ML20150B250
| ML20150B250 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 06/23/1988 |
| From: | Norrholm L Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20150B252 | List: |
| References | |
| NUDOCS 8807110417 | |
| Download: ML20150B250 (12) | |
Text
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UNITED STATES
~i NUCLEAR REGULATORY COf.. MISSION d
WASPINGTON, D. C. 20555 g
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June 23, 1988 COMNONWEALTH EDISON COMPANY AND IOWA-ILLIN0IS GAS AND ELECTRIC COMPANY DOCKET N0. 50-254 QUAD CITIES NUCLEAR POWER STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.109 License No. DPR-29 1.
The Nuclear Regulatory Cornission (the Commission) his found that:
A.
The application for amendment by Commonwealth Edison Company (thelicensee)datedJanuary 29, 1988, 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, l
the provisions of the Act, and the rules and regulations of the l
Commission; C.
There is reasonable assurance (1) that the activities authorized by thi.c amendment can be conducted without endangering the health and se"ety of the public, and (ii) that such activities will be conduc.9d in
.w aliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and si.fety of the public; and l
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements l
have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license l
amendment, and paragraph 3.8 of Facility Operating License No. DPR-29 is hereby amended to read as follows:
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B.
Technical Specifications The Technical Specifications contained in Appendix A and B, as revised through Amendment No.109, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specificatiors.
3,.
This license amendment is effective as of the date of its issuance.
FORTHENKLE GULAT9RY COMMIS' ION
/r>7" M
Leif J.
orrh Im, Acting Director Pro.iect directorate III-2 Division of Reactor Projects - III, IV, V and Special Projects
Attachment:
Changes to the Technical Specifications Date of Issuance:
June 23,1988
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ATTACHMENT TO LICENSE AMENDMFNT NO. 109 FACILITY OPERATING-LICENSE NO. DPR-29 DOCKET NO. 50-254 Revise the Appandix A Technical Specifications by removing the pages identified below and inserting the attached pages.
The revised pages are identified by the' captioned amendment number and contain marginal linas indicating the area of change.
REMOVE INSERT-1.1/2.1-2a 1.1/2.1-2a 3.2/4.2-5a 3.2/4.2-Sa 3.2/4.2-12 3.2/4.2-12
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p-t 0U40 CIVIES OPR-29 The definit.ons used above for the APRM scram trip apply. In tre event of operation with a manicum fraction limiting cower density (NFLPD) greater than the fraction of rated power (FRP), the setting shall be modified as f ollows:
5 1 -(.58WD + 50)
FR' MFLPD The definitions used above for the APRM scram trip apply.
The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than 1.0 in which case the actual operating value will be used.
This may also be performed by increasing tne APRM gain by the inverse ratio, MFLPD/FRP, wttch l
accompitshes the same ce;ree wf protection as reducing the trip setting by FRP/MFLPD.
C.
Reactor low water level scram setting shall be 144 inches above the top of the active fue1* at nore41 cperating conditions.
D.
Reactor low water leve*. ECC5 initiation shall be 1 84 inches above l
the top of the active fuel
- at normal operating conditions.
E.
Turbine stop valve scram shall ce i 10% valve closure from full open.
F.
Turbine control valve fast closure scram shall initiate upon actuation of the f ast closure solencic valves which trip the turbine centrol valves.
G.
Main steamline isolatten valve closure scram shall be i 10% valve closure from full open.
H.
Main steamline low-pressare int-tiation of main steam 1tne isolatien valve closure shall be 1 825 0519
' Top of active fuel 15 defired to be 360 inches above vessel zere (See Bases 3.2).
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Anenament No.
109 1.1/2.1-:a
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3.2 LIMITING CONCITION FOR OPERATION SASE$
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 This set of terminates operator errors before they result in serious consequences.
specifications provides the limiting conditions of operation for the primary system isolation function. initiation of the emergency core cooling system, control rod block anti standby gas treatmsnt systems. The objectives of the specifications are (1) to assure the effectiveness of the protective instrumentation when raquired 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 When necessary, prescribe the trip settings required to assure adequate performance.
one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations, Some of the settings on the instrumentation that initiates or an,qoftrols core and containment cooling have tolerances explicitly stated where the high-c 1p, values are both critical and may have a substantial effect un safety. It should be nr.ed that the setpoints of other instrumentation, where only the high Or LOW end Of the setting has a direct bearing on safety, are chosen at a legal away from the normal operating range to prevent inadvertent actuation of the safety system involved and exposure to asnorssi situations.
Isolation valves are installed in those lines that penetrate the primary containment and must be isolates turing a loss-of-coolant accident su that the rsdiation dose limits are not exceeded du:ing an accident condition. Actuation of these valves is initiated by the protective irstrumentation which serves the condition for which isolation is required (this instrumentation is shown in Table 3.2.1).
Such instrumentttton must be available whenever primary containment integrity is required. The objective is to isolate the primary containment so that the guidlines of 10 CFR 100 are not exceeded during an accident.
The instrumentation which initiates primary system isolation is connecter! in a r!Jat bus Thus the discussion given in the basis for specificatton 3.1 is applicable arrangement.
here.
The low reactor levet instrumentation is set to trio at > 8 inches on the level instrument (top of active fuel 15 defined to the 360 inches above vessel zero) and after allowing for the full power pressure drop across the steam dryer the low-level trip is at 504 inches above vessel zero. Or 144 inches above the top of active fuel. Retrofit 8x8 fuel has an active fuel length 1.24 inches longer than earlier fuel designs.
This trip initiates However, present trip setpoints were used in the LOCA analyses *.
closure of Group 2 and 3 primary containment isolation valves but does not trip the recirculation perips (ref erent.e 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 maximm break: the setting is therefore adequate.
The low low reactor level instrumentation is set to trip when reactor water level is l
1444 inches above vessel zero (with top of active fuel defined as 360 inches above vessel zero. -59 inches is 44 inches above the top of active fuel). This trip intttates closure of Group 1 primary containment isolation valves (reference SAR Section 7.7.2.2) and also activates the ICC subsystems, starts the emergency diesel generator, and trips This trip setting level was chosen to be low enough to prevent the recirculation pumps.
spurious operation but high enough to initiate ICCS operation and primary system tsolattun so that no melting of the tuel cladding will occur and so that postaccidentFor cooling can be accomplished and the guidelines of 10 CFR 100 will not be exceeded.
the !omplete circumferential break of a 28-inch recirculation line and w'.th the trip setting given above, ECCS initiation and primary isolation are initiated and in time to The instrumentation also covers the full spectrum of breaks meet the above criteria.
and meets the above criteria.
The high-drywell pressure instrumentation is a backup to the water level instrumentation in addition to initiating ECCS. it causes isolation of Group 2 isolation valves.
- and, For the breaks discussed above. this instrumentation will initiate ECCS operation at about the same time as the low low water level instrumentation: thus the results given above are applicable here also. Group 2 1 solation valves include the drywell vent, Hign-drywell pressure activates only these valves purge and sump isolation valves.
because hiph drywell pressure could c cur as the result of non-saf ety-related causes such as not purging the drpell air during start-up. Total system isolation is not desiracle for these condittons. and only the valves in Group 2 are required to close.-
The low low water level instrwentation initiates protection for the full spectrum of loss-of-coolant accidents and causes a trip of Group 1 primary system isolation valves.
Loss of coolant accident analysis for Dresden Units 2 & 3 and Quad Citics Units 1 &
2, NEDO-24146A. Aort 1, 1979 0723B/0336Z 3.2/4.2-54 Amendment No.109
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QUAO-C!?IES OPR-29 TABLE 3.2-2 IN57RUMENTAT!0N THAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS Minimum Number of Operable or Tripped Instrume chanaals Trio Function Trie tevel settina R ema r k s 4
Reactor low low 184 inches above 1.
In conjunction with low-water level top of active fue1*
reactor pressure initiates core spray and LPCI.
2.
In conjunction with high-drywell pressure 120-second time delay and low-pressure core cooling interlock initi-ates auto blowdown.
3.
Initiates st arting of diesel generators.
4(8}
High-drhl)1 (3}
12.5 psig 1.
Initiates core spray, LPCI.
HPCI. and 5BGT5.
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2.
In conjunction wit?. 10w low water level. 120-second time I
delay. and low-pressure core cooling interlock initiates auto blowdown.
3.
Inittates starting of diesel generators.
4.
Initiates isolation of control room ventilation.
2 Reactor low 300 rstgip1350 psig 1.
Permissive for opening core pressure spray and LPCI admission valves.
2.
In conjunction with low low reactor water level initiates core spray and LPCI.
Containment spray Prevents inadvertent operation of interlock containment spray during accident conditions.
2 3}
2/3 core height 12/3 core height 4 33 containment 0.5 psigipil.5 pstg high pressure 2
Timer auto 1120 seconds In conjunction with low low blowdown reactor water level, high-drywell pressure, and low-pressure core cooling interlock initiates auto blowdown.
4 tow pressure 100 psigip1150 psig Defers APR actuation pending con-core cooling pump firmation of low-pressure core discharge pressure cooling system operation.
2/ BUS (5)
Undervoltage on 3045 1 5% volts 1.
Initiates starting of diesel emergency buses generators.
2.
Permissive fLr starting ECCS pumps.
3.
Removes nonessential loads from buses.
4 Bypasses degraded voltage t ime r.
- Top of active fuel is defined at 360* above vessel zero for all water levels used in the LOCA analysis 07238 3.2/4.2-12 Amenorent No.109
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'o UNITED STATES
~,,
NUCLEAR REGULATORY COMMISSION o
- y WASHINGTON, D. C. 20555
%.....,o COMMONWEALTH EDIS0N COMPANY AND IOWA-ILLIN0IS GAS AND ELECTRIC COMPANY DOCKET NO. 50-265 QUAD CITIES NUCLEAR POWER STATION, UNIT 2 AMENDME.'IT TO FACILITY OPERATING LICENSE Amendment No. 105 License No. DPR-30 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Commonwealth Edison Company (the licensee) dated January 29, 1988, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I;
(
B.
The facility will operate in conformity with the application.
l the provisions of the Act, and the rules and regulations of the l
Comission:
1 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 Ui) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the common defensc 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 Comission's regulations and all applicable requirements have been satisfied.
l 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPR-30 is hereby amended to read as follows:
r B.
Technical Specifications The Technical Specifications contained in Appendix A and B, as revised through Amendment No.105, are hereby incorporated in this license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the dete of its issuance.
FORTHEJd_ EAR EGULATORY COMMISSION d
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Leif d'. Pt rr I ctin Director Project Directorate III-2 Division of Reactor Projects - III, IV, V and Special Projects
Attachment:
Changes to the Technical Specifications Date of Issuance:
June 23, 1988 l
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ATTACHMENT TO LICENSE AMENDMENT N0.105 FACILITY OPERATING LICENSE NO. DPR-30 DOCKET NO. 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 and contain marginal lines indicating the area of change.
REMOVE INSERT 1.1/2.1-2a 1.1/2.1-2a 3.2/4.2-Sa 3.2/4.2-51 3.2/4.2-12 3.2/4.2-12 i
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Quad CETIES OPR.30 The definitions used above for the APRM scram trip apply. In the event of operation with a maximum fraction Itatting power density (MFLPD) greater than the fraction of rated power (FRP), the setting shall be modified as follows:
51 (.58WD + 50)
FRP MFLPD The definitions used above for the APRM scram trip apply.
The ratio of FAP to MFLP0 shall be set equal to 1.0 unless the actual operating value is less than 1.0, in which case the actual oper4 ting value will be used.
This adjustment may also be performed by increasing the APRM gain by the inverse ratio. MFLPD/FRP. which accomplishes the same cegree cf protection as reducing the trip setting by FRP/MFLPD.
C.
Reactor low water level scram setting shall be 144 inches above the top of the active fuela at normal operating conditions.
D.
Reactor low water level ECCS initiation shall be 1 84 inches above l
the top of the active fuel
- at norpul operating conditions.
E.
Turbine stop valve scram shall be i 10% valve closure from full open.
I F.
Turbine control valve fast closure scram shall initiate upon actuation of the fast closure solenoid valves which trip the turbine control valves.
G.
Main steam 1tne isolation valve closure scram shall be i 10% valve i
closure from full open.
l t.
Main steam 11ne low-pressure int-ttation of main steam 1tne isolation l
valve closure shal' be 1 825 psig.
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' Top of active fuel is defined to be 360 inches above vessel zero (See i
Ps'es 3.2).
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1 07298 1.1/2.1 24 Amenoment No.
105 l
DPR-30 3.2 LIMITING CONDITION FOR OPERATION BASES o
In addition to reactor prGtection instrumentation which initiates a reactor scram, protective Instrumentation has been provided which lettiates 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 Ilmiting conditions of operation for the primary system is)1ation function, initiation of the emergency core cooling system, control rod block and standby gas treatment systems. The objectives of the specifications are (1) to assure the effectiveness of the protective instrumentation when required by preserving its capability to tolerate a single f ailure 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 performance. When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations. Sone of the settings on the instrumentatton that initiates or controls chr$ 0W d$ltaib liat and cooling tave tolerances explicitly stated where the high an ue re Doth critical and may have a substantial effect on safety. It should be noted that the setpoints of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the norpul operating range to prevent inadvertent actuation of the safety system involved ar.d exposure to abnorwul situations.
Isolation valves are installed in those Itnes that penetrate the pr' mary sontainment and must be isolated during a loss-of-coolant accident so that the radiation dose Itmits are not exceeded during an accident condition. Actuation of these valves is initiated by the protective inst umentation which serves the condition for which isolation is required (this instrumentation is shown in Table 3.2.12 Such instrumentation must be available whenever primary l
containment integrity is required. The objective is to isolate the primary containnent 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 basis for Specification 3.1 is applicable here.
The low reactor level instrumentation is set to trip at > 8 inches on the level instrument (top of active fuel is defined to be 360 inches above vessel zero) and after allowing for the full power pressure drop across the steam dryer the low-level trip is at 504 inches above vessel zero, or 144 inches above the top of active fuel, Retroftt 8x8 fuel has an active fuel length 1.24 inches longer l
than earlier fuel designs. However, present trip setpoints were used in the l
LOCA analyses (NE00-24146A. April 1979). This trip initiates closure of Group i
2 and 3 primary containment 1 solation valves but does not trip the I
rectrculation pumps (referense SAR Section 7.7.2).
For a trip setting of 504 l
inches above vessel zero (144 inches above top of active fuel) and a 60-second valve closure time, the valves will be closed before perforation of the cladding occurs even for the maximum break: the setting ts trerefore adequate.
The low low reactor level instrumentation is set to trip when reactor water level is 1 444 inches above vessel zero (w?th top of active fuel defined as 360 l
inches above vessel zero. -59 inches is 84 inches above the top of active fuel). This trip inttlates closure of Group 1 primary containawnt isolation valves (referenca 5AR Section 7.7.2.2) and also acttvates the ECC subsystems, starts the emergency diesel generator, and trips the recirculation pumps. This trip setting level was chosen to be low enough to prevent spurious operation but high enough to initiate ECCS operation and primary systest isolation so that no melting of the fuel cladding will occur and so that postaccident cooling can be accompitshed and the guidelines of 10 CFR 100 will not be exceeded. Fce tre complete circumferent tal break of a 28-tnch rectreulation line and with the trtp setting given above. ECCS initiation and primary isolation are initiated and in time to meet the above criteria. The instrumentation also covers the full spectrum of breaks and reets the above Criteria.
The high-drywell pressure instrumentation is a backup to the water level instrumentation and, in addition to initiating ECCS. it causes isolation of Grouc 2 1 solation valves. For the breaks discussed above, this instrumentation will initiate ECCS operation at about the same time as the low low water leve1
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instrumentation; thus the results given above are applicable here also. Group 2 isolation valves include the drywell vent, purge and sump 1 solation valves.
High-drywell pressure activates only these valves because high drywell pressure could occur as the result of non-safety-related causes such as not purging the drywell air during start-up. Total system isolation is not desirable for these conditions, and only the valves in Group 2 are required to close. The low low water level instrumentation initiates protection for the full spectrum of loss-of-coo
- ant acetdents and causes a trip of Croup 1 primary system isolation valves.
072'd/02252 3.2/4.2-Sa Amendment No. 105
QUAD CITIES DPR-30 TABLE 3.2-2 INSTRUMENTATION THAT INITIATts OR CONTROL 5 THE CORE AND CONTAINMENT COOLING SYSTEMS Minimum Number of Operable or Tripped Instrumep),
Channa11 w Trie Function Trio Level intting Recurks 4
Reactor low low 184 inches above 1.
In conjunction with low-l water level tcp of active fue1*
reactor pressure initiates core spray and LPCI.
2.
In conjunction with high-drywell pressure 120-second time delay and low-pressure ccre cooling interlock initt-ates auto blowdown.
3.
4 Initiates starting of diesel generators.
12.5 pstg 1.
Initiates core soray, LPCI.
aI83 High-drhl11 [33 HPCI and SSGTS.
pressurel J.
2.
In conjunction with low low water level.120-stcond time delay, and low-pressure core cooling interlock initiates auto blowdown.
3.
Initiates starting of diesel generators.
4.
Initiates isolation of control room ventilation.
2 Reactor low 300 psigip1350 psig 1.
Permissive for opening core spray and LPCI admission pressure valves.
2.
In conjuncticn with low low-reactor water lavel tnitiates core spray ano LPCI.
Containment spray Prevents inadvertent operation of interlock containment spray curing accident c or.di t ion s.
2 2/3 core height 12/3 core height 4
containment 0.5 psigsp11.5 psig high pressure 2
Timer auto 1120 secor.ds In conjunction with low low reactor water level. high-drywell blowdown pressure, and low-pressure core cooling interlock initiates auto blow-down.
4 Low-pressure 100 psigipil50 psig Defers APR actuation pending con-Core cooling pump firmation of low-pressure core discharge pressure cooling system operation.
2/ BUS (5)
Undervoittge on 3045 i Si volts 1.
Initiates starting of diesel emergency buses generators.
2.
Permissive for starting ECCS pumps.
3 Removes nonessential loads fran buses.
4.
Bypasses degraded voltage timer.
- Top of active fuel is defined at 360* above vessel zero for all water levels used in the LOCA analysis.
07248 3.2/4.2-12 Amendment No.105
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