ML20197D062

Forwards Tl Wilson, Large Break LOCA in Midland Plant, Per Rj Mattson Request for Trac Calculational Assistance.Results of Large & Small Break LOCA Analyses Presented to NRR & RES on 820412.Rept on Small Breaks Will Be Forwarded

ML20197D062
Person / Time
Site:	05000000, Midland
Issue date:	05/10/1982
From:	Demuth N LOS ALAMOS NATIONAL LABORATORY
To:	Shotkin L NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
Shared Package
ML19301C663	List: ML20197D062 ML20197D803 ML20210Q716
References
FOIA-86-110 Q-7-82-204, NUDOCS 8605140116
Download: ML20197D062 (17)

Text

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IN AEPLY pf7Em 70 K556 Los Alamos NationalLaboratory u m svo, Los Alamos.New Mexico 87545 (505) 667-2021 or retc,,,,,,

t'TS 843-2021 Energy Division Dr. Louis M. Shotkin Analytical Models Branch Division of Accident Evaluation Mail Stop 11305S US Nuclear Regulatory Commission Washington, DC 20555

REFERENCE:

MEMORANDUM FROM R. J. PATTSON, NPR/DSI TO 0. F. BASSFTT, RES/DAE, "RFOUEST FOR TRAC CALCULATIONAL ASSISTANCE", DATED CCTOBFR 23, 1982.

Uc have completed the Midland Best-Estimate LOCA Analyses that are outlined in the reference. The enclosed informal report by Tim Uilson describes the results for the_large-break portion of this task.

Results f rom both the large-and small-break LOCA analyses were presented to representatives from NRR and RES on April 12, 1982. We are planning to issue an informal report (technical note) on the small-break LOCA analyses later this month and to amplify further the description of the large-break LOCA analyses before issuing a technical note on this subject, probably in July.

The large-break LOCA analyses described in the enclosed document were performed with TRAC-PD2 and a generic model of P&W lowered-loop plants.

This model was modified to reflect plant-specific initial and boundary conditions supplied by Mr. Walt Jensen (NRR/DSI). Although we modified the axial zoning of the vessel in the TRAC,model to better represent the large-break behavior, we did not alter the coarse radial and azimuthal zoning because of resource and schedular constrafnts. Thus, while the results are believed to adequately describe the main characteristics of large-break behavior, some of the detailed thermal-bydraulic phenomena may be obscured with the coarse noding.

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8605140116 860421 PDR FOIA PEDROS 6-110 PDR

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Dr. Louis M. Shotkin May 10, 1982 Q-7-82-204 I hope this information provides a satisfactory basis for responding to the request from NRR.

Sincerely, Ne ' son S. DeMuth LWR Safety Analysis NSD:dco Enc: as cited xc w/ enc:

T. Lee, NRC/DAE F. Odar, NRC/DAE R. J. Mattson, NRC/DSI T. P. Speis, NRC/DSI B. Sheron, NRC/DSI J. Guttman, NRC/DSI W. Jensen, NRC/DSI N. Lauben, NRC/DSI W. Minners, NRC/DSI J. H. Scott, EP/NP, MS-F671 J. F. Jackson, Q-DO, MS-J561 R. A. Haarman, Q-6, MS-G777 T. L. Wilson, Q-6, MS-G777 L. L. Smith, Q-7, MS-K556 N. S. DeMuth, 0-7, MS-K556 C. J. E. Willeutt, Q-7, MS-K556 R. Henninger, Q-7, MS-K556 J. Elliott, Q-7, MS-K556 C. E. Watson, Q-7, MS-K556 cRMO (2), MS-A150 File (NSD) i e

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LARGE BREAK LOCA IN THE MIDlhND PLANT

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by

' T. L. Wilson Los Alamos National Laboratory A postulated 2004 double-ended guillotine (DEG) break in the cold leg of I

10o0 A in the Midland B&W reactor was simulated with the TRAC-PD2 computer code.

Best-estimate boundary and initial conditions were used. The ourpose of this analysis was to quantify the conservatism in the applicant's licensing analyses.2 In this analysis, TRAC calculated a ceak cladding temperature of I2 7/ *"

962 K in the hottest rod, and all fuel rods were quenched by 123 s.

The two-loop system (Fig.1) is modeled with 80 three-dimensional finite-difference cells in the reactor vessel and 162 one-dimensional cells in the rest of the system for a total of 234 cells. The Midland reactor is a 2 by 4 (2 hot legs, 4 cold legs) lowered-loop plant. Each loop consists of a steam generator, hot-leg piping, cold-leg piping, two reactor coolant pumps, a high-pressure injection system (HPIS), a low-pressure injection system (LPIS),

and one accumulator. Looo A contains a pressurizer that connects to the h'ot leg. The break is located 3 m from the vessel on a cold leg of loop A.

The cold legs of the unbroken loop (B) are combined for calculational efficiency.

The secondary side of each steam generator is attached to the main feedwater inlet, the auxiliary feedwater inlet, and a long pipe to the steam outlet with a side connection to a safety valve that vents to the atmosphere.

We represented the safety valves with a FILL component in which the mass flow out is a function of secondary pressure. Geometry and other plant data were obtained from the Midland Final Safety Analysis Report (FSAR).3

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The vessel was model~ed with four azimuthal segments, two radial rings, and ten axial levels. The core region was represented by four levels. The upper plenum was modeled with two levels permitting the vent valves to be above the not and cold leg connections. Two axial levels below the downcomer were included to allow reasonable predictions of lower plenum sweepout during blowdown. The model included connections from the upper head to each hot leg to simulate the upper-head circulation observed by B&W in their flow tests.

These connections were needed because a core barrel modeled in TRAC with only a single radial ring cannot model adequately the flow through the iJpper plenum, into the upper-head, and back down into the upper plenum.

The core was modeled with four average rods and four hot rods. Peaking factors of 1.7,1.9, 2.1, and 2.3 were used for the hot rods.

These spanned the range of expected values.

In this report " hot rod" refers to the rod with a peaking factor of 2.3.

Before initiating the break, a steady-state operating condition was obtained.

Input variables are listed in Table I.

Table II gives the design and TRAC-calculated conditions for steady state.

The calculated. temperature (589 K) and pressure (15.15 MPa) at the vessel outlet are close to the design values of 590.3 K and 14.98 MPa, respectively.

The break was placed midway in the 6-m-long pipe between the vessel and one of the main coolant pumos in loop A.

The two pipe segments.(3 in each) adjacent to the break were modeled witti 9. cells each, graded from 1 m (away from the break) to 0.1 m (next to the break). The flow in each segment was i:alculated in a fully implicit manner to maintain a reasonable time step during the high-velocity blowdown phase. The break is modeled as a pressure boundary condition. The bruk ;,ressure during the first 1 s of the transient,A g

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was decreased linear.ly from the steady-state value to atmospheric pressure (0.1 MPa). Thereafter the break pressure rose stowly to 0.328 MPa before leveling off at 0.31 MPa, simulating the pressure in the containment building (Fig.2).4 Table III lists the event sequence calculated for this transient.

'The total break flow (Fig. 3) increased to 32 000 kg/s in the first second before falling off to 5000 kg/s after 14 s.

Accumulator flow began at 14.2 s, and the LPIS was activated at 20.6 s (Fig. 4). Until the end of blowdown (25 s), most of the liquid injected by the emergency core cooling (ECC) system was forced out the break. The core average pressure during and af ter blowdown is shown in Figs. 5 and 6.

The core liquid and lower plenum liquid were swept out by 12 s and 20 s respectively (Fig. 7). The lower plenum remained dry until the end of the blowdown. Flow through the vent valves (Fig. 8) was significant only during blowdown when it accounted for as much as 10% of the flow leaving the core barrel region. The pressurizer emptied by 30 s and, because it was connected to the broken loop, did not contribute to refilling the vessel. From 25 s to 36 s very little break flow occurred and the lower plenum was refilled by accumulator and LPI flow.

By 33 s reflood had started. The core liquid oscillated during reflood in response to rapid boiling of liquid as it came into contact with the fuel rods.

Between 35 s and 55 s the resulting pressure oscillations (Fig. 6) forced liquid up the downcomer and out the break (Fig. 3) in three major surges. The effect on vessel refilling is shown in Fig. 9 where total liquid mass falls in the 35--55 s interval. The vessel filled steadily thereaf ter.

The peak cladding temperatures for the average and hot rod during blowdown were 700 K and 953 K.

The peak temperatures for the entire transient were 812-707 K and 962 K, respectively. Between peaks the average rods cooled 125 K because a significant amount of liquid was entrained in the axial vapor flow. /,15

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The highest temperature found in any average rod is shown in Fig.10, and Fig.11 shows the temperature history of two locations on the hot rod corresponding to the blowdown peak and absolute peak locations.

The reactor trip also tripped the feedwater system, which decreased linearly to zero in 17 s.

The water level in the loop A steam generator (Fig.12) climbed from a steady-state level of 5.5 m to 6.5 m in 20 s.

During this time, the steam outlet pressure rose to 7.4 MPa in loop A and 7.3 MPa in loop B (Fig.13), which opened the atmospheric relief valves ( ARV). From 2--9 s, the ARVs in each loop released 1200 kg of steam, including 'the 15%

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turbine bypass condenser flow (Fig.14).

Sunnarizing, the TRAC calculations predict a maximum cladding temperature of 962 K in the hot rod with all rods quenched by 123 s.

The blowdown phase lasted 25 s and the lower plenum was swept out by 20 s.

By 33 s reflood had started and the core began to refill. Most of the liquid supplied by the pressurizer and half that supplied by the accumulators was forced out'the break; hence, core refilling was acconplished mostly by the LPIS.

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TABLE I PARAMETERSFORLARGE-BREAKAh!ALYSIS Power: 2452 MWth w

Fuel Rods:

e axial power profile 0.39, 1.18, 1.12, 1.17, 0.48 (5 levels bottom to top) e power peaking f actors 1.7, 1.9, 2.1, 2.3 Liquid Activation Pressure Temperature System MPa/ PSIA K/0F HPIS 11.13/1615 288.7/60

(+ 10 s delay)

LPIS 1.41/205 288.7/60 Accumulator 4.24/615 305.4/90 Cold-leg-break location: 3 m from vessel (loop A)

Turbir-bypass and ARV set pointA7.067 MPa (1025 psia)

%2.

x,._ -

Accumulator check valve set point:

e opens if static AP <0.1385 MPa (20.09 psia) e closes if static AP >0.065 MPa (9.43 psia)

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TABLE II STEADY-STATE CONDITIONS Design TRAC Reactor Power 2452 MW 2452 N th th 0

0 Vessel outlet temp.

590.3 K (603 F) 589 K (600.5 F) 0 Vessel in'et temp.

563.7 K (555 F) 562.6 K (553 F)

Vessel outlet press.

14.98 MPa (2173 psia) 15.15 MPa,(2197 psia)

AP across core 0.441 MPa (64 psia) 0.421 MPa (61 psia)

Total AP of one loop 0.729 MPa (105.7 psia) 0.739 MPa (107.2 psia) 4 4

Total primary side flow 1.654 x 10 kg/s 1.648 x 10 kg/s 6

6 (131.3 x 10 lb/h)

(130.8 x 10 lb/h)

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TABLE III TABLE OF EVENTS Time (s)

Event 0.0 Start DEG break ramp Pressurizer begins to empty 0.17 Reactor trips,' power decay initiated.

Atmospheric Rel.ief Valves open Main FW tripped, TSV tripped 1.0 200% DEG break 1.17 TSV closed 4.0 Blowdown peak temp for average rod: 700 K 10.3 HPI initiated (both loops) 14.2 Accumulator Flow begins (both loops) 15.2 Blowdown peak temp for hot rod:

953 K 17.2 Main FW stops 20.0 Lower plenum dry 20.6 LPI initiated (both loops) 25.0 Blowdown ends 30.0 Pressurizer empties (< 0.1 m) 31.9 Peak temp for hot rod: 962 K, average rod:

707 K 33.0 Reflood begins 36.0 Accumulators empty 76.0 Average rods quenched 123.0 Hot rod quenched l15

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REFERENCES 1.

" TRAC-PD2 An Advanced Best-estimate Computer Program for Pressurized Water Reactor loss-of-Coolant Accident Analysis," Los Alamos National Laboratory report LA-8709-MS/NUREG/CR-2054.

2.

NRC Memorandum from R. J. Mattson, NRR/DSI, to 0. E. Bassett, RES/DAE, " Request for TRAC Calculational Assistance," dated October 23, 1981.

3.

" Midland Plant, Units 1 and 2.

License Application, Amendment 33:

Final Safety Analysis Report," Consumers Power Co., Jackson,

~~-

Michigan (November 16,1977).

4.

G. W. Johnson, F. W. Childs, and J. W. Broughton, "A Comparison of

'Best Estimate' and ' Evaluation Model' LOCA Calculations: The BE/EM Study," Idaho National Engineering Laboratory report PG-R-76-009 (December 1976).

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