ML19263C402

Forwards Addl Info Re Proposed Amend to License DPR-29 to Support Reload 4.Info Involves Correction to Water Level, Refs to Reflect Longer Fuel Length of 8x8R Retrofit Fuel & Operation in Coastdown Mode

ML19263C402
Person / Time
Site:	Quad Cities
Issue date:	02/14/1979
From:	Reed C COMMONWEALTH EDISON CO.
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 7902220065
Download: ML19263C402 (10)

Text

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Commonwealth Edison 07e First National Plaza._ Chicago._ lltfnois Address Reply to: Post Othce Box 767 Chicago, llhncis 60690 February 14, 1979 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

Quad-Cities Station Unit 1 Proposed Amendment to License and Appendix A, Technical Specifications, for Facility Operating License DPR-29 NRC Docket No. 50-254 Reference (a): C. Reed letter to Director of NRR dated November 20, 1978

Dear Sir:

Reference (a) transmitted a proposed amendment to the License and Appendix A, Technical Specifications, to Facility Operating License DPR-29 to support core reload No. 4 at Quad-Cities Station Unit 1.

Subsequent to that transmittal and in telephone conversations with the NRC Staff, additional changes to support core reload No. 4 have been identified. The changes include corrections to water level references which reflect the longer fuel length of the 8x8R retrofit fuel and revision of Paragraph 3.C on Page 4 of the License to limit operation in the coastdown mode to 40% power. These changes are identified in Enclosure I and have received on-site and off-site review and approval.

Since these changes are additions to or revisions of a previous submittal currently under review, Commonwealth Edison has determined that an additional fee remittance in accordance with 10 CFR 120 is not required.

Please address any questio!.- concerning this matter to this office.

\

d0 7902220065

Commonwealth Edison NRC Cocket No. 50-254 Director of Nuclear Reactor Regulation February 14, 1979 Page 2 Three (3) signed originals and thirty-seven (37) copies of this transmittal are provided for your use.

Very truly yours, A W f ~ "A' Cordcll Reed Assistant Vice-President enclosure SUBSCRIBED and SWORN to before me this \ '\ , day of \,' -

, 1979.

v s ,

x.

Notary Public

i e

ENCLOSURE I QUAD-CITIES UNIT 1 NRC DOCKET NO. 50-254

t .

l DPR-29

/

-" 3. This Itcense shall be deemed to contain and is subject to the

• b

• conditions specified in the following Coestission regulations 1

in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, 50, Sectier, 4^.f.1 of Port 40, Sections 50.54 and 50.59 of Part and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, resuistions and isand ordersto subjcet of the Comission now or hereaf ter in ef fect; the additional conditions specified or incorporated below:

A. Maxt'rms Power level I

Cocesonvralth Edison is authortred to eperate Qusd-Cit te=

Unit NO. I at power levels not in excess of 2511 etegawatts (the rmat ) .

B Technical Specifications The Technical Specifications contained in Accendices A and B, as revised throagh Amendment No.149, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

i C. Restrictions

' Reactor power level shall be limited to maintain

" pressure margin to the safety valve set points during the worst case pressurization transient.

The magnitude of the power limitation, if any, and.the point in the cycle at which it shall be applied is specified in the Reload No. 4 licensing submittal for Quad Cities Unit No. I (NEDO 24145)

Plant operation shall be limited to the operating plan described therein. Subsequent operation in the coastdown mode is pennitted to 40% power.

D. Equalizer Valys Restriction The valves in the equalizer piping between the recirculation loops shall be closed at all times during reactor cperatien.

! 4. This license is effective as of the date of issuance, and shall '

expire at midnight, February 15, 2007.

l l

clocures: Appnndices A and B-- FOR THE ATCHIC ENERCY CCNMISSION '

L Technical Specificat. ions

[

aIeofIssuance: December 14, 1972 8 //,g.,r.<[

A. Clan. bus eo. Deputy nirec tor for Reac tor Prnlect s Directorate et Licensing 9

-- . - . . *s ALPMA e0PP

• 0000018 - S AL f a27 QUAD-CITIES DPR-29 l

7 D. ' Reactor Water Imel(Shutdawn Condition) curse in Figure 2.1-2, at which point Whenever the reactor is ir. tl e shutdown condi- the actual peaking factor value shall be tion with irradiated fuel in the reactor vessel, the water level shall not be less than that corre-i sponding to 12 inches above the to of the LTPF = 3.06 (7 x 7 fuel assemblies) active fuel

• vhen it is sea ed in 3.0.) (8 x 3 fuel assemblies) .

Oe com.

2. APRM Flux Scram Trip Setting (Re-

fueling or Startup and Hot Standby Mode)

When the reactor mode switch is in the Refuel or Startro Hot Standby posi-tion, the APRM seam shall be set at less than or equal to 15% of rated neutron flux.

3. IRM Flux Scram Trip Setting

' The IRM flux tcram setting shall be set at less than or equal to 120/125 of full scale.

i

4. When the reactor mode switch is in the

, , startup or run position, the reactor shall J not be operated in the natural circula.

tion flow mode.

B. APRM Rod Block Setting Thc AP1M rod block setting shall be as shown in Figure 2.1-1 and shall be:

j S s (.65W + 43)(LTPFiTPF)

The definitions used above for the APRM scram trip apply.

C. Reactor low water level scraim setting shall'be 2 144 inches above the top of the active fuel

• at '

normal operating condit.ans.

D. Reactor low water level ECCS initiation shall be 84 inches ( + 4 mches/-0 inch) above the top i of the active fuel

• at normal operating conditions.

E. Turbine stop valve scram shall be s 10% valve closure from full open.

F. Turbine control velse fast closure scram shall initiate upon actuation of the fast closure sole-noid valves which trip the turbine control valves.

( . -

G Main steamline iso!ation valve closure scram

• Top of active fuel is shall be s 10% valve closure from full open.

defined to be 360 inches above vessel zero (see H. Main steamline low-pressure initiation of main Lases 3.2), steamline isolation valve closure shall be 2 850 psig.

_ . - . . . . - . . . - . . . . . . . . ..- .. n-. . . . -. .a k

QUAD-CITIES DPR-29

! C. settings which maintain equivalent safety margins when the total peak factor (TPF) exceeds the LTPF.

Specification 3.5J established the LHGR maximum which cannot be exceeded under sleady power operation.

B. Core Thermal Power Limit (Reactor Pressure <800 psia)

At pressures below 800 psia, the core elevation pressure drop (0 power, O flow)is greater than 4.56 psi.

At low powers and flows this pressure differentialis maintained in the bypass region of the core. Since the pressure drop in the bypass region is essentially all elevation head. the core pressure drop at low powers and flows will always be greater than 4.56 psi. Analyses show that with a flow of 28 x 1021b/hr I

bundle flow, 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 105 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 h1Wt. 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 800 psia is conservative.

C. Power Transient During transient operation the heat flux (thermal power-to-water) would lag behind the neutron flux due i

to the inherent heat transfer time constant of the fuel, which is 8 to 9 seconds. Also, the limiting safety l

system scram settings are at values which will not allow the reactor to be operated above the safety limit 8

during normal operation or during other plant oper,ating situations which have been analyzed in detail.

t In addition, control rod scrams are such that for normal operating transients, the neutron flux transient 6 ,- is terrninated before a significant increase in surface heat flux occurs. Scram times of each control rod are l checked each refueling outage, and at least every 32 weeks,50% are checked to assure adequate inseration times. Exceeding a neutron flux scram setting and a failure of the control rods to reduce flux to less than the scram setting within 1.5 seconds does not necessarily imply that fuel is damaged; however, for this specification, a safety limit violation will be assumed any time a neutron flux scram setting is exceeded

for longer than 1.5 seconds.

i If the scram occurs such that the neutron flux dwell time above the limiting safety system setting is less than 1.7 seconds, the safety limit will not be exceeded for normal turbine or generator trips, which are the most severe normal operating transients expected. These analyses show that even if the bypass system i

fails to operate, the design limit of h1CPR = 1.06 is not exceed:d. Thus, use of a 1.5-second limit provides additional margin.

The co puter provided has a sequence annunciation program which willindicate the sequence in which scratas occur, such as neutron flux, pressure, etc. This program also indicates when the scram setpoint is cleared. This will provide information on how long a scram condition exists and thus provide some measure of the energy added during a transient. Thus, computer information normally will be available for analyzing scrams; however, if the computer information should not be available for any scram analysis, Specification 1.1.C.2 will be relied on to determine if a safety limit has been violated.

During periods when the reactor is shut down, consideration must also be given to water level requirements due to the effect of decay heat. If reactor water level should drop below the top of the active i

fuel during this time, the ability to cool the core is reduced. This reduction in core. cooling capability could lead to elevated cladding temperatures and cladding perforation.The core will be cooled sufficiently to prevent cladding melting should the water level be reduced to two-thirds the core height. Establish. .

l ment of the safety limit at 12 inches above the top of the fuel *provides adequate margin.This level will rz be continuously monitored whenever the recirculation pumps are not operating.

• Top of active fuel is ciefined to be 360 inches above

, vessel zero -(see Bases 3.2). '

1.1/2.1-5

~

i QUAD-CITIES DPR-29 f 1ABLE 3.14 /

, NOTES FOR TABLES 3.11. 3.12. AND 3.14

, 1. There sha!! be tno operable trp systems or one operable and one troped system for each functon.

2. If the first column cannot be met for one of the tre systems, that try system shall be troped. If the first column cannot be met for both try systems, the appropriate actons listed below shall be taken:

A. Initiar inserton e of operable rods and complete oserton of all operable rods withe 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

I j B. Reduce power level to IRM range and place mode switch a the Startup/ Hot Standby positon within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

i C. Reduce turbine load and close main steamline isolaton valves within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

3. An APRM willbe corisdered moperable rf there are fewer than 2 LPRM oputs per levelor there are less than 50% of the ncrmalcomplement of LPRM's to an APRM.

4. Permissbie to bypass, with control rod block for reactor protecton system reset n refuel and shutdown positions cl the reactor mode switch.

5. Not requred to be operable when prrnary contaoment otegrity is not requaed.

6. The desti permits closure of any one loe without a scram berg citiated.

, 7. Automatically bypassed when reactor pressure is <1060 psig l 8. The + Brinch tro poot is the water level as measured by the instrumentation outsde the shroud. The l f 3 water level osde the shroud will decrease as power rs cereased to 100% n comparison to the level outsde the shrcud to a maxrnum cf 7 oches. This is due to the pressure drop across the steam dryer.

i Therefore, at 100% power, an indicaten of + 8 och water level will actuafly oe + 1 inch oside the shroud.1 inch on the water level i.nstrumentation is 2 504" above vessel zero. (see Bases 3.2) i 9. Permisstle to bypass when fast stage turbine pressure is less than that which corresponds to 45%

rated steam flow. (<400 psi)

10. Tres upon actuaton cf the fast <lo:ure soleno.d which trips the turboe control vahres.

, 11. The APRM downscale tre functon is automaticany bypassed when the IRM nstrumentaton is cperable and not high.

I 12. Channel shared by the reactor protecton and contanment isolaton system.

e. .

3.1/4.1-11

QUAD-CITIES DPR-29

, 3.2 LIMmNG CONDITIONS FOR OPERATION BASES

' In addition to reactor protection instrumentation which initiates a reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious consequences. This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the emergency core cooling system, control rod block, and standby gas treatment systems. The objectives of th: specifications are (l) to assure the effectiveness of the protective instrumentation when required by preserving its capability to tolerate a single failure of any component of such systems even during periods when portions of such systems are out of service for maintenance, and (2) to prescribe the trip settings required to assure adequate perremance.

When necessary one channel may be made inoperable for briefintervals to conduct required functional tests and calibrations. Some of the settings on the instrumentation that initiates or controls core and containment cooling have tolerances explicitly stated where the high ar.d low values are both critical and may have a substantial effect on safety. It should be noted that the setpoints of other instrumentation, when only the high or low end of the setting has a direct bearing on safety, are chosen at a icvel away from the normal operating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situations.

Isolation valves are installed in those lines that penetrate the primary containment and must be isolated during a loss-of-coolant accident so that the radiation dose limits are not exceeded during an accident condition. Actuation of these valves is initiated by the protective instrumentation which senses the conditions for which isolation is required (this instrumentation is shown in Table 3.2-1 ). Such instrumentation rnust be availabic wht.ncver primary containment integrity is required. The objective is to isolate the primary containment so that the guidelines of 10 CFR 100 are not exceeded during an accident.

The. instrumentation which initiates primary system isolation is connected in a dual bus arrangement. Thus the discussion given in the bases for Specification 3.1 is applicable here.

The low-reactor water level instrumentation is set to trip at /8 inches on the level instrument (top of active fuel is defined to be 3(,0 inches above vessel rero) af ter allowing for the full power pressure drop across the ste.u dryer the low level trip is at 504 inches above vessel zero, or 144 inches above top of active fuel. Retrofit 8x0 fuel has an active fuel length 1.24 inches longer than earlier fuel designs, however, present trip setpoints were used in the LOCA analysis (tEDO 241461. This trip initiates closure of Group 2 and 3 prinary containment isolation valves but does not trip the recirculation pumps (reference SAR, Section 7.7.2). For a trip setting of 504 inches above vessel zero and a 60-second valve closure time, the valves will be closed before perforation of the cladding occurs even for the maximum breaks the setting is, therefore, adequate.

The low-low reactor level instrumentation is set te, trip when reactor water level is 444 inches above vessel sero (with top of active fuel defined an 360 inches '

above vessel sero, -59' is 84 inches above the top of active fuel) .

This trip initiates closure of Group i primary containment isolation valves (reference SAR Section 7.7.2.2) and also activates the ECC subsystems. starts the emergency diesel generator, and trips the recirculatia pumps. This trip seuing level was chosen to be high enough to prevent spurious operation but low enough to initiate ECCS operation and primary system isolation so that no melting of the fuel cladding will occur and so that postaccident cooling can be accompished and the guidelines of 10 CFR

- ' 100 will not be exceeded. For the complete circumferential break of a 28-inch recirculation line and with the trip setting given above. ECCS initsation and primary system isolation are initiated and in time to meet the ebove criteria (reference SAR Sections 6.2.7.1 and 14.2.4.2 ). The instrumentation aho covers the full spectrum of breds and meets the above criteria (reference SAR Section 6.2.7.1).

%c high.drywell pressure instrumentation is a backup to the water levelinstrumentation and. in addition to initiating ECCS. it causes isolation of Group 2 isolation valves. For the breals discussed above. this instrumenta-tion willinitiate ECCS operation at about the same time as the low low water kvel instrumentation; thus the results given above are applicable here also. Group 2 isolation valves include the drywcli vent, purge, and sump isolation valves.High-drWell pressure activates or.!y these valves because high drywc!! pressure could occur as the result

• of non-safety related causes such as not purging the drywell air during surtup. Total system isolation is not desirable for these conditions, and only the valves in Group 2 are required to close. The low imv water level (C

1. ~ -

instrumentation initiates protection for the full spectrum orloss-of-coolant accidents and causes a trip of Group 1 primary system isolation valves.

. 12/4.2-5

i QUAD-CITIES DPR-29 TABLE 3.21 INSTRUMENTATION THAT INITIATES PRIMARY CONTAINMENT ISOLATION FUNCTIONS MWmum Number of Operable er Tripped lastrument

! Cnannel:08 lastruments Trip Level Setting Action'D 4 Reactor low water * >l44hches above top of A l active fuel

• 4 Reactol low low water 284oches above top of A active fuel

• 4 Hth drywell pressure'5) s2 psigm A 16 Hgh flow man steamfoc* s120% of rated steam flow B 16 Hgh temperature can s200'F B steamloe tunnel 4 Hgh radation man 57 x normal rated power 3 steamloe tunnefU bacitground i 4 Low man steam pressure) 2850 pst B g 4 Hgh flow RCIC steam!me s300% of rated steam fiow C

, a 16 RCIC ! doe area high s200*F 0 temperature 4 Hth flow HPCI steamine 5300% of rated steam flow D

! 16 HPCI area hgh temperature s200*F D Notes L Whenever preery contamment utegnty rs requeed. there shall be two operable or tripped systems for each function, except for icw pressure man steamine which only need be available a the Run positen i 2. Action:If the irst column cannot be met for one of the trip systems. that tr y system shall be tripped 11 the irst column cannot be met for both trip systems, the approprete actons listed telow shall be taken:

A. wete an ederly sautdown and have the reactor a cold shutdoan condition m 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> B. btste an orderly load reducten and have reactor m Hot Standby nithe 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

C. Close nolaten vaives a Rcic system.

D . Close isolaten valves a HPCI subsystem

3. Need r9t be operable when primary contamment stegrity a not required

4. The notaten trip stnal es bypassed when the mode switch is e Refuel or Startup/ Hot Shutdown a

i n3 instrumentatea a!so rsolates the control room ventdaton system t f That sgnal also automatcally closes the mechanical vacuum pump discharge Ime esolaton valves

• Top of active fuel is defined as 360" above vessel zero for all water levels used in the LOCA analysis (see Bases 3.2) .

{ 3.2/4.2-I I

QUAD-CITIES t

DPR-29 TABLE 3.22 7

. INSTRilMENTATION THAT INITIATES OR CONTROLS THE CCEE AND CONTAINMENT COOLING SYSTEMS tilsieurn Number of Operable er Tripped lastrarnent ChannetfD Trip Function Trip Lcrel Setting Remarks

, 4 Reactor bw bw 284 behes ( + 4 inches /-0 och) 1. In conjunction with bw-reactor pressure

, water level above top of actue fuel

• bitiates core spray and LPCI.

2. In conjunction with high-drywell pressure 120 second trne delay and low-pressure core cochng interbck mitiates auto bbwdown.

3. Initiates HPCI and RCIC.

.l . Initiates startog of desel generators.

4* Hgh-drywell s2psg 1. Initiates core spray, LPCI, HPCI. and pressure (23,(3) SGTS.

2. In conjunction with low bw water level, 120-second tune delay, and bw-pressure core coolog hterlock initiates auto bbwdown.

1 3. Initiates startog of diesel generators.

4. Initiates notation of control room ventilation.

2 Reactor low 300 psigsps350 psig 1. Permissive for opening core spray and LPCI pressure admission vanes.

2. In conjunction with bw bw reactor water level bitiates core spray and LPCI.

Contabment spray Prevents hadvertent operation of contantnent

, bterlock spray during accident conditons.

208 2/3 core height 22/3 core height 4 03 contanment 0.5 pds;st.5 psig high pressure 2 Timer auto s120 seconds in conjunction with low bw reactor water blowdown level, high-drywell pressure, and Icw-pressure core cooling interlock inittates auto blow-down.

4 Low-pressure core 75 psigsps100 psg Defers APR actuation pendog confimation of j coolog pump dis- low. pressure cure coolog system operaton.

i charge pressure i

2 Undervottage on N/A 1. Inities starting of desel generators.

emergency buses 2. Permissive for startog ECCS pumps.

3. Removes nonessential Icads from buses.

}

• Top of active fuel is defined as 360" above vessel zero for all water levels used in the LOCA analysis.

3.2/4.2-12