ML20235E732: Difference between revisions

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February 25, 1972 1
February 25, 1972 1
                          /
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ACRS Members C0!frAINMENT PRESSURE CALCULATION CAPABILITY Attached are four items for background information relative to the subject of an independent capability by the Regulatory Staff to calculate containment pressure. This matter is scheduisd for discussion at the March 1972 ACRS Heating.
ACRS Members C0!frAINMENT PRESSURE CALCULATION CAPABILITY Attached are four items for background information relative to the subject of an independent capability by the Regulatory Staff to calculate containment pressure. This matter is scheduisd for discussion at the March 1972 ACRS Heating.
M. 2. Gaske Assistant to Executive Secretary Attachments:                                           ,
M. 2. Gaske Assistant to Executive Secretary Attachments:
: 1.     REG:RSR-72
1.
: 2.     DRD6T Ltr.,/2/3/71
REG:RSR-72 2.
: 3.     REC:RSR-74, 1/31/72
DRD6T Ltr.,/2/3/71 3.
: 4.     AJN Ltr., 1/23/71 w/ trip                                                                     ,
REC:RSR-74, 1/31/72 4.
report
AJN Ltr., 1/23/71 w/ trip report
                                                                                                                                                            )
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                # **                                  UNITED STATES
UNITED STATES
                      - /%                 ATOMIC ENERGY COMMISSION WASHINGTON, D.C. 20545
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(     ,, ,, &                                                      RECENEC November 24, 1971 iW2 FEB 25 #4 11 3!i U.S. AIDHN ttunut COMM.
ATOMIC ENERGY COMMISSION WASHINGTON, D.C. 20545
(
RECENEC November 24, 1971 iW2 FEB 25 #4 11 3!i U.S. AIDHN ttunut COMM.
ADVISORY COMMITTEE ON P.EACTOR SAFEGUARDS Milton Shaw, Director, Division of Reactor Development and Technology CONTAINMENT RESPONSE ANALYTICAL MODEL DEVELOPMENT - REG:RSR-72 The regulatory staff has immediate need for increased analytical capabili-tiei regarding containment pressure-temperature response analysis. At this tine, the staff is limited to making independent analyses of the peak pres-rar following a LOCA in dry PWR containments and BWR pressure suppression attainments.
ADVISORY COMMITTEE ON P.EACTOR SAFEGUARDS Milton Shaw, Director, Division of Reactor Development and Technology CONTAINMENT RESPONSE ANALYTICAL MODEL DEVELOPMENT - REG:RSR-72 The regulatory staff has immediate need for increased analytical capabili-tiei regarding containment pressure-temperature response analysis. At this tine, the staff is limited to making independent analyses of the peak pres-rar following a LOCA in dry PWR containments and BWR pressure suppression attainments.
In order to adequately evaluate recent applications for certain commercial nuclear power plants, we require additional analytical capabilities for con-telnment response analyses in the following specific areas, starting with the most urgently needed items.
In order to adequately evaluate recent applications for certain commercial nuclear power plants, we require additional analytical capabilities for con-telnment response analyses in the following specific areas, starting with the most urgently needed items.
: 1. Modification of CONTEMPT-PS to more correctly handle the long term pressure-temperature transients in a containment following a LOCA.
1.
: 2. Development of an analytical code to calculate pressure build-up in reactor cavities and other enclosures.
Modification of CONTEMPT-PS to more correctly handle the long term pressure-temperature transients in a containment following a LOCA.
: 3. Deveicpaent of a code to calculate the pressure-temperature transients in ice condenser containments.
2.
: 4. Modification of CONTEMPT-PS to include bypass flow between the (rywell and wetwell.
Development of an analytical code to calculate pressure build-up in reactor cavities and other enclosures.
: 5. Modification of CONTEMPT-PS to handle pressure transients in the air annulus in dual containments.
3.
As the above needs are evaluated, you may determine that some of these tasks can best be performed under the ANC Regulatory Technical Assistance Program even though the tasks are of a code development nature.             Should this be the i
Deveicpaent of a code to calculate the pressure-temperature transients in ice condenser containments.
4.
Modification of CONTEMPT-PS to include bypass flow between the (rywell and wetwell.
5.
Modification of CONTEMPT-PS to handle pressure transients in the air annulus in dual containments.
As the above needs are evaluated, you may determine that some of these tasks can best be performed under the ANC Regulatory Technical Assistance Program even though the tasks are of a code development nature.
Should this be the i
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                                        .    .                                              o 1
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                                                                                                                            \
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case, we would have no objections since, as wasy.pointed coupled to analytical development efforts.
We would be happy to discuss these needs with your staff and provide additi specific details as may be required.
We   would specific      be happy details  as may to bediscussrequired.these needs with your staff and provide additi
b' Edson G. Case, Director Division of Reactor Standards Members of Steering Committee cc:
                                                                                        ,  b' Edson G. Case, Director Division of Reactor Standards cc:  Members of Steering Committee on Reactor Safety Research                                                                     ;
on Reactor Safety Research G. M. Kavanagh C. K. Beck M. B. Biles P. A. Morris H. D. Bruner.
G. M. Kavanagh C. K. Beck M. B. Biles P. A. Morris H. D. Bruner.                                                                                 ,
A. J. Pressesky W. H. Beach, OCM G. O. Bright, WRSPO R. F. Fraley, ACRS (18) i
A. J. Pressesky W. H. Beach, OCM G. O. Bright, WRSPO R. F. Fraley, ACRS (18) i
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Latest revision as of 05:50, 3 December 2024

Forwards Background Info Relative to Subj of Independent Capability by Regulatory Staff to Calculate Containment Pressure.Subj Scheduled for Discussion at ACRS Mar 1972 Meeting
ML20235E732
Person / Time
Issue date: 02/25/1972
From: Gaske M
Advisory Committee on Reactor Safeguards
To:
Advisory Committee on Reactor Safeguards
Shared Package
ML20234E460 List: ... further results
References
FOIA-87-40 ACRS-GENERAL, NUDOCS 8707130066
Download: ML20235E732 (1)
v • d • e


Text

_ _-___-__ _-__

-6 9

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l CU C,,, _;J't! ' ~.g

a S. ca j b

February 25, 1972 1

/

ACRS Members C0!frAINMENT PRESSURE CALCULATION CAPABILITY Attached are four items for background information relative to the subject of an independent capability by the Regulatory Staff to calculate containment pressure. This matter is scheduisd for discussion at the March 1972 ACRS Heating.

M. 2. Gaske Assistant to Executive Secretary Attachments:

1.

REG:RSR-72 2.

DRD6T Ltr.,/2/3/71 3.

REC:RSR-74, 1/31/72 4.

AJN Ltr., 1/23/71 w/ trip report

)

J 8707130066 870623

![\\,hspyS$4TP PDR FOIA THOMASB7-4O PDR

. g, b>a RD 14-2 ACRS omer >

SURNAME >

.....kO.8..ke,/,Cah

..-.[.2,5.[,7,2.,,

2 DATr >

5 Mie tRev.9-53) AECM 0240

n. s, aovraxurrr enerroso omer, me o cos-s.w

1 O

o 4

UNITED STATES

- /%

ATOMIC ENERGY COMMISSION WASHINGTON, D.C. 20545

(

RECENEC November 24, 1971 iW2 FEB 25 #4 11 3!i U.S. AIDHN ttunut COMM.

ADVISORY COMMITTEE ON P.EACTOR SAFEGUARDS Milton Shaw, Director, Division of Reactor Development and Technology CONTAINMENT RESPONSE ANALYTICAL MODEL DEVELOPMENT - REG:RSR-72 The regulatory staff has immediate need for increased analytical capabili-tiei regarding containment pressure-temperature response analysis. At this tine, the staff is limited to making independent analyses of the peak pres-rar following a LOCA in dry PWR containments and BWR pressure suppression attainments.

In order to adequately evaluate recent applications for certain commercial nuclear power plants, we require additional analytical capabilities for con-telnment response analyses in the following specific areas, starting with the most urgently needed items.

1.

Modification of CONTEMPT-PS to more correctly handle the long term pressure-temperature transients in a containment following a LOCA.

2.

Development of an analytical code to calculate pressure build-up in reactor cavities and other enclosures.

3.

Deveicpaent of a code to calculate the pressure-temperature transients in ice condenser containments.

4.

Modification of CONTEMPT-PS to include bypass flow between the (rywell and wetwell.

5.

Modification of CONTEMPT-PS to handle pressure transients in the air annulus in dual containments.

As the above needs are evaluated, you may determine that some of these tasks can best be performed under the ANC Regulatory Technical Assistance Program even though the tasks are of a code development nature.

Should this be the i

[M7/ 7#M ^

u i p ng

i

-i.

f-j o

I Milton Shaw

\\ l l

4 case, we would have no objections since, as was pointed coupled to analytical development efforts.

y.

We would be happy to discuss these needs with your staff and provide additi specific details as may be required.

b' Edson G. Case, Director Division of Reactor Standards Members of Steering Committee cc:

on Reactor Safety Research G. M. Kavanagh C. K. Beck M. B. Biles P. A. Morris H. D. Bruner.

A. J. Pressesky W. H. Beach, OCM G. O. Bright, WRSPO R. F. Fraley, ACRS (18) i

).,e,,... h h A _ g w.

.*. -..,,.. a _-

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