ML20045D304

Forwards Rev to 930618 Submittal Re Fourth Phase of Revised ABWR Tier 1/ITAAC Matl

ML20045D304
Person / Time
Site:	05200001
Issue date:	06/21/1993
From:	Hackford N GENERAL ELECTRIC CO.
To:	Joshua Wilson Office of Nuclear Reactor Regulation
References
MFN-098-93, MFN-98-93, NUDOCS 9306280197
Download: ML20045D304 (3)

Text

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i To: Jerry Wilson, Section Chief Standardization Projt et Directorate IRC - Mail Stop 11-H-3 From: Nore Hackford/ Tony James GE- Nuclear Energy.

San Jose

Subject:

June 18th ITAAC Submittal - REVISION

Reference:

MFN 098-93, Docket STN 52-001 The ten copies dated 6/18/93 addressed to the Document Control Desk of 'US-NRC sent to you Friday contained the fourth phase of revised ABWR Tier 1/ ITAAC material. These ten copies each have a two-page revision (copies attached).

Please replace pp. 1 and 4 of Section 2.14.1 with the enclosed revised pages.

Thanks.

Norm Hackford' b A i

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! .2.14' Containment and Environmental Control Systems O 2.,4., eri- rv Cent in-ent vs 1 -

Design Description The Primay Containment System (PCS) encompasses:

(1) A reinforced concrete containment structure with an internal steel liner. The structure includes various penetrations, equipment hatches

and personnel access locks.

(2) Structures inside the primary containment which partition the containment into dqwell and wetwell regions, provide equipment <

support, radiation protection, and components for operation of the ABWR pressure suppression containment.

Figure 2.14.1 shows the basic configuration and scope.

The steel lined reinforced concrete containment structure attached to a reinforced concrete basemat provides the primary containment pressure barrier and is classified as AShfE Code Section III, Division 2.The Reactor Pressure Vessel (RPV) support pedestal and a diaphragm floor partition the containment p volume into drywell and wetwel! egions. The RPV support pedestal is a steel d structure with concrete fill material. The diaphragm floor is a reinforced concrete structure. Other major internal structures within the containment are the reactor shield wall, lower dqwell personnel and equipment access tunnels 5 and the dywell equipment and piping support structure. These internal structures are steel fabrications.

1 Penetrations through the containment pressure boundary include; the drywell  !

head closure, equipment hatches to both upper and lower drywell regions, j personnel locks into upper and lowei drywells, a combined personnel access and i equipment hatch into the wetwell and piping and electrical penetrations. These l pressure boundary appurtenances are steel structures classified as AShiE Code {

Section III, Disision 1, Class hic.

The containment design pressure is 3.16 kg/cm 2g. The design temperatures for ,

the dgwell and the wetwell are 171 C and 104 C respectively.The maximum )

calculated pressures and temperatures for the design basis accident are less than these design conditions. The primary containment pressure boundary including penetrations and isolation valves, has a leak rate equal to or less than 0.5% per day (excluding main steamline isolation valves (htSIV) leakage) of the containment gas mass at the maximum calculated containment pressure for the design basis accident.

6/18/93 1 2.14

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j Table 2.14.1 Primary Containment System ,

g Inspections, Tests, Analyses and Acceptance Criteria Design Commitment inspections, Tests, Analyses Acceptance Criteria

1. The basic configuration of the PCS is as 1. Inspections of the as-built system will be 1. The as-built PCS conforms with the basic shown on Figure 2.14.1. conducted. configuration shown on Figure 2.14.1.

2. The ASME Code pressure boundary 2. A structural integrity test (SIT) will be 2. The results of the SIT of the pressure components of the PCS retain their conducted on the pressure boundary boundary components' conform with the integrity under internal pressures that will components of the PCS per ASME Code requirements of the ASME Code.

be experienced during service. requirements.

3. The maximum calculated pressures and 3. Analyses of the design basis accident will 3. The maximum calculated pressures and temperatures for the design basis accident be performed using as-built PCS data. temperatures are less than design are less than design conditions. conditions

4. The primary containment pressure 4. An integrated leak rate test of the as-built 4. The primary containment pressure boundary including penetrations and primary containment will be conducted. boundary including penetrations and isolation valves has a leak rate equal to or isolation valves has a leak rate equal to or less than 0.5% per day (excluding MSIV less than 0.5% per day (excluding MSIV

. leakage) of the containment gas mass at leakage) of the containment gas mass at t the maximum calculated containment the maximum ca:culated containment pressure fcr the design basis accident. pressure for the design basis accident.

5. The design differential pressure of the 5. An SIT will be conducted of the as-built 5. An SIT report exists concluding that the as-diaphragm floor between the drywell and diaphragm floor with the drywell pressure built diaphragm floor is able to withstand wetwell is 1.76 kg/cm2 in the downward greater than wetwell pressures by 1.15 the design differential pressure.

direction. times the design differential pressure

6. The water volume in the suppression pool 6. Analyses of the as-built PCS will be 6. The water volume in the suppression pool including the vents is equal to or greater performed. including the vents is equal to or greater than 3580 m3. than 3580 m 3

7. The horizontal center line of the SRVDL 7. Irupections of the as-built SRVDL 7. The horizontal center line of the SRVDL quencher arms are located at or below the quenchers will be conducted. quencher arms are located at or below the elevation of the center row of horizontal elevation of the center row of horizontal vents in the suppression pool. vents in the suppression pool.

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