LD-94-017, Forwards Responses to Questions Raised by ACRS ABB-CE Std Plant Designs Subcommittee at 940209 Meeting Re Sys 800+

Forwards Responses to Questions Raised by ACRS ABB-CE Std Plant Designs Subcommittee at 940209 Meeting Re Sys 800+

ML20078K339
Person / Time
Site:	05200002
Issue date:	03/01/1994
From:	Brinkman C ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY, ASEA BROWN BOVERI, INC.
To:	Coe D Advisory Committee on Reactor Safeguards
Shared Package
ML20078K227	List: LD-94-017, Forwards Responses to Questions Raised by ACRS ABB-CE Std Plant Designs Subcommittee at 940209 Meeting Re Sys 800+
References
ACRS-2924, LD-94-017, LD-94-17, NUDOCS 9502090188
Download: ML20078K339 (56)

Text

-

a a ABB ASEA BROWN BOVERI March 1,1994 LD-94-017 Mr. Douglas Coe Advisory Committee on Reactor Safeguards U.S. Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, MD 20814

Subject:

ABB-CE Responses to ACRS Questions on System 80+m

Dear Mr. Coe:

This letter provides responses to questions (Enclosure) raised by various members of the Advisory Committee on Reactor Safeguards (ACRS) ABB-CE Standard Plant Designs Subcommittee at the meeting held on February 9,1994. I believe that the attached responses will clarify the issues raised by members of the Subcommittee.

If I can be of further assistance regarding these matters, please do not hesitate to call me, or Mr. Stan Ritterbusch of my staff at (203) 285-5206.

Very truly yours, COMBUSTION ENGINEERING. INC.

f, n-l - -

C. B. Bri an, Acting Director Nuclear Systems Licensing

Enclosures:

As stated cc: T. Wambach (NRC)

P. Lang (DOE)

ABB Combustion Engineering Nuclear Power Co cuc.nnFnn - ~ '~- 1000 Peaspee A uc Te+eonone (203i 6881911 9502090188 940310 QSj"cg Bg , ,, 2 Qy20al,9,8S 99 eg9512 EN WSOA PDR ACRS 2924 PDR

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Enclosure to LD-94-017 ABB Combustion Engineering System 80+ Standard Plant Design Responses to ACRS ABB-CE Standard Plant Designs Subcommittee Questions (February 9,1994 meeting)

_ __ _ _ _ . _ _ _ _ _________m

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l Responses to ACRS ABB-CE Standard Plant Designs Subcommittee Questions (February 9,1994 meeting)

Ouestion 940209-1:

What is the impact of the higher core mass flow rate resulting from the lower inlet temperature for System 80+ relative to that in Palo Verde on core vibrations? I

Response

Flow-induced vibratory loads are proportional to the coolant velocity heads in the core region.

Control of flow-induced (are vibrations is accounted for in the System 80+ design by limiting the core mass flow rate for post-core full-power operation such that that the coolant velocity head for the System 80+ core is enveloped by the maximum coolant velocity head in currently operating ABB-CE plants. This ensures that the flow-induced vibratory loads in the System 80+ core are no greater than those resulting from current operational experience.

Flow-induced core vibrations are not a problem in currently operating ABB-CE plants.

1 D Ouestion 940209-2:

Is ABB-CE using ATHOS I or 11 to support design certification? Please provide a copy of the applicable code reference manual. Additionally, more information is needed on the ABB-CE approach for evaluating shell side flow induced vibration, including critical cross-flow velocity. The ACRS would also like ABB-CE to address the upper hot side tube bundle dryout phenomenon, experienced at Palo Verde, to show that they understand it well enough to preclude it from occurring in the System 80+ steam generator.

Response

ABB-CE uses the ATHOS 2 code for steam generator evaluations supporting design certification. A copy of the code manual is provided herewith for your information.

ABB-CE made a presentation to the ACRS in September 1990, on Flow Induced Vibration Analysis with respect to the Palo Verde economizer corner tube wear phenomenon. Since that time, ABB-CE was required, as part of our contract for Yonggwang (YGN) Units 3 & 4 to perform a proof test of design changes implemented. The attached sketch illustrates the model concept utilized, which was very successful in that no tube vibration greater than 1 mil was observed. The results apply to System 80+ as well due to geometric similarity to the YGN design. The test results will be discussed at a future ACRS meeting (possibly April 5 &

6,1994) using data, test model photographs and available videos. The Korean Institute of Nuclear Safety (KINS) has accepted the results of this test series as demonstrating satisfactory resolution of this issue.

With regard to the upper hot side tube bundle dryout issue, which occurred in 1993 at Palo Verde Unit 2, ABB-CE is in the process of developing a System 80+ Steam Generator White Paper which includes a discussion of this topic. The associated ATHOS analysis results for System 80+ are quite favorable and the white paper is expected to be completed within a few weeks. ABB-CE will provide a copy of the completed white paper to the ACRS and a presentation to the ACRS will be made at the upcoming April 5 & 6,1994 meeting, if desired.

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i SECTION I

. 1 INTRODUCTION I

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M ENGINEERING SECTION I INTRODUCTION t

The tables and figures provided describe and illustrate Combustion Engineering's capability in three-dimensional, two-phase flow field analysis for heat exchangers in general and PWR steam generators in particular.

In several years of developnent, the ATHOS(l) code has evolved to an econcrnical tool for developing an understanding of flow related problems in PWR steam generators. In recent R&D work at CE, ATHOS has been used in the early stages of design definition for replacement steam generators to provide compara-

[ tive evaluations of design alternatives.

In the discussion and figures which follow, the ATHOS code is summerized and typical output results are providc; to illustrate capabilities.

E (1) Singhal, A. K. , et al., "ATHOS - A Computer Program for Thermal-Hydraulic Analysis of Steam Generators. Volme 1: Mathematical annd Physical Models and Methods of Solution. Volune 2: Programer's Manual.

Volune 3: User's Manual," EPRI-NP-2689-CCM, October 1982.

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PREDECESSORS HELIOS (1975) AND CALIPSOS (1978) l l DEVELOPED BY CHAM UNDER CE/KWU DIRECTION l

I DEVELOPMENT SPONSORED BY EPRI

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DEVELOPMENT PERFORMED BY CHAM CE TECHNICAL ADVISOR 10 EPRI FIRST VERSION (URSULA) RECEIVED JUNE, 1979 GENERAL RELEASE (ATHOS) VERSION, 1982 I MODIFIED BY CE TO INCORPORATE TUBE BUNDLE REPAIRS AND INFLUENCE OF SLUDGE DEPOSITS ON THERMAL-HYDRAULICS VERIFICATION SIMULATIONS BY C-E AND OTHERS I

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j. brief sunmary of solution procedure, code output, and gecr.1etry modeled.

I III.1 CAPABILITY 0 THREE DIMENSIONAL, WO-PHASE FLOW DISTRIBUTION ANALYSIS 0

STEADY STATE AND TRANSIENT O

HO E ENEOUS AND ALGEBRAIC SLIP O

POLAR AND CARTESIAN COORDINATES 0

PRIMARY AND SECONDARY FLUIDS COUPLED THRU HEAT TRANSFER 0

CHOICE OF HEAT TRANSFER CORRELATIONS

" 0 TYPICAL H0 DEL CONTAINS 1000 CELLS i

TYPICAL CELL 18" X 14" X 30

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^ I:0RE DETAILED HODELS POSSIBLE 0

ALTr0 MATED INPUT PREPARATION I O NUMERICAL AND GRAPHICAL OUTPUT VECTOR AND SCALAR PLOTTING, SPLIT FLOW ECONOMIZER MODEL, l AND TUBE PLUGGING / SLEEVING CAPABILITIES DEVELOPED BY CE.

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l SECTION III, CODE DESCRIPTION (CONTINUED) l III.2 SOLUTION PROCEDURES E The ATHOS solves the mass, momentum and energy equations for each node.

Empir ical correlations are used for solid-to-fluid friction, heat transft", and slip between two phases.

0 ".ne finite-difference equations are obtained by integrating the partial differential equations over control volunes.

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0 An implicit, iterative, successive-substitution method is employed.

All flow variables except pressure are computed slab-by-slab I (i.e. at one horizontal plane at a time), while the pressure is calculated for the whole calculation domain simultaneously. Cal-culation steps are repeated until convergence.

The convergence criterion checks on the error residuals of all finite-difference equations, in each control cell. l

- l 0 The overall solution procedure is the same for all flow models, j l

- The solid-to-fluid frictional resistances are calculated by using empirical correlations for axial flow and crossflcw over tubes, and for l losses due to sudden contraction and expansion in concentrated resist-ance devices such as' tube support plates.

I 0 The .Dittus-Boelter correlation is used for the primary-side film l

l heat transfer coefficient.

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SECTION III, CODE DESCRIPTION _ (CONTINUED) l Three options are provided for the secondary-side convective and boiling heat transfer coefficients. These are:

--the Rohsenow, Collier-Pulling correlation;

-the tellouche-Zolotar (EPRI) correlations; and

--Dingee and Chastain for axial convective flow, Dwyer for cross convective flow, and a Rohsenow (power law) type correlation for boiling.

In the algebraic slip model, the drift flux velocity is calcu-lated by using the basic forms of Zuber-Findlay, with the distri-bution parameter and other constants as recomended by Lellouche and Zolotar.

III.3 COMPUTATIONAL RESULTS I

The ATHOS output includes a summary of geometry data, physical properties of primary and secondary fluids and tube metal, empirical correlations and other parameters used to obtain a converged solution.

For each node in computational domain the code provides the following information:

SECONDARY FLUID:

1HREE COORDINATE VELOCITIES AND MASS VELOCITIES I O PRESSURE O

ENTHALPY h

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SECTION III, CODE DESCRIPTION (CONTINUED)

Figure III.4.2 is the ATHOS model of a U-tube steam generator and I depicts the vertical and horizontal nodalization. A typical steam generator is modeled using approximately 1000 nodes. A more detail model of a particular steam generator region is possible.

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ATHOS APPLICATIONS At Combustion Engineering the ATHOS code has been applied to study, detailed thermal-hydraulic conditions of steam generators, investigate and optimize various design options and predict steam generator performance under various

. conditions which include tube bundle repairs, fouled tube support crevices and sludge deposits. The code has also been used to determine the influence of local thermal-hydraulic conditions on sludge deposition, tube denting and other f

l phenomena experienced by steam generators.

Examples of ATHOS applications are presented.

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I gggggUSTION E806INEERING SECTION IV.1 I THERMAL-HYDRAULIC ANALYSIS IV.1 EERMAL-HYDRAULIC ANALYSIS

. The ATHOS computer program has bee ~n routinely used to determine the thennal-hydraulic characteristics of various C-E, Westinghouse, and KWU steam generator designs. Figures IV.1.1 through IV.1.4 depict the secondary fluid velocities and quality at various locations inside the  ;

steam generator. The location and parameters are user selected options.

Figure IV.1.1 Velocity vectors (R,0) at 10 inches above the tubesheet. i The figure shows secondary fluid velocities just above the tubesheet.

This infonnation can be used to identify stagnant or low velocity zones on the tubesheet. These zones usually coincide with sludge deposition and associated problems.

Figure IV.1.2 Velocity vectors (R,2) in the middle of hot side, i

This figure illustrates secondary fluid velocities in a vertical plane.

Figure IV.1.3 Secondary fluid quality at 10 inches above the tubesheet.

Figure IV.1.4 Secondary fluid quality in the middle of hot side.

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SECTION IV.2 DESIGN OPTIMIZATION AND I COMPARATIVE EVALUATION IV.2 DESIGN OPTIMIZATION AND COMPARATIVE EVALUATION The ATHOS code has been utilized to evaluate the effects of various design and operational options on steam generator perfonnance.

Figures IV.2.1 through IV.2.7 illustrate some of the examples.

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Figure IV.2.1 Axial mass velocity distributions 16 inches above hot side tube-sheet with and without flow baffle.

This figure illustrates the influence of the flow distribution baffle, one of the steam generator design options. 'Ihe code can be used to optimize flow distribution baffle design for the desired thermal-hydraulic conditions above the tubesheet.

Figure IV.2.2 Fluid quality on the hot side tubesheet with and without flow baffle.

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Figure IV.2.3 Velocity / quality distribution above tubesheet.

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Figure IV.2.4 Flow distribution in split flow economizer.

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The figure illustrates flow distribution inside a split flow economizer, one of the steam generator design options.  ;

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, Figure IV.2.5 Comparison of axial velocities at 14 inches above the tube-j sheet with 80/20 and 60/40 (hot / cold) feedwater flow splits.

l l Figure IV.2.6 Comparison of secondary fluid quality at 10 inches above the tubesheet with 80/20 and 60/40 (hot / cold) feedwater flow splits.

Figures IV.2.5 and IV.2.6 illustrate the effects of feedwater flow split on secondary fluid thermal-hydraulics above the tubesheet. Feedwater flow split is one of the design / operation options for a desired flow conditions.

Figure IV.2.7 Comparison of axial velocities at 14 inches above the tubesheet for two different feedwater temperatures.

This figure demonstrates the influence of feedwater temperature on fluid conditions above the tubesheet.

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HOT COLD HOT COLD 80/20 FEEDWATER FLOW SPLIT 60/40 FEEDWATER FLOW SPLIT VEL. RANGE: -0.9 TO 5.1 FT/SEC VEL. RANGE: 0.14 to 4.46 FT/SEC BAND SCALE VEL. RANGE (FT/SEC) BAND SCALE vel.. RANGE (FT/SEC) 1 -0.9 TO -0.3 1 -0.14 TO 0.32 3 0.3 TO 0.9 3 0.78 TO 1.24 5 1.5 TO 2.1 5 1.70 TO 2.16 7 2.7 TO 3.3 7 2.61 TO 3.07 9 3.9 TO 4.5 9 3.54 TO 4.0 FIGURE IV.2.5 COMPARISON OF AXIAL VELOCITIES AT 14 INCHES AB0VE THE TUBESHEET WITH 80/20 AND 60/40 (HOT / COLD) FEEDWATER FLOW SPLITS.

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Figure IV.3.1 Hot side quality contours for extremes of flow baffle crevice fouling.

As the baffle becomes completely fouled, it alters flow through the baffle. The flow redistribution changes quality distribution dramati-I- cally as illustrated in this figure.

1 Figure IV.3.2 Comparison of hot side quality distribution for various tube plugging and sleeving configurations.

This figure shows the change in the hot side quality distribution for as-designed steam generator through three different tube plugging and sleeving configurations. Most of the changes are due to reduced heat transfer surface at plugged locations.

Figure IV.3.3 Comparison of axial' velocities 14 inches above the tube-sheet and sludge heights.

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[ eemmesnon)smamasama SECTION IV.3 PERFORMANCE PREDICTIONS (CONTINUED)

A comparison between axial velocities 14 inches above the tubesheet and sludge height is presented in this figure. Note that the maximm sludge deposition occurs when velocities are either negative (downwards) or very low positive (upward).

Figure IV.3.4 Graphical comparison of tubesheet sludge pattern with thermal-hydraulic conditions.

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I Figure IV.3.5 Comparison of steam generator thermal-hydraulics with denting distribution.

I This figure depicts a correlation between denting and thermal-hydraulics.

Proper materials and tube support designs may help minimize such phenomenon. i l

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Ouestion 940209-3 Describe the criteria used to size the flow holes in the lower grid flow plates.

Response

A debris-resistant bottom spacer grid to prevent fuel failure caused by debris is one of the fuel design enhancements included in the System 80+ standard plant reactor core. This grid design, termed the Guardian grid, is compatible with the System 80 reactor core design and has already been implemented in reload fuel batches at Palo Verde. No changes to the fuel assembly lower end fitting or to the reactor internals were required to accommodate the Guardian grid design in the System 80 reactor core.

The Guardian bottom spacer grid design has special features (e.g., tabs, arches) intended to capture and retain debris in a harmless location in the fuel assembly. The presence of these special features results in smaller flow passages between the fuel rods and grid strips as compared to those in the conventional ABB-CE grid design. However, it is difficult to characterize the flow passages by a specific dimension since they are not uniform in geometry. The geometry of the special features was established through a combination of considerations:

a) I!ydraulic - Since the Guardian fuel was expected to be co-resident in some cores with non-Guardian fuel, there were limitations on the pressure drop differential created by the special features.

b) Fabrication - The special features could not be permitted to become a new source for debris, so they were limited in size to be compatible with the formability of the grid strip material.

c) Debris Screening - The combination of the basic grid geometry and the special features was required to block and retain a significant percentage of what was considered to be representative debris.

The actual sizes of the flow passages for the Guardian grid design were first established by performing analyses and reviewing the test data and fabrica on results from earlier designs.

The proposed Guardian design was then built in full-size prototype form and installed on a full-scale (except for length) fuel assembly. The assembly was next subjected to a flow test at representative flow velocity to measure pressure drop and to perform a debris injection test.

The injection test examined the effect of approximately 100 debris particles, including metallic wires, shavings, and chips of various sizes. The types and size ranges of the particles were selected to be representative of debris observed at ABB-CE and other plants. The testing confirmed that the Guardian grid captured and retained more than 90 percent of the injected debris.

Question 940209-4:

Explain the apparent discrepancy between the statements in Chapters 4 and 12 of CESSAR-DC regarding the use of cobalt-based materials for equipment in contact with the primary system coolant. The statement in Chapter 12 could preclude the use of current materials (indicated in Chapter 4) since current materials probably would not meet the 0.02 wt.% limit

' indicated in Chapterl2.

Resnonse:

The System 80+ design maintains tight limits on cobalt content for austenitic stainless steels and nickel base alloys in contact with primary system coolant which are significantly reduced compared to the limits imposed on previous NSSS designs. Limits on cobalt impurities in previous NSSS specifications for austenitic stainless steels and nickel base alloys were 0.2 ,

wt.% and 0.1 wt.%, respectively. For System 80+, the limit on cobalt content for austenitic stainless steel materials, including weld deposited stainless steel cladding, and for nickel base alloy is 0.05 wt.%. For nickel base alloy steam generator tubing, a cobalt impurity level of 0.015 wt.% or less is specified. These limits represent the minimum cobalt impurity levels )

that are reasonably achievable in commercial production of these types of materials. More restrictive limits would require unnecessarily expensive special melt processes by material producers and would force the user into purchasing heat lot quantities (i.e.15 to 20 tons of material) for each type of material and product form specified.

Other sources of cobalt introduced into the primary system are materials used for wear )

resistance applications. These include hardfacing on reactor internals and valves and pins and l

latches used in control element drive mechanisms (CEDMs). Cobalt base materials have been used for these applications because of their wear resistance. However, low-cobalt or cobalt-free materials will be selected in place of the cobalt base alloys where proven alternative materials exist for the specific application to reduce the amount of cobalt introduced into the primary system Based on the above, Paragraph B.1. of CESSAR-DC Section 12.3.1.3 will be revised to change the limit on cobalt content of primary system materials for equipment in direct contact l with the primary coolant to 0.05 wt.% or less, and to permit exceptions to this limit for cases l

where no proven alternative material exists. This revision eliminates the apparent discrepancy between the statements in Chapters 4 and 12 regarding the use of cobalt base materials for equipment in contact with the primary system coolant. Further, the limits specified are consistent with the revision of Section 5.2.7.1 of the ALWR URD regarding cobalt limits which is being processed by EPRI.

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inspection activities and' reemos the anticipates personnel i

l ==ia-mi- entenaed meresco 1mapins la high Lation areas will be used, Weastee possible, to mini =ise the z. -- of amin +====aa required. une lighting 1 riatures are to =inime == par ====1 esposure during l

=mine'a===ma. Wheem gggggggg age 15 annotemnoe with

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7. spent mael Pool Decca+==ination . ,

, system so+ provados the aspebility to see high pressure domines =1i==a water for tas deonataminstdan the spent emel pool. Altmenstive methods of decon*'=imties, such as uns of a striggebla costing, may be evaluates Dy the operator,  !

as y. 2 1.

32.s.1.s ammmun amen ammtzal scuros term control is as lagpartant aspeak of the System to+"

I dessen. une ro11 ewing easien saatur== reenos the overs 11 does '

dhpa to operation, maim * - , and inspection activities.  !

Fast Perfarmanam A.'

the system eM" design femteres assure low primary system sources with improved fuel clad leakage performanon of less than 0 1% feel clad failures, as unl1 as an extended fuel  ;

oyale.

3. carrosion Product control system so+* desdem $aolases design teetures that reduce carrosion preenom proemotion in the priumry system.

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she system so+= ten opmotries primary system materials.wi aozzesico zetes and very low cobalt impurities 0 w/o for egoipment in direct contact with the cool e EM4 Westvigg toe 1(Ms.

The passemos of ant _ - . . - nasjr x M __

a protaams with het clos in the curresst generation oC sumoleer . En the Systems se+ design, the remotor com pump bearings will be designed to

=i=i-i- the praecm or antimony.

stama gaaerator tubes are fabriantes to relieve stresses te seense etzees oorr=4 = arocking. shis will reemos the Whiitty og tube plugging activities ana furtteur reemos main &-=a= esposures.

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Ouestion 940209-5:

Clarify and explain the differences between the SCU and DNB convolution methods.

Resnonse:

Statistical Combination of Uncertainties (SCU) is a methodology used by ABB-CE to combine the various uncertainties in COLSS/CPC setpoint analyses. It is based on the -

observation that not all uncertainties occur in the most adverse direction at the same time.

Modified SPU (MSCU) is an improvement on the SCU methodology which reduces excess conservatism in the DNBR overall uncertainty factors for COLSS and CPC, primarily through the statistical combinntion of several uncertainty components previously applied deterministically. The results of the SCU and MSCU methodologies are used in the calculation of penalty factors for the setpoint analyses.

DNB convolution is a methodology used in the assessment of plant transient analyses CESSAR-DC, Section 15.0.4). In this methodology, a fuel pin histogram of DNBR versus i number of fuel pins with that DNBR calculated for a transient is statistically convoluted with -

the probability of DNB at each value of DNBR to determine the number of fuel pins predicted to fait due to DNB. The DNB convolution methodology results in a probability-weighted sum of pins predicted to experience DNB during the transient, i

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Ouestion 940209-6 Please provide information on the effectiveness of Heated Junction Thermocouple (HJTC) in reactor water-level measuring applications. Specific points ofinterest are: the dynamic sensor response to a change in water level (the time constant), and a brief history of the frequency and modes of failure. 1 Resnonse:

The HJTC differential temperature response time starts when the sensor is uncovered and extends until the AT setpoint value signaling an uncovered condition is reached. The uncovered condition signifies a lowering of the reactor coolant level. The response time varies from 3 to 25 seconds, depending on the reactor coolant pressure and whether the sensor is covering or uncovering. When the sensors are covering, the response time is independent of reactor coolant pressure and sensor position, and varies from 3 - 5 seconds. When the sensors are uncovering, the response time of the top sensor is also l

independent of pressure, and varies from 5 - 9 seconds. The response time for uncovery of the 1 remaining sensors varies from 8 - 11 seconds at 200 psig reactor coolant pressure to 21 - 25 seconds at 2000 psig.

The HJTC probe frequency and modes faihtre history is as follows. Failures for the HJTC i probe have occurred on two subcomponents of the probe. The first subcomponent is the heater lead wire. The original design of the HJTC probe utilized a small diameter oxygen free high conductivity (OFHC) copper lead to connect the electrical connector to the heater.

Processing steps used during the fabrication of the HJTC probe sensors included a high temperature stress relief anneal of each sensor. This high temperature exposure caused  ;

recrystallization and grain growth in the OFHC copper resulting in single grains spanning the l full diameter of the lead wire and very low strength. The large temperature gradient at the l reactor vessel head elevation between the portion inside the reactor vessel and the portion at  ;

the top of the HJTC pressure housing created the potential for these premature failures of the l copper heater lead wire. The cause of this failure was attributed to creep due to thermally l induced tensile loading.

The material of this lead wire was subsequently revised to GLIDCOP@ AL-20, which is a dispersion strengthened copper, and failures of this item were almost completely eliminated. i GLIDCOP@ AL-20 has much higher strength (by a factor of 4 to 5 higher) at system l operating ':rperature and is relatively unaffected by the manufacturing process. l l

The other subcomponent failure was the welded connection between the electrical connector I backshell and the sensor assembly outer sheath. Although handling precautions are provided, l failure of this welded joint resulted from handling problems in the field.

)

While the GLIDCOP@ AL-20 exhibited good service performance, the nature of this material  ;

caused some difficulties with the bending and joining steps in the manufacturing process. An improved HJTC probe design utilizes a larger diameter nickel-clad copper heater lead wire for increased strength. The fabrication sequence is similar to earlier probes including the high I

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Response 940209-6 (continued):

temperature thermal cycle. However, the nickel cladding of the lead wire provides sufficient additional strength even though the copper core is softened. The as-fabricated strength of this heater lead wire material will prevent recurrence of the type of failures experienced with the original design. The improved design also utilizes a much stronger full penetration braze joint  ;

at the connector backshell to sensor sheath joint to minimize the potential for those failures i attributed to handling.

The improved design is expected to eliminate the failure mechanisms associated with the earlier designs, however, there is an insufficient population of installed probes to establish a failure frequency. Twenty HJTC probes of the modified design have been shipped to utilities )

as replacement spares. No failures have been reported for two (2) of these probes installed to j date.

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j Ouestion 940209-7:

What capability does System 80+ have for handling frequency degradation without tripping  ;

the reactor?

Response

There is sufficient reactor core thermal margin in the System 80+ plant to allow a frequency degradation of approximatey 3 Hz before a reactor trip occurs. The Core Protection Calculators (CPCs) will create a reactor trip signal when the RCP shaft speed decreases to approximately 95% ofits normal value (1190 rpm). The allowable 5% decrease prior to reactor trip corresponds to a frequency degradation of 3Hz. l l

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s Ouestion 940209-8:

Clarify the ABB-CE position regarding the improvement in reactor vessel embrittlement obtained by reducing the copper content in the reactor vessel to very low levels.

Response

The copper content for the System 80+ reactor vessel beltline materials has been specified to be as low as practically achievable by modern steel making practice and to be in accordance with the EPRI ALWR Utility Requirements Document. Copper content for the reactor vessel beltline forgings and weld metals is specified to be 0.03% maximum (CESSAR-DC Section 5.2.3.1).

The benefits achieved from the low specified copper content are a minimum influence of '

copper on radiation damage to the beltline materials and a minimum predicted shift in RTm for the end-of-life fluence to the reactor vessel beltline. Low limits on copper content reduce ,

the amount of irradiation damage to reactor vessel beltline materials influenced by copper content. Controls on other impurity elements such as phosphorous, sulfur and vanadium  !

ensure that secondary embrittlement mechanisms associated with these elements will be minimal. The overall objective is to provide the most radiation damage resistant material available based on the current understanding of irradiation embrittlement.

Regulatory Guide 1.99, Revision 2 provides the current recommended methodology for predicting radiation embrittlement of reactor vessel materials. The prediction correlation was i based on an extensive database of surveillance data from power reactors. The database consisted of materials with a range of copper and nickel contents on which the correlations were based. The range of material chemistries did not cover very low copper contents as specified for System 80+, so there is currently no experimental validation of the benefit of -

further decreases in copper content on predicted shift.' However, the trends established by the correlation revealed decreasing effects of irradiation on properties with lower copper and nickel contents. The trends exhibited by the data are addressed for the System 80+ beltline materials by specifying the lowest, practically achievable limit for copper impurities.

Regulatory Guide 1.99, Revision 2 was issued in May 1988. Since that time, additional surveillance data has become available from operating plants and has been added to the database used to develop the prediction correlations. Investigations are currently in progress to re-evaluate the prediction methodologies for ARTm and decrease in upper shelf energy due to irradiation in Regulatory Guide 1.99 Revision 2. Establishing the low copper limits for System 80+ beltline materials will contribute to minimizing any potential impact of a new revision to the Regulatory Guide prediction methodology.

Further, the predicted irradiated properties of the reactor vessel are used as an initial indication of the material condition and properties of the reactor vessel over the life of the plant. These predictions will be periodically supplemented by surveillance data on the actual beltline forging and weld metal materials. The System 80+ surveillance program includes more capsules and scheduled withdrawals than the minimum required by 10CFR50 Appendix

Response 940209-8 (continued):

11 and ASTM E 185 (" Standard Practice for Conducting Surveillance Tests for Light Water Cooled Nuclear Power Reactor Vessels") to ensure that beltline material properties will be regularly verified and compared to predicted properties over the design life of the plant.

Surveillance data can be utilized to adjust the predicted behavior of the beltline materials in accordance with Regulatory Guide 1.99, Revision 2. The combination of tight controls on impurities of the beltline materials, initial toughness properties and results of the surveillance capsule evaluations provide assurance of the reactor vessel beltline material toughness properties for System 80+.

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y f' The use of water systems to fight oil fires, especially in the Diesel Generator rooms was

. questioned (reference was made to work at the University of Maryland). Additionally, are the spray headers just above the Diesel Generators or at ceiling level? The concern is that if they are just above the Diesel Generators, spray may get into the generators and into the " drip-proof" electrical cabinets in the room.

s

Response

Fire suppression for the System 80+ diesel generators is provided by preaction sprinkler systems. Spray nozzles are located at the diesel generator room ceiling and spaced in accordance with NAFPA-13 criteria.

Potential fire hazards in the diesel generator rooms are from the following,1) diesel fuel oil,

2) diesel lubrication oil, 3) generator energized electrical components, 4) auxiliary electrical equipment (i.e., power panel boards), and 5) transient combustibles (i.e., solvents and ordinary combustibles).

Options for fire supression are,1) carbon dioxide, 2) halon,3) wet pipe sprinkler system,4) .

l dry pipe sprinkler system,5) foam water sprinkler system,6) preaction sprinkler system, and

7) alternative /new suppression agents. Each of these options has its own unique considerations as follows: ,

Carbon Dioxide - This would be an effective extinguishing agent, however, it was not selected because it represents an asphyiation hazard and would have to be contained in the protected area.

i Halon - The effectiveness of this agent in suppressing fuel oil or lubricatiuon oil fires with the diesel at operating temperature is doubtful. Additionally, halon production has been curtailed due to environmental concerns.

Wet Pipe - Provides the same suppression effectiveness as a preaction sprinkler l system. The presence of water in the piping presents water damage l concerns for non-fire scenarios such as pipe leak and seismic failure.

Dry Pipe - Suppression capability is less effective than wet pipe and preaction sprinkler systems due to time delay in discharging water.

Foam-water - This system is considered to be effective for oil and electrical fires. A

, , benefit is that foam water systems effect suppression with less water discharge. The foam agents, however, are corrosive and may exacerbate damage due to fire.

Preaction - This system requires actuation of a heat actuated detection device and a heat actuated sprinkler head. The heat actuation detection device is selected to be more sensitive than the sprinkler heads so that the release mechanism is actuated and pipes are filled with water when the sprinkler

)

i heads actuate. Water is considered effective for oil fire suppression as it dilutes and cools burning oil. Water is also effective for electrical fires and fires involving ordinary combustible materials.

,x ,

t Response to 940209-9 (continuedh -

Alternatives - A number of new suppression agents such as halon alternative gas  !

suppressants and water mist systems are under development. However, none of these attematives represent a viable option for the System 80+

design at this time.  ;

i Fires may occur during maintenance, testing, idle periods or when the diesel generator is loaded to supply station auxiliary power loads during a station blackout. Evaluation of the risk during these periods follows:

Maintenance - Potential fire hazards are associated with personnel error and primarily l involve transient combustible materials. Maintenance personnel would j be expected to recognize incipient fires and notify the station fire brigade  ;

to effect early suppression. The safety significance is minimal since

during maintenance the diesel generator is considered out of severice and other emergency electrical power systems such as redundant diesel ,

generator and the site combustion turbine are assured as standby -l emergency power supplies.

Testing - The potential fire hazard is due to equipment failure. During testing station personnel are present and can deenergize equipment and promptly summon the station fire brigade. This would mitigate fire damage. The safety significance is minimal since other emergency poer sources such  ;

as redundant diesel generator and the site combustion turbine are assured as standby emergency power supplies. .

Idle Periods - The potential for fire during idle periods is considered low. Fire would most likely be associated with activities involving transient combustibles.

During idle periods there is no emergency demand for the diesel generator service so that other emergency power sources such as redundant diesel generator and the site combustion turbine can be assured operable.

, Active Diesel - When the diesel generator is loaded and supplying emergency power is when selection of the optimum fire suppression strategy is demanded.

Station personnel will be in attendance to monitor diesel generator performance when it is supplying power at the local control panels in the diesel generator rooms.

The fire protection defense-in-depth philosophy is important for vital station equipment such as diesel generators. During most of the potential fire scenarios stated above the diesel generator room will be occupied so that any fire will be recognized and the station fire brigade summoned during the incipient stage. This early notification should contribute to prompt suppression with manual fire extinguisher and fire hose before equipment damage occurs. However, if the preaction sprinkler system actuates, a fine mist will result in a moisture rich environment (approaching 100% relative humidity). The generator is air cooled but will be enclosed in a drip proof enclosure as defined in NEMA 1-1.25.A. The generator windings and electrical areas have heaters to remove moisture. Short term operation in this I

s ,

Ouestion 940209-10:

What is the design capability of the doors into the Nuclear Annex from the Turbine Building and the ability of a steam line break in the Turbine Building to adversely affect the protective function of those doors?

Response

The effects of a postulated main steam pipe rupture will not impact the Nuclear Annex. The direct access / egress between the Nuclear Annex and the Turbine Building is limited to elevation 130'6" of the Nuclear Annex. Access is gained at this point through an enclosed stairwell in the Turbine Building. The single doorway to the Nuclear Annex from this stairwell enters into an enclosed corridor in the Nuclear Annex and not directly into a safety-related area.

The Nuclear Annex is protected and designed for external events such as tornado generated missiles and the tornado induced differential pressure. Therefore, the doorway at the Nuclea-Annexffurbine Building interface is designed for a total differential pressure of 2.4 psi.

The Turbine Building is an aluminum sided, steel structure. The design of the Turbine Building is not as robust as the Nuclear Annex and, therefore, it will not withstand analogous structural challenges. The differential pressure associated with a postulated pipe break will rupture the aluminum siding prior to reaching the design capability of the Nuclear Annex doors and walls.

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environment is possible if fire does not affect equipment needed to support diesel generator operation.

The potential vulnerability of diesel generators to fire during the time it is supplying emergency power to station auxiliary power equipment is recognized. For the System 80+

design, the preaction sprinkler ssytem is considered the optimum suppression system for potential fire hazards associated with diesel generators.

As a final note, during the February 9,1994 ACRS meeting, d.e Subcommittee members referred to a paper from the University of Maryland concerning suppression of oil fires. Dr. '

Quintiere and Dr. Milke (of the University of Mstyland) were contacted but were unaware of any paper developed by the University on this iss.'e. If a specific reference can be provided ABB-CE will review the document with respect t its applicability to the System 80+ design.

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Question 940209-11 To what extent do Technical Specifications allow the Alternate AC to be used as a backup for the diesel generators?

Response: ,

The System 80+ Technical Specifications allow the Alternate AC (AAC) to replace one diesel generator (DG), provided the AAC has been demonstrated to be operational within the past  ;

seven (7) days. This specification only applies in Mode 5 (Cold Shutdown) and Mode 6  !

(Refuelling), since two (2) DGs (one in each Division) are required to be operable in all other Modes. This. specification was developed as a result of the System 80+ Shutdown Risk Evaluation which concluded that two (2) sources of emergency AC (one in each Division) would enhance safety. The AAC provision provided additional operational flexibility.

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Ouestion 940209-12:

Regarding the design of the containment post-accident radiation monitors: What is the basis  !

for the environmental temperature qualification requirement? l l

Response:  ;

1 SECY-93-087 and 10CFR50.34(f) have been used to obtain guidance with regard to severe l accident equipment survivability requirements for the System 80+ design. Based on credible severe accident sequences, ABB-CE defined the equipment needed for achieving and maintaining a safe shutdown condition for the plant and maintaining containment integrity. l This information is presented in CESSAR-DC Section 19.11.4.4 along with the expected severe accident environment and equipment survivability requirements. These environmental and survivability requirements were established on the basis of experimental data and analytical assessments of credible severe accident scenarios.

In the aforementioned documentation, ABB-CE has identified containment radiation monitoring as an important parameter for severe accident mitigation. For the System 80+

design, the containment radiation monitor is placed outside the cranewall with its sensor and cabling located away from hydrogen sources, at least 10 feet away from hydrogen igniters and protected from potential radiative effects of diffusion flame burning at nearby igniters. These restrictions are based on a review of experimental data and are intended to address potential local burning effects that may occur. The positioning of the radiation monitor will be such that its expected severe accident thermal environment is bounded by the design basis (DB) l environmental qualification (EQ) envelope. As performed for the DB EQ, the severe accident containment environment was also determined on the basis of a global assessment of containment pressure and temperature.

In the event of a failure of the containment radiation monitor, containment radiation levels can be monitored using other sensors located within the subsphere or plant site and by direct containment air sampling using the Post Accident Sampling System (PASS). The PASS is expected to operate during the later stages of a severe accident. Due to the low probability of failure of the containment radiation monitor and its potential backup capability, there is a high level of confidence that the containment radiation monitoring capability would be available throughout the duration of a severe accident.

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ATTACHMENT 4 l i

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i NRC Staff Responses to ACRS Questions on the System 80+ Standard Plant Design '

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t- :g gJ76V Question $ For those plant components identified in the SSAR or Design Description which are not expected to be replaced during a 60 year '

design life, how will it be assured that such components will be

' designed to last 60 years? Will ABB-CE provide any design guidance for when to replace components that are not expected to last 60 years? ,

FOR NRC STAFF: What is the staff's position on the degree to which ABB-CE needs to specify design guidance in both of the above cases? .

Response: (1) As part of the information submitted in the design certification application, ABB-CE specifies the full range (and associated frequency) of normal, transient, test, upset, emergency, and faulted (including seismic) events that may, or are expected to, occur during the 60-year life of the plant. The required ,

information is provided in CESSAR-DC Table 3.9-1 and has been reviewed for acceptability by the NRC staff.

For ASME Code Class I components (including piping), the ASME Boiler and Pressure Vessel Code,Section III requires an evaluation of the suitability of the component for cyclic operation. The component is evaluated for the entire range of loadings identified in CESSAR-DC Table 3.9-1 to assure that it is suitable for the number of cycles assumed for the 60 year life of the component. The staff also notes that the component's '

life shall be reevaluated for any unanticipated transient events or events that exceed the limits and number of cycles of those specified in CESSAR-DC Table 3.9-1. Fatigue monitoring for Class I components will be performed by the COL licensee in accordance with Section 5.7.2.9 of the ABB-CE System 80+

Technical Specifications.

(2) With regard to the components that are not expected or (not designed) to last 60 years, the COL licensee is required to replace those components when their qualified design life is reached. 4 The staff has not changed its position on the design of mechanical components of a 60-year life plant from that of a 40-year life plant.

The staff considers the above positions adequate. No additional design guidance is required by the staff at this time.

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p 2NI' QuestioW For NRC Staff: Please provide your SERs for the TORC Code referenced in the DRAFT FSER on pg. 4-27, the CETOP-D Code referenced in item 3. on pg. 4-28, and for CEN-139-A-P " Statistical Combination of Uncertainties" referenced in item 4. on pg. 4-28.

l l Response: The attachment contains the three SERs requested: Reference 1 for l

TORC; Section 2.3 of Reference 2 for CETOP-D; and Section 2.6 of Reference 2 for SCU.

Reference 1: Letter from K. Kniel (NRC) to A. Scherer (CE), dated September 14, 1994.

Reference 2: Letter from R. Clark (NRC) to W. Cavanaugh, III (APL), dated July 21,1981.

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