ML20034F620

Forwards SSAR Markups Addressing Open Items 5.3.2-1 & 15.3-1 & Confirmatory Items 4.2-3 & 4.4-1 Re Reactor Sys & Mechanical Engineering Branches Outstanding Items,To Support Accelerated Advanced BWR Review Schedule

ML20034F620
Person / Time
Site:	05200001
Issue date:	02/16/1993
From:	Fox J GENERAL ELECTRIC CO.
To:	Poslusny C Office of Nuclear Reactor Regulation
References
NUDOCS 9303040057
Download: ML20034F620 (22)

Text

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February 16,1993 Docket No. STN 52-001

{

i Chet Posiusny, Senior Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of the Nuclear Reactor Regulation

Subject:

Submittal Supporting Accelerated ABWR Review Schedule - Reactor Systems and Mechanical Engineering Branches Outstanding Items

Dear Chet:

Enclosed are SSAR markups addressing: Open Items 53.2-1 and 153-1; and Confirmatory Items 4.2-3 and 4.4-1.

It should be noted that in addressing Open Item 153-1, it was necessary to markup proprietary Page 4B-7. This markup is an addition to an earlier submittal and the proprietary affidavit under which it was originally issued is applicable.

Please provide copies of this transmittal to George Thomas and Dave Terao.

Sincerely, j

l Jack Fox Advanced Reactor Programs cc: Hal Careway (GE)

Norman Fletcher (DOE)

Bob Huang (GE)

Caroline Smith (GE) i 4

2 Ji93-31 9303040057 930216 1

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Rev.c Standard Plant 4.2 FUEL SYSTEM DESIGN 4.2.1 Discussion The fuel to be loaded in an ABWR is any fuel design approved by the USNRC or that meets the cri:eria documented in Appendix 4B. Using these designs will assure that all fuel system design requirements are met.

To demonstrate ABWR system response in this SSAR, a reference core of BP8x8R fuel is used.

This core is shown in Section 4.3; information for this fuel design is provided in Reference 1.

Each utility referencing the ABWR design may have different fuel and core designs which will be 4.2.3 References provided by the COL applicant to the USNRC for information. See Subsection 4.2.2.1 for COL 1.

GE fue! Bundle Designs, NEDE-31152P.

license information.

GE ControlRod Designs, (To be issued The control rods perform the dual function of power shaping and reactivity control.

A discussion of the rod control system components is presented in Section 4.6.

The control rod design to be used in an ABWR is any design approved or that_ meets the criteria documented in AppendijyLC/To demonstrate the pA

[ABWR system response in this report, a control #

rod design of sheathed cruciform array of stainless steel tubes filled with boron. carbide was used. This design is documented in Reference L 2 and shown in Figure 4.21/The control blade design to be used at the plant will be provided by the COL applicant to the USNRC for information. See Subsection 4.2.2.2. for COL license information.

4.2.2 COL License Information 4.2.2.1 Fuel Design The fuel bundle name. and a reference to documentation of the fuel design will be prosided by the utility referencing the ABWR design to the l USNRC for information. (See Subsection 4.2.1).

4.2.2.2 Control Blade Design The control blade model and referenc-to documentation of the control blade design will be provided by the utility referencing the ABWR design to the USNRC for information. (See Subsection 4.2.1).

4.2-1 Amendment 23

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i The reference ABWR control rod dcsign consists of a sheathed cruciform i

array of stainless steel tubes filled with boron carbide (B C) powder. Figure 4.2-1 is 4

an illustration of the reference design. The main structural members of the reference i

design are made of stainless steel and consist of a top handle, a lower transition piece with a control rod drive coupling, a vertical cruciform center post, and four U-shaped absorber tube sheaths. The top handle, lower transition piece and center post are welded into a single skeletal structure. The U-shaped sheaths are welded to the center post, handle, and lower transition piece to form the housing for the absorber rods filled with B C. Rollers at the top and bottom of the control rod guide the control rod 4

as it is inserted and withdrawn from the core. The B C powder in the absorber tubes 4

is compacted to approximately 70 percent ofits theoretical density. The B C is scaled 4

into the absorber tubes by plugs welded into each end, and is longitudinally separate.1 into individual compartments by stainless steel balls. Typical parameters of the 1

reference ABWR control rod design are provided in Table 4.2-1.

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Control Rod Weight (Ib) 180 Absorber Rod - B C 4

Number per control rod 72 Length (in.)

143 Diameter (in.)

.166 Density (% Theoretical) 70 Abssorber Tube - B C 4

Cladding material 304SS Outside diameter (in.)

.220 Wall thickness (in.)

.027 T

Sheath Thickness (in.)

.045 Pin Material PH13-8Mf Roller Material INCONEL x-750 i

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%.o overall plant control c 2 ara teristics, core thermal power limits, etc.A power flow maps 4.4.2 Description of theThermal and for the power range of operation shown L rb wan Hydraulle Design of the Reactor Coolant 4.4 ' r:: used in the system response analyses documented in Section 6.3 and Chapter 15. The System specific power-flow operating map for each plant will be provided to the USNRC for information.

See Subsections 4.4.4.1 and 4.4.4.2 for COL 4.4.2.1 Plant Configuration Data license information. The nuclear system equipment, nuclear instrumentation, and the 4.4.2.1.1 Reactor Coolant System Configuration reactor protection system, in conjunction with operating procedures, maintain operations within The reactor coolant system is described in the area of the operating map for normal Section 5.4 operating conditions. The boundaries on this map are as follows:

4.4.2.1.2 Reactor Coolant System Thermal Natural Circulation Line,0: The operating Hydraulle Data state of the reactor moves along this line The steady-state distribution of temperature, for the normal control rod withdrawal pressure and flow rate for each flow path in the sequence in the absence of recirculation reactor coolant system is shown in Figure 5.1-1.

pump operation.

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4.4.2.13 Reactor Coolant System Geometric Data 102% Power Rod Line or Rated Power l

(Whichever Is Less): The 102% power rod Volumes of regions and components within the line passes through 102% power at 100%

reactor vessel are shown in Figure 5.12.

flow. Some plants may establish an operating map in which 102% power can be Table 4.4-5 provides the flow path length, achieved at lower flow. The operating state height, liquid level, minimum elevations, and for the reactor follows this rod line (or minimum flow areas for each major flow path similar ones) during recirculation flow volume within the reactor vessel and changes with a fixed control rod pattern; recirculation loops of the reactor coolant however, rated power may not be exceeded.

systems.

Steam Separator Limit Line: This line 4.4.2.2 Operating Restrictions on Pumps resuits from the requirements to have acceptable moisture carryover fraction from Expected recirculation pump performance steam separator.

curves are shown in Figure 5.4-3. These curves are valid for all conditions with a normal 4.4.23.2 Other Performance Characteristics operation range varying from approximately 20 percent to 115 percent of rated pump flow.

Other performance characteristics shown on the power-flow operating map are:

Subsection 4.4.2.3 gives the operating limits imposed on the recirculation pumps by cavitation, Constant Rod Lines A, B, C, D E, F: These pump load, bearing design flow starvation, pump lines show the change in flow associated speed, and steam separator performance.

with power change while maintaining constant N

controf rod position.

4.4.23 Power Flow Operating Map Constant Pump Speed Lines 1,2,3,4,5,6, 4.4.23.1 Umits for Normal Operation 7, 8: These lines show the change in flow associated with power changes while A BWR must operate with certain restrictions maintaining RIP speeds at a constant speed.

because of pump net positive suction head (NPSH),

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u s6to u s ABWR PEV C Standard Plant 4.423.5 Flow Control 4.4233 Regions of the Power Flow Map Region I This region defines the system The normal plant startup procedure requires operational capability with the the startup of all RIPS first and maintain at reactor internal pumps running at their minimum pump speed (30% of rated), at their minimum speed (30%). Power which point reactor heatup and pressurization changes, during normal startup and can commence. When operating pressure has been shutdown, will be in this region. The established. reactor power can be increased.

normal operating procedure is to start This powen-flow increase will follow a line within Region I of the flow control map shown in up along curve 1.

Figure 4.41. The system is then brought to the Region 11 This is the low power area of the desired power-flow level within the normal oper-ating area of the map (Region IV) by increasing operating map where the carryover the RIP speeds and by withdrawing control rods.

through steam separators is expected to exceed the acceptable value.

Operation within this region is Control rod withdrawal with constant pump precluded by system interlocks.

speed will result in power / flow changes along lines of constant pump speed (Curves 1 through Region III This is the high power / low flow area 8). Change of pump speeds with constant control of the operating map which the system rod position will result in power /Dow changes is the least damped. Operation within along, or nearly parallel to, the rated flow this region is precluded by SCRRI control line (curves A through F).

(Selected Control Rods Run-In).

Region IV This represents the normal operating zone of the map where power changes can be made, by either control rod 4

movement or by core flow changes, through the change of the pump speeds.

4.423.4 Design Features for Power Row Control 4.42.4 Thermal and Ilydraulic Characteristics The following limits and design features are Summary Table employed to maintain power-flow conditions shown The thermal-hydraulic characteristics are

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in Figure 4.4-1:

prosided in Table 4.4-1 for the core and tables (1) Minimum Power Limits at Intermediate and of Section 5.4 for other portions of the reactor High Core Flows: To prevent unacceptable coolant system.

separator performance, the recirculation system is provided with an interlock to N

reduce the RIP speed.

A N E AT 7.

(2) Pump Minimum Speed Limit: The Reactor Internal Pumps (RIPS) are equipped with enes s)

Anti-Rotation Devices (ARD) which prevent a tripped RIP from rotating backwards. The ARD begins operating at 300 rpm decreasing speed. In order to prevent mechanical wear in the ARD, minimum speed is specified at 300 rpm. However, to provide a stable operation, the minimum pump speed is set at 450 rpm (30% of required).

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4. 4. 2 5 Th e ma t t4d%k e S4*b k 94do m uce As discussed in the Response to Question 100.1, the ABWR design assures the stability performance in the normal operating region is more stable than current operating BWRs by incorporating the following design features:

- (1)

Smaller inlet orifices, which increase the' inlet single-phase-pressure drop, and, consequently, improve the core and channel stability.

l (2)

Wider control rod pitch, which increases flow area, and, consequently, I

reduces the void reactivity coefficient and improves both core and channel stability, and i

(3)

More steam separators, which reduce the two-phase pressure drop, and improve the stability.

In order to reconfirm this conclusion, a stability analysis based on the procedures developed by the BWROG committee on thermal hydraulic stability (Reference 1) was performed for the ABWR. In this analysis, a conservative nuclear conditions, taking into consideration of future core design, were assumed. The results at the most limiting conditions in the normal operating region (i.e.; the intercept of 102 % rod -line with all i

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operating RIPS at their minimum speeds, assuming only 9 out of 10 RIPS are in operation) are as follows.

Core Decay Ratio 0.72, l

Channel Decay ratio 0.36.

4.4 - A These results are also shown in Figure / together with the criteria. From Figure 1, it is confirmed that that ABWR is stable in the normal operating region.

(6-4 Furthermore, automatic logics (Figure [) which prevent plsnt operation in the region with the least stability margin is also implemented. This design is similar to Option I-A, one of long-term solutions considered by the BWROG.

In addition, in order to meet the stability design requirements specified in the i

ALWR Utility Requirements Document, Option III, LPRM based Oscillation i

Power Range Monitor (OPRM), which is also one of long-term solutions i

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considered by the BWROG, will be implemented in the ABWR design, when the OPRM design is approved by the NRC.

As for issues relates to ATWS stability, they are of no concerns to the ABWR design, since the ABWR design has logic to automatically initiate the SLCS, including automatic initia: ion of feedwater run back. Furthermore, the i

ABWR EPG will incorporate any changes recommended by the BWROG.

In summary, the ABWR stability design is consistent with the licensing i

methodology proposed by the BWROG committee on thermal hydraulic stability. The ABWR will be stable in the nonnal operating region.

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conditioning equipment, alarming monitor, signal analysis and data acquisition equipment, and j

calibration equipment.

t 4.4.4 COL License Information t

4.4.4.1 Power Flow Operating Map i

The specific power flow operating map to be i

used at the plant will be provided by the COL i

applicant to the USNRC for information. (See.

Subsection 4.4.2.3.1).

4.4.4.2 hermal Ilmits e

The thermal limits for the core loading at the plant will be provided by the COL applicant to the USNRC for information. (See Subsection l:

4.4.2.3.1).

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2. Fbw s 36%: To assure fbw rate is higher than that of eight RIPS operations i

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ev. c 5.3.2 Pmssure/Temperatum Limits Heatup and Cooldown.

53.2.1 Umit Curves Curve B in Figure 5.3-1 specifies limits for non-nuclear heatup and cooldown following a The pressure / temperature limit curves in nuclear shutdown.

Figure 5.31 are based on the requirements of 10CFR50, Appendix G. The prressure/ temperature Reactor Operation limits look different than SRP Section 5.3.2 because the ABWR temperature limits are based on a Curve C in Figure 5.31 specifies limits appli-more recent revision of Regulatory Guide 1.99.

cable for operation whenever the core is critical except for low-level physics tests.

All the vessel shell and head areas remote from discontinuities plus the feedwater nozzles were 53.2.1.4 ReactorVessel Annealing evaluated, and the operating limit curves are based on the limiting location. The boltup limits Inplace annealing of the reactor vessel, for the flange and adjacent shell region are based because of radiation embrittlement, is not an-on a minimum metal temperature of RT ticipated to be necessary.

33 C. The maximum throughw.!! temperahe.g. plus gradi-ent from continuous heating or cooling at 55.5"C 53.2.1.5 Predicted Shift in RT and per hour was considered. The safety factors Drop in Upper-Shelf Energy (Ipp[ndix G-IV B) applied were as specified in ASME Code, Appendix G, and Reference 2.

For design purposes the adjusted reference nil ductility temperature and drop in the upper-shelf The material for the vessel will be provided er.ergy for BWR vessels is predicted using the pro-cedures in Regulatory Guide 1.99, with the following requirements of RT as determined in accordance with Branch "INcSTnical Position MTEB 5-2: shell and flanges -229 C; The calculations (see response to Question nozzles 29 C and welds - 29 C.

251.5) are based on the specified limits on (

Phosphorous (0.020%), Vanadium (0.05%), Copper See Subsection 5.3.4.3 for COL license (0.08%) and Nickel (1.2%) in the weld material.

information.

In plate material, the limits are Copper (0.05%)

and Nickel (0.73%). Forgings will have the same 53.2.1.1 Temperature Umits for Boltup chemitry as plate but the nickel limit is 1%.

Minimum closure flang'e and fastener tem-The ABWR neutron fluences are low when cornparedD peratures of RT 33 C are required for with the past reactors because of the fact that tensioning at pre $ad condition and during the incorporation of internal pumps increased the degensioning Thus, the limit is -29 C +

annulus between the shourd and the vessel wall.

L 33 C = + 4 C.

mp t A A surveillance program in accordance with ASTM 53.2.1.2 Temperature Umits for ISI Ilydro-E 185 will be used. The surveillance program will static and Irak Pressure Tests include samples of base metal, weld metal and heat affected zone material. Subsection 5.3.1.6 Pressure (measured in the top head) versus tem-provides added detail on the surveillance program.

perature (minimum vessel shell and head metal tem-perature) limits to be observed for the test and 53.2.2 Operating Procedures operating conditions are specified in Figure 5.3 - 1. Temperature limits for preservice and A comparison of the pressure versus temp-inservice tests are shown in Curve A of Figure crature limit in Subsection 5.3.2.1 with in-5.3 1.

tended normal operation procedures of the most severe service level B transient shows that 5.3.2.13 Operating Umits During fleatup, those limits will not be exceeded during any Cooldown, and Core Operation foreseeable upset condition. Reactor operating procedures have been established so that actual transients will not be more severe than those 5.M Amendment 23

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i N ssit r A The evaluation of fast neutron fluence for the ABWR vessel was done using the Oak Ridge National Lab code DOT-4 on a CRAY X-MP Super Computer using an eighth core symmetry fixed source model. The neutron source was based upon a three dimensional nodal fuel model of ABWR for an integrated equilibrium core with a 26 group neutron spectrum. The results shown below are reasonable in comparison to the BWR/6 calculations which were performed with an older version of DOT. In this comparison, the BWR/6 40 year quarter thickness evaluations for the 218-624 plant were compared to the 3

8 plant and the 40 year ABWR values which are shown on line three of tiIc..._.,g;.or In evaluating the relative fluence, the power level and sh 40 year B\\g" vessel water thickness were taken into account. In the case of the gage,rjhickness, the neutron reduction factor was interpolated from the r*tachd figurepluch shows the calculated fast neutron flux for an annular region as a function ofwater thickness. -rke omh.-

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i BWR/6 ABWR 218-624 238-748 Peak Fluence (40yr) 5.5E+18 4.3E+18 2.2E+17 (1/4 t)

Power (MWt) 2894 3579 3926 Bundles 624 748 872 Power Lvl(kw/l) 52.8 54.5 51.3 VesselIR 276.86 302.26 353.06 Shroud OR (cm) 234.95 256.54 280.35 Water Gap (cm) 41.9 45.7 72.7 Neutron Reduction 0.007 0.0044 0.00042 l

Factor for Water Expected Fast 5.5E+18 3.6E+18 3.4E+17 Fluence based upon 218-624 e

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oI 153-t Geners! Doctric Ccepany ABM PROPRIETARYIhTORMATION nAsmAs Standard Plant cw m ne c i

vendor fuel. Mixed core effects are included in (2) Applicability of the bounding BPWS analysis the calculation.

must be confirmed if analysis is required.

When loading GE fuelinto a core with other 4B.13 REFUELING ACCIDEST vendor fuel, GE will use the nuclear libraries ANALYSIS calculated fro the other vendor fuel and the actual fuel bundle parameters of the other The consequences of refueling accidents are vendor fuelin a mired core analysis to confirmed as bounding or a new analysis shall be demonstrate the adequacy of Ihe plant performed and documented when a new fuel design is introduced.

overpressure system.

4B.11 LOSS-OF-COOIAhT ACCIDEST The consequences of the refueling accident are ANALYSIS METHODS primarily dependent upon the number of fuel rods in a bundle. When the number of fuel rods changes, (1)The criteria in 10CFR50.46 is met by either the effect on the refueling accident must be plant-specific or bounding analyses.

determined based on the fuel design information report.

The criteria is currently met by plant exposure

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dependent, bundle / lattice specific MAPLHGR 4B.14 AhTICIPATED TRANSIENT values which must be met during plant operation.

WITHOUT SCRAM In the future, other criteria or bounding analyses may be approved by the NRC.

The fuel must meet either criteria (1) or (2) below:

(2) MAPLHGR adjustment factors are utilized for each design if required for operating flexibility (1) A negative core moderator void reactivity options.

coefficient, consistent with the analyred range of void coefficients provided in Appendix 15E, shall Plaut MAPLHGR adjustment factors for be maintained for any operating conditions above operation in a configuration or region requiring the startup critical condition.

revised MAPLHGR values such as single recirculation loop operation must be confirmed (2) If criterion (1) above is not satisfied, the limiting for each new fuel design. This will be done for events (as described in Appendix 15E) will be each plant prior to the cycle of operation of the evaluated to demonstrate that the plant response new fuel design in that plant, in within the ATWS criteria specified in Appendix 15E.

(3) When GE fuelis loaded into a core with other vendor fuel, and, when no new system analysis is 4B.1S UNACCEPTABLE RESULTS FOR performed, MAPLHGR values for other vendor INFREQUENTINCIDENTS fuel will be those calculated by that vendor.

(UNEXPECTED OPERATION OCCURRENCES)

The LOCA analysis used to calculate fuel MAPLHGR values is independent of the other The following are considered to be unacceptable fuelin the core. Therefore, insertion of GE fuel safety results for infrequent incidents (unexpected into a core with other vendor fuel will not impact operational occurrences):

the results of the LOCA analyses previously performed for the other vendor fuelif no new (1) release of radioactivity which results in dose system analysis is performed.

consequences that exceed small fraction (10 percent) of 10CFR100.

4B.12 ROD DROP ACCIDENT ANALYSIS (2) failure of fuel cladding which could cause changes in core geometry such that core cooling (1) Plant cycle specific analysis results,if analysisis would be inhibited.

required, shall not exceed the licensing limit.

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worst possible location, and the plant is oper-drawn from the reload core analysis as previous-ated with the mislocated bundle. This event is ly presented is applicable to the / SWR initial categorized as a limiting fault based on the core. Hence, no specific analysit me,tred.

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following data:

15.4.7.4 Barrier Performance Expected Frequency: 0.002 events / operating An evaluation of the barrier performance is cycle.

ne made for this event, because it is a mild This number is based upon past experience.

and highly localized event. No perceptible change in the core pressure is observed.

15.4.7.2 Sequence of Events and Systems 15.4.7.5 RadiologicalConsequences Operation 15.4.7.2.1 Sequence of Events An evaluation of the radiological conse-quences is not required for this event, because The postulated sequence of events for the no radioactive material is released for the r

misplaced bundle accident (MBA) is presented in fuel.

Table 15.4-6.

15.4.8 Rod Ejection Accident 15.4.7.2.2 Systems Operatiou I

15.4.8.1 Identification of Causes and Frequency A fuel load,ing error, undetected by in-core Classificatico instrumentation following fueling operations, may result in an undetected reduction in thermal The rod ejection accident is caused by a margin during power operations. For the analysis major break on the FMCRD housing outer tube or reported herein, no credit for detection is taken associated CRD pipe lines. Due to a break of and, therefore, no correctise operator action or this type, the reactor pressure exerted on the automatic protection system functioning is CRD spud pushes down the hollow piston and the ballout with a large force. The shaft screw and I

assumed to occur.

the motor are forced to unwind. A passive brake 15.4.7.3 Core and System Performance mechanism is installed in the FMCRD system to prevent the control rod from moving. The design This event is presented in Subsection S. E4 of the brake is presented in Section 4.6.1. The of Reference 1.

probability of the initial causes, i.e., a CRD pipe line break or housing break, is considered 3

Mislocated bundle analyses are not performed low enough to warrant its being categorized as a' for reloau cores because, based on analysis of limiting fault. Even if this accident does data asailable from past reloads, the probability happen, the brake prevents the control rod from that a mislocated fuel bundle loading error will ejection. Should the brake fail, the check result in a CPR less than the safety limit is valve will serve as a backup brake to prevent sufficiently small (see Reference S.2 58 of the rod ejection.

Reference 1).

15.4.8.2 Sequence of Events and Systems For ABWR initial core, the mismatch of Operation esposures and integrated bundle power between mislocated bundles are less severe than the if a major break occurs on the FMCRD housing, equilibrium cycle. Therefore, the consequence of the reactor pressure will provide forces that a postulated misplaced bundle accident for the could cause the shaft sciew to unwind. The initial core is less severe th n that for the FMCRD brake mechanism prevents the rod from equilibrium cycle. Consequ:. ?ly, the conclusion moving. Therefore, no rod ejection can occur.

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sva REV B Standard Plant control rod from the hollow piston of the FMCRD.

If the control rod is stuck, the separation-de-tection devices will detect the separation of the

] control rod and hollow piston from the ba!! nut of the FMCRD and rod block interlock will prevent further rod withdrawal. The operator will be alarmed for this separation.

There is no basis for the control rod drop l event to occur.

15.43.32 Identification of Operator Actions f

No operator actions are required to preclude this event. However, the operator will be l

notified by the separation-detection alarm if separation is detected.

15.4.9.4 Core and System Performance 15 4 10 COLtsch

• T O ~

'*^

The performance of the separation-detection devices and the rod block interlocks virtually N SOV

• 4*g\\#^

IO*I bt preclude the cause of a rod drop accident.

l 15.43.5 Barrier Performance Evd A^oks t,f COL. a he

.r J fem An evaluation of the barrier perforrnance is

,g, 4g

%p a not made for this accident since there is no 4

circumstance Dr which this event could occur.

Ak C M Mss end 8*

v.,M 15.4.9.6 Radiological Consequences

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m e

L' W. N F-c.. ( se m.b ds The radiological analysis is not required.

U L 5 4.7. '3 ).

1I 15.4J4 References 1

General Electric Standard Application for Reactor Fuel--United States Supplement, j

NEDE-24011-P-A-US, (Latest approved revi-tion).

2.

C. J. Paone and J. A. Woolley, Rod Drop Accident Analysis for Large Boiling Water Reactors, Licensing Topical Report, March 1972 (NEDO-10527, Supplements I and 2).

15 4-10 Amendmerit 2