Text

Forwards Rev Pages for Sys 80+ Design Control Document. Margin Bar Adjacent to Change & Rev Date Will Be Added to Revised DCD Page
	ML20081D252
Person / Time
Site:	05200002
Issue date:	03/17/1995
From:	Brinkman C; ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY, ASEA BROWN BOVERI, INC.
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	LD-95-011, LD-95-11, NUDOCS 9503200178
	Download: ML20081D252 (61)
	v • d • e

r j ABB !

March 17,1995 LD-95411 Docket 52-002 ,

Attn: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001

Subject:

System 80+* Design Control Document Revisions

Dear Sirs:

This letter forwards revision pages for the System 80+ Design Control Document. These revisions are provided to resolve inconsistencies in the DCD noted by the NRC and faxed to C-E on March 8,1995, plus other inconsistencies identified by ABB-CE in subsequent reviews. A margin bar adjacent to each change and a revision date "(2/95)" will be added to each revised DCD page.

Please call me or Mr. Stan Ritterbusch at 203-285-5206 ifyou have any questions.

Very truly yours, COMBUSTION ENGINEERING, INC.

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C. B. Brinkman Director Nuclear Systems Licensing CBB/vap cc: S. Magruder (NRC)

N. Fletcher w/o enclosure (DOE) 200050 ABB Combustion Engineering Nuclear Power pf),

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, o System 80+ Design ControlDocument Table 2.5.3-1 Discrete Indication and Alarm System and Data Processing System (Continued)

Design Commitment Inspections, Tests, Analyses Acceptance Criteria 7.e) Specified parameters that 7. (Continued) 7.e) Specified parameters that provide information to provide information to indicate whether plant indicate whether plant safety functions are being safety functions are being accomplished during and accomplished during and following design basis following design basis 1 accident events. accident events.

( 7.f) Indication of bypassed and inoperable status of 7.f) Indication of bypassed plant safety systems, as and inoperable status of follows: plant safety systems, as follows:

i. Status of plant operating mode related bypasses of i. Status of plant operating the PPS. mode related bypasses of the PPS.

ii. Bypass status of each channel of the PPS. ii. Bypass status of each channel of the PPS.

iii. Bypass and inoperable status of engineered iii. Bypass and inoperable safety feature systems. status of engineered safety feature systems.

7.g) The status of core cooling prior to and 7.g) The status of core following an accident, as cooling prior to and follows: following an accident, as follows:

i. Eubcooling.

i. Subcooling.

ii. Liquid inventory in the reactor vessel above the ii. Liquid inventory in the fuel alignment plate. reactor vessel above the fuel alignment plate.

iii. Coolant temperature at the core exit. iii. Coolant temperature at the core crit.

7.h) Four channels of PPS l status information. 7.h) Four channels of PPS l status informaticn.

j', i) Four channels of status #

and parameter Four channels of status information from the fi) and parameter i I

ESF-CCS. information from the ESF-CCS.

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cuened Design unteriet (.gjcy5) page 2.5 ss i i

System 80+ Design ControlDocument i 1.8 Regulatory Cornpliance, Industry Codes and Standards System 80+ compliance with U.S. NRC Regulatory Guides, Generic Letters, Bulletins, and elements of the Standard Revier/ Plan is documented in this section. Regulatory Guides, the guide date or revision, and System 80+ compliance with applicable Guides are summarized in Table 1.8-1.

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Operational experience information highlighted in Regulatory Bulletins and Letters has been incorporated into the System 80+ design. Generic Letters and NRC Bulletins from 1980 through December,1993 are identified in Tables 1.8-2 and 1.8-3. The applicability of each Generic Letter or Bulletin to System 80+ is assessed, with additional information for applicable issues provided in the referenced sections of this report.

System 80+ deviations from the U.S. NRC Standard Review Plan, NUREG-0800 [ LWR Edition, June 1987], are listed in Table 1.84. SpecHic sections are also identified where further details relevant to each SRP deviation are discussed. Site-specific compliance with individual Standard Review P! m sections is provided in Table 1.8-5.

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Table 1.8-6 identifies the industrial Codes and Standards, and code editions, oked for certification of the System 80 + Standard Design. Where a particular structure, system, component requires a code edition different from that listed in Table 1.8-6, an explanation of su difference is provided in the appropriate text. Other Codes and Standards that are utilized but r , invoked as essential for design certification are incorporated into the individual chapters of thisGafily MLi R3 Revisions to l accepted industry codes applied to System 80+ will be evaluated on a case-by-case basis. ((The

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'V applicability of code editions will be confirmed by the Combined O r (Opiratin]bLicense applicant in the site- l specific Safety Analysis Report.))

Q ASME Section III, Division 1 Code Cases applicable to System 80+ are identified in Table 1.8-7.

Except for N-122-1, these Code Cases are consistent with those identified in Regulatory Guide 1.84, Revision 29, for design and fabrication, or Regulatny Guide 1.85, Revision 29, for materials and testing, that were in effect on July 31,1993. ((Later code cases that may be used for System 80+ piping and piping support design will be provided with the plant-specific information.))

Cross-references to subsections of this report discussing Unresolved and Generic Safety Issues, the Three Mile Island Rule [10 CFR 50.34 (f)], and new NRC policy issues (SECY-93-087) are provided in Tatles 1.8-8,1.8-9, and 1.8-10.

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l System 80+ Desian controlDocument

,9 U Table 1.8-1 NRC Regulatory Guide Applicability Analysis to System 80+ (Cont'd.) l Document Title Date Section RG 1.21 - Measuring, Evaluating, and Reporting Radioactivity in, and Rev.I 11.1; 11.5.1.1; Releases from Nuclear Power Plants 6/74 11.5.2.1 (COL applicant)

RG 1.22 - Periodic Testing of Protr' ;on Systems Actuation Functions 2/72 7.1.2.17; 8.1.4.2 RG 1.23 - Onsite Meteorological Progiams Not Applicable RG 1.24 - Assumptions Used for Evaluating the Potential Radiological 3/72 15.7 Consequences of a Pressurized Water Reactor Radioactive Gas Storage Tank Failure RG 1.25 - Assumptions Used for Evaluating the Potential Radiological 3/72 15.7 Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors RG 1.26 - Quality Group Clas'sifications and Standards Rev. 3 3.2.2; 10.4 2/76 RG 1.27 - Ultimate Heat Sink Rev. 2 9.2.5 y 1/76 m/

RG 1.28 - Quality Assurance Program Requirements 139, 19 Rev. 3 17 and)Spo

( (Design and Construction) 8/85 CENkD-210[;A, b

Rev. 7 (Section III.2.1)

RG 1.29 - Seismic Design Classification Rev. 3 3.2.1; 7.1.2.18; 9/78 10.4.9 RG 1.30 - Quality Assurance Requirements for the Installation, 8/72 RG 1.28 Rev. 3 Inspec*. ion and Testing of Instrumentation and Electrical Equipment applied instead of RG 1.30 RG 1.31 - Control of Ferrite Content in Stainless Steel Weld Metal Rev. 3 5.2.3.4.2.1 4/78 RG 1.32 - Criteria for Safety-Related Electric Systems for Nuclear Rev. 2 8.1.4.2 Power Plants 2/77 RG 1.33 - Quality Assurance Program Requirements (Operation) Rev. 2 Not Applicable 2/78 RG 1.34 - Control of Electroslag Weld Properties 12/72 5.2.3.3.2.2 RG 1.35 - Inservice Inspection of Ungrouted Tendons in Prestressed Rev. 3 Not Applicable Concrete Containment 7/90 (Concrete containment)

RG 1.36 - Nonmetallic Thermal Insulation for Austenitic Stainless Steel 2/73 5.2.3.2.3; 10.3.2.3.4

[O U l RG 1.37 - Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Components of Water-Cooled Nuclear Power 3/73 5.2.3.4.1.2.1; 10.3.6.2 (Note A)

Plants Asywored Design Maten'ai krtadw tion Page 1.3-3

e. O System 80+ Design ControlDocument Chapter 3 Figures (Cont'd.) h Page 3.7-6 Synthetic Time History H2 Spectra vs. Target Spectra for CMS 2 (1 and 2% Damping) ......... ...... ..... . ... ..... .. . 3.7-44 3.7-7 Synthetic Time Mistory H2 Spectra vs. Target Spectra for CM S2 (5 and 7 % Damping) . . . . . . . . . . . . . . . . . . . . . . . . . ....... 3.7-45 3.7-8 Synthetic Vertical Time History Spectra vs. Target Spectra for CMS 2 ( I and 2 % Damping) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-46 3.7-9 Synthetic Vertical Time History Spectra vs. Target Spectra for CM S2 (5 and 7 % Damping) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-47 3.7-10 Synthetic Time History HI Spectra vs. Target Spectra for CM S3 ( 1, 2, 5 and 7 % Damping) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-48 3.7-11 Synthetie Time History H2 Spectra vs. Target Spectra for CM S3 ( 1. 2. 5 and 7 % Damping) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-49 3.7-12 Synthetie Vertical Time History Spectra vs. Target Spectra for CMS 3 (1,2. 5 and 7% Damping) .......... ................ 3.7-50 1 3.7-13 Schematie Diagram of Interior Structure, Shield Building, l FB,CVCS ............................................ 3.7-51 3.7-14 Schematie Diagram ofInterior Structure, Shield Building, .

D G . I . DG 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-52 )

3.7-15 Schematie Diagram of Interior Structure, Shield Building, i EFWI (Horizontal), EFW2 (Horizontal) . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-53 :

3.7-16 Selich.e6 D;ep m ofInterior Structure, Shield Building, l EFW I (Vertical). EFW2 (Vertical) . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . 3.7-54 3.7-17 Schematie Diagram of Interior Structure, Shield Building, CAA, CAB . . . . . . 3.7-55 3.7-18 Finite Element Model of Steel Contaimnent Vessel . . . . . . . . . . . . . . . . . . 3.7-56 3.7-19 Schematie of Combined NI Structures (Elevation Looking South) . . . . . . . . . 3.7-57 3.7-20 Schematie of Combined NI Structures (Elevation Looking West) . . . . . . . . . . 3.7-58 3.7-21 Schematie Diagram of the SASSI Analysis Process Using CMS 2 and CMS 3 Motions . . . .......................................... 3.7-59 3.7-22 Schematie Diagram of the SASSI Analysis Process Using CMSI Motions . . . . 3.7-60 3.7-23 Reactor Coolant System Seismic Analysis Model .................... 3.7-61 3.7-24 Pressurizer Seismic Analysis Model ............................ 3.7-62 3.7-25 Typical Surge Line Seismic Analysis Model ....................... 3.7-63 3.7-26 Reactor internals Horizontal Seismic Analysis Model . . . . . . . . . . . . . . . . . 3.7-64 3.7-27 Reactor internals Nonlinear Horizontal Seismic Model . . . . . . . . . . . . . . . . 3.7-65 3.7-28 Core Seismie Model; One Row of 17 Fuel Assemblies . ............... 3.7-66 3.7-29 Reactor Internals Linear Vertical Seismic Model . . . . . . . . . . . . . . . . . . . . 3.7-67 3.7-30 Reactor Internals Nonlinear Vertical Seismic Model . . . . . . . . . . . . . . . . . . 3.7-68 3.7-31 Core-Support Barrel Upper Flange Finite-Element Model . . . . . . . . . . . . . . 3.7-69 3.7-32 Damping Value for Seismic Analysis of Piping . . . . . . . . . . . . . . . . . . . . . 3.7-70 3.7-33 Proportional Damping . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.7-71 3.8-1 Containment Details . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.8-44 3.8-2 Category I Structures - Typical Feedwater Penetration . . . . . . . . . . . . . . . . 3.8-47 3.8-3 Three-Dimensional ANSYS Containment Model . . . . . . . . . . . . . . ...... 3.8-54 .

3.8-4 Axisymmetric ANSYS Conta_inment Model ...................... 3.8-55 3.8-5 3.8-57 Nuclear Island Structures @tionQ . . . . . . . . . . . . . . . . . . . . . . .

Approved Design Meteriet. Design of SSC Page vi

4 D System 80+ Design ControlDocument

Response

The containment vessel is designed so that integrated leak r:te testing can be performed at design pressure after completion and installation of penetrations and equipment in accordance with the requirement of Appendix J of 10 CFR 50 (see Section 6.2.6).

3.3.46 Criterion 53 - Provisions for Containment Testing and Inspection The reactor containment shall be designed to permit (1) appropriate periodic inspection of all important areas, such as penetrations, (2) an appropriate surveillance program, and (3) periodic testing at containment design pressure of the leak-tightness of penetrations which have resilient seals and expansion bellows.

Response

The absence of insulation on the containment vessel permits periodic inspection of the exposed surfeces of the vessel. The lower portions of the containment vessel are totally encased in concrete and will not be accessible for inspection. It is contemplated that there will be no need for any special in-service surveillance program due to the rigorous design, fabrication, inspection and pressure testing the containment vessel receises prior to operation.

Provisions are made to permit periodic testing at containment design pressure of penetrations which have resilient seals or expansion bellows to allow leak-tightness to be demonstrated (refer to Section 6.2.6).

3.1.47 Criterion 54 - Piping Systems Penetrating Containment Piping systems penetrating primary reactor containment shall be provided with leak detection, isolation, and containment capabilities having redundancy, reliability, and performance capabilities which reflect ,

the importance to safety of isolating these piping systems. Such piping systems shall be designed with a capability to test periodically the operability of the isolation valves and associated apparatus and to determine if valve leakage is within acceptable limits. l 1

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Response

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Piping systems (deseribed c in CESSA)which penetrate containment are designed to provide the required isolation and testing capabilities. These piping systems are provided with test connections to allow j periodic leak detection tests to be performed, in accordance with 10 CFR 50, Appendix J. l The Engineered Safety Features Actuation System circuitry provides the means for testing isolation valve I operability.

For a discussion of penetration design, refer to Section 6.2.4, Containment Isolation System.

For additional related discussion, see the responses to General Design Criteria 55,56, and 57 (Sections 3.1.48 through 3.1.50).

O Appmved Design Materist . Design of SSC page .1.126 l l

System 80+ Design ControlDocwnent where:

N = total number of components, Sj = composite modal damping for mode j,

4

= critical modal damping associated with component i, 4j = mode shape vector,

{M i} = subregion of mass matrix associated with component i, and

[M] = the mass matrix of the system.

For direct integration method, viscous damping proportional to the mass and stiffness matrix is used; thus

[C] = a[ ] + #[g where [C] is e damping matrix, [K] is the stiffness matrix and [M] is the mass matrix. The values of a and # are selected such that the damping in the range of fr%2ency of interest is approximately equal to the damping of the structure.

((Where composite modal damping is used for piping, the input damping for piping elements is in accordance with Table 3.7-1. That is,for the Safe Shutdown Earthquake, the damping is 2.0 percent of critical dampingforpiping ofdiameter s 12 inches and is 3.0 percent of critical dampingforpiping of diameter > 12 inches.))'

3.7.3 Seismic Subsystem Analysis 3.7.3.1 Seismic Analysis Methods The seismic analysis of the Seismic Category I structures, subsystems, and components other than piping is performed by either the response spectrum or time history method as described in Section 3.7.2.1.1 or an equivalent static method described in Section 3.7.3.5.

When analyzed using the response spectrum method, four options are available for the choice of response spectra. These are described in Appendix 3.9A, Section 1.4.3.2.1.2. Appendir. 3.7D shows sample spectra for use in the three options not related to plant specific analysis.

For Seismic Category I piping, each piping system is ideali:.ed as a mathematical model consisting of i lumped masses connected by clastic members. The stiffness matrix for the piping subsystem is determined using the elastic properties of the pipe. This includes the effects of torsional, bending, shear, and axial deformations as well as changes in stiffness due to curved members. Generally, a response spectrum analysis is performed using the envelope of all applicable spectra to account for inertia effects.

The effects of rocking and torsion are implicitly included because the spectra at the support points include motions due to rocking and torsion. The total seismic response of the piping is then calculated 2

NRC Staff approval is required prior to implementing a change in this information; see DCD Introduction O

Section 3.5.

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,. j Svstem 80+ De CongolDocument i

Overview ;

This Appendix contains the results of the coupled Reactor Coolant System (RCS) seismic analyses. ,

e tf Table 3.7A-1 contains the RCS coupled frequencies and modes (refer to Figures 3.77tand 3.7-J3'for i directions and joint locations) with rigid supports at the building RCS support interfaces.

Table 3.7A-2 contains the seismic loads due to the envelope of all SSE soil cases.

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1 AmM Deewn Acerend- Design of SSC % 2.7A.1

System 80+ oesian controlDocument e All exposed areas after installation shall befieldpainted (or coated) in accordance with the applicable portion of Section M3 ofReference 17.

e The quality assurance requirementsforpainting (or coating) ofstructural steel shall be in accordance with Reference 18 as endorsed by Regulatory Guide 1.54, "Quulity Assurance Requirementsfor Protective Coatings Applied to Water Cooled Nuclear Power Plants".))2 Welding activities associated with Seismic Category I structural steel components and their connections shall be accomplished in accordance with written procedures and shall meet the requirements of AWS Dl.1 (Reference 25). The visual acceptance criteria shall be as defined in NCIG-01 (Reference 24).

3.8.4.5.3 Concrete and Steel Structures in addition to satisfying the load combinations for structural adequacy against the design loadings, the load combinations to ensure safety factors against overturning, sliding, and flotation are checked to ensure overall stability of Seismic Category I structures. The following events are checked as a minimum:

e The overturning about the toe of the foundation supported on soil.

e The foundation sliding on soil.

e Floating of the foundation base mat.

e ne containment vessel slipping in the lower concrete support dish.

e The containment vessel overturning about the edge of the lower concrete support dish.

e ne interior structure concrete slipping inside the containment vessel.

The safety factors which must be satisfied during any of these events are shown in Appendix 3.8A, Section 5.2.4. Safety factors which meet or exceed these criteria have been demonstrated in all analyses.

No increase in allowable stresses under service load conditions due to normal or severe load combinations is permitted due to wind loadings as identified in NUREG-0800, NRC Standard Review Plan, Section 3.8.4, Part 11.5.

Welding activities associated with the Holdup Volume Tank, In-Containment Refueling Water Storage Tank (IRWST), Emergency Feedwater Tank, Refueling Ccvity and Spent Fuel Pool liners shall be accomplished in accordance with the requirements of the American Welding Society (AWS) Structural j Welding Code, Dl.1 (Reference 25). The welded seams of the liner plates shall be spot radiographed !

where accessible, liquid penetrant and vacuum box examined after fabrication to ensure the liners do not leak. The acceptance criteria shall meet the acceptance criteria stated in Article NE-5200,Section III, Division of the ASME Code.

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NCR Staff approval is required prior to implementing a change in this information; ne DCD Introduction Section 3.5.

Aptwered Design Metenel Design of SSC Page 3.8-30

System 80+ Design ControlDocument n

11.1.2 Description (a)

There are two Diesel Fuel Storage Structures; one on each side of the N dear Island.

The main reinforced concrete structure is approximately 25 ft high,63 ft long and 44 ft wide founded on a 2' - 3" thick reinforced concrete mat located 12'-6" below the grade elevation of 90'-9". The walls and the roof are 2' - 3" thick. There is a two foot thick center reinforced concrete wall that divides the structure into two separate bays. Each bay encloses a diesel fuel oil tank, a tank vent, a sump with a sump pump, and necessary piping. The bays are separated from each other and from the equipment room by three-hour rated fire barriers (i.e.,2 ft thick walls). A steel platform at elevation 89'-3" surrounds each of the fuel tanks. The outside doors are protected against tornado missiles by a concrete missile barrier.

There is also an attached outside Seismic Category 11 equipment room that is approximately 10 ft high, 12 ft long and 28 ft wide founded on a 15" reinforced concrete mat. The equipment room is a steel framed structure with insulated metal siding and a metal deck roof.

The Diesel Fuel Storage Structure shall be located a minimum of 50 feet from any hydrogen storage area to preclude loading to the structure from a potential hydrogen burn.

11.1.3 Elevations e El. 78'-3" Pottom of base mat for the main structure e El. 91'-9" Top of base mat for the equipment room structure

El. 91'-9" Top of steel platform

El.103'-3" Top of roof 11.1.4 Codes and Standards The codes and standards applicable to Seismic Category I buildings shall be met for the Diesel Fuel Storage Structure including the equipment room.

II.I.S Loads in addition to the minimum design loads requirements of Section 5.1 of this appendix, the following additional specific load requirements shall 'oe met. Should conflicting values occur between this section and Section 5.1 of this appendix, the values specified in this section apply.

11.1.El Dead IAad (D)

The foundation slab shall be designed to include the reactions imparted by the steel fuel tank support frames. The weight of each tank and oil is approximately 402 kips. (The site specific SAR shall verify the tank volume is adequate for the diesel generators purchased, such that they meet their design criteria.) :

The tank support frame is not covered by this c'riteria and shall be designed in accordance with the rules 1 of Reference ASME Section 111, Division , Subsection NF.

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System 80+ Design ControlDocument

El. 91'-9" Top of the first floor (1 foot above grade)

El. 73'-9" Bottom of basemat 11.2.4 Codes and Standards The codes and standards applicable to Seismic Category I buildings shall be met.

11.2.5 Imads In addition to the minimum design loads requirements of Section 5.1 of this appendix, the following additional specific load requirements shall be met. Should conflicting values occur between this section and Section 5.1 of this appendix, the values specified in this section apply.

11.2.5.1 Dead Imad (D)

The weight of each heat exchanger when full of water is approximately 250 Kips excluding the heat exchanger saddle and leg supports. The heat exchanger support is not covered by this criteria and shall be designed in accordance with the rules of ASME Boiler and Pressure Vessel Code,Section III, Division j

[Il Subsection NF.

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11.2.5.2 IJve Load (L)

The CCW Heat Exchanger Stmeture shall be designed for the following live loads.

kJu Live Load

Fan and Air Inlet Room 150 psf

Roof 100 psf

First floor 150 psf

Basemat 250 psf 11.2.5.3 Temperature Loads (T,)

The normal concrete surface operating temperature within the building ranges from 60*F to 90'F. The ambient temperature range outside of the building shall be assumed to range from -10*F to 100*F (See Section 5.1.1.5 of this appendix). Site specific provisions may be taken to minimize the effects of the structural temperature gradient produced by these conditions.

11.2.5.4 Seismic Loads (E')

The seismic accelerations shall be as specified in the Table 3.8A-3.

11.2.5.5 Internal Flooding The stmeture sump is designed to collect water due to flooding resulting from a potential rupture of the CCW cr Station Service Water (SSW) piping.

9' Anmed Duign stan'at. Duign d ssC hee 3.8A-44

System 80+ Design ControlDocument 11.4 3 Elevations Service water pump structure elevations are site specific.

11.4.4 Codes and Standards The codes and standards applicable to Seismic Category I buildings shall be met.

11.4.5 leads in addition to the minimum design loads requirements of Section 5.1 of this appendix, the follow;ng specific additional load requirements shall be met. Should conflicting values occur between this section and Section 5.1 of this appendix, the values specified in this section apply.

11.4.5.1 Dead Imad (D)

The weight of each Station Service Water (SSW) pump is dependent on site specific considerations.

11.4.5.2 Live lead (L)

The SSW pump supports are designed for thrust loads per vendor drawings.

litm Live Load 3 Concrete Floors (by COL) psf

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Roof % (by COL) psf f

Operating Conditions: f Normal water level ; lEl. (Note 1)

Extreme low water level El. (Note 1)

Maximum water level'(flood) El. (Note 1)

N Note 1: These elevations will be established based on site specific data.

11.4.53 Temperature leads (T,)

The normal concrete surface operating temperature within the building is site specific. The ambient temperature range outside of the building shall be -10'F to + 100'F (Section 5.1.1.5 of this appendix).

Site specific provisions may be taken to minimize the effects of the structural temperature gradient produced by these conditions.

11.4.5.4 Seismic Loads (E')

The seismic response of the structure is site specific.

11.4.5.5 Flooding Flood loads on the Service Water Pump Structure shall include internal flooding and hurricane induced wave forces.

9 Approved Design Material. Design of SSC Page 3.8A-48

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System 80+ onkn contror Doesmaant Tame 7.2-5 Plant Protection Sy3 tem Failure Modes and Effects Analysis (Cont'd.)

No. Name Pasure Mede Cause Syungdoms and Leest Effects Meibed of Desecties Intweese Cearpementing Remiarts and Other Effects Effect LW PPS fortsding Depesideae Petures Freensees

23) Insedacet a) FtP Omarway l ims of hierenal Ims of CEAC, Bestehen, IfL aasi A -- - . susanate via ITP's in sedundans PPC None DPS performs argnet vandaime Ten Peas off power suggey, ITP smars or dem to DPS. taas of Tem peacessor PFC channels channels vesnais avadaMr. based on semainbg fedundens Pecessor faihwes (heedwe.e er PPC opereers Modele as R3P if seinsin self heale mams of Bestshie and LCL channelinformation IChanart A software) er dse control ersasferred frasa MCP. das " via Pmcessors courinae as Typeral) DIAS. DPS annuncisees via opersee based on tem niid links which hah beahh staans '. of

. dam input frone PPC esecueba. ' arians. Permdie operaer's seudule N im use sneveBlance and seming by at RSP PPC Maineenance opersoor, and erse panel remams avadahle.

b) ITP Oseeway.1 Psiluees Ginedware Erroneous das seceived by DPS, Enomeone status indration BistsNe and LCL RPS and ESFAS LCLs PosuMe spencess chanel snpa and fads on er softwest)caess RSP. PPC Operators Module observed, periedie processors caneusse es conversed so Icutef 2 IfL iniciadoe in channel A.

ervoersus esa receives erreaeous dem if in use. surveillance sad sewing by operne Also 3<hannet coincidence.

Ersoneses dem transmined en eperneos. Possibas sedendency(44 channel in Waalde sad tfL pecessors annonciationof bisisNe snps bypass).

and LCL responses to eneneous data via DtAS.

c) ITP Geewey-2 Same as 23 m) Isss of Ttf. beenMe,IfL and Same as 23 al BismMe and tfL RP3 and ESFA3 comudence DIAS and PCS perforse signal fa#s off TTP mams or dem to DFAS. Imse of processors cominue as logic senisms in 2eus.cf.3 validaten based on vining FPS dem to PCS. IAss of MCP operses based en les valid logic. sedundantchanneti _

PPC ereemors undele, en resynnee data input fmm PPC es opermoes actions for variable operneers niedule. 3-seapoints and bypasses. chsenel sedundancy(4th channelim bypess)enance and een penet seemins EsaiaMe.

d) ITP Genewey-2 same as 23 b) Enoneous data seceived by DIAS, Same as 23 b) Sane as 23 b) Same as 23 b) Same as 23 b)

PFC eperefors module and PCS seceive ermneous dain.

e) Ten Processor Same es 23 m) Autoenetic eums within chsenet Annunciating automatic by PPC channel A bietsNe Sasne as 23 c) Autoenetic tessing of sesnaining PPC fads off enabled. Inabilny es perfonn trip ITP's self lieshb sensus checke end LCL processors channelcan continue encerf i - 4' ,

channet bypasses frem MCP or R3P in redundant PPC channels ein connnue em operaer with 7ffecsed FrP sen)nenstica los!c, due operators modules. DIAS and DPS. Periodic test valid arip channel 8 so lack of aess permissavt,hrocessor.

ensineenance and wenust tests. bypass states earil changed 4

via PFC mainernance and nest panel. Alse 3<bannel

[e redundancy (4e channelin bypess).

Approveer Desko Aforarial- bee- _ ^ eien and s~eereal popef.ygg

?L m-

System 80+ Design ControlDocument i

l e Exhaust hoods are provided for 'each sample sink to ensure that leakage of any gases will be exhausted from the sample room.

l 4

Sample sinks are provided to collect all spillage. ;

The routing of high pressure and temperature sample lines outside the reactor containment is not considered hazardous because of the limited flow capacity.

9.3.2.4 Inspection and Testing Requirements 93.2.4.1 Inspection N s.,,,_ . s,/, k During the fabrication of the components and during the installation of the systems lass 2 corfponents and systems are examined to the requirements of ASME Section III, Article C-5000. x Class 4--s components and systems are examined to industry standards. After installation the system is examined for correct routing of piping, placement of hangers, insulation where utilized, and the sample vessels are removed and reinstalled to test the functioning of the disconnects.

9.3.2.4.2 Testing After installation is completed the system is hydrostatic and leak tested to the requirements of ASME Section III, Article NC-6000. Before system operation, valves are operated and observed to function properly, and cooling flow to and from the sample heat exchangers is observed to function. After the sampled system is pressurized, sample flow is observed to meet minimum requirements and instrumentation is observed to function. After the sampled system is pressunzed and heated, the adequacy of the sample heat exchangers to cool the sample flow is observed.

93.2.5 Instrumentation Requirements Local pressure, temperature, and flow indicators are provided to facilitate manual operation, and to verify sample conditions before samples are drawn. Temperature and pressure indication of sample streams downstream of each sample cooler are provided. Flow indication is provided for every sample line.

Alarms are provided, as appropriate, based upon inlet sample temperature, continuous parameter monitoring requirements, and the potential for transients which demand swift corrective action (s).

Radiation inonitors are provided for continuous monitoring of reactor coolant and steam generator blowdown samples.

A boronometer is provided for continuous monitoring of the reactor coolant boron concentration.

Continuous analyzers (as defined by Table 93.2-1) monitor specific water quality conditions in the secondary plant.

93.2.5.1 Process Radiation Monitor The process radiation monitor provides a continuous recording in the control room of reactor coolant gross gamma radiation and specific fission product gamma activity, thus providing a measure of fuel cladding integrity. A high alarm is annunciated in the control room. local and remote samples in the ANwned Design Atatorial Aum1ery Systems Pare 9.316

i Svstem 80+ Dessen conenar Docenent ;

i

~

. Table 9.3.4-1A Reactor Coolant Operating Ilmits Pre Core Het Initial Core Load Startup & Power Shutdown and i Analysis Phnctionalstij and Criticality Operation Refueling i pH (77'F) 3.8 - 10.4 4.5 - 10.5 4.5 - 10.5 3.8 - 10.5 .

Conductivity [ Note 2] [ Note 2] [ Note 2] [ Note 2] l Hydrazine 30 50 ppmI31 30-50 ppmI31 1.5 x Oxygen ppa'1 l -

(max. 20 ppm) i Ammonists j 0-50 ppm 0 50 ppm 0-2 ppm - 0 - 2 ppm l Dissolved Gas - -- [ Note 6) < 10 cc STP/kg(H2O) ;

Lithium 1-2 ppm 0.2-2.2 ppm 0.2-2.2 ppm -

Hydrogen -

[ Note 7] 15-50 cc (STP)/kg <5 cc STP/kg(H 2O)

(H 2O)tsj Ox3 gen s0.1 ppm $0.1 ppmI 9 $0.1 ppm -

Suspended Solids tsj <0.35 N <0.35 ,A < 0.35 -

2 ppm A ) 2 ppm 76*3 y 2 ppm )*6*(y) s0.15 ppm ~ s0.15 ppm' Chloride 50.15 ppm s 0.15 ppm-Fluoride so.15 ppm s0.15 ppm so.15 ppm s 0.15 ppm .

Boron s Refueling s Refueling s Refueling s Refueling l Concentration Concentration Concentration C w nion Sulfate tsj s 0.1 ppm s .05 ppm s .05 ppm s 0.1 ppm ;

Notes: ,

181 Special hot conditioning limits: Temperature > 350' for 7-10 days. i t

(2) Consistent with pH additive concentration.

ISI Hydrazme is maintamed at 30-50 ppm any time the RCS is less than 150*F.

I'l Prior to exceedir- 150*F during heatup or below 400*F during cooldown.

(5) This parameter is used for problem diagnosis. !

I'l Prior to a depressunzation shutdown, reduce total gas to < 10cc(STP)/kg (H2 O) to limit the possibility .

for explosive mixtures.

r71 During the transition from post-core to operating, hydrogen should be maatamed in the 15 to 22cc(STP)/kg ,

(H 2O) range to minimize degassmg requirements in case the reactor plant must be shutdown and depressunzad.

I'l Hydrogen should be ,5ec(STP)/kg (H 2O) before secunng the reactor coolant pumps.

IM Not applicable during core load.

{

[*] %e, aboewd Acac\dn 4 A3 6 b Z4 ffe i fmb1 br T b N koM b *b Voc A:.N1 brsh codd6.u.

Anmenf Deepn Asetenet Awnaery Systenn Page 9.347

_ _ ______ _ _ _ _ _ _ _ _ _ __ _ _ ___ __ _ __ j

I System 80+ Design ControlDocument Table 9.3.4-1B Reactor Coola'n t Detailed Startup and Power Operation Specifications Analysis RangeU3 Normal Abnormal Immediate Shutdown l

pH 4.5 - 10.5 - -

Conductivity [ Note 2] - -

Hydrazine, ppm 1.5 x 02 ppmts) _ _

(max. 20 ppm)

Ammoniat41, ppm 0-2 - -

Lithium, ppm 0.2 - 2.2 tsj _ _

liydrogen, cc (STP)/Kg II2 0 Power Operation 25 - 50 15 - 25 s5 Startup 15 - 25 - -

Oxygen, ppm s 0.1 > 0.1 < 1.0 Suspended Solid /4, ppm s 0.35 - -

Chloride, ppm s 0.15 >0.15 >1.5 Fluoride, ppm s 0.15 >0.15 > 1.5 Boron, ppm < Refueling - -

Concentration Sulfatel '3, ppm 5 .05 - -

Notes:

D3 This table expands upon operation specifications as depicted on T ble 9.3. lA.

[2]

Consistent with additive concentrations.

pl Prior to exceeding 150*F during startup.

I'3 This parameter is used for rapid problem diagnosis, tsj Consistent with plant lithium management program.

Approved Design Material Aun6ery Systems Pope 9.3-68 ,

System 80+ Design contror Document Table 9.3.4-5 Chemical and Volume Control System Parameters e-Paramater Value Normal letdown and purification flow 100 gpm Normal charging flow (to RCS) 90 gpm Normal charging mini-recirculation flow 35 gpm Normal seal injection flow 26 gpm Reactor coolant pump controlled bleedoff (4 pumps) 16 gpm Normal letdown temperature at loop 556*F Normal charging temperature at loop 445'F lon exchanger operating temperature 120*F O

1 i

1 Appemd Design Meterial AunGary Systems Page 9.3-80

System 80+ Design ControlDocument i

, )

( )

An exhaust air system rated for higher capacity than the supply air system, comp!' e te with full V filter train, two 100% exhaust fans and associated ductwork for each division.

I e Safety-related mechanical equipment room cooling units. 1 I

ne safety-related mechanical equipment room cooling units consist of a cooling coil with recirculation fan and dampers to remove heat generated within the space. A recirculation cooling unit is provided in addition to a once-through ventilation system because the served areas are potentially contaminated.

Applicable areas include the following: ' ' I

_]% % fekq l h,afeguard c_omponent Arcaninclu'd ng' afety/njection pump rooms,)ihutdown S f >oling pump dontainment/ pray pump rooms, fuel fool Heat % rooms, motor-driven and steam-driverpmergency feed [ ate r pump rooms,$1utdown poolingfleat'Xrooms, pontainmentppray H'eatgrooms, Shutdown doolmgpeat:X/ rooms, fenetration rooms, and associated piping and valve galleries.hd Y re . T T 9.4.5.2.1 }mm% ~

ponent-se r Description The safety-related mechanical equipment room cooling units consist of chilled water cooling coil, direct-drive centrifugal recirculation fan, and dampers and controls to achieve the desired operation. The chilled water coils are served from the essential chilled water system.

The safety-related mechanical equipment room ventilation units contain intake filters, direct-drive centrifugal supply and exhaust fans, and dampers and controls to achieve the desired operation. There O are heating and cooling coils to temper the outside air as required.

O 9.4.5.2.2 System Operation The Subsphere Building Ventilation System in composed of two divisionally separate, fully redundant ventilation systems each capable of being provided outside air by one 100% capacity supply unit and two 100% capacity supply fans per division. The supply air is filtered and then conditioned as needed by the heaters and cooling coils. There are two 100% capacity exhaust fans provided per division. Air exhausted is monitored by a radioactive gaseous detector sampling the air in the exhaust duct upstream of the exhaust filter train. The exhaust air is continuously processed through one 100% exhaust filter train complete with particulate filters and carbon adsorbers prior to discharge into the atmosphere.

Additional monitoring of the exhaust air is provided in the unit vent. Supply and exhaust fans are electrically interlocked such that the building will always remain under a slight negative pressure to direct all releases through the exhaust filter train. In the event of a loss-of-coolant accident, the ventilation equipment will continue to operate normally as long as offsite power is available. On LOOP, the exhaust fans will be powered from the Class IE diesel generators.

Normal operation of the safety-related mechanical equipment room cooling and ventilation units is as required to maintain space temperatures. The cooling systems will operate based on heat load as indicated by room temperature. In the event of a LOCA or DBA, all units are automatically started and will operate throughout the event.

9.4.5.3 Safety Evaluation The safety-related mechanical equipment room cooling systems consist of two completely redundant, (nV) independent full-capacity systems. Division I cooling system serves Division I essential mechanical equipment rooms, and Division Il cooling system serves Division II essential mechanical equipment Approved Desiger Meterial- AusiEery Systems foge 9.4-21

System 80+ Design ControlDocument Relief valves on the compressor discharge line and on the air receiver tanks protect the starting air system from overpressurization.

9.5.6.2.2 Component Description The starting air compressors are powered from a non-Class IE motor control center. During a loss of offsite power, the starting air compressors are powered from the alternate AC (AAC) power source.

Each compressor discharges compressed air and the heat of compression is removed by a water-cooled aftercooler. The component cooling water system provides cooling water on the tube side.

To minimize the accumulation of moisture, the diesel engine starting air system is equipped with a multi-stage drying and filtering unit located in line between the aftercooler and the receiver tank to supply air with a dewpoint at least 10'F lower than the lowest expected ambient temperature. The air is first passed through a cyclone-type moisture separator and is filtered before entering one of two alternating desiccant drying towers (alternating between active and regeneration cycles). The air is then filtered a second time before entering the receiver tank.

To minimize fouling of the starting air valves or filters with contaminants, drip-traps are provided on the cyclone-type moisture separator and the air dryer pre-filter to collect any oil carryover. Drains are also provided on the aftercooler and air receiver tanks. Periodic blowdown of the drip-traps and drain valves will minimize the buildup of contaminants in the starting air system. Strainers are provided upstream of the starting air solenoid valves to prevent rust carryover to the diesels.

Two starting air receiver tanks for each diesel engine provide storage capacity which is sufficient to allow five successful engine starts without the use of the compressor.

((The COL Applicant will provide specific diesel generator starting air system interface requirements.))

9.5.6.3 Safety Evaluation 6 %puQk The Diesel Generator Engine Starting Air Sys em is ANSI ass h the starting air compressor through the desiccant drying towers, and ANSIsClass 3 from the starting air receiver tank inlet check valve to the engine connections. The diesel engine and engine mounted components are constructed in accordance with IEEE Standard 387. The starting air aftercooler, which uses component cooling water on its tube side and the starting air receiver tank are designed in accordance with the requirements of the ASME Boiler and Pressure Vessel Code,Section III, Class 3.

The starting air compressor and the starting air dryer are designed in accordance with the requirements of the applicable codes.

9.5.6.4 Inspection and Testing Requirements System components and piping are tested to pressures designated by appropriate codes. Inspection and functional testing are performed prior to initial operation; thereafter, the system will be tested in accordance with the technical specifications.

O 3

COL information item; see DCD Introduction Section 3.2.

Apswered Design Material- Auxilary Systems Page 9.5-62

System 80+ Design ControlDocument Chapter 19 Contents (Cont'd.) h Page 19.11 Severe Accident Phenomenology and Containment Performance for the System 80 + PWR . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.11-1 19.11.1 Introduction .......................................... 19.11-1 19.11.2 Scope ................... ............... .......... 19.11-1 19.11.3 System 80+ Design Features for Severe Accident Mitigation . . . . . . . . . . . . 19.11-2 19.11.4 Severe Accident Phenomenology . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.1 1 -27 19.11.5 System 80+ Containment Performance Analyses ................. 19.11-133 e.o.9 summeg a J N&s 6ns , , , . . . . . .

e. u-96 Appendix 19.11 A Failure in the Presence of a Steam Generatoj(Partially Filled with_.LiquidjRepresentative Calculations Kegardmg a s stem Thermally Induced Steam Generator Tube Creeps. . . . . . . . . . . . 19.11 A-1 Appendix 19.1IB Bounding Analyses for DCH for the C-E Evolutionary PWR . . . . 19.11B-1 Appendix 19. llc Assessoast of the De-Entrainment Capability of the System 80+

Reactor Cavity . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19. llc-1 Appendix 19.1ID Two Cell Adiabatic Equilibrium Model for Direct Containment H e ating . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.11D-1 Appendix 19.11E Methodology for the Calculation of Containment Pressure Following a Hydrogen Burn . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.llE-1 Appendix 19.1IF Reactor Vessel Lower Head Failure Area . . . . . . . . . . . . . . . . 19.11F-1 Appendix 19.1IG Calculation of the Effect of 100% Oxidation of the Active Cladding on "in Vessel" and Containment Temperature . . . . . . . . . . . . . . 19.11G-1 Appendix 19.llH Comments on the Construction and Application of the System 80+

Containment Fragility Curve . . . . . . . . . . . . . . . . . . . . . . . . 19.11H-1 Appendix 19.11J Description of S80SOR System 80+ Source Term Methodology . 19.1IJ-1 Appendix 19.11K Hydrogen Mitigation System . . . . . . . . . . . . . . . . . . . . . . . . 19.11K-1 Appendix 19.llL Reactor Cavity Ultimate Static and Dynamic Pressure Capacity Calculation Methodology .......................... 19.11L-1 19.12 Containment Response Analysis . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.12-1 19.13 Consequence Analysis .......................... ........ 19.13-1 19.14 Containment Response Sensitivity Analyses ..................... 19.14-1 19.15 Summary of PRA-Based Design Insights . . . . . . . . . . . . . . . . . . . . . . . 19.15-1 19.15.1 Special Design Features . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.15-2 19.15.2 Internal Events Risk Profile Insights . . . . . . . . . . . . . . . . . ......... 19.15-8 19.15.3 External Events Risk Profile Insights . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.15-32 ;

19.15.4 Shutdown and Low-Power Operation Risk insights . . . . . . . . . . . . . . . . . . 19.15-43 j If.15.5 Use of PRA in the Design Process . . . . . . . . . , . . . . . . . . . . . . . . . . . 19.15-48 19.15.6 Risk Significant SSCs for Consideration in the RAP and Other Activities . . . . 19.15-51 19.15.7 Use of PRA to Support Certification Activities . . . . . . . . . . . . . . . . . . . . . 19.15-52 19.16 Re ferences . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.16-1 l

Appendix 19A Design Alternatives for the System 80+ Nuclear Power Plant .......... 19A-1 l l

Approved Design Afetenet Probablistic lbsk Assessment PageR l

I

Svstem 80+ Denim, CanalDocument .

1

] ~ Chapter 19 Tables 1

Page I i

19.7.5.1-1 Components in Seismic Fault Tree Models . . . . . . . . . . . . . . . . 19.7-3 19.7.5.1-2 Seismic Fragilities For System 80+ Structures and NSSS !

Components ................................... 19.7-14 ;

19.7.5.3-1 Summary of HCLPFs for Seismic Sequences . . . . . . . . . . . . . . . 19.7-15 i 19.7.5.3-2 Seismic Margins Sensitivity Analyses . . . . . . . . . . . . . . . . . . . . 19.7-16 .!

19.7.5.4-1 Comparison of Component HCLPFs for Rock and Soil- _

'i S ites . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... 19.7-17 !

19.7.5.4-2 Structure and Major NSSS Component HCLPFS for l Rock and Soil Sites . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.7-?!

19.7.5.4-3 Summary of HCLPFS for Seismic Sequences for Soil .;

Case B 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.7-22 ;

19.7.5.4-4 Summary of HCLPFs for Seismic Sequences for Bl.5 i Soil S ite . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.7-23 i 19.7.5.4-5 Summary of HCLPFS for Seismic Sequences for Soil l S ite B2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.7-24 i 19.7.5.4-6 Summary of HCLPFS for Seismic Sequences for Soil ;

Site B3.5 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.7-25 !

19.7.5.4-7 Summary of HCLPFs for Seismic Sequences for Soil !

S ite B4 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.7-26 !

19.8.1-1 Frequency of Core Damage from Shutdown Events . . . . . . . . . . . . 19 8-3 19.8.1-2 T&. of Leading CDF Sequences, Internal Events During Shutdown . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.8-3

19.8.1-3 Comparison of Shutdown PRAs . . . . . . . . . . . . . . . . . . . . . . . . 19.8-3 j 19.8.1-4 Examples of System Dependencies for the Shutdown l yCling System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.8-3 19.11.3.1-1 8 Axigymmetric Ultimate Stress Pressure Values . . . . . . . . . . . . . 19.11-148 f 19.11.4.1.1-1 Melt Composition at VB Following a Station Blackout . . . . . . . . 19.11-149 l 19.I1.4.1.1-2 Comparison of Various Exothermic Reactions Associated i with DCH Processes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.11-150 l 19.11.4.1.1-3 A Summary of Low Temperature Debris Dispersal !

Simulant Experiments . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.11-151 19.11.4.1.1-3B Summary of DCH/HPME Experiments . . . . . . . . . . . . . . . . . . 19.11-153- ;

19.11.4.1.1-4 Initial Conditions for "Two Cell" DCH Pressure Cal culation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.11-154 i 19.I1.4.1.1-5 Predicted HPME Pressures (psia) Using *2 CELL" DCH ;

Model . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.1 1 - 154 !

19.11.4.1.1-6 Conditional Containment Failure Probability Associated with DCH . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 19.11-155' 19.11.4.1.2-1 Summary of Subjective Containment Conditiont.! Failuie l Probability Due to "In-Vessel" Steam Explosion . . . . . . . . . . . . 19.11-155 19.11.4.1.2-2 TNT Equivalent Loadings for Various Mass Discharges into a Subcooled Liquid Pool (Efficiency = 3%, Initial Superheat = 5040'R) .. . . . . . . . . . . . . . . . . . . . . . . . . 19.1 1 -156 i

19. l l .4.1.2-3 A Probability Distribution for the Corium Mass Involved in E VS E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.11-156 Anma auton nonaw . neseseuse arek Aueamme neue a

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Svseant 80+ Dealen Coneof C:: d- }

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a i :.< 19.7 External Events Analysis l AA 4 7 l 19.7.1 Qualitative External Event Evaluation j t

his Section Intentionally Blank.- i 19.7.2 Tornado Strike Analysis

. His Section Intentionally Blank. j 19.7.3 Internal Fire Analysis {

t This Section Intentionally Blank. !

19.7.4 Internal Hood Analysis .

t This Section Intentionally Blank. l 19.7.5 Seismic Margins Assessment {

i 19.7.5.1 Methodology

((The COL must confirm the use of seismically robust electromechanical relays in the engineered safety ;

features actuation and control systems [ COL Item 19-10)))]l. !

19.7.5.2 Seismic Event Sequences l This Section Intentionally Blank. 1 19.7.5.3 Seismic Margins Raults and Insights !

I

((De COL applicant will be required to verify that key assumptions for structures, systems and j components considered in the Seismic Margin Assessment (SMA) are valid for the as-built plant l conditions [ COL Item 19-4]. His will include evaluation of High Confidence of low Probability of -l Failure (HCLPF) values for structures which house non-safety related equipment relied upon in the SMA 1 evaluations such as the combustion gas turbine. He verification process will include a seismic {

walkdown, including development of detailed procedures [ COL' Item 19-4] to ensure that as-built :

conditions ccaform to the assumptions used in the SMA and to assure that proper anchorage for j equipment has been provided and that the potential for seismic spatial system *demanoe does not exist.))l l

s the seismic analysis is being redonc, there is a need to augment the internal events model to the l Nxtent possible, by explicit inclusion of stracturaland other passive failures that were excluded from the j internal events model. His is COL Item 19-2j As part of the model development for the Seismic j Margins Analysis, the internal events model was updated to include structural and passive failures. His !

completes COL action item 19-2. I !

l 1 COL information item; see DCD Introduction Section 3.2.

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System 80+ Design ControlDocument Contents (Cont'd.) Page 4.7.1 Radioactive Gas Waste System Failure . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.8A-87 4.7.2 Radioactive Liquid Waste System Leak or Failure . . . . . . .............. 19.8A-87 4.7.3 Postulated Radioactive Releases Due to Liquid Containing Tank Failures . , . . . . . 19.8A-87 4.7.4 Fuel Handling Accident . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19. 8 A -8 7 4.7.5 Spent Fuel Cask Drop Accidents . . . . . .......................... 19.8A-87 5.0 Applicability of Chapter 6 LOCA Analyses to Lower Modes of Operation ... 19.8A-88 5.1 Issue . . . . .. . . . .. . . . . .. . . . . . . .. . ...................... 19.8A-88 5.2 Acceptance Criteria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.8A-88 5.3 Discussion .................... .................. .... 19.8A-88 5.3.1 Description of LOCA Scenario . . . . . . . ......................... 19.8A-90 5.3.2 Selection of Reference Plant Parameters and Conditions for Mode 4 Analysis . . . . 19.8A-91 5.3.3 Ana'ysis Computer Codes . . . . . . . . . . . . . . . . . .................. 19.8A-91 5.3.4 LOCA Analysis for Mode 4 .................................. 19.8A-92 5.3.5 Results of LOCA Analysis for Mode 4 . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.8A-92 5.4 Resolution ......... ..................... ............. 19.8A-94 6.0 Applicability of Chapter 6 Containment Analyses ..................... 19.8A-95 6.1 Introduction ........ ................................... 19.8A-95 6.2 Loss of Coolant Accidents (LOCAs) ............................. 19.8A-95 6.3 Main Steam Line Breaks (MSLBs) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19. 8 A-96 6.4 Inadvertent Operation of Containment Heat Removal Systems . . . . . . . . . . . . . . 19.8A-96 6.5 Con cl us ion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.8A-97 7.0 System 80+ Design Features for Simplicity of Shutdown Operations . . . . . . . . . 19.8A-97 7.1 Introduction ............................................ 19.8A-97 7.2 Discussion ............................................. 19.8A-97 7.2.1 Technical Specifications for Reduced Inventory ...................... 19.8A-97 7.2.2 Shutdown Cooling System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.8A-98 7.2.3 Containment Spray System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.8A-98 7.2.4 Component Cooling Water System .............................. 19.8A-98 7.2.5 Station Service Water System . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.8A-98 7.2.6 Electrical Distribution System . . . . . . . . . . . . . . . . . . . . . . .......... 19.8A-99 7.2.7 Nuplex 80+ Advanced Control Complex .......................... 19.8A-99 7.2.8 Reduced Inventory Instrumentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.8A-100 7.2.9 Containment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19. 8 A- 101 7.3 Conclusion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19. 8 A- 102 8.0 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19. 8 A-102 9.0 References .................... . . . . . . . . . . . . . . . . . . . . . . . 19. 8 A- 103 AppewL A +e Aprosdk i+8A . - . - . .

19.euxi-i A r k B 1o AppendN t9 86 . . .

i9.s are3-1 Apfwoved Design Meterial Probablistic Risk Assessment Pageir

ii System 80+ Oesian contmlDocument h 9.0 References For Appendix 19.8A

1. Letter LD-92-038, C. B. Brinkman (ABB-CE) to D. M. Crutchfield (NRC) dated March 25, 1992.

2. Memo Letter, " Summary of Meeting Held on December 18,1991 Regarding Shutdown Risk,"

T. V. Wambach (NRC), dated January 30,1992.

3. NUREG-1449, " Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States," Draft Report dated February 1992.

4. USNRC Generic Letter No. 88-17, " Loss of Decay Heat Removal," dated October 17, 1988.

5. USNRC AEOD Special Report, " Review of Operating Events Occurring During Hot and Cold Shutdown and Refueling," dated December 4,1990. !

6. NUREG-0800, USNRC Standard Review Plan, Revision 1, July 1987. /'

7. (Jaquith, RMProbabilistic Risk Assessment for the System 80+ Standard Design,"

~

Combusilon Engineering Inc., DCTR-RS-02, January 1991. '

8. Letter from James H. Wilson (NRC) to E. E. Kintner (EPRI) dated September 5,1991.

9. Brockhold, G. (EPRI), Memo to ALWR Utility Steering Committee, ALWR-92-18, January 15,

' 1992.

10. This Reference was deleted.

i

11. ANSI /ANS-58.8-1984. I

12. CEN-373-P, Volume 1, "Re:listic Small Break LOCA Evaluation Model, Calculationai Models,"

April 1988.

Volume 2, " Application of Evaluation Model," December 1988. '

Volume 2 Supplement 1-P, " Application of Evaluation Model to Calvert Cliffs Units 1&2,"

September 1989. !

Volume 3, " Computer Program Input and Output Description," December 1988.

13. Lette LD-88-030,"Submittalof Realistic SBLOCA Evaluation Model," A. E. Scherer(ABB CE) !

to J. A. Norberg (NRC), April 27,1988.

(

C 3- %

14. Letter LD-88-155, " Submittal of Volumes 2 and 3 of Combustion Engineering's Realistic Small Break Loss-Of-Coolant-Accident Evaluation Model," A. E. Schere; (ABB CE) to J. A. Norberg ;

(NRC), December 9,1988. '

O ;

Approved Design Meteriel hobab6saic Nsk Asseesmer,t Pope 19.8A.103

. . l System 80+ Design controlDocument 1

Response 440.86(f)

(f) The table of assumed initial conditions (see Chapter 15) has been expanded to include the worst initial conditions for each event considering events occurring at all modes of plant operation.

This expanded table is presented in Section 4.0 of Appendix 19.8A.

SRXB Question 440.91 Most of the Chapter 15 and 6.3.3 (LOCA) analyses are performed based on the event being initiated at full power operation. The staff requires that C-E provide an assessment on the consequences of the transients and accidents initiated at low power levels or lower modes of plant operation such as shutdown operations. This is required to demonstrate that the analyses performed m CES R- e the bounding cases 'ar all modes of plant operation.

Response 440.91 This assessment is prouded in the System 80+ Shutdown Risk dvaluation in Appendix 19.8A. Section 4.0 of the Appendix contains the ass went of Chapter 15 events and Section 5.0 contains the assessment of Section 6.3.3 events. Also see thc sesponse to RAI 440.16(j).

SRXB Question 440.109 (15.6.3)

Provide the results of an analysis for the potential boron dilution event during the recovering phase following a SGTR when backfill from the seconday system through the ruptured steam generater occurred.

Response 440.109 The System 80+ Emergency Operations Guidelines include steps to prevent backfill from the secondary system through the ruptured steam generator by maintaining a positive pressure difference between the primary and secondary systems. (See Section 2.1 of Appendix 19.8A.) Therefore, boron dilution should not occur and has not been analyzed. A further note is made that backfdl is not necessary to prevent overfilling of the larger System 80+ steam generator as the result of a SGTR event.

SRXB Question 440.129 For the purpose of advanced reactor design reviews, the term shutdown risk encompasses operation when the reactor is suberitical or in transition between suberiticality and power operations up to five percent RATED THERMAL POWER; i.e., for the System 80+ design between MODE 6 and MODE 2 where the modes of operation are defmed in Chapter 16, Table 1.1-1. 1 l

It should be noted that the NRC risk evaluation program is an ongoing study where the projected l completion date is scheduled for the end of the 1991 calendar year. At such time, additional concerns not addressed in Generic Letter (GL) 88-17 will be emerging and additional measures may be required of the System 80 + design in order to ensure there is an adequate level of protection for the public health and safety during shutdown and low power operations. The following shutdown risk RAls are i supplemental to RAI 440.16 issued in December of 1990, and the corresponding responses should be l specifically addressed in Chapter 20 and the associated sections,gESSAR-Dp l l

i Awved Design besterial Probabliscic Risk Asseesanent Page 19.8A(A]-20 I

.= = . .

r Svstem 80+ Derlnn ConkelDocannent SRXB Question 440.140 How will plant instrumentation be designed to operate properly during shutdown operations, and what I key parameters will be monitored? The answer to this question should address particularly the issues of instrument availability and appropriate ranges for shutdown measurements, including:

A description for any deviations, bases and justifications of the deviations, from the Generic i Letter 88-17 recommendations that each plant provide two independent RCS level indications, ;

two independent core exit temperature measurements, the capability of continuously monitoring ,

decay heat removal (DHR) system performance whenever a DHR system is being used for cooling the RCS, and visible and audible indications of abnormal conditions in temperature, level, '

and DHR system performance.

e Indicate what instruments identified in resolution of GSI-99 (see Chapter 20) and Section 5.4.7 are classified as safety related and therefore within the scope of environmental qualification and quality assurance criteria. For those that are not so classified, provide a description of the quality assurance program that will be used to provide reliable instruments with accurate information in the expected ranges of shutdown measurements that will be enhance operator confidence in the instrument, and the training program for operators to understand and interpret and the data '

provided by the instruments.

Response 440.140 The response to this item is provided ie Sections 2.1,2.2,2.3 and 2.8 of Appendix 19.8A.

SRXB Question 440.141 Emergency operating procedures (EOP) are generally developed for use in normal plant operation ;

assuming the availability of the plant systems. During shutdown, however, many systems will be out for maintenance and the plant is m a different configuraticn. What is the CE System 80+ emergency ,

procedure guidelines for the development of enhanced EOPs to provide better operator guidance for !

shutdown and low power operation? Are there any additional entry conditions to address shutdown and ,

lower power operations? What wii! the scope of EOP enhancement be, and will they be symptom- or l cvent+riented? '

Response 440.141 A comprehensive assessment of risks associated with shutdown operations was performed for the System 80 + design and the results are summarized in Appendix 19.8A. Abnormal and accident events postulated to be initiated during shutdown operation were evaluated and means to prevent, detect and monitor them .

and to mitigate consequences were developed.

Based on this analyses, guidance was developed to respond to abnormal events initiated from shutdown conditions. The guidance is intended to be used for development by the plant owner / operator of j appropriate procedures to govern shutdown operations. It is not intended to formally incorporate guidance for shutdown accident response into the EOPs, since the number of possible initiators and contributing occurrences is extremely large. The appropriate approach for reducing shutdown risks is O to adequately plahutage activities and sequences, have sufficient means for monitering the availability and performance o systems, and to use the diverse and redundant means incorporated in the design to mitigate consequen s. These are addressed in Appendix 19.8A. Abnormal operating procedure or t

Anwovent Dennyn atetonal hebahnseio niek Assersment Page 19.aNA]-27 f

t SVstem 80+ Deslan ControlDocument i

Appendix B to Appendix 19.8A :

l' Procedural Guidance to Support Reduced RCS Inventory Operations This Appendix Intentionally Blank 4

l b

19.B0Eed-t

- - . ,. _ . m .. .._, g) ,

E System 80+ Design contrar Document For intermediate pressure transients, the potential for an induced RCS hot leg failure will be conservatively neglected (See Section 19.ll.4.1.1.3) _indu ed steam generator tube rupture is also assumed to be negligible. [Ae-likelihond 19.11.4.1.1.7.2 Quantification of DCHSTREN :DCII Containment Failure Probability Given a high RCS pressure and that RV lower head failure occurs into a " dry" reactor cavity, the containment failure probability is conservatively based on the results of a high avel decomposition of the DCH event into its principal contributors. Peak pressures were established for a variety of conditions and a weighted containment failure probability was established. The process was performed separately for DCH events that are initiated from PDS where the RCS leak rate is governed by a cycling relief valve (CRV) and for the high and intermediate pressure state where a continuous RCS leak is present. Results for this assessment were established using the "two cell" DCH model discussed in Section 19.11.4.1.1.5.2.

19.11.4.1.2 Rapid Steam Generation 19.11.4.1.2.1 In-Vessel Steam Explosions (IVSEs) 19.11.4.1.2.1.1 Description of Phenomena The concept of a fuel induced " steam explosion" within the reactor pr:ssure vessel refers to a phenomenon in which molten fuel rcpidly fragments and transfers its energy to the coolant resulting in steam generation, the development of shock waves and the acceleration of large RV internal masses with possible mechanical damage and failure of he RV. As a consequence of such explosions, there was a concern that missiles would be generated that might contact and locally penetrate the containment and allow for early rad'ation release to the environment. This containment failure mode was initially considered in the Reactor Safety Study (WASH-1400) as the alpha-mode failure. Recent assessments of steam explosion phenomena have suggested that IVSE "do not provide a credible threat to the integrity of either the primary system or containment" (Reference 117).

19.11.4.1.2.1.2 Parameters Affecting In Vessel Steam Explosions (IVSE)

For a steam explosion to produce a threat to the containment, the interaction process must have the following:

1. Sufficient corium mass

2. Favorable geometrical configuration

3. High energy conversion

4. Triggering mechanism

5. Production of a sufficiently energetic missile These issues were investigat~d by the Steam Explosion Review Group (SERG) as they apply to steam explosions within the RV (Reference 126). Most members of the review group believed IVSE could occur but that the probability of producing a containment threatening IVSE was on the order of 104 (See Table 19.11.4.1.2-1). This conclusion was reached despite the expression of differing opinions Approved Design Afsterist ProbabiEstic Risk Assessmerst page 19.1143

System 80+ Der'an ControlDocument 7,

The FITS program lasted several years and included over 100 experiments. Fuel masses varied

() between 2 and 20 Kg. The majority of the FITS experiments used thermitically generated iron-alumina, and the remaining tests used thermitically generated corium.

These tests indicated that energetic fuel - coolant interactions were possible for corium. Thermal energy conversion ratios were found to be in the 0 to 3% range with an uncertainty factor of about 2 (Reference 132). A typical plot of conversion efficiencies established from the FITS A and B test series is presented in Figure 19.11.4.1.2-1. These tests indicate a median conversion efficiency is about 1.5%.

Parametric studies performed as part of FITS also provided the following:

For coherent melt deliveries (that is the melt is delivered in one mass) into water at ambient temperature and pressure 32 explosions were observed out of 37 tests (Reference 132). Thus, the probability of a spontaneous steam explosion under these conditions can be established at 0.86.

Similar experiments conducted with saturated water and at ambient pressure indicated a steam explosion probability of 0.24 (4 observed explosions out of l~i Msts).

en-The influence of pressu on the probability of a spontaneous steam explosion was established in FITS-C (Reference 1 . No spontaneous steam explosions were observed for all five FCI tests conducted at ambient water temperature and a pressure of 5 bars (75 psia), thus, experimentally supporting the position that high pressure steam explosions are extlemely unlikely events.

O Additional experiments were conducted to ascertain the importance of melt delivery on the steam h explosion process. While under certain circumstances, coherent melt deliveries were observed to produce steam explosions, a pre <lispersed delivery of fuel debris was not.

19.11.4.1.2.1.3.3 Ispra High Pressure Experiments in Reference 126, Dr. Mayinger referred to a steam explosion test program performed by EURATOM to establish the influence of system pressure on steam explosions. While details of this test series are unknown, Dr. Mayinger noted that a general conclusion drawn from these experiments was that initiation of steam explosions at pressures greater than about 300 psia require very strong detonative triggers.

19.11.4.1.2.1.4 Significance of IVSE to System 80+

Based on a review of available steam explosion data and analyses, it appears that sufficient information is available to conclude that the probability of containment failure resulting from a corium-coolant interaction (CCI) event is very low (on the order of 0.001 or less). Much of these assessments considered typical PWR geometries analogous to that of System 80+ and are, therefore, considered applicable to System 80+.

19.11.4.1.2.1.5 Applicability of IVSE to the PRA ,

1 The above information is believed to be generally applicable to the System 80+ PRA. Therefore, for the purpose of System 80+ risk assessment, the containment failure caused by an IVSE was taken to have a very small mean containment failure probability for severe accidents where the RCS fails at pressures less than 250 psig and an order of magnitude lower when the RCS fails at high pressures.

This probability is defined in the PRA as variable VB-ALPHA.

Appm.s ouiso, untenet Probasanic nisk An nmart pas, rs. s s.es ;

1 System 80+ Design ControlDocument I i'

n

( ; reactor vessel, further attempts to increase the strength of the lower cavity to accommodate ;

d severe accident pressure are not necessary.

This feature of the System 80+ cavity design guarantees that steam explosion loadings in the reactor cavity (even those that fail the cavity lower walls) will not be sufficient to induce a failure of containment integrity.

2. Even if significant RV motions are postulated, motions of the RCS are restricted via several mechanical restraints which prevent excessive motion of major structures such as the steam generators. If the steam generators remain well constrained (as is currently expected) piping connected to the generators will not fail. Structural integrity of RCS connected piping is also expected.

3. Should a failure of the RV supporting structure occur, and the motion of piping connected to the RCS results in a piping failure, the failure will likely be contained "in containment".

Thus, containment integrity will not be compromizd. This conclusion is primarily a result of the high flexibility associated with the smaller RCS connecting piping and the tendency to design piping penetrations stronger than the connecting piping.

19.11.4.1.2.2.3 Summary of Experimental Evidence

/

A discussion of steam explosion experiments is provided in Subsection 19.ll.4.lh q

C,r In addition to these experiments, steam explosions have also been observed in SANDIA HPME experiments performed as part of SPITS and HIPS test program (P.eference 158), investigating pressurized melt ejection into the water pools. This program included five flooded cavity tests. All tests were observed to produce steam explosion loads characterized by short duration (several milliseconds), high amplitude pressure pulses that typically disassembled the test apparatus ,

(Reference 158). These tests injected an iron-aluminum melt into small relatively enclosed cavities.

These tests were not conducted within the framework of the Severe Accident and Scaling Methodology (SASM) effort and therefore were not assessed to be prototypical of steam explosion loads within PWR cavities.

Additional DCH tests have been performed with flooded cavities as part of the 1:10 scale IET DCH !

program performed at Sandia (IET tests 8A and 8B). These tests injected a corium melt simulant into l a one half flooded Zion cavity. Test IET-Ba injected water from a vessel initially around 150 psia.

Test IET-8B was conducted from a nominal initial pressure of about 900 psia. DCH effects were clearly suppressed. Steam explosions were observed for both tests. The experiment with low pressure core melt injection resulted in a single delayed steam explosion raising the cavity pressure to about 600 psia. The high pressure melt ejection transient produced several smaller (200-400 psi) l steam explosions (Reference 190). l i

A recent experiment performed as part of the Beta Core-Concrete Interaction Program in Germany l (BETA V6.1) has produced a steam explosion that lifted a 7 ton cylinder several meters off its l foundation (Reference 185). While the precise details of this experiment are unavailable, loads were i produced which were equivalent to a 3% conversion efficiency of the corium thermal energy into kinetic energy (Reference 185). A review of this test performed by G-E (Reference 186) suggested D

(b that the loading that developed during Beta V6.1 were due to a rapid quasi-steady pressurization of the KWU reactor cavity. The KWU reactor cavity is relatively tight with a free volume less than i

Anwered Desigen Material. Probab5stic Rusk Assessmerrt Pope 1A 1149

System 80+ D,esign ControlDocument p

19.11.4.1.2.3.2.2 RCS Conditions at VB The energy and mass associated with the RCS steam / water discharge following VB will establish the increment in containment loading due to direct mass and energy addition into the containment. This containment pressurization process is analogous to the containment pressurization following design basis pipe breaks.

19.11.4.1.2.3.2.3 Mass and Superheat of Corium Debris Following VB, steam will be generated in the process of quenching the corium debris. In this process the stored energy from the corium is transferred to the water which in turn is vaporized.

Experimental data on corium quenching indicates that the quenching process exhibits maximum heat fluxes of up to 30 Mw/m 2for short time periods.

19.11.4.1.2.3.2.4 Availability of Water The amount of corium that can be quenched is dependent on the availability of water. If insufficient water is available, quenching will not be complete and steam generation will be limited.

19.11.4.1.2.3.2.5 Contribution Due to Exothermic Reactions In the process of quenching, the metallic portions of the corium debris may release considerable quantities of energy as a consequence of oxidation. As a result of the rapid debris cooldown it is p expected that oxidation will be limited to less than 50% of the molten metallic material.

b 19.11.4.1.2.3.3 Significance to System 80+

The peak containment pressures resulting from rapid steam generation events following a System 80+

RV lower head breach are summarized in Table 19.11.4.1.2-4 for selected severe accident scenarios. These scenarios include a station blackout, a "V" sequence (interfacing systems LOCA) l ar.d a Large LOCA in a containment without :vailability of containment sprays. These events typically span the range of interest for estimating post VB rapid steam generation pressure spikes.

For the first two scenarios, the initial containment pressure will be less than 25 psia, for the last sequence the LOCA was assumed to result in a design basis challenge to the containment resulting in an initial pressure of 67 psis. (This assumption neglects any passive cooling of the containment due l to heat transfer to the containment shell and structures within the containment. Design basis analyses and MAAP calculations suggest that these heat sinks will contribute to a reduction in containment pressure of about 20 psi in a three hour time interval prior to VB.) The containment is subsequently pressurized by a combined high pressure steam release and steam generation due to a rapid quenching of the corium debris. For this study the corium mass quenched comprised 65% of the total core mass (including support structure) and was initially discharged into the containment at 2800*K. The mass and mass distribution used for this study were extrapolated from results of the NRC sponsored SASM l activity presented in Table 19.11.4.1.1-1. To ensure that all steam generation modes were accounted {

for, it was also assumed that 50% of the quenching debris oxidizes during the quenching process.

The resulting bounding containment pressures were established for these scenarios. Results of this l analysis are presented in Table 19.11.4.1.2 () Analogous analyses were performed for the Large l Break LOCA. However, in this analysis containment sprays were not credited and the mass and energy of steam released from corium quench,m at VB was negligible. The peak pressure calculated (V) using this methodology was below 98 psia. While these loadings are above the design basis i

4 Apywored Den &n hintenal . ProbabiEstic Risk Assensmont Page 19.11-55

System 80+ Design ControlDocument n

(') detonations, if they should occur, may potentially pose a threat to containment integrity and to the continued operaGon of mitigative equipment.

l 19.11.4.1.3.2.2 Parameters Affecting IIydrogen Detonation Two classes of hydrogen detonations are typically distinguished: (a) detonation via direct initiation by high explosives and (b) Deflagration-to-Detonation Transition (DDT) resulting from an energetic burn in a confined obstructed geometry. Hydrogen detonations are influenced by (1) hydrogen concentration, (2) presence of inenents, (3) the ignition source and (4) system geometry (scale and configuration).

19.11.4.1.3.2.2.1 Hydrogen Concentration Experimental evidence has indicated that under favorable geometrical conditions a hydrogen detonation in dry air is possible at values of hydrogen concentration as low as 9.5 v/o. Detonations actually produced at these low hydrogen concentrations require a hot, dry mixture and the use of explosive charges (See Reference 132). Based on this observation, the National Research Council reached the conclusion that mixtures of 9 to 11 v/o hydrogen might be detonable. In practice detonations at this low hydrogen concentration are not considered credible in a post severe accident LWR environment. Even at hydrogen concentrations of 13% by volume a substantial energy source would be required to directly initiate a detonation (See Section 19.11.4.1.3.1.2.2).

Another mechanism for producing a detonation involves flame acceleration. Flame acceleration p3 occurs due to turbulerce induced by fans, structural roughness, obstacles, or changes in geometry.

y/ Flame acceleration is only imponant for mixtures that can be classified as highly flammable. Flame acceleration which results in sonic propagation of a detonation front undergoes a deflagration to detonation transition (DDT) and requires concentrations greater than 12% in dry air. The lowest concentration for which DDT has been observed is 15% (See Reference 138), and even then only in dry air and with ideal geometric conditions.

19.11.4.1.3.2.2.2 Ignition Source Direct initiation detonation of lean hydrogen mixtures (below I /o) in an open containment would require a trigger of more than 10 MJ (See Figure 19.11.4.1.3,). In contrast the energy required to initiate a deflagration is more than 10 orders of magnitude lower than that for detonation. 'Iherefore, without an appropriate energy source hydrogen detonations are not possible.

19.11.4.1.3.2.2.3 Steam Inerting of Containment The presence of steam in the containment atmosphere can decrease the potential for, and severity of a hydrogen detonation. Experiments performed to date suggest that volumetric steam concentrations greater than about 30% will render even a stoichiometric mixture of hydrogen and oxygen in a non-detonatable state.

19.11.4.1.3.2.2.4 Geometry Geometrical features can have an important influence on the potential for hydrogen detonation. In

( hydrogen mixtures which spontaneously undergo DDT, the ability of the system to detonate is

( dependent on the level of confm' ement and presence of obstacles. Typically, open geometries are not Approved Desogn Msterial- Probabilstic Stish Assessment Pope 19.1165

o ,

System 80+ Design ControlDocument the steam generator enclosures. Geometric class 4 structures were associated with the reactor cavity, letdown and regenerative heat exchanger rooms, HVAC header and the IRWST. Such geometries are unfavorable to DDT. Mapping the flammability and geometric classifications on the Table 19.11.4.1.3-6 matrix indicates that the System 80+ containment to be primarily a class 5 containment. This classification implies that the potential for a DDT is highly unlikely to impossible.

For selected accident sequences with significant hydrogen discharges to the IRWST the local mixture classification within the IRWST may formally increase, producing an overall resultant class 4 detonation ranking. This would imply that in a dry atmosphere these mixtures can potentially detonate, but the process is unlikely. The presence of modest amounts of steam expected in the IRWST would likely render the mixture inert to detonation. Since the IRWST is expected to be steam inerted (and oxygen starved), this ranking is likely overly pessimistic, and the mixture is closer to a class 5 system.

It should be noted that the Sherman / Berman Ranking Scheme was developed for dry air-hydrogen mixtures. The addition of steam Sas a profound effect on the detonation potential. Even small amounts of steam (on the order of 1 volume percent) will be sufficient to increase the minimum detonatable hydrogen concentration to above 15 v/o. Furthermore, above 30 v/o steam concentrations, hydrogen mixtures will be inert to detonations. These features are also not fully considered in the current ranking procedure.

19.11.4.1.3.2.4.2 Direct Detonation of Hydrogen Within the System 80+ Containment A second source of hydrogen detonation can arise from direct ignition of a flammable mixture.

Direct ignition detonation typically requires an explosive charge within a highly flammable containment atmosphere.

p sources typically availableReference 132 compared in PWR containments. the energy This figure is reproduced required as Figure 19.11. for4.a ).detona M

i From this figure it can bc clearly seen that containment ignition sources have energies which are more than three orders of magmtude lower than that necessary to detonate a 13 v/o dry hydrogen mixture in an unconfined geometry. On the other hand all ignition sources (even those of 10 orders of magnitude lower strength) are sufficient to cause a deflagration.

Based on the above work and supporting analyses presented in Reference 137, the possibility of detonation within the System 80+ containment is considered remote. Direct initiation of a hydrogen detonation would be improbable within the System 80+ containment while initiation of a deflagration during a severe accident is virtually certain. Similarly, an assessment of the intrinsic flammability and geometric features of the System 80+ containment indicates the potential for DDT is highly unlikely to impossible.

19.11.4.1.3.2.4.3 Role of Hydrogen Mitigation System For severe accident application the purpose of the hydrogen igniter is to respond to NRC concerns (References 114 and 116) with regard to hydrogen control during a severe accident. Specifically, the installation of hydrogen igniters is intended to satisfy 10CFR50.34(f)(2)(ix) " Additional TMI-Related Requirements". This rule requires "a hydrogen control system that can safely accommodate hydrogen generated by the equivalent of a 100 percent fuel-clad metal water reaction. The system must also nsure that uniformly distributed hydrogen concentrations in the containment do not exceed 10 percent ay volume . . .* A detailed description of the hydrogen igniter system can be found in Section 6.2.5 and in Appendix 19.llK. It is expected that the hydrogen risk due to detonation is negligible even without the presence of the HMS, as indicated via the Sherman / Berman Ranking assessment. The Approved Design Material- ProbabMstic Risk Assessmerrt Pope 19. r1-68 4

System 80+ Desinn ConvolDocwnent 19.11.4.1.4.2 Cavity Overpressure Failure 19.11.4.1.4.2.1 Description of the Phenomena Fonowing a HPME, large quantities of steam and corium are discharged into the lower portion of the reactor cavity. This discharge can potentially challenge the integrity of the reactor cavity and thereby threaten containment integrity. Cavity overpressurization can potentially result in a structural failure of the reactor cavity and associated RV supports. Failure of the RV supports can produce excessive motions in the RCS and steam generators potentially failing a containment penetration or producing an -

unisolable breach in piping exiting the containment.

Potential sources of cavity overpressurization include the EVSE event and the energetic failure of the RV lower head. The EVSE induced failure of the cavity and or reactor internal supports is considered in Section 19.11.4.1.2. This section considers localized cavity pressurization induced by steam pressurization of the reactor cavity space immediately upon RV lower head failure.

19.11.4.1.4.2.2 Significance to S/ stem 80+

System 80+ is expected to withstand cavity pressurization events following RV lower head failure.

This capability of System 80+ arises from the robustness of the System 80+ reactor cavity design which includes (1) a Rh --setor cavity wall strength and (2) a large reactor cavity volume.

The post severe accident cavity pressurization performance of the System 80+ design was evaluated analytically for a simulated high pressure superheated steam blowdown following a postulated breach in the RV lower head. Analyses were performed using the ABB-CE DDIF Mod 7 (Reference 141) cavity pressurization computer code, in this analysis a multi-compartment representation (see Figure ?9.11.4.1.4-1) was assembled to provide a detailed simulation of the reactor cavity.

Pressui2ations were established using a spectrum of RV lower head failure sizes ranging from the equivalent of a lower head instrument tube failure to a large creep failure of the RV lower head.

Results of these indicate System 80+ cavity loadings to be below 100 psid.

MAAP analyses for similar transients indicate that the cavity pressure 6se associated with a best-estimate HPME event caused by a single ICI tube failure, will be under 20 psi. For either evaluation, predicted loads are below the cavity wall design pressure values of approximately 188 psid (ultimate strength of 235 psid) and consequently will not challenge cavity or containment integrity.

19.11.4.1.4.2.3 Application to the PRA The cavity overpressurization induced containment failure is not considered a credible threat to containment integrity. However, for the purpose of completeness this failure mechanism has been included in the PRA supporting logic models with a very small probability.

19.11.4.1.4.3 Rocket Induced Containment Failure 19.11.4.1.4.3.1 Description of Phenomena til The issue of rocket induced containme'nt failure was formally addressed in the Oconee Level 3 PRA O performed by NSAC (Reference 346). In this assessment, Battelle Columbus reviewed the potential for containment failure due to an in-containment reactor vessel ' lift-off" following_the failure of the RV lower head. The rocket analogy resulted in the authors of Reference /jX identifying rocket QLp Anweeent Demon menonier. hosehnee*c niek Assenemere rope 1s.11-71

System 80+ Design Control Document

(' Experiments pertinent to debris coolability are summarized in Table 19.11.4.2.2-1. Additional details on these experiments as they relate to the ALWR are presented below and in Reference 150.

19.11.4.2.2.2.3.1 Debris Configuration It has been shown by several investigators that the morphology of quenched debris depends upon the relative amounts of liquid debris and water present. Breakup of debris jets can occur if the water depth is sufficient. Otherwise, channeling and accumulation of debris can occur. Small particulate debris breakup (less than 1 mm) is typically not conoucive to debris cooling in that packed debris beds of low porosity exhibit a steam / water counter-current flow phenomenon which makes water penetration difficult. Conversely, high porosity beds of modest decay powers typical of that associatedwith decay heat at times greater than 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months after shutdown, should be easily coolable by an overlying water pool. Experimental evidence indicates that a mixture of both particulate and a continuous phase occur.

Experiments of particular note are simulated corium drop experiments performed by Benz (Reference Ip q59) as well as the Corium Water Thermal Interaction (CWTI) and Corium-Coolant Mixing (CCM) experiments performed by Spencer (Reference 148). In the Benz experiments molten steel or uranium dioxide changes were dropped into an interaction vessel containing excess water and the resultant debris fragmentation was measured. Based on these experiments the smallest average particle diameter was about 2 mm with more than 60% of the debris being greater than 4 mm.

The CWTl tests covered a range of experimental conditions. Of particular interest were tests CWTI-7 p through CWTI-10 which investigated the fragmentation of a uranium zirconium oxide and stainless

( steel mixture which entered a water filled interaction vessel in circular jet. An examination of the debris indicated that the oxide debris was in the form of an internally porous (about 50% porosity) mass.

In the CCM experiment, the presence of deeper water pools resulted in a greater extent of melt break-up and particulate formation giving rise to the collection of loosely bound debris and internal porosity. Characteristic particle sizes ranged between 1 and 5 mm.

Based on these tests, ARSAP (Reference 150) concluded that for prototypic debris and representative debris / water volumes, debris fragmentation would be limited and the majority of the debris will form a continuous porous slab.

Recent experiments conducted by NRC and EPRI suggest the potential for a continuous crust to form at the upper portion of the corium melt. In these tests where crusts were observed the heat transfer appeared to be impacted by its presence. The WETMET and WETCOR experiments (Reference 191 and 192) indicated that prior to crust formation heat transfer from the corium debris would proceed at a rate commensurate with nucleate boiling. Once a crust forms the heat flux from the upper surface was observed to reduce to a stable, heat transfer between 300 and 400 kw/m2 ,

The MACE test series conducted by EPRI (Reference 193) also indicated a stable crust formation.

These tests demonstrated high initial heat fluxes (about 3.5 Mw/m2 ) followed by a decrease in heat 2 2 removal from 600 Kw/m down to about 150 kw/m . The reduction in heat transfer was attributed to the development of an upper crust that separated from the corium melt as the corium simulant O quenched. The ability of the crust to anchor on the test section sidewalls, and the influence of the U anchored crust configuration on the test results are still under investigation as part of the MACE program.

~

Anwormt Design Metenat- hobabilstic Risk Assessment Page 19.11-83

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System 80+ Design ControlDocument

) The MACE tests simulate the corium debris as a mixture of UO 2 , ZrO 2

and Zr. To date, the MACE tests have provided mixed results regarding debris coolability. While the MACE scoping test only established a maximum stable heat flux of 0.6 Mw/m2 which decayed in time, it was noted that the details of the test facility may have contributed to providing an insulating debris surface. When the crust was not present during the test, heat fluxes were in excess of 2 Mw/m2 The MACE scoping test involved a corium charge of about 300 lbm over an arca slightly less than I square foot. The collapsed depth of the debris wa> initially 15 cm. The crust stability was aided by the sidewall design.

By the end of the test, a 2 - 5 cm thick crust had formed (Reference J9d).

l%3 The most recent MACE test involved 960 lbm of simulated corium concrete attack (Test MIB) in a 4 square foot test facility. The test indicated substantial debris quenching and a long duration vigorous heat removal of 2 Mw/m 2was observed. Six hours into the test, concrete erosion was noted to be between 15 and 20 cm and the erosion rate had reduced to I cm/hr. At this time corium quenching was observed. A power reduction step was investigated in order to establish the debris erosion rates 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after shutdown. As with the previous MACE test an anchored bridge crust formed during the experiment. This crus appears to have separated from the melt as downward erosion progressed.

This behavior is considered a facility design issue and will not be prototypical of actual reactor scales.

During the initial contact the heat transfer from the crust allowed a heat flux of upwards of 600 2

Kw/m . As the melt receded from their crust this heat flux decreased in time.

19.11.4.2.2.2.4 Melt Spreading The ability of the corium melt to spread within a water pool was investigated by Greene (Reference A 157). These experiments were used to establish a correlation between a nondimensional spreading V thickness against a nondimensional spreading number. Applying typical ALWR geometric data into this correlation (see Reference 150), suggests that the spreading of the corium debris can be expected to be relatively complete for the ALWR.

Reference 150 also indicates that even if full spreading is not realized, the ability to remove heat via the sides of the debris bed would enhance debris coolability. Therefore, while full debris spreading is expected, it is not required in establishing corium coolability.

To support the assessment of coolability by the ARSAP Program, the MELTEPREAD-1 computer program (Reference 204) was used to calculate the spreading of core materials inside the cavity of various advanced reactors following a localized lower head failure. MELTSPREAD-1 models the j transient speading of reactor materials over a concrete or steel lined concrete substrate accounting for melt interactions with both the substrate and overlaying water. T%ese analyses assumed the cavities ;

were Gooded with water. The released corium was calculated to spread to form a layer having a I uniform upper surface over the full cavity. l i

19.11.4.2.2.2.5 Debris Power The debris power has been noted to have considerable impact on the debris coolability, and the concrete erosion rate and penetration profiles. These items are discussed below. I Debris Coolability O

V Based on debris bed heat and simulated corium - concrete attack experiments, the ability to quench corium debris is strongly dependent on the heat production rate within the corium pool (See Section i

i 19.11.4.2.2.2.3 above). Typically, an overlying pool of water at atmospheric conditions has been Approved Design Material _MbaWistic Risk Assessment Pope 19.11-85

System 80+ Design ControlDocument l

l Thermodynamic analysis and experimental evidence indicate that iodine, cesium and the less volatile radionuclides released from the fuel during core damage accidents in LWRs will behave primarily as aerosols. A substantial fraction of these aerosols deposit on RCS surfaces or within water reservoirs, )

especially for sequences with long residence times. !

i Experimental evidence of aerosol RCS retention processes is provided by the LACE and Marviken :

aerosol transport tests and the INEL severe core damage test series and LOFT test FP-2 (Reference i 154). In general, the experiments consistently demonstrated high levels of fission product deposition within the RCS with significant levels of deposition noted within the first few meters of the source.

Based on NUREG-1150 expen judgement elicitation, fission product release from the RV (FVES) was classified for various PWR sequences as shown in Table 19.11.4.3.2-3. The expert elicitation indicated that high pressure sequences were estimated to have nearly complete fission product retention (low releases to containment) while for low pressure sequences about 50% of the fission products released from the core were retained in the RCS. Table 19.11.4.3.2-3 is consistent with the summary of the fission product transmission / retention in the RCS was provided by Reference 195.

The complete isotopic distribution of fission product transmission characteristics may be found in Reference 195. The FVES factors defined in Table 19.11.4.3.2-3 are considered applicable to the S stem 80+ evolutionary PWR and were consequently maintained in S80SOR.

19.11.4.3.2.1.2.2 Application to MAAP Calculations /Modeling The MAAP code models "in vessel" fission product transport in subroutine FPTRNP. Details of FPTRNP can be found in Reference 203.

19.11.4.3.2.1.3 Fission Products Released During HPME 19.11.4.3.2.1.3.1 Applicable to S80SOR HPME occurs when the reactor vessel lower head fails while the RCS is at high pressure. In past PRAs for existing PWRs, HPME releases have been credited for rapidly introducing' considerable quantities of fission products directly into the containment upper atmosphere./The estimated quantification for this process has been presented in Table 5.17 of Reference 19). Discharges of volatiles (noble gases, cesium and iodines) from the reactor vessel are assumed to be similar for NUREG-il50 Reference Plants and System 80+. However, since the System 80+ includes a debris retentive cavity, HPME discharges for this design are expected to have lower dispersal fractions than for the other plants. Since detailed information on fission product distribution for System 80+ is not available, the HPME fission products released for the Te, Sr , Ba, Ru, La and Ce groups were conservatively applied to the System 80+ source term. The HPME discharge fraction (FDCH) for S80SOR are presented in Table 19.11.4.3.2-4.

19.11.4.3.2.1.3.2 Application to MAAP Calculations /Modeling The role of DCH induced containment failure is negligible for the System 80+ PWR as discussed in Section 19.11.4.1. Consequently, MAAP analyses investigating the System 80+ post-VB aerosol content following HPME events were not performed. All release class assessments of fission product releases from containment following a DCH induced containment failure are established solely via the S80SOR methodology.

Apowomi Design Metenel- 94obabWstic flish Assessment f*sge 19.11-94

System 80+ Design ConkolDocument l

.l

, V 19.11.4.3.2.2.2 Pool Scrubbing Processes (v) I.

Water pools provide an excellent means for scrubbing fission products. System 80+ utilizes this scrubbing concept in two facets of the design. First, all fission product releases discharged into the IRWST will be scrubbed by either a subcooled or saturated water pool which provides an average scrubbing depth of about 6 feet. Second, if the CFS has been actuated prior to VB all fission products released to the reactor cavity will be scrubbed in a water pool approximately 15 feet in depth. The effective decontamination factors (DFs) for pool scrubbing of overlying water pool are based on the work of Powers (Reference 197). The fission product scrubbing correlations developed by Powers is provided in Reference 197 (See Figure 19.11.4.3 1). Scrubbing of fission products within the IRWST was considered to be similar to the scrubbi associated with the Grand Gulf Suppression pool "Downcomer Vents" (see Reference 200). For ervatism it was assumed that al scrubbing occurred as a result of large bubble releases in saturate water pools instead of the more effective subcooled water pools. This assumption provided a conse ative median pool scrubbing DF applied to IRWST discharges of 6.8.

19.11.4.3.2.3 Significance of Natural Deposition Processes

' A review of LACE (LWR Aerosol Containment Experiments) experiments LA2, LA4 and LA6 has shown that removal of aetosol particles from containment via natural mechanisms in steam atmospheres can be significant (Reference 199). During these three tests aerosols were injected into a preheated steam saturated atmosphere with either pre <xisting or delayed leaks. By comparing the actual leakage to that would be expected without natural deposition mechanisms considered, it was O found that natural settling had accounted for a reduction of aerosol leakage by between 5 and 30.

d Specific values were dependent on the status and timing of the leak. It was concluded that for low leakage steel containments over a time period of 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months , natural aerosol removal process can reduce the total fission product leakage by more than a factor of 10.

19.11.4.3.2.4 Fission Product Retention and Filtering in the Secondary Containment The amulus ventilation system (AVS) filtration subsystems are used in System 80+ to reduce the radioactive aerosols and iodine released during the various postulated accidents. This system is designed as an Engineered Safety Feature (ESF). The AVS can provide considerable help in controlling fission product releases for certain severe accident scenarios where the containment remains intact. This is expected to involve approximately 90 per cent of the severe accident sequences. For much of the remaining events the ESF can serve as a useful filter for the early containment leakage portion of the event, up until the time of containment failure.

The System 80+ AVS is located in the secondary containment. For conditions where power is available to operate the filtration system, considerable quantities of fission products can be removed from the atmosphere. The AVS includes high efficiency particulate air (HEPA) filters in tandem with a charcoal absorber bed. The HEPA filters are designed to remove a minimum of 99% of the particulates entering the system. The charcoal absorber bed is designed to remove a minimum of 95% of elemental iodine and about 95% of organic iodines from the fission product releases. Ten percent of the annulus in-leakage can bypass the AVS filters. Operation of the annulus filtration and ventilation system will enable approximately 90% of the fission products released from the containment to be filtered. The remainder of the fission products are assumed to bypass these filters.

[V; Filtration through the AVS was not considered tr scenarios where the containment building is assumed to fail or for "V" sequences.

Appmved Design Matenal. Probab3stic Risk Assessment Page 13.1199

System 80+ Deslan ConkelDoeumont

) 19.11.4.4.1.3 Summary of Required Instrumentation and Equipment esassmer >

A summary of the minimum instrumentation and equipment necessary to function during a severe accident, consistent with 10CFR50.34(f), is presented in Tables 19.1149 and 19.11-4-fB Instrumentation that will be useful in helping an operator recover from a severe accident include: 1

1. RCS Temperature Monitoring via either 19,//.M4-l ,

HJTC probe UHJTCs, Hot and Cold Leg RTDs, or ,

~

CETs,

2. RCS Pressure Monitoring via either RCS or Pressurizer Pressure Indicators,

3. SG Water Level indicator,

4. IRWST Water Level Indicator,

5. Containment Pressure Indicator,

6. Containment Temperature Indicator,

7. Containment Hydrogen Concentration Monitor, and

8. Containment Radiation Monitor.

Included in the instrument system are the instrument sensor and its associated cables, terminals and junction boxes.

Items 1 thru 3 represent instruments that are required only for "in-vessel" recovery sequences. These instruments are not required to survive the post-VB containment atmosphere. The IRWST water level, containment temperature and pressure sensors and containment radiation and hydrogen .

monitoring will continue to provide useful information to the operator even after VB and should have ,

a reasonable expectation of survivability in a post-VB environment.

Equipment that should be able to function in the severe accident environments include:

1. Containment penetrations, airlock, batch seals, electrical / mechanical penetrations.

2. Containment Sprays / Spray Header, heat exchanger and Piping and associated valves and heat sinks.

3. SDS valves and Actuation Circuitry of this system is intended to be activated prior to core uncovery.

4. SIS and EFWS including valve position Indicators and/or Flow indicators for water delivery flow paths to RCS, containment and steam generator (for "in-vessel" recovery sequences).

5. Hydrogen Mitigation System, including igniters, IRWST vents, associated cabling, transformers and power sources.

Amed ca en ****erw h sasareia ni n an.um.nr rare ss.ss.sor

Systern 80+ Design ControlDocument

,a

(") alignment plate. The temperature difference between the heated and unheated junction thermocouple pairs is a direct indication of the presence or absence of liquid inventory. As a level monitoring instrument, this instrument provides useful information as the core uncovers :md provides confirmation of core recovery. The individual unheated junction thermocouples may also trend the progression of core degradation by monitoring the gas temperature in the reactor vessel upper plenum.

The HJTC probes utilize heated and unheated junction Type K thermocouples. Unlike the CET, the RVLMS thermocouple string is top mounted and does not pass through the core. In accordance with the RVLMS design requirements, several of these thermocouples are calibrated for operation up to 1800*F. Consequently, these instruments will continue to function far into the core degredation process. While not providing a direct indication of core degradation, the probes provide valuable trending information to the operator. These thermocouples are characterized by the vendor to survive beyond 2000*F. Therefore, many of these thermocouples (particularly those located towards the upper part of the string above the UGSSP) will likely survive the 1600*F "in-vessel" upper / head environment following a recoverable severe accident.

Since the HJTC string is top mounted, its junction boxes, and leads will be routed away from active igniters. Therefore, exposure to high temperature diffusion flames are not expected.

19.11.4.4.1.4.4.2 Instruments / Equipment Interfacing with the RCS Instrument / Equipment residing in this category includes the RCS and pressurizer pressure sensors, the p SG level monitor and the Rapid Depressurization valves. The operating 7ime frames and

( ,/ environments for these instruments are summarized in Tables 19.11.4.411,-2 f 19.11.4.4-5B, respectively.

e RCS / Pressurizer Pressure Sensors / SG Level Monitors For the operator to appropriately utilize the plant's resources he must be able to assess the equipment limitations and operate the cquipment properly and trend consequences of his actions. To this end it is expected that the operator may need an indication of RCS/ pressurizer pressure, and SG level.

Monitoring RCS pressure is necessary in trending the RCS depressurization following ;

operator actions taken to either establish feed and bleed conditions or to confirm the pressure is sufficiently low to enter shutdown cooling.

In the event that the operator must depressurize the RCS via the steam generator, the water level in the SG should be tracked to assure the presence of SG secondary side inventory. To accomplish this task the operator must rely on the SG level monitors. l All pressure transmitting devices are located outside of the RCS boundary with the only direct interface being a long length of small diameter tubing connecting the RCS to the high pressure side of the pressure transmitter. The sensor tap is typically filled with low velocity fluid.

This length of pipe provides sufficient heat loss and thermal capacitance to maintain the fluid temperature in the vicinity of the sensor to acceptable levels. Therefore, the "in-vessel" environment will not significantly influence instrument operation.

b Appmvod Desigor Matenal- Probehitstic Risk Assessment Page 19.11119

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l System 80+ oesign controlDocument n

Concrete attack is predicted to be monotonically increasing throughout the event and reaches 100

(]

inches of basemat at the analysis end time of 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months (Figure 19.11.5.4.2.2-3). Extrapolating the erosion predictions suggests a basemat melt-through into the extended foundation will occur in about 8 days.

19.11.5.4.2.2 3 Fission Product Releases A summary of fission product releases is presented in Table 19.11.5.4.2.2-3.

19.11.5.4 3 Small Break LOCA The accident initiator is a small LOCA (0.02 ft 2) coupled with the unavailability of safety injection.

Containment sprays as well as auxiliary feedwater are assumed to be available. The cavity flooding system (CFS) is operational to reflood the core in the first transient analyzed but is unavailable in the second. These sequences are designated as SL-llE and llF.

19.11.5.4 3.1 Small Break LOCA with Wet Casity This sequence is initiated by a small (0.02 ft ) cold leg break. In this scenario, the operator actuates the cavity flood system prior to vessel breach. less of cooling water inventory drops the primary system pressure to the low pressurizer pressure reactor trip setpoint and scrams the reactor at 19.7 seconds. Main feedwater is automatically run back upon reactor trip to match the reactor power.

Emergency core cooling water is unavailable, creating rapid boil-off of coolant inventory resulting in

( core uncovery at 1.06 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . Zirconium oxidation begins shortly thereafter and rises dramatically at

(]/ 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months resulting in 1718 pounds of hydrogen produced in the vessel prior to vessel breach. The fuel heats up and slumps into the lower head, with reactor vessel failure occurring at 3.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . A summary of event timings is shown in Table 19.11.5.4.3.1-1 and selected parameter values are provided in Table 19.11.5.4.3.1-2.

19.11.5.4 3.1.1 Primary and Secondary System Response After the initiation of the event, the system depressurizes to the SIT pressure setpoint. Here the introdue: ion of SIT water helps the core to maintain pressure (Figure 19.11.5.43.1-1) at 500 psia and two phase level (Figure 19.11.5.4.3.1-2) at around 13.5 ft. Initial core uncovery is at 1.06 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

The core is partially recovered by SIT injection.

Core uncovery is followed by rapid zircaloy-water oxidation. The fuel is predicted to rapidly heat up and melt. Support plate failure is predicted to occur at 3.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and RV lower head faMure occurs shortly thereafter.

The secondary side of the steam generators temain full of liquid inventory (Figures 19.11.5.4.3.1-3 and -4). The SG associated with the broken RCS loppgles between the secondary side valve setpoint of 1200 psia and 900 psia (Figure 19.11.5.4.3 whre'the other SG cycles in a very tight band about the relief valve setpoint (Figure 19.11.5.4.3. -6). ._.,

19.11.5.4.3.1.2 Containment Response O

U Containment sprays are actuated shortly after initiation of the transient when the containment pressure reaches the actuation setpoint. The IRWST initially is observed to drop in level as CFS is actuated and the IRWST water is sent to the holdup volume and reactor cavitf. The IRWST level Appmved Design hinterial . Probabrastic Risk Assessmarrt Page 19.11 141

System 80+ Design ControlDocument Concrete attack is predicted to be monotonically increasing throughout the event and reaches approximately 90 inches of basemat at the analysis endtime of 80 hours9.259259e-4 days 0.0222 hours 1.322751e-4 weeks 3.044e-5 months ,(FW l.11.5.4.4.2-2). "

Extrapolating the erosion predictions suggests a basemat melt-through into the extended foundation will occurs in approximately 8 days.

19.11.5.4.4.2.3 Fission Product Releases A summary of fission product releases is presented in Table 19.11.5.4.4.2-3.

19.11.5.4.5 Steam Generator Tube Rupture with Stuck Open MSSV In this sequence a steam generator tube rupture (SGTR) event is analyzed with a failure of RCS coolant make-up systems, except for the safety injection tanks (SITS). The rupture of two tubes in one generator is modeled. Charging and safety injection flows are assumed to be unavailable with no main and auxiliary feedwater to either steam generators. The containment spray system is assumed to be available during the transient for containment pressure and temperature control, t 19.11.5.4.5.1 Primary and Secondary System Response Due to the RCS pressure decrease caused by the tube ruptures, the reactor trips on low pressurizer pressure. The rupture of the tubes coupled with no RCS make-up flow results in a rapid depressurization of the primary system (See Figure 19.11.5.4.5.1-1), with a concomitant decrease in primary system inventory (see Figure 19.11.5.4.5.1-2). As the RCS depressurization continues beyond 600 psia, the SITS start injecting fluid into the primary side. This causes a temporary decrease in RCS mass reduction. Subsequent to emptying of the SITS, the RCS inventory steadily decreases due to termination of SIT flow and continued break flow.

As a result of flow out from the SGs via the stuck open MSSVs, both SGs dry out quickly (at about 9 to 12 minutes) as seen from Figure 19.ll.5 A.5.1-3. The dryout of the SGs results in the loss of the normal heat sink for RCS heat removal. Consequently the RCS beats up and begins to rapidly pressurize (see Figure 19.11.5.4.5.1 1). This pressurization causes the PSVs to open up and relieve RCS mass and energy into the IRWST. The PSVs cycle open and close to remove the decay heat.

The continued operation of the PSVs coupled with primary to secondary break flow depletes the RCS inventory as seen from Figure 19.11.5.4.5.1-2. Subsequently the fuel cladding heats up and hydrogen is generated within the core due to zirconium-water chemical reaction. The fuel rod heat-up results in core damage ari ultimately in the absence of RCS coolant make-up, the reactor vessel fails.

Prior to vessel failure, the operatoi activates the cavity flooding system.

A summary of MAAP predicted key event timings and a summary of key transient parameters are provided in Tables 19.11.5.4.5.1-1 and 19.11.5.4.5.1-2, respectively.

19.11.5.4.5.2 Containment Performance The PSV discharge is ducted via the pressurizer pressure relief piping into the IRWST. The discharge of steam into the IRWST heats up the IRWST water. Subsequent to vessel breach the

, containment atmosphere heats up and is pressurized due to the release of steam and corium from the reactor vessel (see Figure 19.11.5.4.5.1-4). The containment continues to be pressurized due to generation of steam within the cavity. This pressurization is terminated at about 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months when containment sprays are actuated on high containment pressure. Subsequently the containment pressure l and temperature decrease to their initial value (see Figures 19.11.5.4.5.14 and 19.11.5.4.5.1-5).

l Anoroved Design Afsterial- ProbeMstic Risk Assessment Page 19.11144 I

i System 80+ Desima consolDocument Figure 19.11.5.4.5.14 shows that the cavity basemat erosion is insignificant (less than I lach) during this transient. This is due to adequate quenching of the core debris in the reactor cavity.

{

19.11.5.4.5.3 Fission Product Releases A summary of fission product group concentrations in the containment atmosphere at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after vessel breach is provided in Table 19.11.5.4.5.1-3 19.11.5.4.6 V Sequence :

The dominant System,80+ V Sequence consists of an intersystem LOCA (ISL) initiated from a full l shear break in the 16f diameter SCS line occurring within the containment building subsphere. This j event is identified in the PRA as PDS 17. .

In this event all ECCS systems are operable. The failure of the SUS pipe outside of containment results in a gradual transfer of ECCS inventory from the containment to the subsphere. This ultimately results in failure of the ECCS function due to the unavailability of a water source. Details of this transient are discussed below. _

19.11.5.4.6.1 RCS Response Characteristics The ISL represents a large LOCA initiated outside of containment. Consequently the RCS response is similar to that of the large LOCA discussed in Section 19.11.5.4.2. In this case the SCS line break is equivalent to 1.4 square feet. The larger failure area results in a more rapid RCS response. In this

( event the core initially uncovers in 76 seconds (See for example Figure 19.11.5.4.6.1-/ and Table 19.11.5.4.6.{). SIT discharge rapidly temporarily recovers the core.

De ECCS maintains the RCS covered until the IRWST is depleted and suction is lost, to ECCS pumps. A second sustained core uncovery begins at 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . Support plate failure occurs at 13,800 ,

seconds and RV failure is predicted to occur shortly thereafter.

He large failure area results in a rapid system depressurization to near atmospheric pressure which is sustained for the duration of the transient (Figure 19.11.5.4.6.1-/).

M4Q y (19.11.5.4.6.2

'"i :

Containment Response Characteristics j The ISLOCA releases all the RCS and containment liquid inventory into the building subsphere.

Once the RV fails the corium is assumed to fully drop into the dry reactor cavity. Core concrete attack begins immediately. Concrete erosion will ultimately lead to a basemat failure. However, the !

bypass pathway provides a more direct means for releasing fission products to the environment. ;

These MAAP analyses do not credit the water accumulation expected in the subsphere ECCS rooms to scrub fission products leaving the RCS. Furthermore, detailed revolatilization models including the large length of SCS piping are likewise not considered in this demonstration.

y 19.11.5.4.6.3 Fission Product Releases MAAP predicted fission product releases for the V sequence are summarized in Table 19.I1.5.4.6.

A .dge4mAq kff b n$t UNt !! A- $W fk b N'U* ~1 Anwe+ent Dentyre Meteriel- PrebeMineic Stiek Asseesneerst Page 19.11146

l System 80+ Desten controlDocument Table 19.11.3.1-1 Axl ymmetric Ultimate Stress Pressure Values Temperature Yield Stress Pressure (psia)

(* F) 290 mmimum 157 mean 172 maximum 187 350 minimum 153 mean 168 rnaximum 182 450 mmimum 147 mean 160 maximum 174 O

O' AIMweved Design Meterial- PnabebiEsee Risk Assessment p ,y , y g yy,y a

Syster180 + oesten controlDocument m

) Table 19.11.4.3.2-4 Mean and' Median Values of Fraction of Fission Products Species Product Pmsent in the Melt Participating in HPME that is Released .

to Containment in a Dimet Containment Heating Event (FDCH) !

(Taken from Reference 194)

RCS tT)CHA Pressure!31 at Vessel NG I Cs Te Sr Ba Ru La Ce BreaG H 1.0 0.094 0.94 0.025 0.007 0.009 0.015 0.006 0.003 (1.0) (0.80) (0.80) (0.16) (0.06) (0.07) (0.07) (0.02) (0.02) 1 1.0 0.94 0.94 0.016 0.003 0.006 0.01 0.004 0.004 (1.0) (0.80) (0.80) (0.16) (0.05) (0.07) (0.06) (0.02) (0.02)

Ill H & 1 refer to high (>2000fsig) andptermediate (< 1300[sig) RCS[ressure, respectively.

A Mean values are presented in parenthesis.

j Table 19.11.A3:fr Mean and Median Values for the Fractions of Radionuclide Gmup I M2-5 Released During Core-Concate Interaction (FCCI) for P% (Taken from Reference 194) 96 5 1 4 ,

~-

FCCIA Zirconium Cavity Content in the ConditionA Mettm I Cs Te Sr Ba Ru La Ce D H 1.0 0.56 0.05 0.04 2x108 8x104 1x10-3 (0.52) (0.15) (0.13) (0.004) (0.015) (0.02)

D L 1.0 0.5 0.05 0.03 2x10s 7xgoa 9x104 (0.45) (0.13) (0.11) (0.004) (0.015) (0.01)

W H 1.0 0.24 0.02 0.02 3x104 4x104 4.5x104 (0.30) (0.11) (0.10) (0.002) (0.002) (0.01)

W L 1.0 0.23 0.009 0.01 3x104 3x104 4x10 4 (0.28) (0.09) (0.07) (0.002) (0.002) (0.01) til Mean values are presented in parenthesis.

m D & W refer to dry and wet cavity respectively.

I'l H & L refer to high and low Zirconium content in the melt.

Anwoved Design Meterial ProbaWEstic Risk Assessment Page 19.11 165

4 , i System 80+ Design contrat Document l g.S-M Table 19.1133-6 Mean and Median Values for the Fraction of Radionuclide Group I Retained in RCS Released into Containment After Vessel Failure l

Conditions FREVI33 I

I Cs Te One opening after vessel breach 0.04 0.02 0.

(0.11) (0.05) (0.04)

Two openings after vessel breach 0.13 0.095 0.

(0.22) (0.20) (0.12) 01 The mean values are shown in parenthesis.

Table 19.11.4.4-1 Minimum List Of System 80+ Instrumentation Required For Severe Accident Mitigation And Recovery Required Required Instnunent Pre-Vessel BreachUl Post Vessel Breach UlUTCr2) Yes No RCS Pressure or PZR Pressure Yes No SI Flow Yes No EFW Flow Yes No SG Water level Yes No IRSWT Water level Yes Yes Hydrogen Monitors Yes Yes Radiation Monitor Yes Yes Cont. Pressure Yes Yes Cont. Temperature Yes Yes CS Flow Yes Yes l

l Ill Instruments required in this column are used to achieve a safe plant shutdown as per 10CFR50.34(f).

[21 Functionability required for thermocouples located in upper guide structure only.

Alyweved Design Afsterial . Probabrisoe Risk Assessment Pope 19.11166 l

l

System 80+ Deslan ControlDocument o -

(j Table 19.11.4.4-2 Minimum IJst of System 80+ Equipment Required For Sevem{WT]

Accident Mitigation And Recovery Operation Required Operation Required Systan Pre-VB Post VB Safety injection (SI) Yes No Emergency Feedwater System (EFW) Yes No _

Containment Isolation Yes / tio k Safety Depressurization System Yes 'I Cavity Flooding System Yes No i Hydrogen Mitigation System (Igniters) Yes Yes Containment Penetration lategrity Yes Yes Containment Spray (CS) Yes Yes Shutdown Cooling System (SCS) Yes Yes Table 19.11.4.4-3 Maximum "In-Vessel" Pressure / Temperature Conditions Prior to VB Tanperature Pressure ( )I RCS Iacation C ) C. ) Comments Upper / Head (above < 1600'F <2500 psi UIUTCS located within this UGSSP)DI region.

Cold legs (suction / < 700 'I' <2500 P i RCS Pressure discharge) ( ,p Pressurizer <700 for LOCAs <2500 @ Pressunzer pressure tap and Rapid "F Depressurization valve interface

< 1200'for Transients with this region.

with PSV cycling or SDS open

[1] Upper Guide Structure Support Plate.

Table 19.11.4.4-4 Maximum Containment Pressum/ Temperature Conditions Prior to VB Transient Containment Temperature l"/) Containment Pressun (M)

'In Containment" Release <300* F I' '

< 75 y A Sequence f Bypass /SGkRehesse Sequence < 250' F < 30 PS UL L/

bT l

ANwored Design Meterial- ProbabiEsalc Risk Ar Asment Pere 1R 11-167

6 d

Svstem 80+ ' Desien ContalDocument i

1 i

i 45.0 1_ l m - ,

37.5 _

t 30.0 _f i

_ i.

22.5 _

. 15.'0 -

_ e 3

7.5 i

' '''' '''' '''' '''' i 0.0 ' ' ' '

0 " ' ' ' 20 40 60 80 100- ;

1 TIME HR !

i

-5 ? i i

j3 SG Water lael vs Time: Plant Accident Sequence SBOBD-E; Plant Figure 19.11.5.4.1.$

Damage State 242; Statios Black =d with Battery Depletion and Wet !

Cavity L dDeste A8sswdal.Mo6aMReele Aish Assessment paye 73. t 7,247 r

I

System 80+ Design Control Document O;

i 1

0 3s4 pteam r+

gr t) p WD

~O Ap9wevent Design MetM- ProbaMstic Itisk Asseasmerrt

/ y 7 y,

.e . . . ,

svstem 80+ outer canard cocanart i- ~ Appendix 19.11A .

t

/

/~ Failure in the Presence of a Steam Generator N. Partially Mlled with Liquid Representative Calculations Regarding a System 80+

'lhermally Induced Steam Generator Tube Creep Contents Page F

1.0 Introduction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.1 1 A- 1 2.0 Background . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.llA-1 i

3.0 Estimate of Minimum Water Level . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.11 A-1 4.0 Refere nces . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19.ll A-3 Figures 19.11A-1 Average Wall Temperature Versus Repture Time for Steam Generator Tube (Inconel 600)(Reference A1) ............................ 19.11 A-4 I

1 I

~

l l

Amvenf Deniger nietaniel- ProbeMiraio Alek Assessment Popei

o .. .

System 80+ Design controlDocument requiring IRWST entry. Igniters are generally located 7 to 10 feet above floors to allow easy access without impeding personnel passage or becoming a personnel safety hazard.

7.0 Igniter System Verification This section summarizes the MAAP 4 containment analyses performed to assess the System 80+ HMS.

The MAAP 4 code was used to study hydrogen mixing and combustion in the System 80+ containment.

The study assessed the potential for hydrogen build-up in the containment and to calculate the best-estimate response of the HMS.

MAAP 4" contains a state-of-the-art lumped parameter model for containment thermal-hydraulics. The model was specially constructed to simulate natural circulation in advanced light water reactor containments. Key elements of the model used for the hydrogen calculations for System 80+ are as follows:

e Mechanistic, semi-implicit models for gas, water, and energy transport between control volumes, e models for both unidirectional and counter-current flow @ through containment junctions, e stable treatment of water-solid regions; these can develop in System 80+ calculations if the IRWST pool is sub-nodalized or if the cavity flooding system is activated, e flexible modelling of containment heat sinks, and e

advanced modelling of hydrogen combustion. Both non-global burns initiated by the hydrogen O igniters and global burns are treated using a single, unified framework. This model has been successfully compared to a great variety of experimentsM.

7.1 Construction of Plant Model A detailed (23 control volumes, 35 junctions, and 37 heat sinks) containment model was constructed,b

-has b Tiger !-! rd ' ! 2. Considerable effort was taken to minimize artificial mixing which can be caused by the limitations inherent in lumped parameter containment codes.

Calculations were performed using the number and location of the igniters presented in Table 19.llK-4.

In addition the MAAP 4 simulations employed an IRWST vent area of 200 ft2, A recent International Standard Problem (ISP-29) tested the ability of various lumped parameter codes to predict hydrogen concentrations in the HDR containment during experiment Ell.2. The results indicated that all the codes tended to over-predict mixingN: whereas, very little hydrogen was measured below the elevation at which the hydrogen was injected, the codes predicted substantial mixing.

Part of this tendency to over-predict mixing is a consequence of inherent assumptions used in lumped parameter codes, i.e. the fact that control volumes are assumed to be well-mixed can lead to numerical diffusion. However, these problems can at least be minimized by careful construction of the containment model. For example, the effects of numerical diffusien were reduced in this study by employing a relatively large number of nodes.

Approved Design Materis'. Probabilstic flisk Assessment Pope 19.11K-16

System 80+ Design ControlDocument

The structures housing the Station Service Water (SSW) pumps and the Component Cooling

( " )

Water (CCW) heat exchangers was assumed to be as strong as the interior structure of the nuclear anncX.

Seismically induced gross structural collapse of the nuclear annex interior structure was assumed to result in the failure of all safety equipment and lead directly to core damage.

Seismically induced failure of shield building structure was assumed to result in damage to the control room due to falling concrete. This was assumed to cause a transient, and failure of the responding safety systems if control was not transferred to the remote shutdown panel.

It was assumed that the seismically induced failure of the shield building structure would not directly result in failure of the remote shutdown panel.

It was assumed that the dominant seismic failure mode for the containment shell is vertical rotation / overturning. This failure mode was assumed to result in failure of all safety equipment and breach of containment integrity.

lt was conservatively assumed that seismically induced failure of the supports for any major RCS component, such as the RCPs or the steam generators, would result in failure of all RCS piping attached to the component due to excess motion of that component. Given the one-fail-all-fail assumption, it was further assumed that all like components, i.e. RCPs, would fail at the same time. In addition, it was conservatively assumed that as a direct result of the piping failures, the

,q Safety Injection System would be unable to provide sufficient makeup flow for RCS reactivity ls_/ control.

Under the one-fail-all-fail assumption, it was assumed that seismically induced failure of a major PCS pipe would result in failure of all equivalent RCS piping. This would result in a LOCA in Scess of ECCS Capacity.

Consistent with the above two assumptions, it was assumed that there were no seismically induced large LOCA sources.

It was assumed that the occurrence of a seismic event did not automatically result in loss of offsite power or failure of the standby combustion turbine regardless of the "g" level.

The System 80+ Class I ctrical distribution system is provided with protection schemes which conform to the requirements ofIEEE STD-741-1986. The protective schemes are designed to isolate faulted equipment from the rest of the system to minimize the effect of the fault and to maximize the availability of the remaining equipment. The basic schemes consist of ground fault protection, instantaneous overcurrent and timed overcurrent protection. In developing the SMA models, it was assumed that the seismic failure of equipment in the Electrical Distribution System were "open circuit" failures. Implicit within this assumption is the assumption that if a

hot short" failure were to occur, the appropriate circuit interrupter (s) would open on overcurrent

, to prevent " backward" propagation of the fault.

Seismically induced sliding of the nuclear island structure is assumed to result in severing of all

(]

V piping or cabling into the nuclear island. This was conservatively assumed to lead directly to core damage.

Approved Design Afstarial- Pnsbebilstic Risk Assessment Page 19.15-41

~

System 90 + Design ControlDocument

'")

Failure of the standby SCS train for either DHR or feed operation is dominated by railure of control valves and MOVs. An aggressive valve testing and maintenance y ogram on the SCS and g

CSS would reduce shutdown risk. M"

The use of the CVCS to makeup inventory is an important recovery action in reduced inventory ces operation. !! tiso acts as a temporary cooling technique. The operator should have procedures No and training on its use. /M*d

Safety injection in conjunction with bleed is an important means of removing decay heat during shutdown modes. Having two of the four SIS trains available during most shutdown modes is an important new technical specification.

The CSS pumps are used as backup to the SCS pumps. The operator must therefore be properly trained in performing the procedure (s) to align the CSS pumps for operation if the SCS pumps should fail. Again, valve maintenance and testing is important for shutdown risk reduction.

LOCA Insights The major insights for LOCAs during low-power and shutdown modes of operation are listed below:

SCS injection is an important means of makeup following a LOCA. Therefore, the development of appropriate procedures and the training of the operator to perform these procedures are necessary and important in mitigating LOCA evems during low-power and shutdown modes of oper&sion.

n

()

The dominant failure mode for SCS injection is failure of control valves and MOVs. A valve maintenance program is important.

The use of the SIS to provide injection during a LOCA is important. Since manual acation is required, training and procedures are required to properly accomplish this task.

The Chemical and Volume Control System (CVCS) is another important means of makeup following a LOCA and with proper training and procedures this system will most likely be available when required.

For LOCAs located in the containment, the IRWST acts as a sump and makes the coolant available for injection. Procedures are needed to ensure that flow paths are maintained during the outage.

LOOP Insights The insights for loss of offsite power during low-power and shutdown modes of operation are listed below:

The new technical specification for having two of the three standby and emergency generators :

available reduces the CDF. '

The reliability of the two switchyards is important to isk reduction for LOOP. Procedures to

(

v) control maintenance in both these areas at the same time should be considered. l

)

- - o w .,a.,. n .u,, n .u.. . -, e r ,w

\

o 4 e , i l

System 80+ Desian contror Document l 1

7~

l 135. NUREG/CP-0076, "On Two Aspects of Hydrogen Risk," Forestier, A., Goldstein, S.,

" l Proceedings of the Third Workshop on Containment Integrity,1986. '

136. NUREG/CR-5662,(BNL-NUREG-52271), " Hydrogen Combustion Control and Value-Impact Analysis for PWR Dry Containments," Yang, J.W., et. al., May,1991. I 137. HEDL-TME-82-8, "Hydrogea Centrol Systems for Severe Accident Conditions: A State of Technology Report," Hanford Engineering Development Laboratory,1982.

138. NUREG/CR-5275: " FLAME Facility: Effect of Obstacles and Transverse Venting on Flame Acceleration and Transition to Detonation for Hydrogen - Air Mixtures at Large Scale," Sandia, April 1989.

P 139. "The Possibility of Local Detonations During Degraded Core Accidents in the Belleonte Nuclear.

Power Plant," Sherman, M., Berman, M., Nuclear Technology, Vol. 81, April 198{ _

140. Ther-Reference 4atenticd!y B!ri. n " l> c 9laareke n "b ch n= k . . . "

141. CENPD-141, " Reactor Plant Subcompartment Analysis," Combustion Engineering, Inc., March, 1978.

142. NUREG/CR-4551, Vol. 3, Rev.1, Part 1, " Evaluation of Severe Accident Risks: Surry Unit 1,"

Sandia National Laboratory, October,1990.

(D w/ 143. EPRI NP-2953, " Hydrogen Combustion and Control Studies in Intermediate Scale," Torok, R.,

et. al., June 1983.

144. NUREG/L " '/Q, " Experiments to Investigate the Effect of Flight Path on Direct Containment Heating in the Surtsey Test Facility," Allen, M.D., Pilch, M., October,1991.

145. CE-NPSD-74-P, " Evaluation of Design Features which Minimize the Probability of Interfacing System LOCAs for System 80+ Standard Design," Combustion Engineering, Inc., May,1992.

146. DCTR-RS-02,Rev 0, Probabilistic Risk Assessment for the System 80+ Standard Design,"

Combustion Engineering, Inc., January,1991.

147. This Reference Intentionally Blank.

148. EPRI-NP-5127, Spencer, B. W., " Hydrodynamic' 4 Her.t Transfer Aspects of Corium Water Interactions,* March,1987.

149. DOE /ID 10290, " Technical support for the Hydrogen Control Requirement for the EPRI Advanced Light Water Reactor Requirements Document," January,1990.

150. DOE /ID-10278, " Technical Support for the Debris Coolability Requirements for Advanced Light Water Reactrs in the Utility /EPRI Light Water Reactor Requirements Document," June,1990.

151. Bjornsson, H., et. al., "Penetr:Ition of Water inte Hot Rock Boundary of Magna at Grimsvoth,"

Nature, Vol. 295, February 18, 1982.

Approved Design Meterial- Probabinstic flirk Assessmerrt Page 19.16-7 j

q ggstem 80+ oeska coew occament 1

' Appendix 19A '

Design Alternatives i i

t I

This Appendix Intentionally Blank l

f r

4 i

I I

E I

f l

f t

I

h 8 f ,

o 4 , .

System 80+ Design ControlDocument OPERATOR ACTIONS REACTOR TRIP RECOVERY Instructions Contingency Actions

1. d Trip Actions 1.

are=rc per';iober

__ ed.

2. Con irm iiiagiiosis of uncomplicated 2. Rediwnose event and exit to Reactor Trip by verifying Safety appropriate Optimal Recovery Function Status Check acceptance Guideline or to Functional Recovery criteria are satisfied. Guideline.

3. Verify pressurizer level is: 3. Manually coerate PLCS or charging and letdown to restore and maint:dn

a. [2% to 78%l pressurizer level [33% to 52%].

and

b. trending to 133% to 52%]

4. Verify pressurizer pressure is: 4. Manually operate PPCS or pressurizer heaters and spray to control RCS pressure:

a. [2160 to 2370 psia] a. [2225 to 2300 psia]

(

and and

b. trending to l2225 to 2300 psia] b. wititin the Post Accident P-T limits of Figure 4-1.

and

c. within the Post Accident P-T limits of Figure 4-1.

5. Verify steam bypass control system is 5. If condenser vacuum is lost, steam control'ing RCS T,, l551-562*F]. bypass control system is unavailable, or the MSIVs are closed,lhen use the atmospheric dump valves to control RCS T , [551-562*F].

6. Ensure at least one steam generator has 6.

level being maintained or restored in the normal band using main, startup or emergency feedwater.

l

Step Performed Continuously O'

ADM Emergency Operations Gddeines Pege 4-2

.. ..y System 80+ Design controlDocument

. O N/

1 P YES N NO COOLDOWN REQUIRED 7 STEP #19 RCS BORATION 3g MAINTAIN PLANT STABILIZED STEP #21 STEP s20 RCS COOLDOWN STEP 822 >

1P RCS PRESSURF CONTROL STEP #23 1 P RCS INVENTORY CONTROL STEP 824 1 P CONOENSATE NVENTORY STEP #2S q 7 g BYPASS ESFAS SETPOWTS STEP 826 1 P ISOLATE, ANT OR DRAIN S!Ts '

/

h [ STEP #2/ 7 , 28 INITIATE LTOP STL' #29 1P' SCS EMTRY NO YES CONDITIONS SATtSFIEO?

1 r STEP #30 INITIATE SCS OPERATION MONITOR FOR VODING

$NHerT DEPRESSURIZATDN)

STEP #30 STEP #30 ELIMINATE RCS VOIDS STEP S30 Strategy Chart for Loss of OEsite Power Figure 9-12c ADM Eme pency Operations GuideEnos Page SSO

. . . , e System 80+ Desian control Document 41E i e i a i , i i a i i a a a 5 . .

3.5 10 P

B B

D .

5 W 3 10 . .

A B

Time after trip when =0 hoes

, Feedwater is started 2.51/ - 4 -

a E

Q U

1

21/ - tehoes -

B D.

5 G

L 1.5 10 - -

/ ,

5 ~

v 1 10 -

Basis:

Secondary Pressure = 1100 psia j(,D ' Feedwater Temperaturr; = 1200F-N 5 10" - -

1 b 0)

M , , , , , , , , , , , , , , ,

b 0 2 4 6 8 10 12 14 16 18 20 22 24 26 28 30 32 l TDG 0===) From start of Feedwater Typical Feedwater Required for Sensible Heat Rernoval Figure 10-4 I

ADM Emergency operatione Gandonnes Page 10-16

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