ML20010E632

Forwards Addl Info in Response to Request Made to 810721 & 0812 Technical Review Meetings.Info Includes Justification for T-cold Upperhead Temp Assumption in LOCA Analysis & Discussion of Manual Valves in ECCS

ML20010E632
Person / Time
Site:	Wolf Creek, Callaway
Issue date:	09/01/1981
From:	Petrick N STANDARDIZED NUCLEAR UNIT POWER PLANT SYSTEM
To:	Harold Denton Office of Nuclear Reactor Regulation
References
SLNRC-81-083, SLNRC-81-83, NUDOCS 8109080073
Download: ML20010E632 (8)

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SNOPPS Standardized Nuclear Unit Power Plant System 5 Choke Cherry Road Nicholas A. Petrick Rockvi!Ie, Maryland 20850 Executive Director (301) 86M010 September 1,1981 SLNRC 81- 083 FILE: 0290 SUBJ: RSB Review

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j Nf  % l Mr. Harold R. Denton, Director ..L 1 J 0 Office of Nuclear Reactor Regulation 7 U.S. Nuclear Regulatory Commission 9- SEi' O 419815 ,_I2 '

Washington, D.C. 20555 u.s. m g r y m '

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Docket Nos.: STN 50-482, STN 50-483, and STN 50-486 / x' M. '

Dear Mr. Denton:

Technical review meetings were held with the NRC's Reactor Systems Branch on July 21 and August 12, 1981. As a result of the meetings, SNUPPS agreed to provide additional information. This letter contains some of the information requested.

1. Agenda item #14 from the 7/21 meeting concerned justification for the T-cold upper head temperature assumption in the LOCA analysis.

Enclosure A provides the requested information.

2. Agenda item #16 concerned manual valves in the ECCS system which, if mispositioned, would degrade the function of redundant flow trains. The response to this item was included in FSAR Revision 6 (p. 6.3-36). The two valves indicated in that FSAR change were the only two that fell into the NRC's category for requiring locking and control room position indication. However, during our review of this matter, it was determined that a valve in the condensate storage system (V-015 on figure 9.1-12) presented a similar situation. It was decided to add control room position indication to this locked-open valve as well.

3. Agenda item 440.101 concerned the applicability of WCAP 7769 to SNUPPS. Westinghouse Topical Report WCAP-7769, Revision 1, "Over-pressure Protection for Westinghouse Pressurized Water Reactors,"

is applicable to the SNUPPS units and is incorporated in the SNUPPS applications by reference in the FSAR. Tables 2-1 and 2-2 of WCAP-7769, Revision 1, present typical values for various parameters of g each class of Westinghouse-designed nuclear steam supply systems (i.e., 2, 3, and 4 loop). As would be expected, actual values @A within each class vary to some degree due to the specific design details of each plant. SNUPPS parameters, as illustrated on the //

[f attached additions (see Enclosure B) to Tables 2-1 and 2-2 of WCAP-7769, are similar to those provided for the " typical" four loop plants.

8109080073 810901 PDRADOCKOSOOOg A

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SLNRC 81-083

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4. Agenda item 440.205 concerned the applicability of the Diablo Canyon tests to the SNUPPS design. Enclosure C provides the' information requested by the NRC.

V ly yours,

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~ Nicholas A. Petrick RLS/dck/3a18 Enclos es cc: J. K. Bryan UE G. L. Koester KGE D. T. McPhee KCPL W. A. Hansen NRC/ Cal T. E. Vandel NRC/WC

Enclosure A to SLNRC 81- 083 UPPER HEAD T-COLD VERIFICATION In order to achieve upper head temperatures in the T cold zone, bypass flow was diverted into the vessei head region. A study was performed and documented in Reference 1 to determine the amount of bypass flow necessary to achieve T cold conditions in the head. As descirbed in Section 2 of Reference 1, an analytical model for upper head temperature calculation was developed for both UHI and non-UHI plants. To estimate the upper head region fluid temperature with the analytical model, numerous boundary conditions must be known. The boundary conditions used were based on experimental data obtained from a series of three hydraulic tests conducted at the Westinghouse Forest Hills facility.

These tests were the UHI flow distribution test, the 1/7 scale UHI upper internals test and the 1/7 scale 414 flow test.

To provide experimental verification of the analytical model, a 1/5 scale model upper head temperature test was developed as described in Section 3 of reference 1. Results for both UHI and non-UHI plants showed good agreement with analytical predictions. Further confirma-tion of the analytical procedures was obtained by an in-plant head fluid temperature measurement program as described in section 4 #

reference 1. The program included measurements from 2, 3 and 4 loop plants. Both UHI and non-UHI plants were measured. All three types of upper core plate designs (flat, top hat, and inverted top hat) were included as well as both neutron shield configurations (thermal shield and neutron pad). As reported in Section 4 of reference 1, good agree-ment was reached between measurements and the analytical model for the above spectrum of non-UHI plant types. This provides good assurance that the upper head fluid temperatures have been adequately calculated by the analytical model described in Reference 1.

Recent data from a UHI plant (Sequoyah Unit 1), included in the in-plant head fluid temperature program, also shows good agreement between pre-dicted and measured upper head temperatures.

In conclusion, assurance that upper head te nperature can .be maintained in the T cold zone has been provided by a verified analytical technique as described in Reference 1.

References:

1. R.H. McFetridge, D.C. Gar..er, " Study of Reactor Vessel Upper Head Region Fluid Temperature," WCAP-9401 Rev. 1, December 1978.

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Enclosure B page 1 of 2 TABLE 2-1 .

PRESSURIZER Als STEAM GENERATOR SAFETY VALVE RELIEF CAPACITY PRESSURIZER STEAH.CENERATOR Engineered '

Safeguards Number Capacity per Total Number Total Design Power of Safety Valve capacity of Safety Capacity Rating - HWC Valves 1b/hr Ib/hr Valver Ib/hr h

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1. Four-Loop Plants 3216 3 408,000 1,224,000 20 14,619,000 Consolidated Edison Company of New Yo'k Indian Point Nuclear Generating Unit No. 2 Indiana and Michigan 3381 3 420,000 1,260,000 20 17,153,800 Electric Company Donald C. Cook Units No. 1 and No. 2 420,000 1,263,000 20 14,800,000

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Fublic Service Elec- 3500 3 tric and Gas Company Salen Nuclear Gener-ating Station Unita No. 1 and No. 2 20 IB,22M08 l SNUPPS 3674 3 4ZO,000 1,260,000 II. Threa-Imop Planta 3 293,330 879,990 15 11,176,725 Virginia Electric 2546 and Power Company Surry Power Station Units No. 1 and No. 2 2774 3 345,000 1,035,000 15 12,148,647 Duquesne Light-Co.e-I pany saaver Valley Power Station 2300 3 288,000 864,000 12 10,068,845 Carolina Power and Light Cosgany N.s. Robinson ihmit Me, 2

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Enclosure B page 2~of 2 .

TABLE 2-2 i TYPICAL PLANT THERMAL-HYDRAULIC PARAMETERS

! Units 2-Loop 3-loop 4-Loop SNuppS i

Esot Oestput.- Core NWt 1,780 2.652 3,411' 3,4fl SysteenPressure pois 2,250 2,250 2,250 2,260 l 582,800

'l Coelant Flow spa 178,000 265,500 354,000 1

Average Core Mass Velocity 106 lb/hr-ft2 2.42 2.33 2.50 2,6Z Inlet Teeparature *F 54 5 544 552.5 668 6 l

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• F 581 580 588 69/,8 j Core Average Tg Ft 12 12 12 /Z j- Core Length Average Power Density kw/1 102 100 104 /C,5 A 6

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Maximus Fuel Temperature *F <4100 <4200 <4200 44200 Fuel Loading kg/l 2.7 2.6 2.6 2.7 I; Pressuriser Volume Ft3 1000 1400 1800 1800 t

i Pressuriser Volume Ratioed to O./47 Primary System Volume 0.157 0.148 0.148 l' '

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l Peak Surge Bete for Pressuriser 43,2 i Safety Valve 31 sing Transient Ft3/sec 21.8 33.2 41.0

) Pressuriser Safety Valve Plow

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at 2500 pela - +3% Accumulation Ft 3/sec 26.1 36.1 43.3 43,Z j Satie of Sefsty Valve Flow to i Peak Surge Rate 1.197 1.087 1.056 /.00 Fan 11 Power Steen Flow per Loop lb/sec 1078 1076 1038 JOS/

l M aal Ebell-side Steam /07,000

.l Gamerator Unter Maes per Imop lb 100,300 106.000 106.000 1

, l

Enclosure C to SLNRC 81-083 COMPARIS0N OF SNUPPS TO DIABLO CANYON SNUPPS and DIABLO CANYON Unit 1 have been compared in detail to ascer-tain any differences between the two plants that could potentially affect natural circulation flow and attendant boron mixing. Because of the similarity between the plants, it was concluded that the natural circulation capabilities would be similar, and, therefore, the results of prototypical natural circulation cooldown tests being conducted at DIABLO CANYON will be representative of the capability at SNUPPS.

The general configuration of the piping and components in each ractor coolant loop is the same in both SNUPPS and DIABLO CANYON. The eleva-tion head represented by these components and the system piping is similar in both plants.

To compare the natural circulation capabilities of SNUPPS and DIABLO CANYON,-the hydraulic resistance coefficients were compared. The coefficients were generated on a per loop basis. The hydraulic resistance coefficients applicable to normal flow conditions are as follows:

DIABLO CANYON UNIT 1 Ft/(gpm)2 SNUPPS Reactor Core & Internals 7.6 x 10-10 7.2 x 10-10 27.6 x 10-10

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Reactor Nozzles 36.8 x 10-10 RCS Piping RV outlet to SG inlet 4 x 10-10 SG outlet to RCP inlet 10 x 10-10

• RCP discharge to RV inlet 10 x 10-10 RC loop 24 x 10-10 24 x 16-10 Steam Generator 114.4 x 1 -10 122.0 x 10-10 182.8 x 10- 180.8 x 10-10 Flow Ratio: Diablo SNUPPS , (182.8)l/2 ,gg Canyon (180.8)

The general arrangement of the reactor core and internals is the same in SNUPPS and DIABLO CANYON. The coefficients indicated represent the resis-tonce seen by the flow in one loop.

The reacter vessel outlet nozzle configuration for both plants is the same. The radius of curvature between the vessel inler nozzle and down-comer section of the vessel on the two plants is difference. Based on 1/7 scale model testing performed by Westinghouse and other literature, the radius on the vessel nozzle / vessel downcomer juncture influences the hydraulic resistance of the flow turning from the nozzle to the down-comer. The DIABLO CANYON vessel inlet nozzle radius is significantly smaller than that of SNUPPS, as reflected by the higher coefficient for DIABLO CANYON.

• The SNtFPS reactor coolant pumps include wiers. The Diablo Canyon pumps do not include wiers; however, the effect of wiers is negligible (confirmed by tests which indicate a 5 gpm loss of he.d across the wier).

Enclosure C to SLNRC 81- 083 Page 2 The resistance coefficient for the RCS piping for both plants is the same.

Steam generator units were also compared to ascertain any variation that could affect natural circulation capability by changing the effec-tive elevation of the beat sink or the hydraulic resistance seen by the primary coolant. It was concluded that there are no differences in the original design of the steam generators in the two plants that would adversely affect the natural circulation characteristics. Indeed the circulation should be enhanced in the SNUPPS as the water feeds into the hot side.

As indicated, the difference between the total resistance coefficients for the two plants is insignificant. It is expected that the relative effect of the coefficients would be the same under natural circulation conditions such that the natural circulation loop flowrate for SNUPPS would be within two percent of that for DIABLO CANYON.

. The coefficients provided reflect the flowrate and associated heat removal capability of an individual loop in the plant. The comparison, therefore, does not take into coniideration the number of loop avail-able nor the core heat to be removed. An evaluation of the SNUPPS steam relief and auxiliary feedwater systems has been performed to demonstrate that cooling caa be provided via two steam generators following the most limiting single active failure, i.e., the failure of an atmospheric relief valve.

Loop circulation flow is dependent on reactor core decay heat which is a function of time based on core power operating hi" y. Under natural circulation flow conditions, flow into the upper heaa area will consti-tute only a small percentage of the total core natural circulation flow and therefore will not result in an unacceptable thermal / hydraulic impedance to the natural circulation flow required to cool the core.

For typical 4-loop plants (including SNUPPS) there are two potential flow paths by which flow crosses the upper head region boundary in a reactor. These paths are the head cooling spray nozzles, and the guide tubes. The head cooling spray nozzle is a flow path between the downcomer region and the upper head region. The temperature of the flow which enters the head via this path corresponds to the cold leg value (i.e., Tcold) Fluid may also be exchanged be^ ween the upper plenum region (i.e., the portion of the reactor be6 ween the upper core plate and the upper support plate) and the coper head region via the guide tubes. Guide tubes are dispersed in the upper plenum region from the center to the periphery. Because of the nonuniform pressure dis-tribution at the "oper core plate elevation and the flow distribution in the upper plenum. region, the pressure in the guide tube varies from location to location. These guide tube pressure variations create the potential for flow to either enter or exit the upper head region via the guide tubes.

Enclosure C to SLNRC 81- 083 Page 3

To ascertain any difference between the upper head cooling capabilities between DIABLO CANYON and SNUPPS, a comparison of the hydraulic.resis-i tance of the upper head regions was made. These flow paths were con-l sidered in parallel to obtain the following results:

. DIABLO CANYON UNIT 1 SNUPPS j

, Flow area (ft2) 0.77 . 844 Loss ccefficient 1.51 1.45 Overall nydraulic resistance (ft-4) 2.57 2.036 Relative head region flowrate 1.00 1.12 (Based on hydraulic resistance)

As indicated above, the effective hydraulic resistance to flow in SNUPPS is slightly less than DIABLO CANYON. Assuming that the same pressure differential existed in both plants, the SNUPPS head flow rate would be 112 percent of the DIABLO CANYON flow.

It can, therefore, be concluded that the results of the natural circu-lation cooldown tests performed at DIABLO CANYON will be representative of the natural circulation and boron mixing capability of SNUPPS. The results of these tests will be reviewed for applicability. A natural circulation cooldown test will be performed at SNUPPS prior to startup following the first refueling if the DIABLO CANYON prototype test is not completed or does not provide satisfactor" results during the first~

fuel cycle at SNUPPS.

I RLS/dck/3a21

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