ML20136B500
| ML20136B500 | |
| Person / Time | |
|---|---|
| Site: | Dresden, Quad Cities  |
| Issue date: | 08/30/1979 |
| From: | Reed C COMMONWEALTH EDISON CO. |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20136B506 | List: |
| References | |
| NUDOCS 7909060379 | |
| Download: ML20136B500 (55) | |
Text
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Commonwd Edison I
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, 7 }- Address Reply to: Post Office Box 767 J
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- One First National Plaza. Chicago, Illinois
- 4. ' 7:g Chicago, Illinois 60690 August 30, 1979 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Subject:
Quad-Cities Station Unit 2 Proposed Amendment to Pacility Operating License DPR-30 and Appendix A, Technical Specifications, to Implement 10 CFR 50.59 Reload Licensing NRC Docket No. 50-265 Reference (a):
J.
F. Quirk letter to Olan D.
- Parr,
" General Electric Licensing, Topical Report NEDE-240ll-P-A, " Generic Reload Fuel Application", Appendix D, Second Submittal," dated February 28, 1979
Dear Sir:
Pursuant to lO.CFR 50.59, Commonwealth Edison proposes to amend the License and Appendix A, Technical Specifications, to Facility Operating License DPR-30 to support the review of future reloads for Quad-Cities Unit 2 by Commonwealth Edison in accordance with the provisions of 10 CFR 50.59.
These changes are identified in Enclosure I and are based in part on plant analyses summarized in Enclosure II, " Loss-of-Coolant Accident Analysis Report for Dresden Units 2, 3 and Quad-Cities Units 1 and 2 Nuclear Power Stations," NEDO 24146A, 79NED273, April, 1979.
The significant changes to the Technical Specifications include:
1.
The MCPR Safety Limit has been generalized to include core configurations with and without retrofit 8x8 fuel.
As a result of the flatter local power (and CPR) distribution of the 8x8R design, which adversely affects the transition boiling probability distribution, the MCPR Safety Limit for core loading patterns
'containing 8x8R fuel has been increased from 1.06 to 1.07.
The value of 1.06 is still applicable for \\
core loading patterns containing no 8x8R fuel.
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NRC Do at No. 50-265 j
g C:mm:nws:lth Edis:n Director of Nuclear Reactor Regulation August 30, 1979 Page 2 2.
The Rod Drop Accident figure of merit has been changed from 1.3% AK to 280 cal /gm.
This is a consequence of red worths slightly less than 1.3% AK resulting in a peak fuel clad enthalpy of 280 cal /gm for some local peaking conditions.
Similar revisions have previously been reviewed and approved for Dresden Unit 2 Reload 4 Cycle 7.
3.
New MAPLHGR curves have been included which reflect the improved flooding characteristics of the retrofit 8x8 design during a postulated LOCA.
A reanalysis of the ECCS performance (Enclosure II) for the limiting break size LOCA (DBA) has resulted in relaxed MAPLHGR limits.
These relaxed limits are primarily due to the effects of drilled lower tie plates in the retrofit 8x8 fuel.
In addition to providing MAPLHGR limits for fuel to be used in Quad-Cities 2 Reload 4 Cycle 5, Enclosure II also contains MAPLHGR limits for various other retrofit fuel designs of different enrichments.
These have been included in the proposed 10 CFR 50.59 Technical Specification chang-es to avoid later additions should these fuel designs become part of future core loading strategies.
Enclosure II assumes only 156 bundles with drilled lower tie plates.
Quad-Cities 2 Cycle 5 will utilize 180 retro-fit 8x8 fuel bundles with future cycles utilizing a larger number.
Therefore, the MAPLHGR limits presented in Enclosure II should remain conservative for future cycles, maintaining the 10 CFR 50.59 approach to reload licensing.
4.
The LHGR spiking penalty for postulated fuel densification for 8x8 fuel types (standard and retrofit) has been included in the transient analyses by lowering the safety limit by an equivalent amount.
Therefore, no LHGR spiking penalty is required to be applied (for any 8x8 fuel type) to the design limit LHGR of 13.4 kw/ft.
The spiking penalty for all 7x7 fuel types is still required but will not change.
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e Comm:nwrith Edisin NRC Du :ket No. 50-265 Director of Nuclear Reactor Regulation August 30, 1979 Page 3 5.
Reference to the limiting total peaking factor (LTPF) has been eliminated and replaced with reference to the maximum fraction of limiting power density (MFLPD) for adjustment of the APRM flux scram and rod block trip settings.
This proposed change eliminates the need for different LTPFs for different fuel types as a consequence of different bundle heat transfer areas, while providing the same degree of protection (with respect to reduction of the trip settings).
Therefore, the trip settings for the APRM scram and the rod blocks will be reduced by FRP/blFLPD whenever the MFLPD exceeds the FRP (Fraction of Rated Power).
This ratio is equivalent to the ratio of LTPF to TPF (Total Peaking Factor),
i.e.
CTP LTPF =
DLHGR/RCTP/K DLHGR *
=
TPF LHGR/CTP/K LHGR RCTP 1
=
MFLPD Where:
DLHGR = Design Limit LHGR RCTP
= Rated Core Thermal Power K = Constant Also, the proposed Technical Specifications provide for increasing the APRM gains in lieu of an actual reduction in APRM trip set points whenever the MFLPD exceeds the FRP.
This method establishes an initial APRM signal closer to the flow-biased setpoints, and thus has the same effect as reducing the actual scram and rod block setpoints.
For consistency with the LHGR surveillance requirement and the Standardized Technical Specifications, the proposed changes also require that the FRP/MFLPD multiplier be applicable only above 25% rated thermal power.
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NRC Dq et No. 50-265 Camm:nw: lth Edisin Director of Nuclear Reactor Regulation August 30, 1979 Page 4 6.
The Target-Rock safety / relief valve setpoint is being changed to 1115 psig for both the relief and safety functions of the valve.
The relief function setpoint of the valve was changed to preclude multiple relief valve discharges.
The proposed Technical Specification setpoint of 1115 psig for both the safety and relief function will be utilized in future licensing analyses and, therefore, the setpoint change is necessary to ensure the applicability of the licensing analyses.
7.
Because 8x8R fuel has an active fuel length 1.24" longer than previously utilized bundle designs, ambiguity exists in referencing the top of the active fuel for the reactor low water level scram setting and ECCS initiation set-point.
Consequently, the top of active fuel is being defined as 360 5/16" above the reactor pressure vessel zero point.
This corresponds to the actual top of active fuel for pre-retrofit fuel designs.
The reactor low water level scram and ECCS initiation setpoints occur at 143 7/8" and 83 7/8" based on the above definition for the top of active fuel.
These values have been rounded to 144" and 84" in the proposed specifications for realistic surveillance precision and for consistency with i
previous specifications.
8.
A 25 psi margin is currently maintained between the cal-culated peak steam line pressure of the most limiting abnormal operational transient and the lowest spring safety valve setpoint.
The purpose of this margin is to preclude actuation of the spring safety valves during pressurization events with bypnss valve failure.
Such events are very low probability ( <1/ plant lifetime) and are an operational concern rather than a safety consider-ation.
This argument has been presented by General Electric in Reference (a) and is currently being reviewed.
The proposed amendment, therefore, eliminates the scram reactivity license restriction (and consequently eliminates the 25 psi margin to the lowest spring safety valve set-point).
NRC concurrence would not preclude Edison from administratively incorporating pressure margin if future
I E
Docket No. 50-265 e
Ccmm nwsalth Edizen Director of Nuclear Reactor Regulation August 30, 1979 Page 5 assessments indicate it is prudent to do so to avoid a forced outage in the unlikely event of bypass failure and significant safety-valve discharge to the drywell.
9.
The " Generic Reload Fuel Application," NEDE-240ll-P-A, documents the fuel designs, methods, evaluations, codes, generic criteria, test results, and assumptions which are utilized in the reload licensing analyses.
It has been referenced whenever appropriate in the proposed changes (e.g. to replace outdated references and discussions).
10.
The limiting safety system bases for the APRM rod block i
trip setting has been clarified to indicate that the setting provides protection against grossly exceeding the MCPR Fuel Cladding Integrity Safety Limit.
Adequate local protection from a rod withdrawal error is provided t
by the RBM system alone.
Previous wording did not clearly distinguish the functions of the two systems.
These proposed changes have received on-site and off-site review and approval.
Pursuant to 10 CFR 170, Commonwealth Edison has determined that the proposed amendment is Class III.
As such, we have enclosed a fee remittance in the amount of $4,000.00.
For purposes of your schedule, we are requesting that your review be completed no later than October 15, 1979 in order to allow sufficient time to resolve any staff concerns.
Also note that should this amendment be approved, the effective date must be later than the end of the current Cycle 4 (currently scheduled for November 4, 1979) since the LOCA analysis (Enclosure II) is based on the use of retrofit fuel designs with drilled lower tie plates which are not utilized in the current cycle.
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l Crmm:nwulth Edison NRC I~cket No. 50-265 Director of Nuclear Reactor Regulation August 30, 1979 Page 6 f
Three (3) signed originals and thirty-seven (37) copies of-this transmittal are provided for your use.
Very truly yours, C'
.hE+-r-Cordell Reed Assistant Vice-President enclosures 1
SUBSCRIBED and SWO to before e this
, day of oun.k.
, 1979.
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JCLOOERAA Nothry Public O"
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ATTACHMENT I Proposed License Amendment and Technical Specification changes for Quad-Cities Unit 2.
DPR-30 License Page 4 Appendix A Page #
1.0-2 1.0-4 1.1/2.1-1 1.1/2.1-2 1.1/2.1-3 1.1/2.1-4 1.1/2.1-5 1.1/2.1-6 1.1/2.1-7 1.1/2.1-8 1.1/2.1-9 1.1/2.1-10 1.1/2.1-11 Figure 2.1-2 (Deleted) 1.2/2.2-1 1.2/2.2-2 1.2/2.2-3 3 1/h.1-1 3.1/h.1-3 3.1/4.1-5 3.1/h.1-7 3.2/h.2-5 3.2/h.2-6 3.2/h.2-7 3.2/h.2-8 3 2/h.2-11 3.2/h.2-12 3.2/h.2-14 3 2/h.2-15 3 3/h.3-3 3 3/h.3-4 3 3/h.3-8 3.3/4.3-9 3.3/h.3-10 3.3/h.3-11 3.4/h.h-3 3 5/h.5-9 3.5/h.5-10 3.5/h.5-13 3.5/h.5-lh 3.5/h.5-lha Figure 3 5-1 3 5/h.5-17 3.6/h.6-4
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.Q D PR-3(I (4
3.
This tis ense shall be deced to c ont eln an.I to euhjece to the condit tene eye. t tled in the following Crm iselon re gw aftone in 10 Ct a Chap' e r I. Part 20, 5eselon MJ 16 et rest
't).
,ection 40.48..f Part 40. Sections 50.44 enJ 10 39 of'e'ert SO.
and section 10 12 of Part 10. te owbject to all atettieNie provietane ist ehe Ace one t o the tulee, regulets.sie omt ordes s o f t h e s'evrcil e s s on now or hereaftee la ef fect; e...
to.s..hject so the additional condit ione spes tiled or te oryntese. he l.a.
A Preste Power Level Conwoonwea te n Edteon t o authorised to operate Qued titles Unit No. 2 et power levele not in excese of 7381 segewet t e (thermat) 8-TechnicalJoectftcattonj The technli.nl spectfications contained in Appendices A and 8. as revised through Amenenent No. O,are hereby f acorporated in the license. The I tcensee shall operats the facility in accortfante with t 'ie Technical specifications.
3.C Restrf.ctines i
Operation in the coastdown mode is i
O permitted to 40% power.
j Am. 84 3 3/8/18 g
I D. gu a lhe r va l ve He s t r_3 c t,1 on The valve s in the equ liser piping between the Am. 12 4/21/75 circulat ton loops shall be closed at re-reactor operation.
a l l t tre s d u s t r.9 E. Security Plan The licensee shall rnaintain in effect and fully implement all provisions of the Comission-approved physical security plan,
^*- 8 including amencents and changes made pursuant to the authority 2/23'79 of 10 CFR 50.54(p). The approved security plan consists of documents withheld from public disclosure pursuant to 10 CFR 2.790(d), referred to as Quad Cities Nuclear Power Station Units Nos. I and 2 Physical Security Plan dated as follows:
i Plan - November 18, 1977 Revision 1 - May 19.1978 Revision 2 - May 27. 1978 s
Revision 3 - July 28,1978 i
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QUAD-CITIES DPR-30 Limiting Conditions for Operation (l.CO) - The limiting conditions for operation specify the minimum H.
acceptable levels of system performance necessary to assure safe startup and operation of the fac
[
When these conditions are met, the plant can be operated safely and abnormal situations can be safely 7
controlled.
i Limiting Safety System Setting (LSSS) -The limiting safety system settings are settings on instrumenta.
1.
tion which initiate the automatic protective action at a level such that the safety limits will not be exceeded.. The region between the safety limit and these settings represents margin, with normal operation lying below these settings. The margin has been established so that with proper opera I
the instrumentation, the safety limits will never be exceeded.
Logie System Functional Test - A logic system functional test means a test of all relays and contacts of K.
a logic circuit from sensor to activated device to ensure all components are operabic per design intent.
Where possible, action will go to completion; i.e., pumps will be started and valves opened.
Modes of Operation - A reactor mode switch selects the proper interlocking for the operating or L
shutdown condition of the plant. Following are the modes and interlocks provided:
- 1. Shutdown -In this position, a reactor scram is initiated. power to the control rod drive > is removed.
and the reactor protection trip systems have been deenergized for 10 seconds prior to petmimve iN l
manual reset.
Refuel - In this position, interlocks are established so that one control rod only may be withdrawn 2.
when flux amplifiers are set at the proper sensitivity level and the refueling crane is not over the reactor. Also, the trips from the turbine control valves, turbine stop valves, main steam isolation valves, and condenser vacuum are bypassed. If the refueling crane is over the reactor, all rods must be fully inserted and none can be withdrawn.
- 3. Startup/ Hot Standby -In this position,the reactor protection scram aip3. initiated by condenser low vacuum and main steamline isolation va!ve closure, are bypassed, the low prewure main steamfun isolation valve closure trip is bypassed, and the reactor protection system is energized. with IR M and APRM neutron monitoring system trips and control rod withdrawal interlocks in service.
Run - In this position the reactor system pressure is at or above 850 psig,and the reactor protection 4.
system is energized with APRM protection and RMB interlocks in service (excluding the 15r* high c
flux scram).
M. Operable - A system or component shall he considered operable when it is capable of performing its intended function i.. its required rnanner.
N. Operating Operating means that a system or component is performing its intended functions in its required manner.
Operating Cycle -Interval between the end of one refueling outage for a particular unit and the end of O.
the next subsequent refueling outage for the same unit.
P.
Primary Containment Integrity - Primary containment integrity means that the drywell and pressure suppression chamber are intact and all of the following conditions are satisfied:
1.
All manual containment isolation valves on lines connecting to the reactor coolant system or containment which are not required to be open during accident conditions are closed.
1.0-2
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QUAD-CITIES DPR-30 h
Y.
Shutdown - The reactor is in a shutdown condition when the reactor mode switch is in the Shutdown position and no core alterations are being performed.
1.
110: Shutdown means conditions as above, with reactor coolant temperature greater than 212' F.
- 2. Cold Shutdown means conditions as above, with reactor coolant temperature equal to or few than 212 F.
Z.
Simulated Automatic Actuation - Simulated automatic actuation means applying a simulated signal to the sensor to actuate the circuit in question.
BB. Transition Boiling - Transition boiling means the boiling regime between nucleate and film boiling.
Transition boiling is the regime in which both nucleate and film boiling occur intermittently, with neither type being completely stable.
CC. Critical Power Ratio (CPR) - The critical power ratio is the ratio of that assembly power which causes some point in the assembly to experience transition boiling to the assembly power at the reactor condition ofinterest as calculated by application of the GEXL correlation (reference NEDO 109581 DD. Minimum Critical Power Ratio (MCPR) The minimum incore critical puwer ratio corresponding to the most limiting fuel assembly in the core.
O-EE. Suneillance Interval - Each surveillance requirement shall be performed within the specified surveil-lance interval with:
A maximum allowable extension not to exceed 25% of the surveillance interval.
a.
b.
A total maximum combined interval time for any 3 consecutive surveillance intervals not to exceed 3.25 times the specified surveillance interval.
FF.
Fraction of Limiting Power Density (FLPD) - The fraction of limitirg power density is the ratio of the linear heat generation rate (LHGR) existing at a given location to the design LHGR for that bundle type.
GG..
Maximum Fraction of Limiting Power Density (MFLPD) - The maximum fraction of limiting power density is the highest value existing in the-core of the fraction of limiting power density (FLPD).
HH.
Fraction of Rated Power (FRP) - The fraction of rated power is the ratio of core themal power to rated thermal power of 2511 MWth.
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QUAI)-CITilts
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DPR-30 1.1/2.1 17Ul{t. C1. ADDING IVIl(GitlTY I
SAFETY I.IMIT 1.lMITING SAIT.TY SYS1EM SErlING Applicability:
Applicaldtit):
The safety limits establishn! to preserve the fuct The limiting urety sysicm settings appl 3 to trip cladding integrity apply to those variables which settings of the instrurnents and devices which are anonitor tiie fuel thermal behavior.
provided to paesent the fuel cladding integrity safety limits from being escreded.
Objectiie:
Objectise:
The objective of the safety limits is to estahiish The objective of the hmiting safety spiem settings limits below which the integrity of the fuel ciadding is to def ne the level of the proeco variah!cs at v.hich is p eserved.
automatic protective action is initiated to prevent the fuel cladding intertity safety limits from being exceeded.
SPECIFICATIONS i
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A.
Reaclor Preuure > S00 nig and Core l'Iow A.
Neutron Flux Trip Settings i
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> 10% of Rated The existence of a minimum critical The limiting safety system trip settings shall be power ratjo (stePR) less than 3.00 as specified be!ow;
.i for core loading patterns cont ain-Ing no retrofit Bxel fuel (two water
- 1. APRM Flux Seram Trip Setting (!!un rods) or 1.07 for core lo.eding Mode) patterna containirwy r etrofit ex8
. i,
fuel shall constitute violation of When the reactor mode switch,s in the i
t iv. fuel claddirvy integrity safety Run position t!.e Al'RM nus scram limit.
setting shall be as shown in l'i.use t
i 2.1 1 and shall be:
B.
Core Thermal Power Limit (Reactor Prewure s 600 psi;:)
Ss(65W + 55) l p
When the reactor pressure is : 800 psig or with a mnimum setpoint of 120?e for core l'.ow h less than 10". of rateel, the core core 0 w equal to 96 x 10*1b/hr and l
thermal power shall not execed 25'~, of rated 8' '#
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thermal power.
where:
S
= setting in percent f r;ited C.
Power Transient power
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- 1. The peutron Aus shall not execed the g,
g g; g,,
scram setting established in Spenfica.
qutrw to >roduce a rated cone tion 2.l.A for longer then 1.5 seconds flow of Mi million. Ib/hr. In as indica:ed by the proecss computer.
the event of cparation with a maximum fraetton of limitivJ
- 2. When the process computer is out of power density (NFT.I'D) greater service, ilu,s safety hmit shall be as.
than the frectio s or rated sumed to be cueerded if the neutron power (rFP), the Uttilwj thall Hun eseceds the satam setting. estalr be modified ar. follows:
}
v lished by Sgecifie.ition 2.1.A and a
[,y g control roJ stram does riot occur.
s 6 (.sswo + 55) L m.Pos I
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l QUAD-CITIES DPR-30 D.
Resceor Water level (Shutdown Condition) where:
FRP = fraction of rated Whenever the reactor is in the shut-E** #
down condition with irradiated fuel in the reactor vessel, the water 1evel shall not be less than that MFLPD = maximum fraction of corresponding to 12 inches above the limiting power dens-l top of the active fuel
- when it is ity where the limit-seated in the core.
ing power density for each bundle is
- Top of active fuel is defined to be the design linear 360 inches above vessel zero (Sec heat generation rate Bases 3.2),
for that bundle.
The ratio of FRP/MFLPD shall be set equal to 1.0 unless the actu-al operating value is less than 1.0 in which case the actual operating value will be used.
This adjustment may be accom-plished by increasing the APRM gain and thus reducing the flow reference APRM lligh Flux Scram Curve by the reciprocal of the APRM gain change.
fueling or Startup and llot Standby Mode)
When the reactor mode switch is in the Refuel or Startup flot Standby posi-tion, the APRM scram shall be set at less than or equal to 15% of rated neutron flux.
- 3. IRM Flux Scram Trip Setting The IRM flux scram setting shall be set at less than or equal to 120/125 of full scale.
- 4. When the reactor mode switch is in the startup or run position, the reactor shall not be operated in the natural circula.
tion flow mode.
B.
APRM Rod Block Setting The APRM rod block setting shall be as shown in Figure 2.1-1 and shall be:
S s (.oSWo+ 43) l 1.1/2.1-2
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QUAD-CITIES DPR-30 The definitions used above for the APRM scram trip apply. In the event of oper-ation with a maximum fraction limiting power density (MFLPD) greater than the fraction of rated power (FRP), the setting shall be modified as follows:
FRP S S (.65Wo + 43)
PELPD The definitions used above for the APRM scram trip apply.
The ratio of FRP to PJLPD 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 adjustment may be accomplished by increasing the APRM gain and thus reduc-ing the flow referenced APRM rod block curve by the reciprocal of the APRM gain change.
C. Reactor low water level scram set: ting shall be 144 inches above the top of the active fuel
- at normal operating condi-tions.
D. Reactor low water level ECCS initiation shall be 84 inches (+4 inches /-O inch) above the top 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 valve fast closure scram shall initiate upon actuation of the fast closure sole-noid valves which trip the turbine control valves.
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G. Main steamline isolation valve closure scram shall be s 10% valve closure from full open.
H. Main steamline low-pressure initiation of main steamline isolation valve closure shall be P. 850 psig.
- Top 60 inches above vessel zero of ac-tive fuel is defined to be 3 (See Bases 3.2) 1.1/2.1-3
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1 QUAD-CITIES 1
DPR-30 4i 1.1 SAWTY LIMIT BASIS a
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The fuel cladding integrity limit is set such that no calculated fuel demage would occur as a acsul* of an
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abnormal operational t ransient.
Decause fuel damage is not direct ly obscavabic, 1
uted to establish a safety limit such that the minimum critical power ratio (Mcrs:)a s.tep-back appi oach is s o no less th.in the f ue l l claddinJ integrity safety limit.rcptt > the fuci cladding integrity safety linit margin relative to the conditions required to maintain fuct cladding integrity.
- a presents a conservative 5
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The fuel claddirry is one of the physical barn icrs which separate radioactave materials f rom the environs The integrity of this cladding lorr ier is related to its relative f a cedo,n from periosations or crackanq.
Although some coa rosion or use-related cracking rsay occur during the lif e of the cladding. fission Paoluct migration frcxs this source is incrementally cumulat ive and continuously rwar.ureble.
t uel claddang per-forations, however, can result frora thermal stacsses which occur frcxe scactoa operation signific.ent ly ahnvc design conditions and the protection system safet'y settings. Mhile fissson product migration f rois c! >ddi n j q
perforation is Ju=:t ts steasurable as that from u e-related cracking, the th.:rwilly touted cladding uer f os.
ations signal a threshold beyond which still greater thermal stresses my cause gaov al cladding deter
- oration. Therefore, the fuel cladding safety limit i s de f sned wi t h r.ia n g a n to t he c c a t i-ratt cr than ir.crtea nt-tions which would produce onset of transition hoiling (MCPR of 1.0).
cant departure froa the condition intended by design for planned operation.Ths;r,e condition:. represent a trigniti-integrity safety limit Therefore, abnormal operational transient.is established such that no calculated fuel damage would occur as a result of anthe fuct clen i
documented in Reference 1.
Basis of the values derived for this t.afety limit for each fuct type is e
A.
Reactor Pressuro > 800 peig and Core Flow > 10% of Rated Onset of transition boiling results in a decrease in heat trpusfer dso:s the claddia:9 and thernfore elevated cladding temperature and the possibility of claddits faalute. Howcver, the existe nce of critical power, or boiling transition is not a directly observsblo paremeter in an operating react-or.
Therefore, the margin to boilirw) transition is calculated frce plant opetc t ing parrmote t s t.uc h as core power, core flow, feedweter temperature, and core power dastr abut iva, the marg in i m cach 4
fuel assembly is characterized by the critical power ratio (CPH), which a s tha ratio of the b u.JJe power which would produce onset of transition boiling divided by the actual tm41e powcr.
%e minimum value of this ratio for any bundic in the core is the minimum critical p.swer ratio (ncon).
j lt is assumed that the plant operation is controlled to the nominal prottetave r.ctponts via the instrumented variables (rigure 2.1-3).
The MCPR fuel cladding integrity safety )latit has sufficient con;crvatis.m to assure that in thc evesd of an rhnormal operational transient initiated frots the norrSil oper atinq con.lition. :nore than 99.0a j
of the fuci roda in the core are expoeted to avoid boiling transit ion.
I The margin betwen ECN of I
1.0 (onset of transition boiliruj) and thr: safety limit, is derived from a detailed statist acal
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analysim considering all of the uncertainties in anonitoring the core operating state. includsng uncertainty in the boiling trancition correlation (see c.g., ke f ere nce 3 ). accau:.c the boiling transition correlation is bar.ed on a large quantity of full-scale data.
t he r a in a veny high cnn-fidence that operation of a fuel assembly at the ccidition of McM<
- the fuct cledding integrity safety limit would not produce boiling transition.
However, if boiling transition were to cecur, cladding perforation would not be expected.
cl eddlovi g
temperatures would increase to appro::Imately 1130 r, which is balow the perf neation tempee ntua c of the cladding material. This han been verified by tests in the Cchern) ricct ric Tent fu> actor (ct ra),
where similar fuc1 operated above the critical heat flux for a significant pctiod of t ame (3C uin, utes) without cla,dding perf oration.
If reactor pressure should ever execed 1400 psia during normal power operation (the limit of applicability of thu boiling transition correlation), it would be assumed that the fuel cladding integrity safety limit has been violated,
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In addition to the boiling transition limit (MCPR) cperation is const rained to a maximum f.H cn s 17.5 kw/f t for 7 x 7 fuel arul 13.4kw/f t for all Ox0 fuel types. This constraant is estabitshed by Specification 3.5.J.
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QUAD-Cllll3 DPR-30 O
l Specification 3.5J established the LliGR masimum which cannot be esceeded under steady power operation.
B.
Core Thermal Power Limit (Reactor Preuure<800 psia)
At pressures below S00 psia, the core elevation pressure drop (0 power, O flow) is greater than 4.56 psi.
At low powers and flows this pressure differential is maintained in the bypass region of the core. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low powers and flows will always be greater than 4.56 psi. Analyses show that with a flow of 28 x 10'lb/hr bundle flow, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. T hus the bundle flow with a 4.56-psi driving head will be greater than 28 x 10'lb/hr. Full scale ATLAS mst data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. At 25% of rated thermal power the peak powered bundle would have to be operating at 3.86 times the average powered bundle in order to achieve this bundle power. Thus, a core thermal power limit of 25% for reactor pressures below 800 psia is conservative.
1 C.
Power Transient During transient operation the heat flux (thermal power to-water) would lag behind the neutron flux due to the inherent heat transfer time constant of the fuel. which is 8 to 9 seconth Also, the limiting safety system scram settings are at values which will not allow the reactor to be operated above the safety limit during normal operation or during other pla.it operating situations w hich have been analyzed in detail.
In addition, control rod scrams are such that for normal operating transients, the neutron flux transient is terminated before a significant increase in surface heat flux occurs. Control rod scram times are checked as required by Specification 4.3.C.
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.
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 fails to operate, the design limit of MCPR = the fuel cladding intectrity safety limit is not exceeded.
Thus, use of a 1.5 second limit provides additional marcrin.
The computer provided'has a sequence annunciation progrard which will indicate the sequence in which scrams occur, such as neutron flux, pressure, etc. This program also indicates when the scram setpoint is c! cared. 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 he available for any scram analysis, Specilication I.I.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 elrect of decay heat. If reactor water level should drop below the top of the active 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 perIbration. The core will he cooled sut0ciently to prevent cladding melting should the water level be reduce (to two-thirds the cose height I Atabbsh ment of the safety limit at 12 inches above the top of the fuel provides adesguarc nury.in. 'lhis level w di l
be continuously monitored whenever the rcriirulation pumps are not operating.
- Top of the active fuel is defined to be 360 inches above vessel zero (see Bases 3.2).
1.1/2.14 1
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i QUAD-CITIES DPR-30 TABLE 3.24 NISTRUMENTAT10N THAT INITIATES R00 8 LOCK thaunse llaneer of Operatie er Tripped lastrument Channets per Trip SystemIH lastnument irty Level Setting 2
APRM upscale (flow biasF3
$[0.650W + 439 FRP D
MFLPD 2
APRM upscale (Refuel and Startup/ Hot s12/125 full scate Standby mode) 2 APRM downscalea3 23/125 fut scale 1
Rod block monitor upscale (flow bias)m s0.650W + 42m 1
Rod block monitor downscalem 23/125 fun scale 3
IRM downscale m ts>
23/125 full scale 3
1RM upscale
- s108/125 full scale 2*
SRM detector not h Startup position
- 22 feet below core center-Irie 3
iRM detector not in Startup position
- 22 feet below core center.
leie
{
2* "5 SRM upscale s 105 counts /sec 28 SRM downscale*
2 102 counts /sec 1
Hgh water level in scram discharge volume s25 gallons Notes
- 1. For the Startup/ Hot Standby and Run positions of the reactor mode cciccter switch, there shall be two opc able or tripped trip systems for coca func.
tion except, the SRF. rod blochu.
JIU4 upscale and IRM down:cale need not be oporubic in the Itun pmsition, APMM downneale, APR!4 upscale (flow binned),
and EN4 dowrx:cale need not be operabic in the Startup/ Hot Standuy mule.
l Tte HN1 upscale riecd riot be operable at less than 30% rated ther.m1 power.
One chsnriel rny be byparsed above 30% rated thermal power provided t ha t limiting control rod pattern does not exist. For systems wlt.h m,re than a
one channel per trip system, if the first column cannot be met for one of the two trip systems, this condition may exist for up to 7 days provided that during, t hat tiruo the operabic system is functionally tes ted in-mediately and daily thereaf ter; if this condition lasts longer than 7 days the cystem chall be tripped.
If the first column cannot be mot for both trip cystems, the systems shall be tripped.
- 2. W9 is the percent of drive flow required to produce a rated core flow of 90 million lb/hr.
(2511 Wt).
Trip level setting is in percent of rated power J IRM deenscale may be bypassed when it es en res Ismest range 4 Tha funcien is bypassed when the count rate rs 2 t00 CPS.
$ one of tfie four sRtl sputs may be bypassed.
6 Thrs $Rtl functen may be bypassed a the Ingher IRM ranges (isages 8. 9. and 10) when the IRM spscale rod block is operable 1.
Not regsved to be operable ehde perbreeg low power physics tests at atmospheric pressure dereg or efter refueleg at poner levels not to escoed 5 Mwt.
1 This IRM functen occurs when the reactor enode switch es in the Retuelor Startup/ Hot Standby positen 9 The tre es bypassed when the SAM rs Ibuy mserted 3.2/42-14
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QUAD-CITIES DPR-30 0
1ABLE 3.24 POSTACCIDENT MONiiORING INSTRUMENTAil0N REQUIREMENTS'23 lastrument thamnum IIrrber Readest of Opereble Leceties Nester Channelsmm Parameter Unit 2 Provided Range 1
Reactor pressure 902 5 1
01500 psig 2
0-1200 psig 1
Reactor water level 902-3 2
100 inches + 200 inches (0 inches is top of fuel)
- 1 Torus water temperature 902 21 2
0-200* F 1
Torus air temperature 902 21 2
0 600* F Torus water levet.
902 3 1
-25 inches - + 25 inches 2(4) indicator Torus water level.
I 18 inch range sight glass 1
Torus pressure 902 3 1
-5 inches Hg to 5 psig 1
Drywell pressare 902-3 1
-5 inches Hg to 5 osig O to 75 psg 2
Drywell temperature 902 21 6
0 600* F 2
Neutron monitoring 902-5 4
0.110' CPS 2"I Torus to drywell 2
43 psid differential pressure Notes 1.
Instrument channels required during power operation to monitor poetaccident conditions.
2.
Provisions are ma$e for local sampling and monitoring of drywell etmosphere.
- Top of active fuel is defined to be 360 inches above vessel zero (See Bases 3.2).
O 3.2M.2-15
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QUAD-CITIES DPR-30 0
- 3. The control rod drive housing support
- 3. The correctness of the control rod system shall be in place during reactor power operation and when the reactor withdrawal sequence input to the coolant system is pressurized above RWM computer shall be verined after loading the sequence.
atmospheric pressure with fuel in the reactor vessel, unless all control rods Prior to the start of control rod with-are fully inserted and Specification drawal towards criticality, the capabil-3.3.A.I is met.
ity of the rod worth minimizer to properly ful611 its function shall be Control rod withdrawal sequences a.
shall be established so that max-verified by the following checks:
imum reactivity that could be a.
The RWM computer onh.ne diag-added by dropout of any incre-n stic test shall be successfully ment of any one control blade Performed.
would be such that the rod drop accident b.
Proper annunciation of the selec-design limit of 280 cal /qm._is _not exceeded.
tien error of one out-of sequence b.
Whenever the reactor is in the e ntr I r d shah k veri 6ed.
Startup/ Hot Standby or Run c.
The rod block function of the mode below 20% rated thermal l RWM shall be verified by with-power, the rod worth minimizer drawing the first rod as an out-shall be operable. A second opera-of-sequence control rod no more tot or qualified technical person may be used as a substitute for an than to the block point.
h inoperable rod worth minimizer which fails after withdrawal of at least 12 control rods to the fully withdrawn position. The rod worth minimizer may also be bypassed for low power physics testing to demonstrate the shut-down margin requirements of Specification 3.3.A if a noclear engineer is present and verifies the step-by step rod movements of the test procedure.
4.
Control rods shall not be withdrawn 4.
Prior to control rod withdrawal for for startup or refueling unless at least startup or during refueling, verify that two source range channels have an at least two source range channels observed count rate equal to or greater have an observed count rate of at least than three counts per second and these three counts per second.
SRM's are fully inserted.
- 5. During operation with limiting con-
- 5. When a limiting control rod pattern trol rod patterns, as determined by the exists, an instrument functional test of nuclear engineer, either:
the RBM shall be performed prior to w
rawal f e a.
both RBM channels shall be esignated ro45) operable.
and daily thereafter.
b.
control rod withdrawal shall be blocked; or i
l 3.3 / 4.3-3
QUAD-CITIF.S 1)PR-30 the operating power level shall be c.
limited so that the MCPR will re-main above the MCPR fuel cladding integrity safety limit assuming a sin-gic error that results in complete withdrawal of any single operabic control rod.
C.
Scram Insertion Times C.
Scram insertion Times I. The average scram insertion time, ha-
- 1. After refueling outage and prior to sed on the deenergization of the scram operation above 30% power, with re-pilot valve solenoids at time zero. of all actor pressure above 800 psig. all con-operable control rods in the reactor trol rods shall be subject to scram. time power operation condition shall be no measurements from the fully with, greater than:
drawn position. The scram times shall be measured without reliance on the Average Scram control rod drive pumps.
% inserted from insertion Fully Wuhdrawn Times (sec) 5 0.375 20 0.900 50 2.00 90 3.50 p)
The average of the scram insertion times for the three fastest control rods of all groups of four control rods in a two by two array shall be no greater than:
% laserted Front Average Scram Fully withdrawn Times (sec) 5 0.39N 20 0.954 50 2.12 90 3.80
- 2. The maximum scram insertion time
- 2. Fo!!owing a controlled shutdown of fnr 90% inwrtion orany operable con-the reactor, but not more frequently trol rods shall not exceed 7 seconds.
than 16 weeks not less frequently than 32 week intervals,50% of the control
- 3. If Specification 3.3.C.I cannot he met.
rod drives in M pdrant of &
the reactor shall not he made super-redor m sbli k mrd fm k crtucal: if operating, the reactor shall scram times specified in Specification be shut down immediately upon deter-3.3.C. All control rod drives shall have mination that averare scram time is
- g,
defia,ent.
cach year. Whenever all of the control 4.
If Specification 3.3.C.2 annot he met, rod drive scram times have been mea-the deficient control rod shall be con-sured, an evaluation shall be made to 3.3 / 4.3-4
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QUAD CITIES DPH-30 B.
Control Hod Withdrawal 1.
Control rod dropout accidents as discussed ir. McTerence 1 can lead l
to significant core damacc.
If coupling integrity is maintained, the possibility of a rod. dropout accident is climinated. The over-travel position feature provides a positive cheek, as only uncoupled drives may reach this position.
. Nutron instrumentation response to rod movement provides a verification th'at the rod is following its drive. Absence of such response to drive movement would indicate an uncoup!cd j
condition.
- 2. The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the extremely remote event of a housing failure. The amount of reactivity which could be added by this small amount of rod withdrawal, which is less than a normal singic withdrawal l
increment, will not contribute to any damage to the primary coolant system. T he design basis is given in Section 6 6 l. and the design evaluation is given in Section 6.6.3 of the SAR. This support is not l
required if the reactor coolant system is at atmosphericpressure, since there would then be no driving i
j force to rapidly eject a drive housing. Additionally, the support is not required if all control rods are l
fully inserted or if an adequate shutdown margin with one control rod withdrawn has been demonstrated. since the reactor would remain suberitice! even in the event ofcomplete ejection of the strongest control rod.
3 control rod withdrawal and insertion sequences are established to assure that the maximum insequence individual control rod or control rod segments which are withdrawn could not be worth enough to cause the rod drop accident design limit of 280 cal /cm to be exceeded if l
they were to drop out of the core in the manner defined for the rod drop accident. These sequences are developed prior to initial oper-ation of the unit following any refueling outage and the requirement that an operator follow these acquences is superviced l>y the HWM or a second qualified station employce. These sequences are developed l
to limit, reactivity worthn of control rods and l
together with the integral rod velocity limiters and the actic, of the control rod drive system.
limits potential reactivity insertion such that the results of a control ru * *op accident will not escced a maximum fuel energy content of 260 cal /gm The peak fuel enthalpy or 280 cal /gm is below the i
energy content at which rapid fuel dispersal and primary system damJge hase been found to occur based on esperimental data as is discussed in Iteference 2,
l The analysis of the cont of rod drop accident was originally presented in Sections 7.9 3.14.2.1.1and 14.2.1.4 of the SAR. haprmements in analytical capabiliry have allowed a snore terined analysis of the control rod drop accident.
f
.These techniques are described in a topit.al report (Reference 2) and l
two supplements (References 3 and 4).
In addition, a banked position l
withdrawal sequence deceribed in Rcrerence 5 has been developed to i
further reduce incremental rod worths. Method and Imulu for the rod i
drop accident armlyscs are documented in Referenco 1.
By using the analytical modds dewribed in those reports teupled with conservative or worst. case i
j input parameters, it has been sictermined that for power levels less than2@ of rated power, the l
specified limit on insequence control rod or control rod segment worths will limit the peak fuct enthalpy to less than 280 cal /g. Above20% power even single operator errors cannot result in l
l out-of-sequence control rod worths which are sufficient to reach a peal fuel enthalpy of 250 cal /g should a postulated control rod drop accident occur.
The following parameters and worst-case assumptions have boon utilized in the analysis to determine compliance with the 280 cal /gm peak fuel enthalpy. Each core reload will be analyzed to show conformance to the limitin6 parameters.
l
- a. an interassembly local peaking factor (Reference 6),
t I
3 3/h.3-8 h
5 n..
--n m.
e,-nm
-, - - - +
- - ~ ~ - - -,
4 i
QUAD CITIES DPR-30 b.
the delayed neutron fraction chosen for the bounding reactivity curve l
- c. a beginning-of-life Doppler reactivity feedback
- d. scram times slower than the Technical Specification rod scram insertion rate (Section 3 3.6.1) e.
the maximum possible rod drop velocity of 3.11 fps f.
the design accident and scram reactivity shape function, and g.
the moderator temperature at which criticality occurs In most cases the worth of insequence rods or rod segments in con, junction l
with the actual values of the other important accident analysis parameters described above, would most likely result in a peak fuel enthalpy sub-stantially less than 280 cal /g design limit.
l Should a control drop accident result in a peak fuel energy content of 280 cal /g, fewer than 660 (7 x '
- 7) fuel rods are conservatively estimated to perforate.This would result in an offsite dose well below the guideline value of 10 CFR 100. For 8 x 8 fuel, fewer than 850 rods are conservatively estimated to perforate, with nearly the same consequences as for the 7 x 7 fuel case because of the rod power l
differences.
The rod worth minimizer provides automatic supervision to assure that out of sequence control rods j
will not be withdrawn or inserted; i.e., it limits operator deviations from planned withdrawal sequences (reference SAR Section 7.9). It serves as a backup to procedural control of control rod worth. In the event that the rod worth minimizer is out of service when required, a licensed operator or other qualified technical employee can manually fulfill the control rod pattern conformance function of the rod worth minimizer, in this case, the normal procedural controls are backed up by independent procedural controls to assure conformance.
4.
The source range monitor (SRM) system performs no automatic safety system function, i.e., it has no scram function. It does provide the operator with a visual indication of neutron level. This is needed for knowledgeable and efficient reactor startup at low neuuon levels. The consequences of reactivity accidents are functions of the initial neutron flux. The requirement of at least 3 counts per second assures that any transient, should it occur, begins at or above the initial value of 10' of rated power used in the analyses of transients from cold conditions. One operable SRM channel would be adequate to monitor the approach to criticality using homogeneous patterns of scattered contrcel rod withdrawal. A minimum of two operable SRM's is provided as an added conservatism.
- 5. The rod block monitor (RBM) is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power operation. Two channels are provided, and one of these may be bypassed from the console for maintenance and/or testing. Tripping of one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage. This system backs up the operator, who withdraws centrol rods according to a written sequence. The specified restrictions with one channel out of service conservatively assure tnat fuel damage will not occur due to rod withdrawal errors when this condition exists. During reactor operation with certain limiting control rod patterns, the withdrawal of a designated single control rod could result in one of more fuel rods wi th MCPit's less than the MCPR fuel cladding integrity safety limit.During use orsuch patterns, it is judged that testing of the RBM system to assure its operability prior to withdrawal of such rods will assure that improper withdrawal does not occur. it is the responsibility or the Nuclear Engineer to identify these limiting patterns and the designated rods either when the patterns are initially established or as they develop due to the occurrence ofinoperable control rods in other than limiting
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patterns.
3 3A.3-9
QUAD CITIES I
I DPR-30 C.
Scram Insertion Times The control rod system is analyzed to bring the reactor subcritical at a rate fast enough to prevent fuel damage, i.e.,
to prevent the MCPR from becomin6 less than the fuel cladding integrity safety limit.
The limiting power transient is that resulting from a turbine control (Generator Load Rejection) valve closure with failure of the turbine bypass system.
Analysis of this transient shows that the nega tive reactivity rates resulting from the scram with the average response of all the drives as given in the above specification, provide the required protection, and MCPR remains greater than the fuel cladding integrity i
safety limit.
The minimum amount of reactivity to be inserted during a scram is controlled by permitting no more than 10% of the operable rods to have long scram times. In the analytical treatment of the transiems 390 milliseconds are allowed between a neutron sensor reaching the scram point and the start of motion of the control rods. This is adequate and conservative when compared to the typically observed time delay of about 270 milliseconds. Approx-imately 70 milliseconds after neutron riux reaches the trip point, the pilot scram valve solenoid deenergizes. Approximately 200 milliseconds later, control rod motion begins. The time to deenergize the pilot valve scram solenoids is measured during the calibration tests required by Specification 4.1. The 200 milliseconds are included in the allowable scram insertion times specified in Specification 3.3.C.
The scram times for all control rods will be determined at the time of each refueling outage. A representative sample of control rods will be scram tested at increasing intervals following a shutdown.
Scram times of new drives are approximately 2.5 to 3 seconds; lower rates of change in scram times following initial plant operation at power are expected. The test schedule at increasing time intervals I
provides reasonable assurance of detection of slow drives belbre system deterioration beyond the limits I
of Specification 3.3.C.The program was developed on the basis of the statistical approach outlined below and judgment.
j The probability that the mean 90% insertion time of a samp!c of 25 control rod drives will not excced 0.25 seconds of the mean of all drives is 0.99 at a risk of 0.01. If the mean time exceeds this range or the mean 90% insertion time is greater than 3.5 seconds, an additional sample of drives will be measured to verify the mean performance.
h Since the differences between the expected observed mean insertion time and the limit of Specification j
3.3.C greatly exceed the expected range, this sampling technique gives auurance that the limits of Specification 3.3 C will not he exceeded. As further assurance that the limits of Specification 3.3.C will not be exceeded, all operable drives will be scram tested to determine compliance to Specification 3.3.C if the enlarged sample of 50 control rods exceeds 4.25 seconds. The 0.75 second margin to the limit is greater than the maximum capected deviation from the mean and therefore gives assurance that the mean will not exceed the limit of Specification 3.3.C. In addition,50% of the control rods will be checked every
{
16 weeks to verify the performance and for correlation with the sarapling program.
The history of drive performance accumulated to date indicates that the 90's insertion times of new and overhauled drives approximate a normal distribution about the mean which tends to become skewed toward longer scram times as operating time is accumulated. The probability of a drive not exceeding the mean 90% insertion time by 0.75 seconds is greater than 0.999 for a normal distribution. The measurement of the scram performance of the drives surrounding a drive exceeding the expected range of scram performance will detect local variatior.s and also provide assurance that kical scram time limits are not cacceded. Continued monitoring of other drives exceeding the expected range of scram times provides surveillance of possible anomalous perforrnance.
The numerical values assigned to the predicted scram performance are based on the analysis of the Dresden 2 startup data and of data from other ilWICs such as Nine Mile Point and Oyster Creek.
1 The occurrence of scram times within the limits, but significantly longer than average, should be viewed as an indication of a systema' tic problem with contiol rod drives, especially if the number of drives exhibiting uch scram times exceeds eight, the allowable number ofinoperable rods.
I 3.3 / 4 3 -10 4
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I QUAD-CITIES DPil-30 D. Control Rod Accumlators The basis for this specification was not described in the SAR and is therefore presented in its entirety.
Requiring no more than one inoperable accumulator in any nine. rod square array is based on a series of XY PDQ.4 quarter cose calculations of a cold clean core.The worst case in a nine-rod withdrawal sequence resulted in a k, < l.0.Other repeating rod sequences with more rods withdrawn resulted in k > l.0. At reactor pressures in excess of 800 psig. even those control rods with inoperable g
accumulators will be able to meet required scram insertion times due to the action of reactor pressurc.
In addition, they may be normally inserted using the control rod drive hydraulic system. Procedural control will assure that control rods with inoperable accumulators will be spaced in a one in-nine array rather than grouped together.
E.
Reactivity Anomatics During each fuel cycle. excess operating reactivity varies as fuel depletes and as any burnable poison in supplementary control is burned. The magnitude of this excess reactivity may be inferred from the critical rod configuration. As fuel burnup progresses, anomalous behavior in the excess reactivity may be detected i
by comparison of the critical rod pattern selected base states to the predicted rod inventory at that state.
~
Power operating base conditions provide the most sensitive and directly interpretable data relative to core reactivity. Furthermore, using power operating base conditions permits frequent reactivity comparisons.
Requiring a reactivity comparison at the specified frequency assures that a comparison will be made before the core reactivity change exceeds 1% ak. Deviations in core reactivity greater than 1% ok are not expected and require thorough evaluation. A 1% reactivity limit is considered safe.since an insertion of the reactivity into the core would not lead to transients exceeding design conditions of the reactor system.
F.
Economic Generation Control Sprem h
Operation of the facility with the economic generation control system (EGC)(automatic flow control) is limited to the range of 65% to 100% of rated core flow. in this flow range and with reactor power above 20%, the reactor could safely tolerate a rate of change ofload of 8 MWe/sec (reference SAR Section 7.3.5 ).
Limits within the EGC and the flow control system prevent rates orchange greater than approximately 4 MWe/sec. When EGC is in operation this fact will be indicated on the main control room console. The results ofinitial testing will be provided to the NRC before the onset of routine operation with EGC.
References
- 1. " Generic Reload iuel Applica tion", NEDE-24011-P-A" l
- 2. C. J. Paone, R.C.Stirn, and J. A. Wooley,* Rod Drop Accident Analysis for Large BWR *v GE Topical Report l
3 NEDO-10527, Marsh 1972.
- 3. C. J. Paone. R. C. Stirn, and R. M. Young.* Rod Drop Accident Analysis for I.arge RWR'v, Supplement I.
l GE Topi:a! Report NEDO-10527. July 1972.
- 4. J. M. Itavn, C. J. Paone, and R. C. Stirn. ' Rod Drop Accident Analysis for Large BWR*s, Addendum 2 l
Exposed Cores,' Supplement 2, GE Topical Report NEDO 10527 January 1973.
- 5. C. J.
Paone, " Banked position withdrawal sequence," Licensing topical Report NEDO-21231, January, 1977.
- 6. To include the power spike errect caused by gaps between ruel pellets. I
- Approved revialon runnber a t 1.bne reload fuel ana l.ynen are pe rronnmi.
l 3 3/4.3-11
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l QUAD CITIES DPR-30 3.4 LIMITING CONDITIONS FOR OPERATION BASES A.
The design ob,Jective of the standby liquid control ayatem la to provid the capability or bringing the reactor from rull power to a cold, xenon-free shutdown assuming that none of the withdrawn contrcl rods can be incerted.
To meet thia objective, the liquid control system is designed to inject a quantity or baron which produces a concentration of no less than 600 ppm or baron in the
{
reactor core in approximately 90 to 120 minutes with imperfect mixing.
A buran concentration of 600 ppm in the reactor care is required t.o bring the reactor from full power to 3% ak or mare l
cuberitical condition considering the hot to cold reactivity swing, xenon poisoning and an additional margin in the reactor core for imperfect mising of the i
chemical solution in the reactor water. A normal quantity of 3470 gallons of solution hasing a !3 4%
sodium pentaborate concentration is required to meet this shutdown requirement.
The time requirement (90 to 120 minutes) for insertion of the boron solution was selected to override the rate of reactivity insertion due to cooldown of ihe reactor fo!!owing the senon poison peak. I or a required pumping rate of 39 gpm the maximum storage volume of the boron solution is established as 4875 gallons (195 gallons are contained below the pump suction and. therefore, cannot be inserted).
Baron concentration, solution temperature, and volume are checked on a frequency to assure a high reliability of operation of the system should it ever be required. Experience with pump operabihty indicates that monthly testing is adequate to detect if failures have occurred.
The only practical time to test the standby liquid control system is during a refueling outage and by initiation from local stations. Components of the system are checked periodically as described above and make a functional test of the entire system on a frequency of less than once each refueling outage unnecessary. A test of explosive charges fran one manufacturing batch is made to assure that the charges are satisfactory. A continual check of the firing circuit continuity is provided by pilot lights in the control roorn B.
Only one of the two standby liquid control pumping circuits is needed for proper operation of the sprem.
If one pumping circuit is found to be inoperable, there is no immediate treat to shutdown c.ipabihty, and reactor operation may continue while repairs are being made. Assurance that the remaining sptem will perform its intended function and that the reliability of the system is good is obtained by demonstrating operation of the pump in the operable circuit at least once daily. A reliability analysis indicates that the plant can be operated safely in this manner for 7 days.
C.
The solution saturation temperature of 13% sodium pentaborate, by weight,is 59' F. 'lhe solution shall be kept at least 10
- F thove the saturation temperature to guard against boron precipitation. l he 10
- F margin is included in Figure 3.3-1. Temperature and liquid level alarms for the system are annunciated in the control room.
Pump operability is checked on a frequency to assure a high reliability of operation of the system should it ever be required.
Once the solution has been made up, boron concentration will not vary unless more baron or more water is added. Level indication and alarm indicate whether the solution volume has thanged, which might indicate a possible solution concentration change. Considering these factors, the test interval has been established.
3.4/4.4-3
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QUAI)-CITIF.S i
DPR-30 O
cycle by assuring that water can be run thiough the drain lines and i
actuating the air-operated vahes by operation of the following sensors:
I I) loss of air
- 2) esguipmeni drain sinnp high level I
- 3) vault high level i
- d. The condemer pit 5. foot trip cia-cuits for each channel > hall be checked once a month. A lapic j
system functional test shall be per-L formed during each refueling outage.
I.
Average Planar LiiGR I.
Average Planar I.llGR During steady state power operation, the average Daily during ateady atate operation linear heat generation rate (APLilGR) of all the a.bove 25% rated therma 1 power, rods in any fuel assembly,as a function of average the average planar LHGR shall planar exposure, at any axial location, shall not be checked.
exceed the maximum average planar LIIGR h
shown in Figure 3.51 If at any time l during operation it is determined by siorinal sur-veillance that the limiting value for APLilGR is J.
Local I.IIGit being exceeded, action shal1 be initiated within l5 minutes to restore operation to.nthin the pre-Daih hig madpm pr pion scribed limits. If the APLilGR is not returned in above 25% of rated thermal power, the hical LHGR shall be checked.
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 j
corresponding action shall continue until reactor operation is within the presenhed limits.
J.
Iecal LifGR During steady state power operation, the linear 4
heat generation rate (LilGR) of any rod in any fuel assembly at any axial location shall not exceed the maximum allowable LilGR -
If at any time during operation it is determined by normal surveillance that the hiniting value for LilGR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the pre-scribed limits. If the LilGR is not returned to l,
3.s u.s.,
(
I QUAD CITIES DPR-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 4
condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveillance and cor.
responding action shall continue until reactor operation is within the prescribed limits.
Maximum allowable LHGR for all 8XO fuel types is 13 A KW/ft.
For 7X7 and mixed oxide fuel, the raaximum allowable LilGR is as follows:
LHGR_ <LilG R,1 -t aP/P)_(L/L,)
where:
i I
- tilGR,
=, design LilGR 17 5 kW/ft.
=
)
1
( AP/P)_
= masimum power spiking penalty
.035 initial core fuel
=
=.029 reload 1. 7 x 7 fuel
.028 reload I. mixed oxide fuel
=
L,
= total core length 12 feet
=
L
= Axial distance from hottom of core f
i K.
Minimum Critical Power Ratio (MCPR)
K.
kinimum Critical Power Ratio (MCPR)
During steady. state operation MCPR shall be The MCPR shall be determined d.nly durmt treater than or equal to steady-state power operation abme 2.4'; of 1.35 (7 x 7 fuel) rated thermal power.
1.35 (8 x 8 fuel) at rated power and flow. If at any time during operation it is determined by normal surveillance that the limiting value for MCPR is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed limits.
If the steady-state MCPR is not returned to within the prescribed hmits 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 prescrit.ed limits. For core flows other than rated, these nominal values of MCPR shall be increa ed by a factor of kg where k i
g s as shown in Figure 3.5.2.
3.5/4.5-10 1
e
%q----9 g-w-
I
(
i QUAD-CITIF.S DPR diesel generators. All of these systems have been sized to perform their intended function considering the simultaneous operation of both units.
These technical specifications contain only a single reference to the operability and surseillance requirements for the shared safety-related features of each plant. The level of operabihty for one unit must be maintained independently of the status of the other. For example, a diesel (1/2 diesel) which is shared between Units I and 2 would have to be operable for continuing Unit I operation even if Unii 2 were in a cold shutdown condition and needed no diesel power.
Specification 3.5.F.3 provides that should this occur, no work will be performed which could preclude adequate emergency cooling capability being available. Work is prohibited unless it is in accordance with specified procedures which limit the period that the control rod drive housing is open and assures that the worst possib!c loss of coolant resulting from the work will not result in uncovering the reactor core.
Thus, this specification assures adequate core cooling. Specification 3.9 must be consulted to determine other requirements for the diesel generator.
G.
Maintenance of Filled Discharge Pipe If the discharge piping of the core spray, LPCI mode of the RHR, HPCI, and RCIC are not filled, a water hammer can develop in this piping, threatening system damage and thus the availability of emergency cooling systems when the pump and/or pumps are started. An analysis has been done which shows that if a water hammer were to occur at the time emergency cooling was required, the systems would still perform their design function. However, to minimize damage to the discharge systems and to ensure added margin in the operation of these systems, this technical specification requires the discharge lines to be filled whenever the system is in an operable condition.
Specification 3.5.F.4 provides assurance that an adequate supply of coolant water is immediately availa$le to the low-pressure core cooling systems and that the core will remain covered in the event of a loss-of-coolant accident while the reactor is depressurized with the head removed.
H. Condensate Pump Room Flood Protection See Specification 3.5.H.
1.
Average Planar 1.HGR This specification assures that the peak cladding temperature following a postulated design basis loss-of-coolant accident will not exceed the 2200* F limit specified in 10 CFR 50 Appendix K considering the postulated effects of fuel pellet densification.
The peak cladding temperature following a postulated loss-of-coolant accident is primarily a function of the average LilGR of all the rods in a fuel assembly at any axial location and is only secondarily dependent on the rod-to-rod power distribution within a fuel assembly. Since expected local variations in power distrihution within a fuel assembly aflect the calculated peak cladding temperature by less than 120' F relative to the peak temperature for a typical fuel design,the limit on the average planar LilGR is sufficient to assure that calculated temperatures are below the 10 CFR 50 Appendix K limit.
The maximum average planar I.HGR's shown in Figure 3.51 are based on calculations employing the models described in Reference 2. Power operation with LHGR*s at or below those shown in Figure 3.5 1
\\
assures that the peak cladding temperature following a postulated loss-of. coolant accident will not exceed the 2200' F limit. These values represent limits for operation to ensure conformance with 10 CFR 50 and Appendix K only if they are more limiting than other design parameters.
The rnaximum average planar LHGk's plotted in Figure 3.51 at higher exposures result in a calculated peak cladding temperature ofless than 2200' F. However, the maximum average planar LilGR's are 3.5/4.5-13
-m-.--
s-
(
(
QtlAD-CI11ES DPR-30 shown on Figure 3.5-1 as limits because conformance calculations have not been peiformed to justify operation at 1.IIGR's in excess of those shown.
J.
local 1.1IGR This specification assures that the maimt.m linear heat reneration rate in any rod is less than the deugn linear heat generation rate even if fuel pellet densification is postulated. The power spite penalty
,1s discussed in Reference 2 and assumes a linearly increasing variation in asial l
gaps between core bottom and top and assures with 95% confidence that no more than one fuci rod exceeds the design LilGR due to power spiking.
K.
Minimum Critical Power Ratio (MCPR)
The steady state values for MCPR specified in this specification were selected to provide margin to acconuno.
date transients and uncertainties in monitoring the core operating state as well as uncertainties in the critical power correlation itself. These values aho assure that operation will be such that the initial condition assumed for the LOCA analysis plus two percent for uncertainty is satisfied.
For
[
any of the special set of transients or disturbances caused by single operator error or single equipment malfunction,it is required that design analyses initialized at this steady. state operating limit yield a MCPR of not less than that specified in Specification 1.1.A at any time during the transient. assuming instrument trip settings given in Specification 2.1. For analysis of the thermal consequences of these transients, the value of MCPR stated in this specification for the limiting condition of operation bounds the initial value of MCPR assumed to exist prior to the initiation of the transients.
This initial condition, which is used in the transient analyses, wil L pre-clude violation of the fuel cladding integrity safety limit.
Assumptions and methodc used in calculating the required steady state MCPR limit for each reload cycle are documented in Reference 2.
The results apply with increased conservatism while opera ting with MCPR's greater than specified.
h The most limiting transients with respect to MCPR are generally:
a) Rod withdrawal error b) Load rejection or Turbine Trip without bypass c) loss of feedwster heater Seve ra l f actors influence which of these transients results in the l a rges t reduction in critical power ratio such as the specific fuel loading, ex-posure, and fuel type.
The current cycles reload licensing analyses spec-
)
ifies the limiting transients for a given exposure increment for each fuel type.
The values specified as the Limiting Condition of Operation a re con-servatively chosen to bound the most restrictive over the entire cycle for l
e::c h i'ne I 1.ype.
1
- 3. 5/h. 5-14
(
I Q U A D-CITIES DPR.30 0
For core Dow rates less than rated, the steady state MCPR is increased by the formula given in the specifi-cation. This assures that the MCPR will be maintained greater than that specified m Specification I.I.A esen in the event that the motor generatot set specJ controller causes the scoop tube positioner for the Duid coupler to move to the maximum spec <! position.
References
- 1. " Loss-of-Uoolant Analysis Report for Dresden Units P.
3 and Quad Cities Unita 1, I:ue l ea r Power Sta tionc, " Isi.DO-p4146A x,
9 April, 1979
- 2. " Generic Reload Fuel Application," NEDE-240ll-P-A"
- 3. I. M. Jacobs and P. W. Marriott, GE Topical Report APED 'i/36,
" Guidelines for Determining Safe Test Intervals and Hepair Times for Er.gineered Safeguards," April,1969 f
- Approved revision at time of plant operation.
- Approved revision number at time reload fuel analyses are performed.
- 3. 5/h. 5-lha
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30.000 10.000 20.000 PLANAR AVERAGE EXPOSURE (.WD/T)
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- FIGURE 3.5 1 MAUMUM AVERAGE PLANAR LINEAR HEAT GENERATI0ff RATE (MAPLHGR)
V5. PLANAR AVERAGE EXPOSURE (Sheet 1 of M r
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FIGugE 3.5-1 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGo,'
VS, PLANAR AVERAGE EIP05URE (Sheet 2 of 1 )
4 s-
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VERSUS PLA.N AR AVERAGE EXPCSURE und Cities Unit 2 g.
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PWtAA AVE R AGE E k POSU A E luvudhi
. MAX 1%I4 AVERAGE FLM AR LINEAR HEAT GENE 9ATMN RATE UWLHGR)
WEP.3US Pt.AN AR AVERAGE EXPOSURE Figure 3.5-1
(%eet b of h)
(
(
QUAD-CITIES DPR-30 Should the switches at levels (a) and (h) fail or the operator fail to trip the circulating water pumps on alarm at level (S), the actuation ofeither level switch pair at level (c) shall trip the circulating water pumps automatically and alarm in the control room. These redundans level switch pairs at level (c) are designed and installed to IEl.1:
279, ' Criteria for Nuclear Power Plant Protection Spreme As the circulating water pumps are tripped cithen manually or automatically, at level (c) of 5 feet, the masimum water level reached in the condenser pit due to pumping will be at elevation $68 feet 6 inches elevation (10 feet above condenser pit floor elevation 558 feet 6 inches; 5 feet plus an additional 5 feet attributed to pump coastdown).
In order to prevent the RIIR service water pump motors and diesel. generator cooling water pump motors from overheating, a vault cooler is supplied for each pump. I;ach vault cooler is designed to maintain the vault at a maximum 105
- F temperature during operation of its respective pump. For cumple, if diesel generator coohny water pump I/2-3903 starts, its cooler also starts and maintains the vault at 105 F by removing heat supplied to the vault by the motor of pump I/2-3903. If,at the same time that pump I/2-3903 is in operation. RilR service water pump IC starts,its cooler will also start and compensate for the added heat supplied to the vault by the IC pump motor keeping the vault at 105' F.
Each of the coolers is supplied with cooling water from its respective pump's discharge !ine. After the water has been passed through the cooler it returns to its respective pump's suction line. In this way the vault coolers are supplied with cooling water totally inside the vault. The cooling water quantity needed for each cooler is approximately 14 to 5% of the design flow of the pumps so that the recirculation of this small amount of heated water will not affect pump or cooler operation.
Operation of the fans and coolers is required during shutdown and thus additional surveillance is not required.
Watertight vauhs for the ECCS pumps in the reactor building are tested in es>entially the same manner and frequency as described for the condenser pump room vauhs.
I Verification that acceu doors to each vault are closed following entrance by personnel is covered by station operating procedures.
The LHGR shall be checked daily to determine if fuel burnup or control rod movement has caused changes in power distribution. Since changes due to burnup are slow and only a few control rods are moved daily, a daily check of power distribution is adequate.
Ascrage Planar LIIGR At core thermal power levels less than or equal to 25%. operating plant experience and thermal hydraulic anal)ses indicate that the resuhing average planar LIIGR is below the maximum average planar LIIGR by a considerable margin: therefore, evaulation of the average planar Ll!GR below this power level is not necessary. The daily requirement for calculating average planar LilGR above 25% rated thermal power is suHicient since power distribution shifts are slow when there have not been significant power or control nod changes.
I ocal LilGR The LilGR as a function of core height shall be checked daily during reactor operation at greater than or equal to 25% power to determine if fuel burnup or control rod movement has caused changes in power distribution. A limiting LHGR value is precluded by a considerable margin when employing any permissible control rod pattern below 25% rated thermal power.
1 Minimum Critical Power Ratio (MCPR)
At core thermal power levels less than or equal to 25%. the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small. For all designated control rod patterns which may be employe I at this point, operating plant c.sperience and thermal hydraulic analysis indicate that the resulting MCPR value is in excess of requirements by a considerable margin. With this low void content, any inadvertent core flow increase would only place ope:ation in a more conservative mode relative to MCPR.
l I
3.5/4.5-17
~
__m__
(
(
4 QllAll ClliF.S f
DPit-30 O
2.
Hoth the $ ump and ai: sainphnp sys-tems shat! he operahir durmp reactor powcr operation luom and arter the date that one of these sy stems is m.ide j
or found to be inoperah!c for any rea-son. re.ictor power operation is [w -
rniwihle only during the succcedmr 7 days.
{
3.
If the conditions in I or 2 alwe een-not be mer. an orderly shutdow n slull be initiated anst the reactor shall be in a cohl shutdown conduion within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> t
I Safrsy ami HelicI Vahes I.
Safety and Itrlief Yahes 1.
Prior to reactor startup for p.mes op.
A minnnum of I/2 of all salery sahes shall be l
cration. during reactor juwer operat.
bench chetked or replaced with a benth ing conditions. and w benever the reac-checked valve cath refueling outare. 'Ihe pop-tot coolant pressure is pieater than 90 ping point of the safety vahes shall be set as psig and temperature greater than tidlows:
320 ' F. all nine of the safitv valves shall he operable 3 he sidenoid-Nwnber of Valves Serpona tystg) i activated pressure vah cohall be oper-g able as requited by Specification 3.5.D.
~
2 1250 l
2.
If Specification 3M I is not met, the 4
12(,o rextor shall remain shut down until the condition is corsected nr. if in The allowable setpomt earor for each sabe is operation. an orderly shutdown shall i 1%.
be initiated amt the teactor coolant pressure imd temper.nure shall he 1 f vMm slWI ht ddd fm w t p-M 4
""'# '"' ' " "'E ""'"E"
'#I I'#""
I below 90 psig and 320 1; within 24 t
shall be:
{
hours Number of Valves Setynnu (psig)
I I I F" l
j 2
s 1130 2
< 1135 1
{
" Target l{ock combitt.nion s.ifity /tchef valve.
1 F.
Structural Integrity F.
Structural Inregrity The structural integrity of the primary system The nondestmrtive inspections listed in Tahic boundary shall be maintained at the level required 4.61 shall he perfonned as siveified in accor-by the ASMI: lloder and Pressme Vessel Code, dance with Section XI of the AShil\\lloiler and Stetion XI. Rules for Inscrwe Inspection of
. Pressure Vessel Cmic, 1971 1:dition. Summer i
Nuclear Power ' Plant Component s".
1974 1971 Addenda, the tesults obtamed from com-lihtion. Sammer 1975 A ldenda (ASNili Code pliance with this specification will be evaluated Section XI).'
after 5 years and the condusions wdl be te-viewed with the NRC.
r i
l l
3.M 41e-4
~.
s 4
.m..
i Qt'AI)-CITIES Ill'R -30 i
l_
e
=
Refuences
" Generic Reload fuel Applications," NEDE-24011-P-A*
1.
- Approved revision number at time reload fuel analyses are p'erformed.
I i
+
O I>
h 1.1/ 2.1 -6
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. _ __.H QtlAIMTill3 Di'R-30 2.11.lMITING SAFI TY SYSTES1 SETTING liASES The abnormal operational transients applicabic to operation of the units have been analyzed throughout the spectrum of planned operatiag conditions up to the rated thermal power condition of 25 I I MW. In addition. 2511 MW is the licensed maximum steady-state power level of the units. This maximum steady-state power level will never knowingly be exceeded.
Conservatism is incorporated in the transient analyses in estimating the controlling factors, such as void reactivity coefficient, control rod scram worth, scram delay time, peaking factors, and axial power shapes. These factors are selected conservatively with respect to their effect on the applicable transient results as detemined by the current analysis model.
Conservatism incorporated into the transient analysis is documented in Reference 1.
Transient analyses are initiated at the conditions given in this Reference.
I tie absolute vatue of tne voto reactivity coemetent useo in tne anarysis is conservativeiy estimatea to oe anout n.o greater than the nominal maximum value expected to occur during the core lifetime. The scram worth used has been derated to be equivalent to approximately 80"o of the total scram worth of the control rods. Ihe scram delay time and rate of rod insertion allowed by the analyses and 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 reactivity insertion.
The rapid insertion of negatise reactivity is assured by the time requirements for 5% and 2L% insertion. By the time the rods are 607 inserted. approximately 4 dollars of negative reactivity have been inserted, which strongly nuns the transient and accomplishes the desired effect. The times for 50% and 90% insertion are given to assure h
proper completion of the expected performance in the earlier portion of the transient. and to establish the uhimate fully shut down steady-state condition.
This choice of using conservative values of c antrolling parameters and initiating transients at the design power level produces more pessimistic answers th.n would result by using expected values of control parameters and analyzing at higher power levels.
$teady-state operation uithout forced recirculation will not be permitted except during startisp testing. The analysis to support operation at various power and flow relationships has considered operation with either one or two recirculation 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 conserva-tively assumed to ex'ist prior to initiation of the transients.
A.
Neutron Flus Trip Setting 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 chambers provide the basic input signals, the APRM system responds directly to average neutron flux. During transients, the instantaneous rate of heat transfer from the fuel (reactor thermal powet )
is less than the instantaneous neutron flux due to the time constant of the fuel. Therefore during abnormal operational transients. the thermal power of the fuel will be less than that indicated by the neutron flus at the scram setting. Analyses demonstrate that with a 120% scram trip setting. none of the abnormal operational transients analyzed violates the fuel safety limit. and there is a substantial margin from fuel damage.Therefore, the use of flow referenced scram trip provides even additional margin.
1.1/ 2.1 -7
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QUAD-CITIES DPR An increase in the APRM scram trip setting would decrease the margin present before the fuel cladding integrity safety limit is reached. The APRM scram trip setting was determined by an analysis of margins required to provide a reasonable range for maneuvering during operation. Reducing this operating margin would increase the frequency of spurious scrams, which nave an adverse ef fect on reactor safety because of the resulting thermal stresses.
Thus, the APRM scram trip setting was selected because it provides adequate margin for thi fuel cladding integrity safety limit yet allows operating margin that reduces the possibil-ity of unnecessary scrams.
Tha scram trip setting must be adjusted to ensure that the LHGR transient peak is not increased for any combination of max!. mum fraction of limiting power density (MFLPD) and reactor core thermal power. The scram setting is adjusted in accordance with the formula in Specification 2.1. A.1, when the MFLPD is greater than the fraction of rated power (FRP).
The adjustment may be accomplished by increasing the APRM gain and thus reducing the flow referenced APRM High Flux Scram Curve by the reciprocal of the APRM gain change.
2.
APRM Flux Scram Trip Setting (Refuel or StartupAtot Standby Mode)
For operation in the Startup mode while the reactor is at low pressure, the APRM scram setting of 15% of rated power provides adequate thermal margin between the setpoint and the safety limit, 25% of rated. The margin is adequate to accor= odate anticipated maneuvers associated with power plant startup.
Effects of increasing pressure at zero or low void content are minor, cold water from sources available during startup is not much colder than that already in the system, temperature coefficients are small, and ~edatFol rod patterns are constrained to be uniform by operating procedures backed up by the rod worth minimirer. Of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power rise. Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change powet by a signifi-cant percentage of rated power, the rate of power rise is very slow. Generally, the heat flux is in near equilibrium with the fission rate.
In an assumed uniform rod withdrcwal approach to the scram level, the rate of power rise. is no more than 5% of rated power per minute, and the APRM system would be more than adequat'e to assure a scram before the power could exceed the safety limit. The 15% AFRM scram remains active ~ until the mode switch is placed in the Run position. This switch occurs when reactor pressure is greater than 850 psig.
3.
IRM Flux Scram Trip Setting The IRM system consists of eight chambers, four in each of the reactor protection system logic channels. The 1RM is a 5-decade instrument which covers the range of power kvel between that covered by the SRM and the APRM. The 5 decades are broken down into 10 rango, each being one-half a decade in size.
The IRM scram trip setting of 120 divisions is active in each range of the IRM.
For example, if the instrument were on Range 1, the scram setting would be 120 divisions for that range-likewise, if the instrument were on Range 5, the scram would be 120 divisions on that range.
Thus, as the IRM is ranged up to accomodate the increase in power level, the scram trip set-ting is also ranged up.
The most significant sources of reactivity change during the power incrtase are due to control rod withdrawl. In order to ensure that the IRM provides adequate protection against the single rod withdrawal error, a range of rod withdrawal accidents was analyzed. This analysis included starting the accident at various power levels. The most savne case involves an initial condition in which the reactor is just suberitical and W ;?.M system is not yet on scale.
Additional conservatism was taken in this analysis by assuming that the IRM channelC10SeSt to the withdrawn rod is bypassed. The results of this analysia show that the reactor is scrammed and peak power limited to 1% of rated power, thus maintaining MCPR above the fuel cladding integrity safety limit. Based on the above analysis, the IP11 provides protection against local control rod withdrawal errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM.
1.1/2.1-9
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I QUAD-CITIES DPR-30 5
APRM Rod Block Trip Setting Reactor power level may be varied by moving control rods or by varying the recirculation flow rate. The APRM system provides a control rod block to prevent gross rod withdrawal at constant recirculation flow rate to protect against grossly exceeding the MCPR Fuel Cladding Integrity Safety Limit. This rod block trip setting, which is automatically varied with recirculation loop flow rate, prevents an increase in the reactor power level to excessive values due to control rod withdrawal. The flow variable trip setting provides substantial margin from fuel damage, assuming a steady-state operation at the trip setting, over the entire recirculation flow range. The margin to the safety limit increases as the flow decreases for the specified trip setting versus flow relationships therefore the worst-case MCFR which could occur during steady-state operation is at 108% of rated thermal power because of the APRM rod block trip setting. The actual power distribution in the core is established by specified control rod sequences and is monitored continuously by the incore IPRM system. As with APRM scram trip setting, the APRM rod block trip setting is adjusted downward if the maximum fraction cf limit-ing power density exceeds the fraction of rated power, thus preserving the APRM rod block safety margin. As with the scram setting, this may be accomplished by adjusting the APRM gains.
C.
Reactor Low Water Level Scram The reactor low water level scram is set at a point which will assure that the water level used in the bases for the safety limit is maintained. The scram setpoint is based on normal operat-ing temperature and pressure conditions because the level instrumentation is density compensated.
D.
Reactor Low Low Water Level ECCS Initiation Trip Point The mergency core cooling subsystems are designed to provide sufficient cooling to the core to dissipate the energy associated withthe loss-of-coolant accident and to limit f uel cladding temperature to well below the cladding melting tempatal ure to assure that core geometry remains intact and to lir.it any cladding metal-water reaction to less than 1%.
To accomplish their intended function, the capacity of each emergency core cooling system component was established based on the reactor low water level scram setpoint. To lower the setpoint of the low water level scram would increase the capacity requirement for each of the ECCS components. Thus, the reactor vessel low water level scram was set low enough to permit margin for operation, yet will not be set lower because of ECCS capacity requirements.
The design of the ECCS components to meet the above criteria was dependent on three previously set parameters: the maximum break sire, the low water level scram setpoint, and the ECCS initiation setpoint.
To lower the setpoint for initiation of the ECCS could lead to a loss of effective core cooling.
To raise the ECCS initiation setpoint wculd be in a safe direction, but it would reduce the margin established to prevent actuation of the ECCS during normal operation or during normally expected transients.
E.
Turbine Stop Valve Scram The turbine stop valve closure scram trip anticipates the pressure, neutron flux, and heat flux increase that could result from rapid closure of the turbine stop valves.
With a scram trip setting of 10% of valve closure from full open, the resultant increase in surface heat flux is limited such that MCPR remains above the MCPR fuel cladding integrity safety limit even during the worst-case transient that assumes the turbine bypass is closed.
}
F.
Turbine Control Valve Fast Closure Scram The turbine control valve fest closure scram is provided to anticipate the rapid increase in pressure and neutron flux resulting from fast closure of the turbine control valves due to a load rejection and subsecuent failure of the bypass, i.e.,
than the MCPR fuel cladding integrity safety limit for this transient.it prevents MCPR from becoming less 1.1/2.1-9
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QUAD-CIIIES DPR-30 G.
Reactor Conlant low Prewure Initiates Main Steam isolation Valve Closure The low-pressure isolation at 850 psig was provided to give protection against fast reactor depres-surization and the resulting rapid cooldown of the v ssel. Advantage was taken of the scram feature which occurs in the Run mode when the main steamline isolation valves are closed to provide for reactor shutdown so that operation at pressures loher than those specified in the thermal hydraulic safety limit does not occur, although operation at a pressure low than ItSO psig would not necessarily constitute an unsafe condition.
- 11. Main Steamline Isolation to Yahe Closure Scram The low-pressure isol3 tion of the main steamlines at 850 psig was provided to give protection against rapid reactor depressurization and the resulting rapid cooldown of the vessel. Advantage was taken of the scram feature in the Run mode which occurs when the main steamline isolation valves are closed to provide for reactor shutdown so that high power operation at low reactor pressures does not occur, thus providing protection for the fuel cladding integrity safety limit. Operation of the reactor at pressures lower than 850 psig requires that the reactor mode switch be in the Startup position, where protection of the fuel cladding integrity safety limit is provided by the IRM and APRM high neutron flux scrams.
Thus. the combination of main steamline low-pressure isolation and isolation valve closure scram in the Run mode assures the availability of neutron llux scram protection over the entire range of applicabilXy of the fuel cladding integrity safety limit. In addition, the isolation valve closure scram in the Run mode anticipates the pressure and flux transients which occur during normal or inadvertent isolation valve closure. With the scrams set at 10% valve closure in the Run mode, there is no increase in neutron
{
flux.
I.
Turbine EliC Control Fluid 1.ow-Pressure Scram The turbine EllC control system operates using high-pressure oil. There are several points in this oil system where a lost of oil pressure could result in a fast closure of the turbine control valves. This fast closure of the turbine control valves is not protected by the turbine control valve fast closure scram. since failure of the oil system would not result in the fast closure solenoid valves being actuated. For a turbine control valve fast closure. the core would be protected by the APRM and high-reactor pressure scrams.
Ilowever, to provide the same margins as provided for the generator load rejection on fast closure of the turbine control valves. a scram has been added to the reactor protection system which senses failure of control oil preuure to the turbine control system. This is an anticipatory scram and results in reactor shutdown before any signi0 cant increase in neutron flux occurs. The transient response is very similar to that resulting from the turbine control valve fast closure scram. The scram setpoint of 900 psig is set high enough to provide the necessary anticipatory function and low enough to minimize the number of spurious scrams. Normal operating pressure for this system is 1250 psig. Finally, the control valves will not start until the fluid pressure is 600 psig. Therefore, the scram occurs well before valve closure begins.
J.
Condenser 1.ow Vacuum Scram Loss of condenser vacuum occurs when the condenser can no longer handle the heat input. Loss of condenser vacuum initiates a closure of the turbine stop valves and turbine bypass valves which eliminates the heat input to the condenser. Closure of the turbine stop and bypass valves causes a pieisure transient, neutron flux rise. and an increase in surface heat flux.To prevent the cladding safety limit from being exceeded if this occurs, a reactor scram occurs on turbine stop valve closure in the Run mode. The turbine stop valve closure scram function alone is adequate to prevent the cladding safety limit from being exceeded in the event of a turbine trip transient with bypass closure.
l The condenser low vacuum scram is anticipatory to the stop valve closure scram and causes a scram i
N before the stop valves are closed and thus the resulting transient is less severe. Scram occurs in the Run mode at 23.iuch IIg vacuum stop valve closure occurs at 20-inch lig vacuum, and bypass closure at 7-inch lig vacuum.
1.1/2.1-10
- -- - -... -......- _. f..
_, _ __g 1
i QUAD-CITIF.S DPR-30 M
-f r
References 1.
" Generic Reload Fuel Application," NEDE-24011-P-A*
I r.2 1
- Approved revision number at time reload analyses are performed i
4 1
1 i
4 0
I i.
i I
f j%
u 1
I l
i -
j sin. si i
Q.;
Figure 2.1-2 has been deleted
i
(
QtfAD-CITIES DPR-30 1.2/2.1 REACTOR COOLANT SYSTEM SAFETY LIMIT LIMITING SAFETY SYSTEM SE1 TING Applicability:
Applicability:
Applies to limits on reactor coolant system Applies to trip settings of the instruments and pressure.
devices which are provided to prevent the reactor system safety limits from being exceeded.
Objective:
Objective:
To establish a limit below which the integrity of the To define the level of the process variables at which reactor coolant system is not threatened due to an automatic protective action is initiated to prevent overpressure condition.
the safety limits from being exceeded.
SPECIFICATIONS A.
The reactor coolant system pressure shall not A.
Reactor coolant high. pressure scram shall be exceed 1325 psig at any time when irradiated s1060 psig.
]
fuel is present in the reactor vessel.
B.
Primary system safety valve nominal settings shall be as follows:
I valve at 1115psig'"
l 2 valves at 1240 psig 2 valves at 1250 psig 4 valves at 1260 psig orTarget Rc,ck combination safety / relief valve The allowable setpoint error for each valve shall be i1%.
e 1.2 / 2.2 - I
(
s
\\
\\
QUAD-Cll ti:S g
DPM-30 t
O 1.2 sal 1.TY 1.1%11T !!ASI:S The scador toolant system integrity is an important
- arrier in the presentinn of uncontrolled re! case of fmion prc. ducts la is cucntial that the interrsty of this systern be protetted by estabhshing a preuure hmit to be observed for all operating conditions.ind wheneser there is irradiated fucl in the reactor vessel The prc9ure safety lunit of 1325 psig as measured by the vessel stcam space pressure indicator is equivalent to 1375 psig at the kmcst elevation of the reactor coolant systein. ~lhe 1175 p.ig value is derived frora the design pressures of the scauor prenore vesici and coolant system piping. The sc pective deqn preuures are !?50 png at 575* F and !!75 psig at 500' I2. T he prenure safety linnt was chosen as.'te lou er of fl.e prenure transients permitted by the appheah!c dcugn todes ASMI. Boiler and Pressure Vessel Code Section !!! for the pecuure s cuel, ar.d USASI101.1 Code for the rc.ictor coolant sysiem piping The ASMl: lbi!er and Pn worc Vessel Code per mits pressure tranuents up to 10'.i over dedgn pressure (!;0"o x 1250 = 1375 psip, and the 1;hASI Code permits pressure tranuents up to 20';our the design pressure t 120"* x 1175 = 1410 psip) 1he s.ifety limit pressure o:
1375 psig is refcsenced to the lowesi elevation of the pr. mary coolant system. Evaluation performed to assure that this safety limit pressure is not exceeded for any reload is documented in Reference 1.
The design bads for the reactor pressure vem! rnales evident the substantial margin of prmection against failure at the safety pressure hmit of 1375 pug. 'lhe vessel has becn designed for a gener.il mernbrane stress no greater than 26.700 psi at an internal preuure of !? 50 psig; this is a factar of 1.5 below t!.. yield strength of 40.100 psi at $75
- F. At the pressure inmt of I 3 75 psig, the general rnembrane stren wn!! only t>c 29,400 psi, still safel;. below the yield strength.
lhe relationships of stress Icveh to yield strength are comparable for the primary system piping and provide a similar mut n of protection at the estabbshed safety pressme Lmit.
i lhe rannat upcrating p:cssure of the scactor coolant system is IWO psip. l'or de turbi e in orlost of cI ctiicallaad s
transients, the turbine trip scram oi generator load rejection sciam together with the turhs bypass system hmits the pressure to approximately 1100 pse (References 2,3 andl ). E adation. rressme retref vahes have been rioviico to 4
reduce the probabdity of the safety ahes opeutmg m t.he event that the tuibme bypass should ta;!.
Teisliy, the safety va!ves are sired to keep the scactor coolant system preur, t etow 1375 rsig with re credit taken for retef u!ves dunng the postulated full closaic of all MSIVs without ducct N1.s position switch) scram. Credit is taken lor the rieutron flux scram, however.
l lhe indirect flux s: ram and safety vahe actuation, provide adequate margin l
below the peak a!!owable vewi pressure of 1375 ps:g.
Reictor preuuic is continuously monitored in the control room during operation on a 1500 psi full scale pressure recorder.
Refen necs
'I 1
" Generic Reload Fuel Application", NEDE-240ll-P-A*
2 SAR, Section 11.22 3.
Quad Cities 1 Nuclear Power Station firr.t reload license submittal, Section 6.2.4.2, February 1974.
4.
GE Topical Report NEDO-20693, General Electric Boiling Water Reactor No. 1 licensing submittal for Quad Cities Nuclear Power Station Unit 2, December 1974 Approved revision number at time reload analyses are performed.
1.2/2.2-2 i
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I QUAD CITIES DPR-30 2.2 LIMITING SAFETY SYSTEM SETTING BASES In compliance with Section III of the ASME Code, the safety valves mus t be set to open at no higher than 103% of design pressure, and they mus t limit the reactor pressure to no more than 110% of design pressure.
Both the high neutron flux scram and safety valve actuation are required to prevent overpressurizing the reactor pressure vessel and thus exceeding the pressure safety limit.
The pressure scram is available as backup protection to the high flux scram.
Analyses are performed for each reload to assure that the pressure safety limit is not er.ceeded.
If the high flux scram were to fail, a high-pressure scram would occur at 1060 psig.
1.2/2.2-3
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I QUAI)-Cll ti?S IIPit.to r..m GJ 3.1/4.1 ItEACIOlt PROTECflON SYSTICM l.f MillNG CONIH flONS IOR OPfit AllON SURVEll.l.ANCI: Iti'Qriltall:N1S Applicalaht,u Applicabiht):
Applies to lhe instrumentation and.mrelated de-Applies to the surveillante of the instrumentation uces s hhh initiate a reartnr cram.
and associatcJ deuces w hish initiate reactor scram.
Objectisc:
Objecth e:
T.e awure the opetehility of the reaum prowtion To specify the type and frequency af surveillante to spiem.
be applied to the protection instrumentation.
SPECIFICATIONS t (br setroinis. minimum number of trip sys.
A.
Instrumentation ytems shall be functior. ally scms, and minimurn number of instrument tested and calibrated as indicated in Tables channels slut must be operchie for each pnsi-41 1 and 4 I 2 respectisely C. g tion of the reactor mode switch shall be as f
given in Tables 3.I-I through 3.1-4. The system B.
Daily during teactos power operation. the core response times from the openin; of the sensor power distribution shalt le chetted for maximum mntact up to and including the openin; of the fraction of 1imiting. power dens-trip actuator contaus shall not execed 100 ity (MFLPD) and compared with the 3 D d 5-fraction of rated power (FRP)
B* If, during operation, the maximum when operating above 25% rated fraction of limiting power dens-thermal power, ity exceeds the fraction of rated power when operating above 25%
rated thermal power, either:
C.
}Vhen it is ileteimined thJ1 J channel is failed in the unsafe cenditun and Co!urnn I of Ta-
- 1. the APRM scram and rod bles 3.I.1 through 3.13 unnnt be met. that block settings sha11 be trip system must be put in the inpped o,ndition reduced to the values immediately. All other R PS ch.tnneh that mon-given by the equations ator the same variable shall be functionally i
in Specifications 2.l.A.1 tested within R hours. The trip systern with the fa led t-hannel may be untripped for a period of and 2.1.B.
This may be time not to exceed I hout in conduct this i
accomplished by increas-testing. As long as the trip system with the ing the APRM gain and failed thannel (nntains at ! cast one operable g
thus reducing the slope channel monitoring that same vanable. that and intercept point of trip system may be placed in the untripped the flow reference APRM position for short periods of time to allow scram and rod block set-functional testing of all RPS instrument chan-
- 3 tings by the reciprocal nets as specified by Tat!e 4.1.I.The trip ivstem
~
of the APRM gain change, may be in the untripped position for no more than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> per functior.al tot perical for this 2
the power distribution
U "I' chall be changed such g
that the maximum fraction of limiting power density no longer exct:eds the fraction of rate.d power.
3.1/4.1-1
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QUAD-CITIES DPR-30 gallons. As indicated above, there is sufficient volume in the piping to accommodate the scram without impairment of the scram times or amount ofinscrtion of the control rods.This function shuts the reactor down while sufficient volume remains to accommc.date the discharged water and precludes the situation in which a scram would be required but not be able to perform its function adequately.
Loss of condenser vacuum occurs when the condenser can no longer handle heat input. Loss of condenser vacuum initiates a closure of the turbine stop valves and turbine hypass valves, which eliminates the heat input to the condenser. Closure of the turbine stop and bypass valves causes a pressure transient, neutron flux rise, and an increase in surface heat slut To prevent the cladding safety limit from being exceeded if this occurs, a reactor scram occurs on turbine stop valve closure. The turbine stop valve closure scram function alone is adequate to prevent the cladding safety limit from being exceeded in the event of a turbine trip transient with bypass closure The condenser low-vacuum scram is a backup to the stop valve closure scram and causes a scram before the stop valves are closed, thus the resulting transient is less severe. Scram occurs at 23 inches Hg vacuum, stop valve closure occurs at 20 inches Hg vacuum. and bypass closure at 7 inches Hg vacuum.
High radiation levels in the main steamline tunnel above that due to the normal nitrogen and oxygen radioactivity are an indication ofleaking fuel. A scram is initiated whenever such radiation level exceeds seven times normal background. The purpose of this scram is to reduce the source of such radiation to the extent necessary to prevent excessive turbine contamination. Discharge of excessive amounts of radioactivity to the site environs is prevented by the air ejector off. gas monitors, which cause an isolation of the main condenser off gas line provided the limit specified in Specification 3.8 is exceeded.
The main steamline isolation valve closure scram is set to scram when the isolation valves are 10% closed from full open. This scram anticipates the pressure and flux transient which would occur when the valves close. By scramming at this setting the resultant transient is insignificant A reactor mode switch is provided which actuates or hypasses the various scram functions appropriate to the O4 particular plant operating status (reference SAR Section 7.7.l.2). Whenever the reactor mode switc Refuel or Startup/ Hot Standby position, the turbine condenser low-vacuum scram and main steamline isolation valve closure scram are bypassed. This bypass has been provided for ficxibility during startup and to allow repairs l
to be made to the turbine condenser. While this bypass is in effect, protection is provided against pressure or flux increases by the high-pressure scram and APRM 15% scram, respectively, which are effective in this mode.
If the reactor were brought to a hot standby condition for repairs to the turbine condenser, the main steamline isolation valves would be closed. No hypothesized single failure or single operator action in this mode of operation can result in an unreviewed radiological release.
The manual scram function is active in all modes, thus providing for a manual means of rapidly inserting control rods during all modes of reactor operation.
The IRM system provides protection against excessive power levels and short reactor periods in the startup and intermediate pcwer ranges (reference SAR Sections 7.4.4.2 and 7.4.4.3). A source range monitor (SRM) system is aho provided to supply additional neutron level information during startup but has no scram functions (reference SAR Section 7.4.3.2). Thus the IRM is required in the Refuel and Startup/ Hot Standby modes. In addition. protection is provided in this range by the APRM 15% scram as discussed in the bases for Specification 2.1. In the power range, the APRM system provides required protection (reference SAR Section 7.4.5.2).Thus, the IRM system is not required in the Run mode. the APRM 's cover only the intermediate and power range; the IRM's provide adequate coverage in the startup and intermediate range.
The high-reactor pressure, high.drywell pressure, reactor low water level, and scram discharge volume high level scrams are required for the Startup/ Hot Standby and Run modes of plant operation. They are therefore required to be operational for these modes of reactor operation.
f The turbine condenser low. vacuum scram is required only during power operation and must be bypassed to start up the unit.
3.1/ 4.1 -3 i
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QUAD-CITIES DPR-30 CS 4.1 SURVEII.I.ANCE REQUlilEMENTS tlASES A.
The minimum functional testing frequency used in this specification is based on a reliability analysis using the concepts developed in Reference 1. This concept was specifically adapted to the one-out-of.two taken twice logic of the reactor protection system. The analysis shows that the sensors are primarily responsible for the reliability of the reactor protection system.This analysis makes use of
- unsafe failure' rate experience at conventional and nuclear power plants in a reliability model for the system. An
- unsafe failure' is defined as one which negates channel operability and which. due to its nature, is revealed only when the channel is functionally tested or attempts to respond to a real signal. Failures such as blown fuses. ruptured bourdon tubes. faulted amplifiers. faulted cables, etc., which result in 'upscate' or
'downscale' readings on the reactor instrumentation are ' safe' and will be casily recognized by the operators during operation because they are revealed by an alarm or a scram.
The channels listed in Tables 4.1-1 and 4.1-2 are divided into three groups respecting fur.cdonal testing.
These are:
- 1. on-off sensors that provide a scram trip function (Group l);
- 2. analog devices coupled with bistable trips that provide a scram function (Group 2 ); and
- 3. devices which serve a useful function only during some restricted mcde of operation, such as Startup/ Hot Standby. Refuel, or Shutdown, or for which the only practical test is one that can be performed at shutdown (Group 3).
The sensors.that make up Group I are specifically selected from among the whole family ofindustrial on-off sensor; that have earned an excellent reputation for reliable operation. Actual history on this class
-O ef sensers everatint n nuclear rewer ciants s8ews feur faiiures in 422 senser reats. er faiiere rate ef i
0.97 x 10 */hr. During design. a goal of 0.99999 probability of success (at the 50% confidence level) was adopted to awure that a balanced and adequate design is achieved.The probability of success is primarily a function of the sensor failure rate and the test interval. A 3-mnnth test interval was planned for Group 1 sensors. This is in keepmg with good operating practice and satisfies the design goal for the logic configuration utilized in the reactor protection system.
To satisfy the long-term objective of maintaining an adequate level of safety throughout the plant lifetime, a minimum goal of 0.9099 at the 95% confidence level is proposed. With the one-out-of.two taken twice logic, this requires that each sensor have an availability of 0.993 at the 95% confidence level.
This level of availability may be maintained by adjusting the test interval as a function of the observed failure history (Reference 1). To facilitate the implementation of this technique, Figure 4.1 1 is provided l
to indicate an appropria:e trend in test interval. The procedure is as follows:
1.
Like sensors are pooled into one group for the purpose of data acquisition.
- 2. The factor M is the exposure hours and is equal to the number of sensors in a group, n, times the elapsed time T (M = nT).
- 3. The accumulated number of unsafe failures is plotted as an ordinate against M as an abscissa on Figure 4.1-1.
4.
After a trend is established. the appropriate monthly test interval to satisfy the goal will be the test interval to the left of the plotted pointt
- 5. A test interval or I month will be used initially until a trend is established.
Group 2 devices utilize an analog sensor followed by an amplifier and a histable trip circuit. The sensor and amplifier are active components, and a failure is almost always accompanied by an alarm and an l
indication of the source of trouble. In the event of failure, repair or substitution can start immediately.
An 'as-is' failure is one that ' sticks' midscale and is not capable of going either up or down in response l
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s QUAD-CITIF.S DPR-30 SS switches, hence calibration is not applicable; i.e., the switch is either on or off. Hated on the above, no calibration is required for these instrument channels.
B.
The MFLPD shall be checked once per day to determine if the APRM scram requires adjustment. This may l
normally be done by checking the LPRM readings. TIP traces, or process computer calculations. Only a small number of control rods are moved daily, thus the peaking factors are not expected to change significantly and a daily check of the MFLPD is adequate.
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References 1.
I. M. Jacobs.* Reliability of Engineered Safety Features as a Function of Testing Frequency.' Nuc/ car Safety, Vol. 9. No. 4. pp. 310-312. July-August 1968.
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OUAD-CITIES DPR-30 12 IIMITING CONDl'IlO% i OR Ol'8 at ATION !! Wi$
In addition to reactor protection inurumentation s hich initiates a reastor scram. proteoise in,trumentation ha been provided whish imiiates action to mitirate the con equences of accidenn which aic beyond the operator's -
abihty to conitol. or tea mmatn operaior etrun t< fore they result in serious consequerrn. 'l hiuct of spesnications provides the hmiting conditions ofoperation for the primary system isolation function. iniliation of the cmerrens y 3
core coohnt. system, control rod blocL. and 4andl.y ras treatment systems. 'Ihe objecle of the specifications are I
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(I) to sissure the efTectiseneu of the proicctne imiromentation when requirci by preserving its capability to o
tolerate a sing!c failure of any component of such sysicms o en during periods m hen portions of sus h syuenn are out of service for maintenance, and (D in pressribe the trip settings required to anure adcquate performance.
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When necessasy one thannel may be made inoperabic for briefintervals to conduct req tired functional teu and i' )
whbrations Some of the settings on the instrumentation that initiates or controls core.ind containment ciohng hase tolerances emphritly stated where the hi;h and low vatun are both critical and may have a sub tantial ctlect i
l on safety. It should be noted that the setpoints of othen inurumentatioc where only the high or low end of the i
setting has a direct hearing on safety, are thosen at a level away from the normal operating r.inye to present inadscrtent actuation of the safety systern involved and esposure to abnormal situations.
Isolation valves are in talled in those lino that penetraic the primary containment.ind muu be iw'ated during j
a loss-of-coolant auident w that the rad' avion do>c hmits are not esteeded during an an ident conditian. Actaation i
of thoe val.cs is initiateif hy the protective imiturnent ition whish e.oss the omJetiom ti.t u hu h iwlation t-required (this inorumentation iuhou n in Ianle 3.2 1 t Such inutumentation mus he avadable w henoct prunary containment interrisy is r+ ired The objectise is to holate the primary conta ment so that the ruidehno of b
10 Cl R 100 are not eseccJed during an anident 7
p" discuoion given in the hasm for Specification 3.1 is applicable here.The instrumendia.n which The low-reactor water level instrumentation is set to trip at >8 inches on tlE$ level instrument (top of active fuel is defined to be 360 inches above vessel rero) and aftar allowing for the full power pressure drop acrosa the steam dryer the low level trip is inches above vessel rero, or 144 inches above the top of at 504 active fuel. Retrofit 8x8 ; fuel has an active fuel length 1.24 inches longer than earlier fuel designs. However, prcsont trip setpointo were used in the LOCA analyses (NEDO-24146A, April 1979).
This trip initiates closure of Group 2 and 3 primary containment isolation valves but does not trip the recirculation pumps (refer-ence SAR Section 7.7.2).
For a trip' setting of 504 inches above vessel zero (144 inches above top of active fuel) and a 60-second j
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valvo 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 icvel instrumentation is set to trip when reac-tor water 3evel is 44.4. inches above vossol zero (with top of active fuel' defined as 360 inches above vessel zero,
-59 inches is 84 inchds above the top of active fue11. This trip initiates slosure of Group I primary containment isolati
%dvn (reference SAR Settion 7.7.2.2) and aho attivaics the ITC subsystems l
, starts the emergency die.cl rencrator, anJ trips the ruircul.ition pumps. lhim trip setimg lesel was chosen to be high enosyh in proent I
spurious operation but 10.
i enough to initiate LCCS operation.ind primary system isolatinn $n that no meltin; oithe fuel claddmg w tilmur and so that postaccident uvhng sun be accomplist.sJ und ti:e guidefines of 10 CF R 110 wdl not be exceeded. For the complete cirtumferential break of a 28-in(h reetreadation line and with the trip setting risen abme. I C(.% imtutirn and prim.ity system isolation are initiated and in 6me to meet the ahme triteria and meets the above uiteria
. I he inurumentation aho covers the full spectrum of breab The high-drywell prcuore inurunnentation is a battup to the water level inurumentation and, in addition in initiating ECCS. it uusn nnlation of Group 2 i olation valvet l'ur the bre.6 diwuwed alcre, this instrument.e-tion willinitiate ECCS operation at about the ume time as the low low water levelinstrumentation: thus the resnh, if ven above are applicable here ahn Group 2 isol.ition uhes include the drywell vent, purge, and surnp iwlation ulves. liigh.drpell preuure attivates only these vahn because high drywelf pressure could occur as the roult c4 non4al'cty.telated causo such as not puerk the drysell air during startup. Total syvem isolation is not
. e iuhle for these sondi;inas, and only the who in Group 2 are required to close. The low law water lect j
m tetiument.ition initiato pruintion for tbc futbres trum oflowof. coolant accidents and causo a trip of Group Ig g esmary system iw?ar on ulvn.
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5 QUAD-CITIES DPR-30 1 '
\\inimi tube > are previded in the main steamlines as a means of measuring steam now and aho limiting the loss mass inventory from the vessel during a steamline breal accident. In addition to monitoring steam flow, or imirumentation o provided which cames a trip of Group i isolation valves. The primary function of the imtiumemation is to detect a break in the main steamline. thus only Group i valves are clcied. For the worst we accident. main steamline break outside the drywell. this trip setting of 120% of rated steam flow,in conjuncunn with the flow limiters and main steamline vaive closure, limits the mass inventory loss such that fuel is not uncovered, fact temperatures remain less than 1500* F, and release of radioactivity to the environs is wel! 5:!cw 10 CFR 100 guidelines (rcretence SAR Sections 14.2.3.9 and 14.2.3.10).
Temperature. monitoring instrumentation is provide I in the main steamline tunnel to detect leaks in this area.
Trips are provided on this instrumentation and when exceeded cause closure of Group I isolation valvcs. Its setting of 200* F is low enough to detect Icaks of the order of 5 to 10 gpm; thus it is capabic orcovering the entire spectrum of breaks. For large breaks, it is a backup to high. steam flow instrumentation discuwed above. and for small breaks with the resulting small release'of radioactivity, gives isolation before the guidelines of 10 Cl R 100 are exceeded.
High radiation monitors in the main steamline tunnel have been provided to detect gross fuel failure. This j
instrumentation causes closure of Group i valves, the only valves required to close for this accident. With the established setting of 7 times normal background and main steamline isolation valve closure, fission product release is limited so that 10 CFR 100 guidelines are not exceeded for this accident (reference SAR Section i 2.2. l.7 h Pressure instrumentation is provided which trips when main steamline pressure drops below S50 psig. A trip of this instrumentation results in closure of Group i isolation valves. In the Refuel and Startup/Ilot Standby mode-inis trip function is hypassed. This function is provided primarily to provide protection againsi a pressure regul not malfunction which would cause the control and/or bypass valve to open. With the dip set at S50 psig. inventory loss is limited so that fuel is not uncovered and peak cladding temperatures are much less than 1500' F, thus, there g3
() are no fission products available for release other than those in the reactor water (reference SAR Section i 1.2.3 y The RCIC and the llPCI high flow and temperature instrumentation are provided to detect a break in their respective piping. Tripping of this instrumer:tation results in actuation of the RCIC or ofIIPCI isolation valves Tripping logic for this function is the same as that for the main steamline isolation valves. thus all sensors are required to be operable or an a tripped condition to meet the single-failure crituu. I he irip settmp ol.'iN F and 3ts0% of design flow and valve closure tirne are such that core uncovery is prevented and fission product selease is within limits.
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The instrumentation which initiates ECCS action is arranged in a one-out-of-two taken twice logic circuit. Unlike the reactor scram circuits, however, there is one trip system associated with each function rather than the two trip systems in the reactor protection system. The single-failure criteria are met by virtue of the fact that redundant core m,hng functions are provided. e.g., sprays and automatic blowdown and high pressure coolant injection. Ibi-specitication requires that if a trip syitem becomes inoperable. the system which it activates is declared moperable.
For example,if the trip system for core spray A becomes inoperable. core spiav A is declar-d inoperable and the aut-of-service specification of Specification 3.5 govern. This specification preserses the cil'euisenew of the system witt. opect to the single-fadure crueria even during periods when maintenance or testing is being performed.
The control rod block functions are provided to prevent excessive control rod withduwal so that MCPR does not go below the MCPR Fuel Claddi,nct Integrity Safety Limit.
The trip logic for this function is one out of n; e.g., any trip on one of the six APRMi. eight IR M *s four SRM's will result in a rod block. The minimum instrument channel requirements assure su0icient instrumentation to assure that the single-failure criteria are met. The minimum imtrument channel requirements for the RBM may be reduced by one for a short period of time to allow for maintenance, testing, or calibration.
This time period is only-3% of the operating time in a month and does not significantly increase the risk of preventing an inadvertent control rod withdrawal.
3.2 / 4.2 -6
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. g OUAD-CITIES D pit-30 The APRM rod block function in flou blaced and prrrvents a nignificant reduction in tiCPR, enpecially during operation at reduced flow.
The APRM provides gross core protection, i.e.,
limits the grons withdrawl of control rods in the normal withdrawal sequence.
In the refuel and startup/ hot ntandby modes, the APRM rod block function is set at 12% of rated power.
This control rod block provides the same type of protection in the Refuel and Startup/Ilot Standby re is as the APRM flow-bianed rm] block dor c in the hun mode, i.e.,
pres nts control rod withdrawal bef ore a scram is r eached.
The RBM rod block function provides local protection of the core, i.e.,
the prevention of trannition boiling in a local rerrion of the core for a single rod uithdrawal error irom a limiting control rod pattern. The trip point is flow biased. The worst-cane single control rod withdrawal error is analyzed for each reload to assure that, with the specific trip settings, rod withdrawal in blocked before the MCPR reaches the fuel cladding integrity safety limit.
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Below 30% power, the worst-case withdrawal of a single control rod with-out rod block action will not violate tho fuel claddinr1 integrity safety limit. Thus the RDM rod block function in not required below this power level.
The IkM block function provides local as well as gross core protection.
The scaling arrangement is such that the trip setting is less than a factor of 10 above the indicated level.
Analysis of the worst-case accident results in rod block action before MCPR approaches the MCPR fuel cladding integrity safety limit.
A dov.nscale in6tian on an APR M or IR\\f is an mdicate m the instrument 1.as faded or is t.n:
In either case the inurumeni u s!! not respond to changes in inmrol rod motion. and the centrol rod motion is thus prevented. The doun<cale trips are set at 3/125 of fu:t wate.
The SRM rod bbck w ah F 10XPsand the dercuor r.oi f ully inscried assures that the SRM S are rot withdrawn from the cose prior to commenung rod waLJrawal foi star tup lhe stram dmharre v lume high u. tcr level rod blo:L prosides annunciation for opciator rciinn lhe alarm setpoint has been sclttled to pmvide cdnguate time o
to allow determination of the cause of level i icreve and concttisc action prior to automatic scram initiation.
For c!"cctive emergency core conhar for small pipe breats the llPCI system must function. since reactor preuure does not decrease rapidly enough m allow coher core spt.n or ! PCl to operr..e in tune lht autum.uic prsssure relief function is provided as a bad op to the llPCI in Lt.s iwnt the llPCI does not opera'c.1he airangement of the tripping contacts n such as to promic thn funsnon w hen ntcenary and minimite spurious opera 90n 1he inp seuints given in the specihcanon me adequate so murc the ahose critesia are met (refeiuw e SAR Net oon f. 2.(. 3) 1he spculication preserves the c!1ct tisencu of the sprem during periods of mainten.u.tc. Icstina. or (ahbeation and aho ininirniics the ink of inadvertent operan.m. i c, only one instrument (hannel out of w t vire.
lwu air cjutor out.n momton.uc pim ided and, when their imp point is rexhed. cause an isol.uion of the air l
cjestor on7as hue. hotatmn n ininated w ben te:h onn umens rc.n h their high trip pomt or one h.n an upscale trip and the ether a downw are top 1 here o a 15 nunuit Jclay l+tme the air ejector 00 7 6 avdation ialce n timed.
This delay is accounted for 1 9 the to nunute hohh p inne of the od'-gas before it is rc! cased to the (himney.
Hoth imtruments ate requited for uip but the smuurnent asc so designed that any imtrument fadurc gives a dnwnscale uip 't he inp stuinps of the mstrumenn arc set so that the chimney retc.nc sate hmit given in Specil.catinn 3 S A 2 is not cuceded l'our radiation rooniton are prosided in the reauer hudding sentitation ducts which initianc isolation of the reactor hudJmp and operation of the st.mdby r.n ocatment spiem The monitors are located in the reactor building, sentdation theo l he nip forse n a one-out of two for cat h set. and exb set can initiaic a trip independent of the other vt Any upole inp wdi tause she dtsurd xuon. Irtp settinr,s of 2 inR/hr for nu. niton in the ventdation duo.nc h.ned u;nn moi ant noa nul s ennLuion n.d uion and standby p.n ocairnent spicm operation so that the scnnlanon st.n i idenc ute imni rnen m Sp.uth atmu 3 f A 3 is not euccded.Two radunon moniton are providcJ on il e reinchn; tian w hu h untuie notanon of the reaoor building and opcunon of the standby
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gas uraiment spkun lhe inp lorit n ont-out-ot two liip scuing of luo mit/he ihr the mointon on the refuchng uoor.uc bascJ upon minutng nonnat u nidation notunm and standby r,n ocainient spton oper.uinn 12/-t? 7
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.,,o QUAD-CITIES DPR-30 so that none of the activity released during the refueling aa:ident leaves the reactor building via the normal ventilation stack but that all the activity is processed by the standby gas treatment system.
The instrumentation which is provided to monitor the postaccident condition is listed in Table 3.2-4. The instrumentation listed and the limiting conditions for operation on these systems ensure adequate monitoring of the containment following a loss-of. coolant accident. Information from this instrumentation will provide the operator with a detailed knowledge of the conditions resulting from the accident; based on this information he can make logical decisions regarding postaccident recovery.
The specifications allow for postaccident instrumentation to be out of service for a period of 7 days. This period is based on the fact that several diverse instruments are available for guiding the operator should an accident occur, on the low probability of an instrument being out of service and an accident occurring in the 7-day period, and
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cn engineering judgment.
. The normal supply of air for the control room ventilation system comes from outside the service building. In the event of an accident, this source of air may be required to be shut down to prevent high doses of radiation in the i
control room. Rather than provide this isolation function on a radiation monitor installed in the intake air duct, signals which indicate an accident 'i.e., high drywell pressure, low water level, main steamline high flow, or high radiation in the reactor building ventilation duct, will cause isolation of the intake air to the control room. The above trip signals result in immediate isolation of the control room ventilation system and thus minimize any radiation dose.
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QUAD-CITIES DPR-30 TABLE 3.21 INSTRUMENTAll0N THAT INITIATES PRIMARY CONTAINMENT ISOLAil0N FUNCil0NS Busimum Nomest et opwous a Tripped lastrument Chamael/13 lastruments Trip' Level Settlag Action
- 4 Reactor low water 08
>l44 inches above top of A
active fue?
4 Reactor low low water 284inchegabove tcp of A
active fuel 4
Hgh drywell pressurem s2 pspi A
16 Hth flow main steamline*
s120% of rated steam flow B
16 H(h temperature mah s200* F B
steamline tunnel 4
Hgh radiation main s7 x normal rated power B
steamline tunnef5) background 4
Low main steam pressure)
2850 pst B
4 Hgh flow RCIC steamline
$300% of rated steam flow C
16 RCIC twbine area hth s200*F C
temperatwe v
4 H(h flow HPCI steamfine s300% of rated steam flow D
16 HPCI area hth temperature s200* F D
Neles 1.
Whenever prunary contamment stegrity is requned. there shafl be two operable or tripped systems for each functen except he low pressure mam steam!me whch onfy need be avadabte a the Run position.
2.
Acten:If the kst column cannot be met for one of the trip systems, that trip system shaft be tngped.
If the kst column cannot be met for both inp systems, the appropriate actens hsted bebw shall be taken:
A.
hdete an orderly shutdown and have the reactor in Cold Shutdown conditen in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
8.
bitete an orderly load reduction and have reactor e Hot standby within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
C.
Close notation valves a RCIC system.
Q.
Cbse isolaten valves m HPCI subsystem.
1 Meed not be operable when prnnary contamment stegnty rs not requned.
4.
The malaten tnp signalis bypassed when the mode switch is e Refuelor Startup/ Hot Shutdown.
i Thts instrumentaten aise isolates the control room ventdaten system.
6.
This signal also automatcalh cbses the mechancal vacuum pun'p discharge lee isolation vetves.
- Top of active fuel is defined as 360" above vessel zero for all water levels used in the LOCA analysis (See Bases 3.2).
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TABLE 3.22 tilSTRUMENTATION THAT INiilATES OR CONTROLS THE CORE Alto CONTAINMENT COOLING SYSTEMS INaisum 11 ember of Operable er Tripped lostnaneet ghannel:W Trip Function Trip Level Setting Remerks 4
Reactor low low 284 inches ( + 4 inches /-0 inch)
- 1. In conjunction with low-reactor pressure water level above top of actrve fuel
- hitiates core spray and LPCI.
- 2. In conjunction with high-drywell presswe 120-second time delay and bw pressure core cooling interbck initiates auto blowdown.
- 4. Initiates startog of diesel generators.
45 High-drywell s2 psg
- 1. Initiates core spray, LPCI, HPCI. and pressure,m
- scis, m
- 2. In conjunction with low low water level, 120-second trne delay, and low-pressure core coonng interlock initiates auto blowdown.
A
- 3. Initiates starting of diesel generators.
- 4. Initiates isolation of control room ventilaton.
2 Reactor bw 300 pstsps350 psig
- 1. Permissive for opening core spray and LPCI pressure admission va!ves.
- 2. In conjunction with bw low reactor water level hitiates core spray and LPCI.
Containment spray Prevents inadvertent operation of containment hterbck spray during accident conditons.
2m 2/3 core height 22/3 core height 45 contahment 0.5 psigsps!.5 psig high pressure 2
Tcaer auto s120 seconds in conjunction with low low reactor water bbwdown level, high-drywell pressure. and low pressure core cooling interlock initiates auto bbw-down.
4 Low-pressure core 75 psigsps100 psig Defers APR actuation pendog confrmation of cocimg pump dis-low pressure core coolog system operation.
charge pressure 2
lindervoltage on N/A
- 1. Initiates starting of diesel generators, emergency buses
- 2. Permissive for startbg ECCs pumps.
- 3. Removes nonessential bads from buses.
- Top of active fuel.is defined as 360" above vessel zero for all water levels used in the LOCA analyses (See Bases 3.2).
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