowledS e and belief has consistently been held in confidence by the General Electric Company, no public disclosure has been made, and it is not <

available in public sources. All disclosures to third parties have been made pursuant to regulatory provisions of proprietary agreements which provide for maintenance of the information in confidence.

9. Public disclosure of the inforrnation sought to be withheld is .

likely to cause substantial harm to the competitive position of the General Elect.ric Company and deprive or reduce the availability of ,

profit making opportunities because it would provide other parties, ,

including competitors, with valuable information on input parameters and anlysis results of the SAFER /CESTR LOCA analysis ,

methodology, as well as details of current fuel designs that are not available to other parties.

STATE OF CALIFORNIA COUNTY OF SANTA CLARA

)) ss: ,

l Rudolph Villa, being duly sworn, deposes and says:

That he has read the foregoing affidavit and the matters stated therein are true and correct to the best of his knowledge, information, and [

belief.

Executed at San Jose, California, this /2 day of [ ! h *4 , 19 k

., / ~

/ k '

Rudolph Villa General Electric Company l Subscribed and sworn before me this y ohhth/U_198[.

g i

" " ] py )

] 7[ "

PAut A F HUME

  1. $ OfL .hMAL S TARY PUBLIC, STATE @ CALIFOR:11A  !

f43 TAU Pitu: caVFCRMa ( j Win OAV Om (

Me em. sarim m 13, !!:;

- l l

1 l

i i

f i

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