Text

ECCS Large Break Spectrum Analysis Using ENC WREM-11A PWR Evaluation Model
	ML19276D709
Person / Time
Site:	Prairie Island
Issue date:	11/22/1978
From:	Krajicek J; SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:	;
Shared Package
ML19276D707	List: ML19276D706; ML19276D708; ML19276D709; ML19276D711;
References
	XN-NF-78-046, XN-NF-78-46, NUDOCS 7901090231
	Download: ML19276D709 (45)
	v • d • e

I XN NF 78 46 I

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I ECCS LARGE BREAK SPECTRUM ANALYSIS FOR I

PRAIRIE ISLAND UNIT 1 USING ENC WREM-IIA PWR EVALUATION MODEL I

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NOVEMBER 1978 g

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g E(ON NUCLEAR COMPANY,Inc.

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g em a o;

I ISSUE DATE: 11/22/78 XN-NF-78-46 ECCS LARGE BREAK SPECTRUM ANALYSIS I

FOR PRAIRIE ISLAND UNIT 1 USING ENC WREM-IIA PWR EVALUATION MODEL By J. E. KRAJICEK

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Appraved: ' J,l$[hYv/,,,

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Nuclear Safety Engineering Approved: [

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G.A.Sofer,Mfnager

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Nuclear Fuels Engineering

/G.J.Bu5selman,Itanig'er-

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f. S. Ke191odom, i

anager Licensing and C mpliance I

I NON-PROPRIETARY E(ON NUCLEAR COMPANY,Inc.

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NUCLEAR REGULATORY COMMISSION DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLE ASE READ CAREFULLY I

This technical report was derived through research and development programs sponsored by Exxon Nuclear Company, Inc.

It is being sub-mitted by Exxon Nuclear to the USNRC as part of a technical contri-bution to facilitate safety analyses by licensees of the USNRC which utilize Exxe Nuclear-fabricated reload fuel or other technical services provided b-Exxon Nuclear for liaht water power reactors and it is true and corres,t to the best of Exxon Nuclear's knowledge, in fo r ma tion, isrx1 belief. The information contained herein may be used by the USNRC m its review of this report, armi by licensees or applicants before the a USNRC which are customers of Exxon Nuclear in their demonstration of compliante with the USNRC's regulations.

E Without derogating from the foregoing, neither Exxon Nuclear nor any person actirig on its bet.alf:

A.

Makes any warranty, express or implied, with respect to the accuracy, completeness, or usefulness of the infor-mation contained in this document, or that the use of any information, apparatus, method, or process disclosed in this document will not infringe privately ownej rights; or 8.

Assumes any liabilities with respect to the use of, or for darrages resulting from the use of, any information, ap-paratus, method,

.>r process disclosed m this rfocument.

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XN-NF-F00, 766

E XN-NF-78-46 E

TABLE OF CONTENTS Section Page

1.0 INTRODUCTION

AND

SUMMARY

1 2.0 MAJOR REACTOR COOLANT SYSTEM PIPE RUPTURES (LOSS-0F-COOLANT ACCIDENT).......................

4 2.1 IDENTIFICATION OF CAUSES AND ACCIDENT DESCRIPTION....

4 2.2 THERMAL ANALYSIS.....................

5 2.2.1 Method of Analysis................

5 2.2.2 Large Break LOCA Analysis Modeling........

6 2.3 RESULTS.........................

7 2.4 CO NC LU S I ON S.......................

10 3.0 INTERIM UPPER PLENUM INJECTION "Gi;EL AND RESULTS....... 40 3.1 INTERIM UPI.MODEL CHANGES................

40 3.2 INTERIM UPI MODEL RESULTS................

41 4.0 MODEL HISTORY.........................

42 4.1 GENEALOGY OF MODELS................... 42 44

5.0 REFERENCES

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I XN-NF-78-46 I

ii LIST OF TABLES Page Table 1.1 Peak Cladding Temperature Results Prairie Island Unit 1 Reactor with ENC Fuel................

3 2.1 Prairie Island Unit 1 Large Break Events...

12 k.

2.2 Prairie Island Unit 1 Large Break Results..........

13 2.3 Prairie Island Unit 1 2-Loop PWR Data............

14 2.4 Prairie Island Unit 1 Dry Containment Data Containment Physical and Thermal Parameters.........

16 2.5 Heatup Calculated Results Summary Prairie Island 18 Unit 1 Reactor with ENC Fuel................

i LIST OF FIGURES Page Figure 2.1 RELAP4/EM Blowdown System Nodalization for Prairie Island Unit 1 2-Loop PWR..............

19 2.2 Axial Peaking Factor Versus Fuel Rod Length for Prairie Island Unit 1 ECCS Analysis............. 20 21 2.3 Blowdown System Pressure 0.4 DECLG Break..........

2.4 Blowdown Total Break Flow, 0.4 DECLG Break......... 22 2.5 Blowdown Average Core Inlet Flow, 0.4 DECLG Break...... 23 2.6 Blowdown Average Core Outlet Flow, 0.4 DECLG Break..... 24 2.7 Blowdown Hot Chanr.al Inlet Flow, 0.4 DECLG Break......

25 2.8 Blowdown Hot Channel Outlet Flow, 0.4 DECLG Break...... 26 2.9 Blowdown Pressurizer Surge Line Flow, 0.4 DECLG Break.... 27 2.10 Blowdown Intact Loop Accumulator Flow, 0.4 DECLG Break... 28 2.11 Blowdown Hot Rod Cladding Surface Temperature, 29 Node 20, 0.4 DECLG Break.........

2.12 Blowdown Hot Rod Volumetric Average Fuel 30 Temperature, Node 20, 0.4 DECLG Break............

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XN-NF-78-46 LIST OF FIGURES (continued)

Figure Page 2.13 Hot Rod Blowdown Heat Transfer Coefficient, Node 20, 0.4 DECLG Break..

31 2.14 Hot Rod Blowdown Depth of Zirconium - Water Reaction, Node 20, 0.4 DECLG Break..............

32 2.15 Containment Backpressure, 0.4 DECLG Break.......... 33 2.16 Normalized Power, 0.4 DECLG Break.............. 34 2.17 Reflood Core Flooding Rate 0.4 DECLG Break.........

35 2.18 Reflood System Pressure, 0.4 DECLG Break...........

36 2.19 Reflood Downcomer Mixture Level, 0.4 DECLG Break.......

37 2.20 Reflood Core Mixture Level, 0.4 DECLG Break......... 38 2.21 T00DEE2 Calculated Cladding Surface Temperature, 0.4 DECLG Break.......................

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XN-NF-78-46

1.0 INTRODUCTION

AND

SUMMARY

This document presents results of a break spectrum analysis using I

the ENC WREM-IIA PWR ECCS evaluation model(1,2,3) for the Prairie Island Unit 1* Nuclear Power Plant operating at 1650 MWt. The results show that the criteria specified by 10 CFR 50.46 are satisfied with an analysis performed in conformance to Appendix K of 10 CFR 50. A cal-culation to evaluate the impact of injection of ECCS fluid above the core (upper plenum injection, UPI) rather than in the downcomer as modeled using the ENC WREM-IIA ECCS models is also presented. This analysis was done using an interim model which was developed by the NRC(20) staff and modified by Westinghouse. (21)

Guillotine break configurations were calculated for double-ended cold-leg pipe breaks (DECLG) with discharge coefficients of 1.0, 0.6 and 0.4.

Split break configerations of the cold-leg pipe were cal-culated with a break area equal to twice the cross sectional pipe area (DECLS, 8.25 ft.2), then with break areas reduced to 0.6 and 0.4 times this value. The break spectrum analysis was performed for a core composed of Exxon Nuclear Company (ENC) fuel at Beginning-of-Life (B0L) conditions.

The limiting break was calculated to be the 0.4 DECLG break, which resulted in a calculated peak clad temperature (PCT) of 2197 F and a calculated maximum local Zr/H 0 reaction of less than 13 percent.

2 Addition of the interim UPI model calculated result +1F, to the ENC WREM-IIA model result yields in a final PCT value of 2198 F.

All of the calculations in the break spectrum were performed at a core power of 1683 MWt, which is 102 percent of rated power. The analysis

This ECCS analysis applies also to Prairie Island Unit 2 with ENC I

fuel of the same fuel design as Unit 1, Cycle 5.

XN-NF-78-46 supports operation of the Prairie Island Unit 1 plant with a total Linear Heat Generation Rate (LHGR) of 14.03 kw/ft, at a total peaking factor (F ) of 2.21 at rated power. The peak clad temperature versus break discharge coefficient and break size are presented in Table 1.1.

A detailed discussion of the break spectrum results is provided in Section 2.0.

The ENC WREM-IIA model was used for this analysis. The model includes the following computer codes: RELAP4-EM/ ENC 28C for blowdown and hot channel analyses; REFLEX for core reflood analysis; CONTEMPT LT/22, as modified in CSB 6-1(I4) for containment backpressure analysis; T00DEE2/APR78 fnr heatup analysis; and the ENC modified interim NRC UPI model which accounts for UPI-core interaction.

System models and nodalization used for the ENC WREM-IIA computer codes have been pre-sented in the Two-Loop PWR Example Problem Document XN-NF-77-25(A)(16) and the generic PWR ECCS Evaluation Model Update ENC WREM-IIA Document XN-NF-78-30.(3,23)

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I XN-NF-78-46 TABLE 1.1 Peak Cladding Temperature Results Prairie Island Unit 1 Reactor with ENC Fuel F = 2.21 Maximum Power Node Clad Surface Temp.

0 E0BY PCT Results

( F)

PCT TIME PCT Guillotine Breaks (sec)

( F)

DECLG C = 1. 0

1150, 240.

2m.

D I

DECLG C = 0.6 1176.

230.

2117.

D DECLG*

I C = 0.4 1400.

206.

2197.

D Split Breaks 1.0 DECLS 8.25 ft2 1139.

234.

2142.

0.6 DECLS 4.95 ft2 1118.

232.

2114.

0.4 DEC S 3.30 ft 1065.

242.

2142.

A value of +1F must be added to the 0.4 DECLG PCT value to account for UPI effects.

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I XN-NF-78-46 2.0 MAJOR REACTOR COOLANT SYSTEM PIPE RUPTURES (LOSS-OF-COOLANT ACCIDErlT) 2.1 IDENTIFICATION OF CAUSES AND ACCIDENT DESCRIPTION The analysis for large breaks specified by 10 CFR 50.46(15),

" Acceptance Criteria for Emergency Core Cooling Systems for Light Water Power Reactors", is presented in this section.

The results of the loss of coolant accident analysis are shown in Tables 2.1 and 2.2, which indicate

~

compliance with the Acceptance Criteria. The analytical techniques used are in compliance with Appendix K of 10 CFR 50, and are as described in XN-75-41, Volumes I and II, and supplements (I); ENC-WREM-IIA model is described in XN-76-44(I7),XN-76-36(18), XN-NF-78-30( ), XN-NF-78-25(23), and XN-76-27 plus supplements (2)

Except as noted below, the detailed system models are as given in the two-loop PWR example problem report: XN-NF-77-25(A)(16),

For the purpose of loss-of-coolant accident (LOCA) analyses, a LOCA is defined as a hypothetical rupture of the Reactor Primary Coolant System piping, up to and including the double-ended rupture of the largest pipe in the Reactor Coolant System or of any line connected to that system up to the first closed valve.

Should a major break occur, depressurization of the Reactor Coolant System results in a pressure decrease in the pressurizer. A reactor trip signal occurs when the pressurizer low pressure trip setpoint is reached.

Reactor trip and scram were conservatively neglected for the large break analyses.

A Safety Injection System signal is actuated when the appropriate setpoint (high containment pressure) is reached.

These countermeasures will limit the consequences of the accident in two ways:

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I XN-NF-78-46 1.

Reactor trip anc' borated water injection complements void formation in causing rapid reduction of power to a residual level corresponding to fission product decay heat.

2.

Injection of borated water provides heat transfer from the reactor core and prevents excessive clad temperatures.

At the beginning of the blowdown phase, the entire Reactor Coolant System contains subcooled liquid which transfers heat from the core by forced convection cooling. After the break develops the time to departure from nucleate boiling (DNB) is calculated consistent with Appendix K of 10 CFR 50fl4)

Post DNB core heat transfer (both transition and film boiling occurring) is also calculated in accordance with Appendix K of 10 CFR 50. As the core becomes uncovered, both turbulent and laminar forced convection to steam are con-sidered as core heat transfer mechanisms.

When the Reactor Coolant System pressure fails below 715 psia, the accumulators begin to inject borated water.

The conservative assumption is made that accumulator ECC water bypasses the core and goes out through the break until the termination of bypass.

This conservatism is consistent with Appendix K of 10 CFR 50.

2.2 THERMAL ANALYSIS 2.2.1 Method of Analysis For breaks greater than 1.0 ft, the RELAP4-EM code (1,2,17) 2 is used to calculate the transient depressurization of the Reactor Coolant System as well as to describe the mass and enthalpy of flow out of the break.

A specialized calculation (RELAP4-EM/ HOT CHANNEL) is used to calculate cladding temperatures using time dependent boundary conditions in the upper and lower plenum I

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I XN-NF-78-46 volumes from the basic blowdown analysis.

Beyond the point of refill to the bottom of the core, a specialized calculation (REFLEX) is applied to determine the reflooding rate and system conditions. After end-of-bypass (E0BY), the program T00DEE2 is used to calculate peak clad temperatures.

2.2.2 Large Break LOCA Analysis Modeling The Prairie Island Unit 1 nuclear power plant is a 2-loop Westinghouse pressurized water reactor with a dry containment.

The reactor coolant system is nodalized into control volumes representing reasonably homogeneous regions, interconnected by flow-paths or " junctions" I

as described in XN-NF-77-25(A)(16) The nodalized system blowdown model schematic is given in Figure 2.1.

For conservatism, the upper head temperature is taken to be that of the core outlet temperature. One percent of the steam generator tubes were assumed to be uniformly plugged. The unbroken loop was assumed symmetrical and modeled the same as the broken loop except for the break nodalization and the pressurizer.

Pump performance curves characteristic of the Prairie Island pumps as supplied by the NSSS vendor were used in the analysis.

System input parameters are given in Table 2.3.

The evaluation of the E0BY has been updated to reflect available data from cold-leg steam-water mixing studies (5,6,7,8,9,&l0) as shown in Reference 11.

The reactor core is modeled with heat generation rates determined from reactor kinetics equations with reactivity feedback and with decay heating as required by Appendix K of 10 CFR 50. The axial power profile used for the analysis is a chopped cosine curve with an axial peaking factor of 1.384 as given in Figure 2.2.

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_7, XN-NF-78-46 I

The values for the primary coolant system core inlet temperatures and the steam generator secondary side pressure were set based on Prairie Island Unit 1 plant operatine data c'>tained from the utility. The values of the core inlet temperature and the steam gener-ator secondary side pressure are 530 F and 710 psig, respectively.

The containment backpressure for the analysis of the postu-lated LOCA was evaluated in accordance with the discussion presented in XN-75-41, Supplement 5, Section 4.6.

A containment analysis was performed using the computer code CONTEMPT-LT, Version 22, modified as described in Supplement 5, Revision 1, of XN-75-41(I)

The condensing heat transfer coefficient is modeled in accordance with Branch Technical Position CSB 6-1,

" Minimum Containment Pressure Model for FWR ECCS Performance Ev The containment parameters used in the containment analysis to determine the ECCS backpressure are presented in Table 2.4.

2.3 RESULTS Using the ENC WREM-IIA codes, transient system behavior is determined by solving the governing conservation equations for mass, energy, and momentum.

Energy transport, flow rates, and heat transfer are determined from appropriate correlations.

Table 2.1 presents the timing and sequence of events as determined for the large break guillotine configuration with discharge coefficients of 1.0, 0.6 and 0.4 and the split break configuration with break areas of 8.25, 4.95, and 3.30 square feet.

In general, the transient events occur slower for smaller dis-charge coefficients or break sizes.

Table 2.2 presents the peak clad temperatures and maximum metal-water reaction results for the above spectrum of lo cak i a.cs lhts range of break sizes was determined to include the limiting case for peak clad temperature.

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I XN-NF-78-46 The analysis of the loss-of-coolant accident is performed at 102%

of 1650 MWt (1683 MWt).

The core power and other parameters used in the analyses are given in Table 2.3.

Since there is usually margin between the value of the peak linear power density used in this analysis and the value expected in operation, a lower peak clad temperature would be obtained by using the peak linear power density expected during operation.

For the results discussed below, the hot spot is defined to be the location of maximum peak clad temperature. This location is given in Tat,le 2.2 for each break size analyzed.

Figures 2.3 through 2.21 present the results of the analysis for the 1imiting break (0.4 DECLG). Unless otherwise noted, zero time corresponds to the time of break initiation.

The maximum peak cladding temperature of 2197 F (excluding the UPI model correction) was calculated for the double-ended cold-leg guillotine break configuration (C = 0.4) and a total linear heat genera-D tion rate of 14.31 kw/ft (F = 2.21) for ENC fuel (102% of 14.03 kw/ft).

The maximum PCT of 2198 F includes +1F calculated with the UPI model. The maximum local metal-water reaction is less than 13% and the total core metal-water reaction reached will be much less than 1%, all well below the limits set by the criteria of 10 CFR 50.46.

Additional results from the break spectrum analysis are given in Table 2.5.

Table 2.5 shows the maximum power node surface temperature for the limiting break (0.4 DECLG) to be more than 220F higher than the other cases at end-of-bypass (E0BY).

Based on these maximum power node temperatures at E0BY, the maximum temperature is clearly defined, and there is a well defined decreasing temperature trend with increasir.g C D for the guillotine breaks and an increasing temperature trend with a

XN-NF-78-46 increasing break size for the split breaks. Temperature differences between comparable fuel rod nodes are consistent between the limiting break case and the other break cases.

Cladding temperature behavior remains consistent with E0BY temperature results until the reflood rate is calculated to be I

less than one inch per second.

A reflood rate of less than one inch per second is predicted to occur 115 to 120 seconds from the initiation of the postulated LOCA.

The high cladding temperatures asrociated with the limiting break (0.4 DECLG) result in clad rupture (during abiabatic heatup) for the high power node prior to the initiation of reflood.

For the other breaks which exhibit fuel cladding temperatures which are less than those of the 0.4 DECLG case, cladding rupture occurs 30 to 40 seconds after initiation of reflood and at higher axial elevations along the rod.

The rupture node axial location information is also presented in Table 2.5.

It is the lower initial cladding temperatures which delays cladding rupture and causes rupture to occur at higher elevations where the reflood heat transfer coefficients are lower.

This accounts for the higher (but not limiting) PCT for the 1.0 DECLG case and the split breaks shown in Table 2.5.

That is, since reflood rates are essentially the same for all cases, temperature behavior during abiabatic refill and during reflood would be expected to be consistent with the maximum power node temperatures as shown in Table 2.5.

The limiting break is consistent with this expected behavior, but other cases deviate from the expected trend due to migration of the rupture node and associated steam cooling effects around the rupture node. When these phenomena are cen-sidered with the Table 2.2 events, the limiting break for Prairie Island Unit 1 is clearly defined as the 0.4 DECLG case, and the break spectrura shape can be I

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I XN-NF-78-46 explained in terms of the model.

ENC has performed numerous analyses and sensitivity studies on PWR systems using the ENC ECCS evaluation model.

These studies have demonstrated the adequacy of the system nodalization used.

In addition, these studies have shown that for transient conditions similar to those calculated for the Prairie Island reactor during the LOCA, the reactor coolant inlet pipe or cold leg is the worst break location and that BOL fuel condition result in the maximum PCT.

2 Small break (1.0 ft area or less) analyses have not been performed as a part of the Prairie Island ECCS analysis because the ECCS systems provide adequate protection and are capable of limiting the I

cladding thermal transient for small pipe breaks to temperatures well below those of the large breaks. This exclusion of small breaks is based on ENC analysis results from similar Westinghouse reactor designs (24),

NSSS vendor results have also shown that small breaks are not limiting in all these cases including Prairie Island.

2.4 CONCLUSION

S For breaks up to and including the double-ended severance of a reactor coolant pipe, the Prairie Island Emerger.cy Core Cooling System will meet the Acceptance Criteria as presented in 10 CFR 50.46 for the ENC reload fuel of similar design operating in accordance with the LHGR limits noted in Section 1.0.

That is:

1.

The calculated peak fuel element clad temperature does not exceed the 2200 F limit.

2.

The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1% of the total zircaloy associated with the active fuel rod length in the reactor.

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-II-XN-NF-78-46 3.

The cladding temperature transient is tenninated at a time when the core geometry is still amenable to cooling. The hot fuel rod cladding oxidation limits of 17% are not exceeded during or after quenching.

4.

The system long-tenn cooling capabilities provided for previous cores remain applicable for ENC fuel.

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TABLE 2.1 PPAIRIE ISLAND UNIT 1 LARGE BREAK EVENTS Event Time (Seconds)

DECLG DECLG DECLG 1.0 DECLS 0.4DECLg) 0.6DECLg) 2 (3.30 ft (C =1.0)

(C =0.6)

(C =0.4)

(8.25 ft )

(4.95 ft D

D D

Start 0.00 0.00 0.0 0.000 0.00 0.00 Initiate Break 0.05 0.05 0.05 0.05 0.05 0.05 h

Safety Injection Signal 0.50 0.55 0.65 0.50 0.55 0.60 Accumulator Injection, Intact Loop 6.80 7.40 8.80 6.85 7.00 8.55 Accumulator Injection, Broken Loop 0.15 2.50 4.80 0.85 3.10 6.95 Pressurizer Empties 8.80 8.80 8.80 8.80 8.80 8.80 End-of-Bypass 19.34 19.79 21.50 18.70 18.79 19.56 f

Safety Injection Flow, SIS 25.50 25.55 25.65 25.50 25.55 25.60 Start of Reflood 35.36 35.73 37.24 34.69 34.75 35.32 A

P Accumulator Empty, Intact Loop 45.56 47.71 48.13 45.63 45.85 47.52 Peak Clad Temperature Reached 239.50 229.60 205.50 i 233.60 231.60 242.60

TABLE 2.2 PRAIRIE ISLAND UNIT 1 LARGE BREAK RESULTS Event DECLG DECLG DECLG 1.0 DECLS 0.6 DECLS 0.4 DECLS 2

2 2

(C =1.0)

(C =0.6)

(C 0.4)

(8.25 ft )

(4.95 ft )

(3.30 ft )

D D

D Peak Cladding Temperature

OF 2174 2117 2197 2142 2114 2142 Peak Temperature Location, ft 7.00 6.75 7.81 7.00 7.00 7.25 Local Zr/H O Reaction (Max.), %

9.88 8.74 12.34 9.12 8.34 8.50 2

Local Zr/H 0 Location, ft 7.00 6.75 7.50 7.00 7.00 7.25 2

Total H Generation, % of 2

total Zr reacted

<1%

Hot Rod Burst Time, sec 67.24 63.69 36.80 67.60 72.09 87.06 Hot Rod Burst Location, ft 6.75 6.50 6.00 6.75 6.75 7,00 Linear Heat Generation Rate, kw/ft at BOCREC 0.7330 0.7311 0.7227 0.7363 0.7358 0.7329 2

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A value of +1 F must be added to the limiting PCT value te account for upper plenum injection-core interaction effects.

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I XN-NF-78-46 I I

I TABLE 2.3 PRAIRIE ISLAND UNIT 12-LOOP PWR DATA 1650*

Primary Heat Output, MWt 7

6.82 x 10 Primary Coolant Flow, lbm/hr 3

10,309.**

Primary Coolant Volume, ft 2,250.

Operating Pressure, psia 530.

Inlet Coolant Temperature, UF 2364.

Reactor Vessel Volume, ft 3

1000.

Pressurizer Volume, Total, ft 600.

Pressurizer Volume, Liquid, ft I

3 2000.

Accumulator Volume, Total, ft (each of two) 3 1250.

Accumulator Volume, Liquid, ft 714.7 Accumulator Trip Point Pressure, psia I

2 50,985.

Steam Generator Heat Transfer Area, ft 6

3.54 x 10 Steam Generator Sec1ndary Flow, lbm/hr 724.7 Steam Generator Secondary Pressure, psia 277.

Reactor Coolant Pump Head, ft 1190.

Reactor Coolant Pump Speed, rpm 2

78,000.

Moment of Inertia, lbm-ft / rad 27.5 Cold Leg Pipe, I.D., in 29.0 Hot Leg Pipe, I. D., in 31.0 Pump Suction Pipe, I.

D., in

Primary Heat Output used in RELAP4-EM Model = 1.02 x 1650 = 1683 MWt.

- Includes total accumulator and pressurizer volumes.

XN-NF-78-46 I

I TABLE 2.3 (Continued)

I Fuel Assembly Rod Diameter, in*

0.424 Fuel Assembly Rod Pitch, in*

0.556 Fuel Assembly Pitch, in*

7.803 Fueled (Core) Height, in*

144.

2 Fuel Heat Transfer Area, ft 28,851.

2 Fuel Total Flow Area, ft 26.71 Steam Generator Tube Plugging (Assumed uniform) 1%

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ENC fuel parameters I

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I XN-NF-78-46 I

I TABLE 2.4 l

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PRAIRIE ISLAND UNIT 1 DRY CONTAINMENT DATA Containment Physical and Thermal Parameters 6 3 Net Free Volume 1.36 x 10 f t I

Outside Air Temperature

-20 0F Spray Flow 23 g

Trip Setting I

Time Delay Fan Coolers 4

9 Trip Setting 35 Time Delay I

Containment Initial Conditions:

Temperature 90 F Pressure 14.7 psia Relative Humidity 100%

Containment Spray Water:

g Temperature 70 F Flow Rate (Total, 2 pumps) 3600 gpm Fan Air Cooler Capacity (total 4 coolers)

Vapor Temperature (UF)

Capacity (Btu /hr) 7 150 3.10 x 10 170 3.65 x 167 190 4.60 x 107 210 5.55 x 107 230 6.75 x 107 250 8.20 x 107 270 9.60 x 107 8

290 1.11 x 10 8

300 1.18 x 10 I

XN-NF-'78-45 TABLE 2,4 (Continued)

PRAIRIE ISLAND UNIT 1 CONTAINMENT DATA Passive Heat Sink Thermal Conductivity and Volumetric Heat Capacity Data Thermal Volumetric Conductivity Heat Capacity Materials (Btu /hr-ft OF)

Btu /ftJ OF)

Steel 28.0 56.2 Structural Concrete 0.8 32.0 l

Metal Overcoat Paint 0.29 28.0 l

l Metal Primer 1.50 28.0 Concrete Overcoat Paint 0.29 28.0 Concrete Primer 0.29 32.0 f

Containment Passive Heat Sinks THICKNESS SURFgCE j

DESCRIPTION MATERIAL (s)

IN2 FT 1.

Contain cylinder steel 1.5 41,300.

l 2.

Containment dome steel 0.75 32,000.

I i

3.

Reactor vessel and steel 0.25 7,860.

3 refueling canal concrete 12.00 7,860.

E 4.

HVAC ducting steel 0.25 32,000.

I 5.

NSSS supports steel 0.5 44,000.

t 6.

Exposed pipe steel 0.375 C6,800.

7.

Hand rails steel 0.145 1,695.

8.

Grating steel 0.09 12,400.

9.

Conduit and cable trays steel 0.1 6,000.

10.

Accumulators steel 1.44 2,200.

11.

Ductwork steel 0.1875 35,125.

12.

Thick concrete walls concrete 12.0 40,800.

13. Thick concrete floors concrete 6.0 25,070.

14.

Thin concrete walls concrete 3.0 7,570.

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I XN-NF-78-46 TABLE 2 1 Heatup Calculated Results Summary Prairie Island Unit 1 Reactor with ENC Fuel I

Max. Power Node Clad Surf. Temp.

PCT Results Heatup Rupture I

0 E0BY PCT Time PCT Node No.

Guillotine Breaks

( F)

(sec)

( F)

DECLG CD = 1.0 1150.

240.

2174.

11 DECLG CD = 0.6 1176.

230.

2117.

10 DECLG*

CD = 0.4 1400.

206.

2197.

8 Split Breaks 1.0DECgS 1139.

234.

2142.

11 8.25 ft 0.6DECgS 4.95 ft 1118.

232.

2114.

11 0.4 DECLS 3.30 ft2 1065.

242.

2142.

12 I

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A value of +1F must be added to the 0.4 DECLG PCT to account for UPI effects.

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XN-NF-78-46 thl (kl I

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l XN-NF-78-46 I

3.0 INTERIM UPPER PLENUM INJECTION MODEL AND RESULTS 3.1 INTERIM UPI MODEL CHANGES The interim UPI analysis applied to the Prairie Island Unit I reactor follows the approach presented in the DSS SER " Safety Evaluation Report on ECCS Evaluation Model for Westinghouse Two-Loop Plants"( 0)

The computer program representing the NRC Staff model was obtained and veri-fied against the NRC Staff sample results.

The computer program was then modified as requimd to represent the Prairie Island reactor and incorporate I

the base ENC ECCS analysis results. Modifications to the model include the following:

(1)

Capability to use time-varying data from the base ENC WREM-IIA ECCS analysis was added for the following parameters: Containment pressure, satura tion temperature, decay power, reflood rate, and low pressure safety injection (LPSI) flow and subcooling.

(2) The carry over rate fraction value used was 0.7 ( 1, 22)

(3) An upper plenum structures conduction heat transfer conduction model was incorporated in the program.

(4) The clad temperature rise versus flooding rate curve, figure 24 in Reference 20, was replaced with the updated data provided in Figure 2 from Reference 21. The updated curve is more representative of the peak rod power, initial rod temperature, reflood subcooling, and system pressure for a two-loop PWR.

(5) The horizontal entrainment was taken as a constant fraction of the LPSI injection flow (1.6%) (20)

(6) The core flow area was changed as appropriate to reflect all ENC fuel in the Prairie Island Unit I reactor.

I I

XN-NF-78-46 (7) The core heat capacity was revised to be consistent with the fraction of the core being quenched by UPI downflow.

3.2 INTERIM UPI MODEL RESULTS Applying the above model to the Prairie Island reactor with Exxon Nuclear fuel gives the following results:

y ENC WREM-IIA Analysis for Prairie Island 0.4 DECLG Break Revised Analysis with UPI Peak Cladding Temperature Peak Cladding Temperature 0

2197 F 2.21 2198 F These results show continued compliance with 10CFR 50.46 and Appendix K to 10CFR Part 50, and are considered conservative since the adverse effects of UPI steam generation have been considered in reducing reflood rates, while the benefits of UPI steam generation in reducing core temperatures have been neglected.

I I

I XN-NF-78-46 I

4.0 MODEL HISTORY The following section presents the genealogy of the models used in the subject analysis.

4.1 GENEALOGY OF MODELS The Prairie Island Unit 1 ECCS analysis was performed with the following code versions.

Blowdown s RELAP4-EM/ ENC 28C Hot Channel s RELAP4-EM/ ENC 28C Reflood s REFLEX Heatup s T00DEE2/APR78 Containment s CONTEMPT-LT/ VERSION 22 RELAP4-EM/ ENC 28B - The code changes to RELAP4-EM/ ENC 26A to produce RELAP4-EM/ ENC 28B are described in the attachment to a letter to D. F. Ross from G. F. Owsley dated October 1978.

RELAP4-EM/ ENC 28C - Three plot variables were added to RELAP4-EM/ ENC 28B to pennit plotting of fuel related heat slab internal temperatures, i.e.,

pellet surface temperature, clad inside surface temperature, etc.

A causal heat slab variable (time step control) was initialized to permit execution of a RELAP4 case without a core and with zero heat slabs.

These changes do not affect calculated results.

REFLEX - The REFLEX code is described in XN-NF-78-30 (3) which is the ENC WREM-IIA document.

T00DEE2/APR78 - This version incorporated the ENC WREM-II models into ENC master code version, T00DEE2/JAN77.

These code modifications are described in XN-NF-77-27 and in XN-NF-77-58.

These code modifications included logic to facilitate the calculations required in conjunction with the sensitivity XN-NF-78-46 studies required by the NRC for the rupture pressure uncertainty analysis.

The following AVAIL groups were modified or added to T00DEE2:

AVAIL (78)

FLECHT multipliers option 0.0 use FLECHT correlation multipliers 1.0 use FLECHT/ ENC 2-WREM I correlation multipliers 2.0 use FLECHT/ ENC 3-WREM-II correlation multipliers AVAIL (82)

Gas volume in fuel (dishes and cracks)

AVAIL (83)

Heat transfer coefficient multiplier (default 1.0)

AVAIL (84)

Prerupture strain constant (default 0.20)

AVAIL (85)

Blockage model switch 1.0 ENC WREM-I model (must also set AVAIL (78) = 1.0) 2.0 ENC WREM-II model (default valve)

In addition, the code I/O was modified such that additional data at the time of fuel rupture is printed (fraction of flow area blocked, rupture pressure, rupture temperature, temperature and volume of the fuel plenum, and average temperature and volume for the gas in the cracks, dishes, and gap.)

These changes do not affect calculated results.

CONTEMPT-LT/ VERSION 22 - The RELAP4-EM environmental package was added to allow input data to be submitted as free fonnat as a convenience to the user.

None of the code models were changed, and no changes in the calculated results occurred.

I I

XN-NF-78-46 I

5.0 REFERENCES

1.

Exxon Nuclear Company, Exxon Nuclear Company WREM-Sased Generic PWR ECCS Evaluation Model, XN-75-41:

a.

Volume I, July 1975 b.

Volume II, August 1978 c.

Volume III, Revision 2. August 1975 I

d.

Supplement 1, August 1975 e.

Supplement 2, August 1975 f.

Supplement 3, August 1975 g.

Supplement 4, August 1975 I

h.

Supplement 5, Revision 5, Octover 1975 1.

Supplement 6, October 1975 j.

Supplement 7, November 1975 2.

Exxon Nuclear Company, Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model Update ENC WREM-II, XN-76-27:

a.

July 1976 b.

Supplement 1, September 1976 c.

Supplement 2, November 1976 3.

Exxon Nuclear Company, Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model Update ENC WREM-IIA, XN-NF-78-30 August 1978.

I 4.

U.S. Nuclear Regulatory Commission, WREM, Water Reactor Evaluation Model, Revision 1, May 1975 5.

Lilly, G.

P., Mixing of Emergency Core Cooling Water with Steam:

1/3 - Scale Test and Summary, EPRI 294-2 Final Report, June 1975.

6.

Broderick, J. R., Loiselle, V., CoJd Leo Condensation Tests,

o Task C. Steam Water Interaction Tests, CENPD-129, March 1974.

I 7.

Broderick, J. R., Burchill, W. C., Lowe, P. A.,1/5 Scale Intact Loop Post-LOCA Steam Relief Tests, CENPD-63, March 1973 Revision.

8.

Flanigan, L. J., Cudnik, R. A., Denning, R. S., Topical Report I

on Experimental Stuoles of ECC Delivery in a 1/15 Scale Transparent Vessel Model, BMI-1941, November 1975.

9.

Rothe, P. H., Wallis, G. B., Thrall, D. E., Cold Leg ECC Flow Oscillations, EPRI NP-282, November 1976 I

10.

Zender, S.

N., Jensen, M.

F., Sackett, K.

E., Experiment Data Report for Semiscale M00-1 Test S-01-4 and 5-01-4A, ANCE-1196, March 1975.

11. Antonopoulos P.

T., and Husain, A., Method for Calculating End of I

Bypass Time for Yankee Rowe Loss-of-Coolant Accident Analysis, YAEC-ll25, March 19/7.

I I

XN-NF-78-46 12.

Block, J. A., and Wallis, G. B., Effect of Hot Walls on Flow in a Simulated PWR Downcomer During a LOCA, CREARE-TN-188, May 1974.

13.

Block, J. A., and Crowley, C. J., Hot Wall Experiments in Simulated Multiloop PWR Geometry, CREARE-TN-202, February 1975.

14.

U.S. Nuclear Regulatory Commission, Minimum Containment Pressure flodel for PWR ECCS Performance Evaluation, Branch Technical Position CSB 6-1.

15.

10 CFR 50.46 and Appendix K of 10 CRF 50, Acceptance Criteria for g

Emergency Core Cooling Systems for Light Water Cooled Nuclear Power E

Reactors, Federal Register, Volume 39, Number 3, January 4,1974~.

16. Exxon Nuclear Company, Exxon Nuclear Company EC_CS Eva_luation of a 2-Loop Westinghouse PWR With Dry Containment _ _Using the ENC WREM-II ECCS Model - Large Break Example Problem, XN-NF-77-25( A). September 1978.

17.

Exxon Nuclear Company, Revised Nucleate Boili_ng Lockout for ENC WREM-Based ECCS Evaluation Models, XN-76-44, September 1976.

18.

Exxon Nuclear Company, Exxo_n Nuclear _Com3any WREM-Bas _ed Generic PWR ECCS Evaluation Model (ENC-WREM-II

- 4 _oop PWR With ice Condenser, Large Break Example Problem, XN-76-36, August 1976.

19.

Exxon Nuclear Company, ECCS Analysis for the R. E. Ginna_ Reactor With ENC WREM-II PWR Evaluation Model, XN-NF-77-58, December 1977.

20.

U.S. Nuclear Regulatory Commission, Safety Evaluation Report on ECCS Evaluation Model for Westinghouse Two-Loop Plants, Analysis Branch, IHvision of Systems Safety, Office of Nuclear Reactor Regulations, November 1977.

21. Letter from L. O. Mayer to Director of Nuclear Reactor Regulation, February 24, 1978 (Docket No. 50-282 and 50-306).

22.

U.S. Nuclear Regulatory Commi sion, Safety Evaluation Report on Interim ECCS Evaluation Model for Westinghouse Two-Loop Plants, March 1978.

23.

Exxon Nuclear Company, Big Rock Point Example LOCA Analysis Using The Exxon Nuclear Company Non-Jet Pump BWR Evaluation Model - Large Break Example Problem, XN-NF-78-25, Revision 1, September 1978.

24. Rochester Gas and Electric Corporation, letter, L. D. White to D. L. Ziemann (NRC), August 7,1978, Question 5.1.

I

ML19276D709

Contents

Text

1.0 INTRODUCTION

SUMMARY

5.0 REFERENCES

1.0 INTRODUCTION

SUMMARY

2.4 CONCLUSION

5.0 REFERENCES

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ML19276D709

Text

1.0 INTRODUCTION

SUMMARY

5.0 REFERENCES

1.0 INTRODUCTION

SUMMARY

2.4 CONCLUSION

5.0 REFERENCES

Navigation menu

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