ML20064F008
| ML20064F008 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 11/20/1978 |
| From: | Reed C COMMONWEALTH EDISON CO. |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7811270170 | |
| Download: ML20064F008 (109) | |
Text
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Accren Reply to: Post Office Box 767 Chicago, litinois 60690 November 20, 1978 Director of Nuclear Reactor Regulation U.S. Nuclear Regulatory Ccusaission Washington, DC 20555
Subject:
Quad-Cities Station Unit 1 Proposed Amendment to License and Appendix A, Technical Specifications, for Facility Operating License DPR-29 i
to Support Reload No. 4 A.
NRC Docket No. 50-254
.)
Dear Sir:
Pursuant to 10 CFR 50.59, Camaonwealth Ediso.n proposes to amend the License and Appendix A, Technical Specifications, to i
Facility Operating License No. DPR-29 to support core reload No. 4 at Quad-citics Station Unit 1.
These changes are identified in Enclosure I and are based on plant analyses sununarized in Enclosures II, III and IV.
l It is planned, for Reload No. 4, to load 192 8x8R fuel assemblics, 4 of which will have developmental featurds designed for resistance to pellet-clad interaction (Barrier Lead Test
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Aasamblies - BLTAs).
The primary reference for this Quad-Cities Unit 1 Reload 4 Cycle 5 licensing submittal is the General Electric Generic Reload Fuel Application (NEDE-24011).
This new licensing suhaittal format contains similar technical information as previous submittals while deleting all explanatory text.
The attached enclosures that support this reload l
suhaittal are identified as follows:
Enclosure II
" Supplemental Reload Licensing Submittal for Quad-Cities Nuclear Power Station Unit 1 Reload 4", NEDO-24145, 78HED283, September 1978; i
Enclosure III - " Quad-Cities Nuclear Power Station Unit 1 Reload 4 Supplemental Licensing Information for Barrier Lead Test j
Assemblies", NEDO-24147, 78NED285, September 1970; and Enclosurc IV
" Loss-of-Coolant Accident Analysis Report for Dre: don Units 3, i
3 and Quad-cities Units 1, 2 Uuclear Power Stations,* UEDO-24146, i
78HED2G4, September 1978.
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T NRC Docket Ns. 50-254 Direr.itor'of Nuclear Reactor Regulation November 20, 1978 Page 2 The significant changes in this transmittal include a) a new MCPR Safety Limit of 1.07 as a result of the flatter local power (and CPR) distribution of the 8x8R design which, in turn, adversely affects the transition boiling probability distribution, b) a new LTPF of 3.00 for 8x8R fuel as a result of the increased active fuel length (145.24")
i j
and addition of a second water rod, and c) new MAPLHGR curves which reflect the improved flooding characteristics during a LOCA of the 8x8R design (which includes two alternate flow' path holes drilled in the lower tie plate orifice nozzle.)
It should also be noted that a separate MCPR Limiting Condition of operation has been specified for the Barrier Lead Test Assemblies (BLTAs).
These four developmental bundles are virtually identical to the standard retrofit design (8DRB265-L) with the addition of a) pellet-clad buffer materials (two bundles have;
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copper barriers and two have zirconium liners),
b) two segmented rods per BLTA (each consisting of four segments with hafnia-yttria and pellets), and c) pre-pressurization (3 ata. Belium).
Although these four bundles are neutronically treated the same as other reload fuel, a conservative treatment of the effects of pre-pressurization has been used for transient heat transfer assumptions and hence, the more restrictive MCPR LCO.
NRC Docket No. 50-254 Director of Nuclear Reactor Regulation November 20, 1978 Page 3 A reanalysis of BCC8 performance (Enclosure IV) for the limiting break size LocA has resulted in relaxed MAPLBGR limits primarily due to the effects of drilled lower tie plates in the retrofit and BLTA reload fuel.
The sont notable feature of the new analysis is the utilisation of Duane Arnold as the lead plant for several ENR 3's with loop selection logic.
This approach allows units with s mi1ar ECCS characteristics to e
reference the full break spectrum analysis of the lead plant and necessitates detailed calculations for the 1p). ting break size only (in this case the Dah value of 4.18 ft.
It should also be noted that Enclosure IV a,sstanes only 156 drilled bundles which is conservative for the QC1 C5 reload of 192 drilled bundles.
Conclusion The safety impact of the new standard retrofit fuel
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design (which includes incorporation of two larger water rods, a refined enrichment distribution, and 6 inch regions of natural uranium at the bottom and top of the active fuel) has been generically evaluated and specifically evaluated for Quad-cities 1 Reload 4 in Enclosure II.
In all cases, the cumulatin effect of the design changes has resulted in improved margia to established safety limits.
The impact of loading four special test assemblies has also been found to have little safety significance based on the evaluation presented in Enclosure III.
Although the increase enrichment in the central axial fuel sone is expected to increase operating MAPLBGR values, margin to limits is expected to improve since the LOCA analysis, incorporating the effects of drilled lower tie plates, increases the MAPIRGR limits proportionally greater than the expected increase in operating MAPlaGR values.
These proposed changes have received on-site and off-site review and approval.
s NRC Docket No. 50-254 Director of Nuclear Reactor Regulation November 20, 1978 Page 4 Pursuant to 10 CFR 170, Conanonwealth Edison has determined that the proposed amendment is Class III.
As such, we have enclosed a fee remittarece in the amount of $4,000.00 j
For purposes of your schedule, the projected startup date for this unit is approximately 90 days from the date of this transmittal.
Three (3) signed originals and thirty-seven (37) copies of this transmittal are provided for your use.
Very truly yours, h r.s. A. A L h
dell Reed Assistant Vice-President Enclosures SUBSCRIBED and S3fO before me this,N g to
, day of lli / tl DI N L,1978
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Alotary Public/
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ENCLOSURE I QUAD-CITIES UNIT 1 e
E C DOCKET NO. 50-254 W
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s DPR-29
- s -
This license shall be deemed to contain end is subject to the 3.
conditions specified in the following commission regulations in 10 CFR Chapter 1: Part 20, Section 3o.34 of Part 30,
Section 40.41 of ;' art 40, fections 50.54 and 50.59 of Part 50, and section 70.32 of Part 70; is subject to all applicabic provisions of the Act and to the rules, regulations and orders of the Coeritosion now or hereaf ter in ef fect; and te subject to the additional conditions specified or incorporated below:
A.
Marth e Power f.evel Comewnwealth E.2t 4on is authorized to operate Quad-Cities Unit No. 1 ar. power levele not in excese of 2511 megawatts (thermal).
8 Technical Specifications The Technical Specificatioris contained in Appendices A and B, as revised through Amen @ent No.146, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the -
Technical Specifications.
C. Restrictions q_
Reactor power level shall be limited to maintain pressure margin to the safety valve set points during the worst case pressurization transient.
The magnitude of the power limitation, if any, and.the point in the cycle at which it shall bh applied is specified in the Reload No. 4 licensing submittal for Quad Cities Unit No.1 (NE00-24145).
(j Subsequent operation in the coastdown mode is permitted based on the Generic Relond Fuel
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Application (Pg.
5-9 of NEDE-240ll-A) and its 1
subsequent approval (D. G. Eisenhut to R. Gridley letter dated May 12, 1978).
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- 0. Equalizer Valve Restriction The valves in the equalt:er piping be.wcen the recirculati=a loops shall be closed at all times dt: ring reactor cperati:n.
4.
This license is effective as of the date of issuance, and shall, expire at midnight, February 15, 2007.
(
- elocures Appendices A and B--
FOR THE ATOMIC ENERCY CCNMIssIoN '
Technical Specificattens j
ate of Issuance December 14, 1972 8 //a,f.,/ m A. clambuseo. ftputy Directir f or Reac tor Prn,. c t s Directorate et Licensing
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QUAD-CITIES DPR-29 b,
D. Reactor Water Imel(Shutdown Condition) curve in Figure 2.12, at which point I
the actual peaking factor value shall be Whenever the reactor is ir. the shutdown condi.
tion with irradiated fuel in the reactor vessel.
the water level shall not be less than that corre-LTPF = 3.06 (7 x 7 fuel assemblies) sponding to 12 inches above the top of the 3.03 (8 x 8 fuel assemblies) active fuel when it is seated in the core.
3.00 (8x8R fuel assemblies) l
- 2. APRM Flux Scram Trip Setting (Re-fueling or Startup and Hot Standby Mode; When the reactor mode switch is in the Refuel or Startup Hot Standby posi-l tion, the APRM scram shall be set at less than or equal to 15"o of rated neutron flux.
3.
IRM Flux Scram Trip Setting The IRM flux scram s'etting 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 i
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 (.65W + 43)(LTPF/TPF)
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The definitions used above for the APRM scram trip apply.
C. Reactor low water level scrim setting shall~be 2 143 inches above the top of the active fuel at, normal operating conditions.
D. Reactor low water level ECCS initiation shall be 83 inches ( + 4 inches /-0 inch) above the top of the active fuel at normal operating conditions.
E. Turbine stop valve scram shall be s 10"o 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.
G. Main steamline isolation valve closure scram shat! b: s 10"e valve closure from full open.
H. Main steamline low-pressure initiation of main steamline isolation valve closure shall be 2 850 psig.
1.1/ 2.1-2
i QUAD-CITIES DPR-29 1.1 SAFETY LIMIT BASES j
The fuel cladding integrity limit is set such that no calculated fuel damage would occur as a result of an abnormal operational transient. Because fuel damage is not directly observable, a step-back approach is used to establish a safety limit such that the minimum critical power ratio (MCPR)is no less than LO7MCPR > 1.07erpresents l l
a conservative margin relative to the conditions required to maintain fuel cladding integrity.
The fuel cladding is one of the physical barriers which separate radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the protection system safety settings. While fission product migration from cladding perforation isjust as measurable as that from use-related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding sar ty e
limit is defined with margin to the conditions which would produce onset of transition boiling (MCPR of 1.0).
These conditions represent a significant depanure from the condition intended.by design for planned operation.
A.
Resetoe Precure > 800 psig and Core Flow > 10% of Rated Onset of transition boiling results in a decrease in heat transfer from the cladding and therefore elevated cladding temperature and the possibility of cladding failure. However, the existe.;e of critical power, or boiling transition, is not a directly observable parameter in an operating reactor. Therefore, the margin to boiling transition is calculated from plant operating parameters such as core power, core flow, C
feedwater temperature, and core power distribution. The margin for each fuel a,sembly is characterized by the critical power ratio (CPR), which is the ratio of the bundle power whici' would produce onset of transition boiling divided by the actual bundle power. The minimum value of this ratio for any bundle in the core is the minimum critical power ratio (MCPR). It is assumed that the plant operation is controlled to the nominal protective setpoints via the instrumented variables (Figure 2.13).
'The safety limit (MCPR of 1.07 has su#icient conservatism to assure that in [he event of an abnormal l operational transient initiated from the normal operating condition, more than 99.93 of the fuel rods in h.
the core are expected to avoid boiling transition. The margin between MCPR of 1.0 (onset of transition boiling) and the safety limit. L O71s derived from a detailed statistical analysis considering al uncertainties in monitoring the core operating state, including uncertainty in the boiling transition correlation (see e.g., Reference I ). Because the boiling transition correlation is based on a large quantity of fhll-scale data, there is a very high confidence that operation of a fuel assembly at the condition of MCPR = 1.07would not produce boiling transition.
However, if boiling transition were to occur, cladding perforation would not be expected. Cladding temperatures would increase to approximately 1100' F, which is below the perforation temperature of the cladding material. This has been verified by tests in the General Electric Test Reactor (GETR), where similar fuel operated above the critical heat flux for a significant period of time (30 minutes) without cladding perforation.
If reactor pressure should ever exceed 1400 psia during normal power operation (the limit of applicability of the boiling transition correlation 3. it would be assumed that the fuel cladding integrity safety limit has been violated.
In addition to the boiling transition limit (MCPR) operation is constrained to a maximum LHGR: 17.5 kw/n for 7 x 7 fael and 13.4 kw/ft for 8 s 8 fuel. Yhis constraint is established by Specifications 2.1.A.!
and 3.5.J. Specification 2.1.A.I established limiting total peaking factors (LTPF) which constrain LHGR's to the maximum values at 100"o power and established procedures for adjusting APRM scram 1.1/ 2.1-4 1
QUAD-CITIES DPR-29
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settings which maintain equivalent safety margins when the total peak factor (TPF) exceeds the LTPF.
,j SpeciAcation 3.5J established the LHGR maximum which cannot be exceeded under s'teady power l
operation.
b B.
Core Thermal Power Limit (Reactor Pres'ure<800 psia) s At pressures below 800 psia, the core elevation pressure drop (0 power, O flow) is greater than 4.56 psi.
At low powers and flows inis pressure oifferential is maintained in the bypass region of the cere. Since the pressure drop in the bypa:,s region is essentially all elevation head, the core pressure drop at low powers and flows will always te greater than 4.56 psi. An. !yses show that with a flow of 28 x 108 lb/hr bundle flow, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus the bundle flow with a 4.56-psi driving head will be greater than 28 x 10'Ib/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicste that the fuel assembly critical power at this flow is approximately 3.35 MWt. A. 25% of rated thermal power, the peak powered bundle would have to be operating at 3.86 times the average powered bundle in order to achieve this bundle power. Thus, a core thermal power limit of 25% for reactor pressures below 800 psia is conservative.
C Power Transient Durmg 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 seconds. 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 plant operating situations which have been analyzed in detail.
In addition, control rod scrams are such that for normal operating transients, the neutron flux transient
,O is terminated before a significant increase in surface heat fiux occurs. Scram times of each control rod are checked each refueling outage, and at least every 32 weeks,50% are checked to assure adequate inseration times. Exceeding a neutron flux scram setting and a failure of the control rods to reduce flux to less than the scram setting within 1.5 seconds does not necessarily imply that fu;l is damaged: however, for this specincation, a safety limit violation will be assumed any time a neutron flux scram setting is exceeded for longer than 1.5 seconds.
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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
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the most severe normal operating transients expected. These analyses show that even if the bypass system fails to operate, the design limit of MCPR = 1.07 is not exceeded. Thus, use of a 1.5-secon provides additional margin.
The computer provided has a sequence annunciation program which willindicate the sequence in which scrams occur, such as neutron tiux, pressure, ete This program also indicates when the scram setpoint is cleared. This will provide information on how long a scram condition exists and thus provide some measure of the energy added during a transient. Thus, computer information normally will be available for analyzing scrams: however, if the computer information should not be available for any scram analysis, Specilication 1.1.C.2 will be relied on to determine if a safety limit has been violated.
During periods when the reactor is shut down, consideration must also be given to water level requirements due to the effect of decay heat. If reactor water level should drop below the top of the active fbel during this time, the ability to cool tr:e re is reduced. This reduction in core-cooling capability could lead to elevated cladding ternperatures and cladding perforation.The core will be cooled sufficiently to prevent cladding melting should the water level be reduced to two-thirds the core eight. Establish.
ment of the safety limit at 12 inches above the top of the fuel provides adequate margin. This level will be continuously monitored whenever the recirculation pumps are not operating.
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1.1/2.1-5
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QUAD-CITIES DPR-29 I
An increase in the APRM scram :4 setting would decrease the margin present before the fuel cladding integrity safety limit is reached. The APRM scram trip setting was determined by an 4
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analysis of margins required to provide a reasonable range for maneuvering during operation.
Reducing this operating margin would increase the T-equency of spurious scrams, which have an
,l adverse effect on reac ar safety because of the resulting thermal stresses. Thus, the APRM scram trip setting was selectv '.'ecause it provides adequate margin for the fuel cladding inte 1
yet allows operating margin that reduces the possibility of unnecessary scrams.
The scram trip setting must be adjusted to ensure that the LHGR transient peak is not increased for any combination of TPF and reactor core thermal power. The scram setting is adjusted in accordance.
with the formula in Specification 2.1.A.1, when the maximum total peaking factor is greater than the i
limiting total peaking factor.
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 accommodate anticipated maneuvers associated with power plant
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startup. Effects ofincreasing 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 control rod patterns are constrained to be uniform by operating procedures backed up 4
by the rod worth minimizer. Of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power rise. Because the Hux distribution manannted with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change power by a significant 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 withdrawal approach to the scram level, the rate of power rise is no more than 5% of C
rated power per minute. and the APRM system would be more than adequate to assure a scram before the power could e/ceed the safety limit. The 15% APRM scram remains active until the mode i
switch is placed in the Run position. This switch occurs when reactor pressure is greater than 850 P8i -8
- 3. IRM Flux Scram Trip Setting The IRM system consists of eight chambers, four in each of the reactor # protection system logic channels. The IRM is a 5-decade instrument which covers the range of power level between that covered by the SRM and the APRM. The 5 decades are broken down into 10 ranges, each being i
one-half a decade in size.
l The IRM scram trip setting of 120 divisions is active in each range of the IRM. For example, if the l
instrument were on Range I, 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 accommodate the incre.~.se in power level, the scram trip setting is also ranged up.
The most significant sources of reactivity change during the power increase are due to control rod withdrawal. In order to ensure that the IRM provides adequate protection agaimt 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 severe case involves an initial condition in which the l
reactor is just subcritical and the IRM system is not yet on scale.
l Additional conservatism was taken in this analysis by assuming that the IRM channel closest to the withdrawn rod is bypassed. The results of this analysis show that the reactor is scrammed and peak power limited to 1% of rated power, thus maintaining MCPR above 1.07. Based on the above l
analysis, the IRM provides protection against local control rod withdrawat errors and continous withdrawal of control rods in sequence and provides backup protection for the APRM.
.1.1/ 2.1 -8
4 QUAD-CITIES 4
DPR-29 B.
APRM Rod Block Trip Setting Reactor power level may be varied by moving control rods or by varying the recirculation flow rate. The a
APRM system provides a control rod block to prevent roo withdrawal beyond a given point at constant recirculation flow rate to protect against the condition of an MCPR less than 1.0*L 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 withdrr.wal. The flow variable trip setting provides substantial margin from fuel damage, assuming a steady-state operat;on at the trip settint over the entire recirculation flow range. The margin to the safety limit increases as the flow decrease; for the specified trip setting versus flow relationship; therefore the worst-case MCPR which could occur during steady-state operation is at 108% of rated thermal power because of the APRM rod block trip setting. The i
actual power distribution in the core is established by specified control rod sequences and is monitored continuously by the incore LPRM system. As with the APRM scram trip setting, the APRM rod block trip setting is adjusted downward if the maximum total;seaking factor exceeds the limiting total peaking factor, thus preserving the APRM rod block safety margin.
(1 C.
Reactor Iew Water leel Scram The reactor low water level scram is set at a point which will assure that the wat'er level used in the bases for the safety limit is maintained. The scram setpoint is based on normal operating temperature ud pressure conditions because the level instrumentation is density compensated.
D.
Reactor Iew tew Water Level ECCS Initiation Trip Point The emergency core cooling subsystems are designed to provide sufficient cooling to the core to dissipate the energy associated with the loss-of-coolant accident and to limit fuel cladding temperature to well Il below the cladding melting temperature to assure that core geometry remains intact and to limit any cladding metal-water reaction to less than 1% To accomplish their intended function, the capacity ofesch 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 wa.ter level scrar6 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
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parameters: the maximum break size, 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 would be in a sa fe direction, but it would reduce the margin established to prevent actuation of the ECCS during normal operation or during normally expected transients.
j E.
Turbine Stop Vahe 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 1.07 evra during the worst-case transient that assumes the turbine bypass is closed.
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F.
Turbine Control Valve Fast Closure Scram The turbine control valve fast 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 Joad rejection and subsequent fa.ture of the bypass. i.e it prevents MCPR from becoming less than 1.07 for this transient.
For the load rejection from 100% power. the LHGR increases to only 106.5% ofits ts.ted value, which p,
results in only a smzll decrease in MCPR.
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1 QUAD-CITIES DPR-29 i O i
1 1.2/2.1 REACTOR COOLA!:T SYSTEM J
i 4
i SAFETY LIMIT LIMITING SAFETY SYSTEM SETTING 1
l 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 i
system safety limits from being exceeded.
Cbjective:
Objective:
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To establish a limit below which the integrity of the To define the level of the process variables at which reactor coolant system is act threatened due to an automatic protective action is initiated to prevent overpressure condition.
the safety limits from being exceeded.
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SPECIFICATIONS A.
The reactor coolant system pressure shall not A.
Reactor a>olant high-pressure scram shall be O
exceed 1325 psig as any time when irradiated s1060 psig.
fuel is present in the reactor vessel.
B.
Primary systcm safety valve nominal settings shall be as follows:
I valve at 1115 psig" 2 valves at 124ti psig C'-
' ' "' i 250 psig d
4 valves at 1260 psig
" Target Rock combination safety / relief valve The allowable setpoint error for each valve shall be i1%.
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1 1.2 / 2.2-1
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QUAD-CITIES DP' -29 R
l 4
Venturi tubes are provided in the main steamlinc. as a means of measuring steam ilow and also limiting the loss of mass inventory from the venel during a steamline break accident. In addition to monitoring steam flow, instrumentation is provided which causes a trip of Group I isolation valves. The primary function of the 1
instrumentation is to detect a break in the main steamline, thus only Group i valves are closed. For the worst-case accident, main steamline break outside the drywell, this trip setting of 120% of rated steam flow,in conjunction l
with the flow lirr.iters and main steamline valve closure, limits the mass inventory loss such that fuel is not uncovered, fuel temperatures remain less than 1500
- F, and release of radioactivity to the environs is well below 10 CFR 100 guidelines (reference SAR Sections 14.2.3.9 and 14.2.3.10).
Temperature monitoring instrumentation is provided in the main steamline tunnel to detect leaks in this area.
Trips are provided on this instrumentation and when exceeded cause closure of Group 1 isolation valves, its setting of 200
- F is low enough to detect leaks of the order of 5 to 10 gpm; thus it is capable of covering the entire spectrum of breaks. For large breaks, it is a backup to high-steam flow instrumentation discussed above, and for small breaks with the resulting small release of radioactivity, gives isolation before the guidelines of 10 CFR 100 are exceeded.
High rridiation monitors in the main steamline tunnel have been provided to detect gross fuel failure. This
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instrun.entation causes closure of Group i valves, the only valves required to close for this accident. With the establisred 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 12.2.1.7 ).
Pressure instrumentation is provided which trips when main steamline pressure drops below 850 psig. A trip of this instrumentation results in closure of Group i isolation valves. In the Refuel and Startup/ Hot Standby modes this trip function is bypassed. This function is provided primarily to provide protection against a pressure regulator malfunction which would cause the control and/or bypass valve to open. With the trip set at 850 psig, inventory O
loss is limited so that fuelis not uncovered and peak cladding temperatures are much less than 1500' F; thus, there are no fission products available for release other than those in the reactor water (reference SAR Section I l.2.3 ).
The RCIC and the HPCI high flow and temperature instrumentation are provided to detect a break in their respective pipin[t. Tripping of this instrumentation results in actuation of the RCIC or of'HPCI isolation valves.
Tripping logic for this function is the same as that for the main steamline isolation valvgs, thus all sensors are required to be operable or in a tripped condition to meet the single-failure criteria. The trip settings of 200
- F and p!
300% of design flow and valve closure time are such that core uncovery is prevented and fission product release is within L.t:
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 coo!ing functions are provided, e g, sprays and automatic blowdown and high-pressure cooiant injection. The specification requires that if a trip system becomes inoperable, the system which it activates is declared inoperable.
For example,if the trip system for core spray A becomes inoperable, core spray A is declared inoperable and the out-of-service specifications of Specification 3.5 govert This specification preserves the eff'ectiveness of the sptem with respect to the singinfailure criteria even during periods when maintenance or testing is being performed.
The control rod block functions are provided to prevent excessive control rod withdrawal so that MCPR does not approach I.07 The trip logic for this function is one out of n: e.g., any trip on one of the six APRM's, eight IRM's, four SRM's will result in a rod block. The minimum instrument channel requirements assure sufficient instrumentation to assure that the single-failure criteria are met. The minimum instrument 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.
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l
4 QUAD-CITIES DPR-29 The APRM rod block function is dow biased and preants a significant reduction in MCPR. especially during operation at reduced dow. The APRM provides gross core protection,i.e limits the gross core control rods in the l
normal withdrawal sequence. The trips are set so that MCPR is mainained greater than 1.07 The APRM rod block function, which is set at 12"o of rated power. is functionalin the Refuel and Startup/ Hot Standby modes. This control rod block provides the same type of protection in the Refuel and Startup/ Hot Standby modes as the APRM Mow-biased rod block does in the Run mode. ie.,it prevents MCPR from decreasing below I.07during control rod withdrawals and prevents control rod withdrawal before a c: ram is reached.
The RBM rod block function provides local protection of the core,i.e., the prevention of transition boiling in a local region of the core for a single rod withdrawal error from a limiting control rod pattern. The trip point is How biased. The worst-case single control rod withdrawal error has been analyzed, and the results show that with the specined trip settings, rod withdrawal is blocked before the MCPR reaches 1.07,thus allowing adequate margin (Reference 1).
Below 70% power, the worst-case withdrawal of a single control rod results in a MCPR greater than 1.07 without l
rod block action. Thus it is not required below this power level.
The IRM 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 a
block action before MCPR approaches 1.0 7.
A downscale indication on an APRM or IRM is an indication the instrument has failed or is not sensitive enough.
In either case the instrument will not respond to changes in control rod motion, and the contr51 rod motion is thus prevented. The downscale trips are set at 3/125 of full scale.
The SRM rod block with s 100 CPSand the detector not fully inserted assures that the SRM"s are not withdrawn from the core prior to commencing rod withdrawal for startup. The scram discharge solume high water level rod O
block provides annunciation for operator action. The alarm setpoint has been selected to provide adequate time to allow determination of the cause of Icvel increase and corrective action prior to automatic scram initiation.
For c'rective emergency core cooling for small pipe breaks, the HPCI system must function, since reactor pressure does not decrease rapidly enough to allow either core spray or LPCI to operate in time. The automatic pressure relief function is provided as a backup to the HPCI in the event the HPCI does not operate. The arrangement of the tripping contacts is such as to provide this function when necessary and minimize spuriou's operation. Th: trip settings given in the specification are adequate to assure the above criteria are met (reference SAR Section 6.2.6.3 ).
The specification preserves the effectiveness of the system during periods of maintenance, testing, or calibration and also minimizes the risk ofinadvertent operation, i.e., only one instrument channel out of service.
' Two air ejector off-gas monitors are provided and. when their trip point is reached, cause an isolation of the air ejector off gas line. Isolation is initiated when both instruments reach their high trip point or one has an upscai:
trip and the other a downscale trip. There is a ! $-minute delay before the air ejector off-gas isolation valve is closed.
This delay is accounted for by the 30 minute holdup time of the off gas before it is relea*cd to the chimney.
Both instruments are required for trip, but the instruments are so designed that any instrumen' failure gives a downscale trip. The trip settings of the instruments are set so that the chimney release rate hmit given in Specification 3.8.A.2 is not exceeded.
t l
Four radiation monitors are provided in the reactor building ventilation ducts which initiate isolation of the reactor building and operation of the standby gas treatment system. The momtors are located in the reactor building ventilation duct.The trip logic is a one-out-of-two for each set, and each set can initiate a trip independent l
of the other set. Any upscale trip will cause the desired action. Trip settings of 2 mR/hr for monitors in the i
ventilation duct are based upon initiating normal ventilation isolation and standby gas treatment system operation j
so that the ventilation stack release rate limit given in Specification 3.8.A.3 is not exceeded.Two radiation monitors are provided on the refuehng door which initiate isolation of the reactor building and operation of the standby
(..
gas treatment systems. The trip logic is one-out-of two. Trip settings of 100 mR/hr for the monitors on the refueling door are based upon initiating normal ventilation isolation and standby gas treatment system operation
^
l 3.2/ L2-7
QUAD-CITIES DPR-29 so that none of the activity released during the refueling accident leaves the reactor building via the normal i
ventilation stack but that all the activity is processed by the standby gas t.:stment 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 de -"~' knowledge of the conditions resulting from the accident; based on this information he can make logical deciaons regarding postaccident recovery.
The specifications a!!ow 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 probabili:y of an instrument being out of service and an accident occurring in the 7. day period, and on 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 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.
References 1.
GE Topical Report NEDO-24145, " General Electric Boil.ing Water
(
Reactor Reload No. 4 Licensing Submittal for Quad-Cities Nuclear Power Station (Unit 1) ", Section 6.3.3.2, September,1978.
l
.A 3.2 /.l.2 -8
QUAD-CITIF.S DPR-29
. f",
L.
TABLE 3.23 lil5TRUMENTATION THAT INiilATES ROD BLOCK Inimes Non6w of Operstde er Tripped lastrument m
lastrument Trip Leyel setting themasis per Trip $rstea 2
APRM tipscale (f!cw bias 773 so.650W + 43'2' 2
APRM upscale (Refuel and Startup/ Hot s12/125 fun scale Standby mode) 2 APRM downscale"'
23/125 full scale 1
Rod block mon: tor upscale (Acw btas7
50.650W + 42 (2) 1 Rod block rnonitor downscak*
23/125 fun scale 3
IRM downscale m is' 23/125 fun scale 3
IRM upscaW
s108/125 fun scale 2*
SRM detector not in Startup positat*
22 feet below crve center.
Ime 3
IRM detector net in Startup position'8' 22 feet below core center-line 2m
- SRM upscale s 105 counts /sec 2*
SRM downsca'e*
2 102 counts /sec 1
Heh water level in scram disctiarge vclurne s25 gallons Notes 1.
For tire startup/ Hot star:4by and R2n sositens of rne reactor mode selector s itch there shaft te tao operable or tne:ed tna systems t each functen neept the sRM rod blocks. IRM upsca'e aed WM deanwa'e nesd not be eteratie e the Run positen. APdM downscale. APRM vC5ca:e (flow biased) REM ersca'e. and R$M dow# scale seed not be or'ef ab4 m tPe startilp/4l star (by made af the tilst column cannot be m.' lor one of the two tf tp s6tems. this condsten mey es st 1Br 90 to 7 days pfovided that dureg tDat fune the caesabe s6 tem 4 mnCttnally tested imfrediately and dady thereafter, d th s Conditen lasts longer (Pan i deys the srstem shall be tripped if the best Column Canttet be met for both trip systems, the systems inall be tripped.
- 2., W e the seactar receeviatma loop now a pe< cent. Inp level settag is a perceit of rated power 82511 MWtt 1 IRM downscate may te boassed whetut is on its le*est range l
4.'
The functen is byrassed ahen the count rate e 2 t00 CPS.
5.
one of tre tour sRM routs may be bycassed i
& The sPM fumten may be bysassed a the highof IRM ranges trantes 8.1 and 10) enen Je IRM opscala rod bioch is operab;e.
l 1.
Not retened to be opef able enne peflormmg Icw power phncs te;ts at atmosphers pf essure dureg or af ter refueleg at poner levels not to exceed 5 M at.
L The IRM functen occus when the reactar mode switch is e the Refuti er startup/ Mot stancty positen.
1 The try a bypassed when t!!e SAM 4 fuh eserted.
t 3.2/.t.2-14 i
QUAD-CITIES DPR-29 l
1 c.
the operating power level shall be l
limited so thst the LICPR will
)
l remain above 1.07 assuming a sin-gle error that results in complete withdrawal of any single operable control rod.
C.
Scram Insertion Times C.
Scram insertion Times
- 1. The average scram insertion time, ba-
- 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 pu~mps.
% inserted from insertion Fully withdrawn Times (sec) 5 0.375
(].
20 0.900 V
50 2.00 90 3.50 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:
i
% inserted From Average Scram Fully Withdrawn Times (sec) t.
5 0.398 20 0.954 50 2.12 90 3.80 l
- 2. The maximum scram inwrtion time 2.
Following a controlled shutdown of for 90% insertion of any operable con-the reactor. but not more frequently l
trol rods shall not exceed 7 seconds.
than 16 weeks nor less frequently than
- 3. If Specincation 3.3.C.I cannot be met.
weef intervals. 50% of the control i
the reactor shall not be made super-r d drtves in each quadrant of the critical. if operating. the reactor shall react r e te shall be measured for the
[
be shut down immediately upon deter.
Scram times SPecined in Specincation
~
mination that average scram time is 3.3.C. All control rod drives shall have
(
de6cim experienced scram test measurements each year. Whenever all of the control l
4.
If Specincation 3.3.C.2 cannot be met, rod drive scram times have been mea-the dencient control rod shall be con-sured, an evaluation shall be made to
\\
3.3 /.t.3-4 I
l
'i QUAD-CITIES DPR-29 O
6.
d-er-c7 i d i>> d
=<re rr cca ero oos-t c.
a beginning-oflife Doppler reactivity feedback.
j
- d. the rod scram insertien rate shown in Specification 3.a.C, e.
the maximum possible rod drop velocity of 3.11 fps, j
f.
the design accident and scram reactivity shape function, and i
- g. the moderator temperature at which criticality occurs.
In most cases the wcs th ofinsequence rods or rod segments will be substantially less than 0.013 ak.
Further, the addition of 0.013 ak worth of reactivity, as a result of a rod drop and in conjunction with the actual values of the other important accident analysis parameters described above, would most likely result in a peak fuel enthalpy substantially less than 280 cal /g design limit. However, the 0.013 Ak lirr.it is applied in order to allow room for future reload changes and ease of verification without repetitive technical specification changes.
p 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 estiman f.o 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 differences.
~
The rod worth minimizer provides automatic supervision to assure that out of sequence control rods 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 ofservice when required a licensed operator
[
or other qualified technical employee can manually fulfill the control rod pattern conformance l
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 eScient reactor startup at low neutron 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 10-8 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 control r 4 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 red 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 o one of the channels will block erroneous rod withdrawal soon enough to prevent r
fuel damage. This system backs up the operator, who withdraws control rods according to a written l
sequence. The specified restrictions with one channel out of service conservatively assure that fuel l
damage will not occur due to rod withdrawal errors when this condition exists. During reactor I
operation with certain limiting control rod patterns, the withdrawal of a designated single control rod could result in one of more fuel rods with MCPR's less than 1.07.During use of such patterns, it isjudged 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 of the Nuclear Engineer to identify these limiting patterns and the designated rods either when the patterns are initially established or as th:y develop due to the occurrence ofinoperable control rods in other than limiting patterns.
4 3.343-9 l
'a a
t j
QUAD-CITIES
(
DPR-29 O
e 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 becoming less than 1.07.The limiting power transient is that i
i resulting from a turbine stop valve closure with failure of the turbine bypass system. Analysis of this i
transient shows that the negative reactivity rates resulting from the scram with the average response of i
all the drives as given in the above speci6 cation, provide the required protection, and MCPR remains greater than 1.0% Reference I shows the control rod scram reactivity used in analyzing the transients.
l Reference I should not be confuad with the total control rod worth,18% ak, as listed in some l
' tiendments to the SAR. The 18% ak value represents the amount of reactivity available for withdrawal in the cold clean core, whereas the control rod worths shown in Reference I represent the amount of reactivity available for insertion (scram) in the hot operating core. The minimum amount of reactivity to be ingrted during a scram is controlled by permitting no more than 10% cf the operable rods to have long scram times. In the analytical treatment of the transients. 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 flux 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 specined in Specifica, tion 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 provides reasonable assurance of detection of slow drives before system deterioration beyond the limits of Specification 3.3.C. The program was developed on the basis of the statistical approach outlined below and judgment.
The probability that the mean 90% insertion time of a sample of 25 control rod drists will not exceed 0.25 seconds of the mean of all drives is 0.99 at a risk of 0.0!. If the mean time exceeds this range or the
(
meaa 90% insertion time is greater than 3.5 seconds, an additional sample of drives will be measured to verify the mean performance.
Since the differences between the expected observed mean insertion time and the limit of Specification 3.3.C greatly exceed the expected range, this sampling technique gives assurance that the limits of Specification 3.3.C will not be 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 expected deviation from the mean and therefore gises assurance that the mean will not exceed the limit of Specification 3.3.C. In addition 50% orthe control rods will be checked every 16 weeks to verify the performance and for correlation with the sampling program.
The history of drive performance accumulated to date indicates that the 90% 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 variations and also pravide assurance that local scram time limits are not exceeded. Continued monitoring of other drives exceeding the expected range of scram times provides surveillance of possible anomalous performance.
~
The numerical values assigned to the predicted scram performance are based on the analysis of the Dresden 1 startup data and of data from other BWR's such as Nine Mile Point and Oyster Creek.
3.3AL3-10 i
- s e
QUAD-CITIES DPR-29
<C i
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 />. Surveil-lance and corresponding action shall continue until reactor operation is within the prescribed limits.
LHG R,,,
LHGR,1 -( AP/P),,,(L/L )
r
- LHGR,
- design LHGR where-
- 17.5 kW/A, 7 x 7 fuel assemblies
- 13.4 kW/ft,8 x 8R fuel assemblies 8x8 (a P/ P),,,
- maximum power spiking penalty
.035 initial core fuel
.029 reload I, 7 x 7 fuel
.022 reload, 8 x 8 fuel
=
MO eko'id"N#x hk' ht$el assemblies
=
l L,
total core length l
12 feet v
L Axial distance from bottom of core K.
Minimum Critical Power Ratio (MCPR)
K.
Minimum Critical Power Rat,io (MCPR)
During steady state operation MCPR shall be The MCPR shall be determiped daily during greater than or equal to steady-state power operatio~n above 25% of O
"'""" *"~
1,,,,,,,,,,,,
l.29 (8 x 8 fuel) er and(flow. lf at any tithe during at rated pow.32 1
8 x 8 BLTA) operation it is determined by r.ormal surveil-lance that the limiting value for MCPR is oeing exceeded, action shall be initiated within 15 minutes to restore operation to within the pre-scribed limits. If the steady state MCPR is not returned to within the prescribed limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the reactor shall be brought to the cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveil-l lance and corresponding action shall continue until reactor operation is within the prescribed limits. For core flows other than rated, these nominal values of MCPR shall be increased by a factor of k,, where k,is as shown in Figure
^
3.5-2.
1
{
3.5/.t.5-10 l
0 QUAD-CITIES DPM-79 i
1 5
3.5 IlMITING CONDITION FOR OPERATION BASES A. Coce Spray and LPCI Mode of the RHR System This specification assures that adequate emergency cooling capability is available whenever irradi ted fuel is in the reactor vessel.
Besed en the loes-of-coolant analytteal methode deseratacd an oeneral electrie T6pical Report NEco-20s66 and the specirte snelyste in tit:Do 24146, "t.oss-or-coolant Analysts fienort ror DresdenUnite2.3andound-eNt es Units 1, 2 Nuclear Power StationesSeptember 19 cor. coott,., systems provide surrictent cooling to the core to dise1pate the energv eseactated with the loss-or-coolant acc id e nt, to liantt caseeleg,a fuel eladding temperature to less than 220oo, to assure that r
eore geometry remains intact, to limit cladding metal-water reaction to less than 1%, and to li. nit the calculated local motel-water reaction to less than 17 The limiting conditions of operation in Specifica' ions 3.$Al through 3.5A6 specify the combinations of operable subsystems to assure the availability of the minimum cooling systams noted above. No single failure of ECCS equipment occurring during a loss-of-coolant accident under these limiteng conditions of operation will result in inadequate cooling of the reactor core.
Core spray distribution has been shown. in full scale tests of systems similar in design to that of Quad Cities 1 and 2. to exceed the minimum equirements by at least 2M. In addition, cooling effectiveness has been demonstrated at Icw th.in half the rated flow in simulated fuct awembhes with heater rods to durlicate the decay he.it charaiternno of irraduied fuel. the amdent analwn n additionally conservatae m that no cred.: n taken for sprsy coolitig OL t.he reactior core before the internal pressure has fallen to 90 psig.
The LPCI mode of the RHR system is designed to provide emergency cooling to the core by flooding in the event of a loss-of coolant accident. This system functions in combination with the core spray system -
to prevent excessive fuel cladding temperature. The LPCI mode of the RHR system in combination with the core spray subsystem provides adequate cooling for break areas of approximately 0.2 ft up to and h,/
3 l
including 4.38 ft. the latter being the doub'e. ended recirculation line break with the equalizer hne 2
between the recirculation loopseloaedwithout assistance from the high-pressure emergency core ooling subsystems.
The allowable repair times are established so ti:4r the average risk rate for repair would be no greater than the basic risk rate. The method and concept are described in Reference 1. Using the results deseloped in this reference, the repatr period is found to l'e few than half the test interut. This assumes that the core spray subsystems and LPCI constitute a one out of two splem. ho*eser. the combined ellect of the two systems so limit escewne (laddmg temperatute must.slso be cuoudered. the test meersal specmed m Spectrication 15 was 3 months t herefore, an allowable repair period which maintains the basic r:A considering single l'adures should be few than 30 days. and thn specificauon is within thn period. For multiple failures, a shorter intetul is specined, to unprove the assurance that the remaining sp: ems wdl function. a daily test is called for. Although it is recognized that the mformation gaen in Reference i provides a quantitative method to esumate allowable repair times, the lack of opcating data to support the analytical approach prevents complete accept.ince of this method at this time. Therefore, the times stated in the specific items were estabhshed with due regard to judgment.
Should one core spray subsystem become inoperable. the remaining core spray subsystem and the entire L*Cl mode of the RHR sprem. ire avadable should the need for core cooling arise. To assure that the remaining core spray, the LPCI mode os the RHR sy> tem, and the dicsci generators are availaele. they r
are demonstrated to be operable immediately. This demonstrat:on includes a manual inittation of the s
i pumps and associated vah es and diesel generators. Based on judgments of the reliability of the remaining systems, i.e, the core spray and LPCI. a 7-day repatt period was obtained.
3.5/4.5-11
QUAD-CJTIES DPR-29 Should the kws of one RilR pump occur.: nearly full complement of co.c and enntainment cochng equipment is available. Three RllR pumps in coupncinm with the core spray subsystem will perform the core cooling function. Because of the avastahihty of the majority of the core coohng equipment. wheh will be demonstrated to be operable, a 30 day repair period is pstified. If the LPCI mode of the RilR system is not avail.shle. at least two RHR pumps must he avaitante to futhl! the containment cochng Anaction. The 7 day repair period is set on this basis.
B.
RHR Service Water The containment cooling mode of the RHR system is provided to remove heat energy from the containment in the event of a loss-of-coolant accident. For the flow specified, the containment long-terrr.
pressan is limited to less than 8 psig and is therefore more than ample to provide the required heat-removal capability (reference SAR Section 5.2.3.2).
De containment cooling mode of the RHR system consists of two loops, each containing two RIIR service water pumps, one heat exchanger. two RHR pumps. and the associated valves, piping. electrica!
equipment, and instrumentation. Either set of couipment is capable of performing the containment cooling function. Low of one RilR service water pump does not seriously jeopirdire the contamment p
cooling capability, as any one of the remaining three pumps can sainfy the coohng requiremenn Since
(,7 there is some redund.incy left a 30 Jay rep.ur peraxi is adequate Lou of one kiop of the containment cuoling mode of the RHR sprem leaves one remaining system to perform the containment cooling Ametion. The operable system is demonstrated to be operable each day when the above condition occurs.
Based on the fact that when one loop of the contamment cooling mode of the RHR system becomes inoperable, only one system remains, which is tested daily, a 7-day repair period was specified.
e C.
High.Precure Coulant Injection The high pressure coolant injection subsystem is provided to adequately cool the core foe all pipe breaks smaller than those for which the LPCI mode of the RHR system or core spray subsystems can protect the core.
t i
The HPC1 meets this requirement withcut the use of offsite electrical power. For the pipe brea ks for w hich i
the HPCIis intended to function, the core never uncovers and is continuously cooled, thus no cladding damage occurs (reference SAR Section 6.2.5.3 ). The repair times for the limiting conditi,ons of operation were set considering the use of the llPCI as part of the isolation cooling system.
l D, Autontatic Pressure Relier l
The relief valves of the automatic preuure relief subsystem are a backup to the HPCI subsystem. They r
enable the core spray subsystem or LPCI mode of the RilR system to provide protection against the small s
pipe break in the event of flPCI failure by deprewurizing the reactor vewel rapidly enough to actuate the core spray subsystems or LPCI nuste of the RilR sptem. The cure spray subsystem and me LPCI mode of the RilR sptem pros ale suthewns ilow of coolant to hmit fcel claidmg temi<raturesshless than 22000F, to assure that core geometry remains intact, to limit the core wide clad metal-water reaction to less than 1% and to limit l
the calculated local taetal-water reaction to less than 17A Loss of 1 of the relief valves affects the pressure relieving capability and, therefore, a 7 day repair period is specified.
Loss of more than one relief valve significantly reduces I
the pressure relief capability, thus a 24-hour repair l
period is specified based on the HPCI system availability during this period.
E.
~~
\\.
j The RCIC sptem a prosided to supply continuous m ieup water to the reactor core when the reactor j
is isolated from the turhme and when the leedw. ster sptem n riat available. Under these conditions the
)
purnping capacity ot the RCIC ptem a surhuent to mamt.an ine water les el above the core a uhout any other water system in operatmn if the water incl in the reactor sewel Jetreases to the RC!C uut.aticri level, the system automatically starn The spiem may also be manuall,s unnated at any tune.
3.5/05-12
~
QUAD.CI'lll S DPP-29 I
H. Ceedensate Pump Rcom Illoud Protection See Specificat.on 3.5.H.
L herage Planar LHGR This speci6 cation assures that the peak claddine temperature following the postulated design-basis a
loss <f-coolant accident will not exceed the 220CPF limit specified in the 10 CFR 50 Appendix K l considering the postulated effects of fuel pellet denufication.
%e peak cladding temperature fcilowing a postulated loss-of-coolant accident is primarily a function of the average heat. generation rate of all the rods of a fuel assembly at any axial toe.ition and is.only secondarily dependent on the risi.to. rod power distribution within.:n awembly Since expected local variatiora in power distribution withm.: fuel.swemNv airest the calculated peak (laddmp temperature by less th.intZr F relative to the peal temperature for.i tvpnal fuel deugn the hmit on the aserare l
plan r LHGR is sutlicient to.swure th.it calculated temperatures are below the limit. 'l he maumu m average planar LHG R's shown m Hgure 3.5-1 are based on calcul.itions employing the models described a
in Reference 2.
l J.
Iacal LHGR This speci6 cation assures that the maximum linear heat generation rate in any rod is less than the design linesr heat generation rate even if fuel pellet denuticanon is pmtulated. Tlie power ipiLe penahy specified is based on that presented in Reference 3 and awumes a knearly increaung v.ari.ition in anal gaps between core bottom and top and aoures won a 95% confidence that no more than_ene fuel rod exceeds the design linear heat. generation rate due to power spiLing An irradiation growth factor of a
0.25% was used as the basis for determining.1/P in accordance with References 4 and 5.
l K. Minimum Critical Power Ratio (MCPR)
The steady state values for MCPR specified in this specificatmn were selected to provide margin to accommo.
date transients and uncertainties m momtoring the core operating state as well as uncertainties in the crincal power correlation itself. These values also assure that operation wdl be sudi that the intual conditu,n anumed for the IDCA analyss. an MCPR of 1.IN.issatis'ied. For any of the special set of tranuents Jr disturbances caused by smgle operator error or smgle equmment malfunction. it is requued that deugn analyses tmuah/cd at this steady state operstmg lumt yic!d a MCPR of not less than that spectned in SpectScation 1.1.A at any
+
time during the transient, assuming instrument inp settings given m Spee ficanon 2.1. For analysis of the thermal consequences of these transients, the hmituig value of MCPR stated m this specificanon is con.
servatively assumed to exist pnor to the initiation of the transients. The results apply with increased con.
servatism whde operating with MCPR's greater than specitied.
The most limituig transaents with respect to MCPR are generally:
a) Rod withdrawal error b) Turbine trip without bypass c) Loss of feedwater heater Several factors influence which of these transients results m the largest reduction in entical power ratio such as the specific fuelloades exposure, and fuel type. The current cycles re!oad beensmg submittd specifies the limitmg transients for a given exposure uicrement u each fuel type. The salues specified as the Lumtsg Condition of Operauon are conservauvely chosen as the most restrictive over the entire cycle for each tuel type.
s.
l 1
3.5/4.5-14 l
/
QUAD-CITIES DPR-29 For core now rates less than rated. the sicady state MCPR is increased by the formula given in the specincation. This assures that the MCPR will be maintained greater than that specin:d in Specincation I.I.A even in the event that the motor-generator >ct speed controller causes the scoop rube positioner for the nuid coupler to move to the maximum speed position.
s i
I References I.
I. M. Jacobs and P. W. Marritt. GE Topical Report APED.5736. ' Guidelines for Determining Safe, Test Intervals and Repair Times for I:noncered Safeguards / April 19fi9.
2.
" Loss-of-Coolant-Accident Analysis neport for urcsden Units 2, 3 and Quad-Cities Units 1, 2 Nuclear Power Stations," EEDO-24146,
{
September 1978.
l 3.
GE Topical Report NEDM-10735.
- Fuel Densincation Effects on General Electric Boiling Water Reactor Fuel.* Section 3.2.1, Supplement 6. August 1973.
4 J. A. Hinds. GE Letter to V. A. Moore. USAEC.
- Plant Evaluation with GE GEGAP.!!!! December 12.
1973.
5.
g USAEC Report.* Supplement I to the Technical Report on Densincation of General Electric Reactor Fuelst December 14.1973.
, j 3.5/4.S.15
wrn-ey i
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l
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. :pi.
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I 9
==gi:
QUAD CITIES UNITS 1 & 2 E
14.0 FUEL TYPE:
7D212L
..i-4 j.
t
- i:;.t.
- I
-l
.:i.!:b u.! _a _., '
.i
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T QUAD CITIES UNITS 1 & 2 l
. 5 T_
FUEL TYPE:
8D250 j
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.l 10.5 0
10.000 20.000 30,000 PLANAR AVERAGE EXPOSURE (.".D/T)
FIGURE 3.5-1 MAXIMUM AVERAGE PLANAR LINEAR HEAT 3ENERATION RATE (MAPLHGR)
VS. PLANAk AVERAGE EXPOSURE (Sheet 1 of 3)
I l
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- l 2
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i 14.5 j
. - --- -.-- - -- j - --- - i;-- ;- -
- - - = n n
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-FUEL TYPE:
EEIC-PU 4
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11.5 QUAD CITIES UNITS 1 & 2 FUEL TYPE:
6DRB265-L i.
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- - - - +
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i 10.000 20.000 30,000 PLANAR AVERAGE EXPOSURE (N D/T)
FIGURE 3.5-1 MAXIMUM AVERAGE PLANAR LINEAR l
HEAT GENERATION RATE (MAPLHGRi (Sheet 2 of 3}
VS. PLANAR AVERAGE EXPOSURE i
9-
uen-2, s.
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..ii. i ili FUEL TYPE:
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QUAD CITIES UNIT 1 i
i 11 5 - = - - - -
FUEL TYPE:
Barrier LTA i
[
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FIGURE 3.5-1 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) vs. PLANAR AVERAGE tXP05URt (Sheet 3 of 3) y*
j QUAD-CITIES DPR-29 i ~.("l
power operation. From and after the i
date that one of these systems is made
{
or found to be inoperable for any rea-son, reactor power operation is per-missible only during the succeeding 7 days.
- 3. If the conditions in I or 2 above can-not be met, an orderly shutdown shall be initiated and the reactor shall be in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
E.
Safety and Relief Yalves E.
Safety and Relief Valves
- 1. Prior to reactor startup for power op-A minimum of I/2 of all safety valves shall be cration, during reactor power operat-bench checked or replaced with a bench ing conditions and whenever the reac-checked valve each refueling outage. The pop-tor coolant pressure is greater than 90 ping point of the safety valves shall be set as poig and temperature greater than follows:
323' F, all nine of the safety valves shall be operable. The solenoid.
Number of Valves Serpoint (psig) activated pressure valves shall be oper-11 5(1) able as required by Speci6 cation
..D.
2 12 Q
2 1250
- 2. If Specincation 3.6.E.1 is not met, the 4
1260 reactor shall remain shut down until the condition is corrected or, if in The allowable setpoint er,ror for each valve is operation, an orderly shutdown shall il%
be initiated and the reactor coolant MI Mief vh shli k hked for m ps-sure refueling utage. The set pressures 90 n
2
- F wi hours.
Number of Valves Serpoint (psig)
I s 1115(1) 2 s 1130 l
2 s1835
"' Target Rock combination safety / relief valve.
F.
Structurallatepity F.
Structurallaaeyity "Ihe structural integrity of the primary system The nondestructive inspections listed in Table boundary shall be maintamed at the level re-4.61 shall be performed as specified in accor-quired by the ASME Boiler and Pressure Vessel dance with Section XI of the ASME Boiler and Code,Section XI, " Rules for Inservice inspection Presmue Vessel Code,1971 Edition, Summer of Nuclear Power Plant Components", 1974 1971 Addenda. The results obtained from com-Edition, wmmer 1975 Addenda (ASME Code pliance with this specification will be evaluated Section XI).
after 5 years and the conclusions will be reviewed i
with the NRC.
3.6/46-4
--m r
v
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(
ENCLOSURE II QUAD-CITIES UNIT 1 NRC DOCKET NO. 50-254 O
i l
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NED0-24145 78NED283 Class I September 1978 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR QUAD CITIES NUCLEAR POWER STATION
}
UNIT 1 RELOAD 4 C'
Prepared:
" R.T. Hill,' Senior Licensing Engineqr Operating Licenses I
)*
i Approved: A
[OperatingLicensesI E. E'ngel, Mar [a'ger j
NUCLE AR ENE RGY PROJECTS OlvisiON. GENE A AL ELECTRIC COMPANY SAN JOSE, CALIFORNI A 95125 GENERAL h ELECTRIC
~'
NEDO-24145 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT
,j PLEASE READ CAREFULLY
-si j
This report was prepared by General Electric solely for Commonwealth Edison Company (CE) for CE's use with the U.S. Nuclear Regulatory Commission (USNRC)
,.i for amending CE's operating license of the Quad Cities Nuclear Power Station Unit 1.
The information contained in this report is believed by General Electric to be an accurate and true representation of the facts known, obtained or provided to General Electric at the time this report was prepared.
O The only undertakings of the General Electric Company respecting information in this document are contained in the nuclear fuel agreements, as amended, between Commonwealth Edison Company and General Electric Company for nuclear fuel and related services for the nuclear system for Dresden Nuclear Power Station Units 2 and 3, dated December 13, 1965 and nothing contained in this document shall be construed as changing said agreements. The use of this information except as defined by said agreement, or for any purpose other than that for which it is intended, is not authorized; and with respect to such unauthorized use, neither General Electric Company nor any of the contributorstothisdocumentmakesanyrepresentationorwarrantj (express or implied) as to the completeness, accuracy or usefulness of the information contained in this document or that such use of auch information may not infringe i
privately owned rights; nor do they assume any responsibility for liability or damage of any kind which may result from such use of such information.
l
\\
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t NEDD-24145 1.
PLA';T UNIOUE ITEMS (1.C)*
t a.
Plant parameter changes See Appendix A b.
Loading Error See Appendix A I
c.
Loss-of-Coolant Accident Analysis See Reference 1 (pg 5) i d.
Barrier Lead last Assembly (BLTA)
See Reference 2 (pg 5) e.
R (item 4)
Value shown includes effect of B C 4
settling (0.0004 Ak) 2.
RELOAD FUEL BUNDLES (1.0, 3.3.1 and 4.0) i Fuel Type Number Number Drilled Irradiated Initial (7DB212) 128
~
4 Reload-1 (7DB230) 22 (7DB230-STR) 1 (7DB230-Pu) 5 (8DB250) 36 Reload-2 (8DB250) 104 (8DB262) 52 Reload-3 (8DB250) 184 New Reload-4 (8DRB265L) 192 192 Total 724 192
~
3.
REFERENCE CORE LOADING PATTERN (3.3.1)
({j Nominal previous cycle exposure: 15,695 mwd /t.
Assumed reload cycle exposure:
16,100 !Md/t.
t
.i Core loading pattern: Figure 1.
4.
CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTD! WORTH -
NO VOIDS, 20*C (3.3.2.1.1 AND 3.3.2.1.2)
BOC k,gf Uncontrolled 1.106 Fully Controlled 0.945 l
Strongest Control Rod Out 0.980 I
R. Maximum Increase in Cold Core Reactivity 0.0024 with Exposure Into Cycle, ak
(
- ( ) refers to areas of discussion in " Generic Reload Fuel Application,"
j NEDE-24011-P-A, August 1978.
1 l
t
NEDO-24145 5.
STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)
Shutdown Margin (ak) ppm (20*C, Xenon Free) 600 0.045 t
6.
RELOAD UNIQUE TRA';SIENT ANALYSIS INPUTS (3.3.2.1.5 AND 5.2)
EOC EOC-1000 mwd /t l'
Void Coefficient N/A*(c/% Rg)
-5.82/-7.28
-6.57/-8.21 Void Fraction (%)
34.49 34.49
(])
Doppler Coefficient N/A (c/%*F)
-0.229/-0,217
-0.223/-0.212 Average Fuel Temperature (*F) 1203 1203 Scran Worth N/A (S)
-41.29/-33.03
-39.22/-31.38 Scram Reactivity vs Time Figure 2a Figure 2b 7.
RELOAD UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARANETERS (5.2)
EOC 7x7 8x8/8xSR Peaking factors (local, radial and axial) 1.30 J.22 1.56 1.84 1.40 4.40 R-Factor 1.100 1.051 Bundle Power (MUt) 5.304 6.245 Bundle Flow (103 lb/hr) 115.88 107.50 Initial MCPR 1.23 1.29
{'
- N = Nuclear Input Data A = Used in Transient Analysis l
~
t 2
t
1 - --
EEDO-24145 d
?
8.
SELECTED !iARGIN IfLDROVD1ENT OPTIONS (5.2.2) f d
9.
CORE-WIDE TPRiSIENT ANALYSIS RESULTS (5.2.1) 4CPK Fe.er Flow t
Q/A
,1 p
s y
8m6f Plant fransient bresvre
(*)
Q Q
Q.
(pstr' (psia) 7x7 ex'R Resten *e I
i Load Rejection 10C 99 100 264.4 110.7 121) 1246
- 0. l e '
O.22 Figure 3a without Bypass Turbine trip Euc -
100 100 225.0 106.6 121a 12 5 without Byp.se 10v0 $4. t Figure 3ti
}
Less of 145'r 1s0 100 120.7 119.4 991 1045 C.15 0.18 Figure 4 Feeowater
+
Heating FeeJwater ECC 100 100 155.3 107.5 1123 1158 0.05 0.11 Failure Figure Sa
~
Controller Feedwater E0C -
100 100 131.2 105.7 111' 1153 Coctrolier 1000 WJ/t Figure 5b j
Failure 10.
_ LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRL' MENT FAILURE) TRANSIENT SUIDfARY (5.2.1)
Rod sition Rod Block (Feet t.CPR MLEGR (Kw/ft)
Lir:iting Readdng Withdrawn) 7x7 8x8/8x8R 7x7 8x8/8x8R Rod Pattern O
104 3.5 0.10 0.16 16.34 15.30 Figure 6 105 3.5 0.10 0.16 16.34 15.30 Figure 6 106 4.0 0.11 0.18 16.72 15.86 Figure 6
- 107 4.5 0.13 0.20 16.90 16.20 Figure 6 108 6.5 0.17 0.26 15.66 15.46 Figure 6 109 7.5 0.24 0.28 15.26 15.70 Figure 6 i
- Indicates setpoint selected M
E 3
...;-....i. -
~ --
-^-
NEDO-24145 11.
OPERATING MCPR LIMIT (5.2) 1.29 (8x8/8xSk fuel) i 1.23 (7x7 fuel)
I i
- 12. OVERPRESSURIZATION ANALYSIS SD0!ARY' (5.3)
}
l Power Core Flow si v
Plant i
(%)
(%)
(psig)
(psic)
Response
MSIV Closure 100 100 1276 1310 Figure 7 (Flux Scram)
()
13.
STABILITY ANALYSIS RESULTS (5.4)
~
Decay Ratio:
Figure 8 Reactor Core Stability:
Decay Ratio, x /*0 0.56 2
(IC0% Red Line - Natural Circulation Power)
Channel Hydrodynamic Performance Decay Ratio, x /*0 2
(100% Rod Line - Natural Circulation Power) 8x8/8x8R channel 0.14 7x7 channel 0.04 f.
14.
LOSS-OF-COOLANT ACCIDENT RESULTS (5.5.2)
See Reference 1.
i l
4 L
- 15. _LOADII:C ERROR RESULTS (5. 5.4)
Limiting E.ent: Misplaced bundle j
!!CPR:
1.07
)
1 16.
_ CONTROL ROD DROP A'ALYSIS RESULTS (5.5.1) b Maximum incremental control rod worth:
tpk % ok C. <u*dB f
/b' n/ W/8
REFERENCES:
i h
(1)
" Loss-of-Coolant Accident Analysis Report for Dresden Units 2, 3 and Quad Cities Units 1,2 Nuclear Power Stations," NEDO-24146, September 1978.
(?) " Quad Citi es Nuclear Power Station Unit 1 Reload 4 Supplemental Licensing Information For Barrier Lead Test Assemblies," NEDO-24147,' September 1978 O
1 i
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.----v,---
,1 4
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)
u I
sMMMMMMMMMMHs MMMMMMMMMMMMM
- sMMMMMMMMMMMMHs
- MMMMMMMMMMMMMMM
- H M M M M M M M M M M M M M M
- MMMMMMMMMMMMMMM
- MEMMMMMMMMMMMMM
- MMMMMMMMMMMMMMM
- ,HMMMMMMMMMMMM*
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- MMMMMMMMMMM*
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l
- MMMMMMMMM*
- HMMMM*
IIIIIIIIII 01 03 05 07 09 11 13 15 17 19 21 23 25 27 29 31 33 35 37 39 41 43 45 47 49 51 53 55 57 59 FUEL TYPE i
A INITIAL FUEL D=
RELOAD 2 (8062501
=
B R E LOA 01 1TD8:301 E
=
RELOAD 2 4808262)
=
F = RELOAD 3 (8092501 qENTE TEST ROD)
G=
A E LO A O 4180R B265 Ll C
RELOAD 1 (808250)
=
Figure 1.
Quad Cities Unit 1 Reload 4 Desi;;n Reference Core Loading 6
NED0-24145 do 1D 1
Os 35 08 30 0.7 1
25 CONTROL ROD DR W E 06 E
J.
z t
9 2
0 y 20 - e 05 E
s e
s E
2 8
8 04 NOMINAL 15 0.3 -
~
J USED IN 10 -
ANALYSl$
02 -
5 -
G.1 -
I I
g O
1 2
3 4
I Figure 2a.
Scram Reactivity and Control Rod Drive Specification vs Time (EOCS)
I l
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l; 9
1
,1 1
so la 4
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of,
--j 06 2
k CONTROL ROD O
DRIVE ti 20 -5 0.5
.a W
O E
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3
/
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8
=
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0.3 10 USED IN AN A LYSIS 0.2 I
{
I.
O.1 0-0 i
I f
0 1
2 3
4 TIME AFTER SCRAM TRIP (wc}
Figure 2b.
Scram Reactivity and Control Rod Drive Specification vs Time (1000.Wd/t Before EOCS) 0
.-.~..-.....-.-.:.
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I.oad Rejection Without Ilypass, EOC, 99% Power
..-_..m._..
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1 lettTMDt FLtfX 1 WSTI. P%S RTSE (PST) i-2 mT Sierra W4tT FUfX 2 91e ir v4 si it tw 150.
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- 2. NUMBERS INDICATE NUMBER OF NOTCHES WITHOR AWN OUT OF 48 BLANK IS A WITHDR AWN ROD
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EEDO-24145 APPENDIX A i
1' PLANT PAR /JETER CHANCES I.
Pressure Relief Systems (Table 5-4, pg 5-62, NED0-24011) j Safety / Relief valve setpoint (psig) 1115 + 1%
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(...
ENCLOSURE III QUAD-CITIES UNIT 1 NRC DOCKET NO. 50-254 e
(.
NEI)0-24147 78NED285 Class I September 1978 i
i QUAD CITIES M;a. EAR POWER STATION
~..
UNIT 1 RELOAD 4 SUPPLDfENTAL LICENSING INFORMATION FOR BARRIER LEAD TEST ASSEMBLIES l
(w "s.
NUCLE AR ENERGY PROJECTS DIVISION
- GENER AL ELECTRIC CCVPANY SAN JOSE, CALIFORNI A 95125 GENERAL h ELECTRIC
?
TABLE OF CONTENTS a
l 4,
Page 1.
INTRODUCTION 1-1 i
j 2.
SUMMARY
2-1
- l i
3.
MECHANICAL DESIGN
-i 3-1 4
4.
THERMAL HYDRAULIC ANALYSIS 4-1 5.
NUCLEAR CHARACTERISTICS 5-1 j
6.
REACTOR LIMITS DETERMINATION 6-1 7.
REFERENCES 7-1 l
i O
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, _,,,, - - - ~ ~ -
NEDO-24147 LIST OF ILLUSTRATIONS Figure Title Page 2-1 Quad Cities Unit 1 Reload 4 Design Reference Core Loading Showing BLTA Locations 2-2 3-1 Barrier Lead Test Assembly Lattice 3-14 3-2 Barrier Lead Test Assembly U-235 Enrichment Axial Profile 3-15 LIST OF TABLES Page
%.]
Table Title Page 3-1 8DRB265 Reload Fuel Assembly and BLTA Design Parameters 3-13 6-1 CETAB Transient Analysis Initial Condition Parameters 6-4 6-2 Barrier Lead Test Assembly MAPLF.GR versus Exposure 6-4 a'
N, l
v/vi
NEDG-24147 1.
INTRODUCTION This document provides the supplemental information for four Bx8 Barrier Lead j
Test Assemblies (BLTA) which are part of Reload No. 4, Cycle 5, at the Quad i
4 Cities 1 Nuclear Power Station.
The generic design information and safety l
analyses for standard 8x8R fuel given in References 1 and 2 are applicable to the Earrier Lead Test Assemblies except as noted in the following supplemental I
informa tion.
i i
1.1 OBJECTIVES (I
The Barrier Lead Test Assembly Program is part of a larger demonstration prog which is intended to provide early experience in a commercial power reactor with fuel rod designs developed for their potential capability to remedy the pellet-cladding interaction (PCI) fuel rod failure mechanism.
The primary objective of the Barrier Lead Test Assembly Program is to accumulate burnup ahead of a large-scale demonstration to provide assurance that, while the remedy resists PCI, it is not subject to some unforeseen problem that becomes manifest at high burnup.
Further objectives of the program are to help define manufacturing process parameters and provide a source of prototypical lead burnup fuel rods which would be available for testing or destructive examination.
1.2 SCOPE y.
The four Barrier Lead Test Assemblies to be loaded as part of Reload 4 at Quad Cities 1 a e targeted to be operated for at least four full reactor cycles The four bundles are all similar in design (Section 3.1) and will be located in symmetric core locations to obtain similar operating histories Location of the BLTA's in the core has been selected such as to assure relatively high power operation and to minimize the impact of the BLTA's on the operation of Quad Cities 1.
I 1-1
NEDO-24147 The performance of these four BLTA's will be closely monitored by the General Electric Company through the following tasks: (1) specific pre-irradiation characterization measurements will be performed; (2) archive samples will be retained; and (3) an inspection program during each reactor outage will be implemented subject to Commonwealth Edison Company approval and the availability of authorized funds.
One of the tasks may also include the replacement of j
selected fuel rod segments.
The removed segments would be available for testing or destructive examination, thus providing additional performance data.
A detailed post-irradiation examination of the BLTA full-length fuel rods could be performed if the interim inspection program indicates that such an examination would be beneficial.
9 T
O g~
V 1-2
_-~... -.....-
NED0-24147 2.
SUMMARY
The Barrier Lead Test Asse.1blies are the same 8x8 lattice configuration and have the same dimensions as General Electric Company's standard 8x8 retrofit fuel (Reference 1).
Notable BLTA mechanical differences with the 8x8 retrofit fuel (Table 3-1) are the use of two segmented rods, a cladding inside surface which is either lined with high purity zirconium or plated with copper in all fuel rods, and fuel rod prepressurization of three atmospheres.
Enrichments are the same as the 8DRB265L bundle.
The segmented rods use the same enrichment as the rods they replace, but contain no natural UO.
2 The BLTA has been evaluated with specific attention to the noted mechanical differences, and results show that all design requirements are satisfied.
Core locations of the BLIA 's are shown in Figure 2-1.
Evaluations show that the BLTA's, as located, will not limit operation of the core. Safety analyses indicate that there is an insignificant effect on the Quad Cities 1 core characteristics resulting from loading the four BLTA's in the locations shown.
(~ %
v I
2-1
NEDO-24147 sHMMMMs sMMMMMMMEL i
sMMMMMMMMMMMs l
MMMMMMMMMMMMM
~
- sMMMMMMMMMMMME
- MMMMMMMMMMMMMMM rMMMMMMMMMMMMMMM o
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'MMMMM*
l llIIIIIIII i
l 01 03 05 07 09 11 13 15 17 19 21 23 25 27 29 31 33 35 37 39 41 43 45 47 49 51 53 55 57 $3 I
i FUEL TYPE A =
INITI AL FUEL D=
RELOAD 218082501 8
=
(
if = E 9Et T F
RELO AD 3 (808250)
=
TEST ROOl G=
RELOAD 4 (80RB265L)
C RELOAC 1 (808250)
S = S ARRIER LTAI
=
i l
Figure 2-1.
Quad Cities Unit 1 Reload 4 Design Reference Core Loading i
Showing BLTA Locations 2-2 I
^^-
NEDO-24147 3.
MECHANICAL DESIGN 3.1 GENERAL DESCRIPTION 3.1.1 Bundle 4
The Barrier Lead Test Assembly (BLTA) design is structurally the same as the 8x8 retrofit (8x8R) design which is being applied to Quad Cities 1 Reload 4 A limited number of bundle modifications have been made to accommodate barrier cladding on all fuel rods and to provide short fuel rod segments which could be removed for later testing. This section reviews the BLTA design and
(__)provide., a detailed description of the bundle changes.
Table 3-1 lists signifi-cant design parameters of the BLTA and provides a direct comparison with the BDRB265L reload fuel bundle.
The BLTA fuel bundle contains 60 full-length fuel rods, two segmente~d fuel rods of four segments each, and two whcer rods, one of which is also a spacer positioning rod.
The rods _ are spaced and supported in a square (8x8) array by the upper and lower tieplates in the same manner as shown on Figure 2-1 of Reference 1.
The lower tieplate has a nosepiece which supports t'he fuel assembly in the reactor.
The upper tieplate has a handle fcr transEerring the fuel bundle from one location to another, The identifying assembly serial
(_) number is engraved on the top of the handle, and a boss projects fro of the handle to aid in ensuring p.oper fuel assembly orientation.
Both upper and lower tieplates are fabricated from Type-304 stainless steel castings.
Zircaloy 4 fuel rod spacers equipped with Inconel-X springs are employed to maintain rod-to-rod spacing.
Finger springs are also employed with the BLTA design at the lower tieplate to channel interface.
The BLTA fuel assembly outline dimensions are the same as those of the 8DRB265L reload fuel to be used for Quad Cities 1 Reload 4 The BLTA uses identical tieplates, spacers and finger springs as the 8DRB265L fuel bundle.
3-1
~
NEDO-24147 3.1.2 Fuel Rods Three types of fuel rods are used in the BLTA.
In addition to tie rods and standard rods, there are two segmented rods, consisting of four segments each, j
which replace two of the standard rods in the 8DRB265L fuel bundle.
Each fuel rod consists of high-density ceramic uranium dioxide fuel pellets stacked within I
a barrier cladding which is evacuated, backfilled with helium, and sealed with Zircaloy end plugs welded in each end.
The helium backfill pressure is 3 atm at room temperature.
Recent studies have shown that a larger inventory of helium gas (achieved oy using a higher cold internal pressure) improves the gap conductance between fuel pellets and cladding resulting in reductions in fuel temperature, thermal expansion and fission gas release.
( ' operate at effectively lower linear heat generation rates (kW/ft) and are The pressurized rods therefore, expected to yield performance benefit in terms of increased fuel reliability.
The effect of 3 atm prepressurization in the BLTA's it negligible on Quad Cities 1 core performance and plant operation.
The prepressurization (3 atm) selected in this design also results in improved margin to MAPLHGR limits by reducing stored energy.
Experience with prepressurized fuel rods has been obtained in General Electric'c ongoing Pressurized Test Assembly (PTA)
Frogram (Section 3.3).
The eight fueled tie rods in cach bundle have threaded end plugs whthh thread into the lower tieplate casting and extend through the upper tieplate casting.
/_s
\\_-
A stainless steel hexagonal nut and locking tab are installed on the upper end plug to hold the assembly together. These tie rods support the weight of the assembly during fuel-handling operations only when the assembly hangs by the handle; during operation, the fuel rods are supported by the lower *.iepiste.
Except for the inside surface of the cladding and the 3 atm prepressurization, the tie rods are identical to these used in the 8DRB265L fuel bundles.Standard nuts and locking tabs are used for their assembly.
Each of the two segmented rods in each bundle consists of four identical fuel rod segments, an upper extensien plug, and a lower extension plug.
Each seg:ent has a threaded upper end plug and a threaded hole in its lower end plug, which make them ccepletely interchangeable.
The extension plugs are designed to provide the same interface with the tieplates as the end plugs on standard full-length fuel rods.
A complete segmented rod is assembled by 3-2
. -.. -. ~
NEDO-24147 screwing the extension plugs and four identical segments together. The positions of the segmented rods in the BLTA were selected to maintain the diagonal symmetry of the bundle and to minimize the impact of potential flux peaking in adjacent fuel rods. Experience with similar segmented fuel rods has been obtained in three specially designed Segmented Test Rod bundles which have i
been in operation since 1974 (see Section 3.3).
}
The remaining 52 fuel rods in a bundle are basic fuel rods having the same active fuel length as the tie rods. The end plugs of the standard rods have r
pins which fit into anchor holes in the tieplates. An Inconel-X expansion spring located over the top end plug pin of each fuel rod keeps the fuel rods seated in the lower tieplate and allows them to expand axially and independently by sliding within the holes of the upper tieplate. Except for the barrier
(
or liner on the inside surface of the cladding and the 3 atm prepressurization, the BLTA basic fuel rods are identical to the standard 8DRB265L fuel rods.
Standard expansion springs are used with the BLTA rods.
Adequate free volume is provided within each fuel rod and fuel rod segment in the form of a pellet-to-cladding gap and a plenum region at the top of the fuel rod to acco=modate thermal and irradiation expansion of the UO and the 2
internal pressures resulting from the 3 atm helium fill gas, impurities, and gaseous fission products liberated over the design life of the fuel. A plenum spring, or retainer, is provided in the plenum space to minimize movement of
(_
the fuel column inside the fuel rod during fuel shipping and handling.
A
\\_/
hydrogen getter is also provided in the plenum space as assurance against chemical attack from the inadvertent admission of moisture or hydrogenous impurities into a fuel rod during manufacturing.
Standard retainers and getters are used in all full-length rods, and similar configurations are applied to the segmented rod, each of which has its own independent plenum.
3.1 3 Water Reds The two water rods used in the BLTA are identical to the 8x8R fuel bundle water rods, and their functions are the same. One of these is used to position seven Zirealoy-4 spacers. The spacer-positioning water rod is assembled to the spacers by sliding the rod through the spacer cell.
The rod is then rotated so that the tabs bracket the elements of the spacer structure, thereby positioning the spacer in the required axial position.
The rod is prevented from rotatinF to unlock the spacers by the' engagement of its (square) lower end plug with 3-3
_.J__.
NEDO-24147 a squa' re tieplate hole.
Several holes are drilled around the circumference of the water rods near each end to allow coolant water to flow through the rods.
I 3.1.4 Fuel i
l The BLTA uses the same fuel pellets as the 8DRB265L fuel bundle.
The fuel j
consists of high-density ceramic uranium dioxide, manufactured by compacting and sintering uranium dioxide powder into cylindrical pellets with chamfered j
edges. The average UO2 pellet immersi n density is approximately 955 of theoretical density.
The pellet dimensions are given in Table 3-1.
Eight different U-235 enrichments are used in the fuel assemblies to reduce the
()
local power peaking factor (Figure 3-1).
Fuel element design and manufacturing procedures have been developed to prevent errors in enrichment location with a fuel assembly.
The fuel rods within each assembly are designed with character-istic mechanical end fittings and are marked with an enrichment identification for each enrichment.
The BLTA bundle incorporates the use of small amounts of gadolinium as a burnable poison in selected fuel rods.
The gadolinia-urania fuel rods are designed with characteristic extended end plugs, which are the same as used on the 8DRB265L fuel bundle gadolinia-uranium rods.
These extended end plugs permit a positive, visual check on the location of each gadolipium-bearing rod af ter bundle assembly.
(j y
Figure 3-1 shows the location of the various fuel rod types within the Barrier Lead Test Assemblies. With the exception of two segmented rods and either a zirconium liner or a copper barrier on the inside surface of the fuel rod cladding, the BLTA rod types and locations are the same as the 8DRB265L reload bundle.
Axial locations of the natural uranium in the fuel rods are shown in Figure 3-2.
The BLTA fuel rods contain the same enrichments and gadolinia loading as corresponding rods in the 8DRB265L bundle.
The BLTA segmented rods also contain the same enrichment as corresponding rods but the total active fuel length is shorter and no natural UO is used. At both ends of every segmented 2
red fuel column there are 0.5-in, hafnia-yttria pellets, which are placed there to assure that no adverse peaking will occur in the bundle due to these segments.
These flux depressor pellets are used in all three operating Segmerted Test Rod bundles (Section 3.3).
3-4
NEO-24147 3.1.5 Cladding J
Overall dimensions of the BLTA fuel rod cladding are identical to the Quad Cities 1 Reload 4 fuel rod cladding.
The thin copper barrier or zirconium liner on the inside surface displaces the same thickness of Zirealoy-2 material to maintain the same total cladding thickness.
The BLTA tubing with this slight f
reduction in Zirealoy-2 material is still adequate to be essentially free standing j
in the 1000 psi BWR external pressure environment.
,{
The following variations of barrier or liner cladding will be used (one BLTA I
of each type):
(..
(1) Cu-barrier (0.4 mil)
Electroless Cu-plating on Zr-2 cladding (2) Cu-barrier (0.4 mil)
Electroless Cu-platia.g on autoclaved (oxidized) Zr-2 cladding (3) Zr-liner (3.0 mil)
Crystal bar zirconium co-extruded with the Zr-2 cladding (4) 2r-liner (3.0 mil)
Lov oxygen sponge zirconium co-extruded with the Zr-2 cladding The cepper is plated on the inner surface of all the Zirealoy-2 cladding by b
electrolessly plate the inside surface. flowing a series of solution The procedure for plating copper directly onto the Zircaloy-2 cladding starts with annealed tubings after a chemical polish (pickle) and prior to the normal autoclave step.
A similar procedure is used for plating on the autoclaved tubes except that the plating is done after some initial surface preparation and the standard autoclave.
All the completed cladding has the same autoclaved outside surface as standard Bx8R fuel rod cladding.
Rigid inspection techniques, including *.00% ultrasonic testing for flaws, are applied to the tubes prior to plating. After plating, the tubes are nondestructively examined to measure copper thickness, and representative sa=ples are destructively tested to assure good adherence of the copper to the Zircaloy-2.
1 3-5
.I i
The high purity zirconium liner tubes are fabricated by co-extrusion of either the crystal bar or low oxygen sponge zirconium with the Zircaloy-2 cladding.
Extrusion billets of Zircaloy-2 are mac!.ined to accept liners of high purity
]
zirconium as close-fitting sleeves inside them. Extrusion of these composite f.
billets is then performed in a normal manner using standard lubricants and operating parameters. This process produces a high quality tubing, which has 3
a uniform liner thickness and an excellent mechanical bond with the Zircaloy-2.
i As a final process step, the outside surface of the zirconium liner tubes is also autoclaved the same as the production 8x8R tubes.
The same rigid inspection techniques are applied after processing the zirconium liner tubes.
This includes tight dimensional control and ultrasonic flaw detection.
3.1.5 Materials All of the BLTA materials which are exposed to the reactor coolant environment are identical to the 8xBR reload fuel. One new material, copper, is introduced in a 0.4 mil layer on the inside surface of all rods in two of the BLTA's.
Fuel rods in the other two test assemblies have a 3-mil inside layei of high purity zirconium (either crystal bar or low oxygen sponge material), which is also the principal constituent in the Zircaloy-2 cladding.
3.1.6.1 copper The copper barrier is at least 99% pure copper and forms an adherent plating to the Zircaloy-2 cladding. This very thin copper barrier has been disregarded in the mechanical design analysis because of its negligible structural contribution.
The benefit from copper is derived from its ability to protect the Zircaloy-2 inside surface from fission products which could potentially contribute to pellet-cladding interaction (PCI) failures.
Tests conducted to date (Reference 4) have confirmed this ability, and also have thus demonstrated the compatibility of copper with both UO fuel and Zircaloy-2 cladding.
2 3-6
NEO-24147 I.
3.1.6.2 zirconium a
The zirconita liner (either the crystal bar or low oxygen sponge material) is
'j high purity zirconium, which is coextruded with the Zircaloy-2 to form an excel-l lent mechanically bonded cladding. This composite cladding meets the same ten-sile strength requirements as standard 8x8R cladding.
However, the detailed I
l mechanical analysis of the cladding was performed using the actual lower zirco-nium strength properties for the applicable liner region.
This resulted in an effectively thinner cladding than the standard product from a mechanical standpoint.
Values of thermal conductivity, coefficient of thermal expansion, and elongation h'
of Zircaloy-2 were used for the zirconium liner material.
The benefit of the zirconium liner also is derived from its ability to resist PCI failures.
Extensive tests to date on this cladding configuration (Reference 4) have confirmed its PCI benefit as well as the material compatibility of zirconium in the fuel rod environment.
3.2 THERMAL AND MECHANICAL EVALUATIONS The same safety and design evaluations as performed for standard 8x8R fuel (Reference 1, Sections 2.4 and 2.5) have been applied to the BLTA designs.
h models used for these evaluations are also the same as described in Referen The Because the BLTA's use the same bundle components and the same fuel rod dimen-sions as tl.e Bx8R fuel bundle, most of the fbel assembly mechanical analyses are directly applicable to the BLTA's.
Effects due to the unique BLTA cladding, prepressurization and segmented rods have been accounted for in the BLTA safety and design evaluations.
3.2.1 Results from Safety Evaluation The calculated LHGR resulting in 15 plastic strain in the cladding is equal to or greater than 160% of the design maximum steady-state power throughout life for all rod types in the BLTA. This ratio considers the presence of a calcu-lated power spiking allowance for densification being added to the maximum
- LHGR, 3-7
l
,j NEDO-24147
)
3 3.2.2 Results from Design Evaluations 3.2.2.1 Steady-State Mechanical Performance 0
BLTA fuel is designed to operate at core rated power with sufficient design margin to accommodate reactor operations and satisfy the mechanical design applied to 8x8R fuel.
a In order to accomplish this objective, the BLTA fuel is designed to operate at a maximum steady-state linear heat generation rate (LHGR) of 13.4 kW/f t, plus an allowance for densification power spiking Thermal and mechanical analyses demonstrate that the mechanical design criteria are met for the maximum operating power and exposure combination throughout
/ >
\\?
the BLTA fuel life.
Design analyses performed for the BLTA fuel show that the stress intensity limits given in Table 2-6 of Reference 1 are not exceeded during continuous operation with LHGR's up to the operating limit of 13.4 kW/f t nor for short-term transient operation up to 165 above the peak operating limit of 13.4 kW/f t (i.e.,15.6 kW/f t), plus an allowanee for densificati6n power spiking.
Stresses due to external coolant pressure, internal gas pressure, thermal gradients, spacer contact, flow-induced vibration, and manufacturing tolerances were considered.
The maximum internal pressure is applied coincident with the minimum applicable coolant pressure.
~
3.2.2.2 Fatigue Analysis fh
~.-
During fuel life, less than 15% of the allowable fatigue life of the BLTA fuel in consumed.
l 3.2.2.3 Incipient Fuel Center Melting l
The LHGR's required to reach center melting in BLTA fuel rods are at least as high as the values shown in Table 2-4 of Reference 1.
The higher gap conductance due to prepressurization raises the required LHGR's for center melting, while the effect of the copper barrier or irconium liner is negligible.
j 3-8
3.2.2.4 Fuel Cladding Temperatures i
BLTA fuel cladding temperatures, as a function of heat flux, are very similar j
to the stancard 8x8R cladding temperatures shown in Reference 1, Figure 2-13 j
for late-in-life conditions.
The cladding geometries are the same and the effect of the barrier or liner is negligible.
3 2.2.5 Densification Analysis 3.2.2.5.1 Power Spiking Analysis
(,,?
The equation employed to calculate maximum gap size is described in Reference 3 The BLTA active fuel length and use of natural uranium at the ends 6f the fuel column are the same as the Quad Cities 18x8R reload fuel and have the same effect on maximum gap size.
The power increase as a function of axial position (as described in paragraph 2.4.2.1.1 of Reference 1) has been added to tha BLTA license limit LHGR (13.4 kW/ft) and considered in the B' TA desi'gn and safety analyses, wherever applicable.
3.2.2.5.2 cladding creep collapse A creep collapse analysis of BLTA fuel was performed using the same bases as described in paragrtph 2.5.3.1.1 of Reference 1.
The internal pressure due to helium backfill during fabrication was considered.
The higher backfill pressure (3 atm) is beneficial to the BLTA compared to the 8x8R design.
No credit was taken for internal gas pressure due to released fission gas or vola-tiles.
A thinner cladding was assumed for the BLTA creep-collapse analysis to conservatively account for the lower strength of the Zr liner.
Based on the analysis results, cladding collapse was not calculated to occur for the BLTA design.
3.2.2.5.3 Increased Linear Heat Generation Rate BLTA design changes have no effect en the conclusion for 8x8R fuel (Reference 1) that the pellet decrease in length due to densification is less than the increase in length due to daermal expansion.
3-9
.j NEDO-24147
- s
((
3.2.2.5.4 Stored Energy Determination The effects on stored energy due to densification are accounted for in the LOCA evaluation.
s I
3.3 FUEL OPERATING AND DEVELOPMENTAL EXPERIENCE 4
s The basic structure of the Barrier Lead Test Assemblies and the configuration of their fuel rods are the same as the standard reload fuel for Quad Cities 1 Reload 4 (Reference 2).
The unique BLTA features:
(1) ccpper barrier or zirconium liner cladding, em
()
(2) segmented rods, and (3) prepressurization have all been included in other fuel assemblies which are currently in operation and which have been extensively evaluated through both analysis and testing.
The program ur.ier which the BLTA's are being administered also includes a task that provides operating experience and other supporting test data (Reference 4)
Power reactor experience with fuel having the above features has been obtained through the use of three Segmented Test Rod bundles which have been operating since 1974.
More than 350 segments, of the same basic design as the 8 segnents
(
.)
in each BLTA, have been irradiated in these three test bundles.
The segments contain a variety of different fuel design concepts, including both copper barrier cladding and zirconium liner cladding. Wall thicknesses of the barrier segments have ranged from 28 mils to 34 mils (BLTA rods are 32 mils thick) and prepressurization has ranged from 1 atm to 17 atm of helium.
All recent replacement barrier segments have been prepressurized to 3 ats.
A complete summary of the status of all copper barrier and zirconium liner segments may be found in Reference 4.
Several Cu-Barrier and Zr-Liner segments have been removed from the Segmer.ted Test Rod bundles, and eight of these segments have been ramp tested to 18 kW/ft or higher in the General Electric Test Reactor (GETR).
The first two seg=ents tested had burnups between 4000 and 5000 mwd /t; the remaining segments ranged from 7200 to 9200 mwd /t when ramp tested. None of the segments experienced 3-10
i a pellet-cladding interaction failure.
One of the early tests of a copper barrier fuel rod did result in failure caused by substantial center melting, 1
which strained the cladding beyond its yield point.
The other copper barrier i
segment in the first test series also experienced extensive fuel center melting but survived the ramp test.
This experience base will continue to be expanded upon as the Segmented Test Rod Program proceeds, and will continue to provide I
information supporting BLTA operation.
In addition to this power reactor operating experience with segmented rods, a number of other supporting tests are also described in Reference 4 Prepressurization of General Electric BWR fuel rods with helium to 3 atm has
(
been extensively studied in preparation for a design change, which is now planned to be implemented during 1979 (Reference 5).
Part of this preparation has included the operation of a Pressurized Test Assembly (PTA) in Peach Bottom Unit 3 hince April 1977.
This fuel assembly has the same 8x8R ruel. rod geometry as the BLTA's and contains 24 rods that are prepressurized to 3 atm (Reference 6).
A visual examination of the assembly was performed at the Peach Bottom site in April 1978.
The mechanical integrity of the PTA was confirned with no abnormal conditions observed.
Further interim examinations of the PTA will be performed contingent on the availability of the hael as influenced by plant operation.
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NDO-24147 e
Table 3-1 BDRB265 RELOAD FUEL ASSDfBLY AND BLTA DESIGN PARAME"fERS i
. i 8DRB265L BLTA Fuel Assembly Geometry 8x8 8x8 Rod Pitch (in.)
0.64 0.64 Water to Fuel Volume Ratio 2.75 2.75 2
Heat Transfer Area (ft )
94,9 94,3 Weight of UO2 (Kg) 200.5 199.3 Weight of U (Kg) 176.8 175.7 Average Enrichment (w/o U-235) 2.65 2.65 b~'
Finger Springs Yes
- Yes Full-Length Fuel Rods Number 62 60
- Active Fuel Length (in. )
145.24 145.24 Gas Plenum Length (in.)
9.48
' 9.48 Fill Gas Helium Helium Fill Gas Pressure (atm) 1 3
Fuel Rod Segments Number
~
0 8
Active Fuel Length (in. )
N/A 29.6 3 Gas Plenum Length (in. )
N/A 5.5 Fill Gas N/A Helium Fill Gas Pressure (atm)
N/A 3
Fuel Material Sintered UO Sintered UO 2
2 Pellet Diameter (in. )
0.410 0.410 Pellet Length (in.)
0.410 0.410 Pellet Immersion Density (% TD) 95.0 95.0 Cladding Material Zr-2 Zr-2/Cu or Zr-2/Zr Outside Diameter (in. )
0.483 0.483 Thickness (Including Barrier or Liner, if any) 0.032 0.032 3-12 L
-'~ '
Table 3-1 8DRB265 RELOAD FUEL ASSEMBLY AND 8LTA DESIGN PARAMETERS jl 8DRB265L BLTA
{
Copper Plating Thickness N/A 0.0004 in.
l{
or l
Zirconium Liner Thickness N/A 0.003 in.
I Water Rod Material Z r-2 Zr-2 Outside Diameter (in. )
0.591 0.591 Thickness 0.030 0.030 Spacers Material Zr-4 with Zr-4 with Inconel Inconel X-750 springs - X-750 Springs Number per bundle 7
7 Fuel Channel Material Zr-4 Zr-4 Outside Dimension (in. )
5.438 5.438 Wall Thickness (in.)
0.080 0.080 O
' Includes natural Uranium (12 in. per full-length rod).
3-13
BLADE a
l 9
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6 5
4 4
4 4
4 5
i 5
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i 5
4 2
Gd 2
2 2
4 4
2 2
2 1
1 Gd 2
4 Gd 2
2S HO 1
1 2
2 4
2 1
HO 1
1 Gd 2
2 (v
4 2
1 1
1 1
1 2
4 2
Gd 1
Gd 1
2S 3
5 4
2 2
2 2
3 4
AODTYPE U-235 wt%
NUMBER OF RODS 1
3.8 2
14 3.0 3
19 2.4 4
2 2.0 5
14 1.7 6
4 1.3 Gd 1
3.0 H0 6
7 WATER RODS 2
25 4 SEGMENTED) 3.0 2
64
- SPACER POSITIONING Figure 3-1.
Barrier Lead Test Assembly Lattice 3-14
NEDO-24147 FULL LENGTH SEGMENTED FUEL ROD FUEL ROD t
4 l
TCP OF FUEL COLUMN i
t 6 in.
(B)
AL (C)
(o)
II ID) 0 0.5 (C) 0.5 U
(D) 133 24 in.
(A)
N A
(C) 29 63
~.
If N) i IC)
(D)
\\'
6 in.
(B)
A l
BOTTOM OF FUEL COLUMN (A) 2.82 I, AVER AGE IN A LL ROCS.
(B) 0.717, ALL RODS EXCEPT 2 SEGMENTED ROCS.
(C) 3.0 3, ALL SEGMENTS (D) H AFNI A YTTRI A PELLETS.
I Figure 3-2.
Barrier Lesd Test Assembly U-235 Enrichsent Axial Profile 3-15/3-16 t
NEDO-24147 4.
THERMAL-HYDRAULIC ANALYSIS Discussion of steady-state hydraulic models is presented in Section 4 of Refer-ence 1.
The BLTA is treated the same as an 8x8R bundle in the Quad Cities 1, Cycle 5 thermal-hydraulic analysis. Pressure drops through BLTA and standard 8x8R bundles are considered to be the same because identical bundle components and the same fuel rod geometries are used. Relative pressure drop effects due to the two segmented rods in each BLTA are negligible.
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NCDO-24147 6
5.
NUCLEAR CHARACTERISTICS The zirconium liners have no impact on the BLTA nuclear performance.
The copper barriers do not affect bundle rod-to-rod power distributions, but do cause j
a small (less than 0.2%) decrease in bundle reactivity.
i I
In the nonfueled regions of the BLTA's, the segmented rods cause local increases in the maximum peak-to-average rod powers. However, this local effect is counter-acted by decreases in total bundle power in the same regions.
In the natural vranium regions of the bundles, the enriched fuel of the segmented rods experience large increases in peak-to-average rod powers.
However, since these are low bundle power regions, the local increases are not significant.
In summary, nuclear analyses have shown that the effects of the two segmented rods and either the Zr-liner or the Cu-barrier cladding in each BLTA will have no significant effect on the nuclear performance of the BLTA's relative to the 8DRB265 reload fuel in Quad Cities 1.
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5-1/5-2 i
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NEDO-24147 6.
REACTOR LIMITS DETERMINATION 6.1 GETAB LIHITS 6.1.1 GETAB Transient Results The most severe transient for a Barrier Lead Test Assembly from rated conditions is a load rejection with failure of the bypass valves.
4 This event results in a maximum ACPR of 0.25.
Addition of this ACPR to the safety limit MCPR gives the minimum initial MCPR to avoid violating the safety limit MCPR during the most severe transient involving a BLTA.
(,
6.1.2 MCPR Operating Limit Based on a safety limit MCPR cf 1.07 for this cycle, the operating limit for the Barrier Lead Test Assemblies is 1.32.
This is 0.03 higher than the 8x8R fuel assemblies because of a conservative treatment of the effects of pre-pressurization on transient performance of the BLTA design.
A more comprehen-sive analysis comparing 3 atm and 1 atm initial pressures has shown an insig-nificant ( 0.01) difference in ACPR (Reference 5, Section 4.2.2).
6.1.3 Transient Analysis Initial Condition Parameters p)
The values used as initial input conditions for the BLTA transient analysis s.
are shown in Table 6-1.
6.2 STABILITY ANALYSIS The addition of four Barrier Lead Test Assemblies to the Quad Cities 1 core t
will have a negligible impact on the reactor core stability and the channel hydrodynamic performance as compared to 8DRB265L reload fuel bundles (Re ference 2, Section 13).
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i h
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.O 4
NEDO-24147 6.3 ACCIDENT EVALUATIONS 5
6.3.1 Loss-of-Coolant Accident Results
!t i
An emergency core cooling system (ECCS) analysis has shown that the effects of A
a loss-of-coolant accident on the Barrier Lead Test Assemblies results in a i
slight increase in MAPLHCR and lower peak cladding temperatures (PCT) compared to the 8x8R reload bundle.
This improvement is caused by the higher initial pressurization of the BLTA, which increases the gap conductance and reduces the stored energy in tse fuel rods.
O Results of the ECCS analysis are presented in Table 6-2.
The potential effect that the barriers could have on the overall c"ladding strength and, consequently, the calculated PCT, was considered by running an additional analysis which neglected the structural strength supplied by the 3 mil zirconium liner. Results of this separate analysis indicated the potential for a maximum of 10 fuel rod perforations at an exposure of 30,000 mwd /t for the highest power bundle.
No perforations were predicted for the copper barrier fue,1 where no significant reduction in cladding strength is expected to occur due to the thin-ness of the copper plating.
6.3.2 Loading Error Accident The analysis and results of a loading error involving either a rotated BLTA or a misplaced BLTA in an 8x8 or 7x7 location in Quad Cities 1, Cycle 5 are conserva-tively represented by the 3DR3265L reload bundle loading error results (Reference i
- 2. Section 15).
t l
6.3.3
_ Control Rod Drop Accident Since the nuclear characteristics of the BLTA are nearly identical to the standard reload bundle, the analysis and results of the control rod drop accident for the i
reload (Reference 2, Section 16) is considered applicable to the BLTA fuel.
6-2
NEDO-24 947 Table 8-1 GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS i
BLTA j
Peaking Factors (local, radial, axial)
(1.22, 1 79, 1.40)
R-Factor 1.051 l
Bundle Power, MWt 3
6.082 Bundle Flow, 10 lb/hr 108.57 Initial MCPR 1 32 t
Table 6-2
(..
BARRIER LEAD TEST ASSDiBLY MAPLHGR, PCT, OIIDATION FRACTION VERSUS EXPOSURE EXPOSURE MAPLHGR PCT OIIDATION (mwd /t)
(W/ft)
( F)
FRACTION 200 11.6 2171 0.032 1,000 11.6 2178 0.033 5,000 12.0 2198 0.034 10,000 12.1 2195 0.033 15,000 12.0 2200 0.033 20,000 11.7 2187 0.032 25,000 11.0 2115 0.056 30,000 10.5 2029 0.042 6-3/6-4
~
.o NEDO-24147 7.
REFERENCES 1.
" Generic Reload Fuel Application," NEDE-24011-P-A, August '979.
f 2.
" Supplemental Reload Submittal for Quad Cities Nuclear Power Station Unit 1 Reload 4," NEDO-24145, September 1978.
3.
V. A. Moore, letter to I. S. Mitchell, " Modified GE Model for Fuel Densifica-tion," Packet 50-321, Edwin O. Hatch Reactor, Unit 1, March 22,1974 4
" Demonstration of Fuel Resistant to Pellet-Cladding Interaction First Semi-annual Report,
(~.
July - Dece=ber 1977," GEAP-2377, February 1978.
L.
5.
R. B. Elkins, " Fuel Rod Prepressurization, Amendment 1," (Proprietary).
Licensing Topical Report, NEDE-23786-1-P, May 1978.
6.
" Pressurized Test Assembly Supplemental Information for Reload-1. Licensing Amendment for Peach Bottom Atomic Power Station Unit 3 Reanalysis Supplement,"
NEDO-21363 4, Supplement 4, January 1977 O
%s 7-1/7-2
h 4
1 ENCLOSURE IV QUAD-CITIES UNIT 1 l
NRC DOCKET NO. 50-254
~
1 1
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.s NEDO-24146 78NED284 Class I September 1978 i
LOSS-OF-COOLANT ACCIDENT ANALYSIS REPORT FOR DRESDEN 1: NITS 2, 3 AND QUAD CITIES UNITS 1, 2 NUCLEAR P0b'ER STATIONS I
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NUCLE AR ENE RGY PROJECTS OIVISION e GENE R AL E LECTRIC COYPANY SAN JOSE, CALIFORNI A 95125 l
GENERAL h ELECTRIC
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TABLE OF CONTENTS F
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Page 1.
INTRODUCTION 1-1 i
2.
LEAD PLANT SELECTION 2-1 3.
INPUT TO ANALYSIS 3-1 4.
LOCA ANALYSIS C0!!PUTER CODES 4-1 4.1 Results of the IX!B Analysis 4-1 j
4.2 Results of the SCAT Analysis 4-1 4.3 Results of the SAFE Analysis 4-1 4.4 Results of REFLOOD Analysis 4-2 4.5 Results of the CHASTE Analysis 4-3 4.6 Methods 4-3 5.
DESCRIPTION OF !!0 DEL A%'D INPUT CHANGES 5-1 6.
CONCLUSIONS
~
6-1 7.
REFERENCES 7-1 l
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I 111/iv 1
L
NEDO-24126 LIST OF TABLES Table Title Fjgij; 1
Significant Input Paraneters to the Loss-of-Coolant Accident 3-1 2
Summary of Break Spectrum Results 4-5 3
LOCA Analysis Figure Summary - Non-Lead Plant 4-6 4A MAPLilGR Versus Average Planar Exposure 4-7 4B MAPLHGR Versus Average Planar Exposure 4-7 4C MAPLHGR Versus Average Planar Exposure 4-8 4D MAPLHGR Versus Average Planar Exposure 4-8
()
4E MAPLHGR Versus Average Planar Exposure 4-9 4F MAPLHGR Versus Average Planar Exposure 4-9 4G MAPLHGR Versus Average Planar Exposure 4-10 4H MAPLHCR Versus Average Planar Exposure 4-10
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NEDO-24146 LIST OF ILLUSTRATIONS i
Figure Title g
- j 1
Water Level Inside the Shroud and Reactor Vessel Pressure Following a Maximum Recirculation Line Suction Break, LPCI Injection Valve Failure, Break Area = 4.18 ft2 (LBM) 6-3 2
Peak Cladding Temperature Following a Maximum Recirculation Line Suction Break, LPCI Injection Valve Failure, Break Area = 4.18 ft2 (LBM) 6-4 3
Fuel Rod Convective Heat Transfer Coefficient During Blowdown at the High Power Axial Node for a Maximum Recirculation Line Suction Break (4.18 ft )
6-5 2
4 Norr.alized Core Average Inlet Flow Following a Maximum
( }:.
Recirculation Line Suction Break (4.18 ft )
6-6 2
5 h*isimum Critical Power Ratio Following a Maximum Recirculation Lina Suction Break (4.18 f t )
6-7 2
6 Variation with Break Area of Time for Which Hot Node Remains Uncovered 6-8 i
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vil/viii
NED0-24146 1.
INTRODUCTION The purpose of this document is to provide the results of the loss-of-coolant accident (LOCA) analysis for the Drer. den Units 2, 3.and Quad Cities Units 1, 2 Nuclear Power Stations (D2,3/QC1,2) with a partial core loading of reload fuel with holes drilled in the lower tieplates. The analysis was performed using approved General Electric (GE) calculational models.
This reanalysis of the plant LOCA is provided in accordance with the NRC requirement (Reference 1) and to demonstrate conformance with the ECCS acceptance criteria of 10CFR50.46. The objective of the LOCA analysis con-(}
tained herein is to provide assurance that the most limiting break size, break location, and single failure combination has been considered for the plant.
1 The required documentation for demonstrating that these objectives have been satisfied in given in Reference 2.
The dogumentation contained in this report is intended to satisfy these requirements.
The general description of the LOCA evaluation models is contained in Reference 3.
Recently approved model changes (Reference 4) are described in References 5 and 6.
These model changes are employed in the new REFLOOD and CHASTE computer codes which have been used in this analysis.
In addition, a model which takes into account the effects of drilling alternate flow path
({])
holes in the lower tieplate of the fuel bundle and the use of such fuel bundles in a full or partial core loading is described in References 7, 8, and 9.
This model was also approved in Reference 4.
Also included in the reanalysis are current values for input parameters based on the LOCA analysis reverification program being carried out by GE.
The specific changes as applied to D2,3/QC1,2 (partial drill) are discussed in more detail in later sections of this document.
Plants are separated into groups for the purpose of LOCA analysis (Refer-ence 10). 'sithin each plant group there will be a single lead plant analysis which provides the basis for the selection of the most limiting break size yielding the highest peak cladding temperature (PCT). Also, the lead plant analysis provides an expanded documentation base to provide added insight into evaluation of the details of particular phenomena.
The remainder of 1-1
NEDO-24146 the plants in that group will have non-lead plant analyses referenced to the
~1ead plant analysis. This document contains the non-lead plant analysis for l
D2,3/QC1,2 which are now BVR/3's in the BWR/4 with loop selection logic group of plants and is consistent with the requirements outlined in Reference 2.
The same models and computer codes are used to evaluate all plants. Changes to these models will cause changes in phenomenological responses that are similar within any given plant group. The dif ference in input parameters are not expected to result in significantly different results for the plants within a given group.
Emergency core cooling system (ECCS) and geometric dif ferences between plant groups may result in different responses for different groups but within any group the responses will be similar. Thus, the C.
i lead plant concept is still valid for this evaluation.
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MEDO-24146 2.
LEAD PLANT SELECTION Lead plants are selected and analyzed in detail to permit a more comprehen-g sive review and eliminate unnecessary calculations.
This constitutes a
}
generic analysis for each plant of that type which can be referenced in I
subsequent plant submittals.
.f The lead plant for D2,3/QC1,2 with drilled fuel is Duane Arnold. The justifi-cation for categorizing D2,3/QC1,2 in this group of plants is the same as for Pilgrim and the lead plant analysis for this grouo is presented in Reference 11.
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MEDO-24146 is d
.8 3
3.
INPUT TO ANALYSIS
]
A list of the significant plant input parameters to the LOCA analysis is
)
presented in Table 1.
t j
~
Table 1 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS 4
4 Plant Parameters Core Thermal Power 2578 MWt, which corresponds to 102% of rated power Vessel Steam Output 9.96 x 106 lbm/h, which' corresponds 4
1 to 102% of rated power Vessel Steam Dome Pressure 1020 psia Recirculation Line Break Area 4.18 ft (DBA) for Large Breaks - Suction Number of Drilled Bundles.
156 Fuel Parameters:
Peak Technical Initial Specification Design Minimum i
Linear Heat Axial Critical Fuel Bundle Generation Kate Peaking -
Power Fuel Type Geometry (kW/ft)
Factor Ratio
- i A
A.
7D212 - No Gad 7x7 17.5 1.57 1.2 B.
7D212L 7x7 17.5 1.57 1.2.
C.
7D230 7x7 17.5 1.57 1.2 D.
EEIC - Pu 7x7 17.5 1.57 1.2 E.
8D250 8x8 13.4 1.37 1.2 I
F.
8D262 8x8 13.4 1.57 1.2 G.
8DRB265L 8x8 13.4 1.57 1.2 H.
Barrier LTA 8x8 13.4 1.57 1.2 i
To account for the 2% uncertainty in bundle power required by Appendix K, the SCAT calculation is performed with an MCPR of 1.18 (i.e.,1.2 divided by 1.02) l for a bundle with an initial MCPR of 1.20.
I S
f 3-1/3-2
.. _.. _., _ _ _. _ ~.. _ _ _ _ _. _ _ _
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NED0-24146 i
l 4
LOCA ANALYSIS COMPUTER COPES f
4.1 RESULTS OF THE LAMB ANALYSIS This code is used to analyze the short-term blowdown phenomena for large postu-lated pipe breaks (breaks in which nucleate boiling is lost before the water level drops and uncovers the active fuel) in jet pump reactors. The LAMS output (core flow as a function of time) is input to the SCAT code for cal-t culation of blowdown heat transfer.
The LAMB results presented are:
o Core Average Inlet Flow Rate (normalized to unity at the beginning of the accident) following a Large Break.
4.2 RESULTS OF THE SCAT ANALYSIS This code completes the transient short-term thermal-hydraulic calculation for large breaks in jet pump reactors. The GEXL correlation is used to track the boiling transition in time and location. The post-critical heat flux heat transfer correlations are built into SCAT which calculates heat transfer
~
coefficients for input to the core heatup code, CHASTE.
(%
%-)
The SCAT results presented are:
Minimum Critical Power Ratio Following a Large Break.
e l
e Convective Heat Transfer Coefficient following a Large Break.
4.3 RESULTS OF THE SAFE ANALYSIS This code "ih used primarily to track the vessel inventory and to model ECCS performance during the LOCA. The application of SAFE is identical for all break sizes. The cede is used during the entire course of the postulated acci-dent, but after ECCS initiation, SAFE is used only to calculate reactor system pressure and ECCS flows, which are pressure dependent.
l 4-1
- j The SAFE results presented are:
Water Level inside the Shroud (up to the time REFLOOD initiates) and e
j Reactor Vessel Pressure
}
e 4
'i 4.4 RESULTS OF REFLOOD ANALYSIS i
s This code is used across the break spectrum to calculate the system inveatories after ECCS actuation. The models used for the design basis accident (DBA) application ("DBA-REFLOOD") was described in a supplement to the SAFE code description transmitted to the USNRC December 20, 1974. The "non-DBA REFLOOD"
{)
analysis is nearly identical to the DBA version and employs the same major assumptions. The only differences stem from the fact that the core may be partially covered with coolant at the time of ECCS initiation and coolant levels change slowly for smaller breaks by comparison with the DBK.
More pre-cise modeling of coolant level behavior is thus requested principally to determine the contribution of vaporization in the fuel assemblies to the flow limiting (CCFL) phenomenon at the upper tieplate. The counter current differences from the DBA-REFLOOD analysis are:
i (1) The non-DBA version calculates core water level more precisely than the DBA version in which greater precision is not necessary.
~
k),
(2) The non-DBA version includes a heatup model similar to but less detailed than that in CHASTE, designed to calculate cladding tem-perature during the small break. This heatup model is used in calculating vaporization for the CCFL correlation, in calculating swollen level in the core, and in calculating the peak cladding temperature.
The REFLOOD results presented are:
Water Level inside the Shroud e
Peak Clad'ing Temperature and Heat Transfer Coefficient for breaks e
calculated with small break methods 4-2
NED0-24146
"}
i l
4.5 RESULTS OF THE CRASTE ANALYSIS 1
h This code is used, with suitable inputs from the other codes, to calculate the j
fuel cladding heatup rate, peak cladding t5mperature, peak local cladding
'I oxidation, and core-wide metal-water reaction for large breaks. The detailed
.l fuel model in CRASTE considers transient gap conductance, clad swelling and rupture, and metal-water reaction. The empirical core spray heat transfer and l
channel vetting correlations are built into CHASTE, which solves the transient heat transfer equations for the entire LOCA transient at a single axial plane in a single fuel assembly.
Iterative applications of CRASTE determine the maximum permissible planar power where required to satisfy the requirements of
()'
10CFR50.46 acceptance criteria.
The CRASTE results presented are:
Peak Cladding Temperature versus time e
Peak Cladding Temperature versus Break Area e
Peak Cladding Temperature and Peak Local Oxidation versus Planar e
Average Exposure for the most limiting break size m()
e Maximum Average Planar Heat Generation Rate (MAPLHGR) versus Planar Average Exposure for the most limiting break size 4
A sum =ary of the analytical results is given in Table 2.
Table 3 lists the figures provided for this analysis. The MAPLHGR values for each fuel type for D2,3/QCl,2 are presented in Tables 4A through 4H.
4.6 MEIH0DS In the fol' lowing sections, it will be useful to refer to the methods used to analyze DBA, large breaks, and small breaks. For j et-pump reactors, l
these are defined as follows:
a.
DBA Methods. LAM 3/ SCAT / SAFE /DBA-REFLOOD/ CHASTE. Break size: DBA.
4-3
NEDO-24146 1,i n,;
1
- I.
b.
Large Break Methods (LBM). LAMB / SCAT / SAFE /non-DBA REFLOOD/ CHASTE.
2 Break sizes:
1.0 ft 1 A < DBA.
b.
Small Break Methods (SBM). SAFE /non-DBA REFLOOD. Heat transfer i
i coefficients: nucleate boiling prior to core uncovery, 2
25 Btu /hr-ft _.F after recovery, core spray when appropriate. Peak l
cladding temperature and peak local oxidation are calculated in
}
non-DBA-REFLOOD. Break sizes: A i 1.0 f t j.
1 O
O T
a e
lI e
e
- 4 i
.)
~
4-4
.i h
Table 2 St*!O!ARY OF BREAK SPECTRC! RESl'LTS e Break Size e Location Cor e-Wid e Peak Local Metal-Water e Single Failure PCT (*F)
Oxidation (%)
Reaction (%)
e 4.18 ft' 2200(1) 11.6 0.20 e Recire Suction LPCI Injection Valve e
{"s'N 1.
PCT from CHASTE 9
S e
l l
l e
O 4-5 l
l
~~
' i Table 3 i
LOCA ANALYSIS FIGURE SIMMARY - NON-LEAD PLA'iT
' f.$
Large Break Methods s.l Maximum
[';
Suction Break DBA
?
(LPCI Injection Valve Failure) i 4
(4.18 ft2)
I !-
i t
- Water Level Inside Shroud and Reactor Vessel Pressure 1
Peak Cladding Temperature 2
k)
Heat Transfer Coefficient 3
1 Core Average Inlet Flow 4
Minimum Critical Power Ratio 5
Peak Cladding Temperature of the Highest Powered Plane Experiencing Boiling Transition 2
4 Variation with Break Area of Time for Which Hot Node Remains Uncovered 6
O i
i i
J 9
e 4-6 9
4 5
,--n.
m-,
,.,.w
--,...y
-,,-.c-,-. - -,, - -
..,w...-,,-+ve,-e-,.,,,,,_n,--..,..,-.
..e,,,
s.--
- *-. _,. ~ ~ ~
f Table 4A MAPLHCR VERSUS AVERAGE PLANAR EXPOSURE i
t PLANT: Dresden - 3 FUEL TYPE:
7D212 - NO GAD t
Average Planar Exposure MAPLHGR PCT 0xidation (mwd /t)
(kW/ft)
(*F)
Fraction 5,000 14.8 2192 0.036 12,500 14.7 2198 0.061 4
22,500 14.2 2189 0.082 30,000 12.5 1994 0.033 Table 4B MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE PLANT: Dresden 2/ Quad Cities 1,2 FUEL TYPE:
7b212L Average Planar Exposure MAPLHGR PCT Oxidation
(!!Wd /t)
(kW/ft)
(*F)
Fraction 200 14.0 2198 0.642 1,000 14.1 2195 0.040 72
\\-
5,000 14.7 2198 0.037 10,000 14.7 2197 0.036 15,000 14.4 2198 0.081 25,000 13.2 2063 0.023 30,000 12.1 1938 0.015 l
j 4-7 A
3
.. _ ~ -. -
.j iable 4C MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE j
PLANT: Dresden 2,3/ Quad Cities 1 FUEL TYPE:
7D230 i
Average Planar Exposure MAPLHCR PCT 0xidation (mwd /t)
(kW/ft)
(*F)
Fraction 200 14.3 2198 0.034 1,000 14.5 2197 0.034 5,000 14.7 2198 0.032 CT 10,000 14.5 2193
- 0.032 15,000 14.0 2198 0.074 20,000 13.7 2197 0.073 25,000 13.6 2198 0.070 Table 4D MAPLHGR VERSUS AVERAGE PLANAR E.VOSURE PLANT: Quad Cities 1 FUEL TYPE: EEiC - PU Average Planar Exposure MAPLHGR PCT 0xidation
_ (mwd /t)
(_kW/ft)
(*F)
Fraction 200 14.1 2198 0.042 1,000 14.3 2194 0.040 5,000 14.6 2197 0.039 l
10,000 14.2 2198 0.083 15,000 13.5 2198 0.112 20,000 13.3 25,000
~
2195 0.114 13.2 2195 0.116 l
4-8
i Table 4E MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE i'
PLANT: 'Dresden 2,3/ Quad Cities 1,2 FUEL TYPE: 8D250
.q Average Planer 4
l Exposure MAPLHGR PCT 0xidation
]
(mud /t)
(kW/ft)
(*F)
Fraction s
{
200 11.2 2106 0.024 1,000 11.3 2109 0.024 5,000 11.9 2169 0.029
.g 10,000 12.1-2179
. 0.029 15,000 12.2 2198 0.031 20,000 12.0 2199 0.031
~
25,000
'11.5 2148 0.027 30,000 10.6 2020
- 0.017 Table 4F MAPLHCR VERSUS AVERAGE PLANAR EXPOSLTE i
PLANT: Dresden 2,3/ Quad Cities 1,2 FUEL TYPE: 8D262 Average Planar i j Exposure MAPLHGR PCT 0xidation (mud /t)
(kW/ft)
('F)
Fraction i
200 11.1 2104 0.024 1,000 11.3 2107 0.024 5,000 11.9 2166 0.029 10,000 12.1 2175 0.029 15,000 12.2 2199 0.031 20,000 12.0 2199 0.032 25,000 11.6 2157 0.028 30,000 10.7 2042 0.019 t
4-9
,m..
---,-,.m. -.,...., - -.
-v
j
- N EDO-24146 Y
Table 4G MAPLHGR VERSUS AVERAGE PLANAR EXPOSLTE J.
PLANT: Dresden 2,3/ Quad Cities 1,2 FUEL TYPE:
8DRB265L Average Planar Exposure MAPLHGR
. PCT 0xidation-(mwd /t)
(kW/ft)
(*F)
Fraction 200 11.6 2189 0.035 1,000 11.6 2188 0.034 5,000 11.8 2198 0.034 10,000 11.9 2196 0.033
-15,000 11.9 2198 0.034 20,000 11.7 2195 0.034
~
25,000 11.3 2154 0.030 30,000 10.7 2075
, 0.023-Table 4H MAPLHGR VERSUS AVERAGE PLANAR EXPOSt%E PLANT: Quad Cities 1 FUEL TYPE:
Barrier LTA Average Planar Exposure MAPLHGR PCT 0xidat' ion
(! fwd / t)
(kW/ft)
(*F)
Fraction 200 11.6 2171 0.032 1,000 11.6 2178 0.033 5,000 12.0 2198 0.034 f
10,000 12.1 2195 0.033 15,000 12.0 2200 0.033 20,000 11.7 2187 0.032 25,000,
11.0 2115 0.056 30,000 10.5 2029 0.042 4-10
- x _ ~ : _=.=
7 NEDO-24146' 4
5.
DESCRIPTION OF MODEL AND INPUT CHANGES The only change between this ECCS analysis and the previous D2,3/QC1,2 analysis (Reference 12) is the use of the partial drill model. A description of this model is presented in Reference 13.
Approval for the use of this model is given in Reference 14.
.' j The addition of an alternate bypass flow path via holes drilled in the lower j
tieplate tu the BWR 3's provides them with the same bypass 'flowpaths as I
the BWR4's (i.e., same as BWR4's with core plate holes plugged and holes 1
drilled in the lower the plates).
Since the BWR3's do not currently include C,4 1
the LPCI modification, the ECC systems and the core configuration are the same as the BWR4 non-mod plants. The primary difference between these two groups of t
l' plants is that SWR 3's have a lower power density than BWR4's.
Thus, in this analysis credit is taken for flow through holes in the fuel assembly lower t
tieplates as in the lead plant (Duane Arnold).
O O
i I
4 1
i 1
5-1/5-2 l
.- + -. -.. - -
,__.-,-_.m
.-m
-...__--,.,_-,--.,.____.__._..m.
.,...,-.,,,-,7,_%
.--,-,.-y-,w._--,.-,y--,,
-. -,, - - - _ _ ~. _,..,.
=
NED0-24146 6.
CONCLUSIONS I
The LOCA analysis in accordance with the requirements of Reference 2, for non-lead plants with loop selection logic and fuel bundles with drilled lower tieplates in a full or partial core loading, is presented in Figures 1 through 5.
This analysis is for the maximum recirculation line suction break which is the most liniting break for this plant.
The characteristics that determine which is the most limiting break are:
(a) the calculated hot node reflooding time, (b) the calculated hot node uncovery time, and (c) the time of calculated boiling transition.
The time of calculated boiling transition increases with decreasing break size, since jet pump suction uncovery (which leads to boiling transition) is deter-mined primarily by the break size for a particular plant. The calculated hot node uncovery time also generally increases with decreasing break size, as it is primarily determined by the inventory loss during the blowdown.: The hot node reflooding time is determined by a number of interacting phenomena such as
(
depressurization rate, counter current flow limiting and a combination of available ECCS.
The period between hot node uncovery and reflooding is the period when the hot node has the lowest heat transfer. Hence, the break that results in the longest l
period during which the hot node remains uncovered results in the highest calcu-lated PCT.
If two breaks have similar times during which the hot node remains l
uncovered, then the larger of the two breaks will be limiting as it would have an earlier boiling transition time (i.e.,
the larger break would have a more severe LAMB / SCAT blevdewn heat transfer analysis).
Figure 6 shews the variation with 1 eak size of the calculated time the hot node remains uncovered for D2,3/QCl,2.
Based on these calculations, the DBA was 6-1
7 FEDO-24146 determined to be the break that results in the highest calculated PCT in the 1.0 ft to DBA region. Although the 34% DBA break has a slightly longer total j
core uncovered time, the resulting PCT is less than the DBA.
This is due to I
a much later boiling transition time associated with the 34% DBA and a later core uncovery time which results in a slower heatup rate due to a lower decay i
heat.
l For breaks smaller than 1.0 ft the calculated PCT's will be less than those calculated in the 1.0 ft to DBA break range, as has been demonstrated for the lead plant (Reference 11) for D2,3/QC1,2.
O~
The single failure evaluation showing the remaining ECCS following-an assumed failure and the effects of a single failure or operator error that causeu any manually controlled, electrically operated valve in the ECCS to move to a position that could adversely affect the ECCS are presented in Reference 15.
e 9
0 6-2
h O
O I
l 1200 80 800 3
40 2
m
~
3
.J g
TAF WATER LEVEL n
w w
m E
.4 ts i
f E
2 w
u m
e ta N
4 o
5 E
7 400 sap o
20 i
i I
J l
VESSE L PRESSURE l 0
0 100 200 yw) 4og g 0 TIME (soc,i v.
e Figure 1.
Water Level Inside the Shroud and Reactor Vessel Pressure Following a Recirculation Suction Line llreak, LPCI Injection Valve Failure, fireak Area = 4.18 ft2 (L13M)
. ~.
..... ~
2500 t
HIGH POWER AXI AL -
I HIGHEST POWER AXI AL PLANE I f
- LOWEST AXIAL PLANE TO EXPE RIENCE CPR 1.0 PRIOR 2000 TO JET PUMP UNCOVERY
/
- SPR AY COOLING
/
C
/
U w
rf g is00 5
/
n
/
w>
/
W i
/
- 5 O
= = " "
- HIGH POWER AXIAL 9
.n P s
PLANE UNCOVERED o
5 s
[
u s-v v
4 D
O 500 -
n
- ONSET OF SOILING TRANSITION O
2 4
6 10 20 40 00 100 200 400 TIME (sed) 9 Figure 2.
Peak Cladding Temperature Following a 4.18 f t' Recirculation Line Break, LPCI Injection Valve Failure (DBA) i
- JET PUMP UNCOVERY
- ONSET OF BOILING TRANSITION if I P k
10.000 f
I n
3 10W s..
b.
5*
27 i
9i 15 85 mO
- =
9 2 100
<a e>
>E
<E E3
- HIGH POWER AXIAL PLANE UNCDVERED w>
h Iw gg A
.)
z5 y
8 l
I 10 --
- ONSET OF LOWER PLENUM FLASHING l
l n
10 20 30 i
TIME (m)
Figure 3.
Fqti Rod Convective Heat Transfer Coefficient During Blowdown at the High Power Axial Node for a 4.18 ft2 Recirculation Line Suction Break (DBA) 6-5
.......u...
O-O i
t l
14 4
l
- LCWER PLENUM F LASHING 3
10 i
if l
E 3
m w
i E
O Uo 06 w
]
JET PUMP UNCOVERY -
Q e
?
I f
M i
z b
H>
1 1I o2 -
i
-O2 l
0 4
8 9*
12
.e t
'16 20 24 TIME AFTER BRE AK forc) 4 f
Figure 4.
Normalized Core Average Inlet Flow Following a 4.18 ft Recirculation 1,ine Suction Break (DBA) 1 i
.i
-n
~.
20 l
18 1
16 t
i j
1.4 9
JET PUMP bNCOVERY --
7
- NOTE 1 x
E 1.2 4
2 10 E
5 8
e h
3 A
j 08 i
7 5
06 NOTE 1: CPR = 1 AT SPACER 2 9 53 kw/ft AVERAGE PLANAR LINEAR HE AT GENER ATION R ATE ( APLHGR);
85% OF MAXIMUM AVERAGE PLANAR LINEAR HE AT GENERATION RATE O4 -
(MAPLHGH) l 02 -
I i
l l*
o I
2 3
4 5
6 y
TIME (soc)
I Figure 5.
Minimum Critical Power Ratio Following a 4.18 ft Recirculation Line Suction Break (DBA)
,