ML20214G440

Requests Assistance in Review of Ability of Mark II Containment to Accomodate Suppression Pool Hydrodynamics for Listed Plants

ML20214G440
Person / Time
Site:	Nine Mile Point, Susquehanna, Columbia, Limerick, LaSalle, 05000000, Zimmer, Shoreham, Bailly
Issue date:	08/11/1975
From:	Par O, Parr O Office of Nuclear Reactor Regulation
To:	Heineman R Office of Nuclear Reactor Regulation
References
CON-WNP-0008, CON-WNP-8 NUDOCS 8605220133
Download: ML20214G440 (7)

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Docket File LWR-3 File F. Williams S. Miner 4 '

M. Rushbrook

. .. , . , , , .., R. E. Heineman, Director, Division hf hs[ ems Safety TECHNICAL -ASSISTAN' s REQUESTS Your assistance is requested for the following:

PLANT NAMES: Shoreham Nuclear Power Station UnitsNo.1, Limerick Generating Station Units No.1 &

. , , , . . . . . . . . No. 2, William H. Zimer Nuclear Power Station, Unit 1. Bailly Generating Station Nuclear 1, LaSalle County Station Units No.

1 and No. 2. Susquehanna Steam Electric Station Units No. I and No. 2, WPPSS Nuclear Project No. 2, Nine Mile Point Nuclear Station, Unit No. 2. -

DOCKET NUMBERS: 50-322, 50-352 & 50-353, 50-358, 50-367, 50-373 & 50-374, 50-387 & 50-388, 50-397,

)' 50-410 ,

RESPONSIBLE BRANCH: Light Water Reactor Branch No. 3

, , CONTACT: Sydney Miner (x27243)

TECilNICAL REVIEW BRANCHES: Containment Systens Branch , Structural EdgineeridgtBrinchrifiecfianic31 Engineering 1

Branch TARGET DATE FOR C0fiPLETION: End 1FourlttitQuarted il977

, DESCRIPTION OF REQUEST: Background information and documentation provided for the review are described in

. Enclosure 2.

. The review of the ability"of the Mark !!

containment to acconanodate suppmssion pool Hydrodynamics will consist of:

1. A generic review of the DFFR and the Mark II containment support program.

2. A revic'.1 of' each applicant's DAR either as a separate request for assistance or as part of the OL review.

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. , i AUG 1 1 1976 Request for Assistance

.This request for assistance covers part 1 of the review as outlined above. We mquest that you review:

(a) The Mark II Containment Forcing Function Infonnation Report (DFFR) NED0 21061 and NED 21061P and provide a written report on the adequacy of the analytical methodology, loads margins and load combinations for use in evaluating the ability of the Mark II containments to accommodate the suppressi pool hydrodynamic loads that may occur durin a postulated loss-of-coolant accident and during safety / relief valve discharges.

(b) The " Mark II Containment Support' Program Report," NEDO 21297 and provide a written mport on the adequacy of the support progra to provide confirmation of the analytical

. methodology, loads and margins in the DFFR.

(c)'The results of containment support program (outlined in NED0 21297) to assum that the an'a lytical methodology, loads, margins and load combinations provided in the DFFR are adqquately conservative.

Schedule For items (a) and (b) above,

1. Revised DFFR and responses to previously transmitted questions am due from Mark II owners September 1,1976. <

2. Feeting witii nark II owners Sept.1976.

3. CSB, SAB and MEB provide questions for transmittal to Mark II owners Oct. 4,197:

4. Mark II owners respond to llov. 15, 1976 questions. (Item 2 abwe)

5. Meeting with Mark II owners Dec. 1976.

6. CSB, W an ! "E" provide written evaluati-January 17, 1976.

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11. Heinenan 11 576 Schedule For item (c), above, The schedule for the support program stretcht through the September of 1977 with no exact dates given for completion of some of the activities (enclosure 2). However, the LOCA related activities are scheduled for completion prior to January 17, 1977. Then fore the evaluation of this portion of the support program should be included in your written evaluation of the DFFR.

The safety / relief valve related activities are scheduled to be completed late in September of 1977. . The review and written evaluation for this portion of the program should be completed by the end of the fourth quarter of 1977.

(fE O. L Olan D. Parr, Chief Light Water Reactors Branch No. 3 Division of Pruject Management cc: W. Mcdonald, MIPC R. DeYoung S. Varga H. Berkow l

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ENCLOSURE 1 In April,1975 letters were sent to all' the Mark II owners requesting information relative to the ability of the Mark II containment to acconinodate suppression pool hydrodynamic loads that may occur during a postulated loss-of-coolant accident (LOCA) and during safety / relief valve discharges. In response, the Mark II owners provided a schedule and program for resolving the effects of the LOCA and safety / relief valve hydrodynamic loads. The program consisted of the following major elements:

1. A Dynamic Forcing Functions Report (DFFR) which would provide the basis for the forcing functions on the Mark II containments due to safety / relief valve discharges and loss-of-coolant accident (LOCA).

2. Plant unique reports or Design Assessment Reports (DAR) based on the DFFR.

3. Mark II Supportina Program. ,.

In addition, as a result of our inquiries, a Mark II owners group was fonned primarily to coordinate the activities on the Mark II owners taward developing resolutions to the pool dynamic and safety /mlief valve issues.

On October 24, 1975, the General Electric Company submitted for the Mark II owners group a design basis document designated as the " Mark II Containment Forcing Function Information Report (DFFR) NED0 21061 and NEDE 21061P. The first quarter of 1976 most of the Mark II owners submitted their plant unique reports (DAR's).

On June 9,1976 the General Electric Company submitted for the Mark II owners the " Mark II Containment Supporting Program Report, NED0 21297, which described the test and analysis program for verifying the analytical techniques, and margins in the DFFR. The schedule for submittal of the test and analytical results are also included.

In June 1976 the General Electric Company also submitted for the Mark II owners the following documentation on the results from Phase 1 of the "4 T" tests:

1. " Mark !! Suppression Test Program - Phase I Tests," NEDE and NED0 13442P-01 providing the results of Phase 1 of the "4 T" Pool Swell Test.

2. " Mark II Phase I - 4 T Test Application Memorandum" which briefly compares the results of the 4 T tests (Phase I) with the assumptions and analysis presented and discussed in the DFRR.

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AUG 11 1376 Based on a preliminary review of the DFFR and of the various plant specific reports' we found that there was a considerable variation 14 the manner in which the DFFR was applied by each Mark 11 awner in his Design Assessment Report (DAR). Subsequently, after a series of meetings, the Mark II owners agreed to update the DFFR so that it can be used to a greater extent as generic to all Mark II containments.

The revised DFFR is scheduled for submittal September 1,1976. In addition, questions generated as a result of the DFFR review and meetings with the Mark II owners were transmitted to the Mark II owners with a request that they be responded to by September 1,1976.

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, g ENCLOSURE 2 MARK II SUPPORTING PROGRAM Activity Report Program Activity Type Completion Date A. LOCA Related Activities

1. "4T" Pool Swell Test Test Received Phase I Report (NEDE-13442P-01 & HED013442-01)

Phase II & III Report Test September 1976

2. DFFR Pool Swell Velocity Analysis November 1976 Breakthrough Model Confinnation

3. Impact Tests on Pool Test Received Internal Structures (NEDE-13426P & NEDC-20989-2P)

4. Quality Impact Model Analysis Received *

(NEDE-13426P & NEDC-20989-2P)

B. Safety / Relief Valve Related Activities

1. Relief Valve Pipe Clearing Analysis May 1977 Model for Quencher (Confirmation of Model)

2. Relief Valve Pipe Clearing Analysis Received Model for Ramshead (NED0-20942,NEDE-20942P, NED0-21062 & NEDE-21062P)

3. In-Plant Test of 5/R Test End 4th Quarter 1976 Valve Discharge Loads During Consecutive Actuations (Ramshead)

4. DFFR Relief Valve Analysis Late 2nd Quarter 1977 Consecutive Actuation Transient Analysis Confirmation (Ramshead & Quencher)

5. In-Plant Test of S/R Test September 1977 Valve Discharge Loads During Consecutive Actuations (Quencher)

6. Safety / Relief Valve Steam Analysis End 4th Quarter 1976 Discharge Thermal Mixing

{p,y. (Confirmation of Model)

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AUG 1 I 137f C. Mark 11 Su3 porting Program Report Received Report (NE h-2129i)

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