ML20024G801
| ML20024G801 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 08/20/1974 |
| From: | NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20024G796 | List: |
| References | |
| NUDOCS 9104300427 | |
| Download: ML20024G801 (37) | |
Text
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'u'u o Innediate - Irr*cdiate means that the required action will be initiated as soon as practicable O-DO D.
considering the safe operation of the unit and the imp-tance of the required action. o
>O CO E. Instrtm. cat Functional Test - An instrument functional tet merins the injection of a simulated DA instrtment channel response, starm, and/or N signal into the primary sensor to verify the proper initiating action.
ON_ ua 00 F. Instrument Calibrstion - An instrument calibration nean, the adjustment of an instrument signal "U output so that it corresponds vithin acceptable range, s'ccuracy, and response time to a known value (s) of the parameter which the instrunent monitors. Calibration shall eccocrpass the entire U instrument including actuation, alarm or trip. Femponsa time is not part of the routine instrtment calibration but will be checked once per cycle. G. Liniting Conditiccw for Operation (IEO) - De limiting cenditions for operation specify the minimun acceptable levels of system performance necessary to amare safe startup and cperation of the facility. When these conditions are met, the plant can be operated safely and abnormal situations can be safely controlled. II. Limiting Safety System Setting (LSSS) - %e limiting safety system settings are settirrs on instru-c entation which initiate the automatic protective actien at a level such that the safety limits will not be exceeded. The region between the safety limit and these settings represents margin with normal operation lying below these settings. ne margin has been established so that with proper cperation of the instrumentation, the safety limits will never be exceeded. I. Minimum Critical Heat Fluz and Power Ratios in-core ratio of critical heat flux (that
- 1. Minimum Critical IIeat Flux Ratio (MCHFR) - The lowest heat flux which results in transition boiling) to the actual heat flux.
- 2. Minimum Critical Power Ratio (MCPR) - We lowest in-core ratio of critical power (that power which cetises some point in the assembly to expericince the enset of transition boiling) to the bundle puer.
J. Mode - %c reactor mode is that which is established by the mode-selector switch. K. Operable - A system or component shall be considered operable when it is capable of performing its intended function in its required manner. L. Operating - Operating means that a system or cerpenent is perforning its required functions in its required manner. M. Operating Cvele - Interval between the end of one refueling outage and the end of the next subsequent refueling outage. 1.0 _
l. I 4
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3.0 Litf111NC. CONDITIO?G TOR O?ERATION 4.0 SURVEILIANCE REQUIEDID.'T3 - l 4 I ! Recir culation System I. Fecirculation System il. i I. Except as specified in 3.5.1.2 below, whenever 1. Once per month, when irristed fuel is in the { reactor with reactor coolant temperature greate. ; 1rradiated fuel is in the reactor, with reactor 4 0 coolant temperature greater than 212 F and both than 212 F and both reactor retirculetion l reactor recirculation pumps operating, the pumps operating, the recirculation systen cross l recirculation system cross tie valve interlocks tie valve interlocks shall be demonstrated to j i shall be opersble. be operable by verifying that the cros tie valves cannot be opened using the normal control
. 2. The recirculation systen cross tie valve icter- switch. ! locks may be inoperable if at least one cre'ss tie valve is maintained fully closed. 2. When a recirculation system cross tie valve interlock is inoperable, the position of at l 1 east one fully closed cross tie valve.shell be recorded daily. ;
4 i t t 3.5/4.5 jogg sty t
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Bnnes Continued 3 5: G. Energency Cooling Availability he purpose of Specification G is to assure that sufficient core cooliry, equipment is available at all timos. It is during r= fueling outagen that enjor naintenance is perforred and during such tire that all core and contairment cooling subsystems rny be out of service. Spacification 3 5.G.3 allows all core and containmant coo'ing subsystems to be Inopamble prmrided no work is being done which has the potenttal for draining t e reactor vessel. D un events requiring corn cooling are precluded. Specification 3 5.G.4 mcognizes that concurrent with control rod drive enintenance during the refuelins; outage, it eny be necessary to drain the suppression chnnber for maintensnee or for the inspection required by Specification h.7. A.I. In this situation, a sufficient inventory of water is enintained to assure adequate com cooling in the unlikely event of lor.s of contml rod drive housins; or instru=ent thimble seal integrity. H. Deleted T. Recirculation System he capacity of the thergency Core Coolant System is betsed en the potential consequences of a double ended recirculation line break. Such a break involves 3.9 sq. ft. when the cross tie valves are closed and 5.3 sq. f t. when the cross tie valves are open. Specification 3.11.A is based on an ECCS evaluation assuming a break area of 3.9 sq. f t.; the limitations of 3.11.A do not apply to the larger break area. Derefore, at least one cross tie valve must remain closed with two pterp operation to reduce the potentini break area. De cross tie valve is aUowed to be open during one p*my operation. Widt only one pu=p, rated power cannot be achieved. Under these conditions, the erpec*ed peak clad tenperature during a loss of coolant accident is less than that for two pump operation with the cross tie valve closed. 3.5 BASES 113 REV
Bases 4.5: The testing interval for the core tnd contain-ant moliri rystems is bared en a quantitative reliability analysis, judment, and p racticality. The core renlin; ,ystems h,vc not been designed to be fully testable during creration. For exa. ple, the core rpray final dission valves do not open until reacter pressure has fallen to 450 psig; thus, during operation even if hirJi dryvell pressure vare simulated, the final valves would not open. In the case of the HPCI, automatic initiation during power operation vo*21d [ result in pur ping cold water into the reactor vestal, which is not desirable. The systems can be automatically actuated during a refueling outage and this will be done. To increase the availability of the individual coeponents of the core and containment cooling systems, the cocponents which make up the system, i.e., instrumentation, pu ps, valve operators, etc. , are tested more frequently. The instrumentation will initially be functionally testad once per mnth until a trend is established t and thereafter according to Figure 4.1 (see Section 3.1/h.1) with an interval net greater than three months. The pumps and motor-operated valves are tested each conth to assure their operability. The j combination of a simulated automatic actuation test each refuelirg cycle arri monthly tests cf the pu. rs ani valve operstors is deemed to be adaquate tastin6 of these syntans. l ) With cotcponents or subsystems out-of-service, overall ccre and contairnent coolin6 reliability is main-tained by demonstrating the operability of the resaining cooling equipment. The degree of operability l to be demonstrated depends on the nature of the reason for the out-of-service equipment. For routine 1 out-of-service periods caused by preventative maintenance, etc., the pump and valve operability checks ] will be performad to demonstrate operability of the remainin6 cocoonents. However, if a failure, design deficiency, etc., caused the out-of-service period, then the deconstration of operability should be 4 thorou6h enough to assure that a similar probim does net exist on the recaining cmponents. For example, l if an out-of-service period were caused by failure of a rurp to deliver rated capacity due to a desip ! deficiency, the other pumps of this type right be rubjectad to a flow rate test in addition to the l. operability checks. l ] 4 e s 4.5 IdSES ,
__ _ _ - _ . --_ _ - _= 4 i l 3.0 LUf1 TING CmDITIOf:S FOR OPERATIONS 4.0 SURVEILIANCE REQUIRDfD.TS 3.11 REACTOR FUEL ASSDiBLIES 4.11 RFACTOR FUEL ASSI2tBLIES l Applicability Applicability i Ihe Limiting Conditions for Operation The Surveillance Requirements apply to associated with the fuel rods apply to the parameters which monitor the fuel i those paraecters which tronitor the fuel rod operating conditions. i red operating cenditions. Objective Objective {_ The objective of the Limiting Conditiot-s The objective of the Surveillance Require- ents } for Operation is to assure the perfor- is to specify the type and frequency of surveil-i 1 mance of the fuel rods. lance to be applied to the fuel rods. 3 Specificatiens Specifications j A. Average Planar Linear IIcat Geneta- A. Average Planar Linear IIcat Genera-tion Rate (APUIGR) ation Rate (AFUiGR) During steady state power operatien, The APUIGR for each type of fuel as a
- the APUIGR for each type of fuel as function of average planar exposure shall a function of average planar exposure be determined daily during reactor operation shall not exceed the limiting value at 2 257 rated thermal power.
shown in Figures 3.11-1. If at any time during steady state power opera- ~ 3 tion it is determined that the limit-1 ing value for APUIGR is being execcJ-j ed, action shall be taken imediately to restore operatien to within the q prescribed limits. 3.11/4.11 189 8 REV l J
~4 . 0 LTMITING CONDIT1r*4S FOR OPEPATION 4.0 st*RVEILIANCE REQUIPEME'iTS 7 Incal UIGR n. Tocal U?CR During steady state power operation, the linear The local U1GR as a function of core height shall heat generation rate (U!GR) of any rod in any be checked dMily during reactor operatien at C fucI assembly at any axial location shall not 237 of rated thermal power.
exceed the maximum allowable UICR as calculated by the following equation: UIGR S DIGR 1-max d g P max LT j 1 UIGR d = Design U1GR
= 17.5 kw/ft for 7x7 fuel
= 13.4 kv/ft for 8x8 fuel f i l
i max = Maximum power spiking penalty ( Pi
= 0.026 for 7x7 f-e7
= 0.021 for 8x8 fuel LT = Total core length = 12 ft L = Axial position above bottom core If at any time during steady state power operation it is deternined that the limiting value of UfGR is being exceeded, actien shall be taken ir=tedi-ately to restore operation to within prescribed a its.
l' 189 C 3.11/4.11 EU
l I l t t 3.0 LIMITING CONDITImS FOR OPEPATIW 4.0 SliRVEILTAtiCE REQUIREMENTS I f l C. Minimum Critical Power Ratio (MCPR) C. Minirem Critical Power Ratio (MCFR) ! During steady state power operation, MCPR shall be determined weekly l MCPR shall be 1 1.19. If at any during reactor power operation at f time during steady state power 2 257. rated thenaal power. t operation it is detetufned that the , i limiting value for MCPR is being l exceeded,-action shall be taken irraediately to restore operati<m within the prescribed limits. I i i 1 i I i 1 i ! 4 I 4 f i l 'l i i d !
' i i I i
l 4 ; i L j i 1 189 D i 3.11/4.11 REV [ T
Bases 3.11 . A. Average planar Linear 11 eat Generation Rate (APUICR)
- This specification assures that the peak cladding tenperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in the 10CFR50, Appendix K.
The peak cladding temperature following a postulated loss-of-coolant accident is prLnarily a i function of the average heat generation rate of al.' ' ' rods of a fuel assembly at any axial location and is only dependent secondarily on the ro. rod power distribution within an assembly. Since exp.-cted local variations in power distribution within a fuel assarbly affect ,' the calculated peak cladding temperature by less than + 20"F relative to the peak tecperature for a typical fuel design, the limit on the average linear heat generation rate is sufficient to assure that calculated temperatures are within the 10CFR50 Appendix K limit. The limiting T value for APIRGR is given by this specification. It is recognized that ApulGR is a calculated parameter that is not continually monitored and alarmed directly during core power distribution changes. If at the time of the calculation it is found that the limits are being exceeded, there is always on action which will return the average planar UlGR to within prescribed limits, namebr power reduction. Under most circumstances, this will not be the only
- alternative. Therefore, the only way to have a reportable Abnormal Occurrence is to knowingly allow operation beyond the prescribed limits without taking the necessary action to restore the average planar UIGR to within prescribed limits.
, B. Local IRGR This specification assures that the linear heat generation rate iu any rod is less than the design linear heat generation if fuet pellet densification is postulated. The power spike penalty specified is based on the analysis presented in Section 3.2.1 of Reference 1 and in References 2 and 3, and assumes a Ifnearly increasing variation and axial gaps between core bottom and top and assures with .. a 9n confidence, that no more than one fuel rod exceeds the design linear heat generation rate due to power spiking. 3.11 BASES 189 E REV
Bases 3.11 teontinued) t It is recognized that IJIGR is a calculated parameter that is not continually nonitored and alarrrd directly during core power-distribution changes. If at the time of the calibration it is found that the limits are being exceeded, there is always an action which will return the Li!GR to within prescribed limits, namely power reduction. Under rost circumstances, this will not be the only alternative. Therefore, the only way to have a reportable Abnomal Occurrence is to knowingly allow operation beyond the prescribed limits without taking the necessary action to restore the IJIGR to within prescribed limits. C. Minimum Critical Power Ratio (MCPR) The ECCS evaluation presented in Reference 4 assumed the worst steady state MCPR to be 1.19. This value is chosen such that the onset of transition boiling is avoided in operational transients. Since lower values of MCPR have not been considered in the ECCS evaluation, operation at that condition is not pemitted. 1 It is recogntzed that NCPR is a calculated parameter that is not continually renitored and i alamed directly during core powe r distribution and thermal-hydnulic changes. If at the - time of the evaluation it is found that the linits are being exceeded, there is always an action i which will return the MCPR to within prescribed 11r'its, namely power reduction. Under most circumstances, this will not be the only alternative. There fore, the only way to have a reportable Abnomal Occurrence is to knowingly allev operation beyond the prescrlh 1 limits without taking the necessary action to restore the MCpR to within prescribed limits. Re fe rences.
- 1. " Fuel Densification Effects in General Electric Boiling Water Reactor Fuel," Supplements -
6, 7, and 8, NEDM-10735, August, 1973.
- 2. Supplement 1 to Technical Report on Densificatien of General Electric Reactor Fuels, December ~ '
14,1974 (USAEC Regulatory Staff)
- 3. Cocaunication: V A Moore to I S Mitchell, '%dified CE Model fo- Fuel Densification,"
Docket 30-321, March 27, 1974
- 4. "Monticello Nuclear Generating plant Loss-Of-Crolant Accident Analysis Conformance with 10 CFR 50 Appendix K, August 1974," L 0 Mayer ('SP) to J F O' Leary, August 20, 1974.
2.11 BASES 189 F REV
Bases 4.11 The APUIGR and local flIGR shall be checked daily to detemine if fuel burnup, or control rod movement has caus J chaages in power distribution. Since changes due to burnup are slow, and only a few control rods are removed daily, a daily check of power distribution is adequate. For a 11:niting value to occur below 257. of rated themal power, an unreasor ably large peaking factor would be required, which is not the case for operating control rod sequences. At core themal power levels less than or equal to 25%, the reactor will be operating at minimum recirculation i pt-np speed and the moderator vold content wili be very small. For all designated control rod patterns 1 whici may be employed at this condition, plant operating experience and thermal hydraulic analysis indicated that the resulting MCPR value is in excess of requirements by a considerable margin. With this low void content, any inadvertent core flow increase wou' ' only place operation in a more conservative mode relative to MCPR. The weekly requirement for calc .M - r '. shove 252 rated thermal power is sufficient since power distribution shifts are very slow when there rne not een significant power or control rod changes. The APUICR and local UICR are more sensitive to tocal .c changes tlun MCPR, They will airo generally reach limiting values prior to MCFR reaching la Mmitir -lu , In chis way APUICR and local UlGR serve as early indicators of the MCPR behavior. 1ht - .a.1. is s.unsistent to require a calculation of APulGR and local UlGR on a more frequent basis than itCPR. t t [ i , I 4.11 BASES 189 C REV l
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1 i I EX111151T C 1 I 1 MONTICELLO NUCLl%R GENERATING PLANT l LOSS-OF-COOIANT ACCIDENT ANALYSES ) CONF 0lNANCE WITil 10CFR 50 APPENDIX K l AUGUST 1974 i l l l' 1.
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2 Discussion Pres (nted in the following document are the results of the loss-of-coolant accident analysis of the Monticello Euclear Power Station. The analysis was performed using General Electric calculational models which are consistent with the requirements of Appendix K of 10 CFR part 50. A complete discussion of each code employed in the analysis is presented in Section 11 of Reference 1. Input to the Analysis A list of the significant plant input parameters to the loss-of-coolant accident analysis is presented in Table 1. Table 1 SIGSIFICANT INPUT PARAMETERS TO Tile LO2-OF-C001ET ACCIDENT ANALYSIS PLAST PARAMETEF.S: 1703 Mvt which corresponds to Core Thermal Power . . . . . . . . . . . . . . 102 '. Licensed core power Vessel Steam Output . . . . . . . . 6,913 X 10' tbm/h which corresponds to 10 2 , ^, of Licensed core power Vessel Steam Dome Pressure . . . . . . . .. . . 1040 psia Design Basis Recirculation Lint Break Area . . . . 3.0 ft 2 FULL PARAMETERS: Peak Technical Initial Specification Design Minimum Linear Heat Axial Critical Generation Rate Peaking Power Fuel Lundle Fuel Iype Geometry (kv/ft) Factor Ratio 7x7 17.5 1.57 1.19 Initial Core 7x7 17.5 1.57 1.19 Reload 1 8x8 13.4 1.57 1.19 Reload 2 i A more detailed list of input to each model and its source is presented in Section 11 of heterence 1. 1
3 Results of the Analysis The results of the analysis are presented in the order in which they are calculated. The prese,*ation of the results is divided intoThese four major portions portions are: according to the model . rom which the output is obtained. A, Calculated by the Short-Term Thermal-Hydraulic Model (LAMB) B. Calculated by the Transient Critical Power Model (SCAT) C. Calculated by the Long-Term Thermal-Hydraulic Model (SAFE) D. Calculated by the Core Heatup Model (CRASTE) A summary of the results is presented in Table 2. At the MAPLHCR* employed in the analysis, the most severe pipe break yields a calculated peak cladding temperature less than or equal to 2200 F, a calculated maximum local metal-water reaction less than or equal to 17% and a calculated core-wide metal-water reaction less than or equal to 1%. Compliance with the 10CFR50.46 criteria for coolable geometry and long-term cooling has been shown in ELference 1. The reactor is therefore fully in conf ormance with 10CFR50.46 and Appendix K with operations at the MAPulGR used in the analysis. Details of the MAPLHCR values used as a function of fuel exposure are given in Figure (s) D5 for each fuel type in the reactor. These values, if more limi ting than other design parameters , represent limits for operation to ensure conformance with 10CFR50.46 and Appendix K.
- Maximum (Bundle) Average Planar Linear Heat Generation Rate Table 2
SUMMARY
OF RESULTS OF THE MONTICELLO NUC1IAR P0k'ER STATION APPENDIX K LOSS-OF-C00LAST ACCIDENT ANALYSIS Design BLsis Break Highest Temperature Maximum Metal- Intermediate break Single Water Reaction Break Core Average Local PCT ( F) Area (ft2) Failure PCT ( F) l l l LPCI Inj. l valve 2200 0.3 % 7% Failure (a) 1700 0.07 HPCI (b) (c) .ysters Available - 2CS t HPCI + ADS
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1 l i . A, APPEND 1X K SHORT-TEIN THEPMAL-HYDPAt11C ANALYSIS , l l General D(scription of the I AMb Code The 1AMB code is a model which is used to analy:e the short-term thermodyre ic and thermo-hydraulic behavior of the coolant in the vessel during a postuted loss-of-coolant accident. In particular, LAMD predicts the core flow, core inlet i i f enthalpy and core pressure during the blowdown prior to the end of lower plenum l flashing ( ~ 20 seconds). For a detailed description of the model and a discussion l regarding sources of input to the model refer to the " LAMB Code Documentation" l portion of Section 11 of Reference 1. I Results of IAMB Analysis j Presented in this section are results of the loss-of-coolant accident analysis which are calculated by LAMB. These results include: Figure Parameter Core Average Inlet Flow Rate (Normalized to unity i at the beginning of the accident) following a Design Basis Accident A-1 Core inlet Enthalpy following a Design Basis Accident A-2 l Core Average Pressure following a Design Basis Accident ^~3 l These results are input to the SCAT code discussed in Section E. B. APPENDIX R TRANSIENT CRITICAL POWER ANALYSIS ( Ceneral Description of the SCAT Code The SCAT code is used to evaluate the short-term thermal-hydraalic response of the coolant in the core during a postulated loss-of-coolant accident. In particular l the convective heat transfer process in the thermally limiting fuel bundle is analyzed I during the transient. For a detailed description of the model and a discussion f regarding sources of input to the model refer to the " SCAT Code Documentation" portion l of Section II of Reference 1. t l Results of the SCAT Analysis I Presented in this section cre results of the loss-of-coolant accident analysis which l are calculated by SChT. These results include: t Figure l Parameteg Minimum Critical Power Ratio following a Design B-1 Basis 1.ccident Convective heat Iransfer Coef ficient following a Design Basis Accident B-2 These results are used as input to the CHASTE code discussed in Section D, l
C. AlTINDDiJ 1m TERM THEPMAl-HYDMUuc ANALYSIS General Description of SAFE Code The SAFE code is a model which is used to analyze the long-term thermodynamic l behavior of the coolant in the vessel during both small and large breaks. Since the I calculational proceiure of the loss-of-coolant accident analysis dif fers depending on whether or not a break is classified as "small" or "large", it is appropriate to distinguish between the two classifications of breaks. A small break is defined as tt. size break for which nucleate boiling heat transfer exists in the core until the heat fluxes are below the pool boiling critical power condition. This occurs approximately 20 to 25 seconds after the break. For small breaks, core heatup is therefore based solely on the uncovery and recovery of the fuel and the duration of spray cooling all of which are predicted by the SAFE core. For the "large" break analysis, the IA.MB and SCAT codes are employed to determine the time of boiling transition and the post-boiling-transition convective heat transfer during the blowdown. The SAFE code calculates the uncovery and reflooding of the fuel and the duration of spray cooling. For a detailed description of the model and a discussion regarding sources of input to the model refer to the " SAFE Code Documentation" portion of Section II of Reference 1. I i Results of the SAF Analysis Presented in this tection are results o f the loss-of-coolat.t accident analysis j which are calculated by SAFE, These results include: l Figure Parameter I Water Level inside the Shroud and Reactor Vessel Pressure following a Design Basis Accident C-1 l Fuel Rod Convective Heat Transfer Coefficient C-2 following a Design Basis Accident Water Lcvel inside the Shroud and Reactor Vessel Pressure following a Small Break of the Recirculation C-3 Line P. APU; DIX K CORE PEATUP ANALYSIS General bescription of CHASTE code The transient thermal response of the core to a loss-of-coolant accident calculated by CliASTE can generally be broken down into four stages: (1) fuel pin temperature redistribution; (2) fuel rod bundle temperature redistribution; (3) metal-water reaction heatup; and (4) core standby cooling system effects. Phenomena occurring during these stages that are considered in the analysis are described below. Fuel Pin Temperature Fedistributien Following a reactor shutdown, a large heat source is still available within the core in the form of sensible heat in the fuel. This is represented by the temperature profile in the fuel rod. Initially, the temperature profile is steep because of the
high power generation rates during nonal operatior . lollowing the shutdown, the sensibic heat in the fuel will be redistributed by thenaal conduction within the fuel and cladding and by convection and radiation in the gap between fuel and cladding, with the amount of heat removed being dependent on surface conditions. At the end of three or more fuel time constants (fuel thernal time constant is about 8 to 10 seconds), the radial temperature profile in the fuel pin is almost flat, consistent with the low fission product decay power generation. Puel Rod Bundle Temperature Redistribution As the cladding temperature increases and the core coolant void fraction approaches unity, radiant heat transmission between rods and the channel wall tends to equalize the temperature of all rods at a given axial position. The total energy in the core continues to increase during this period due to continuing fission product decay. Metal-Water Reaction Ucatup The fuel pin cladding is made of an alloy of Zircaloy, which reacts with steam at high temperatures. The zircaloy-steam checmical reaction rate is exothermic and strongly dependent upc.. the reaction temperature. The tamperature dependence is exponential and the rate of reaction becomes significant at cladding temperatures in the range of 22000r or higher. Dmercency Core Cooline System (ECCST Effects Redundant emergency core cooling systems performance for a given LOCA is dependent t upon the conditions of the accident. The core cooling systems will provide suf ficient cooling to prevent excessive cladding heatup. The primary purpose of the core heatup analysis is to determine the effectiveness of the emergency core cooling systems. For a detailed description of the CHASTE model and a discussion regarding sources of input to the model refer to the " CHASTE Code Documentation" portion of Section 11 , of Reference 1. i l Results of CHASTE Analysis l Presented in this section are results of the loss-of-coolant accident analysis which are calculated by CRASTE. These results include: l Parameter Figure l l Feak Cladding Temperature following a Design l Basis Accident D-1 peak Cladding Temperature following a small Break of the Recirculation Line D-2 Peak Cladding Temperature versus Break Area D-3 Peak Cladding Temperature versus Planar Average Exposure D-4 Maximum Average Planar Linear Heat Generation versus Planar Average Exposure D-5 Figure D-4 shows the calculated peak cladding temperature as a function of i
~ _ _. .- - - - . .
exposurt if the fuel bundle is operated at the average planar linear heat generation rate plotted in Figure D-5. For discussion purposes, Figures D-4 and D-5 can be separated into three ranges of average planar exposure:
- 1. No ECCS Limit In many cases, (that is, over a range of planar average exposures), the fuel has the capability to operate at the peak technical specification linear heat generation rate (Table 1) and the calculated peak cladding temperatures during the postulated accident are less than 2200 F. In general, this occurs early in the life of the fuel when the gap conductance is high and the internal fuel rod pressure is low. Over this exposure range, no maximum ALHGR is plotted in Figure D-5 and the fuel can be operated without a power restriction in conformance with Appendix K.
The resulting maximum temperatures and maximum oxidation percentages are plotted in Figure D-4.
- 2. ECCS Limit Applies In this exposure range, the calculated temperature would exceed 22000 F if the fuel were operated at the peak technical specification LHCR, so it is necessary to reduce the average planar power to meet Appendix R requirements.
The resulting maximum average planar linear heat generation rate (HAPLHGR) is plotted in Figure D-5 and the resulting temperatures (2200 F) and exidation percentages are shown in Figure D-4.
- 3. Fuel Depiction Limits Apply After the initial cycle of operation, depleted fuel assemblies operate at lower power relative to fresh reload fuel. Consideration of this effect has been included in the GEGAP-III gap conductances, which are input as initial conditions in analysis of the postulated LOCA by employing conservative estimates of the maximum duty of the individual fuel rods in the highest power assembly (see Section 4.23 of Reference 2). For consistency with the GEGAP-III calculation of gap conductance and rod fission gas inventory, the GEGAP-III assumed exposure history for the peak power rod in the bundle was employed to determine the peak LHGR late in life. The average planar power was determined f rom the peak LHCR and the highest local peaking factor. The average planar po'wer used in the CHASTE calculation is shown in Figure D-5, and the resulting temperatures and oxidation percentages are shown in Figure D-4. The apparent limit imposed by this part of the curve should be of no practical concern because it is improbable that any particular fuel type l would be capable of operating at or above the presented power level. The planar power used in the analysis is, however, shown as a limit because a conformance calculation has not been made in excess of that value.
1
e 8-REFERENCES 1, General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50 Appendix K, NED0-20566 (to be issued)
- 2. GEGAP-Ill: A Model for the Prediction of Pellet-Cladding Thermal Conductance in Bk'R Fuel Rods, NEDO-20181, November 1973, t
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