ML19354C423

LOCA Analysis Rept for Facility, Class 1,Revision 1

ML19354C423
Person / Time
Site:	Monticello
Issue date:	12/31/1980
From:	GENERAL ELECTRIC CO.
To:
Shared Package
ML112970812	List: ML112970811 ML112970813 ML19345F249 ML19354C423
References
80NED296, NEDO-24050-1, NEDO-24050-1-R1, NUDOCS 8102170019
Download: ML19354C423 (37)

Text

NEDO-24050-1 80NED296 Class I Rev. 1 December 1980 LOSS-OF-COOLANT ACCIDENT ANALYSIS REPORT FOR MONTICELLO NUCLEAR GENGATING PLANT

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I NUCLEAR POWER SYSTEMS DIVISION e GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNI A 95125 GEN ER AL $ ELECTRIC g/ b 1 n o a !8

IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT

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Please Read Carefully The only undertakings of General Electric Company respecting informtion in this document are contained in the contract betueen Northern States Pouer Company and General Electric Company and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than Northern States Pouer Company or for any purpose other than that for uhich it is intended, is not authorised; and uith respect to any unauthorised usc, General Electric Company makes no representation or oarranty, and assumes no liability as to the cor"pleteness, accuracy, or usefulness of the inforr"ation contained in this document.

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NED0-24050-1 CONTENTS l

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1. INTRODUCTION 1-1 l

2. LEAD PLANT SELECTION 2-1

3. INPUT TO ANALYSIS 3-1

4. LOCA ANALYSIS COMPUTER CODES AND RESULTS 4-1 l

i 4.1 Results of the LAMB Analysis 4-1 4.2 Results of the SCAT Analysis 4-1 4.3 Results of the SAFE Analysis 4-1 4.4 Results of the REFLOOD Analysis 4-2 4.5 Results of the CHASTE Analysis 4-3 4.6 Methods 4-4

5. DESCRIPTION OF MODEL AND INPUT CHANGES 5-1

6. CONCLUSIONS 6-1

7. REFERENCES 7-1

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l NEDO-24050-1 TABLES l

_ Table Title Pm Significant Input Parameters to the Loss-of-Coolant Accident Analysis 3-1 2 Summary of Break Spectrum Results 4-5 l

l l 3 LOCA Analysis Figure Summary - Non-Lead Plant 4-5 4A MAPLHCR Versus Average Planar Exposure (8DB219) 4-6 4B MAPLHGR Versus Average Planar Exposure (8DB250) 4-6 4C MAPLHGR Versus Average Planar Exposure (8DB262) 4-7 4D MAPLHGR Versus Average Planar Exposure (8DRB265) 4-7 4E MAPLHGR Versus Average Planar Exposure (8DRB282) 4-8 4F MAPLHGR Versus Average Planar Exposure (P8DRB265) 4-8 4G MAPLHGR Versus Average Planar Exposure (P8DRB482) 4-9 4H MAPLHGR Versus Average Planar Exposure (P8DRB284) 4-9 l

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i NEDO-24050-1 ILLUSTRATIONS Figure Title Pm la Water Level Inside the Shroud and Reactor Vessel Pressure Following a 1.4 ft2 Recirculation Line Suction Break, LPCI Injection Valve Failure (34% pBA) (LBM) 6-4

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I lb Water Level Inside the Shroud and Reactor Vessel Pressure Following a 4.0 ft2 Recirculation Line Suction Break, LPCI Injection Valve Failure ODBA) 6-5 2a Peak Cladding Temperature Following a 1.4 ft Recirculation Line Suction Break, LPCI Injection Valve Failure, Break Area = 34% DBA (LBM) 6-6 2b Peak Cladding Tempeature Following a 4.0 ft Recirculation Line Suction Break, LPCI Injection Valve Failure CDBA) 6-7 i

3a Fuel Rod Convective Heat Transfer Coefficient During 2

j Blowdown at the High Power Axial Node Following a 1.4 ft Recirculation Line Suction Break, LPCI Injection Valve

! Failures (34% DBA) 6-8 3b Fuel Rod Convective Heat Transfer Coefficient During l j

! Blowdown at the High Power Axial Node for a 4.0 ft2 Recirculation Line Suction Break, LPCI Injection Valve Failure CDBA) 6-9 )

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4a Normalized Core Average Inlet Flow Following a 2.4 ft Recirculation Line Suction Break (60% DBA) 6-10 l l

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) 4b Normalized Core Average Inlet Flow Following a 4.0 ft

, Recirculation Line Suction Break (DBA) 6-11 '

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l 2 5a Minimum Critical Power Ratio Following a 2.4 f t Recirculation Line Suction Break (60% DBA) 6-12

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' 2 Sb Minimum Critical Power Ratio Following a 4.0 ft Recirculation Line Suction Break (DBA) 6-13 6 Variation with Break Area of Time for Which Hot Node Remains Uncovered 6-14 l

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1. INTRODUCTION The purpose of this document is to provide the results of the loss-of-coolant accident (LOCA) analysis for the Monticello Nuclear Generating Plant (Mc nticello) with a partial core loading of reload fuel with holes drilled in the lever tie-plates. The analysis was performed using approved General Electric (GE) calculational models.

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This reanalysis of the plant LOCA is provided in accordance with the NRC i t

j requirement (Reference 1) and to demonstrate conformance with the ECCS accep- [

j tance criteria of 10CFR50.46. The objective of the LOCA analysis contained I herein is to provide assurance that the most limiting break size, break loca-j tion, and single failure combination has been conaidered for the plant. The required documentation for demonstrating that these objectives have been

, satisfied is given in Reference 2. The documentation 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 j and CRASTE computer codes which have been used ir. this analysis. In addition, a model which takes into account the effects of drilling alternate flow path i l

holes in the lower tieplate of the fuel bundle and the use of such fuel bundles l in a full or partial core loading is described in References 7, 8, and 9. Thio 1

model was also approved in Reference 4.

l The specific changes as applied ao Monticello are discussed in more detail in later sections of this document. This analysis takes credit for the effects of prepressurization of the fuel rods in appropriate fuel types.

l Plants are separated into groups for the purpose of LOCA analysis (Reference 10).

Within each plant group there will be a single lead plant analysis which pro-vides the basis for the selection of the most limiting break size yielding the -

highest peak cladding temperature (PCT). Also, the lead plant analysis pro-vides an expanded documentation base to provide added insight into evaluation 1-1

NED0-24050-1 of the details of particular phenomena. The remainder of the plants in that group will have non-lead plant analyses referenced to the lead plant analysis.

This document contains the non-lead plant analysis for Monticello, which is now a BWR/3 in the BWR/4 group of plants with loop selection logic, and is con-sistent with the requirements outlined in Reference 2.

The same models and computer codes are used o evaluate all plants. Changes to these mode)s will cause changes in phenomenological responses that are i similar within any given plant group. The difference in input parameters is not expected to result in significantly different results for the plants within .. given group. Emergency core cooling system (ECCS) and geometric differences between plant groups may result in different responses for diffc:

ent groups but within any group the response will be similar. Thus, the lead plant concept is still valid for this evaluation.

NEDO-24050-1

2. LEAD PLANT SELECTION Lead plants are selected and analyzed in detail to permit a more comprehensive review and eliminate unnecessary calculations. This constitutes a generic analysis for each plant of that type which can be referenced in subsequent plant submittals.

The lead plant for Monticello with drilled fusi is Duane Arnold. The justifica-tion for categorizing Monticello in this group of plents and the lead plant analysis for this group is presented in Reference 11.

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NEDO-24050-1

3. INPUT TO ANALYSIS A list of the significant plant input parameters to the LOCA analysis is presented in Table 1.

Table 1 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLA!C ACCIDENT ANALYSIS Plant Parameters:

l l Core Thermal Power 1703 MWt, which corresponds to 102% of rated core power Vessel Steam Output 6.91 x 106lbm/h, which corresponds to 102% of rated core power Vessel Steam Dome Pressure 1040 psia Recirculation Line Break Area 1.4 ft2 (34% DBA), 4.0 ft2 (DBA) for Large Breaks - Suction Number of Drilled Bundles 292 Fuel Parameters:

Peak Technical Initial Specification Design Minimum Lf.near Heat Axial Critical Fuel Bundle Generation Rate Peaking Power Fuel Type Geometry (kW/ft) Factor Ratio

• A. 8DB219L 8x8 13.4 1.57 1.24 B. 8DB250 8x8 13.4 1.57 1.24

[ C. 8DB262 8x8 13.4 1.57 1.24 D. 8DRB265L 8x8R 13.4 1.57 1.24 E. 8DRB282 8x8R 13.4 1,57 '1.24 F. P8DRB265L P8x8R 13.4 1.57 1.24 G. P8DRB282 P8x8R 13.4 1.57 1.24 H. P8DRB284LB P8x8R 13.4 1.57 1.24 l

• To account for the 2% uncertainty in bundle power required by Appendix K to 10CFR50, the SCAT calculation is performed with an MCPR of 1.22 (i.e. ,

1.24 divided by 1.02) for a bundle with an initial MCPR of 1.24.

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NEDO-24050-1

4. LOCA ANALYSIS COMPUTER CODES AND RESULTS 4.1 RESULTS OF THE LAMB ANALYSIS This code is used to analyze the short-term blowdown phenomena for large postulated pipe breaks (breaks in which nucleate boiling is lost bsdure the water level drops and uncovers the active fuel) in jet pump reactors. The LAMB output (core flow as a function of time) is input to the SCAT code for calculation of blowdown heat transfer.

The LAMB results presented are:

e 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 int.o SCAT which calculates heat transfer coefficients for input to the core heatup cude, CHASTE.

The SCAT results presented are:

o Minimum Critical Power Ratio following a Large Break, o Convective Heat Transfer Coefficient following a Large Break, l

4.3 RESULTS OF THE SAFE ANALYSIS This code is used primarily to track the vessel inventory and to model ECCS performance during the IOCA. The application of SAFE is identical for all break sizes. The ccde is used during the entire course of the postulated accident, but after ECCS initiation, SAFE is used only to calculate reactor system pressure and ECCS flows, which are pressure dependent.

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NEDO-24050-1 1

j The SAFE results presented are:

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! e Water Level inside the Shroud (up to the time REFLOOD initiates) and i

Reactor Vessel Pressure

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j 4.4 RESULTS OF THE REFLOOD ANALYSIS j

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< This code is used across the break spectrum to calculate the system inventories i

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" ,

l 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 DBA. More precise modeling of coolant level behavior is thus requested principally to determine the contribution of vaporization in the fuel assemblies to the I counter current flow limiting (CCFL) phenomenon at the upper tieplate. The differences from the DBA-REFLOOD analysis are:

1. The non-DBA version calculates core water level more precisely than f the DBA version in which greater precision is not necessary.

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l 2. The non-DBA version includes a heatup model similar to but less detailed than that in CHASTE, designed to calculate cladding temper-ature during the small break. This heatup model is used in calcu- r lating vaporization for the CCFL correlation, in calculating swollen level in the core, and in calculating the peak cladding temperature.

The REFLOOD results presented are:

e Water Level inside the Shroud e Peak Cladding Temperature and Heat Transfer Coefficient for breaks calculated with small break methods 2

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NED0-24050-1 4.5 RESULTS OF THE CHASTE ANALYSIS This code is used, with suitable inputs fram the other codes, to calculate the fuel cladding heatup rate, peak claddiag temperac ve, peak local cladding oxidation, and core-wide metal-water reaction for large breaks. The detailed fuel model in CHASTE considers transient gap conductance, clad swelling and rupture, and metal-water reaction. The empirical core spray heat transfer and channel vetting correlations are built into CHASTE, which solves the transient heat transfer equations for the entire LOCA transient at a single axial plane l

in a single fuel assembly. Iterative applications of CHASTE determine the maximum permissible planar power where required to satisfy the requirements of 10CFR50.46 acceptance criteria.

The CHASTE results presented are:

e Peak Cladding Temperature versus time e Peak Cladding Temperature versus Break Area e Peak Cladding Temperature and Peak Local Oxidation versus Planar Average Exposure for the most limiting break size e Maximum Average Planar Heat Generation Rate (MAPLHGR) versus Planar Average Exposure for the most limiting break size A summary of the analytical results is given in Table 2. Table 3 lists the l

figures provided for this analysis. The MAPLHGR values for each fuel type l in the Monticello core are presented in Tables 4A through 4H.

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NEDO-20050-1 4.6 METHODS In the following sections, it will be useful to refer to the methods used to analyze DBA, large breaks, and small breaks. For j et-pump reactors, these are defined as follows:

a. DBA Methods. LAMB / S CAT / S AF E/ D BA-R EFLOOD / CHAS TE. Break size: DBA. l

b. Large Break Methods (LBM). LAMB / SCAT / SAFE /non-DBA h2 FLOOD / CHASTE.

Break sizes: 1.0 ft2 < A < DBA.

c. Small Break Methods (SBM). SAFE /non-DBA REFLOOD. Heat transfer coefficients: nucleate bei?!ng prior to core uncovery, 25 Btu /hr-ft *F after recovery, core spray when appropriate. Peak cladding temperature and peak local oxidation are calculated in non-DBA-REFLOOD. Break sizes: A < 1.0 ft .

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l NED0-24050-1 Table 2

SUMMARY

OF BREAK SPECTRUM RESULTS e Break Size Peak Cladding Core-Wide e Location Temperature Peak Local Metal-Water e Single Failure (PCT)-(*F) Oxidation (%) (' action (%)

e 1.4 ft (34% DBA) 2200 f1) 3.6 0.21 e Recirc Suction e LPCI Inj ection Valve e 4.0 ft (DBA) 2138( } Note 2 Note 3 s Recirc Suction e LPCI Injection Valve

1. PCT from CHASTE

2. Less than most limiting break (3.6%)

3. Less than most limiting break (0.21%) ,

Table 3 LOCA ANALYSIS FIGURE

SUMMARY

- NON-LEAD PLANT Large Break Methods Limiting Maximum Suction Break Suction Break (LPCI Injection (LPCI Injection l Valve Failure) Valve Failure)

(1.4 ft2)(34% DBA) (4.0 ft2)(DBA) l Water Level Inside Shroud and la lb Reactor Vessel Pressure Peak Cladding Temperature 2a 2b Heat Transfer Coefficient 3a Jb Core Average Inlet Flow 4a 4b Minimum Critical Power Ratio 5a Sb Variation with Break Area of Time for 6 Which Hot Node Remains Uncovered l

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NEDO-24050-1 Table 4A MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Plant: Monticello Fuel Type: 8DB219L Average Planar Expo sure MAPLHGR PCT Oxidation (mwd /st) (kW/ft) (*F) Fraction  ;

l 200 11.4 2094 0.024 I 1,000 11.5 2104 0.025 5,000 11.9 2163 0.029 10,000 12.0 2198 0.032 ,

15,000 11.9 2198 0.032 l 20,000 11.8 2194 0.032 25,000 11.3 2125 0.026 30,000 10.2 1982 0.016 35,000 9.7 1906 0.012 40,000 9.1 1829 0.009 Table 4B MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE i

Plant: Monticello Fuel Type: 8DB250 4 Average Planar Exposure MAPLHCR PCT 0xidation (mwd /st) ,kW/ft)

( (*F) Fraction 200 11.2 2101 0.024 1,000 11.3 2107 0.024 5,000 11.9 2174 0.030 10,000 12.1 2192 0.031 15,000 12.1 2198 0.032 20,000 11.9 2199 0.032 25,000 11.5 2157 0.028 30,000 10.6 2029 0.018 35,000 9.6 1897 0.011 40,000 9.0 1812 0.008 4-6

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NEDo-24050-1

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l Table 4C l

l MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE p I'

• Plant: Monticello Fuel Type: 8DB262 Average Planar i Exposure MAPLHGR PCT Oxidation l (mwd /st) (kW/ft) (*F) Fraction #

200 11.1 2100 0.024 l 1,000 11.3 2105 0.024 l 5,000 11.9 2172 0.029 l 10,000 12.1 2189 0.031 15,000 12.1 2198 0.031 20,000 12.0 2199 0.032 25,000 11.6 2161) 0.029 ,

1 30,000 10.5 2052 0.020 '

l 35,000 9.8 1926 0.013 40,000 8.9 1798 0.008 45,000 8.0 1676 0.004 ,

50,000 7.3 1566 0.003 4

Table 4D

! MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE  ;

Plant: Monticello. Fuel Type: 8DRB265L Average Planar l

Exposur,e MAPLHGR PCT Oxidation j (mwd /st) (kW/ft], (*F) Fraction l

200 11.5 2198 0.036 1,000 11.6 2198' O.036 l

j 5,000 11.7 2198 0.035 10,000 11.8 2197 0.034 15,000 11.7 2200 0.035 l

20,000 11.6 2196 0.035 25,000 11.3 2173 0.032 30,000- 10.7 .2093 0.025 i

35,000 10.2 2013 0.019-l 40,000 9.6 1928 'O.014 l

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NEDO-24050-1 Table 4E MAPLHGR VERSUS AVERAGF PLANAR EXPOSURE Plant: Monticello Fuel Type: 8DRB282 Average Planar Exposure MAPLHGR PCT Oxidation (mwd /st) (kW/ft) ('F) Fraction 200 11.2 2160 0.032 l 1,000 11.2 2155 0.032 5,000 11.6 2198 0.035 10,000 11.7 2198 0.035 15,000 11.7 2198 0.035 20,000 11.5 2195 0.035 25,000 11.3 2167 0.032 30,000 11.1 2142 0.00?

35,000 10.4 2055 0.022 40,000 9.8 1972 0.016 Table 4F MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Plant: Monticello Fuel Type: P8DRB265L Average Planar Exposure MAPLHGR PCT 0xidation (mwd /st) (kW/ft) (*F) Fraction _

200 11.6 2186 0.034 1,000 11.6 2193 0.035 5,000 11.8 2198 0.034 10,000 11.9 2196 0.034 15,000 11.9 2199 0.034 20,000 11.8 2194 0.034 25,000 11.3 2141 0.028 30,000 10.7 2057 0.021 35,000 10.2 1971 0.016 40,000 9.6 1871 0.011 4-8

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NEDO-24050-1 I

i Table 4G MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Plant: Monticello Fuel Type: P8DRB282

) Average Planar

! Exposure MAPLHGR PCT 0xidation i

___ _( mwd /st) (kW/ft) (*F) Fraction i 200 11.2 2146 0.031  !

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1,000 11.2 2146 0.030 5,000 11.8 2199 0.035 10,000 11.9 2198 0.034

15,000 11.8 2198 0.034

! 20,000 11.7 2194 0.034 25,000 11.3 2135 0.028

! 3C,000 11.1 2097 0.025 35,000 10.4 2006 0.036 4 40,000 9.8 1913 0.026 l

l Table 4H i MAPLHCR VERSUS AVERAGE PLANAR EXPOSURE Plant: Monticello Fuel Type: P8DRB284LB Average Planar Exposure MAPLHCR PCT 0xidation 1 (mwd /st) (kW/ft) (*F) Fraction 200 11.4 2172 0.033 1,000. 11.4 2168 0.032 5,000 11.8 2198 ,

0.035 i 10,000 11.9 2198 0.034 15,000 11.9 -2198 0.034 20,000 11.7 2193 0.034 25,000 11.4 2156 0.030 30,000 10.8 2057 0.022 35,000 10.2 ~1965 0.016 40,000 9.5 1867 0.011 4-9/4-10 l

NED0-24050-1 .

5. DESCRIPTION OF MODEL AND INPUT CHANGES The only change between this ECCS analysis and the previous Monticello analysis (NED0-24050) 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.

The addition of an alternate bypass flow path via holes drilled in the lower tieplate to the BWR/3s provides them with the same bypass flowpaths as the BWR/4s (i.e., same as BWR/4s with core plate holes plugged and holes drilled in the lower the plates). Since the BWR/3s do not currently include the LPCI modification, the ECC systems and the core configuration are the same as the BWR/4 non-mod plants. The primary difference between these two groups of plants is that BWR/3s have a lower power density than BWR/4s. Thus, in *.his analysis credit is taken for flow through holes in the fuel assembly lower tieplates as in the lead plant (Duane Arnold).

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NED0-24050-1

6. CONCLUSIONS

! The LOCA analysis results in accordance with the requirements of Reference 2 for non-lead plants with fuel bundles with drilled lower tieplates in a full or partial core loading are presented in Figures la through Sa for the limiting suction break (34% DBA) and Figures lb through 5b for the maximum suction break (DBA).

The characteristics that determine which is the most limiting break area at the DBA location are:

a

l. the calculated hot node reflooding time;

2. the calculated hot node uncovery time; and

3. 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 calcalated 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.

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! The period between hot node uncovery and reflooding is the period when the hot i

l 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 cal-culated PCT. If two breaks have sLnilar times during which the hot node remains 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 I

( severe LAMB / SCAT blowdown heat transfer analysis).

Figure 6 shows the variation with break size of the calculated time the not node remains uncovered for Monticello. The refloocing tir~ used for Figure 6 is conservative, since, in the analysis, intermediate short recoveries that f

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NEDO-24050-1 occurred in the range of 2 ft (50% of DBA) and 1.1 ft (27% of DBA) are conservatively ignored. This results in three peaks in the uncovered time versus break area plot. CHASTE PCT was calculated for these three peaks (48%, 42%, and 34% of DBA) in addition to the DBA and 1 f t . The 34% of DBA, 1.4 f t , with a LPCI-injection valve failure was determined to be the most limiting break.

Drilled lower tie plates in the fuel bundles provide an additional passage for core spray water accumulated in the bypass region, because of uper tie-plate CCFL, to refill the lower plenum and reflood the bundles earlier rela-tive to non-drilled bundles.

The benefit gained in early reflooding time due to a partial drill dimi,1 aes with decreasing break size since: (1) CCFL decreases with decreasing break area, and (2) the contribution of HPCI to water level accumulation increases for smaller breaks. Hence, the limiting break for Monticello shifts from 40%

of DBA when no drilling is _ assumed to 34% of DBA when partial drilling is assumed.

The conservative approach of using the 60% DBA LAMB / SCAT results with the 34%

DBA SAFE /REFLOOD results for calculations for the 34% DBA was used in all calculations for the analysis to determine the MAPLHCR's in Tables 4A through 4H.

l The DBA (the complete severence of the recirculation piping) results are shown on Figures Ib through 5b. The most significant change in these results from the previous analysis is that the reflooding time decreases from approximately l

260 seconds to approximately 194 seconds. This is due to the input and model changes described in Section 5. l

. The single-failure evaluation showing the remaining ECCS following an assumed l

failure and the effects of a single failure or operator error that causes any manually controlled, electrically operated valve in the ECCS to move to a position that could adversely affect the ECCS are presented in Reference 12.

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NEDO-24050-1 Low-flow multipliers were derived for application to this LOCA analysis, consistent with Reference 15. A 0.94 multiplier should be applied to all MAPLHGRs, for core flow less than 90%, rated flow, and a 0.91 multiplier should be applied for flow less than 70% rated, l

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HIGH POWER AX1AL PLANE REFLOODED HIGHEST POWER AXI AL PLANE "o "

-- LOWEST AX1AL PLANE TO 2.0 -

EXPERIENCE CPR = 1.0 5 PRIOR TO JET PUMP UNCOVER Y t /

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Figure 2a. Peak Cladding Temperature Following a 1.4 ft Recirculation Line Suction Break, LPCI Injection Valve Failure, Break Area = (34% DBA) (LBM)

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7.0 REFERENCES

1. Letter, Dennis L. Ziemann (NRC) to L. O. Mayer (NSP), "Re: Monticello Nuclear Generating Plant," dated March 11, 1977.

2. Letter, Darrell G. Eisenhut (NRC) to E. D. Fuller

" Documentation of the Reanalysis Results for the Loss-of-Coolant Accident OCA) of Lead and Non-Lead Plants." June 30, 1977.

3. General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10CFR50 Appendix K, NEDO-20566, dated January 1976.

4. " Safety Evaluation for General Electric ECCS Evaluation Model Modifica-tions," letter from K. R. Coller (NRC) to G. G. Sherwood (GE), dated April 12, 1977.

5. Letter, A. J. Levine (GE) to D. F. Ross (NRC) dated January 27, 1977,

" General Electric (GE) Loss of Coolar* accident (LOCA) Analysis Model Revisions - Core Heatup Code CHAS T'5.d

6. Letter, A. J. Levine (GE) to D. B. Vassallo (NRC), dated March 14, 1977,

" Request for Approval for Use of Loss of Coolant Accident (LOCA)

Evaluations Model Code REFLOOD05."

7. " Supplemental Information for Plant Modification to Eliminate Significant In-Core Vibrations," Supplement 1, NEDE-21156-1, September 1976.

8. " Supplemental Information for Plant Modification to Eliminate Significant In-Core Vibrations," Supplement 2, NEDE-21156-2, January 1977.

9. Letter, R. Engel (GE) to V. Stello (NRC), " Answers to NRC Questions on NEDE-21156-2," January 24, 1977.

10. Letter, G. L. Gyorey (GE) to V. Stello, Jr., dated May 12, 1975,

" Compliance with Acceptance Criteria for 10CFR50.46."

7-1

NEDO-24050-1

11. Letter, Lee Liu ('.EP&L) to Edson G. Case (NRC), Letter No. IE-77-1453, dated July 29, 1977.

i

12. Letter, L. O. Mayer (NSP) to D. L. Ziemann (NRC), "Monticello Nuclear Generating Plant, Docket No. 5-263, License No. DPR-22, Transmittal of ECCS Analysis," dated July 9, 1975.

13. Letter, R. E. Engel (GE) to D. G. Eisenhut (NRC), " Loss of Coolant Accident Analysis Methods for BWR 2/3 With Drilled Lower Tieplates (NEDE-24094)'," dated January 17, 1978.

14. Letter, D. G. Eisenhut (NRC) to R. L. Gridley (GE) dated June 7, 1978.

15. Letter, R. L. Gridley (GE) to D. G. Eisenhut (NRC), " Review of Low Core Flow Effects on LOCA Analysis for Operating BWR's", dated September 28, 1978. 1

)

7-2